Vol 66, No 4 (2024)

The work examines the state of ¹⁰⁶Ru in solutions simulating technological media and liquid radioactive waste from nuclear power plants with a VVER reactor, as well as the influence of the physicochemical state of the radionuclide on the efficiency of ion exchange and ultrafiltration treatment methods. During the study, the regions of existence of ionic and nonionic states of the radionuclide were studied in solutions of complex composition. In addition, the size distribution of particles containing ¹⁰⁶Ru were established. The formation of complex compounds of ¹⁰⁶Ru with polyborate ions in the neutral and slightly alkaline pH region has been shown, which leads to a decrease in the efficiency of decontamination of liquid radioactive waste from ruthenium radionuclides.

The thermal stability of dibenzo-21-crown-7 (DB21C7) and its solution in 1,2-dichloroethane (DCE) after contact with nitrogen dioxide, which is a product of nitric acid degradation, was studied. It is shown that exothermic processes accompanied by the release of gaseous products are observed in inert and oxidizing atmosphere. The compounds formed after contact with nitrogen dioxide were determined by gas chromatography–mass spectrometry. Differences in gaseous products in the temperature range of 300°С were noted for the initial crown ether and for DB21С7 and its solution in DСE after contact with NO₂.

The optimal conditions for the synthesis of a granular sorbent based on mixed Mn(III, IV) oxide by the interaction of aqueous solutions of MnSO₄ and KMpO₄ in an alkaline medium were determined: the molar ratio Mn²⁺/MnO₄ is 1.70–1.80; the pH of the reaction mixture is 11.0–12.5; the calcination temperature is 220°C. For the sorbent obtained under optimal conditions, the values of the distribution coefficient (K_d) ⁹⁰Sr in 0.01 M CaCl₂ solution, the static exchange capacity for calcium, the hydromechanical strength of granules, as well as the dependence of K_d ⁹⁰Sr on the concentration of sodium and calcium ions were determined. It is shown that the resulting sorbent has higher sorption characteristics with respect to strontium compared with known sorbents. A technology has been developed for the production of pilot batches of sorbent, named MDM. Examples of the use of MDM sorbent for the purification of various types of liquid radioactive waste from strontium and radium radionuclides are given.

An automated method has been developed for the synthesis of [N-methyl-¹¹C]choline, a radiopharmaceutical (RP) for the diagnosis of cancer using positron emission tomography (PET). The synthesis was carried out on a home-made module, using combined technology of on-line ¹¹C-methylation processes and solid-phase extraction methods. The radiochemical yield of [N-methyl-¹¹C]choline was 80% (based on the activity of the methylating agent, [¹¹C]CH₃I, decay corrected), which ensures the production of several clinical doses of radiopharmaceutical in one batch. [N-methyl-¹¹C]choline was obtained with a radiochemical purity of more than 99% and an amount of 2-dimethylaminoethanol (the main chemical impurity) of 0.06 mg/mL, which meets the requirements of the Russian and European Pharmacopoeia.

Radiation resistance of diesel fuel with the addition of various percentages of toluene was studied. The experiments were carried out for a long time to study the postpolymerization processes. The kinetics of processes during irradiation of pure diesel fuel was studied at the temperature T = 20°C and dose rate P = 0.07 Gy/s in the range of absorbed doses D = 15–150 kGy, and a mixture of toluene with diesel fuel was irradiated within the absorbed dose range D = 24–90 kGy at a toluene concentration of 1. 3. and 5 vol %. Analysis by gas chromatography–mass spectrometric (GC/MS) was performed, and the density, viscosity, and iodine number of the diesel fuel before and after irradiation at various absorbed doses were determined. The kinetics of postpolymerization processes after the end of irradiation shows that the rate of the process and its share in the total polymerization depend on the irradiation time, initial mixture density, and dose. By adding additives (antirads), one can choose the composition of diesel fuel that will better withstand radiation exposure. It is necessary to find the optimal concentration of toluene in the composition of diesel fuel, at which the viscosity and density will not change with an increase in the absorbed dose.

The content of global ^239,240Pu in Caspian sea water (1996–2056) was calculated using the sorption–diffusion model of the radionuclide uptake by bottom sediments with the distribution coefficient K_d = 50 × 10³ and diffusion coefficient D = 0.1 × 10^–7 cm²/s. The ^239,240Pu global fallout on the sea was assumed to be equal to the experimental value for the mid-latitude belt of Russia, 60 Bq/m². At the plutonium fallout density on the sea surface of 58 Bq/m², its inventory in the sea Q as of the year 1964 is 21.9 TBq. In 1996, the experimentally determined ^239,240Pu concentration in Caspian sea water was ~20 μBq/L, and the calculation by the model gives 17.8 μBq/L. During the ~ 30-year migration of global ^239,240Pu, about 93% of the radionuclide passed from the aqueous phase to bottom soils. According to the calculations, the ^239,240Pu concentration in the seawater in the period 1996–2056 will decrease from 17.8 to 10.5 μBq/L, and the ^239,240Pu inventory in the seawater, from 6.3 to 3.7%, of the fallout value. The results were verified using an independent method for ⁹⁰Sr monitoring in Caspian sea water and determining the ^239,240Pu/⁹⁰Sr concentration ratio in the water. The results of estimating the ^239,240Pu concentrations in the seawater (2017–2020) by these two methods reasonably agree with each other.