A continuous evaporation experiment was carried out, in order to verify the ruthenium (Ru) volatilization mechanism, for a simulated high-level liquid waste (HLLW). The feed solution contained 2.65M (10³ mol/m³) of nitric acid and 0.097M fission product compounds (FPs) including 0.008M RuNO3+. The operation pressure was 12, 700±200 Pa and the operation was continued for 44 h. The total volume of the feed solution was 13.0×10^-3m³ and the volume of the final evaporator solution was 0.90×10^-3m³. The concentration ratios of elements in distillate to concentrate initially decreased with the evaporator operation. Then the ratios reached plateau for Cs and Ce, with values of about 7×10^-7. The ratio for Ru was 200% of those for Cs and Ce at 15 h after the operation began, and decreased to 110% toward the end of operation. The difference in ratios was quantitatively explained by the formation rate of volatile Ru in simulated HLLW solution taking into account the changes of activity and concentration of nitric acid caused by salting out effect of FPs.

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A new rate equation is derived for uranium isotope exchange reaction between gaseous UF₆ and solid UF₅ by assuming the total number of UF₅ molecules on the particle surfaces to depend on time. The reaction parameters included in the equation are determined from the experimental results, and are compared with the previous ones. The rate equation given in this work satisfies the related isotopic mass balance, and includes explicitly the terms related to the UF₆ density and the mean size of UF₅ particles, both of which cause an important effect on the reaction. Since the rate equation derived in this work facilitates the simulation of the reaction under various conditions, the long term behavior of a simulated exchange reaction is studied under the condition considered to be close to that in a recovery zone of the MLIS process.

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Laboratory centrifugation tests using irradiated fuel dissolution juices claim that hydrosoluble polymer adjunction before centrifugation, might increase the clarification efficiency. In fact, it may be enhanced from 50% at the reference state to 99% after flocculant addition. The flocculant effect consists in increasing the aggregate initial average size and decreasing the initial number of aggregates, but also in improving the floc resistance to the shear field of the centrifugal separator. An experimental process has been perfected to transpose the laboratory results to an industrial scale. The flocculant shall only be selected if the average size of the shear produced microflocs lies higher than the average size of the initial colloidal aggregates.

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Experiments were performed to study the permeation of hydrogen isotopes through a membrane of palladium-silver alloy in the temperature range of 353 to 673 K using the co-current double tube method. It is observed that permeation is limited by the diffusion process of hydrogen atoms in the bulk of membrane in the larger pressure difference range between inner and outer channels. Whereas the recombination reaction of hydrogen atoms on the outer surface of membrane is the rate controlling step in the lower pressure difference range below 10³Pa. There exists a wide intermediate range where the both processes cannot be ignored. The isotope effect ratio and the separation factor in permeation are also quantified. The permeation behavior of multi-component hydrogen isotopes through membrane of palladium-silver alloy is discussed applying observations in this study.

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Axisymmetric concentration profiles of 3-component isotope separating thermal diffusion column with continuous feed and draw-offs were calculated by solving the convection-diffusion equations using Newton iterative method, where the flow variables were specified as input. An approximate expression of thermal diffusion coefficient in multi-component mixture in terms of binary thermal diffusion factors and binary diffusion coefficients was used for the analyses.
Analytical results were compared with experimental data of 3-component natural abundance argon 36-38-40 isotopes separated in a 950 mm-height, inner 15 mmφ tube column with a 0.3 mmφ hot wire in the axis. Validity of the analyses and the approximation in the thermal diffusion coefficient was ensured by close agreement with experimental results.

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Sodium pool fire code, SOFIRE II, written for the constant value of stoichiometric combustion ratio and heat of reaction is used to compute the buildup of pressure and temperature in a containment. In the SOFIRE II model, for the formation of a mixture of Na₂O and Na₂O₂ in the sodium pool, the input stoichiometric combustion ratio and heat of formation values need to be varied to corresponding values admissible for the mixture. In the present work, the SOFIRE II one-cell model is revised and the present version SFIRE1C (Sodium FIRE 1 Cell model) accounts for the formation of Na₂O in an early stage of the fire and shifts to the formation of Na₂O₂ at a later stage. Thus SFIRE1C computes in a more realistic manner the reaction products which are formed in the pool. The model for sodium oxide aerosol release is also modified in this version, by incorporating a more appropriate aerosol release rate equation. The calculated values using the SFIRE1C one-cell model are compared with sodium pool fire experimental results.

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A computer code system DSOCEAN has been developed for assessing the collective dose of Japanese due to radionuclides released to the ocean from a spent nuclear fuel reprocessing plant. This computer code system uses a box model which represents the transfer of radionuclides between boxes of seawater into which the ocean around Japan is divided. The code system consists of a series of three interlinked main computer codes, which estimates the exchange rates of radionuclides between the boxes, the radionuclide concentrations in each box, and the collective dose from various exposure pathways, respectively.
By using DSOCEAN, two calculations were carried out to estimate the collective dose from a liquid effluent. One is associated with a routine release of radionuclides from a hypothetical reprocessing plant. The other is an application of the code system to disposal of liquid radioactive waste to the surface water of the ocean. The calculated results identified important radionuclides and exposure pathways.

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It is recognized in the source term evaluation of light water reactors that large uncertainties still remain in the fission products aerosol behavior in reactor coolant system during severe accidents⁽¹⁾. Especially the aerosol size distribution has a great impact on the deposition behavior onto the inner surface of pipe wall through several mechanisms such as gravitational sedimentation and impaction. The present analytical models⁽²⁾⁽³⁾ used in source term evaluation consider the growth of aerosol size due to steam condensation onto aerosol particles. However the effect of condensation is neglected in superheated steam. In an early stage of severe accident sequence the generated FP aerosol will be transported to primary circuit by superheated steam. Therefore it is essential to evaluate the aerosol size distribution under superheated steam. In the present experiments, the CsI gas and aerosol behaviors at a PWR hot leg inlet was simulated and the effect of difference in carrier gas, nitrogen and steam, on the CsI behavior in pipe was investigated.

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The separation methods of transuranium elements (TRUs) from high-level radioactive liquid waste (HLW) are classified into two categories, such as the wet processes and dry processes. A new partitioning process which combines the wet process and dry process has been developed by Toshiba Corporation. Figure 1 shows this process which combines oxalate precipitation and electrorefining. The oxalate precipitation yields good TRU recovery and electrorefining gives good TRU separation from HLW. There are three steps in the process. First, TRUs, lanthanides and alkaline earth elements whose oxalates are lower in solubility are recovered from HLW by the addition of oxalic acid. The previous paper⁽¹⁾ shows the experimental study which satisfies the requirement of 99.9% recovery of simulated TRUs from simulated HLW by the oxalate precipitation. In the second step, theseoxalates are converted to chlorides. In the final step, the TRUs are recovered on a cathode by electrorefining after dissolving the chlorides in eutectic KC1-LiC1 salt.
The TRU oxalates were converted to chlorides by inducing reactions with CoCl₂, HCl or CCl₄ at 500 °C^(2)-(4). The materials of apparatus must be using corrosion resistance at high temperature. Further, it is necessary for tetravalent TRU chlorides such as NpCl₄, whose vapor pressures are high⁽⁵⁾, to be recovered. Therefore, the conversion from oxalates to chlorides satisfies the requirements of no use of corrosive gas and lower reaction temperature.
This note describes an experimental study to confirm the new partitioning process, which is found to satisfy the requirement for conversion of simulated TRU oxalates to chlorides without corrosive gas and at lower temperature.

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