Hydride embrittlement resistance of Zircaloy-4 and Zr-Nb alloy cladding tubes and its implications on spent fuel management
Sangbum Kim, Joo‐Hee Kang, Youho Lee
Abstract
In the spent fuel storage phase, nuclear fuel cladding is subjected to increased embrittlement owing to a large amount of hydride precipitation. This study compares the differences in the hydrogen-induced cladding embrittlement of cold work stress-relief annealed (CWSR or SRA) Zircaloy–4 and Zr-Nb alloy cladding with ring compression test at the temperature of the spent fuel pool, which is approximately 40 °C. Experiments demonstrate that an abrupt ductile to brittle (DTB) transition occurs at the critical hydrogen content of 560 and 490 wppm for Zircaloy-4 and the tested Zr-Nb cladding tubes, respectively. Even beyond the critical hydrogen content, sufficiently high cladding ductility with the offset strain >10% is maintained up to ∼ 90 MWd/kgU for both cladding materials on the rod-average basis. Extensive EBSD analyses coupled to thermodynamic modeling demonstrate that this is primarily due to the slightly larger grain diameter of Zircaloy-4 tube, which reduces the number of available sites for inter-granular hydride precipitation. Reduced inter-granular hydride precipitation prevents the extent of hydride interlink, thereby improving the hydride embrittlement resistance. Nevertheless, the tested Zr-Nb alloy cladding presents an extened discharge burnup limit for abrupt DTB transition owing the reduced in-core cladding oxidation rate. The presented understanding of microstructural effect on hydride interlink and resulting embrittlement may provide a basis for understanding the general hydride embrittlement phenomena of Zircaloy cladding which include, but not limited to, wet storage, dry storage, and post-accident ductility of high burnup Zircaloy cladding.