Volume 59, Nº 4 (2017)

Formulation of the forming paste for preparing unsupported granulated triuranium octoxide U₃O₈ was developed. The mixture contains U₃O₈ and (NH₄)₂U₂O₇ in a weight ratio from 1: 1 to 1: 1.5, and also water in an amount of 25–40 wt % relative to the sum of the dry components. Calcination of crude granules at 800–1100°С results in formation of strong ceramic granules with 57–45% open porosity, consisting of U₃O₈. Granulated U₃O₈ efficiently catalyzes the conversion of ammonium nitrate to nitrogen in an NH₄NO₃–H₂O vapor–gas mixture with the addition of ammonium carbonate. The absence of support makes the catalyst composition more definite and simplifies its regeneration.

Physicochemical principles of preparation of U(VI) carbonate solutions in the step of oxidative dissolution of U₃O₈ and UO₂ in the Carbex process are considered. Carbonate solutions with the U(VI) concentration higher than 100 g L^–1, suitable for subsequent final purification of uranium by extraction, can be prepared under the conditions of formation of U(VI) carbonate–peroxide complexes in the course of dissolution with prevention of hydrolysis of U(VI) compounds. The behavior of impurities simulating some fission products in the course of oxidative dissolution was studied, and the decontamination factors of U(VI) from the chosen simulated fission products were determined.

N,N,N′,N′-Tetra-2-ethylhexylglutaramide (TEHGA) was used as a new extractant for the extraction of U(VI) and Th(IV) from nitric acid solutions. Toluene was found to be the most suitable diluent for TEHGA. The extraction of nitric acid was also studied. The influence exerted on the distribution ratio (D) of U(VI) and Th(IV) by the concentrations of HNO₃, TEHGA, and LiNO₃ as salting-out agent, and also by the equilibration time, temperature, and kind of diluent was examined. Good U–Th separation can be achieved using 2–3 M HNO₃. The results obtained show that U(VI) and Th(IV) are mainly extracted as UO₂(NO₃)₂·2TEHGA and Th(NO₃)₄·TEHGA, respectively. The IR spectra of the extracted species were analyzed. The thermodynamic functions of the process were calculated. Back-extraction of U(VI) and Th(IV) from the organic phases was also studied.

The effect of γ-irradiation of tert-butylthiacalix[4]arene (TCA) solutions in m-nitrobenzotrifluoride (NBTF) and tetrachloroethylene (TCE) on the extraction of ²⁴¹Am from alkaline carbonate solutions was studied. TCA itself remains stable upon γ-irradiation of its solutions in NBTF to a dose of 200 kGy, but the diluent undergoes strong degradation. The radiation resistance of TCA in TCE is considerably lower: A dose of 70 kGy causes complete degradation of TCA. In the TCA–TCE–aqueous phase system, sulfate ions appear upon γ-irradiation as the final product of the extractant radiolysis. A large number of γ-radiolysis products of TCE and TCA were detected by HPLC and GCMS. The products of radiolysis of TCA in TCE, compared to the initial extractant, have lower molecular mass and higher polarity. The results show that chlorinated diluents are not promising diluents for thiacalixarene in extraction processing of alkaline high-level waste.

A novel procedure was suggested for studying the phase distribution of radioactive iodine in the liquid–gas system. Without redox reactions, the degree of iodine transfer into the gas phase is maximal for sulfuric and nitric acid solutions at pH ≤ 1.5. In weakly acidic, neutral, and alkaline solutions, the iodine transfer into the gas phase is insignificant (degree of transfer <0.05). Conditions for stabilization of iodine species (presence of redox reagents, solution acidity) were found. The most efficient purification of ⁹⁹Мо to remove iodine radionuclides is observed in sulfuric acid solutions, where the probability of formation of IО₃^– anions, exhibiting the highest affinity for titanium hydroxide, is minimal. Silver-modified Termoxid sorbents are suitable for selective sorption of iodine only from alkaline solutions (0.3–0.5 M NaOH), with the best result obtained for Т-5(Ag) (iodine distribution coefficient 2000–3000 mL g^–1). An additional step of purification of alkaline molybdenum concentrates to remove iodine on Т-5(Ag) sorbent was suggested; it allows the residual iodine content of the molybdenum concentrate to be decreased by a factor of 20.

Esomeprazole was labeled with ^99mTc in high (up to ~98.0%) radiochemical yield. The optimum conditions are as follows: pH 8, 50 μg of SnCl₂·2H₂O, 30 min, and 2 mg of the substrate. The complex is stable for 8 h. The reaction mixture was separated by gel chromatography using such eluents as NaCl solution and, phosphate, citrate, and carbonate buffer solutions. Free ^99mTcO₄^– and the complex were also efficiently separated by reversed-phase HPLC, paper chromatography, and electrophoreses. Intravenous biodistribution studies of ^99mTc-esomeprazole complex showed high uptake in the stomach ulcer, reaching about 30.5% ID/g at 1 h post injection. Such a high ^99mTc-esomeprazole uptake makes this agent promising for stomach ulcer imaging.

The main sources of formation of liquid radioactive waste (LRW) containing seawater are determined, and the main problems arising in management of such waste are analyzed. Sorption methods for removing long-lived Cs and Sr radionuclides from highly mineralized (>1 g L^–1) LRW are determined. The main physicochemical and sorption characteristics, advantages, and drawbacks of candidate sorbents for removing Cs and Sr radionuclides are described. Examples of using SRM and VS-5 chemical reaction sorption materials developed for removing Sr from LRW with the mineralization of up to 60 g L^–1 are given. The results of studying composite materials based on BaSiO₃ and resorcinol–formaldehyde resins, intended for removing Cs and Sr radionuclides from seawater, are analyzed. Composite sorbents of such type efficiently remove Cs and Sr radionuclides from seawater. Processes developed by the authors and brought into practice at various plants of the Far East for treatment of multicomponent LRW formed in the course of operation, repair, and decommissioning of nuclear-powered surface ships and submarines are described.

The parameters of the diffusion–sorption interaction of rocks with Sr, Cs, U, Np, Pu, and Am isotopes dissolved in simulated groundwater were determined by in- and through-diffusion methods for monolithic rock samples (gneisses, dolerites, lamprophyres) from the Yeniseisky site of the Nizhnekansy massif. Despite heterogeneous mineral composition and structure of the rocks, these parameters mainly depend on the physicochemical properties of the element and on the open porosity of the rock. The effective diffusion coefficients D_e vary in the order Sr = Cs ≫ U > Np > Am ≥ Pu, dividing these elements into two groups, the difference between which is 1.5–2 orders of magnitude. With respect to the apparent diffusion coefficient D_a, the elements are also subdivided into two groups, Sr = Cs > Np ≥ U ≫ Am > Pu, differing by 5 orders of magnitude because of different sorbability of the elements: Pu > Am ≫ Cs = Sr > U > Np. The end members of this series differ in the sorption capacity coefficient of the rocks, α = D_e/D_a, by 4.5 orders of magnitude.