[en] Full text: In the processing of monazite ore to separate uranium, thorium and rare earth elements from the ore, uranium is produced in the form of ammonium diuranate or yellow cake. This yellow cake requires further purification to remove thorium and rare earth impurities from the cake. The purification of uranium by solvent extraction was investigated using 5, 10 and 15% tributyl phosphate (TBP) in kerosene and 3, 5 and 10% di(2- ethylhexyl) phosphoric acid (D2EHPA) in kerosene as the extractants. The prepared uranium feed solution was in 4 N HNO₃ with the presence of thorium and cerium as the impurities. The distribution ratio of uranium in D2EHPA was found to be higher than that of TBP and the distribution ratios in both extractants increased with the increasing concentration of extractants. For the feed of uranium concentration less than 20,000 mg/L, the extraction efficiencies of 10% D2EHPA and 15% TBP were higher than 90%. The separation factor of U-Th in TBP was in the average of 20-50 while this factor in D2EHPA was lower than 0.1. The separation factors of U-Ce in both extractants in the average were higher than 100. Using nitric acid of 1.0 - 2.5 N as the scrubbing solution, the purity of the extracted uranium in the TBP extractant could be further increased

[en] Room temperature ionic liquids are being investigated as possible substitutes to high-temperature molten salts in non-aqueous reprocessing application. The head-end process to the oxide electrowinning involves the dissolution of the spent oxide fuel in molten salt medium. This was usually carried out by chlorination of the oxide fuel present in high temperature molten salt or ionic liquid medium. However, such procedure is usually slow, tedious and incomplete in ionic liquid medium. To facilitate the dissolution of uranium oxide, we explored the feasibility of adding a small quantity of nitric acid to ionic liquid (1-butyl-3-methylimidazolium chloride (C_4mimCl m.p 340 K)) medium for the first time, and studies on the electrochemical behavior of uranium in the resultant solution, which is also unknown. The results obtained from dissolution of U_3O_8 and electrochemical behavior of U(VI) in C_4mimCl has been reported in this paper

[en] This paper described the method of determination for O/U ratio of UOsub(2+x) sample in which the amount of U(IV) and total uranium amount by constant coulometric titration. Using this method, the O/U ratios of powdered uranium oxide, fuel pellet and doping pellet of Gd₂O₃-UO₂ were determined. It is found that O/U ratio for U₃O₈ is 2.666+- 0.003 and the precision is 0.11%. The deviation between six O/U ratios obtained by different methods is discussed. (Author)

[en] It is presented the data of facilities of atomic energy complex of the Republic of Kazakstan. At the present time Republic of Kazakstan has 7 uranium mines, 2 plants producing uranium oxide (Aktau and Stepnogorsk), 1 plant on processing and production of fuel pellets for WWR-K and RBMK reactors (Ulbinskij metallurgical plant, Ust-Kamenogorsk), 1 atomic power plant with BN-350 reactor which is in the frame of Mangyshlak Atomic Energy Complex (MAEC) and National Nuclear Centre. Atomic power plant with BN-350 reactor has been producing of electric power up to 125 MWt and fresh water up to 10000 ton/day and night. Common supplies and resources of uranium in Kazakstan composed 1168000 ton. Common volume of uranium mining in Kazakstan in 1992 composed 3000 ton. The only in the Former Union production of beryllium seized all technological process ranging from processing of concentrates to getting of final production. The Republic of Kazakstan has all conditions for creation of developed atomic energetics

[en] Oxidation process of sintered (UO_2+ZrO_2) pellets in order to obtain sintered pellet oxidation data to produce U_3O_8 powder has been carried out as a preliminary study of fuel recycle simulation with AIROX process. The sintered (UO_2+ZrO_2) pellets with variation of ZrO_2 concentration from 0 – 1% were heated at temperature 300 - 500 °C for 0.5 – 2 hours to produce U_3O_8 powder. From the process above, the oxidation efficiency and the powder characterization, i.e. the density, were determined. The results showed that the oxidation process occurred at temperature above 400 °C after 0.5 hour in which grayish black U_3O_8 pellets had changed into brownish black U_3O_8 powder. Oxidation efficiency of 100% was achieved after 2 hours at ZrO_2 concentration of 0.4%, while the highest density was obtained at ZrO_2 concentration of 0.2%. (author)

[en] Microwave processing is emerging as an innovative heating method for many traditional ceramics, advanced ceramics and ceramic composites. The perceived advantages of microwave heating over conventional heating include expectations for uniform heating, better properties of the product, greater throughput, low processing time and greater energy efficiency. A number of actinide and rare earth oxides useful in nuclear industry have been interacted with microwaves of frequency 2450±50 MHz to see the feasibility of observable heating effects. This study will explore scientific understanding in development of microwave processing of the oxides involving drying, calcining, binder burnout, annealing, joining, melting, sintering, powder synthesis, dissolution etc. The materials considered were dried by microwave heating to constant weight ensuring quantitative removal of moisture before making experimental observations. The study was carried out by varying the incident microwave power flux, amount of the oxides, interaction time and observable temperature differences were measured precisely. The rise in the temperature in oxides as U₃O₈, UO₂, and PuO₂ under the experimental conditions was significantly large. On the other hand, ThO₂, and rare earth oxides La₂O₃, Gd₂O₃, Sm₂O₃ and Dy₂O₃ showed insignificant temperature rise. The order of rise in temperature of the various oxides was expected as per their material dielectric properties (ε'' dielectric constant, ε'' dielectric loss factor and loss tangent tanδ = ε''/ε'). However it is to be noted that the rise in temperature depends upon many other factors such as the incident microwave power, microwave frequency, amount of the substance and interaction time, insulation, thermal conductivity of the oxide, variation of dielectric properties with temperature and the critical temperature (T_c). ThO₂ and rare earth oxides under investigation have shown weak microwave absorption behavior. (author)

[en] Experimental results of U₃ O₈ and U O_2+x (0.106 ± x ± 0.28) pellets microwave sintering are presented, as well as indications on the effects of sample configurations, initial densities, efficiency of atmosphere purification system and thermal insulation, related to the final density values. The furnace, operating at 650 W and 2.45 GHz, with uranium oxide pellets inside a quartz tube or aluminium crucible, with additional Fiberfrax insulation, led the material up to the intermediate sintering stage, the final stage reached sintering U O₂. 106 18 minutes, in commercial purified nitrogen. The adopted experimental procedure has semi-industrial application potential. (author)

[en] The performance evaluation studies of three indigenously manufactured amperometers were carried out for uranium determination in U₃O₈ standard solution. Three analysts independently carried out ten determinations each of uranium by Biamperometric end point detection method with each of these amperometers. Two-way ANOVA test was performed on the data using MATLAB-7.0 software to test whether the variation in the observed data is statistically significant. No significant bias in the data produced by three analysts with three independent instruments was observed. The results obtained show that the indigenous amperometers are suitable for uranium determination in fuel materials and for nuclear material accounting, with high precision and accuracy. (author)

[en] Oxidation-reduction process of uranium oxide in a number of cycles has been investigated to determine the influence of oxidation-reduction cycle on the density and grain of U_3O_8 and UO_2 powders obtained from nuclear fuel recycling. In the first cycle, UO_2 powder obtained from the oxidation-reduction process of UO_2 sintered pellets was heated at various temperatures and time to form U_3O_8. The U_3O_8 powder was then reduced at temperature 850 °C for 2 hours in hydrogen atmosphere to convert it back to UO_2. The oxidation-reduction process was the second cycle. The UO_2 powder obtained was reoxidized to form U_3O_8, and the U_3O_8 was reduced back to UO_2 in the third cycle. In the fourth cycle, UO_2 obtained from the third cycle was oxidized to form U_3O_8. The densities and photographs of the two types of powders as the products of oxidation-reduction process were obtained. The results showed that the cycles of oxidation-reduction of uranium oxide influence the density and grain of U_3O_8 and UO_2 powders. In the third cycle, the powder density of U_3O_8 as the oxidation product is higher than those obtained in the first, second and fourth cycle. Similar trend prevails for UO_2 powder, the reduction product, up to the third cycle. Meanwhile the grains of both U_3O_8 and UO_2 powders tend to crowd together in the first cycle but then become smaller and separate in the following cycles. (author)

[en] Radiochronometric dating of yellow cake has been carried out. The aim of this study was to obtain data on the age of yellow cake for compiling a database for nuclear forensics library. The radiochronometer used was ²²⁸Th/²³²Th. ²³²Th is thorium which naturally co-exists in yellow cake and will reach a decay equilibrium to ²²⁸Th. In the yellow cake production process, however, the equilibrium is disturbed so that the 'time' will be zeroed again. The test samples were yellow cake from Cogema, from PTBGN mining in Kalan, and by-products from Petrokimia Gresik. The steps of processes include dissolving the sample, separating thorium from uranium by cation exchange chromatography using dowex 50W-X8 resin, electrodeposition of thorium and isotopes analysis using an alpha spectrometer. Age determination was performed by calculating the ²²⁸Th/²³²Th activity ratio and decay formula iteration. Based on the analysis of the ²²⁸Th and ²³²Th isotopes using an alpha spectrometer and the results of age calculations, the age of Cogema yellow cake was not able to be determined because it does not contain ²³²Th. The age of Petrokimia Gresik yellow cake is 9.90 years with a bias of 20.35 years and the age of PTBGN yellow cake is 12.85 years with a bias of more than 20.15 years when compared to the known estimated production time. The bias obtained is higher when compared to the results determined by previous researchers by the same method, i.e., a bias of less than 10 years. (author)