Hydride growth mechanism in zircaloy-4: Investigation of the partitioning of alloying elements - Mouton, Isabelle; Chang, Yanhong; Chakraborty, Poulami; Wang, Siyang; Stephenson, Leigh T.; Ben Britton, T.; Gault, Baptiste (Materialia, 2021)
The long-term safety of water-based nuclear reactors relies in part on the reliability of zirconium-based nuclear fuel cladding. Yet the progressive ingress of hydrogen during service makes zirconium alloys subject to delayed hydride cracking. Here, we use a combination of electron back-scattered diffraction and atom probe tomography to investigate specific microstructural features from the as-received sample and in the blocky-α microstructure, before and after electrochemical charging with hydrogen/deuterium followed by a low temperature heat treatment at 400 °C for 5 h followed by furnace cooling at a rate of 0.5 °C/min. Specimens for atom probe were prepared at cryogenic temperature to avoid the formation of spurious hydrides. We report on the compositional evolution of grains and grain boundaries over the course of the sample's thermal history, as well as the ways the growth of the hydrides modifies locally the composition and the structure of the alloy. We observe a significant amount of deuterium left in the matrix, even after the slow cooling and growth of the hydrides. Stacking faults form ahead of the growth front and the segregation of Sn at the hydride/matrix interface and on these faults. We propose that this segregation may facilitate further growth of the hydride. Our systematic investigation enables us discuss how the solute distribution affects the evolution of the alloy's properties during its service lifetime. [Display omitted]