Waste Form Materials Research Articles

A finite difference informed random walker (FDiRW) solver for strongly inhomogeneous diffusion problems

In nature, many complex multi-physics coupling problems exhibit strong diffusivity inhomogeneity. For instance, in the context of radionuclide absorption by porous wasteform materials within a flowing waste stream, the difference of species’ diffusivity in solid and liquid phases spans by 3 ∼ 8 orders of magnitude. To solve the diffusion equations with strongly inhomogeneous diffusivity, traditional discretization-based methods, such as the Finite Difference Method (FDM), require infinitesimally small-time steps (<10-10) as high spatial resolutions are employed in most microstructure evolution processes, leading to prohibitively high computational costs. This work developed an integrated numerical approach (FDiRW: Finite Difference informed Random Walk) to tackle this challenge. The idea is that utilizing the Random Walk concept, the fast diffusion is modelled as a superposition of point source’s solution for a concentration distribution while FDM is used to obtain the point source’s solution at each node. A mesh-coarsening algorithm is developed to generate an exclusive coarse mesh for FDiRW approach to maximize its efficiency. The effectiveness of the coarse mesh-based FDiRW approach is validated by benchmarking Finite Difference solutions. Numerical results demonstrated that FDiRW achieves a remarkable 1000x computational efficiency improvement over FDM while preserving desired accuracy for a medium-sized model of 192×192×192 grids. As models scale up, a floating-point operations (PLOPs) analysis of the FDiRW algorithm reveals that its computational complexity grows quadratically in terms of the number of nodes employed in computation.

Radionuclide Solid:liquid partitioning in an aged, reducing-grout wasteform recovered from a disposal facility

The Saltstone Disposal Facility on the Savannah River Site in South Carolina disposes of Low-Level Waste in a reducing-grout waste form. Reducing grout is presently being evaluated as a subsurface disposal waste form at several other locations in the United States, as well as in Europe and Asia. The objective of this study was to collect core samples directly from the Saltstone Disposal Facility and measure desorption distribution coefficients (Kd; radionuclide concentration ratio of saltstone:liquid; (Bq/kg)/Bq/L)) and desorption apparent solubility values (ksp; radionuclide aqueous concentration (moles/L)). An important attribute of this study was that these tests were conducted with actual aged, grout waste form materials, not small-volume simulants prepared in a laboratory. The reducing grout is comprised of blast furnace slag, Class F fly ash, ordinary portland cement, and a radioactive salt waste solution generated during nuclear processing. The grout sample used in this study underwent hydrolyzation in the disposal facility for 30 months prior to measuring radionuclide leaching. Leaching experiments were conducted either in an inert (no oxygen) atmosphere to simulate conditions within the saltstone monolith prior to aging (becoming oxidized) or they were exposed to atmosphere conditions to simulate conditions of an aged saltstone. Importantly, these experiments were designed not to be diffusion limited, that is, the saltstone was ground finely and the suspensions were under constant agitation during the equilibration period. Under oxidized conditions, measured Tc Kd values were 10 mL/g, which was appreciably greater than the historical best-estimate value of 0.8 mL/g. This difference is likely the result of a fraction of the Tc remaining in the less soluble Tc(IV) form, even after extensive oxidation during the experiment. Under oxidized and reducing conditions, the measured Ba and Sr (both divalent alkaline earth metals) Kd value were more than an order of magnitude greater than historical best-estimate values of 100 mL/g. The unexpectedly high Ba and Sr Kd values were attributed to these radionuclides having sufficient time to age (form strong bonds) in the sulfur-rich saltstone sample. Apparent ksp values under reducing conditions were 10−9 mol/L Tc and 10−13 mol/L Pu, consistent with values measured with surrogate materials. Measured apparent Ba, Sr, and Th ksp values were significantly greater than historical best-estimates. The implications of the generally greater Kd values and lower ksp values in these measurements is that these cementitious waste forms have greater radionuclide retention than was previously estimated based on laboratory studies using surrogate materials. This work represents the first leaching study performed with an actual aged, reducing-grout sample and as such provides an important comparison to studies conducted with surrogate materials, and provides high pedigree data for other programs around the world evaluating reducing grouts as a wasteform for subsurface nuclear waste disposal.

A comprehensive XPS investigation of the effect of aqueous corrosion on the surface of glass-zirconolite composite material (Fe–Al–BG–CaZrTi2O7)

Extensive X-ray photoelectron spectroscopy (XPS) analyses of iron–aluminum borosilicate glass (Fe–Al–BG), zirconolite (CaZrTi2O7), and Fe–Al–BG–CaZrTi2O7 composite material have demonstrated that a hydrated layer has been formed on the surface of these materials after exposure to water. The surface chemistry of Fe–Al–BG–CaZrTi2O7 composite material as a potential nuclear wasteform candidate was observed to change in a similar fashion to the borosilicate glass as a current wasteform material after these materials are exposed to water. The XPS measurements suggest that the aqueous corrosion of the Fe–Al–BG–CaZrTi2O7 composite material is a result of ion exchange, hydrolysis, and redox reactions of metal–oxygen bonds with water at the surface of the Fe–Al–BG–CaZrTi2O7 composite material. Moreover, the XPS results proved that the composition and chemical environment of the precipitated layer on the surface of the Fe–Al–BG–CaZrTi2O7 composite material after exposure to water remained unchanged between 270 and 365 days of exposure to water. The stability of the surface layer during this time frame could potentially protect the surface of the Fe–Al–BG–CaZrTi2O7 composite material from further corrosion.

Phase field-volumetric lattice Boltzmann model of ion uptake in porous nuclear waste form materials under continuous flow

The flow field within the mesopores of sorbent particles plays a crucial role in radionuclide diffusion and ion uptake kinetics, thus, impacting the overall performance of porous nuclear waste form materials. To fundamentally understand the influence of microstructures and material properties on the radionuclide absorption and retention processes requires a coupled multi-physics model that considers the advection and diffusion within the flow field, the reaction at liquid-solid interfaces, and finally, the solid-state diffusion within a complex nanoporous medium. This study employs the volumetric lattice Boltzmann method (VLBM) to accurately and efficiently calculate the steady state velocity field inside the mesopores of sorbent particles. The obtained velocity field is then utilized to calculate the advection of ions in the steady flow. A phase field (PF) model of ion uptake is used to describe the reaction occurring at the solid-liquid interface and diffusion inside the porous medium. The integrated PF-VLBM model is verified in terms of the mass conservation and numerical efficiency and validated qualitatively with experimental observation data. Then, it is applied to study the influence of thermodynamic and kinetic properties, as well as flow field conditions on the ion uptake kinetics. The numerical results demonstrate that the ion uptake kinetics in porous particles has three distinct stages, which is in agreement with the observations in continuous flow experiments. In the first stage, the kinetics is predominantly controlled by the flow field and ion diffusivity in the liquid phase. The kinetics in the second stage is primarily governed by ion diffusivity in the solid phase. In the third stage the system reaches a dynamic equilibrium with a net zero uptake flux at the interface. It is also found that porous structures significantly affect the efficiency and capacity of ion uptake. The simulation results can help to understand the physics behind the observed ion uptake kinetics in experiments and to facilitate the development of constitutive equations that can account for heterogeneous microstructures in engineering performance codes.

Влияние защитного покрытия на процессы коррозии металлических контейнеров для иммобилизации высокоактивных радиоактивных отходов

The article explores nickel-based protective coatings and their influen e on high-temperature changes occurring on the inner surface of steel containers designed for HLW immobilization provided their interaction with the melts of model waste form materials. It evaluates the corrosion behavior of container surfaces with a nickel-based protective coating depending on its application method, i.e., galvanic and chemical. The study indicates the factors governing the corrosion depth, i.e., the depth to which the container steel surface may get altered depending on 4 different waste form material compositions. The study demonstrates high efficien y of the galvanic method in terms of nickelbased protective coating application and the major influen e that high sodium and calcium content has on the depth of corrosion penetration into container surface.

Process Development of Zirconolite Ceramics for Pu Disposition: Use of a CuO Sintering Aid

Zirconolite-structured ceramics are candidate wasteform materials for the immobilisation of separated Pu. Due to the refractory properties of zirconolite and other titanates, removing residual porosity remains challenging in the final wasteform product when utilising a conventional solid state sintering route. Herein, we demonstrate that the addition of CuO as a sintering aid increases densification and promotes grain growth. Moreover, zirconolite phase formation was enhanced at lower process temperatures than typically required (≥1350 °C). CuO addition allowed an equivalent density to be reached using process temperatures of 250 °C lower than the undoped composition. At 150 °C lower than the undoped zirconolite, the addition of CuO resulted in a favourable microstructure and phase assemblage, as confirmed via X-ray diffraction and scanning electron microscopy. Secondary phases of CaTiO3 and Ca0.25Cu0.75TiO3 were observed at some processing temperatures, which may prove deleterious to wasteform performance. The use of a CuO sintering aid provides an avenue for the further development of the thermal processing of ceramic wasteform materials.

Факторы выщелачивания фосфатной матрицы РАО в условиях глубинного захоронения

The article presents the results of experiments on the interaction of model phosphate waste form materials of various phase compositions containing simulators of RW elements with leaching solutions simulating groundwater and changes in its composition due to the impact produced by bentonite barrier material, radiolysis and alkaline plume from Portland cement concrete at temperatures of 25, 90 and 120°С under static water mode considered typical for deep RW disposal facilities. It provides numerical data on the equilibrium saturation concentrations of leaching solutions saturated with barrier and waste form components and simulators of RW elements during phosphate waste form leaching. The paper also discusses the degree of influence produced by the main leaching factors (the phase waste form composition, elevated temperature, the consequences of radiolysis and groundwater interaction with barrier and structural materials) on the concentrations of elements in RW phosphate waste form leachates under the disposal conditions.

Review of Rare-Earth Phosphate Materials for Nuclear Waste Sequestration Applications.

Several materials have or are currently being investigated for nuclear waste sequestration applications, including crystalline ceramic oxides, glasses, and glass-ceramic composites. Rare-earth phosphates have been investigated extensively for this application owing to the range of structures that the hydrous or anhydrous versions can adopt as well as the fact that naturally occurring rare-earth phosphates have been found to contain U or Th. The purpose of this mini-review is to discuss (generally) the properties that must be considered when identifying nuclear wasteform materials and (more specifically) the structure and properties of rare-earth phosphates with special attention being given to the resistance of these materials to radiation-induced structure damage. The last section of the mini-review contains an introduction to the development of glass-ceramic composite materials that contain rare-earth phosphate crystallites dispersed in a glass matrix. These composite materials have been suggested to be superior to using just glass or ceramic materials for nuclear waste sequestration applications owing to improved waste loading capabilities.

Effect of charge and anisotropic diffusivity on ion exchange kinetics in nuclear waste form materials

Motifs of radionuclide absorption particles (absorbers or fillers) for the nuclear waste treatment usually have crystal structures with nano-sized tunnels for fast ion diffusion. On the other hand, the accumulation of ions generates an electric field which affects the ion diffusion. Therefore, the strong anisotropic kinetic properties and the electric field affect ion exchange kinetics, especially in an assembly of these particles which may block the fast diffusion tunnels for each other. In this work, we developed a mesoscale model to describe the effect of anisotropic kinetic properties and electric field on ion exchange kinetics. With the model we simulated the effect of anisotropic diffusivity, electrochemical potential, particle morphology, size and aggregation on ion exchange kinetics during batch experiments. It is found that (1) ion exchange results in charge accumulation inside the particle because different ions have different diffusivities and different diffusion driving forces. The electric neutrality is not preserved; (2) the electric field speeds up the slower diffusion ions (either uptake or release) and slows down the faster diffusion ions to reduce charge accumulation and maintains charge neutrality; (3) the ion exchange becomes very slow at the later stages because (a) diffusion driving force (electrochemical gradient, Cs+ concentration in the solution and Na+ inside the particle) continuously decreases, as observed in batch experiments; and (b) the diffusivity decreases due to the Cs+ concentration dependence of ion diffusivity. This can explain the observed experimental phenomena that the radionuclide capacity usually cannot be reached; (4) for needle-like particle and their clusters, both surface area and diffusion distance along the direction with the fastest diffusivities have significant impact on the ion exchange kinetics.

Phase Evolution in the CaZrTi2O7-Dy2Ti2O7 System: A Potential Host Phase for Minor Actinide Immobilization.

Zirconolite is considered to be a suitable wasteform material for the immobilization of Pu and other minor actinide species produced through advanced nuclear separations. Here, we present a comprehensive investigation of Dy3+ incorporation within the self-charge balancing zirconolite Ca1–xZr1–xDy2xTi2O7 solid solution, with the view to simulate trivalent minor actinide immobilization. Compositions in the substitution range 0.10 ≤ x ≤ 1.00 (Δx = 0.10) were fabricated by a conventional mixed oxide synthesis, with a two-step sintering regime at 1400 °C in air for 48 h. Three distinct coexisting phase fields were identified, with single-phase zirconolite-2M identified only for x = 0.10. A structural transformation from zirconolite-2M to zirconolite-4M occurred in the range 0.20 ≤ x ≤ 0.30, while a mixed-phase assemblage of zirconolite-4M and cubic pyrochlore was evident at Dy concentrations 0.40 ≤ x ≤ 0.50. Compositions for which x ≥ 0.60 were consistent with single-phase pyrochlore. The formation of zirconolite-4M and pyrochlore polytype phases, with increasing Dy content, was confirmed by high-resolution transmission electron microscopy, coupled with selected area electron diffraction. Analysis of the Dy L3-edge XANES region confirmed that Dy was present uniformly as Dy3+, remaining analogous to Am3+. Fitting of the EXAFS region was consistent with Dy3+ cations distributed across both Ca2+ and Zr4+ sites in both zirconolite-2M and 4M, in agreement with the targeted self-compensating substitution scheme, whereas Dy3+ was 8-fold coordinated in the pyrochlore structure. The observed phase fields were contextualized within the existing literature, demonstrating that phase transitions in CaZrTi2O7–REE3+Ti2O7 binary solid solutions are fundamentally controlled by the ratio of ionic radius of REE3+ cations.

Cs3Bi2I9-hydroxyapatite composite waste forms for cesium and iodine immobilization

Perovskite-based ceramic composites were developed as potential waste form materials for immobilizing cesium (Cs) and iodine (I) with high waste loadings and chemical durability. The perovskite Cs3Bi2I9 has high Cs (22 wt%) and I (58 wt%) content, and thus can be used as a potential host phase to immobilize these critical radionuclides. In this work, the perovskite Cs3Bi2I9 phase was synthesized by a cost effective solution-based approach, and was embedded into a highly durable hydroxyapatite matrix by spark plasma sintering to form dense ceramic composite waste forms. The chemical durabilities of the monolithic Cs3Bi2I9 and Cs3Bi2I9-hydroxyapatite composite pellets were investigated by static and semi-dynamic leaching tests, respectively. Cs and I are incongruently released from the matrix for both pure Cs3Bi2I9 and composite structures. The normalized Cs release rate is faster than that of I, which can be explained by the difference in the strengths between Cs−I and Bi−I bonds as well as the formation of insoluble micrometer-sized BiOI precipitates. The activation energies of elemental releases based on dissolution and diffusion-controlled mechanisms are determined with significantly higher energy barriers for dissolution from the composite versus that of the monolithic Cs3Bi2I9. The ceramic-based composite waste forms exhibit excellent chemical durabilities and waste loadings, commensurate with the state-of-the-art glass-bonded perovskite composites for I and Cs immobilization.

Synthesis and characterisation of Ce-doped zirconolite Ca0.80Ce0.20ZrTi1.60M0.40O7 (M\u2009=\u2009Fe, Al) formed by reactive spark plasma sintering (RSPS)

Reactive spark plasma sintering has been utilised as a high-throughput processing route for the synthesis of two simulant zirconolite wasteform materials, targeting Ca0.80Ce0.20ZrTi1.60M0.40O7 (M = Fe3+ and Al3+). Materials were processed under 15 MPa uniaxial pressure, with heating/cooling rates of 100 °C/min to 1320 °C, maintained under vacuum. Despite moderate yield (> 80 wt%) of zirconolite-2M, a considerable Ce-rich perovskite phase was formed in both formulations, attributed to complete reduction of the Ce inventory to Ce3+, as determined by Ce L3-edge XANES analysis. The composition charge balanced with Al3+ was favoured on the basis of lower accompanying perovskite fraction.Graphical abstract

Phase and microstructure evolution of 0.2SiO2/Zr1-xNdxSiO4-x/2 (0 ≤ x ≤ 0.1) ceramics

Zircon (ZrSiO4) ceramics were widely used in the field of nuclear waste immobilization because of their excellent properties. However, the SiO2/ZrSiO4 ceramics were usually obtained in the preparation of ZrSiO4. The actinides immobilization behavior of SiO2/ZrSiO4 ceramics is unclear. Herein, Nd-doped 0.2SiO2/Zr1-xNdxSiO4-x/2 (0≤x ​≤ ​0.1) ceramics were fabricated and the effects of Nd-doping and sintering temperature on their phase and microstructure evolution were investigated. The obtained ceramics were characterized by XRD, SEM, BSE and EDS. Results show that 0.2SiO2/Zr1-xNdxSiO4-x/2 ceramics with 0≤x ​< ​0.04 are ZrSiO4 phase; while with 0.04≤x ​≤ ​0.1 are coexistence of ZrSiO4 and Nd2Si2O7 phases. The solubility limit of Nd3+ in 0.2SiO2/ZrSiO4 is about 4 ​at%. The lattice volume of ZrSiO4 increased with increasing Nd-doped content and sintering temperature. Furthermore, higher Nd-doped content and sintering temperature are beneficial to the densification of the obtained ceramics. The results demonstrate that the 0.2SiO2/ZrSiO4 ceramics can be employed as waste form materials to immobilize trivalent actinides.

Machine learning-enabled prediction of chemical durability of A2B2O7 pyrochlore and fluorite

Pyrochlore-structure type and its derivative in a general formula A2B2O7 (A = rare earth elements and actinides; B = Ti, Sn, Zr, Hf, Pb, Si, etc.) display excellent structural flexibility and rich crystal chemistry as promising nuclear waste form materials capable of immobilizing actinides and fission products. It is essential to understand these materials’ chemical durability and element release of radionuclides in order to evaluate their performance in near-field environment. However, it is a formidable grand technological challenge to experimentally perform durability testing across hundreds of thousands of possibilities resulting from their extreme compositional complexities due to cation substitutions at both A and B-sites. In this work, we demonstrate a machine learning approach to determine the key materials parameters and structural characteristics governing the leaching behaviors from a small set of selected compositions as model systems, enabling a science-based prediction of their chemical durability that can be extended to a wide range of chemical compositions. The combination of four key structural characteristics and materials parameters, including ionic radius size difference rA-B, ionic potential difference EB-A, electronegativity difference χB-A, and lattice parameter α, creates features an optimized prediction of the chemical durability. Two machine learning models, linear regression and Kernel ridge regression models, are trained on the randomly-split training dataset derived from the experimentally-determined elemental release rates, and subsequently tested on the testing dataset. The predicted leaching rates from both machine learning models show an excellent agreement with the experimental data, demonstrating the feasibility of rapidly evaluating the material properties of new compositions. These results highlight the immense potential of synergizing informatics through machine learning-based models and well-controlled experiments of selected model systems to accelerate materials design and discovery with optimized compositions and performance of promising materials for effective nuclear waste management.

Perovskite-Derived Cs2SnCl6-Silica Composites as Advanced Waste Forms for Chloride Salt Wastes.

Advanced materials and processes are required to separate halides and fission products from complex salt waste streams associated with the chemical reprocessing of used nuclear fuels and molten salt reactor technologies for immobilization into chemically durable waste forms. In this work, we explore an innovative concept using metal halide perovskites as advanced host phases to incorporate Cs and Cl with very high waste loadings. Wet chemistry-synthesized Cs2SnCl6 powders from CsCl salt solutions are successfully encapsulated into a silica matrix to form a composite using low-temperature spark plasma sintering with tunable Cs and Cl loadings up to 31 and 26 wt %, respectively. Chemical durability testing of the composite waste forms by semi-dynamic leaching experiments demonstrates that an incongruent leaching mechanism dominates the release of Cs and Cl. The metal halide perovskite-silica composite waste forms display exceptional chemical durability with the long-term release rates of Cs and Cl comparable to or outperforming the state-of-the-art waste form materials but with significantly higher waste loadings. The scalable synthesis of the metal halide perovskite from wet chemistry processes opens up new opportunities in designing advanced waste forms for salt wastes with very high waste loadings and exceptional chemical durability for the sustainable development of advanced fuel cycles and next-generation reactor technologies.

The formation of stoichiometric uranium brannerite (UTi2O6) glass-ceramic composites from the component oxides in a one-pot synthesis

Brannerite glass-ceramic composites have been suggested as suitable wasteform materials for high-actinide content wastes, but the formation of glass-ceramic composites containing stoichiometric uranium brannerite (UTi2O6) has not been well-studied. Uranium brannerite glass-ceramic composites were synthesised at by a one-pot cold-press and sinter route from the component oxides. As a comparison, two further samples were produced using an alkoxide-nitrate route. A range of compositions with varying molar ratios of uranium and titanium oxides (from 1:2 to 1:3.20) were synthesised, with a range of different heat treatments (1200 °C for 12–48 h, and 1250 °C for 12 h). All compositions were analysed by X-ray diffraction, scanning electron microscopy, energy dispersive X-ray spectroscopy, and X-ray near-edge spectroscopy, and found to contain UTi2O6 as the majority crystalline phase forming within a glass matrix of nominal stoichiometry Na2AlBSi6O16. In compositions with UO2:TiO2 ratios of 1:2 and 1:2.28, particles of UO2 were observed in the glass matrix, likely due to dissolution of TiO2 in the glass phase; this was prevented by the addition of excess TiO2. This work demonstrates the suitability of this system to produce highly durable wasteforms with excellent actinide waste loading, even with a simple one-pot process. Some grains of brannerite consist of a UO2 particle encapsulated in a shell of UTi2O6, suggesting that brannerite crystallises around particles of UO2 until either the UO2 is fully depleted, or the kinetic barrier becomes too large for further diffusion to occur. We propose that the formation of brannerite within glass-ceramic composites at lower temperatures than that for pure ceramic brannerite is caused by an increase in the rate of diffusion of the reactants within the glass.

Hot Isostatically Pressed (HIPed) fluorite glass‐ceramic wasteforms for fluoride molten salt wastes

AbstractMolten pyroprocessing salts can be used to dissolve used nuclear fuel from a reactor allowing recovery of the actinides. Previously, ANSTO have demonstrated hot isostatically pressed (HIPed) sodalite glass‐ceramic wasteforms for eutectic (Li,K)Cl salts containing fission products, but this system cannot be used for the analogous molten alkali fluoride salts (eg, FLiNaK), which have utility in the application of the next generation of nuclear reactors. In this work, a novel glass‐ceramic composite wasteform has been prepared by HIPing, as a candidate for the immobilization of fission product‐bearing FLiNaK salts. The wasteform has been tailored to immobilize the high fluoride content of the waste within fluorite, whereas the waste alkali elements are incorporated in a durable sodium aluminoborosilicate glass, with total waste loadings of ~17‐21 wt% achieved. It was also demonstrated that the speciation of Mo‐ and Sb‐simulated fission products was altered by adding Ti metal due to a controlled redox environment. The resulting candidate wasteform has been studied by X‐ray diffraction and scanning electron microscopy, including the HIP canister‐wasteform interaction zone, and its performance assessed via leaching studies using the PCT and ASTM C1220 leaching protocols. Dr Vance very much enjoyed the challenge of wasteform design for problematic nuclear wastes, for which fission product‐bearing FLiNaK salts are a clear example. His ability to hone in on a wasteform solution with viable waste loadings that meet performance requirements was testament to his nearly 40 years experience in nuclear waste immobilization. The samples discussed in this work represent the last wasteform materials that he prepared.

Immobilization of cesium and iodine into Cs3Bi2I9 perovskite-silica composites and core-shell waste forms with high waste loadings and chemical durability

Cs3Bi2I9, a defect perovskite derivative, is a potential host phase to immobilize iodine and cesium with high waste loadings. In this work, two strategies were explored to form Cs3Bi2I9-silica composites and a core-shell structure in order to improve chemical durability of waste form materials meanwhile maintaining high waste loadings. Cs3Bi2I9 loadings as high as 70 wt.% were incorporated into a silica matrix to form silica-ceramic composites, and 20 wt.% Cs3Bi2I9 was encapsulated into silica to form a core–shell structure by low temperature spark plasma sintering. Chemical durability of the composite and core-shell waste forms was evaluated by semi-dynamic leaching experiments, and Cs and I were incongruently released from waste form matrices. A BiOI alteration layer formed, acting as a passivation layer to reduce the release of radionuclides. The long-term iodine release rate was low (30 mg m−2 day-1) for the 70 wt.% Cs3Bi2I9–silica composite leached in deionized water at 90 °C, which can be further reduced to 5 × 10−3 mg m−2 day−1 for the 20 wt.% core-shell structure. This work highlights a robust way to immobilize the highly mobile radionuclides with high waste loadings through encapsulation into durable matrices and a surface passivating mechanism that can greatly reduce the elemental transport from waste form materials and significantly enhance their chemical durability.

Sintering of Metakaolin-based Na-Pollucite ceramics and their immobilization of Cs

Metakaolin-based Na-Pollucite ceramics (high-temperature sintered Na-pollucite ceramics: HTNP), with a target composition close to Pollucite (CsAlSi2O6), were fabricated at sintering temperatures above 1000 °C to immobilize Cs as waste form material. The sintering behaviors of Metakaolin-based Na-Pollucite ceramics were investigated, in particular with respect to sintering temperature, grain size, and microstructure. The results indicate that major phases of the pollucite ceramics consist of pollucite phase and a small fraction of other mineral phases like Albite, SiO2, and Al2O3, and the microstructures greatly depends on the molar Cs/Na ratio. Furthermore, with increasing Cs content in the pollucite ceramics, the sintering temperature increases gradually, and the pollucite grain size decreases accordingly. On the other hand, to avoid the high-temperature volatilization of Cs-ions during sintering, low-temperature sintered Na-pollucite (LTNP) ceramics were prepared below 950 °C by introducing H3BO3 into the metakaolin-based Na-Pollucite ceramics. After the LTNP ceramics were sintered using a low-temperature liquid-phase process, even when the sintering temperature was decreased to 650 °C, they remained structurally as pollucite phase. In addition, the sintering-temperature and grain-size dependency on the molar Cs/Na ratio in LTNP ceramics is consistent with that in HTNP ceramics. Leaching in the PCT (Product Consistency Test) indicates that the leaching concentrations of HTNP ceramics are, in total, lower than in LTNP ceramics. However, the LTNP and HTNP ceramics samples, with the molar Cs/Na ratios 1:1.5, 1:0.67, and 1:0.25 show better immobilization abilities for the Cs ion. This indicates that, besides the pollucite phase formation, the immobilization of the Cs ion correlates also with dense microstructures, grain size, and good crystallization. More importantly, low sintering temperatures and the fabrication of Cs crystalline ceramics are the important factors to improve the Cs ion immobilization.

Crystallization of Rare-Earth Phosphate-Borosilicate Glass Composites Synthesized by a One-Step Coprecipitation Method

Glass-ceramic composites are heterogeneous materials that have been suggested for use as nuclear waste form materials. A one-step coprecipitation method for synthesizing rare-earth phosphates (REPO...