Radiochemistry最新文献

This study was aimed to prepare immunoradiometric (IRMA) system using solid-phase magnetizable cellulose particles coupled with luteinizing hormone (LH) to determine LH in human serum. The LH-IRMA system was designed through preparation of LH polyclonal antibody (anti-LH), ¹²⁵I-LH tracer, and LH standards. Production of anti-LH was carried out through immunization of three Balb/C mice with LH-BSA immunogen. Magnetizable cellulose particles were activated using 1,1’-carbonyldiimidazole (CDI) and coupled with purified anti-LH. Optimization and validation of LH-IRMA were undertaken. The results obtained show that the local solid-phase magnetic particle LH-IRMA is a sensitive, specific, precise, and accurate method for measuring the LH concentration in human serum and can be used as a diagnostic tool for some disorders such as infertility in females and impotence in males.

The results of the sorption of ¹³⁷Cs from the clarified phase of accumulated high-level alkaline radioactive waste of the Mayak Production Association with sorbents based on modified nickel ferrocyanide (Fersal) and resorcinol–formaldehyde resin (RFS-I) in batch and flow conditions are presented. In batch mode, the distribution coefficient of ¹³⁷Cs on Fersal and RFS-I sorbents is 2300 and 730 cm³/g, respectively. In flow mode, the volume of the passed solution before the ¹³⁷Cs breakthrough for Fersal and RFS-I sorbents is 140 and 85 column volumes, and the maximum purification factor is 10⁴ and 10³, respectively. To desorb cesium from the Fersal and the RFS-I, 8 M HNO₃ and 1 M HNO₃ solutions, respectively, were used. A decrease in the sorption characteristics of the RFS-I sorbent during the sorption of cesium from high-level alkaline radioactive waste was discovered. A conclusion was made about the possibility of using the Fersal sorbent for the recovery of ¹³⁷Cs from high-level alkaline radioactive waste of the Mayak Production Association.

The extraction of the tricarbonate complex of uranyl [UO₂(CO₃)₃]^4– from aqueous solutions on clay powders from kaolin clays of the Kampanovskoye deposit and from bentonite clays of the 10th Khutor and Dinozavrovoe deposits, as well as their mixtures, was investigated. The studies were carried out with clay powders, both untreated and treated with water, 0.5 M Na₂CO₃ and NaNO₃ solutions, and 2 M NaOH solutions. It was shown that the [UO₂(CO₃)₃]^4– complex is not sorbed on clay materials from aqueous solutions under static conditions. The filtration of an aqueous solution of [UO₂(CO₃)₃]^4– through columns with clay mixtures was established to enable the extraction up to 87% of uranium from the amount passed through the column.

Nanodiamonds (NDs) are promising carriers of radionuclides in radiopharmaceuticals for the use in nuclear medicine. In this work, we investigated the influence of the properties of NDs suspensions, including the sizes of their aggregates, in aqueous solutions with different pH on their binding of medical isotopes ⁹⁰Y and ¹⁷⁷Lu and determination of conditions for fabricating promising conjugates for further in vivo studies. It was shown that sorption is influenced by the composition of the solution, which determines the properties of the NDs surface, while the forms of cations of the studied nuclides in the solution do not affect the sorption value.

The extractants diphenyl-N,N-dioctylcarbamoylmethylphosphine oxide and diphenyl-N,N-diisobutylcarbamoylmethylphosphine oxide for the SNF reprocessing were tested. The conditions for extraction and separation of uranium and TPE–REE were determined. A flowsheet was proposed for reprocessing SNF from high-temperature gas-cooled reactors. Upon dynamic testing, at least 99.9% of uranium and plutonium and at least 99.5% of americium and rare earth elements were separated. The TPE + REE and U + Pu fractions were recovered. The U + Pu fraction contained approximately 5% Am, while the TPE + REE fraction had less than 0.1% U and Pu.

In this work, we studied the sorption of scandium, whose isotopes ⁴⁴Sc and ⁴⁷Sc are explored for diagnostics and therapy in nuclear medicine, by aggregates of commercial (TAN, STP) and oxidized nanodiamonds (ox-STP) from aqueous solutions. The sorption capacity of the studied NDs for scandium was determined; 100 μg of NDs is found to be sufficient for the sorption of 1 GBq of ⁴⁷Sc, which is equivalent to the activity of isotopes used in therapy. It was shown that the supposed mechanism for the binding of Sc(III) to nanodiamond aggregates is chemisorption, and the chemical composition of the nanodiamond surface affects the sorption efficiency to a greater extent than the speciation of scandium in solution. The obtained data on Sc(III) sorption are correlated with the sizes of nanodiamond aggregates; it is shown that sorption under experimental conditions does not depend on the size of the aggregates. For the further development of radiopharmaceuticals based on ⁴⁷Sc, optimal carriers for this isotope—TAN and ox-STP—were suggested.

The article presents the results of studies on composition optimization, properties, and laboratory dynamic tests of extraction mixtures based on p-isononylcalix[6]arene for the alkaline HLW purification to remove ¹³⁷Cs. The distribution of ¹³⁷Cs among the products of the technological scheme is presented and the raffinate purification factor to remove ¹³⁷Cs are determined. It was revealed that, in the case of using nitric acid as a stripping solution, nitrosation of p-isononylcalix[6]arene is possible in the technological process. The composition of the resulting product was determined. To prevent nitrosation of p-isononylcalix[6]arene, formic acid was used in the stripping stage, which ensured stable operation of the extraction setup for 48 h while ¹³⁷Cs extraction from model alkaline HLW was at a level of 99.7%.

The possibility of tritium introduction into graphene oxide (GO) by tritium thermal activation method was demonstrated. It was established that, in order to produce the highest possible specific radioactivity, thin films of GO with a thickness of 5.6 mg/m² must be treated with tritium atoms. The experiment conducted at 77 K showed a number of advantages. GO was processed with tritium atoms, the resulting specific activity of [³H]GO reached 2.6 Ci/mg in term of the weight of the initial GO (0.7 Ci/mg after removal of the labile tritium). Specific energy release of [³H]GO with this specific activity is 22.3 W/kg, which is quite sufficient for its application as a component of a nuclear battery.