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JNC TN8400 2001-022, 60 Pages, 2001/03
JNC-TN8400-2001-022.pdf:1.31MB
A numerical simulation code for the TRUEX (Transuranium Extraction) process was developed. Concentration profiles of americium and europium were calculated for some experiments of the counter current extraction system those were carried out in CPF (Chemical Processing Facility) by using the code. Calculation profiles were in agreement with the experimental results. Operational conditions were also examinted for the americium recovery experiment by the TRUEX process carried out in the Plutonium Fuel Center. It was shown that lowering the concentration of nitric acid in the scrub solution and decreasing the flow rate of solvent and strip solution was effective for improving the performance of the stripping step and reducing the volume of the waste solution. In order to find the optimum conditions for various experiments, this simulation code was modified to calculate the concentration profiles of other metal elements such as zirconium and iron and the effect of oxalic acid on the extraction behavior of the metal elements. The calculated concentration profiles of americium and europium were varied by this modification. In the experiment at CPF, the calculations were carried out to obtain recovery ratio of americium in the product stream with the amount of oxalic acid added to the process. This calculation result showed that it was possible to improve the performance of decontamination of fission products by increasing oxalic acid concentration added to the process. The calculation was also carried out for finding the optimum conditions of oxalic acid concentration added to the europium recovery process.
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JNC TN8440 2000-013, 179 Pages, 2000/04
JNC-TN8440-2000-013.pdf:10.31MB
The plutonium conversion development facility (PCDF) has been operated for 17 years and about 12 tons plutonium-uranium mixed oxide (MOX) powder has been converted since operation started in 1983. The first maintenance program for aging of apparatus was carried out from 1993 to 1994. The calcination-reduction fumace, liquid waste evaporator had been dismantled and renewed. The second maintenance program was carried out form 1998 to 1999. The microwave ovens, powder blender, ventilation control panel and so on were dismantled and renewed. Large volume radioactive wastes were generated during this maintenance such as the furnace, the filter casings and glove boxes. These wastes were too large to be packed into the waste container and these wastes were polluted by MOX powder unfixed on these surface. SO cutting and packing operation for these wastes and recovery of MOX powder from them were carried out. In this report, the method of this cutting and packing operation, the radioactive exposure to the operators in this operation, the estimation of nuclear material quantity migrated to filters, the evaluation of re-floating factor of radioactive material, etc. were discussed.
; Koyama, Tomozo; Funasaka, Hideyuki
JNC TN8400 2000-016, 188 Pages, 2000/03
We summarized the conditions and results of all dissolution experiments (bench scale experiments (dissolution of sheared fuel pins) and beaker scale experiments (dissolution of a few sheared fuels pieces) of the irradiated fast reactor fuels, which were carried out in the Chemical Processing Facility (CPF). The fabrication and irradiation conditions of the dissolved fuels were also put in order.
Wada, Hiroaki
JNC TN8400 2000-015, 37 Pages, 2000/03
This report describes the study done within the period of time when I was postdoctoral research worker at Japan Nuclear Cycle Development Institute. The report includes two parts as follows. (1) Exact Solution of Electric Transitions for High Energy photons. Technologies for creating high-energy 
beams have been rapidly developed. These advancements make the research using high-energy -rays more important. The electric transition rates for high-energy -rays were formulated. The electric multipole fields were treated strictly in the process of calculating the electric transition rates and the nuclear states were taken as the harmonic oscillator wave functions. (2) Production of the isomeric state of 
Cs in the thermal neutron capture reaction 
Cs(n, )Cs. In order to obtain precise data of the neutron capture cross section of the reaction Cs(n, )Cs, the production probability of isomer state 
Cs was measured in this work. The 1436 keV -ray emitted from both of 
Cs and Cs was measured. A production ratio of Cs to (Cs and Cs) was deduced from time dependence of peak counts of 1436keV -ray. The probability of the production of CS was obtained as 0.75
0.18 and this value revised the effective cross section upwards 92%. The effective cross section 
and the thermal neutron capture cross section 
were obtained as =0.290.02 b and =0.270.03 b with taking into account the production of 
Cs.
; Koyama, Tomozo; Funasaka, Hideyuki
JNC TN8400 2000-014, 78 Pages, 2000/03
JNC-TN8400-2000-014.pdf:2.13MB
We investigated the factors which affected the dissolution of U and Pu to the nitric acid solution with the fragmentation model, which was based on the results of dissolution experiments for the irradiated fast reactor fuels in the Chemical Processing Facility(CPF). The equation that gave the fuel dissolution rate was estimated with the condition of fabrication (Pu ratio (Pu/(U+Pu))), irradiation (burn-up) and dissolution (nitric acid concentration, solution temperature and U+Pu concentration) by evaluating these effects quantitatively. We also investigated the effects of fuel volume ratio to the solution in the dissolver, burn-up and flouring ratio of the fuel on the f-value (the parameter which shows the diffusion and osmosis of nitric acid to the fuel) in the fragmentation model. It was confirmed that the fuel dissolution rate calculated with this equation had better agreement with the results of dissolution experiments for the irradiated fast reactor fuels in the CPF than that estimated with the surface area model. In addition, the efficiency of this equation was recognized for the dissolution of unirradiated U pellet and high Pu enriched MOX fuel. It was shown that the dissolution rate of the fuel slowed down at the condition of the high U-Pu concentration dissolution by the calculation of the dissolution behavior with this equation. The dissolution of the fuel can be improved by increasing the nitric acid concentration and temperature, but from the viewpoint of lowering the corrosion of the dissolver materials, it is desirable that the f-value is increased by optimizing the condition of shearing and stirring for the improvement of dissolution.