Researchers at Idaho National Laboratory and the Massachusetts Institute of Technology have conducted a three-dimensional study of irradiated U-10Zr—uranium alloyed with 10 percent zirconium—across key radial regions, providing new insights into how the material swells, transfers heat, and interacts with the fuel cladding.

U-10Zr was extensively tested in historical sodium-cooled fast reactors and is now attracting attention for use in next-generation advanced reactors.