Search Results (59)

Complex oxides with the general formula Gd₂(Ti_1−xZr_x)₂O₇ are promising candidates for radioactive waste immobilization due to their capacity to withstand radiation by dissipating part of the free energy driving defect creation and phase transitions. In this study, samples with varying zirconium content (xZr = 0.00, 0.15, 0.25, 0.375, 0.56, 0.75, 0.85, 1.00) were synthesized via the sol–gel method and thermally treated at 500 °C to obtain nanosized powders mimicking the defective structure of irradiated materials. Synchrotron-based techniques were employed to investigate their structural properties: High-Resolution X-ray Powder Diffraction (HR-XRPD) was used to assess long-range structure, while Pair Distribution Function (PDF) analysis and Extended X-ray Absorption Fine Structure (EXAFS) spectroscopy provided insights into the local structure. HR-XRPD data revealed that samples with low Zr content (xZr ≤ 0.25) are amorphous. Increasing Zr concentration led to the emergence of a crystalline phase identified as defective fluorite (xZr = 0.375, 0.56). Samples with the highest Zr content (xZr ≥ 0.75) were fully crystalline and exhibited only the fluorite phase. The experimental G(r) functions of the fully crystalline samples in the low r range are suitably fitted by the Weberite structure, mapping the relaxations induced by structural disorder in defective fluorite. These structural insights informed the subsequent EXAFS analysis at the Zr-K and Gd-L₃ edges, confirming the splitting of the cation–cation distances associated with different metal species. Moreover, EXAFS provided a local structural description of the amorphous phases, identifying a consistent Gd-O distance across all compositions. Full article

This study investigates the mechanical and microstructural performance of three alkali-activated geopolymer formulations, constituted of metakaolin (MK), blast furnace slag (BFS), and a ternary blend of MK, BFS, and fly ash (MIX), for the immobilization of simulated radioactive liquid organic waste (RLOW). Thermal ageing tests were performed to evaluate geopolymer durability, including fire exposure (800 °C) and climatic chamber cycles (from −20 to 40 °C). Characterization through thermogravimetric analysis (TGA), compression tests, and scanning electron microscopy with energy-dispersive spectroscopy (SEM-EDS) was carried out to assess material degradation after thermal ageing. Preliminary results showed substantial strength and microstructural degradation in oil-loaded specimens after cyclic climatic ageing, while fire-exposed blank matrices retained partial mechanical integrity. BFS matrices exhibited the best thermal resistance, attributable to the formation of Ca-Al-Si-hydrate (C-A-S-H) gels. These findings support the use of optimized geopolymer formulations for safe RLOW immobilization, while contributing to the advancement of knowledge on sustainable and regulatory-compliant direct conditioning technology. Full article

In the context of the disposal of spent radioactive fuel, heat-emitting radionuclides such as Cs and Sr are of utmost concern, as they have a major influence on the distance at which disposal galleries should be spaced apart and, thus, the cost of a disposal facility. Therefore, certain scenarios investigate the partitioning and transmutation of spent fuel to optimize the disposability of both Cs- and Sr-rich waste streams and the remaining fractions. In this study, the Cs- and Sr-rich waste stream, a nitrate-based solution, was immobilized in metakaolin and blast furnace slag-based alkali-activated matrices. These matrices were chosen for immobilization because they are known to offer advantages in terms of durability and/or heat resistance compared with traditional cementitious materials. The goal of this study is to develop an optimal recipe for the retention of Cs and Sr. For this purpose, recipes were developed following a design-of-experiments approach by varying the water-to-binder ratio, precursor, and waste loading while respecting matrix constraints. Leaching tests in deionized water showed that the metakaolin-based matrix was superior for the combined retention of both Cs and Sr. The optimal recipe was further tested under accelerated leaching conditions in an ammonium nitrate solution, which revealed that the leaching of Cs and Sr remained within reasonable limits. These results confirm that alkali-activated materials can be effectively used for the immobilization and long-term retention of heat-emitting radionuclides. Full article

Several recent studies have evaluated technologies of spent nuclear fuel processing specifically for solidifying transuranic (TRU) waste as a by-product of fission. Of the TRU group, plutonium and the minor actinides will be responsible for the bulk of the radiotoxicity and heat generation of spent nuclear fuel in the long term (300 to 20,000 years). In this study, we investigated magnesium potassium phosphate (MKP)-based compounds as host waste forms for the encapsulation of inactive trivalent Nd and Sm as analogues of the minor trivalent actinides, Am and Cm. Waste forms were fabricated under ambient atmospheric conditions by adding 5 wt.% of substances containing Nd or Sm via the following two routes: powder oxides and aqueous solutions of nitrate salts. Waste form performance was established using strength and aqueous medium leaching tests of MKP-based specimens. The MKP materials were analyzed by X-ray diffraction (XRD), scanning and transmission electron microscopy (SEM and TEM), energy-dispersive X-ray spectroscopy (EDS), and Raman spectroscopy. The waste forms exhibited a compressive strength of ≥30 MPa and were durable in an aqueous environment. The leachability indices for Nd and Sm, as per the ANS 16.1 procedure, were 19.55–19.78 and 19.74–19.89, respectively, which satisfy the acceptable criteria (>6). The results of the present room temperature leaching study suggest that MKPs can be effectively used as a host material to immobilize actinides (Am and Cm) contained in TRU waste. Full article

Among the many possible applications of iron phosphate glasses, one of them is that they are promising materials in waste vitrification, particularly for radioactive waste. In vitrified form, waste elements should be permanently immobilized in a glass network as they are susceptible to harsh environmental conditions. The self-diffusion of the vitrified material species may limit the potential usefulness of the glasses. This paper presents the possibility of using molecular dynamics simulations to study this process and the substitution of SrO into an iron phosphate glass network. It was evidenced that the self-diffusion mechanism differed significantly depending on whether the glass was in a solid or liquid state. The proposed method also offered a relatively easy prediction of glass characteristic temperatures, such as transformation and flow. We also observed, and here describe, an aggregation process of the glass elements that may drive their crystallization. The obtained results are discussed in light of the experimental and theoretical structural feature literature data. Full article

Rare earth (RE) aluminosilicate glasses exhibit several favorable chemical, mechanical and thermal properties. As such, they are considered to be model systems for long-half-life actinides and are candidate containment materials for long-term immobilization of radioactive wastes. The aim of the present study was to investigate the effect of the substitution of sodium oxide on the glass transition temperature and structure of lanthanum aluminosilicate glasses. The primary objective was to elucidate the relationship between the substitution of Na₂O for La₂O₃ on the Tg reduction and structural characteristics of lanthanum aluminosilicate glass, including identifying changes in the main Qn species and local environments of Si and Al. The structure of SiO₂–Al₂O₃–La₂O₃–Na₂O glasses has not been studied previously, and, thus, this investigation is the first to assess the structural changes occurring when La₂O₃ is substituted by Na₂O. Three glasses were prepared with general composition (mol.%): 55SiO₂–25Al₂O₃–20M₂O_n (M = La or Na; n = 3 or 1). Glass G1 contains 20 mol.% La₂O₃; in G2, 15 mol.% of La₂O₃ was substituted by 15 mol.% Na₂O; and Glass G3 contains 20 mol.% Na₂O. The glasses were characterized by DSC to determine glass transition temperatures. As expected, as Na is substituted for La, T_g decreases substantially. Structural studies were carried out by FTIR spectroscopy, ²⁹Si, and ²⁷Al MAS NMR. As Na is substituted for La in these aluminosilicate glasses, the main goals that were achieved were the identification of Qⁿ species and also changes in the local environments of Si and Al: {Qⁿ_Si(mAl)} and {Qⁿ_Al(mSi)}. Full article

In this paper, we study the influence of densified microsilica and colloidal nanosilica admixtures on the mechanical strength and the microstructural characteristics of special mortars used for immobilizing radioactive concrete waste. The experimental program focused on the replacement of cement with micro- and/or nanosilica, in different proportions, in the basic composition of a mortar made with recycled aggregates. The technical criteria imposed for such cementitious systems, used for the encapsulation of low-level radioactive waste, imply high fluidity, increased mechanical strength and lack of segregation and of bleeding. We aimed to increase the structural compactness of the mortars by adding micro- and nanosilica, all the while maintaining the technical criteria imposed, to obtain a cement matrix with high durability and increased capacity for immobilizing radionuclides. The samples from all the compositions obtained were analyzed from the point of view of mechanical strength. Also, micro- and nanosilica as well as samples of the optimal mortar compositions were analyzed physically and microstructurally. Experimental data showed that the mortar samples present maximum compressive strength for a content between 6 and 7.5% wt. of microsilica, respectively, for a content of 2.25% wt. nanosilica. The obtained results suggest a synergistic effect of micro- and nanosilica when they are used simultaneously in cementitious compositions. Thus, among the analyzed compositional variants, the mortar composition with 3% wt. microsilica and 2.25% wt. nanosilica showed the best performance, with an increase in compressive strength of 23.5% compared to the control sample (without micro- and nanosilica). Brunauer–Emmett–Teller (BET) analysis and scanning electron microscopy (SEM) images highlighted the decrease in pore diameter and the increase in structural compactness, especially for mortar samples with nanosilica content or a mixture of micro- and nanosilica. This study is useful in the field of recycling radioactive concrete resulting from the decommissioning of nuclear research or nuclear power reactors. Full article

Magnesium Potassium Phosphate Cements (MKPCs) are considered a good alternative for the immobilization of aluminium radioactive waste. MKPC composition and moisture curing conditions are relevant issues to be evaluated. The corrosion of pure aluminium (A1050) and AlMg alloys (AA5754) with 3.5% of Mg is studied in MKPC systems prepared with different MgO/KH₂PO₄ (M/P) molar ratios (1, 2, and 3M) and moisture curing conditions (100% Relative Humidity (RH) and isolated in plastic containers (endogenous curing)). The Al corrosion potential (E_corr) and corrosion kinetic (i_corr and V_corr) are evaluated over 90 days. Additionally, the pore ion evolution, the matrix electrical resistance, the pore structure, and compressive strength are analysed. The corrosion process of Al alloy is affected by the pH and ion content in the pore solution. The pore pH increases from near neutral for the 1M M/P ratio to 9 and 10 for the 2 and 3M M/P ratio, increasing in the same way the corrosion of pure Al (AA1050) and AlMg alloys (AA5754). The effect of Mg content in the alloy (AA5754) becomes more relevant with the increase in the M/P ratio. The presence of phosphate ions in the pore solution inhibits the corrosion process in both Al alloys. The MKPC physicochemical stability improved with the increase in the M/P ratio, higher mechanical strength, and more refined pore structure. Full article

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. Full article

Glass-containing materials are widely considered among the most reliable materials for the immobilization of radioactive waste materials. The present work considers the synthesis of glass–ceramic and glass crystalline composite materials based on borosilicate glasses. The synthesis of glass–ceramic materials was carried out by a gradual temperature decrease, followed by crystallization for several hours. Sintering of crushed samples with crystalline components was carried out as an alternative procedure. Porous glasses were produced from glass melts by quenching. After impregnating the resulting porous materials with aqueous solutions of cesium nitrate, compaction of the glass was carried out to form glass crystalline composites. The thermochemical characteristics of the parent glasses were determined using the differential scanning calorimetry method. The phase composition and structure of the glass-containing materials were determined using X-ray phase analysis, X-ray spectral microanalysis, and Raman spectroscopy. Full article

Conditioning of radioactive waste generated from the operation of medical institutions, nuclear cycle facilities, and nuclear facilities is important for the safety of the environment. One of the most hazardous radionuclides is radioactive cesium. There is a need for more effective solutions to contain radionuclides, especially cesium (Cs+). Geopolymers are promising inorganic materials that can provide a large active surface area with adjustable porosity and binding capacity. The existence of nanosized zeolite-like structures in aluminosilicate gels was shown earlier. These structures are candidates for immobilizing radioactive cesium (Cs⁺). However, the mechanisms of their interactions with the aluminosilicate framework related to radionuclide immobilization have not been well studied. In this work, the influence of alkaline cations (Na⁺ or K⁺) and the aluminosilicate framework structure on the binding capacity and mechanism of interaction of geopolymers with Cs⁺ is explored in the example of a sodalite framework. The local structure of the water molecules and alkaline ions in the equilibrium state and its behavior when the Si/Al ratio was changed were studied by DFT. Full article

A novel solid-phase synthetic approach was developed to produce a mineral-like composite ceramic based on strontium titanate (SrTiO₃) and yttrium titanate (Y₂Ti₂O₇) matrices for immobilizing radionuclides such as ⁹⁰Sr and its daughter product ⁹⁰Y, as well as lanthanides and actinides, via reactive spark plasma sintering technology (SPS-RS). Using XRD, SEM, and EDS analyses, the sintering kinetics of the initial mixed oxide reactants of composition Y_xSr_1–1.5xTiO₃ (x = 0.2, 0.4, 0.6 and 1) and structure-phase changes in the ceramics under SPS-RS conditions were investigated as a function of Y³⁺ content. In addition, a detailed study of phase transformation kinetics over time as a function of the heating temperature of the initial components (SrCO₃, TiO₂, and Y₂O₃) was conducted via in situ synchrotron XRD heating experiments. The composite ceramic achieved relatively high physicomechanical properties, including relative density between 4.92–4.64 g/cm³, Vickers microhardness of 500–800 HV, and compressive strength ranging from 95.5–272.4 MPa. An evaluation of hydrolytic stability and leaching rates of Sr²⁺ and Y³⁺ from the matrices was performed, demonstrating rates did not exceed 10⁻⁵–10⁻⁶ g·cm⁻²·day⁻¹ in compliance with GOST R 50926-96 and ANSI/ANS 16.1 standards. The leaching mechanism of these components was studied, including the calculation of solution penetration depth in the ceramic bulk and ion diffusion coefficients in the solution. These findings show great promise for radioactive waste conditioning technologies and the manufacturing of radioisotope products. Full article

The operation of nuclear power plants generates a large amount of low- and intermediate-level radioactive waste liquid. Zeolite-rich geopolymers, which are synthesized under hydrothermal conditions from industrial waste fly ash, can effectively immobilize radioactive nuclides. In this study, the synthesis law of zeolite-rich geopolymers and the adsorption/desorption performances of radioactive nuclide Cs⁺ were researched using XRD, SEM and ICP. The results show that the increase in curing temperatures and NaOH concentrations leads to the transformation of Y-type zeolite to chabazite and cancrinite at low NaNO₃ concentrations. However, at high NaNO₃ concentrations, NaOH above 2 M has no obvious effect on the phase transformation of the main zeolite of chabazite and cancrinite. In the adsorption and desorption experiment of Cs⁺ on the chabazite/garronite-rich geopolymer, it was found that the adsorption of Cs⁺ in the low initial concentration range is more suitable for the Freundlich equation, while the Langmuir equation fits in the adsorption process at the high initial concentration range. Moreover, the desorption kinetics of Cs⁺ are in good agreement with the pseudo-second-order rate equation. Thus, the adsorption of Cs⁺ on chabazite/garronite-rich geopolymers is controlled by both physical and chemical reactions, while desorption is a chemical process. Full article

Among the many radiochemical problems, the search for new materials and technologies for the immobilization of radioactive waste remains relevant, and the range continues to change and expand. The possibility of immobilizing the spent chloride electrolyte after the pyrochemical processing of the mixed uranium-plutonium spent nuclear fuel of the new fast reactor BREST-OD-300 on lead coolant into glass-composite phosphate materials synthesized at temperatures of 650–750 °C was studied. The structure of the obtained samples was studied using XRD and SEM/EDS methods. It has been shown that the spent electrolyte simulator components create stable mixed pyrophosphate phases in the glass composite structure. The materials were found to have high hydrolytic stability. This indicates the promise of using phosphate glass composites as materials for the reliable immobilization of the spent electrolyte. Full article

This paper reports a method for the fabrication of mineral-like SrMoO₄ ceramics with a powellite structure, which is promising for the immobilization of the high-energy ⁹⁰Sr radioisotope. The reported method is based on the solid-phase “in situ” interaction between SrO and MoO₃ oxides initiated under spark plasma sintering (SPS) conditions. Dilatometry, XRD, SEM, and EDX methods were used to investigate the consolidation dynamics, phase formation, and structural changes in the reactive powder blend and sintered ceramics. The temperature conditions for SrMoO₄ formation under SPS were determined, yielding ceramics with a relative density of 84.0–96.3%, Vickers microhardness of 157–295 HV, and compressive strength of 54–331 MPa. Ceramic samples demonstrate a low Sr leaching rate of 10⁻⁶ g/cm²·day, indicating a rather high hydrolytic stability and meeting the requirements of GOST R 50926-96 imposed on solid radioactive wastes. The results presented here show a wide range of prospects for the application of ceramic matrixes with the mineral-like composition studied here to radioactive waste processing and radioisotope manufacturing. Full article