Uranium silicide fuels proved over decades their exceptional qualification for the operation of higher flux material testing reactors with LEU elements. The application of such fuels as target materials, particularly for the large scale fission Mo-99 producers, offers an efficient and economical solution for the related facilities. The realization of such aim demands the introduction of a suitable dissolution process for the applied U3Si2 compound. Excellent results are achieved by the oxidizing dissolution of the fuel meat in hydrofluoric acid at room temperature. The resulting solution is directly behind added to an over stoichiometric amount of potassium hydroxide solution. Uranium and the bulk of fission products are precipitated together with the transuranium compounds. The filtrate contains the molybdenum and the soluble fission product species. It is further treated similar to the in-full scale proven process. The generated off gas stream is handled also as experienced before after passing through KOH washing solution. The generated alkaline fluoride containing waste solution is noncorrosive. Nevertheless fluoride can be selectively bonded as in soluble CaF2 by addition of a mixture of solid calcium hydroxide calcium carbonate to the sand cement mixture used for waste solidification. The generated elevated amounts of LEU remnants can be recycled and retargeted. The related technology permits the minimization of the generated fuel waste, saving environment, and improving processing economy. 1. Introduction Particularly for the large scale producers, the conversion of the production targets for fission Mo-99 presents a serious challenge for keeping economical conditions for operating their plants. The uranium enrichment dropping from ~90% to ~19.8% demands modifications on process operation to compensate for the resulting loss in output. Evaluations based on keeping the production process proven since decades unchanged and just increasing the amount of processed targets are not realistic in general. The dominant reasons are limitations on efficient irradiation positions in the available research reactors plus drastically increased processing and waste costs. The idea of maintaining the current production process [1–8] by increasing fuel densities of the targets exploiting the progress in target technology from actually ~1？gU/cm3 for highly enriched uranium (HEU) to approximately 2.6？gU/cm3 for low enriched uranium (LEU) targets can be classified as a compromise. Such compromise is appropriate for several small- and medium-scale facilities but not
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