NRA research on fuel cladding ductility loss at operational temperature due to radial Zr hydrides
Zirconium alloys are widely employed as the materials for nuclear fuel cladding in light water reactors mainly due to their good corrosion resistance and low thermal neutron-absorption cross section. Since fuel cladding works as a physical barrier to prevent the release of radioactive materials, it is crucial to maintain cladding ductility to ensure its integrity over its entire life.
Zirconium alloys are known to be embrittled by hydrogen absorption and Zr hydride precipitation during operation. While these phenomena had been extensively studied, the conditions for the occurrence of macroscopic hydrogen-induced ductile-brittle transition at operational temperatures (about 300°C) with quantitative consideration of precipitated Zr hydrides' morphology remained unclear.
In the presented study , tube burst tests and post-test analyses on the high-burnup fuel cladding samples were performed to investigate the effect of precipitation and reorientation of Zr hydride on the ductility of fuel cladding at 300°C. The results quantitatively demonstrated that circumferential permanent strain to failure decreased as the hydrides became longer, more densely precipitated, and more aligned in the radial direction. Furthermore, the observations indicated that approximately 1% of the circumferential strain serves as a threshold for identifying the macroscopic hydrogen-induced DBT of the zirconium alloy cladding.
 A. Yamauchi and K. Ogata, J. Nucl. Mat. Sci., 57 (2020), 301–311
Nuclear Regulation Authority (NRA)