Inert matrix and thoria fuel irradiation at an international research reactor

A major issue in the public debate on nuclear power, is how to break down the large plutonium stockpiles. Different concepts have been developed during the last years to burn plutonium. Two such concepts are stabilised zirconia based inert matrix (IM) and thoria (T) fuels. By using of IM fuels a larger fraction of plutonium could potentially be consumed without breeding new plutonium in comparison with todays MOX fuels. The aim of the presented study is to measure the general thermal behaviour of IM, inert matrix doped with thoria (IMT) and thoria under irradiation conditions similar to those in current light water reactor (LWRs). Of particular interest are the fuel thermal conductivity (and its degradation with burnup), fission gas release (FGR), fuel densification and fuel swelling. The irradiation is performed under HBWR conditions and a target burnup of ò≠400-450 kW d cm-3, which is equivalent to ò40-45 MW d kg-1 for the MOX fuel, is envisaged. Among other things considerably higher operating temperatures in the IM and IMT rods have been observed compared with those in the thoria fuel. The higher temperatures, which were caused by the lower thermal conductivity of IM, result in higher FGR of the IM and IMT fuel. This work gives the obtained results after 6 cycles (671 days) of irradiation.