Radiochemistry最新文献

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.

Prerequisites for the development and examples of pilot-scale and industrial installations for combined electrochemical and ultrasonic decontamination of metal surfaces (fuel rods, other items, fragmented metal waste), implementing the process developed by Alexandra Plus LLC and St. Petersburg State Institute of Technology, are considered. The installations, their main parameters, and operation results are described. The combined decontamination is advantageous in the efficiency and treatment rate over conventional immersion in decontaminating solutions or solely ultrasonic treatment.

This study investigates the natural radioactivity of ⁴⁰K, ²²⁶Ra, and ²³²Th in valley sediments from Wadi Al-Maia located in Egypt’s Eastern Desert. Radiation monitoring was conducted using high-resolution gamma-ray spectrometry with a NaI(Tl) detector. The activity concentrations of ²²⁶Ra, ²³²Th, and ⁴⁰K were measured. Based on these measurements, several radiological hazard parameters were calculated, including the absorbed dose rate (ADR), radium equivalent activity (Ra_eq), gamma index (I_γ), annual gonadal dose equivalent (AGDE), annual effective dose equivalent (AEDE_out and AEDE_in), and the external and internal hazard indices (Hₑₓ and Hᵢ_n).The average values of these parameters were 347 nGy/h (ADR), 724 Bq/kg (Ra_eq), 2.73 (I_γ), 2471 μSv/year (AGDE), 425 μSv/year (AEDE_out), 1700 μSv/year (AEDE_in), 1.96 (H_ex), and 2.47 (H_in), all of which exceed the worldwide recommended limits. These results indicate that the study area exhibits elevated natural radioactivity levels and potential radiological health risks. The findings can serve as a baseline for future monitoring of natural radionuclide contamination and radiological risk assessment in the Eastern Desert of Egypt.