Post-irradiation examination of uranium-7 wt% molybdenum atomized dispersion fuel

Ann Leenaers, Sven Van Den Berghe, Edgar Koonen, C. Jarousse, Francois Huet, M. Trotabas, M. Boyard, Sébastien Guillot, Leo Sannen, Marc Verwerft

    Research outputpeer-review

    159 Scopus citations


    Two low-enriched uranium fuel plates consisting of U-7wt%Mo atomized powder dispersed in an aluminum matrix, have been irradiated in the FUTURE irradiation rig of the BR2 reactor at SCK · CEN. The plates were submitted to a heat flux of maximum 353 W/cm2 while the surface cladding temperature is kept below 130°C. After 40 full power days, visual examination and profilometry of the fuel plates revealed an increase of the plate thickness. In view of this observation, the irradiation campaign was prematurely stopped and the fuel plates were retrieved from the reactor, having at their end-of-life a maximum burn-up of 32.8% 235U (6.5% FIMA). The microstructure of one of the fuel plates has been characterized in an extensive post-irradiation campaign. The U(Mo) fuel particles have been found to interact with the Al matrix, resulting in an interaction layer which can be identified as (U,Mo)Al3 and (U,Mo)Al3. Based on the composition of the interaction layer it is shown that the observed physical parameters like thickness of the interaction layer between the Al matrix and the U(Mo) fuel particles compare well to the values calculated by the MAIA code, an U(Mo) behavior modeling code developed by the Commissariat à l'énergie atomique (CEA).

    Original languageEnglish
    Pages (from-to)39-47
    Number of pages9
    JournalJournal of Nuclear Materials
    Issue number1
    StatePublished - 1 Oct 2004

    ASJC Scopus subject areas

    • Nuclear and High Energy Physics
    • General Materials Science
    • Nuclear Energy and Engineering

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