Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

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dc.contributor.authorPasqualini, EEen_AU
dc.contributor.authorRobinson, ABen_AU
dc.contributor.authorPorter, DLen_AU
dc.contributor.authorWachs, DMen_AU
dc.contributor.authorFinlay, MRen_AU
dc.date.accessioned2020-03-26T06:22:33Zen_AU
dc.date.available2020-03-26T06:22:33Zen_AU
dc.date.issued2016-10-01en_AU
dc.description.abstractNuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U–(7–10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry–4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction—either from fabrication or in-reactor testing—and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm3, 3.8E+21 (peak). © 2016 Elsevier B.V.en_AU
dc.identifier.citationPasqualini, E. E., Robinson, A. B., Porter, D. L., Wachs, D. M., & Finlay, M. R. (2016). Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding. Journal of Nuclear Materials, 479, 402-410. doi:10.1016/j.jnucmat.2016.07.034en_AU
dc.identifier.govdoc9013en_AU
dc.identifier.issn0022-3115en_AU
dc.identifier.journaltitleJournal of Nuclear Materialsen_AU
dc.identifier.pagination402-410en_AU
dc.identifier.urihttps://doi.org/10.1016/j.jnucmat.2016.07.034en_AU
dc.identifier.urihttp://apo.ansto.gov.au/dspace/handle/10238/9260en_AU
dc.identifier.volume479en_AU
dc.language.isoenen_AU
dc.publisherElsevieren_AU
dc.subjectZircaloyen_AU
dc.subjectResearch reactorsen_AU
dc.subjectTest reactorsen_AU
dc.subjectFuel-cladding interactionsen_AU
dc.subjectFuel platesen_AU
dc.subjectArgentine CNEAen_AU
dc.subjectATR Reactoren_AU
dc.subjectAluminiumen_AU
dc.subjectFabricationen_AU
dc.subjectHeterogeneous effectsen_AU
dc.titleFabrication and testing of U–7Mo monolithic plate fuel with Zircaloy claddingen_AU
dc.typeJournal Articleen_AU
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