Microstructural and micromechanical characterization of Cr diffusion barrier in ATR irradiated U-10Zr metallic fuel

Chromium (Cr) has been recognized as a promising diffusion barrier candidate to mitigate the Fuel Cladding Chemical Interaction (FCCI) failure in metallic fuels for sodium-cooled fast reactor. This paper, for the first time, conducted an in-depth post-irradiation examination of the microstructure and composition evolution, and micromechanics of the Cr diffusion barrier in U-10Zr fuel/HT9 cladding irradiated in the Advanced Test Reactor (ATR) to 8.7% burnup at an averaged Peak Inner Cladding Temperature of 540–550 °C. Transmission Electron Microscope (TEM) characterization confirmed the preferential intergranular diffusion of Zr and U in the Cr diffusion barrier, suggesting that grain boundaries served as fast path for the diffusion of Zr and U into the Cr diffusion barrier. The interaction zone is dominated by nano crystalline α-ZrCr₂ Laves phase. Despite the interaction, there is no microcracks being observed in the preserved Cr diffusion barrier and HT9 cladding, serving as a good barrier to mitigate FCCI under the studied in-reactor irradiation conditions. High density cavities in uniform distribution are observed inside Cr grains, nano particles contains Cr, Mn, and O are confirmed by Electron Energy Loss Spectroscopy (EELS) analysis. However, it is unclear whether the cavities will become an issue to the barrier integrity at higher burnup. In-situ Scanning Electron Microscopy (SEM) micro-tensile testing uncovered mechanical softening in the HT9 cladding nearing the Cr diffusion barrier, possibly due to the coarsening of lath structure and carbides precipitates.