Inhibition of uranium dissolution in mixed uranium thorium dioxide under geological disposal conditions

A safety case supporting the direct disposal of spent mixed oxide (MOx) nuclear fuels in a deep geological repository (DGR), or of plutonium waste processed into a disposal MOx, must consider not only the local redox conditions but also the impact of the additional actinide on the oxidation and subsequent dissolution of the uranium dioxide matrix in the event of container breach. The higher alpha radiation field surrounding plutonium containing fuels increases the production of radiolytic oxidants close to the surface of the fuel and which could enhance the oxidative dissolution of uranium compared to UO 2 fuels. Conversely, doping uranium dioxide with less soluble actinides, has the potential to stabilise the uranium dioxide matrix. Thorium is analogous to plutonium in the latter respect. Model MOx fuel pellets, of lower radioactivity, containing uranium and thorium dioxide have been fabricated and their dissolution studied under conditions relevant to geological disposal. The environment of a deep repository with a clay host rock is simulated by the imposition of anoxic conditions and use of dearated synthetic Callovian Oxfordian ground waters, containing 0.5 mmol.L −1 hydrogencarbonate ions, doped with 0.1 mmol.L −1 hydrogen peroxide. The uranium release and peroxide consumption over time are compared in static batch experiments with UO 2 , ThO 2 and U 0.75 Th 0.25 O 2 pellets and to blank experiments. In experiments with UO 2 and U 0.75 Th 0.25 O 2 pellets, dissolution is independent of hydrogen peroxide concentration, rather the dissolution is controlled by the conduction of holes. The homogeneous doping of uranium dioxide with 25% thorium dioxide has reduced the normalised uranium release by at least a factor of five when compared with the dissolution of undoped uranium dioxide pellets. Hole conduction appears more limiting in the case of mixed uranium thorium oxide pellets due to Th(IV). The results presented here support the proposition that despite the increased radioactivity of MOx fuel, disposal of MOx would not have a solubility greater than UOx (uranium oxide) spent fuel under these conditions. Laboratory simulation of radiolytic production has been improved by development of a method by which a syringe pump continuously adds ground waters doped with hydrogen peroxide in batch systems. This overcomes the time limitations of conventional batch dissolution with peroxide that decomposes within a period of several days. Uranium dissolution of mixed uranium thorium dioxide pellets is similar in the static batch experiments with and without continuous additions of hydrogen peroxide. In this regime where uranium dissolution is independent of hydrogen peroxide concentration, the dissolution yield is not a useful parameter to compare between different experimental methods.