The reprocessing of spent nuclear fuel (SNF) of nuclear power plants leads to the generation of radioactive waste (RW), including high level waste (HLW). HLW contains hazardous long-lived fission products and actinides, therefore, the creation of effective methods of handling it is the main task of the final stage of the closed nuclear fuel cycle. At present, HLW is immobilised in Na–B–Si or Na–Al–P glasses, which does not meet the requirements of safety and economic efficiency. In particular, the low content of waste (3–20 wt %) in the glasses worsens the use of the deep storage space for HLW. Over time, the glasses crystallize with increasing solubility of the matrix in groundwater. The solution to the problem of handling HLW is related to the separation of radionuclides onto fractions. Many of these techniques have already been tested on real liquid waste from SNF reprocessing. As a result, fractions of elements can be obtained for immobilisation in optimal matrices with crystalline (mineral-like) phases. Compounds with structures of pyrochlore, zirconolite, monazite, brannerite, perovskite, britholite, murataite, garnet are suitable for incorporating actinides. For Cs and Sr, phases with the structures of hollandite, pollucite, perovskite, and langbeinite are proposed. Glass crystalline materials (glassceramics) are optimal for isolating wastes of the complex composition. Given the methods for the synthesis of HLW matrices and promising technologies for their industrial production are considered.

Electron backscatter diffraction (EBSD) has been used for more than 30 years for analyzing the structure of minerals and artificial substances. In recent times, EBSD has been widely applied for investigation of irradiated nuclear fuel and matrices for the immobilization of radioactive waste. The combination of EBSD and scanning electron microscopy (SEM/EDS) methods allows researchers to obtain simultaneously data on a specimen’s local composition and structure. The article discusses the abilities of SEM/EDS and EBSD techniques to identify zirconolite polytype modifications and members of the polysomatic murataite–pyrochlore series in polyphase ceramic matrices, with simulations of Pu (Th) and the REE-actinide fraction (Nd) of high-level radioactive waste.

The structural and chemical properties of zirconolite (ideally CaZrTi₂O₇) as a host phase for separated REE–actinide-rich wastes are considered. Detailed analysis of both natural and synthetic zirconolite-structured phases confirms that a selection of zirconolite polytype structures may be obtained, determined by the provenance, crystal chemistry, and/or synthesis route. The production of zirconolite ceramic and glass–ceramic composites at an industrial scale appears most feasible by cold pressing and sintering (CPS), pressure-assisted sintering techniques such as hot isostatic pressing (HIP), or a melt crystallization route. Moreover, we discuss the synthesis of zirconolite glass ceramics by the crystallization of B–Si–Ca–Zr–Ti glasses containing actinides in conditions of increased temperatures relevant to deep borehole disposal (DBD).

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.

The thermal fields for a borehole repository of the Cs–Sr and REE–actinide fractions of radioactive waste are calculated based on the waste content, the size of the matrix block, and the predisposal storage time. It is concluded that such repositories are promising for the REE–actinide fraction included into crystalline matrices with a block diameter of 0.1–0.2 m.

Structural properties and water dissolution of six sodium–aluminum–phosphate (NAP) glasses have been investigated before and after irradiation by a gamma-ray source based on ⁶⁰Co. Two of these samples were of simple composition, and four samples had a complex composition with radionuclide simulants representing actinides, fission, and activated corrosion products. Samples of the simple composition are fully vitreous, whereas samples of the complex composition contained up to 10 vol.% of aluminum–phosphate, AlPO₄, and traces of ruthenium dioxide, RuO₂. Based on the study of pristine and irradiated glasses, it was established that the radiation dose of 62 million Gray had practically no effect on the phase composition and structure of samples. At the same time, the rate of leaching of elements from the irradiated samples by water was decreased by about two times.