Sergey V. Yudintsev

Materials based on Ca–REE silicate with an apatite–britholite structure are candidate matrices for isolating the REE–actinide fraction of nuclear fuel reprocessing waste. Britholite has a simple composition and high capacity for REEs and actinides; it is stable in hot solutions. Leaching rates of REEs (actinide simulants) by water and brine are 10^–7–10^–9 g/(cm² day). Literature data on the properties of ceramics and glass-ceramics containing britholite are analyzed, including composition, behavior in solutions and under irradiation, information on natural mineral analogs, and methods for the industrial production of britholite matrices for immobilization of the REE–actinide fraction.

Britholite, a Ca-REE silicate sample with an apatite structure, a potential matrix for isolation of the REE-actinide fraction of wastes from spent nuclear fuel processing, is produced by cold crucible induction melting. Britholite has a simple composition, high capacity for actinides, and is stable in heated aqueous solutions. For the first time, the leaching rates of Nd in water and brine at 200–250°C are determined to be 10^–3–10^–5 g/(m² day).

The choice of efficient methods for the immobilization of high-level waste (HLW) resulting from the reprocessing of spent nuclear fuel (SNF) is an important scientific and practical task. The current policy of managing HLW within a closed nuclear fuel cycle envisages its vitrification into borosilicate (B-Si) or alumina–phosphate (Al-P) glasses. These wasteforms have rather limited waste loading and can potentially impair their retaining properties on devitrification. The optimal solution for HLW immobilization could be separating radionuclides into groups using dedicated capacious durable matrices. The phases of the Nd₂O₃–ZrO₂–TiO₂ system in this respect are promising hosts for the REE (rare earth elements: Nd, Ce, La, Pr, Sm, Gd, Y) –MA (MA: Am, Cm) fraction of HLW. In this manuscript, we present data on the composition of the samples analyzed, their durability in hot water, their behavior under irradiation, and their industrial manufacturing methods.

The choice of efficient methods for immobilization of high-level waste (HLW) resulting from reprocessing of spent nuclear fuel (SNF) is an important scientific and practical task. The current policy of managing HLW within closed nuclear fuel cycle envisages its vitrification into borosilicate (B-Si) or alumina-phosphate (Al-P) glasses. These wasteforms have rather limited waste loading, and can potentially impair their retaining properties on devitrification. The optimal solution for HLW immobilisation could be by separating radionuclides into groups using dedicated capacious durable matrices. The phases of the Nd2O3 – ZrO2 – TiO2 system in this respect are promising hosts for REE – MA (MA = Am, Cm) fraction of HLW. We present herewith data on composition of samples analyzed, their durability in hot water, behavior under irradiation, and industrial manufacturing methods.

Separation of high-level waste (HLW) onto fractions simplifies their isolation in matrices and disposal. One of these fractions consists of rare earths (REE) and minor actinides (MA = Am, Cm). Among the rare earth elements there are isotopes 144Ce, 147Pm, 151Sm, and 154,155Eu with half-lives of up to 93 years, decay of the nuclides will cause heating of the matrices with HLW. It has been shown that preliminary storage of the REE–MA matrix for 10 years or more will reduce the content of REE radioisotopes and their contribution to temperature increase in geological repository.

High-level waste (HLW) separation into fractions simplifies isolation in matrices and disposal. One of these fractions consists of rare earths (REE) and minor actinides (MA = Am, Cm). Among the rare earth elements there are the isotopes ¹⁴⁴Ce, ¹⁴⁷Pm, ¹⁵¹Sm, and ^154,155Eu with half-lives of up to 93 years; decay of the nuclides will cause heating of the matrices with HLW. It has been shown that preliminary storage of the REE–MA matrix for ten years or more will significantly reduce the content of REE radioisotopes and their contribution to temperature increase in the geological repository.

The current policy of managing high-level waste (HLW) derived in the closed nuclear fuel cycle consists in their vitrification into B-Si or Al-P vitreous forms. These compounds have rather limited capacity with respect to the HLW (5–20 wt%), and their properties change over time due to devitrification of the glasses. Cardinal improvement in the management of HLW can be achieved by their separation onto groups of elements with similar properties, followed by their immobilization in robust waste forms (matrices) and emplacement in deep disposal facilities. One of the possible fractions contains trivalent rare-earth elements (REEs) and minor actinides (MAs = Am and Cm). REEs are the fission products of actinides, which are mainly represented by stable isotopes of elements from La to Gd as well as Y. This group also contains small amounts of short-lived radionuclides with half-lives (T_1/2) from 284 days (¹⁴⁴Ce) to 90 years (¹⁵¹Sm), including ¹⁴⁷Pm (T_1/2 = 2.6 years), ¹⁵⁴Eu (T_1/2 = 8.8 years), and ¹⁵⁵Eu (T_1/2 = 5 years). However, the main long-term environmental hazard of the REE–MA fraction is associated with Am and Cm, with half-lives from 18 years (²⁴⁴Cm) to 8500 years (²⁴⁵Cm), and their daughter products: ²³⁷Np (T_1/2 = 2.14 × 10⁶ years), ²³⁹Pu (T_1/2 = 2.41 × 10⁴ years), ²⁴⁰Pu (T_1/2 = 6537 years), and ²⁴²Pu (T_1/2 = 3.76 × 10⁵ years), which should be immobilized into a durable waste form that prevents their release into the environment. Due to the heat generated by decaying radionuclides, the temperature of matrices with an REE–MA fraction will be increased by hundreds of centigrade above ambient. This process can be utilized by selecting a vitreous waste form that will crystallize to form durable crystalline phases with long-lived radionuclides. We estimated the thermal effects in a potential REE–MA glass composite material based on the size of the block, the content of waste, the time of storage before immobilization and after disposal, and showed that it is possible to select the waste loading, size of blocks, and storage time so that the temperature of the matrix during the first decades will reach 500–700 °C, which corresponds to the optimal range of glass crystallization. As a result, a glass–ceramic composite will be produced that contains monazite ((REE,MA)PO₄) in phosphate glasses; britholite (Ca_x(REE,MA)_10-x(SiO₄)₆O₂) or zirconolite ((Ca,REE,MA)(Zr,REE,MA)(Ti,Al,Fe)₂O₇), in silicate systems. This possibility is confirmed by experimental data on the crystallization of glasses with REEs and actinides (Pu, Am). The prospect for the disposal of glasses with the REE–MA fraction in deep boreholes is briefly considered.

Highly radioactive materials classified as high-level nuclear waste (HLW) of atomic power engineering should be disposed of deeply underground in special geological disposal facilities (GDFs), which can be of either shaft or borehole type. The advantages of borehole-type GDFs result from smaller volumes of mining operations, a simpler construction technology, shorter construction time and cost. This allows us to consider them as an alternative to shaft-type GDFs. The parts of the boreholes in which waste containers should be placed can be both vertical and horizontal. Computer simulation of the migration of radionuclides from a group of parallel horizontal boreholes into the biosphere made it possible to conclude that horizontal GDF boreholes have significant advantages over vertical ones. We determined a forecast of ²⁴¹Am migration by a method of mathematical modelling of ²⁴¹Am release from vitrified HLW disposed of in several horizontal drillholes. The maximum concentrations of americium in the near-surface groundwater above the repository are calculated depending on the number of boreholes, the depth of their location and the distance between them, the permeability of rocks and the time of waste storage prior to disposal. Influence of different conditions on the safety of a GDF of borehole type is estimated. Calculations show that the heat generated by HLW causes a weaker groundwater convection near horizontal boreholes compared to vertical boreholes of the same capacity. In addition to that, at an equal thickness of the rock layer separating the HLW from the surface, the geothermal temperature of the host rocks in the near field of a horizontal borehole will be lower than the average geothermal temperature near a vertical borehole. As a result, the rate of radionuclides leaching from the waste forms by groundwaters will also be lower in the case of horizontal boreholes.

This article considers tasks needed to ensure sustainable development of the Russian nuclear industry, which are aimed for solving the problems of the initial and final stages of the nuclear fuel cycle, from reproduction of the uranium mineral-resource base to handling of liquid high-level waste of spent nuclear fuel reprocessing. The trends in the development of the mineral-resource complex are analyzed in connection with the projected growth of electricity generation at nuclear power plants. New approaches to the allocation and substantiation of prospecting areas based on the models of mineral-forming systems of uranium deposits using remote-sensing data, GIS modeling, and neural-network technologies are proposed on the example of the territory of southeastern Transbaikal within the framework of current prognostic and mineragenic studies. A detailed analysis of trends in solving the problems of the back end stage of the nuclear fuel cycle is carried out. The use of phosphate glass ceramics with monazite is proposed for isolation and subsequent disposal of the REE–actinide fraction of high-level waste.

The tasks of ensuring the sustainable development of the nuclear energy industry in Russia are considered, which are aimed at solving the problems of the initial and final stages of the nuclear fuel cycle from the reproduction of the mineral resource base of uranium to the management of high-level liquid waste from the reprocessing of spent nuclear fuel. An analysis was made of the trends in the development of the mineral resource base of uranium in connection with the predicted growth in electricity generation at nuclear power plants. Based on the example of the territory of southeastern Transbaikalia, within the framework of topical mineralogical studies, new approaches to the allocation and justification of prospecting areas are proposed based on models of mineral-forming systems of uranium deposits using Earth remote sensing data, GIS modeling and neural network technologies. A detailed analysis of trends in solving the problems of the final stages of the nuclear fuel cycle has been carried out. For isolation and subsequent disposal of the REE-actinide fraction of high-level waste, it is proposed to use phosphate glass ceramics with monazite