The article describes the State of Art in HLW vitrifi ation at the PA Mayak’s radiochemical production facility. It presents some conceptual and engineering designs that may further advance the vitrifi ation technology at the enterprise considering various types of liquid HLW. The article discusses the key differences between the promising evacuable small-sized melter and the units previously operated at the enterprise. It overviews its evolvement stages and the results of tests run at PA Mayak in 2018—2020 on the prototypes of evacuable small-sized direct Joule-heated melters designed for borosilicate glass melting. The paper overviews a wide range of the implemented studies and the melters’ operational experience. It focuses more specifi ally on the third improved prototype of an evacuable small-sized melter with its designs developed by PA Mayak in 2022 based on the previously tested prototypes.

Relevance. The need to obtain experimental data on dependence of thermoviscosimetric properties of the examined borosilicate glasses on temperature, in order to solve the problem of developing a low-melting composition of borosilicate glass for a small-scale melter designed by Mayak Production Association. Aim. To get the most complete characterization of the glasses under study using thermoviscometric measurements. Methods. Mathematical modelling, simplex-lattice design. Results and conclusions. One of the most effective methods of experiment planning, which allows, with a relatively small number of experiments, obtaining a mathematical model reflecting the dependence of the studied property of the mixture on the content of components in it, is the simplex experiment planning method. To simplify the solution of such a task, a group or groups of components are usually distinguished, in which the ratio of the mass fraction of the components does not change. In this article, three groups of components are taken – simulators of high-level radioactive waste, sodium and boron oxides, and glass frit. The authors have identified the most promising area, where the mass fraction of glass frit is from 67.5 to 75%, the mass fraction of sodium and boron is from 25 to 32.5% and the mass fraction of high-level waste components is from 0 to 20%. The paper presents the results of thermoviscometric measurements of 15 melts of low-melting borosilicate glasses of different compositions at temperatures ranging from 700 to 1,200°C. Using simplex-based mathematical models of experiment planning, domains of dependence of thermoviscosimetric characteristics of glass melts on their compositions were created. The authors identified the compositions of glass melts that are found within the viscosity limits determined as allowable for glass melt discharge during operation of the removable small-scale melter of Mayak Production Association design.

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.

The present work describes the results of the removal of cesium by sorbents of various classes from highly mineralized alkaline solutions simulating the clarified phase of storage tanks with high-level radioactive waste (HLW) of the Mayak Production Association. Within the scope of the performed works, inorganic sorbents of the Clevasol® and Fersal brands, as well as resorcinol-formaldehyde ion-exchange resins (RFRs: RFR-i, RFR-Ca, and Axionit RCs), were used. The sorbents’ characteristics under both static and dynamic conditions are presented. The Fersal sorbent has demonstrated the best sorption characteristics in the series of sorbents under study. The disadvantage of inorganic sorbents is the loss of mechanical strength upon cesium desorption, which complicates their repeated use. It has been demonstrated that RFRs, despite their lower selectivity towards cesium and adsorption capacity, can be used many times in repeated sorption-desorption cycles. The latter makes RFRs more technologically attractive in terms of the total volume of decontaminated HLW. However, RFRs tend to be oxidized during storage, which results in the formation of carboxyl groups and a decrease in sorption characteristics—this must be further taken into account in the real processes of liquid radioactive waste (LRW) management.

The article overviews the research implemented at PA Mayak from 1986 to 2021 to develop a method enabling the processing of heterogeneous alkaline HLW. It presents relevant flow charts developed under this research, the reasoning behind them and their evolution stages. The study reflects the broad spectrum of research performed and the scaling-up experience associated with the implemented experiments. The study demonstrates the need for further optimization of the flow chart adopted in the designs of a new vitrification complex to reduce the volume of generated solid Class 1 radioactive waste forms and to increase the processing capacity.

Cement mortars and compounds prepared in the course of the development of a process for solidification of the solution fraction of alkaline high-level waste (HLW) accumulated at the Mayak Production Association were studied. Only materials commercially available in Russia were included in dry mix formulations. The spread, bleeding, final setting, and heat release were determined for model inactive cement mortars, and the compression strength, frost resistance, and water resistance, for the compounds. The 137Cs leach rate was determined for the cement compounds with real solidified high-level solutions. The formulation meeting the whole set of regulatory and technological requirements to cemented radioactive waste, provided that the compound is placed in limited-volume containers (e.g., in NZK-type containers), was developed.

Zirconium is one of critical high-level waste components that get into the glass and affect its properties. There are various zirconium sources such as partial corrosion damage of fuel rod clads in the course of SNF reprocessing or dissolution of the zirconium-containing refractory material of the glass-making furnace in the glass melt. Furthermore, zirconium is present in SNF as a fission product. Dissolution of zirconium in aluminoborosilicate glasses developed as a host material for high-level waste vitrification was studied by electron microscopy. The zirconium distribution between the glass and crystalline phases was studied. The optimum component ratio ensuring maximal incorporation of zirconium into the glass-containing fraction of the host material was determined. The results obtained can serve as a basis for adjusting the synthesis parameters of glass-containing borosilicate host materials used for immobilization of zirconium-containing high-level waste.

Выбрана область оптимальных матричных составов на основе доменного шлака с добавками бентонита и метакаолина для отверждения модельного щелочного раствора высокоактивных отходов. Исследования проведены с использованием математического метода планирования эксперимента на симплексе. Определены основные показатели шлакощелочных компаундов, зафиксированные в нормативных требованиях к цементированным радиоактивным отходам. Получены данные о механической прочности исходных образцов цементных компаундов и образцов после циклов замораживания— оттаивания, а также о скорости выщелачивания 137Cs из цементного компаунда. Исследована механическая прочность цементного компаунда после длительного пребывания в воде. The area of optimal matrix compositions based on blast furnace slag with bentonite and metakaolin additives for solidification of model alkaline solution of high-activity waste has been selected. The investigations were carried out using the mathematical method of simplex experi-ment planning. The main parameters of slag-alkali compounds fixed in the regulatory requirements for cemented radioactive waste have been determined. The data on mechanical strength of initial samples of cement compounds and samples after freezing—thawing cycles are obtained. The data on the rate of 137Cs leaching from the cement compound were obtained. The mechanical strength of cement compound after a prolonged stay in water was investigated.