Annealing tests of in-pile irradiated oxide coated U–Mo/Al–Si dispersed nuclear fuel
Source abstract
Abstract
U-Mo/Al based nuclear fuels have been worldwide considered as a promising high density fuel for the conversion of high flux research reactors from highly enriched uranium to lower enrichment. In this paper, we present the annealing test up to 1800°C of in-pile irradiated U-Mo/Al-Si fuel plate samples. More than 70% of the fission gases (FGs) are released during two major FG release peaks around 500°C and 670°C. Additional characterisations of the samples by XRD, EPMA and SEM suggest that up to 500°C FGs are released from IDL/matrix interfaces. The second peak at 670°C representing the main release of FGs originates from the interaction between U-Mo and matrix in the vicinity of the cladding.
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Research context
OpenAlex topics: Nuclear Materials and Properties · Nuclear reactor physics and engineering · Radioactive element chemistry and processing
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References (46)
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