Bentonite clay is the most promising material with great waterproof properties, which is considered to hold a great potential as an engineering sealing barrier in deep geological repository use. However, the other properties of the said material after contact with groundwater may be of a huge concern in terms of longterm disposal of radioactive waste in a number of circumstances. This article analyses possible results of such interactions and their consequences. Samples of borosilicate, aluminophosphate, and iron phosphate glass, containing rare earth elements (REE) of the composition Ce0,12La0,19Nd0,31O as simulators of radioactive waste, were synthesized. Solutions of model underground water, imitating the water of the Nizhnekansky granitoid massif that passed through clay engineering barriers, were investigated. The analysis of the structure of glasses after contact with model bentonite, zeolite, and kaolin contact solutions at 120 ?, simulating the heat release of alpha-emitters immobilized in glass, was carried out. It was shown that glasses, upon contact with model solutions, were subject to the formation of a corrosion layer as well as depletion of the glass phase in structure-forming elements. Significant structural rearrangement after leaching in all samples was determined by IR spectrometry, while the smallest changes were typical for samples leached in bidistilled water and kaolin contact solution. This information could be crucial for modeling the long-term behavior of nuclear glass in various scenarios of geological repository degradation or accidents.

Ceramic samples with the following composition (wt %): 50 TiO2, 10 CaO, 10 MnO2, 5 Al2O3, 5 Fe2O3, 10 ZrO2, 10 Ln2O3 (Ln = La, Ce, Nd, Ho) or 10 СеО2, were studied by X-ray photoelectron spectroscopy. According to the data of X-ray phase analysis and scanning electron microscopy, they consist of murataite, zirconolite, and perovskite. In smaller quantities there are crichtonite, pyrophanite-ilmenite, and rutile. Ce3+ dominates in cerium samples: the Ce3+ : Ce4+ ratio is 3 : 1 and does not depend on the method of adding the element to the charge—in the form of СeО2 or Се2О3. All ceramics are dominated by Fe3+, its fraction is 92–94 rel %, while manganese is represented only by Mn3+ cations.

A new type of high-level waste (HLW) is generated during pyrochemical reprocessing of mixed nitride spent uranium–plutonium nuclear fuel. Such waste is a spent electrolyte, which is a mixture of chloride salts containing approximately 25.7 wt.% LiCl + 31.6 wt.% KCl + 4.1 wt.% CsCl + 5.1 wt.% BaCl2 + 3.8 wt.% SrCl2 + 29.7 wt.% LaCl3, and its immobilization in reliable matrices is an actual radiochemical problem. The structure and hydrolytic stability of sodium aluminoironphosphate (NAFP) glass and a low-temperature mineral-like magnesium potassium phosphate (MPP) matrix, which are promising for spent electrolyte immobilization in the presence of hydrogen peroxide solutions simulating natural water radiolysis products, were studied in this work. The structure of the samples was studied using the SEM-EDS method. It was shown that the initial samples of NAFP glass after leaching in hydrogen peroxide solutions are prone to precipitation of crystalline phases on the surface, which are mainly represented by a mixture of sodium–iron–aluminum pyrophosphates. It was established that the leaching rate of structure-forming components of NAFP and MPP matrices generally increase, but remain at a low level, meeting modern requirements for HLW immobilization. This confirms the effectiveness of the studied matrices for the industrial use of the spent electrolyte.

Glass samples containing 3 wt % of KTcO4 have been synthesized in the system, mol %: 40Na2O–10Al2O3–10Fe2O–40P2O5. The glass samples have been studied by X-ray photoelectron spectroscopy. It was found that the degree of technetium oxidation is Tc(IV), which differs from the behavior of rhenium under similar conditions: about 83% of Fe is present as Fe(III), and the rest, as Fe(II). A high stability of the lass against leaching at 90°C was demonstrated in the half-dynamic test (State Standard R 52126–2003): the rate of Tc leaching from the glass is about 3 × 10–6 g/(cm2 day).

Leaching of Np and Pu in water from samples composed of two titanate phases Nd2Ti3O9 (75 wt %) and Nd2Ti2O7 (25 wt %) is investigated. The rates of leaching actinides decrease with time at T = 90°C and are equal to 10–8 g/(cm2 day) for Pu and 7 × 10–7 g/(cm2 day) for Np on the 28th day of the experiment. The higher value of neptunium is likely related to its existence in the samples in valence states IV and V, while plutonium is in states III and IV. It is shown that the acidity of the solution has a significant effect on the leaching rate. Under near-neutral conditions typical of deep repositories, matrices for actinide immobilization based on REE titanates can be considered as corrosion resistant. The causes of the formation of a perovskite-like phase with Nd2Ti3O9 composition are discussed. They can be related to both the high temperature and the reducing conditions of the synthesis process.

The effect of γ-irradiation with a dose of 62 MGy on the stability in water of two types of sodium aluminophosphate glasses was studied: glasses of a simple composition (GS) containing only frame elements (O, Na, Al, Fe, P) and a modified one (GM) with the addition of simulators of radioactive waste components (Cr, Mn, Ni, Ru, La, U). Leaching was carried out at 25 and 90°C with water replacement on the 1st, 3rd, 10th, 14th, 21st, and 28th days from the experiment onset. The dissolution rate of glasses, normalized by the matrix elements (Na, Al, Fe, P), after irradiation remains unchanged or slightly decreases. The leaching rate of radionuclide simulators in the vast majority of cases decreases, but for Cr, Ni, and La in experiments at 90°C, it slightly rises. The temperature curve of glass leaching intensity is described by the Arrhenius formula. An increase in temperature from 25 to 90°C bring about an increase in the leaching rate of all elements from glass matrices both before and after irradiation by 0.5–1.5 orders of magnitude. For both experimental temperatures, the leaching rate of elements decreases with time, which is explained by the protective layer formation on the surface of the samples.

The search for matrices and technological solutions for the reliable immobilization of volatile radionuclides and high-level waste (HLW) components is an actual radiochemical problem. Methods of obtaining of sodium alumino-iron phosphate (NAFP) and iron phosphate (FP) glass composite materials synthesized at temperatures of 450–750 °C, their structure and hydrolytic stability were investigated in this paper. The structure of the samples was studied by XRD and SEM-EDS. It was shown that, in the case of FP materials, the phase composition varies depending on the synthesis temperature, while NAFP materials have a complex multiphase composition at all crystallization temperatures. It has been established that the samples of the obtained glass composite materials have a high hydrolytic stability. At the same time, FP material obtained at 650 °C are the most stable, which makes this medium-temperature method of synthesis promising for the immobilization of volatile HLW components.

Samples of sodium aluminum iron phosphate glass were synthesized containing rhenium as a surrogate of technetium from radioactive waste. The phase composition, structure and water resistance of the obtained glasses were investigated. It was shown that the samples with the rhenium oxide inclusion up to 2.88 wt % are X-ray amorphous and homogeneous, and their anionic motif corresponds to the glassy one. It was found that the oxidation state of rhenium in the obtained glasses is Re(VII), and 97% of iron is in the form of Fe(III) and 3% in the form of Fe(II). The glass was established to be highly resistant to leaching at 90°C. The rate of Re leaching from glasses in accordance with the international product consistency test (PCT) and the semi-dynamic test of State Standard GOST R 52126-2003 is no more than 6  ×  10–6 and 3  ×  10–6 g/(cm2 day), respectively.

Immobilization of spent electrolyte–radioactive waste (RW) generated during the pyrochemical processing of mixed nitride uranium–plutonium spent nuclear fuel is an acute task for further development of the closed nuclear fuel cycle with fast neutron reactors. The electrolyte is a mixture of chloride salts that cannot be immobilized directly in conventional cement or glass matrix. In this work, a low-temperature magnesium potassium phosphate (MPP) matrix and two types of high-temperature matrices (sodium aluminoironphosphate (NAFP) glass and ceramics based on bentonite clay) were synthesized. Two systems (Li0.4K0.28La0.08Cs0.016Sr0.016Ba0.016Cl and Li0.56K0.40Cs0.02Sr0.02Cl) were used as spent electrolyte imitators. The phase composition and structure of obtained materials were studied by XRD and SEM-EDS methods. The differential leaching rate of Cs from MPP compound and ceramic based on bentonite clay was about 10−5 g/(cm2·day), and the rate of Na from NAFP glass was about 10−6 g/(cm2·day). The rate of 239Pu from MPP compound (leaching at 25 °C) and NAFP glass (leaching at 90 °C) was about 10−6 and 10−7 g/(cm2·day), respectively. All the synthesized materials demonstrated high hydrolytic, mechanical compression strength (40–50 MPa) even after thermal (up to 450 °C) and irradiation (up to 109 Gy) tests. The characteristics of the studied matrices correspond to the current requirements to immobilized high-level RW, that allow us to suggest these materials for industrial processing of the spent electrolyte.

The influence exerted by the composition of ceramics with imitators of radioactive wastes (Nd, U) on their stability against leaching in water at 90 and 150°C has been studied. Raising the content of TiO2 in a ceramic with Nd led to an increase in the fraction of perovskite and decrease in the amount of murataite. Samples with perovskite have the lowest stability in solution, especially at 150°C. Making smaller the fraction of perovskite lowers the rate of Nd leaching by a factor of 2 (90°C) and 5 (150°C). The weaker influence exerted by temperature on the U leaching from the samples is observed because more stable pyrochlore or zirconolite appear in the samples instead of perovskite.

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.

The effective isolation of radioactive waste (RW) from the environment is the main problem for the further development of nuclear power.The main phases in titanate-based ceramics are perovskite, rutile, zirconolite and murataite.Murataite grains have a zonal structure with high content of rare earth elements at the center of structure and low content of these at edges, that precludes their leaching in contact with a solution.Murataite-based ceramics containing simulated rare earth elements of high level waste (HLW) were produced via melting of oxide mixtures in a resistance furnace at 1500°C.All samples were composed of mainly murataite and minor perovskite, crichtonite, zirconolite, and pyrophanite/ilmenite phases.Thus, murataite is the dominant host phase for a sample containing zirconium oxide.All samples were analyzed by scanning electron microscopy with an energy dispersive X-ray spectroscopy.Elemental leaching rates from the ceramic with low perovskite content were lower by one order of magnitude then leaching rates for high perovskite content.

The iron-free and iron-bearing (Np, Pu)-containing sodium-aluminum-phosphate glass samples were studied by X-Ray photoelectron spectroscopy. Oxidation states of neptunium, plutonium and iron, and elemental composition of the surfaces of the samples studied were determined. In both iron-free and iron-bearing glasses Np is present predominantly as Np(V) on the surface and Np(IV) in the bulk. Pu(III) dominates over Pu(IV)/(V) in the iron-free glass while oxidized forms of Pu prevail in the iron-bearing glass.

The phase composition, structure, elemental distribution, and hydrolytic stability of quench-hardened and tempered sodium-aluminum-phosphate (SAP) and sodium-aluminum(iron) phosphate (SAIP) glass materials containing 10 wt.% (above 100%) oxides of rare earth elements (REE) were studied. It is shown that all materials obtained by quench-hardening melts (except La-containing melts) were x-ray amorphous, while slowly cooled (tempered) melts partially or completely crystallized with different phases of orthophosphates being released. The rate of leaching of elements from tempered SAP glasses determined at 90°C is approximately one or two orders of magnitude higher than then for the quench-hardened glasses, and the leach rate of elements from quench-hardened SAIP glasses is lower than from quench-hardened SAP glasses.

Samples of aluminum (iron) phosphate glasses containing weighable amounts of rare earth and transuranium elements were synthesized. The quenched glasses obtained are X-ray amorphous, and in the course of annealing they undergo partial crystallization with segregation of a poorly soluble phase of monazite structure. However, the hydrolytic durability of the glasses remains on the level meeting the requirements to immobilized high-level waste (no more than 10 –7 g cm –2 day –1 ). Plutonium in the surface layer of the glass samples occurs in oxidation state IV, and neptunium, mainly in oxidation states IV and V.