Radiochemistry最新文献

The interaction of U(VI) with natural and synthesized hydroxyapatite Ca₅(PO₄)₃(OH) is of interest for the retention of uranium migrating in the environment from various sources of anthropogenic pollution. Another potential application of hydroxyapatite is its use as a carrier for radiopharmaceuticals. In this case, in addition to the parameters of the uranium sorption onto the sorbent surface, it is necessary to study the influence of the structural features of hydroxyapatite nanoparticles on the ²³⁰U retention on the sorbent surface. This study deals with the kinetic parameters and mechanisms of uranium sorption onto three textural modifications of hydroxyapatite, with the nanopID morphology before and after sorption, and with the strength of uranium retention on hydroxyapatite. The patterns obtained are of interest from both medical and radioecological viewpoints.

The possibility of using activated needle petroleum coke (NPC), a carbon nanomaterial with the specific surface area of approximately 2500–3500 m²/g, as a sorbent for recovering radionuclides from aqueous solutions was examined. The physicochemical properties of NPC and the distribution coefficients (K_d) of actinides, fission products, and radium on NPC samples in nitric acid, neutral, and weakly alkaline solutions, including those with high salt content, were determined. High salt concentrations favor an increase in K_d, which is untypical of carbon nanomaterials. NPC shows promise for recovering and concentrating uranium, americium, and rare earth element isotopes from contaminated groundwater, river water, and seawater and from solutions of neutralized liquid radioactive waste.

The efficiency of the lanthanide(III) extraction from nitric acid solutions with bis(diphenylphosphorylmethylcarbonylamino)alkane (bis-CMPO) solutions in organic solvents significantly increases in the presence of an ionic liquid, benzyltriethylammonium di-2-ethylhexyl sulfosuccinate ([N_222Bn]SSu), in the organic phase. The lanthanide(III) distribution ratios in the extraction from 3 M HNO₃ solution with bis-CMPO solutions in dichloroethane increased by more than two orders of magnitude in the presence of [N_222Bn]SSu. Lanthanide(III) ions are extracted with mixtures of bis-CMPO and the ionic liquid in organic solvents via the cation-exchange mechanism. The influence of the structure of bis-carbamoylmethylphosphine oxides, kind of organic diluent, and HNO₃ concentration in the aqueous phase on the efficiency of the metal ion extraction into the organic phase containing the ionic liquid is discussed.

The results of long-term (2015–2025) monitoring of dissolved ¹³⁷Cs activity concentration in water of Inkyozaka pond located in close vicinity of the Fukushima Daiichi NPP are presented. Regular seasonal variations of dissolved ¹³⁷Cs in pond water were observed concurrently with seasonal fluctuations of water temperature, which confirms their relation to the temperature dependence of the radiocesium desorption from sediments to solution. The activation energy of radiocesium desorption, obtained from the observational data, is 18.8 kJ/mol, which is in agreement with similar laboratory and field investigations. An empirical exponential model and a semiempirical diffusion model of long-term dynamics of dissolved ¹³⁷Cs in water of Inkyozaka pond were suggested. The modeling results qualitatively describe the seasonal fluctuations of the dissolved ¹³⁷Cs concentrations in the pond water.

The efficiency of the high-temperature treatment (HTT) of mixed uranium–plutonium nitride spent nuclear fuel (MUPN SNF) to separate the fuel compound from the cladding and remove volatile fission products from the fuel was evaluated. The high-temperature treatment of SNF allows separation of the fuel compound from the cladding. The degree of removal of cesium nuclides from the fuel compound does not exceed 7%, whereas tritium, radiocarbon, and ruthenium are removed from the fuel compound to 94.2, 89.9, and 95.5%, respectively, relative to their initial content in SNF.

Procedures for decontamination of aramid-based personal protective equipment from combined alpha and beta contamination were developed at the Mayak Production Association. Only decontamination formulations containing reductants capable of reducing plutonium to the lower oxidation state (III) appeared to be effective. It was demonstrated that after five decontamination cycles based on the developed procedures, physicomechanical and protective properties of the aramid fabrics remained unchanged.

Thermodynamic modeling of changes in the chemical and phase composition of uranium–zirconium carbonitride nuclear fuel U_0.9Zr_0.1(C_0.495N_0.495O_0.01)_0.995 containing an oxygen impurity depending on the temperature (1600–2000 K) and burnup (1–10%) was performed. Under these conditions, accumulation of fission products in the fuel leads to the formation of a heterogeneous system consisting of a solid solution based on uranium–zirconium oxycarbonitride containing lanthanides and of separate condensed carbide, intermetallic, and oxide phases, of cesium and its iodide, of barium and its telluride, of strontium, and of technetium. The concentrations of these phases at the indicated temperatures and burnup levels were calculated.

The paper deals with the specific features of the operation of molten-salt reactors (MSRs) as applied to the off-gas treatment systems for the localization of the main radiotoxic nuclides of iodine, tritium, and radioactive noble gases (RNGs). The domestic off-gas treatment system of a research MSR (RMSR), adopted by the Rosatom State Corporation, is compared to that of the Oak Ridge National Laboratory (ORNL, the United States, MSRЕ project). It is shown that the domestic research and development involved in the RMSR gas treatment strategy surpasses the world’s level owing to quantitative localization of molecular tritium from the gas phase, followed by its localization. This ensures radioecological safety of the reactor facility operation. The production rate of hazardous gaseous radionuclides (tritium, radioactive noble gas isotopes) in RMSR is calculated.

The possibility of the ¹³⁷Cs recovery by sorption onto ferrocyanide sorbents using a crossflow membrane filtration installation was examined. The steps of the sorption, washing, desorption, and sorbent regeneration were developed. This approach allows using the unsupported sorption material (nongranulated sorbent). Under comparable conditions, the sorption onto the nongranulated sorbent using a crossflow filtration installation occurs more efficiently than in the classical process of the column sorption onto the granulated sorbent. Namely, the process kinetics is better, the capacity is higher, and the effective filtration cycle is potentially higher. The contamination of the target cesium concentrate with products of the sorption material degradation is excluded.

Uranium is extracted, separated, and recovered using 2,8,14,20-tetramethyl-11,23-dinitrocavitand-5,17-dihydroxamic acid (NMCDH) in the presence of diverse ions. Uranium forms a yellow complex with NMCDH in n-octanol with the maximum absorbance at 450 nm and molar absorptivity of 5.0 × 10⁴ L mol^–1 cm^–1. It follows Beer’s law in the concentration range of 0.48–5.76 µg mL^–1. The extracted complex is directly analyzed using ICP-MS with considerably higher sensitivity (detection limit 0.5 ng mL^–1). The extraction constants of the uranium(VI)–NMCDH complex are determined. This method enables recovery of 99.95% pure uranium from monazite sand and phosphate rock; it is also effective for preconcentrating and detecting uranium in environmental samples.