Radiochemistry最新文献

The reaction of nitric acid with formic acid, including that in uranium-containing solutions, was studied. Empirical equations for the dependence of the induction period (τ) on the reactant concentrations and temperature were determined. The dependence of τ on the formic acid concentration is exponential. The effect of uranium concentration on τ in denitrated solutions becomes noticeable only at temperatures lower than 60°C. The main factor affecting the denitration completeness is the [formic acid] : [NO₃]¯ molar ratio. Starting from the molar ratio of ≥3, uranium-containing solutions are denitrated at 90°C in an hour almost quantitatively. The resulting uranyl formate is partially precipitated. The initial stage of the reaction is accompanied by vigorous gas evolution. At 90°C, ~80% of the gas volume is released in this stage (about 10 s), whereas at 40°C, only ~10%.

The composition of the gas phase over uranium–plutonium nitride U_0.8Pu_0.2N_0.9950_0.005 containing oxygen impurities and fission products after fast neutron irradiation to a burnup of 13.6% heavy atoms (h.a.) was subjected to a thermodynamic analysis in the temperature interval 900–2000 K. In this interval, the accumulation of fission products in the fuel leads to the formation of a multicomponent gas phase containing the following main elements and compounds: fuel species (Pu, PuO, PuN, U, UO, UN), highly volatile species (Cs, N₂, Ba, Sr, CsI, BaI, SrI, I), volatile species (Te, Pd, BaO, NdO, LaO, SrO, CeO), low-volatile metals (Nd, Mo, Y, Tc, La, Ce, Zr), and low-volatile nitrides (NdN, LaN, CeN, YN, ZrN). The partial pressures of these components of the gas phase over U_0.8Pu_0.2N_0.9950_0.005 at the 13.6% burnup as functions of temperature were calculated.

The oxidation of UN and UC in a field of microwave radiation in air at atmospheric pressure was studied. The influence of the crucible material on the extent of oxidation of UN and UC was examined. Under the influence of a microwave field with a power of 800 W and a frequency of 2.45 GHz, under definite conditions, heating of UN and UC to ~993 K is observed with their oxidation in air to U₃O₈. For fast (15–20 min) and safe (without fires and explosions) oxidation of UN and UC, crucibles made of quartz and carbon ceramics are the best.

Experimental data on the extraction of nitric acid, europium, and americium with 40% TBP in C13 in the presence of salting-out agents such as sodium, magnesium, iron, and aluminum nitrates were obtained. The nitric acid and nitrate ion concentrations ensuring efficient recovery of transplutonium and rare earth elements were chosen. Based on the data obtained, the existing program for mathematical modeling (Statics, Dynamics program complexes) was adjusted. To check the adequacy of program operation, computations were made for the experimentally tested flowsheet of the high-level raffinate partitioning using the extraction system based on TBP in paraffin in the presence of a salting-out agent. Comparison of the calculation results with the results of cell-by-cell analysis of samples from the first block shows that the deviations are within the measurement uncertainty of the analytical equipment. The possibility of predicting the operation of the extraction cascade at varied parameters (phase flow rates, number of steps, salting-out agent concentration) was demonstrated.

The solubility of plutonyl(VI) and uranyl(VI) nitrates in nitric acid solutions taken in the ratios Pu/(Pu + U) = 0.3 ± 0.015 (main data), Pu/(Pu + U) = 0.25 ± 0.025, and Pu/(Pu + U) = 0.5 ± 0.025 (single compositions) at +25, +15, and +5°C (±3) was determined by the method of isohydric crystallization. The calculated solubility isotherms of plutonyl(VI) and uranyl(VI) nitrates are presented in a wide range of nitric acid concentrations.

The sorption treatment of aqueous solutions to remove TBP on the Stirosorb polymer sorbent in the batch and dynamic modes was studied. Stirosorb polymer sorbent allows removal of TBP from aqueous media at a solution temperature not exceeding 70°C to the residual organic matter content of no more than 10–11 mg/dm³ at a load of ~0.9 g TBP/g sorbent. A solution containing U(VI) and Pu(IV) is decontaminated from phosphorus-containing compounds on Stirosorb sorbent without solution depletion of U(VI) and Pu(IV).

The possibility of introducing deuterium into dopaquine, a compound consisting of quinone and dopamine fragments, was examined. The isotope exchange can be carried out with deuterated water in the presence of trifluoroacetic acid (TFA) and HCl. The best results were obtained with TFA at 80°С. To activate isotope exchange, catalysts based on iridium or palladium (5% Pd/Al₂O₃, cycloocta-1,5-dienyliridium(I) 1,1,1,5,5,5-hexafluoropentane-2,4-dionate) were added to the reaction mixture in addition to the acid component. The preparative synthesis of [D]dopaquine was carried out by isotope exchange with deuterated water (D₂O : TFA 5 : 1, 80°С, 3 h). [D]Dopaquine was obtained in 20% yield with the deuterium content in the interval 1.2–1.4 atoms per molecule. The deuterium content in the dopamine fragment is 1.2–1.3 atoms per molecule. The quinone fragment contained approximately 4–7% of the isotope.

A digital model was developed and used for simulating the discharge of a borosilicate glass melt via the inductively heated drain of the removable small-size heater designed by the Mayak Production Association. The thermal calculation of the small-size furnace for high-level waste vitrification with borosilicate glass in the course of glass melting and discharge was made, and the calculation results were compared to the experimental data. The maximal difference in the absolute values of the temperature of furnace parts in the places of the thermocouple installation does not exceed 100°С, which corresponds to approximately 10% of the acting temperatures. Taking into account the assumptions made, this result can be considered satisfactory.

The results of the development and implementation of a tangential filtration procedure for the treatment of mother liquors from ammonia precipitation of uranium are presented. The tangential filtration installation at RT-1 plant was put into operation in 2019. The productive capacity of the installation with respect to permeate is 450 dm³/h on the average. Up to 50 m³ of the mother liquor can be processed in one filtration cycle. The decontamination factor is 10³ for α-emitters and 10² to 10³ for β-emitters. The resulting permeate corresponds in its activity to the LLW category.

Procedures were developed for measuring and calculating the radionuclide composition of the cesium–strontium fraction of radioactive waste from spent nuclear fuel reprocessing. The optimum conditions for recovering specific components and separating them from components interfering with their determination were found. The content of the following controlled radionuclides was determined experimentally: ¹⁴С, ⁹⁰Sr, ⁹⁹Tc, ¹²⁶Sn, ¹²⁹I, ^135,137Cs, ^{234–236,238}U, ^238–242Pu, ²³⁷Np, ^241,243Am, and ^244–246Cm. The content of ⁶³Ni, ⁹³Zr, ⁹⁴Nb, ¹⁵¹Sm, and ^242mAm was determined by calculations. The radionuclide composition was determined in the solution before solidification of the cesium–strontium fraction. The composition of the solidified cesium–strontium fraction in the form of the borosilicate glass was determined by the calculation from the activity concentrations of the radionuclides in the solution before solidification and process characteristics (solution volume, glass frit weight, separation factors, and factors of incorporation into the glass).