首页|Hydride embrittlement resistance of Zircaloy-4 and Zr-Nb alloy cladding tubes and its implications on spent fuel management
Hydride embrittlement resistance of Zircaloy-4 and Zr-Nb alloy cladding tubes and its implications on spent fuel management
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NSTL
Elsevier
& nbsp;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 & nbsp;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 ~& nbsp;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 hy-dride 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. (C) 2021 The Authors. Published by Elsevier B.V.& nbsp;
NSTL
Elsevier
