Model borosilicate glasses of four compositions with different contents of the cesium–strontium fraction are synthesized. The following properties of glass are studied: melting temperature, glass density, compressive strength, homogeneity, and thermal and chemical resistance. Based on the obtained experimental data, technical requirements are developed for samples of a solidified short-lived fraction of high-level waste (HLW).

Procedures were developed for measuring and calculating the radionuclide composition of the cesium–strontium fraction of radioactive waste from spent nuclear fuel reprocessing. The optimum conditions for recovering specific components and separating them from components interfering with their determination were found. The content of the following controlled radionuclides was determined experimentally: 14С, 90Sr, 99Tc, 126Sn, 129I, 135,137Cs, 234–236,238U, 238–242Pu, 237Np, 241,243Am, and 244–246Cm. The content of 63Ni, 93Zr, 94Nb, 151Sm, and 242mAm was determined by calculations. The radionuclide composition was determined in the solution before solidification of the cesium–strontium fraction. The composition of the solidified cesium–strontium fraction in the form of the borosilicate glass was determined by the calculation from the activity concentrations of the radionuclides in the solution before solidification and process characteristics (solution volume, glass frit weight, separation factors, and factors of incorporation into the glass).

A method is proposed for quantifying the specific activities of radioecologically significant radionuclides, including 14C, 99Tc, and 129I, in vitrified high-level wastes. The conditions for the separation and purification of these radionuclides from potential interfering components are optimized and the yield coefficients and purification factors are determined. The quantification of 14C entails a three-fold distillation in CO2 followed by the measurement of the 14C counting rate in the purified solution by liquid scintillation spectrometry. 129I was determined by dissolving a sample in HNO3, five-fold extraction, and the subsequent measurement of the 129I counting rate in the purified solution using liquid scintillation spectrometry. The determination of 99Тс included dissolving a sample in the presence of ClO–, double extraction-chromatographic separation of 99Тс with an adsorbent impregnated with methyltrioctylammonium nitrate, followed by the determination of its activity by inductively coupled plasma–mass spectrometry. We have applied this approach to a simulator of vitrified high-level waste, labeled with radionuclides, and achieved satisfactory results. This approach will be instrumental in the analysis of the accumulated vitrified high-level wastes at the “Mayak” Production Association (Mayak PA). The results will be important for modeling the engineering safety barriers of a deep disposal site for radioactive wastes.

This article presents the results of studies of 15 low-melting borosilicate glasses of different compositions using the simplex-based method of mathematical planning. For each glass, the melting temperature and uniformity, as well as the rate and degree of leaching are determined in accordance with GOST (State Standard) R 52126–2003 [1] and NP-019-2015 [2]. Mathematical models are built based on the data obtained. As a result of the research, the most promising area for further research and development of the composition of borosilicate glass for a removable small-sized melter designed by Mayak Production Association is identified.

A method is proposed for quantifying the specific activities of radioecologically significant radionuclides, including 14 C, 99 Tc, and 129 I, in vitrified high-level wastes. The conditions for the separation and purification of these radionuclides from potential interfering components are optimized and the yield coefficients and purification factors are determined. The quantification of 14 C entails a three-fold distillation in CO 2 followed by the measurement of the 14 C counting rate in the purified solution by liquid scintillation spectrometry. 129 I was determined by dissolving a sample in HNO 3 , five-fold extraction, and the subsequent measurement of the 129 I counting rate in the purified solution using liquid scintillation spectrometry. The determination of 99 Тс included dissolving a sample in the presence of ClO – , double extraction-chromatographic separation of 99 Тс with an adsorbent impregnated with methyltrioctylammonium nitrate, followed by the determination of its activity by inductively coupled plasma–mass spectrometry. We have applied this approach to a simulator of vitrified high-level waste, labeled with radionuclides, and achieved satisfactory results. This approach will be instrumental in the analysis of the accumulated vitrified high-level wastes at the “Mayak” Production Association (Mayak PA). The results will be important for modeling the engineering safety barriers of a deep disposal site for radioactive wastes.

The impurities contained in lead and lead-based alloys, which are widely used in various branches of industry, i.e., nuclear, medical, electrical engineering, etc., affect their physicochemical properties which necessitates developing of the reliable method for the impurity determination. Photometric, spectral, and chemical — spectral methods used to address this problem are labor-intensive and do not always have the required sensitivity. A method of inductively coupled plasma mass spectrometry (ICP-MS) coupled with High Matrix Introduction (HMI) technology has been proposed as alternative easy to use procedure designed to be more sensitive. The Agilent HMI Sample Injection System provides inline dilution of the sample aerosol (supplied from the spray chamber to the burner) with pure argon. This method of sample introduction provides for analysis of the solutions with a solute content of up to 1% and higher. The aerosol dilution reduces concentration of the matrix and solvent at the inductively coupled plasma interface without conventional dilution. In this case, the matrix suppression of impurities is almost eliminated and CeO+/Ce+ is reduced to 0.2%, while the typical CeO+/Ce+ ratio for the Agilent 7500 mass spectrometers is 1 – 2%, but no more than 3%. We present application of this method to the analysis of Mg, Ca, Fe, Cu, As, Ag, Sn, Sb, Bi in lead by an Agilent 7500cx ICP-MS with preliminary acid digestion of lead samples in a microwave autoclave. The use of the HMI system made it possible to exclude the stage of sample dilution, reducing the possibility of sample contamination with a diluent, and to determine the content of impurities in a highly concentrated matrix at a level of 10–4 – 10–5 %. The efficiency of the method, as well as the possibility of using multi-element standard solutions prepared with 1% nitric acid for analysis of the samples with high lead content is shown.

— We have prepared multicomponent phosphate and borophosphate glass systems containing considerable concentrations of high-level waste components critical to the vitrification process, such as iron, chromium, nickel, and sulfur, which imitate the composition of the high-level waste in storage tanks. We have studied the glass preparation temperature, viscosity, phase formation, thermal stability, and hydrolytic stability of the materials. The lowest rate of cesium leaching from quenched boron-free aluminophosphate glasses was on the order of 10 –5 g/(cm 2 day).