Under a set of assumptions for Japan's future nuclear power generation program, necessities for spent fuel management are assessed up to the year 2050. Within a foreseeable time frame, plutonium recovered from reprocessing will mainly be recycled to light water reactor (LWR) plants as “Pu-thermal” fuel. From the viewpoint of spent fuel management, the following observations are suggested as important:
(1) In the middle-term up to 2010-2020, spent fuel arising which will exceed the domestic reprocessing capacity must be dealt with by storage. Amount of spent fuel to be stored can be forecasted with a high degree of precision.
(2) In the longer-range up to 2050, significant uncertainty exists depending on number of factors, such as nuclear power generation and actual operation of reprocessing. Forecasted requirement of spent fuel storage varies from none to uniformly increasing up to 25,000tHM by 2050.
(3) MOX spent fuel resulted from plutonium recycling in LWRs must be stored by the time when the second domestic reprocessing plant will be completed and can accept those fuel. This means careful planning and management will be needed for varying quantity and quality in the spent fuel accumulation in the near future depending on strategies to be undertaken.
(4) Theoretically, it is suggested that there may be the optima in duration and capacity in storage. Thus, deliberation must be maintained in implementing services for spent fuel, such as reprocessing and/or storage, in a way the whole system does not by far depart from those optimality.

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  The conventional cask storage facility has a high ceiling. In order to shorten the construction period and reduce the cost, a new type of cask storage facility is proposed. Its ceiling height is made low and the stack is set up on the opposite side of the air inlet through the storage area, giving what we call a “stack-type facility”. The flow pattern in the storage area of this facility is then different from that of the conventional facility and the heat removal performance of a stack-type facility has been studied. Simulation experiment by using a 1/5 reduced scale model was performed, and thermal-hydraulic phenomena of the facility were investigated. Following results were obtained by this study:
1) The ceiling height hardly influences the heat removal characteristic of the casks. On the other hand, the ceiling temperature is so much influenced by the ceiling height.
2) The stack height directly influences the heat removal of the casks. The stack height by which heat generation of casks could be safely removed was estimated.
3) There are two kinds of flows in the storage area of the facility, one is a upward flow induced by buoyancy force on the cask surface, and the other is a horizontal flow induced by the stack effect. These flows contribute to the cooling of casks.

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  We investigated three structural materials which may not contain a high level of boron or enriched boron, such as (1)Borated stainless steel(B-Stainless Steel), (2) Composite material(three-layered clad plate composed of B-Stainless Steel/Cu/B-Stainless Steel structure), and (3) Borated aluminum alloy, Applicability of these material for baskets was studied, several material tests were conducuted, and the material data which was required for the standarization, etc. were obtained. Also, matters to be considerd were descrived. As a result of above study, by using above 3 materials with cask baskets the designs of casks and cask baskets which corresponded to high burnup and MOX spent fuels could be specifically indicated.

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  A new alternative method for separation of Np in the first eo-decontamination step is proposed. It comprises two steps, namely, preconditioning of Np valence state in the dissolved solution of spent fuel by NO gas bubbling in HNO₃ medium to produce HNO₂, which is considered as salt-free process to convert Np(VI) to Np(V) and stabilization of Np(V) with urea, finally, the demonstrative counter current cascade extraction of Np(IV) and Np(V) in a miniature mixer-settler was carried out. The batch experiments show that Np(V) produced after conditioning may be slowly oxidized again to Np(VI) during standing time. Addition of urea in the HNO₃ solution might enhance the stability of Np(V). On the other hand, the solvent extraction by 30% TBP/kerosene could greatly accelerate the oxidation rate of Np(V). The chemical flow sheet study at 25°C shows that, more than 98% of Np could be routed into HLLW if urea is added in the HNO₃ solution. The operating temperature has great influence on the kinetics of Np(V) oxidation. If operation temperature races to 36°C and urea is not added, about 38% of Np will go along with U and Pu into organic phase. The behavior of Np(IV) during extraction shows great accumulation in the middle stages of battery.

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  Bulk compositional, mineralogical and physical properties of weathered basaltic ash soil (“Andisol”) derived mainly from Mt.Fuji were studied. Mineralogical studies revealed that the dominant primary material and weathering products are volcanic glass, allophane and halloysite and the sequence of weathering is volcanic glass→allophane→10Å halloysite→7Å halloysite. X-ray fluorescence analysis indicates that the relative elemental mobilities during the weathering is Na, Ca>K>Mg>P>Si>Ti, Fe>Al>Mn. The trends of soilwater chemistry (H₄SiO₄ concentration) with depth were calculated based on dissolution -precipitation kinetics-fluid flow coupling model. In order to calculate the trends, the data on present-day annual rainfall, solubility of basalt glass, porosity and specific weight of soil, deposition rate of volcanic ash and grain size of volcanic glass were used. The calculated results were compared with analytical trends of soilwater chemistry. From this comparison the dissolution rate constant of basalt glass was estimated to be 10^-9.4 - 10^-9.2 (mole Si m^-2 s^-1). This value is consistent with previous experimental dissolution rate constant of basalt glass reported in the literature.

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  It is necessary to visualize and characterize pore geometries in rocks for accurate estimation of underground flow system. Recent technological development prodides three dimensional observation of internal structure of rock with high resolution. X-ray computerized tomography (CT) medical scanner, optical microscope observation of void space dyed with colored resin, Confocal Scanning Laser Microscope (CSLM), mercury intrusion porosimetry and Atomic Force Microscope (AFM) were used as tools to visualize pore networks. By comparison of each advantage and disadvantage, it is concluded that micro-focus X-ray CT system is the most effective for three-dimensional pore visualization, and the usage of AFM combinated with the data of pore size distribution obtained by mercury intrusion porosimetry is the most suitable for the permeability estimation.

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  It is necessary to evaluate the influence of colloids on radionuclide transport in geologic environments relevant to safety assessment of radioactive waste geological disposal. This report summarizes the recent researches from the viewpoint of colloids generation in radioactive waste disposal system, influence of the colloids on the radionuclide transport and transport model. The colloid of radionuclide could be originated in the waste dissolution. It could be retained in the engineered barrier system. The influence of humic and biological colloids on radionuclide migration should be estimated, thus the researches on the interaction of organic matter, microbe and radionuclide are required. Colloid generation from bentonite and cement material has been investigated, however the data are not sufficient to discuss the significance in radionuclide transport. To simulate the radionuclide migration behavior under existence of colloids, the modeling of the equilibrium among three phases (radionuclide-colloid-rock) is important.

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