The high pH condition of aqueous solution in the radioactive waste repository can be produced by dissolution of hydrated cement. Many studies have been reported to clarify the mechanism to maintain high pH for a long term, however, these can not represent all phenomena related to the dissolution process of the cement hydrates because of the lack of solid phase analyses. We have studied not only the aqueous phase but the change of the solid phase simultaneously. We have studied the applicability of the permeability test method using centrifugal force to percolate pore water. By using this method, we have investigated the dissolution phenomena of hydrated cement within acceptable period of experiment.
  As a result, the solid phase analysis indicated the clear distinction between altered and unaltered area and the existence of a dissolution front was identified. On the altered area of the solid phase, it was confirmed that Ca(OH)₂ was dissolved. An obvious change of microstructure was observed. The relation between the volume of percolated water and concentration of the liquid phase composition contained in percolated water through a specimen of hydrated cement was found to be quite similar to that of the test data by Pfingsten et. al.[1] and consequently the dissolving process of cement was found to be identical.

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  In order to develop the dissolution process of irradiated fuels under mild conditions, the dissolution behavior of UO₂ pellet with clad in HNO₃ solutions has been examined under various conditions. The dissolution rate at the boiling point (b.p.) was confirmed to be almost the same as that at 90°C and such a phenomenon was found to be due to the effect of HNO₂, because the concentrations of HNO₂ in the dissolvent solution at 90°C is maximally 50 times that at b.p. The UO₂ pellets containing 13 elements as simulants of fission products(FP) were dissolved in HNO₃ solutions at various temperatures to examine the behavior of undissolved residues under the mild dissolution conditions. It was found that a total amount of undissolved residues at 90°C is less than that at b.p. and that main components of undissolved residues in the dissolution of simulated fuels are almost the same as those of irradiated fuels. Furthermore, corrosion rates of materials such as SUS304ULC and SUS310Nb were examined in the HNO₃ solutions containing simulated FP ions at various temperatures. It was clarified that the corrosion rates of these materials at 90°C are almost 1/10 those at b.p. and that the retardation of corrosion rate is attributed to the effect of lowering temperature and HNO₂. From these results, it is expected that the dissolution of irradiated fuels becomes possible under milder conditions than those in conventional reprocessing processes and this results in improvements of confidence and safety in the dissolution process.

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  It has already passed 20 years since the Tokai reprocessing plant started its operation by Japan Nuclear Cycle Development Institute (JNC; former Power Reactor and Nuclear Fuel Development Corporation (PNC)) as the first reprocessing plant in Japan. The Tokai reprocessing plant has been operating with the mechanical head-end process and the solvent extraction (PUREX). While JNC had a lot of experiences such as the dissolver failure etc., the reprocessing amount achieved to 936 tons.
  On the other hand, while the fast breeder reactor (FBR) development was carried out for realization of plutonium utilization, JNC started the development of the FBR fuel reprocessing technology. The FBR fuel reprocessing has different characteristics from LWR reprocessing as follows, wrapper tube, high plutonium content, high fission products content by high burn-up. Therefore, JNC has been developing the reprocessing equipment, which satisfied following requirements, wrapper tube removal process (dismantling process), high capacity and compact design at severe critical condition, corresponds to increase of the specific activity and insoluble sludge.
  And also JNC has been started the advanced reprocessing technology with not only PUREX process but also advanced process, so as to meet various social needs, such as improvement of safety and economics, effective use of resource, decrease of environmental burden and improvement of non-proliferation.

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  The chop and reach method used in conventional reprocessing facilities has several disadvantages as follows,
(1)A large shearing machine is used for chopping fuel assemblies. It has significant problems for long-term operation and maintenance.
(2)Dissolution rate of fuel pellets with zircalloy cladding is very slow. The dissolver is operated in severe conditions.
  Hence, a new head-end process was developed to improve disadvantages in conventional one. This process consists of several steps, such as the dismantling of fuel assemblies without a large shearing machine, the chemical decladding of fuel, the removal of iodine from the fuel by using voloxidation technique and the dissolution of fuel under milder conditions.
  In this paper, the recent research on the development of proposed process was reported and its applicability to the practical use was reported.

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