Yttrium stabilised zirconia inert matrix fuel irradiation at an international research reactor

Different concepts have been developed during the last decade to transmute transuranium elements (TRU) using uranium-free inert matrix fuels (IMF) in a once-through-cycle to reduce the amount of TRU in the nuclear waste. For today’s LWRs yttrium stabilised zirconia (YSZ) and other oxides like alumina, spinel or ceria have been proposed as inert matrix materials. By employing IMF, a larger fraction of plutonium can potentially be consumed in comparison with MOX fuels without breeding new plutonium. The aim of the presented study is to measure the general thermal behaviour of YSZ-based IMF under irradiation conditions similar to those in current LWRs in direct comparison to standard MOX fuel. Of particular interest are the fuel thermal conductivity (and its degradation with burnup), fission gas release (FGR), fuel densification and fuel swelling. A secondary aim is the direct comparison of the fuel performance between YSZ-based IMF and MOX fuel. The irradiation is performed under HBWR conditions and has reached an average assembly burnup of ∼300 kW d cm−3 until the end of 2004, which is equivalent to ∼29 MW d kg−1 for the MOX fuel.