A new thermo-chemical and electrolytic hybrid hydrogen production system in lower temperature range is newly proposed by the Japan Nuclear Cycle Development Institute (JAEA) to realize the hydrogen production from water by using the heat generation of sodium cooled Fast Breeder Reactor (FBR). The system is based on sulfuric acid (H 2 SO 4 ) synthesis and decomposition process developed earlier (Westinghouse process), and sulfur trioxide (SO 3 ) decomposition process is facilitated by electrolysis with ionic oxygen conductive solid electrolyte to reduce the operation temperature 200-300• C lower than Westinghouse process. SO 3 decomposition with the voltage lower than 0.5 V was confirmed in the temperature range of 500 to 600• C and theoretical thermal efficiency of the system evaluated based on chemical reactions was within the range of 35% to 55% under the influence of H 2 SO 4 concentration and heat recovery. Furthermore, hydrogen production experiments to substantiate the whole process were performed. Stable hydrogen and oxygen production were observed in the experiments, and maximum duration of the experiments was about 5 hours.

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The Japan Atomic Energy Agency has been developing centrifugal contactors for solvent extraction to apply to next-generation reprocessing plants. The centrifugal contactor has some attractive advantages such as more compact design and shorter liquid residence times than conventional contactors. Many kinetic studies using a miniature centrifugal contactor have been carried out worldwide. However, there are few engineering-scale studies in which stage efficiencies, transition behavior of concentration profiles, and robustness under maloperation conditions have been comprehensively discussed for a contactor cascade system. In this study, we carried out extraction and stripping tests of an engineering-scale centrifugal contactor cascade system based on a flowsheet of 10 kg-HM/h using uranyl nitrate solution. As a result, the stage efficiencies on uranium extraction and stripping were quite high, nearly 100% for extraction and 97-98% for stripping. The uranium concentration profiles became stable within 10 minutes for both extraction and stripping sections. No overflow or entrainment was observed under normal operation during the extraction and stripping tests. During the stripping test, it was estimated that an increase in temperature of the feed stripping solution from 308 to 333 K or a decrease in the flow rate ratio of the organic to aqueous phase from 1.0 to 0.8 corresponded to the distribution capacity of the two contactors. The maloperation test, in which a motor at a stage of the contactor cascade system was intentionally stopped, showed that the system could maintain stable operation with no emergency shutdown following the installation of at least two additional stages.

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