Although subchannel codes are used for the thermal-hydraulic analysis of fuel bundles in nuclear reactors from the former, many composition and empirical equations based on experimental results are needed to predict the two-phase flow behavior. When there are no experimental data such as the reduced-moderation water reactor (RMWR), therefore, it is very difficult to obtain highly precise predictions. The RMWR core adopts a hexagonal tight lattice arrangement with about 1mm gap between adjacent fuel rods. In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the flow characteristics. Thus, we start to develop a predictable technology for thermal-hydraulic performance of RMWR core using advanced numerical simulation technology. As part of this technology development, we are developing advanced interface tracking method to improve conservation of volume of fluid. In this paper, we describe a newly developed interface tracking method and examples of the numerical results. In the present results, the error of volume conservation in the bubbly flow is within 0.6%.

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In the dry reprocessing process for the spent nuclear fuels by the oxide electro-winning method, the oxygen ion exists in the molten NaCl-2CsCl due to oxygen gas purging of the melt for the MOX electrolysis deposition. It is thought that this oxygen ion influences the chemical form of rare earth ions in the molten NaCl-2CsCl, however the reaction behaviors of these rare earth ions with the oxygen ion has not been understood well.
In this study, dry air was purged into the molten NaCl-2CsCl containing SmCl₃ at 923K, and the behavior of Sm³⁺ in the melt was measured by the spectrophotometry and the differential pulse voltammetry. It was confirmed to generate the samarium oxychloride, SmOCl, by the reaction between H₂O in air purging reacted and SmCl₃ in the molten NaCl2CsCl. The amount of SmOCl generated increased with the amount of H₂O in air, and 37.5% of the amount of H₂O was consumed to generate SmOCl in this experiment.

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Colloidal effect is one of the major factors to enhance the transport of radionuclide in geosphere. In this study, the impact of kinetic reactions of radionuclide sorption onto both the fracture wall and colloid on the transport behavior of radionuclide was investigated by carrying out the series of transport experiment of the radionuclide with clayey colloid through a sample of fractured rock. In the experiments, Cs and colloids derived from bentonite were used as moderately sorbing radionuclide and typical natural colloid in the concentration of 1.0×10^-6g/l(7.5×10^-9mol/l)and 3.510^-3g/l, respectively. A granodiorite with the artificial the single fracture (dimension: 50×50×0.5mm)was used. The results of the transport experiments were analyzed by numerical simulation code of radionuclide transport COLFRAC which can simulate radionuclide transport based on either equilibrium or kinetic reaction of radionuclide sorption onto fracture wall and colloid. Consequently, it was indicated that results of the calculation strongly suggests the radionuclide sorption onto colloidal particles. It was suggested that kinetic sorption data are required with the measurements. The experimental results also indicated that the activity of radionuclide will be retarded by colloid filtration on the fracture.

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A design study program on the Gas Turbine High Temperature Reactor (GTHTR300) is being carried out in Japan Atomic Energy Research Institute (JAERI). Design of structures, systems and components which have fundamental engineering safety functions, such as shutdown of reactor, maintaining the coolability of reactor, prevention of the release of radioactivity, has progressed in this program. Coolability must be maintained even when a depressurization accident occurs which is the most severe event with loss of cooling function. The passive cooling system by using natural circulation of air was chosen because of simplicity of the system and its low cost. Cooling panels consisting of rectangular ducts were adopted after detail investigation on the capability of heat removal and on the structural integrity under high pressure condition during depressurization accident. This is an original design which is different from cooling panels in High Temperature Engineering Test Reactor (HTTR) of JAERI. Results of design study on performance of heat removal and structural integrity of passive cooling system of the GTHTR300 are described in this paper.

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In blasting decontamination, alumina or steel grit has been used as conventional blast material. The problem with such existing materials is the decreasing of blasting performance during the repeated use due to their diminishing hard- ness. Consequently, a lot of waste grit is generated as secondary waste.
In order to solve this problem, airconia has been selected because of its high strength and spherical shape. A repeating test in which blasting is executed 300 times and a decontamination test were performed using actual waste materials to confirm the applicability of zirconia grit.
More than 90% of zirconia grit kept the initial performance after being used 300 times. Zirconia exhibited durability superior to that of alumina which was broken in the course pf several repetitions of use. Moreover, the decontamination performance of zirconia grit has been confirmed to be equivalent to that of alumina grit.
From the results, zirconia grit is expected to be applied to the decontamination of metal wastes.

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Large-scale melting tests of simulated miscellaneous solid wastes were conducted to characterize solidified products (50L-drum size). Two heating modes were adopted in the tests: hybrid heating and induction heating modes. In the former, wastes were heated with both an induction furnace and a plasma torch, and in the latter, an electrically-conductive crucible was used with the induction furnace. Visual observation, chemical analysis and radioactivity measurements were conducted to the solidified products. It was found that the radioisotope distribution of solidified products was almost homogeneous. There was no remarkable deterioration in the solidified products subsequent to the leaching test.

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Target plate model (35mm×25mm×100mm) of divertor device for fusion reactor was fabricated using low activated and light material, Si/xSiC Metal Matrix Composite (MMC) and high strength and thermal conductive SiC. This model has bonding structure of armor tile (25mm×25mm×10mmt) of Hot Pressed (HP) and Hot Isostatic Pressed (HIP) Si/SiC (10wt%)-MMC and substrate (25mm×25mm×100mm) of SiC/BeO (1wt%) with water cooling channel (φ15mm).
Armor tile material, Si/10SiC-MMC shows high thermal conductive and low resistivity with over 200MPa peak bending strength at high temperature. This material behaves super plasticity over peak temperature corresponding to peak bending strength and low temperature brittleness of this material was improved to degrade DBTT to 673K by optimization of Hot Press (HP) and Hot Isostatic Press (HIP) conditions.
Substrate body material, SiC/1BeO behaves major characteristics of high bending strength over 500MPa, high density of 3.2g/cm³ and high thermal conductivity of 270W/m·K at room temperature.
Target plate model with water cooling by 293K and 2MPa pressurized water at 1m/s was exposed under 5MW/mm² ×30s high heat flux H⁺ ion beam and showed very good cooling performance and no damage was found after this test.

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For the purpose of comparing the mental attitudes toward nuclear power generation held by the public between localities containing nuclear installation sites and those of urban communities, a survey has been conducted by mail and by direct interviews. The survey covered Fukushima, Niigata and Fukui Prefectures, representing the former group of localities, whereas Tokyo and Osaka, representing the latter. The mail survey revealed differences between the two groups of localities in their image of nuclear power and in their perception of its present status, as well as in their sense of values in daily life. The interview survey indicated that the factors behind the differences included regional characteristics such as temperaments of the resident population, and effects on the private life style brought by nuclear installation. Narrowing down this gap of mental attitude between the two groups of localities should call indispensably for better mutual understanding. In bringing this about, consideration should be given to particular local characteristics, including popular mentality and personal view of life, as well as image and knowledge of energy sources held by the individuals.

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[in Japanese]

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