Fission products speciation in nuclear fuel: Synthesis and characterisation of mixed oxide (U,Pu)O2 SIMfuel

Abstract

In the context of the nuclear fuel recycle, U1-yPuyO2±x Mixed Oxides (MOx) are currently the most studied fuels. These materials have to meet accurate physico-chemical properties, known to change drastically during irradiation with the formation of fission products (FP). Investigating the FP behaviour is thus crucial to predict the fuel properties evolution during irradiation. Model uranium-based materials called SIMfuel have been developed in the last decades to overcome the high radiotoxicity of UO2 spent fuel, and to enable separation-of-effect studies on its otherwise overwhelmingly complicated chemistry. In this work, a fabrication route for U1-yPuyO2±x SIMfuel (SIMMOx) has been developed, in order to study the speciation of soluble FP (Ce, La, Nd, Sr, Y, Zr) inside spent MOx fuel, and their interaction with the U1-yPuyO2±x matrix. The resulting material is representative of irradiated MOx fuel with Pu content of 21 g/g% (Pu/(U+Pu)=26 g/g%) and a burnup of 13 at.%.XRD and EPMA show a material representative of real spent fuel (U,Pu)O2 matrix: globally monophasic, and compliant with the homogeneity requirement of irradiated MOx. Raman microscopy confirmed the lack of secondary phases, showing that all the FP successfully entered the (U,Pu)O2 solid solution. Pu-enriched zones (<30 μm) contain higher FP concentration, showing that this phenomenon, already observed in irradiated fuel, is due also to thermodynamics and not only to burnup, as previously assumed.XAS analysis supplied crucial information such as the oxidation states, the O/M ratio, and the prevalent chemical form of each actinide, and FP. A charge compensation mechanism between the (U,Pu)O2 matrix and the FP in solution has been observed.