Irradiation behavior of U–Nb–Zr alloy dispersed in aluminum

Three U–Nb–Zr alloys (U–5Nb–3Zr, U–6Nb–4Zr, and U–9Nb–3Zr) were included in a screening irradiation test of low-enrichment aluminum matrix dispersion fuels. Fuel particles made from these alloys reacted readily with aluminum during fuel fabrication and post-fabrication annealing, resulting in large...

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Bibliographic Details
Published in:Journal of nuclear materials 2001-11, Vol.299 (2), p.175-179
Main Authors: Meyer, M.K, Hofman, G.L, Wiencek, T.C, Hayes, S.L, Snelgrove, J.L
Format: Article
Language:English
Subjects:
Applied sciences
Energy
Energy. Thermal use of fuels
Exact sciences and technology
Fission nuclear power plants
Installations for energy generation and conversion: thermal and electrical energy
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Summary:Three U–Nb–Zr alloys (U–5Nb–3Zr, U–6Nb–4Zr, and U–9Nb–3Zr) were included in a screening irradiation test of low-enrichment aluminum matrix dispersion fuels. Fuel particles made from these alloys reacted readily with aluminum during fuel fabrication and post-fabrication annealing, resulting in large fuel plate thickness increases. Under irradiation, the behavior of U–5Nb–3Zr (wt%) alloy based fuel was poor at 41 at.% 235U burnup, showing indications of incipient breakaway swelling. The post-irradiation microstructural characteristics of U–6Nb–4Zr based fuel were somewhat better than those of U–5Nb–3Zr, but is marginal at 70 at.% burnup. U–Mo based fuels generally show less reaction on fabrication and better fuel performance characteristics during irradiation.
ISSN:0022-3115
1873-4820
DOI:10.1016/S0022-3115(01)00676-6