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Zirconolite (CaZr_xTi_3-xO₇) has been widely studied as a candidate ceramic wasteform for immobilising actinide-rich wastes; this arises from its ability to accommodate actinides (U/Pu) in its crystal structure for long timeframes (~10⁶-10⁹ years) in geological environments. The addition of glass to zirconolite to fabricate glass-ceramics increases the flexibility to accommodate heterogeneous actinide-rich wastes while simplifying processing conditions; the latter was studied in this research. Cerium-bearing zirconolite glass-ceramics were fabricated by targeting zirconolite (Ca_0.8Ce_0.2ZrTi_1.6Al_0.4O₇; Ce as actinide surrogate) with varying amounts (0-100 vol.%) of glass (NaAl_1.5B_0.5Ca_0.7Ti_0.2Si₂O_8.6), followed by sintering at 1250℃-1450℃ / 6-24 h / air, to systematically investigate the effect of glass content on the phase assemblage (including composition) under various processing conditions. The calcined precursors possessed small particle sizes (< 25 μm) and this helped to minimise the influence of particle size on microstructural development during sintering for all samples. Thermogravimetric analysis and differential scanning calorimetry of the calcined precursors showed that glass addition (25-50 vol.%) lowered the zirconolite crystallisation temperature (1283℃ to 1250℃). X-ray diffraction, scanning electron microscopy, and X-ray energy dispersive spectroscopy analyses showed that glass addition lowered the sintering temperature (1320℃ to 1270℃) required to fabricate a near phase-pure glass-ceramic while minimising secondary phase formation (CeO₂/ZrO₂) with > 90% of cerium addition preferentially partitioned into zirconolite. These data showed that increasing the glass content (0-100 vol.%) and sintering temperature (1270°C to 1320°C) resulted in notable changes to zirconolite and glass compositions which was reflected in minor but systematic changes to zirconolite lattice parameters. X-ray photoelectron spectroscopy and X-ray absorption near edge spectroscopy analyses showed a predominance of Ce⁴⁺ incorporation as well as increasing amounts of Ce³⁺ due to auto-reduction at higher sintering temperatures (1270°C-1450°C). Overall, this research demonstrated that glass addition has the potential to simplify processing requirements of candidate glass-ceramic wasteforms for actinide immobilisation.

Aeschynite (ATiNbO₆, A = light rare earth elements) ceramics have been suggested as candidate wasteforms for high minor actinide content wastes. Whilst the possibility of high actinide loading and high aqueous durability of ceramic wasteforms, including the aeschynite structure, is attractive, they are often limited in chemical flexibility compared to glass wasteform materials. The use of glass-ceramics or glass-ceramic composite materials, where the actinides are preferentially partitioned into a specific ceramic phase and other elements are immobilised by the glass phase, can allow for immobilisation of the target radionuclides whilst also being able to safely manage the complex waste streams produced in reprocessing and clean-up operations.

This work examines the feasibility of forming aeschynite glass-ceramic composites, (Ce/Nd)TiNbO₆ in Na₂Al_2-xB_xSi₆O₁₆ glasses at varying glass:ceramic ratios, in a simple one-heat-treatment process. Preliminary work shows that, although aeschynite phases form well at 1200 °C in 50 wt.% Na₂AlBSi₆O₁₆ glass, secondary phases are also present in the produced materials. These include fergusonite (REENbO₄) and CeO₂, with no Ti-containing secondary phases observed. Given previous reports of the very poor resistance to amorphisation of the fergusonite structure, its presence in a wasteform is expected to be deleterious, particularly where amorphisation-induced swelling would lead to cracking of the glass phase in materials such as these.

Following studies performed on the impact of the glass composition (within the borated-albite glass system, Na₂Al_2-xB_xSi₆O₁₆) on the relative stability of zirconolite (CaZrTi₂O₇) formation against the formation of undesirable phases, glass-ceramic composites with varied glass composition have been produced. Similarly, the addition of an excess of TiO₂ has been shown to improve the final phase assemblage of brannerite (UTi₂O₆) glass-ceramic composites, so changes in ceramic-phase stoichiometry have also been examined, from ATiNbO₆ to ATi_1.3NbO₆. The produced materials have been characterised by X-ray diffraction, scanning electron microscopy, and Raman microscopy.

In some cases, nuclear wastes can be treated with ion-exchange materials to remove specific radionuclides from solution via cationic exchange. A promising inorganic ion exchanger, crystalline silico-titanate (CST) or IONSIV^®, has been employed within column systems to successfully remove Cs-137 from contaminated aqueous systems with high specificity. Once the Cs-137 has been incorporated within the IONSIV^® structure, the ion exchange material itself becomes radioactive waste and requires immobilisation within a nuclear wasteform for disposal. The current study investigated a series of advanced glass-ceramic wasteform design concepts for the immobilisation of Cs-loaded IONSIV^®. Key to the wasteform design strategy was to produce highly durable phases, a maximized waste loading and to provide a flexibility in the wasteform such that it can treat IONSIV^® wastes with various Cs-loadings. Non-radioactive Cs-loaded IONSIV^® was used to produce a flexible pollucite glass-ceramic wasteform following the addition of small quantities of glass formers and Fe. X-ray diffraction, scanning electron microscopy, and X-ray energy dispersive spectroscopic analyses showed successful formation of the pollucite glass-ceramic. Waste loadings of ~80 wt.% were achievable with Cs incorporated into the pollucite ceramic phase as targeted. The design also allowed for a wide variation in the Cs-loading of the IONSIV^® exchange material. The chemical durability of the pollucite glass-ceramic was assessed using the Product Consistency Test (PCT) ASTM C1285 standard protocol and evaluated in comparison to a glass wasteform designed for IONSIV^® immobilisation but with a relatively lower waste loading.

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 talk 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. This talk will cover the synthesis of anhydrous and hydrous rare-earth phosphates and the effect of single- and dual ion irradiation on the structure of rare-earth phosphates.

Due to various properties of interest, several phosphate-based ceramics have been considered for long as potential matrices for the long-term conditioning of radionuclides including actinides, in a deep geological storage. Among these ceramics, monazite-derived phases (Ln,An^III)PO₄ have been widely studied due to their high structural flexibility and remarkable chemical durability.

However, while the direct incorporation of An(III) has been already reported using both wet and dry synthesis methods, the incorporation of An(IV) is limited to only few dry studies due to problems associated to wet chemistry precipitation of initial precursors. Nevertheless, wet synthesis methods can lead to significant improvements in terms of chemical homogeneity and sintering capability, which can enhance the chemical durability of materials during leaching tests.

This work was devoted to the elaboration of synthetic monazites doped with tetravalent actinides (An = U, Th) by coupled substitution in order to prevent the formation of cationic vacancies. This mechanism consists in substituting two Ln(III) cations by one An(IV) and one M(II) cations.

The incorporation of An(IV) in the monazite structure was first performed using rhabdophane as starting precursor of Nd_1-2xCa_xAn_xPO₄ (x ≤ 0.1). The protocol based on hydrothermal conditions was optimised by varying the Ca/An stoichiometric ratio in order to prepare single phase samples. The thermal conversion of rhabdophane into monazite-cheralite was investigated under inert atmosphere and in air. It was followed by direct sintering in air for various holding times and heating temperatures. This one-step process allowed the direct preparation of dense pellets of monazite-cheralite. Sintering maps were further built in order to master the final microstructure of the ceramics. Finally, dissolution experiments were performed for various temperatures and leaching media to establish multiparametric dissolution rate laws. Preliminary leaching experiments revealed chemical durability as high as those obtained for undoped monazites.

The United Kingdom holds a substantial inventory of PuO₂, forecast to reach approximately 140 teHM (tonnes equivalent heavy metal) upon completion of reprocessing. This material presents a unique decommissioning prospect for which there is a need to develop a robust management strategy. Prompt immobilisation and disposal within a geological disposal facility (GDF) is a promising route towards ultimate disposition, yet in order to safely underpin the safety case for the geological disposal of Pu, it is necessary to understand the long term evolution of candidate wasteform materials in simulated repository environments. Moreover, there is a need to develop suitable wasteform materials capable of co-accommodating Pu, prescribed quantities of neutron poisoning species, trace processing impurities and transition metal cations capable of providing charge balance for non-stoichiometric compositions. Several baseline wasteform formulations derived from zirconolite, pyrochlore and fluorite-type matrices have been proposed on the basis of high chemical durability, radiation stability and moderate ease of processing. Herein, this talk will provide an overview in recent advances in the formulation refinement and fundamental characterisation of candidate wasteform materials for UK Pu. This includes detailed scoping trials aiming to characterise the incorporation of a representative U, Th and Ce surrogate fraction within zirconolite and pyrochlore phases, fabricated by conventional sintering (CPS), hot isostatic pressing (HIP) and reactive spark plasma sintering (RSPS).

Zirconolite (CaZrTi₂O₇) ceramics are proposed as a promising host-matrix for the immobilization of plutonium stockpile. In the literature, zirconolite ceramics are generally synthesized by conventional solid-state methods. Solid-state synthesis of Zirconolite ceramics require higher temeperatures and also leads to the formation of minor quantities of a chemically less durable perovskite (CaTiO₃) ceramic as a secondary phase. In order to prevent the formation of perovskite phase and to reduce the synthesis temperature, a low-temperature synthetic method such as the sol-gel method has been used in this study to synthesize a phase pure zirconolite ceramic. The following zirconolite compositions has been synthesized via sol-gel methods at lower synthesis temperatures: CaZrTi₂O₇, Ca_1-xCe_xZrTi_2-xFe_xO₇, and Ca_1-xCe_xGd_xZrTi_2-2xFe_2xO₇. The as-synthesized ceramics were characterized using powder XRD, SEM-EDS, and TEM. In order to study the effect of radiation on the structure of zirconolite ceramics, in-situ ion irradiation of synthesized ceramics was performed and the structural response was monitored using in-situ TEM.

Plutonium (Pu) wastes generated in the nuclear fuel cycle have significant radiotoxicity and require long-term immobilization. Zirconolite is one of the key mineral phases in the original titanate-based Synroc formulation and has long been considered a radiation tolerant and chemically durable host phase for Pu wastes. However, the Pu-bearing wasteform and the metallic canister used for containment of the waste material during hot isostatic pressing (HIPing), can potentially interact. This interaction may produce phases that are different to the bulk wasteform material or possibly with different chemical composition to that of the bulk, due to the potential for elemental diffusion across the canister-wasteform interface. In the present work, Pu-bearing zirconolite-rich full ceramic wasteforms were produced via HIPing in research-scale HIP canisters. Two different canisters, made from stainless steel (SS) and nickel (Ni), were used for this study to compare the effects of canister material on the wasteform-canister interaction. Scanning electron microscopy combined with energy dispersive X-ray spectroscopy (SEM-EDS) was utilized to perform a detailed investigation of the elemental compositions of the phases formed over the canister-wasteform interaction zone. These were then compared with phase compositions from regions near the center of the HIP canister. The wasteform sample HIPed in the SS canister showed ~100-120 µm of interaction zone dominated by high temperature Cr diffusion from SS to the wasteform with the Cr predominantly incorporated into the durable zirconolite phase. The wasteform sample HIPed in the Ni HIP canister showed almost no interaction zone with only minor diffusion of Ni from the canister into the wasteform near the interface. Though the HIP canister-wasteform interaction extends to ~120 µm when using a SS HIP canister, this translates to ∼0.2–0.3 vol% for an industrial-scale HIPed wasteform.

Vitrification is an internationally significant process for the disposition of nuclear waste. For several international vitrification projects, including of legacy nuclear wastes from the Hanford Site in the US, salt formation during operation of continuous melters is of substantial concern. The formation of molten salts during vitrification is detrimental due to 1) melter corrosion, 2) volatile release, 3) providing a conductive path between the melter heating elements (causing a short), and 4) segregation of waste components into nondurable water-soluble phases. As such, in-situ process monitoring is a critical technology for successful vitrification. Several in-situ process technologies for glass melters are fairly developed; however, the lack of in-situ surface salt formation detection methods presents a risk to vitrification at the Hanford Site Waste Treatment & Immobilization Plant Project (WTP). While proposed previously, millimeter wave (MMW) radiometry and interferometry are demonstrated for the first time for in-situ detection of salt formation in simulated nuclear waste glass melts. The experimental radiometer and interferometer setup uses the optical properties of the melt and a dual receiver at millimeter wavelengths to elucidate melt activity through assessment of emissivity and surface height changes. A series of previously characterized glasses designed to supersaturate sulfate (such as Na₂SO₄), chloride (such as NaCl), and fluoride (such as NaF) salts were analysed using the MMW radiometer and interferometer. providing insightinto volatile losses, fining, salt formation, salt identity, crystallization, and general emissivity properties of a heterogeneous melt. Thermal analysis, raman spectroscopy, X-ray diffraction, optical microscopy, and literature assessment of dielectric properties were utilized to verify observations from MMW radiometry and interferometry. The proposed contribution demonstrates MMW radiometery and interferometry as an useful method for in-situ salt detection to enable successful nuclear waste vitrification efforts.

Ongoing cleanup of radioactively contaminated seawater generated during the Fukushima
disaster involves the use of ion-exchange materials, which selectively adsorb radioisotopes
in solution. ⁹⁰Sr is amongst the most dangerous of these radioisotopes because its chemical
similarity to Ca promotes the bioaccumulation in bones and teeth, resulting in prolonged
internal exposure. Recyclable absorbents, such as granular sodium titanate (GST), avoid the
secondary waste generation associated with single-use adsorbents by allowing ⁹⁰Sr to be
eluted from the structure and precipitated into a desired compound, such as SrF₂, before the
adsorbent is reused. This waste precipitate requires immobilisation in a suitable wasteform
prior to final disposal. Iron-phosphate glasses are promising materials for nuclear waste
immobilisation due to their excellent chemical durability, high solubility limits, and low
melting temperatures.
In the present work, the influence of SrF₂ additions on the structure, thermal properties, and
phase formation of iron-phosphate glasses was investigated. SrF₂-loaded iron-phosphate
glasses were melted at 1100 °C, vitreous wasteforms were obtained for all compositions up
to 15 mol% SrF₂ loading, as confirmed by X-ray diffraction (XRD). Raman spectroscopy
revealed SrF₂ additions contributed to depolymerisation of the glass network. Differential
scanning calorimetry (DSC) measurements indicated increased SrF₂-loading raised the glass
transition temperature (T_g) and crystallisation temperature (T_c) of the glasses. Fluorine
retention was confirmed by compositional analysis, including electron probe micro-analysis
(EPMA) and X-ray fluorescence (XRF). Crystallisation of iron-phosphate compounds,
identified using XRD and Raman spectroscopy, was found to occur with increased SrF₂
loading. Iron-phosphate crystals were deficient in Sr with respect to the bulk glass, as
confirmed by SEM-EDS and EPMA. These findings suggest iron-phosphate glasses
represent a promising candidate for the vitrification and immobilisation of SrF₂.

Glass objects are reacted with aqueous solutions for multifarious commercial and scientific purposes. Understanding how glasses behave in solutions has been described by theory using a linear dissolution rate model, in which the rate decreases linearly as the silica concentration in solution increases. Such models are especially relevant for glasses that will be used as a waste form for disposal of nuclear waste. We melted a series of fourteen glasses in the system CaO-MgO-Al₂O₃-B₂O₃-SiO₂ in order to test how multicomponent glasses react primarily using single pass flow through (SPFT) systems. Four of these glasses contain three components, seven are four-component, and three are five-component glasses. All glasses contain 60 mol.% SiO₂, except the last five-component glass, which contains 50 mol.% SiO₂. This last composition acts as a good analogue for nuclear waste glasses. Experiments were conducted at pH 7.5 or 9.5 at 75 °C and glass structure was characterized by Solid State NMR. We also conducted flowing experiments in which the solution contained the stable isotopes ²⁶Mg, ³⁰Si, and ⁴³Ca to understand the mechanism of dissolution. After reacting in solutions at low (no or 10%), medium (50%), and high (100%) silica saturation levels for up to eight months, glass wafers were submitted for analyses by SIMS. Together, the dissolution data indicate that multicomponent glasses dissolve non-linearly with respect to dissolved silica with the steepest change occurring at the low SiO₂ concentrations. The data will be used to discuss the mechanism of reaction by a diffusion or dissolution-reprecipitation model.

The Low-Activity Waste (LAW) fraction of Hanford tank waste will be converted to glass at the Waste Treatment and Immobilization Plant (WTP) and disposed on the Hanford site. The durability of LAW glasses has been researched extensively for decades to satisfy contract requirements. To date, most LAW glass durability data has been generated via Product Consistency Test (PCT) and the Vapor Hydration Test (VHT) on non-radioactive simulant glasses fabricated via crucible melts. Non-radioactive glasses were chosen due to ALARA and cost reasons with confidence that radioactive waste glasses would have similar durability behavior through understanding of glass corrosion. To reduce the risk of significant differences in laboratory test response data between WTP melter waste glass and its associated simulant glass, PCT, VHT, and EPA1313 durability tests were performed on actual and simulated LAW glasses fabricated using identical laboratory-scaled melters and with the same procedures, equipment, and location. Actual and simulant glass durability results (including normalized B, Na, Tc, and Re releases) generated from the durability experiments are presented and statistically compared relative to experimental uncertainty.

A comprehensive understanding of the long-term durability of radioactive waste glasses is necessary to ensure the accuracy of future predictive modelling of waste glass alteration, and therefore the integrity of any future disposal solutions. The UK currently has plans to permanently dispose of its radioactive waste, including the vitrified by-products of an extensive reprocessing programme, underground in a deep geological disposal facility (GDF). The potential resumption of alteration of radioactive waste glasses in groundwater remains a source of uncertainty in the performance of any such proposed GDF and could result in a resumed rate of alteration approaching that of the most rapid initial rate. Recent research in France and the US has confirmed the prevalence of the resumption of alteration phenomenon in numerous simulant waste glass and simplified analogue compositions by ‘seeding’ glass dissolution vessels with zeolites at high pHs and temperatures to induce the effect artificially. The disparate composition of UK waste forms prevents a direct comparison with these international standards. However, this study has found not only that this phenomenon occurs in simulant UK waste glass compositions when seeded with natural analcime crystals, but also under a wide variety of conditions, including at free pH (unbuffered deionised water). No resumption was observed at 40 °C, however, at 90 °C, resumption was observed at free pH (pH_(25°C) ~ 9.7), as well as starting pH_(25°C) artificially raised with KOH to 11, 11.5, 12, and 12.5, with varied results suggesting the existence of complex buffer mechanisms. The time of seeding relative to the initiation of the experiments was varied to no effect, suggesting that the resumed rate is not closely related to alteration layer thickness. Variations in the initial seed-to-glass surface area ratio also had no effect, suggesting that a much smaller amount of the seed could induce the same effects.

For several decades, ceramic matrices have been studied as potential materials for the specific immobilization of long-lived radionuclides such as iodine, cesium and actinides. Several of them have been considered and optimized on the basis of the existence of natural analogues. The first step in the search of a radwaste ceramic concerns the phase's ability to immobilize the targeted radionuclides and their daughter products. With this in mind, systematic studies are being carried out to analyze the various substitutions required for effective long-term immobilization. While the initial studies reported in literature involved dry chemistry routes, more recent studies involving wet processes have been developed. The latter make it possible to improve synthesis conditions, particularly in terms of homogeneity, radionuclide incorporation rate and reactivity of the initial powders. This can induce improved sinterability. Several phosphate, silicate and oxide ceramics have been recently prepared in this way, using sol-gel, direct precipitation or hydrothermal processes.

Since radionuclides must be immobilized as monoliths, optimized precursors must be densified by melting or, more frequently, sintering. To achieve such a densification, the powders are generally shaped by uniaxial pressing at room temperature, then the pellets are subjected to a high-temperature calcination step. Several microstructural parameters must then be considered, such as grain size (thus grain boundaries density), absence of secondary phases that could alter the chemical durability of the final ceramics, or presence of pore networks that could contribute to accelerated leaching rate.

Chemical durability during leaching or weathering tests under long-term storage conditions is one of the most important properties for validating the use of a radwaste matrix. This study must usually include a kinetic component, allowing the determination of multiparametric dissolution law. This is obtained by independently varying several parameters such as solution pH, temperature and concentrations of active species at the solid/liquid interface. Under these conditions, qualification tests such as MCC1 do not provide sufficient knowledge of the system to allow long-term extrapolation of the long-term behavior of the ceramic material. This kinetic study must be complemented by a thermodynamic approach aimed at describing saturation phenomena. This includes the identification of potential secondary phases likely to act retroactively on the long-term behavior of ceramics. Coupling structural and chemical characterization of these phases with speciation calculations in solution enables us to assess their solubility product and thus their capacity to delay radionuclide migration. As the confined elements are radioactive, irradiation phenomena play an important role both during the elaboration and sintering stages and during leaching tests, notably through radiolysis phenomena in solution and at the solid/solution interface.

This presentation will focus on several examples to illustrate the various steps involved in qualifying a specific radwaste matrix. We will also discuss the pitfalls that can lead to poor knowledge of the system under study, and impact extrapolation of the long-term behavior of the ceramic prepared.

The idea of immobilizing radionuclides in crystalline host materials was put forward 70 years ago. Since then, continuous research has been conducted on a wide variety of crystalline materials that are considered as possible host matrices. However, many challenges remain, owing, e. g., to the complex chemistry of nuclear waste streams and the exceptionally high requirements regarding physical and chemical long-term stability.

Monazite (LnPO_4, Ln = La-Gd) has long been considered as one of the most promising crystalline host materials for long-term storage of radionuclides, especially actinides. The main reasons for this are its chemical flexibility, its excellent aqueous durability and its low recrystallization temperature, which allows for rapid self-healing of radiation induced damages. It has been shown that monazites can accommodate large amounts of trivalent actinides within their crystal structure. However, the incorporation of tetravalent dopants via coupled substitution with divalent cations has proven challenging, even though natural monazite is known to contain significant amounts of Th and U (combined up to 27 wt-%).

To facilitate assessments with respect to selection criteria such as chemical flexibility, radiation resistance and aqueous durability, efforts are made to identify inactive surrogate-models. The use of cerium as a surrogate for tetravalent actinides will be discussed for monazite-type phases based on the solid solution La_1-x(Ca,Ce)_xPO₄ which was extensively studied using powder and single crystal XRD, electron imaging techniques including EPMA, SEM and TEM as well as spectroscopic measurements including Raman, TRLFS, EXAFS and in-situ XAS experiments. Based on these findings the synthesis of active monazites containing up to 50 % Th⁴⁺ was successfully performed as shown by PXRD measurements.

The poster will focus on irradiation studies of monazite-type ceramic pellets from the solid solution La_1‑xCe_xPO₄. Monazite is known for its remarkable ability to recover from radiation damage by a combination of low recrystallisation temperatures (~570 K) and low activation energies for thermal annealing (<3 eV) as well as irradiation-induced recrystallisation which was observed both from external irradiation and self-irradiation.

While various studies have been published investigating the effects of radiation on the monazite structure, the impact of disorder introduced by solid solutions has not yet been studied extensively. For this reason, various compositions of the aforementioned solid solution were irradiated with Au ions at two different fluences and subsequently analysed with SEM, gracing incidence XRD and Raman spectroscopy in order to gain a better understanding of their radiation stability and recrystallisation properties.

The safe disposal of nuclear waste is one of the intergenerational issues which needs to be solved. A potential route to effectively immobilize radionuclides could be realized by their incorporation into crystalline solid phases in future radioactive waste repositories. In particular, the immobilization of specific waste streams containing minor actinides (Np, Am, Cm) or plutonium in crystalline solid phases may be advantageous compared to glass matrices, which may be less resistant to leaching and disintegration [1-3]. Due to their radiation stability and chemical and structural flexibility, monazite-type compounds are considered suitable matrix materials [4]. To better understand structural changes due to radiation damage, synthetic monazite single crystals with different chemical compositions (La, Nd, Pm, Sm)PO₄ were synthesized by a high-temperature (flux method). Irradiation experiments were performed at the UNILAC beamline of GSI Helmholtz-Centre Darmstadt using 1.7 GeV Au ions and fluences of up to 1e13 ions/cm². The irradiated single crystals were characterized by Raman spectroscopy, secondary electron microscopy and single crystal X-ray diffraction. The irradiation of monazite with 1.7 GeV Au ions results in an embrittlement of the crystals and the formation of a glassy surface layer of about ~48 μm thickness, which correlates well with the projected range of ~44 µm according to SRIM-2013 calculations [5]. The irradiation results in a significant broadening of the Raman modes and further changes in the lattice dynamics. X-ray diffraction experiments revealed the amorphization of the surface layer.

The presentation gives an overview of the structural changes of La monazite single crystals under swift heavy ion irradiation at ion fluences of 5e11 ions/cm² ,1e12 ions/cm² and 2e12 ions/cm². After irradiation, cross sections of the single crystals were prepared and additionally polished with an Ar ion mill to investigate the surface damage along the path of the fast heavy ions using optical light microscopy and Raman spectroscopy. The methods used show a strong surface damage within the projection range of the gold ions due to color change, increase of the FWHM of the Raman band and decrease of crystallinity.

[1] Donald et al. (1997) J. Mater. Sci. 32; [2] Ewing (1999) PNAS, 96; [3] Lumpkin et al. (2006) Elements, 2; [4] Schlenz et al. (2013) Z. Kristallogr. Cryst. Mater. 228; [5] Ziegler et al. (2010) Nucl. Instrum. Methods Phys. Res. B 268

The authors acknowledge the BMBF for financial support in the project No. 02NUK060.

Monazites are rare earth phosphates that are potential host matrices for the immobilization of actinides in high-level radioactive waste streams. This is due to their ability to incorporate various cations through different substitution mechanisms as well as their radiation resistance as observed in natural monazite mineral samples. In this study, LnPO₄ monazite ceramics and single crystals doped with 500 ppm Eu^III as a luminescent probe were irradiated with heavy ions to simulate the recoil of daughter products that occurs during alpha decay of the actinides. More specifically, irradiation experiments were conducted either with 14 MeV Au ions at fluences ranging from 5×10¹³ – 1×10¹⁵ ions/cm² or with swift 1.7 GeV Au ions at fluences of 5×10¹¹ – 2×10¹² ions/cm².

Irradiated monazite ceramics were analyzed with electron microscopy (SEM), vertical scanning interferometry (VSI), grazing incidence diffraction (GID), Raman spectroscopy, and luminescence spectroscopy to probe long- and short-range order of the monazite microstructure.

SEM micrographs and VSI data show clear damage of the irradiated regions of the ceramics, in the form of swollen grains and enlarged grain boundaries. GID images and powder patterns reveal diffuse scattering and amorphous contributions in irradiated samples. Solid solution compositions show larger damage than corresponding monazite endmembers, while polycrystalline and single crystal samples show a similar level of amorphization. In the local coordination environments, Raman spectra of irradiated samples display a shoulder on the ν₁ peak, indicating disruption in the vibrational modes of the phosphate tetrahedra. Luminescence data illustrate ion-irradiation-induced changes in the local LnO₉ polyhedral environment in the monazites. Integrated excitation spectra show a difference in the intensity and position of the excitation peak with irradiation. Especially single crystal data show a systematic decrease of the local site symmetry of the Eu³⁺ cation, and a general broadening of emission spectra, indicative for reduced local order following amorphization.

When placed underground for long-term storage (~10^6 years), phases containing radioactive species are exposed to elevated temperature, T, and pressure, P. Corresponding values can reach ~400˚C and 0.4 GPa for repositories situated 15 km underground. Moreover, formation of He bubbles as a result of undergoing α-decay generates regions where P can reach 10 GPa. Temperature and pressure are powerful thermodynamic parameters that can influence structural and physical properties of materials. These factors, therefore, have to be considered when evaluating performance of nuclear phases to be placed in underground repositories for eternity. Corresponding experimental simulations of accelerated aging can be achieved by subjecting candidate materials to extreme conditions. This would allow to conclude on their long-term stability under harsh conditions. We illustrate this approach on Zr-based ceramic materials as hosts for actinide elements.

Studied phases were Y-stabilized ZrO₂ (YSZ), ZrSiO₄ and GdZr₂O₇ doped with tetravalent ions. Zr-based materials can incorporate substantial amount of actinide elements, as was shown for instance of YSZ. In addition, corresponding tetragonal and cubic YSZ modifications exhibit excellent phase stabilities at elevated T. While HP induced phase transformation in tetragonal YSZ, concentration of the incorporated Th ions remained constant up to at least 12 GPa. Contrary, application of HP induced discharge of incorporated Th atoms in ThxZr1-xSiO₄ system. Nevertheless, stable compounds in this system have been identified and corresponding formations regions were found to be strongly influenced by synthetic conditions. While GdZr₂O₇-based phases were found to be stable at lower P, complete amorphization was observed at P > 40 GPa. Corresponding behavior will be illustrated on synchrotron radiation diffraction experiments with in situ application of extreme T and P. We propose that studies under extreme conditions to be included in a standard protocol for evaluation of materials to be placed for long-tern in underground nuclear waste storage.

In modern UO₂ nuclear fuels, often known as advanced fuels, the use of transition metal (TM) elements as dopants, such as Cr, have been shown to increase the in-reactor fuel performance over traditional non-doped variants. The improved fuel performance arises from enhanced grain growth phenomena of the fuel microstructure. These are dependent upon the chemistries of the dopants during high temperature sintering a part of fuel fabrication, in which specific conditions, i.e. the oxygen partial pressure and temperature, directly control the grain growth. Despite this and other advances in the science behind Cr- and other TM-doped UO₂ modern nuclear fuels, significant paucities of information remain regarding the mechanism for incorporation and formation of secondary phases. Pertinently prior to this investigation, there was no definitive conclusion as to the valence state and local environment of Cr and other TMs elements within the fuel matrix. As this presentation will demonstrate, this paucity of understanding originates from the complexity of chemical states adopted in the bulk ceramic material, which has commonly been investigated in literature. To ameliorate this and to provide novel investigative direction of the TM chemistry in UO₂ fuels, we have fabricated single crystals of Cr/Mn/V/Fe-doped UO₂ that are representative of the bulk fuel material. By studying these comparatively against the bulk material using a variety of advanced spectroscopic techniques (HERFD-XANES, EPR, EXAFS), we have been able to resolve the chemistries of these TM dopants conclusively and unambiguously in both fresh and spent fuel. Our results further corroborate previously thermodynamic models proposed for some of these materials. The results of this investigation will be discussed in detail in this contribution, focusing on the chemistries of particularly Cr, in addition to Mn, V and Fe and compared against current literature.

Recent studies have suggested that crystalline ceramic matrices, such as monazites and zirconia (ZrO₂) have a high potential to be used as immobilization matrices for radioactive waste. At room temperature, zirconia has a monoclinic (m) structure. At higher temperatures, tetragonal (t) and cubic (c) structures can be stabilized. The phase stabilization can also be achieved at ambient conditions by incorporating oversized cations. In addition, several metastable phases (t′, t′′, κ, and t*), can be formed for doped zirconia materials. Out of the several structural polymorphs, especially the cubic structure shows high radiation tolerance, which is important for host matrices containing radioactive elements. In the current study, cerium has been used as an analog for plutonium as these f-elements have identical cation radii and can be stabilized in the trivalent and tetravalent oxidation states. The zirconia samples were co-doped with a small amount of Eu(III) to allow for luminescence spectroscopic analyses of the solid phases. In a first step, the co-precipitation route was applied to synthesize Ce-doped zirconia samples over a wide Ce-concentration range. The phase composition of the samples was investigated with X-ray diffraction, and showed that the radiation tolerant cubic phase was stabilized only for samples with Ce concentrations above 75 mol% . At lower dopant concentrations, a mixture of different phases were present, including monoclinic in a low doping concentration range, tetragonal and tetragonal double prime phases appearing for intermediate Ce-concentrations. The latter phase was detected only by Raman spectroscopy, showing the presence of a defect band at 526.5 cm^-1. In addition, luminescence spectroscopy revealed structural changes in terms of different Eu environments in the t´´ and c samples. To stabilize the cubic phase for low tetravalent doping concentrations, trivalent yttrium (Y) was incorporated as a co-dopant. XRD and Raman analyses show that the cubic phase was stabilized when the concentration of Y was higher than 15 mol%. Finally, using the same co-precipitation route, a series of uranium-doped zirconia samples was synthesized. XRD investigations show a phase transformation from monoclinic to tetragonal and orthorhombic with increasing uranium doping. Identical to the Ce-doped samples, the pure cubic phase was stabilized only in the presence of Y for concentrations higher than 15 mol%. Discerning the crystal structure is crucial to understanding the properties of these phases. Although the binary zirconia systems with only one dopant show different phase compositions for Ce and U, the scenario changes when adding a trivalent co-dopant such as yttrium, which stabilizes the cubic phase both in the presence of uranium and cerium.

Preliminary solubility results for the pure cubic phase of uranium/cerium-doped zirconia co-doped with yttrium will be shown in the poster session.

The advent of modern scanning laser-based Raman spectrometers have allowed for increasingly sophisticated microstructural analysis with high spatial resolution. Fluorescence, long considered a nuisance in these measurements, is now being applied as a direct output signature from these experiments. Rare earth elements (REE) are particularly suited as probe ions for these experiments, being important fission products and having f-f transitions split by local crystal field, allowing investigation of localized phenomena such as partitioning and atomic environment. Laser excitation wavelength selection combined with limited bandwidth gratings makes certain ions highly sensitive to given experimental conditions (e.g., Pr³⁺: 455 nm light with 50-3500 cm^-1 grating, allowing investigation of 456-541 nm light; Sm³⁺, Dy³⁺: 532 nm light with 86-6000 cm^-1 grating, allowing investigation of 534-781 nm light; Nd³⁺, Yb³⁺: 785 nm light with 50-3500 cm^-1 grating, allowing investigation of 788-1054 nm light). In this talk, we will briefly summarize the physics of these experiments, then provide several examples of its use as applied to nuclear materials. Examples include glass-ceramics with multiple rare-earth containing phases, radiation damage in ceramics and natural analogues, and doping of fuels and fuel surrogates, as well as assessment of purity of raw materials. Though REE are particularly suited to these methods, investigation of other metals in various matrices is possible, such as the luminescent transitions of Cr³⁺ and its sharp R-line used in ruby lasers.

Positrons, after implantation into materials, rapidly thermalize and then annihilate with electrons. In the presence of vacancies and other open volume defects positrons trap there prior to annihilation. Conservation of energy and momentum leads to Doppler broadening of the annihilation line which carries information about the annihilation site. Combined with the use of variable energy beams this enables the identification and assessment of vacancies as a function of depth down to about 5 micrometers from the surface. Nuclear materials such as fuel elements, waste, or glasses for vitrification of such waste are exposed to energetic particle irradiation from fission products or self-irradiation of the nuclear materials. Defects including vacancies are generated and alter the material properties. Helium can accumulate as bubbles in voids. Point defects are generated. In this presentation I will discuss the benefits of positron annihilation spectroscopies in nuclear materials and what can be learned. Examples include recent work on the leaching of ISG glasses used for waste vitrification as well as defects generated by ion implantation into oxides and metals.

Various glass and ceramic waste forms have been proposed for high-level nuclear waste (HLW) immobilisation and geological disposal in countries such as the United Kingdom, United States and France. One of the major concerns related to geological disposal is the corrosion of the waste form due to groundwater inlet in these underground facilities and the role radiation damage has on the corrosion of these materials. In the literature, a limited amount of research has been performed on calculating the surface corrosion rate of these glass and ceramic waste forms, largely due to the lack of experimental techniques that can quantifiably measure the surface topography of the material in-situ during an aqueous corrosion experiment. Therefore, new, and novel state-of-the-art experimental techniques are required to study the surface corrosion behaviour of potential glass and ceramic host materials for HLW immobilisation. In this study, ceramic phases of Zirconolite and Perovskite (CaZrTi₂O₇ and CaTiO₃), a glass-ceramic material and a glass material (International Simple Glass 2) were irradiated using Xe²⁺ ions in the Microscope and Ion Accelerator for Materials Investigation (MIAMI-2) Facility at the University of Huddersfield. The samples were then corroded in aqueous solutions of deionised water and a one molar solution of NaOH with in-situ topographic measurements taken using High-Speed Atomic Force Microscopy (HS-AFM) and Interferometry to precisely study surface corrosion rates of these four different irradiated materials in different pH solutions. Post-corrosion TEM was additionally carried out on the corroded samples to offer complementary sub-surface measurements to the above-surface measurements provided by HS-AFM and Interferometry. These results therefore allow us to study both diffusion and dissolution during aqueous corrosion.

Several different types of ceramics have been proposed as potential candidates for the management of radioactive wastes/isotopes found at the back end of the nuclear fuel cycle. Long-term management of such radioactive wastes will involve encapsulation and geological disposal in a specifically engineered geological disposal facility (GDF). Self-irradiation damage, helium bubble formation, and corrosion in conditions typical of a GDF are expected to alter their physiochemical properties and corrosion potentially impacting the release of the radioactive elements into the biosphere. A fundamental understanding of the contribution of such factors towards the overall corrosion is thus an important part of the safety assessment and confidence building. In the majority of the cases, short-term experiments are conducted under controlled conditions on simple to complex model systems, and results are extrapolated to time scales more representative of the expected lifetime of the GDF (hundreds of thousands of years). Luckily, several of these synthetic materials have almost exact natural radioactive analogs, and their studies, both, in terms of their radiation response and weathering provide invaluable data and information for the validation of the short-term controlled experiments.

In this presentation we aim to present results from novel methodologies to study short-term corrosion in almost real-time and combine these with the studies of natural analogies, which have gone weathering for millennia, to develop a more coherent picture of how radiation damage and corrosion could affect the materials used for the conditioning of radioactive elements.

Nuclear fuel reprocessing and waste treatment facilities generate significant quantities of off-gas, which contain volatile radioactive and hazardous elements and compounds that must be captured and safely disposed of. To do that, an efficient and integrated off-gas treatment system is required to meet stringent regulatory requirements for operation, monitoring, and emissions control. The specific design and configuration of this system vary depending on the industry and process. However, a common theme is the utilization of solid sorbent materials to efficiently remove contaminants from various gas streams. There are a large number of sorbents at various stages of development that are being investigated and studied to capture mercury and iodine. The downside is that most of them were not tested under relevant process conditions. This presentation will review available sorbents for iodine and mercury against criteria for deployment in off-gas systems, addressing their performance in different environments and possible disposition pathways. Also included will be a discussion of examples of off-gas system designs and flow sheets from nuclear reprocessing facilities and the Hanford Vitrification Plant.

The deep geological repository concept for radioactive wastes is based on the confinement of the radioactivity over long periods of time by a multiple barrier system. Cementitious materials are used as part of the barriers in most of the repository concepts developed internationally (e.g., as backfill, tunnel lining, or in shaft seals and plugs). Although the behaviour of safety-relevant radionuclides in cementitious environments has been investigated extensively in the last decades, the impact of organic degradation products, originating from organic waste components or from superplasticisers in cementitious materials, on the migration of radionuclides under highly-alkaline, cementitious conditions is not yet fully understood. Therefore, the objective of this work, carried out within the framework of EURAD WP CORI (Cement-Organic-Radionuclide Interaction), was to fill knowledge gaps in the understanding of the impacts of the presence of phthalate (C₈H₄O₄²⁻; degradation product from plasticisers in PVC) and tri-methyl-amine (TMA; degradation product of ion exchange resins) on the migration behaviour of ²⁴¹Am and ¹⁵²Eu in cementitious barriers.

In this context, hardened cement pastes (HCP) were prepared with a water/cement ratio of 0.40 from a composite cement (CEM V/A 42.5N; Calcia, Rombas). The uptake and diffusion of ²⁴¹Am and ¹⁵²Eu in HCP was studied under anoxic conditions in the presence and absence of organics. In the absence of organics, a strong retention of both radionuclides on HPC was observed (R_d values between 10⁵ and 10⁶ dm³ kg^-1). In contrast, at phthalate concentrations exceeding ~10^-3 M, a reduction in the uptake of ²⁴¹Am and ¹⁵²Eu on HCP by several orders of magnitude was observed. This reduction in sorption could be the consequence of the decalcification of calcium silicate hydrates (C-S-H), the main sorbing phase in cementitious materials, due to the increasing formation of Ca-phthalate complexes in solution. These results indicate an increase in the mobility and diffusion of ²⁴¹Am and ¹⁵²Eu in cementitious barriers with increasing phthalate concentrations.

Acknowledgements

The EURAD-CORI project leading to this application has received funding from the European Union’s Horizon 2020 research and innovation programme under grant agreement No 847593.

To dispose of radioactive waste in the radioactive waste repository, the radioactive waste should satisfy the waste acceptance criteria of the repository which differ according to the site of the repository. Among the waste acceptance criteria, a leaching rate of the radionuclides in the waste is one of the main criteria which is directly related to the isolation of the radionuclides from the biosphere. However, the leaching rate of the radionuclides varies followed by the test conditions of the leaching test.

According to the chemical environment of the leachant, the chemical form of the radionuclides varies from precipitate to ion. For instance, cobalt exists as a cobalt ion in the H₂O system with a pH lower than about 9 while cobalt exists as cobalt hydroxide when the pH of the leachant is higher than 9. In addition, the adsorption of the nuclides differs followed by the nuclide concentration which affects the leaching rate. However, several studies prepared waste specimens with high concentrations compared to low-level waste to induce the measurable concentration of the leached nuclides. Therefore, the leaching behavior of the nuclides according to the test condition should be compared to avoid both over- and underestimation of the leaching rate.

In this study, the leaching behavior of Cs, Sr, and Co under several leachant types and concentrations is estimated. The cement-solidified specimens containing single Cs, Sr, and Co were manufactured. The leaching test following ANS 16.1 was performed by applying deionized water and cement-saturated groundwater. As a result, a leachability index difference according to the leachant type and nuclide concentration was discussed. The result of this study is expected to be background data that helps understand the actual leaching behavior of the Cs, Sr, and Co in the low- and intermediate level waste repository.

The management of nuclear waste is a major concern for the nuclear industry and society as a whole. Liquid nuclear waste requires special attention due to its potential for environmental contamination and the long half-life of the most common nuclides, such as Cs-137, Sr-90, and Eu-152. Several studies have addressed the use of Alkali Activated Materials (AAMs) for the immobilization of radioactive waste containing the before-mentioned nuclides. These studies have shown that AAMs can effectively immobilize these elements by forming stable phases that incorporate them into the structure of the material. The incorporation of these elements into the AAMs reduces their release and enhances their long-term stability, making them suitable for long-term storage. However, their integration into AAMs can also affect the properties of the encapsulation matrix. It is essential to understand the effect of these radionuclides on the properties of AAMs to ensure that the resulting material meets the necessary criteria for long-term storage. Furthermore, when compared to conventional water technologies, nanomaterials show great promise in removing heavy metals and radioactive ions from water because of their capacity to integrate different properties, creating multifunctional systems. In particular, CeO₂ nanoparticles have proven to be effective free-radical scavengers, providing defense against chemical, biological, and radiological abuse. In this study, inorganic polymers (IP) of different structural compositions were synthesized, and doped with different combinations of CsNO₃, Sr(NO₃)₂, Eu(NO₃)₃, and CeO₂ nanoparticles. The IP samples were developed from copper slag and a sodium silicate solution. Samples were tested on their microstructural (Scanning Electron Microscopy, Energy-Dispersive X-ray Spectroscopy) as well as their physicochemical (X-ray Fluorescence, Calorimetry, Iron oxidation state) properties in order to assess the influence the dopants have on the alkali-activated structures. Furthermore, the ability of IPs to retain the contaminants was tested with an up-flow percolation test.

Calcium silicate hydrate (C-S-H) is formed as a secondary mineral under the condition saturated with groundwater around radioactive waste disposal sites. The C-S-H is also a main component of cementitious materials and significantly adsorbs cationic radionuclides. However, it is considered that if the structure of C-S-H is altered, for example, by containing Mg from groundwater or host rock, the sorption characteristics for radionuclides may also be changed. Thus, in this study, the sorption behavior of Cs, including Cs-134 and C-137 in high-level radioactive wastes, to Mg-containing C-S-H is experimentally evaluated as a fundamental study.

The Mg-containing C-S-H was synthesized with a (Ca+Mg)/Si molar ratio of 0.4 – 1.6 and Mg content of 0 – 20% to Ca amount, by mixing CaO, SiO₂, Mg(NO₃)₂, NaOH to adjust pH, and ultra-pure water in given amounts. The sorption experiment was carried out by simultaneously adding CsCl solution to be 1.0 mM to synthesize the Mg-containing C-S-H without any drying process. The liquid/solid weight ratio was 20 mL/g, and the total volume of the solution was 30 mL. The curing was 7 and 42 days at 298 K with shaking at 120 strokes/min. As the results, the sorption ratio slightly decreased with increasing the (Ca+Mg)/Si molar ratios. Furthermore, the Raman spectra suggested that the incorporation of Mg into the C-S-H structure decreases the sorption site by facilitating the polymerization of the silicate chain. However, high sorption distribution coefficients of K_d= 5.5 – 10.2 mL/g were estimated in (Ca+Mg)/Si=0.8 with Mg-content up to 20%, as an example of secondary mineral. Moreover, the K_d for all samples exceeded the K_d of 0.04 – 0.4 mL/g for the plutonic rocks. This suggests that C-S-H contributes to the immobilization of Cs without decreasing its sorption performance, even if C-S-H incorporates Mg into its structure.

Bentonite has strongly desired properties for its use in an engineered barrier system – swelling and sealing capabilities, high sorption capacity, for containment and sorption of migrating radionuclides, resulting in low permeability. A geological transformation of smectite (the main component of bentonite) to illite, also known as illitization, has been widely studied and occurs with high temperatures, pressures, and an external K⁺ source. However, this is detrimental to the barrier system as the barrier loses its most critical properties and can easily transport radionuclides to the environments in the far field. We have performed extensive research at Sandia National Laboratories to determine what physical and chemical conditions result in the formation of mixed-layer illite/smectite and complete transformation to illite. Our Parr Vessel reaction studies entail the use of different cation exchanged smectite clays (Na⁺, Cs⁺, K⁺), high temperature (200 °C), various reactor solutions, times ranging from 7 to 112 days and liquid to solid ratios of 100, 500 and 1000. The solid reaction products were analyzed with XRD – air dried and ethylene glycolated mounts – and SEM-EDS. The clay recovered from the reactors was also Na-exchanged to determine any K⁺ fixation within the clay interlayer. Analysis from ICP-OES was also used to characterize the liquid chemistry from the hydrothermal reactions. The data recovered reveal illitization of smectite occurring in as little as 28 days, regardless of Na⁺ or K⁺ cations initially in the interlayer. Preliminary results indicate Cs⁺ prolongs the transformation of smectite to illite, forming mixed-layer illite/smectite after 28 days. Understanding the mechanism of illitization will further help to inform the performance assessment in the design of an engineered barrier system.

SNL is managed and operated by NTESS under DOE NNSA contract DE-NA0003525. SAND2023-03806A

In a zero power reactor (ZPR) loaded with salt fuel, the thermal energy released during operation is so small that the fuel remains solid at room temperature with very low burnup and heat rate. Nuclear power plant (NPP) spent fuel is currently reprocessed and recycled at end-of-life to preserve resources and for the reduction of future burden from wastes. Actinides maybe recycled using advanced processes of separation through various routes; PUREX or pyro-processing. The considered zero power salt reactor is in the form of salt fast reactor with high energy neutron flux during operation. Density is one of the critical thermo-physical properties of any reactor. Determining the density of the salt system (NaCl-UCl₄) is very important to evaluate salt fuel-reactivity and behaviour of the core. Actinides and fission products inventories at the end-of-life of reactor are then significant. Presenting the analytical methods of measuring the densities of salt components using multi-scale approaches of X-ray Diffraction (XRD) for nm features, amorphization ratio or any defects, and Scanning Electron Microscopy (SEM) for µm pores in the salt fuel. The emphasis is on a salt mixture with composition of (NaCl-47mol%+UCl₄-53mol%). Densities were measured by changing compositions along with the identification of the complex phase; Na₂UCl₆. Results obtained were in good agreement with the ideal mixed phase (heterogeneous) density model, thereby establishing that XRD and SEM are important techniques to measure the densities of salt fuels. High density fuel in a reactor enhances the reactivity as well as the average neutronic flux. This work provides the salt density measurement which can be used to correct the reactivity of the fuel at end-of-life and for other utilisations. Actinides (Pu, MA) and fission products inventories at the end-of-life are then insignificant in fraction. In these conditions fuel material may be seen as a zero nuclear waste.

In order to determine the durability of nuclear waste glass in a specific environment, different leaching tests have been developed over the years, aiming at investigating different types of leaching mechanisms occurring at different timescales after the first contact with water. The Single-Pass Flow-Through Test (SPFT) method is commonly used to determine the maximum glass dissolution rate, as in this setup the elements released by the glass are carried away from the sample, preventing the saturation of the solution. These rates can be considered as characteristic properties of a particular glass composition and are basic data requested to describe the properties of the reference waste glasses in the expected disposal environment. In many countries, including Belgium, this environment will be conditioned by the presence of alkaline cementitious materials, increasing the pH of the percolating ground water. To assess the chemical durability of the Belgian reference glasses SON68, SM539 and SM513 under hyper-alkaline conditions, maximum glass dissolution rates were determined at 30 °C, using a KOH solution and a synthetic young cementitious water (YCWCa) with a pH of 13.5, corresponding to young ordinary Portland concrete. In both leaching solutions, the highest dissolution rate was determined for the SM539 glass, which contains the highest amount of Al, while similar rates were found for SM513 and SON68 glasses, whose compositions are comparable. For all glasses, the maximum dissolution rates in YCWCa were lower than in KOH due to the presence of Ca, which causes the formation of a slightly protective layer. The dissolution rates in YCWCa were similar to those measured at 30 °C in static tests in which glass was altered in YCWCa in presence of Ordinary Portland Cement (OPC) with a cement to glass ratio of 1.

During the operational lifetime of a geological disposal facility, groundwater of variable composition may ingress and interact with vitrified radioactive waste, leading to leaching of elements. The rate of this dissolution process is understood to be influenced by elements dissolved in the contacting solution; however, since groundwater is a complex mix of many elements, elucidating the mechanism by which these elements influence dissolution is challenging.

The single pass flow through (SPFT) methodology was used to investigate the effect of individual groundwater cations on the forward dissolution rate of simulant UK high level waste glass. Solutions containing chloride salts of lithium, sodium, potassium, magnesium, calcium, and strontium were flowed over the glass sample until steady state conditions were reached. The forward dissolution rates were compared as a function of each cation element. A series of monolithic static dissolution tests were conducted in parallel. The same series of alkali metal and alkaline earth metal chloride salt solutions were used to study the role and behaviour of the added ions in the formation of an alteration layer during the residual rate.

The implementation of an engineered multi-barrier approach for nuclear waste disposal, to mitigate the release of radionuclides over the operational lifetime of a geological disposal facility (GDF), requires a detailed understanding of the interactions between steel canisters, engineered backfill and natural barriers. In particular, the reaction between Fe – present both in waste containers and within the vitrified waste itself – and silica in vitrified wastes is of interest as it has previously been shown that this element may enhance dissolution of glass in aqueous solutions.

Within a geological environment Fe is present trifold; within the environment in Fe-rich minerals, within the matrix of vitrified waste from aqueous HLW feedstocks, and in the cannister material. The interactions between these Fe sources and the wasteform will hold different influences and impact the long-term durability of the wasteform to varying extents.

In this study, we describe the relationship between Fe present within the glass and dissolution behaviour. A simple five oxide borosilicate glass series, with a formulation based on that of the UK high level waste sodium aluminosilicate glass, MW25, was produced to study the effect of Fe content on glass structure and glass durability, with Fe additions ranging from 0 to 5 mol %. The structure of the glasses, as a function of Fe content, was determined using Raman spectroscopy, XRF, and XAS analysis, with decreasing T_g and depolymerisation of the glass network with increasing Fe content. The dissolution of the glasses was determined using SRCA, PCT and MCC durability testing, utilising solution analysis by ICP-OES. Finally, the bioavailability of Fe within the glass network was tested in two simplified subsurface microbial systems to ascertain if glass-microbe interactions can affect the dissolution behaviour of Fe containing glasses.

Understanding the long-term behavior of nuclear waste glasses prior to storage in a near-surface disposal facility is important as this will assist in assuring that the release of radionuclides from the disposal facility will meet regulatory limits. Archeological and natural samples are of prime importance in this mission as they can be used to validate performance assessment models and will assist in public and regulatory acceptance of the disposal site. Here, we discuss five different near-surface sites where archeological and natural samples have been altered by rain and/or groundwater in the environment for hundreds to thousands of years. The selected sites offer a range of characteristics including average temperature, rainfall, and microbial activity. The chemistry of the samples also varies both in silica content, amongst other oxides, and in heterogeneity, in terms of the abundance of amorphous and crystalline fractions. In addition, we used standard laboratory tests, including the product consistency test (PCT), the vapor hydration test (VHT), and the EPA Method 1313 test, to alter archeological samples and we have compared those results with the corrosion of vitrified archeological materials excavated from one of the sites, a ~1500-year old Iron Age Swedish hillfort, Broborg. We compare characterized site samples with corrosion characteristics generated by standard laboratory durability test methods. Results show that the surficial layer of the Broborg samples resulting from VHT displays some similarities to the morphology of the surficial layer formed over longer timescales in the environment.

Glass is used in the UK, as in many other countries, to immobilise the high activity waste liquors resulting from spent fuel reprocessing. Vitrification is also under consideration for some lower activity waste streams. Understanding glass behaviour in subsurface environments is important to support the safety case for disposal of these wastes in a geological disposal facility. As borosilicate glasses have only been manufactured in the last century most experiments to understand glass dissolution rates and mechanisms have typically been conducted at elevated temperatures and increased surface areas in order to obtain measurable results in a short time period. Long-term glass alteration experiments are rare, as are those that consider glass exposed to natural environmental conditions. An experiment was established at the Ballidon limestone quarry, Derbyshire, in 1970 to investigate modern and archaeological glass alteration under mildly alkaline conditions: limestone rich sediment, pH 9.7-8.2. The study has since been extended to include US, UK and Russian nuclear waste glass compositions, samples of which were removed after 16 -18 years of burial. Here, analysis is presented from a variety of nuclear waste type glasses buried at Ballidon including UK 'Mixture Windscale' type glasses, iron phosphate glasses and US Low Activity Waste borosilicate compositions. Even after a relatively short burial time (<20 years) at low temperatures (average 8 ^oC) alteration layers were visible on most glass types. Study of these layers by electron microscopy, EPMA and microfocus X-ray absorption techniques has revealed their chemistry, morphology and interaction with the surrounding sediment. Results give insight into both the corrosion mechanisms of glasses in complex natural environments and the fate of rare earth elements (representing radionuclides) contained within these glasses.

Whilst most laboratory based tests are conducted under static, sterile, closed system conditions, studies of glasses exposed to natural conditions at Ballidon offer insight into glass behaviour in complex open systems with changing geochemistry, influence from nearfield mineralogy and geomicrobiology. Microbial community analysis conducted at the time of site excavation, supported by laboratory based experiments, shows the probable direct or indirect influence of microbiological processes on the corrosion of glasses at the Ballidon site. Similarly, studies of the adjacent sediment and glass alteration layers reveals the transfer of elements to and from the surrounding minerals.

We report the results of massive parallel molecular dynamics simulations of high-energy radiation damage in phosphate glasses. This damage is created by overlapping multiple 70 keV collision cascades. We quantify different aspects of radiation-induced structural changes including at different stages of damage development, including coordination numbers, cluster sizes and density. The overall trend is that radiation damage causes polymerisation of the phosphate network and the loss of small and isolated clusters. However, the details of this response varies with different glass compositions. This polymerisation indicates that the disparate network of strong Fe-O bonds is weakened, which will subsequently weaken the material’s resistance to radiation as phases of phosphate and iron become separated. Furthermore, the degree of recovery in these simulations is far diminished compared to simulations of low energy cascades. This qualitative difference in material response between cascade energies is an important consideration for the deployment of iron phosphate glasses for nuclear waste encapsulation.

Dose rates within vitrified high-level waste (HLW) initially being of the order of 10² Gy/s are high enough to cause concern on the role of radiation effects on long-term retention of radionuclides and performance of glasses. A significant part (~ 20 to 40%) of the deposited energy in glass, which is of the order of about 4×10⁹ Gy for commercial HLW, is caused by the beta radiation of decaying radionuclides. Experiments within electron microscopes have revealed effective flow of silicate glasses under the electron irradiation including direct visualisation of quasi-melting and flow of vitreous material which is characteristic to its molten state. These experiments raise the question of such effects within vitrified HLW although the dose rates of experiments reported were much higher compared with those specific to HLW. An analysis of the nature of radiation induced flow of glasses and quantitative assessments of irradiation parameters causing flow and potential thresholds (which must not ever be reached in nuclear waste immobilisation practices) are evidently needed. The report analyses the nature of flow both for non-irradiated and irradiated glasses accounting for generation of flow defects in form of broken chemical bonds both by thermal fluctuations and absorbed radiation which can be in form of particles and/or photons. The activation energy of flow Q_H is typically high and constant in glasses (Arrhenius type flow) below the glass transition temperature T_g, however it starts to diminish above the T_g, further decreasing finally achieving its low value Q_L characteristic for melts at the crossover temperature T_A = kT_m, where k = 1.1 ± 0.15, and T_m is the melting (liquidus) temperature regardless of the type of silicate glass-forming liquid. Depending on the temperature and dose rate of radiation the major source of flow defects can be either thermal fluctuations or ionising radiation. The radiation breaks chemical bonds generating flow defects (termed configurons) and modifies the temperature dependence of flow by shifting the low activation energy regime (Q_L)and crossover temperature (T_A) to lower temperatures. Moreover, at high dose rates of radiation T_g can abruptly decrease, thus effectively transforming the glass into a liquid. The equation of viscosity of glasses in radiation fields derived reveals the critical parameters of radiation and enables parametrical estimation of threshold values which separate the liquid-like (molten state characterised by Q_L) from solid-like (glassy state characterised by Q_H) behaviours. The report presents numerical estimations for threshold dose rates and show that these were of the order of ~ 2 10⁶ Gy/s and higher reaching up to ~ 4 10⁹ Gy/sin the experiments with effective quasi-melting of silicate glasses under electron beam irradiation, whereas currently synthesised HLW glasses are characterised by several orders of magnitude lower dose rates below 10³ Gy/s.

Numerous spent fuels in Japan are cooling for future reprocessing. The cooling time of spent fuel from shutdown to reprocessing is a key factor affecting the heat generation rate of vitrified waste and the relevant waste management. The advantage for extending cooling time is to decrease in decay heat of fission products to mitigate the thermal constraints of storage facility and final repository of vitrified waste. Contrarily, ²⁴¹Am is generated by ²⁴¹Pu decay in spent fuel during the interim storage and is a concern for the long-term heat source in vitrified waste.

The present study numerically investigated the trade-off relationship between the decay in fission products and the ²⁴¹Am buildup from the waste management perspectives. For spent fuel with a typical burnup of 45 GWd/MTU, the decay heat in vitrified waste decreases by half in about 10-year cooling and by a quarter in 40-year cooling. This will shorten the storage period of vitrified waste before final disposal. Alternatively, the surplus heat capacity in repository system up to the bentonite buffer limit temperature of 100 °C allows higher waste loadings in vitrified waste. Thereby the waste volume (i.e., number of canisters) and the repository footprint (m²/MTU) can be reduced by up to 13%. However, the disadvantage of ²⁴¹Am buildup owing to the extended cooling time was revealed that the surface temperature of vitrified waste at the time of glass dissolution after disposal will exceed the 60 °C predicted in the safety case. The temperature-dependent glass dissolution rate and the subsequent influences on post-closure safety assessment will also be discussed along with results for high-burnup vitrified waste.

This work was carried out as a part of the basic research programs of vitrification technology for waste volume reduction supported by the Ministry of Economy, Trade and Industry, Japan (Grant Number: JPJ010599).

The disposal of spent nuclear fuel (SNF) in deep geological formations in combination with a resilient multi-barrier concept is the preferred option for the safe isolation of highly radioactive wastes in Germany, as well as several other countries (e.g., Belgium, Finland, Sweden, Switzerland). Nevertheless, an intrusion of ground water, intertwined with the failure of canisters and loss of SNF cladding integrity must be considered in the long-term evaluation of a deep geological repository. An assessment of SNF performance in the repository system requires a thorough process understanding of the dissolution rates, the individual radionuclide source terms as well as the alteration processes of the waste form. The dissolution process of SNF can be described in two steps: (i) a fast, initial release of radionuclides, which segregated to accessible structures of the SNF during reactor operation (ii) a slower, long-term release, originating from the dissolution of the fuel matrix itself.

In the present study, we show results obtained from ongoing leaching experiments with medium (46.9 GWd/t_HM) and high burn-up (50.4 GWd/t_HM) UO_X SNF, in highly alkaline, simulated cement water solutions under anoxic and reducing conditions induced by H₂ overpressure. Both SNF were irradiated in commercial nuclear power plants in Germany and Switzerland during the 1970s and 1980s. For the volatile radionuclides, such as the fission gases, ¹²⁹I and ¹³⁷Cs, a rapid, initial release is observed, in comparison to radionuclides assigned to the SNF matrix, e.g., ⁹⁰Sr, ²³⁸U or ²³⁹Pu. However, the initially rapid release of volatile fission products significantly slows down throughout the experiments, unaffected by neither the pH of the leachant solution nor the presence of reducing H₂, albeit a continuous release is observed. In addition, the data obtained in the current leaching experiments are compared to previous experiments conducted at KIT-INE with UO_X and (U,Pu)O_X fuels.

Spent Nuclear UOX fuel (SNF) leaching experiments were conducted in order to investigate the (fast) release of some of the most critical radionuclides with respect to long-term safety. Previously, samples with a burnup of 55 MWd.kg^-1HM have been leached in the bicarbonate solution (pH ≈ 9) used as reference leaching medium in the framework of the “FIRST-Nuclides” European program. These experiments were conducted without hydrogen, in anoxic conditions. More recently, samples of the same fuel were leached in the same type of bicarbonate solution and in a highly alkaline solution (pH 13.5) under reducing conditions imposed by hydrogen, using pressurized autoclaves at 40 bar. The latter experimental conditions are representative of a deep geological repository conditioned by the presence of cementitious materials imposing a high pH. In presence of hydrogen, the uranium concentration remained stable around 10^-7 M, whereas in anoxic conditions the concentrations increased with time. The Tc concentration was initially lower with hydrogen than in anoxic conditions in bicarbonate solution, but increased with time in reducing conditions to reach similar concentrations. A stable Tc concentration was reached with hydrogen only at high pH. The leached fraction of Sr in bicarbonate solution was higher in anoxic conditions than in reducing conditions, and (in reducing conditions) higher in bicarbonate than in the high pH solution.The released fractions of Cs and I were similar in anoxic and reducing conditions in bicarbonate solution and similar in bicarbonate and high pH solution in reducing conditions. The leached fraction of iodine was similar or slightly lower than the total fission gas release including the fission gas release during the leaching in reducing conditions, but this could not be confirmed for anoxic conditions.

This work presents experimental data of the release of Mo, Tc, Rh and Ru metallic particles from high-burnup spent nuclear fuel (63 MWd/kgU) at two different pH values, 8.4 and 13.2. The release of these elements from SF to the solution is around two orders of magnitude higher at pH=13.2 than at pH=8.4. The high Mo and Tc release at high pH would indicate that both elements would not be congruently released with uranium, as it has been pointed out in some release experiments, and would have an important contribution to the IRF, with values around 5%. On the other hand, Ru and Rh release could be explained by oxidation processes favoured at high pH.

The high release of such elements at high pH could be the consequence of the dissolution of the metallic inclusions contained in the fuel through an oxidative dissolution mechanism. Experimental data has been treated by a semi empirical model to evaluate the relative importance of the contribution of different sources on the release of Mo, Tc, Ru and Rh to deduce both the localization in the fuel and the oxidation state of the elements released to the solution as a function of time.

Understanding the leaching behavior of spent nuclear fuel is crucial for the safety assessment of deep geological repositories where spent nuclear fuel will be disposed. Consequently, numerous studies have been carried out on UO₂-based fuels aiming to determine dissolution rates as well as understanding the dissolution mechanisms. However, new types of nuclear fuels containing additives are currently being introduced in commercial reactors to improve reactor performance and reduce fuel cycle costs. Before their use on a larger scale, these fuels must be shown to be acceptable as a waste form for direct disposal in the intended repository environment. These new fuels with additives such as chromia (Cr₂O₃) have an impact on the UO₂ microstructure, e.g., enlarging fuel grain size, which might affect properties relevant to the safety assessment.

The main goal of this investigation is to gather data on the leaching behavior of fuels doped with Cr₂O₃ under relevant repository conditions. A sample consisting of spent fuel fragments is leached inside an autoclave in simplified, synthetic granitic groundwater (10 mM NaCl and 2 mM NaHCO₃) under H₂ overpressure at Studsvik’s Hot Cell Laboratory. The concentration of radionuclides of interest in the aqueous solution is monitored for 1 year as a function of time by sampling and measurement by Inductively Coupled Plasma Mass Spectrometry. In addition, the composition of the gas phase is analyzed by Gas Mass Spectrometry to detect potential air intrusion and monitor the release of fission gas from the fuel. The fuel sample was irradiated in a commercial pressurized reactor (PWR) to a local burnup of 59 MWd/kgU. The leaching data from the Cr₂O₃-doped fuel experiment is presented and compared to commercial standard UO₂ fuel.

Plutonium and minor actinides are generated in nuclear fuel in non-negligible amounts during the normal operation of nuclear reactors. Proper management is mandatory due to the associated long-term radiation hazards. Moreover, reuse and recycling would be beneficial in terms of the circular economy. Different durable plans for safe storage in geological repositories as well as recycling strategies in advanced nuclear reactors are currently under consideration, but also alternative uses are investigated. Research at Joint Research Centre (JRC) in Karlsruhe is performed to support the safety and safety assessment of these applications.

One of the most promising solutions to reduce the amount of plutonium and minor actinides is their use and/or transmutation in dedicated reactors. In particular, the development of safe production processes for minor actinides-bearing fuels is one of the critical tasks for transmutation technology. In this context, a novel synthesis route, the hot compressed water decomposition of oxalate, was used to prepare homogeneous (U,Pu)O₂, (U,Pu,Am)O₂ and (U,Am)O₂ fuel samples. The relative advantages and the drawbacks of the method are discussed in comparison to established methods, and the main scientific findings are summarised.

In search of chemically stable americium compounds with high power densities for space applications, a number of ceramic materials was prepared and characterized in our Minor Actinide (MA) laboratory. Such ceramics are foreseen as power sources for space applications like Radioisotope Thermoelectric Generators (RTGs), and they have to endure extreme conditions including high vacuum, temperatures, and radiation fields. We summarize and compare the results of different americium ceramics synthesised at JRC Karlsruhe, with respect to their americium content, crystallographic stability in terms of swelling and amorphization resulting from self-irradiation due to alpha decay, stability under vacuum and different atmospheres, the behaviour of the ²³⁷Np decay product, presence of a natural analogue, etc.

Part of the Am-ceramic compounds synthesized for space applications are also suitable as waste forms in the management of plutonium and minor actinides. Thus, our systematic studies on compounds of Pu and Am (such as monazite, pyrochlore, or zircon-like) provide useful information on the long-term stability of various candidate ceramic waste forms under internal irradiation.

Nearly all advanced nuclear fission concepts are based on closed fuel cycles, and today the reprocessing of spent nuclear fuel has reached maturity in terms of uranium and plutonium recycling. Research remains necessary, however, to address the problem of the minor actinides (Am, Np, Cm). For Am, partitioning followed by transmutation is since long proposed. At SCK CEN, heterogeneous transmutation of Am in a dedicated Accelerator Driven System is being investigated since several decades. Specific to this concept is that the transmutation uses (U, Am)O₂ targets with elevated Am concentrations. Although the concept is straightforward, the practical problems related to the fabrication of such (U, Am)O₂ targets remain challenging. Recently, advances have been made in the so-called infiltration route in which porous uranium oxide microspheres are loaded with a nitric acid solution of Am followed by calcination and sintering. The present contribution reports recent advances in tailoring the porosity of uranium oxide spheres, their loading with inactive surrogate infiltrant (Nd³⁺), and first results with active infiltrant (Am³⁺).

The study of uranium oxides at different conditions is of paramount importance in the nuclear field, especially regarding characterization of the spent nuclear fuel behavior in dry storage scenarios. This work presents an assessment of the structural evolution occurring during the oxidation of the UO₂ spent fuel matrix into U₃O₇ in air. In particular, we report the results of X-ray diffraction and Raman spectroscopy analyses obtained on a variety of powdered samples prepared in order to cover a specific stoichiometry range in UO_2+x, with x varying from 0 to 0.30. The oxidation degree of each sample is confirmed by thermogravimetric analysis. Over the hyperstoichiometry range UO_2.00-UO_2.20 three structure transitions are detected, giving rise to three distinct regions associated with consecutive structural rearrangements. As for the UO_2.24-UO_2.30 range, the appearance of a tetragonal distortion and its increasing presence with the increase in oxidation degree is observed. These outcomes improve the understanding of non-stoichiometric uranium oxides, what can be used as a basis for further research on the stability of doped UO₂ matrices, such as ATF (Accident Tolerant Fuel) matrices, under in situ conditions, simulating both interim and final disposal.

_{Oxidation of uranium dioxide (UO2) nuclear fuel occurs during accident scenarios and storage conditions. The excess oxygen is incorporated into the fluorite structure and the resulting atomic-scale defect configuration significantly influences important bulk properties such as thermal conductivity and fission gas release. Previous experimental and modelling efforts have proposed distinct oxygen defect cluster configurations; however, most characterization techniques lack sensitivity to the local atomic structure or the oxygen sublattice and the resulting data cannot be used to validate predicted defect clusters. Here, we present results on single-phase UO2+x systems (x = 0.07 and 0.15) combining advanced experimental and modelling techniques to create high fidelity atomistic models of the oxygen defect clusters. In situ high-temperature neutron total scattering measurements with high sensitivity to the oxygen sublattice were performed at the Nanoscale-Ordered Materials Diffractometer (NOMAD) instrument at the Spallation Neutron Source (Oak Ridge National Laboratory). The data acquired at 600 °C and 1000°C were analyzed via Reverse Monte Carlo modelling techniques which consider both the long- and short-range structures. The analysis reveals evolving behavior as a function of oxygen content with simple clusters in the low O:M regime (UO2.07) and more complex, extended defects for higher oxygen concentrations (UO2.15). Our findings have implications in improving and validating potentials for Molecular Dynamics simulations to advance larger fuel performance codes.}

In safety assessments for the deep geological disposal of high-level nuclear waste, the unlikely ingress of water and corrosion of spent nuclear fuel (SNF) need to be considered. For many ceramics, grain boundary dissolution plays an important role, which can continuously increase the reactive surface area and sometimes even leads to a disintegration of the microstructure. Model systems were developed to enable a detailed study of single effects occurring during SNF corrosion. The most important phase in this aspect is the UO₂ matrix. Here we present a detailed investigation of UO₂ based model systems which were doped with Nd₂O₃ or Gd₂O₃.
Polycrystalline UO₂ ceramic pellets were synthesized using a co-precipitation method, as well as Nd₂O₃ and Gd₂O₃-doped samples. Doping levels ranged from 0.5 wt.-% to 4 wt.-% of the respective dopant, representing the fission products of SNF. These were characterized by the determination of physical, chemical, structural, and microstructural properties via density measurements, XRD, SEM and EDX. A focus was put on the comprehensive analysis of grain size, grain shape and pore size.
Pure UO₂ ceramics are characterized by a monomodal grain size distribution. The doping changes the average grain size for Nd₂O₃ from 11 µm (pure UO₂) to 9 µm at 4 wt.-% Nd₂O₃ and from 11 µm to 10 µm at 4 wt.-% Gd₂O₃. At lower dopant concentrations, clusters of small grains with low dopant content tend to accumulate in addition to larger grains with a higher dopant composition. This effect decreases with increasing dopant concentration. As a result, the grain size distribution of the samples becomes bimodal between 1 wt.-% and 2 wt.-%, while it is nearly homogeneous at 0.5 wt.-% and 4 wt.-%. In conclusion, there is no simple linear relationship between doping level and microstructural changes in the studied systems.

Recent research has shown that fission products and actinides present in UO₂-based spent nuclear fuel significantly improve its resistance to oxidation in air and to oxidative dissolution in aqueous media compared to pure UO₂. However, the mechanisms behind these retardation effects are not yet fully understood. In this study, we have developed thermodynamic models for the oxidation of pure UO₂ as well as Ln- and Pu-doped UO₂ solid solutions in air, with reference to measured data on the oxygen partial pressure at equilibrium.

The doped systems are primarily distinguished from the pure UO₂-UO₃ system by an additional degree of freedom that allows for a decrease in the total free energy by redistributing dopants between the MO₂, M₄O₉ and M₃O₈ phases. Our modelling shows that the cubic phases tend to be significantly enriched in Ln or Pu, while the M₃O₈ phase tends to have the smallest fraction of the impurity component considered. The achievement of a large O/M ratio in a doped system at thermodynamic equilibrium requires this fractionation to be developed. Consequently, the equilibrium oxidation of doped UO₂ is necessarily coupled with the transport of Ln or Pu between the constituent oxides. Thus, the slow diffusion of Ln or Pu within any of the relevant phases is proposed to be the cause of the enhanced resistance to oxidation.

The UO₂-ZrO₂ system has been considered as a fuel additive for accident tolerant fuels due to increased resistance to oxidation and corrosion in high temperature and humid environments. In literature, at higher doping levels (>20 at.% Zr) one or more zirconia phases have been found in addition to (U_1-yZr_y)O_2-x phases which may help to stabilize against oxidation. However, it has also been shown that the addition of several percent of Zr, maintaining a single (U_1-yZr_y)O_2-x phase, can accelerate oxidation in air at temperatures ~200° C. Under normal reactor conditions a small fraction of Zr from Zircalloy cladding can migrate into the fuel pellet and into the UO₂ matrix, potentially forming phases within the outer edge of fuel pellets which are more susceptible to oxidizing. Presented in this study are two series of (U_1-yZr_y)O_2-x materials synthesized via a nitrate coprecipitation reaction. A low-doped series containing 0.1-1 at.% Zr in 0.1% increments, and a high-doped series containing 2-20 at.% Zr in 2% increments. The doped ammonium diuranate materials were calcined either in air to form U₃O₈ or in H₂/Ar to reduce to UO₂ prior to pressing pellets and sintering. A portion of the sintered pellets were then powdered and reoxidized to U₃O₈ allowing the comparison of two sets of doped U₃O₈ processed at low (800° C) and at high (1700° C) temperature. The evolution of phases present across the doping range is shown by Rietveld refinements of X-ray diffraction patterns and compared with thermal analysis. Defect signatures are shown by Raman and infrared spectroscopy. Select samples are analyzed using electron microscopy and in-situ Raman mapping during oxidation, previously shown to correlate well with thermal analysis results.

Among all the fission products formed in UO₂-based spent nuclear fuels, molybdenum is one of the most abundant due to its high fission yield. Although its radiotoxicity is low, it has been studied during the last years because of its relevance on the fuel oxidation and other fission products migration. In fact, the oxygen potential of Mo/MoO₂ is very similar to that of the fuel, hence, the excess oxygen created during fission could be neutralized by the oxidation of metallic Mo to Mo(IV), buffering the oxidation of UO₂. Therefore, the distribution of Mo between metallic particles and dissolved in the UO₂ matrix as MoO₂ is of great importance. In this work, electrochemical experiments were performed to study the influence of molybdenum on the oxidation of UO₂.

UO₂ and Mo powders were mixed and compacted at 700 MPa into pellets, which were then sintered at 1740ºC for 4 hours in a reducing atmosphere (5%H₂/95%Ar). Microstructure characterization of the pellets by SEM evidenced the formation of Mo channels throughout the whole UO₂ pellet, whereas no Mo was found inside the UO₂ matrix. The Mo-doped UO₂ pellet was used in electrochemical experiments as a working electrode. Ag/AgCl (3M KCl) and a Pt wire were used as a reference and counter electrodes, respectively. Test solutions were prepared at pH 10 with NaCl 0.1 mol·dm^-3 in the presence of NaHCO₃, Na₂SiO₃ and/or CaCl₂. The corrosion process was studied by performing cyclic voltammetry, potentiostatic experiments and corrosion potential experiments.

Preliminary results indicate that the presence of Mo significantly decrease the reactivity of UO₂, when compared to that of non doped UO₂. XPS analysis will be performed on the electrode after potentiostatic experiments, to determine the surface oxidation state of both U and Mo by the deconvolution of the U4f band and the Mo 3d band.

Particle analysis is one of the key-techniques used in the field of nuclear safeguards. Beyond traditional uranium isotopic ratio measurement, other methodologies are implemented to better characterize nuclear materials. Among them, age dating at the particle scale enables to determine the time elapsed since the last chemical step of separation/purification or enrichment, for example through the ²³⁰Th-²³⁴U radiochronometer. During this work, uranium-thorium mixed oxide microspheres were synthesized as potential reference materials for nuclear safeguards using a wet chemistry route. The hydrothermal conversion of aspartate precursors at T = 433 K led to mixed dioxide micro-particles with controlled spherical morphology and size, up to 5 mol.% in thorium. In order to remove impurities, densify the micro-particles, and control the chemical form of the final compounds, heat treatments were performed under various atmospheres. Nearly stoichiometric (U,Th)O₂ dioxides were obtained under reducing conditions (Ar-4%H₂) while U₃O₈-based samples were formed under air, with thorium incorporated in the structure up to 2 mol.%. Last, the homogeneity of the cation distributions in the samples was evaluated by various methods, including PERALS α-scintillation counting, as well as X-EDS and LG-SIMS analyses of individual particles, leading to consistent results. Particularly, the relative external reproducibility (2σ) of the ²³²Th⁺/²³⁸U⁺ ion ratios measured at the particle scale remained below 10%, paving the way to use these mixed oxide particles in the field of nuclear safeguards.

The safeguards laboratories at Forschungszentrum Jülich provide the International Atomic Energy Agency (IAEA) with well-defined microparticulate uranium oxide reference materials for mass spectrometric verification measurements to support a sustainable and reliable quality control system for particle analysis in nuclear safeguards. For specific applications, such as chronometrical measurements further development of analytical methods including the quality control process for the evaluation of analytical data itself as well as of novel mixed uranium oxide microparticulate reference materials is required. But due to the extremely low production quantity of the microparticles (few µg) the characterization of doped uranium oxide microparticles is very challenging. Therefore, a co-precipitation method was adjusted to produce doped bulk-scale materials as “internal reference materials” which can be investigated by state-of-the-art analytical techniques to unravel the structural incorporation mechanism of relevant dopants, such as lanthanides, Th, and Pu, into uranium oxide. The determination whether a solid solution or segregated phases are formed in dependence on the chemical properties, ionic radii as well as the amount of tri- and tetravalent dopants will provide essential information about the applicability of these mixed compounds as reference materials in nuclear safeguards. Regarding the transferability to the particle production process in Jülich, the phase transformation from UO₃ to U₃O₈ is of particular interest. Therefore, the pristine materials (doped ammonium diuranate) were investigated with TG-DSC to identify the temperature of the phase transformation of UO₃ to U₃O₈ for the doped materials. Subsequently, the materials were calcined at the identified temperatures and structurally characterized with XRD, Raman, and IR spectroscopy. This presentation will provide an insight regarding the incorporation of tri- and tetravalent dopants, such as lanthanides, Th and Pu, into the uranium oxide structures applying ex situ and in situ techniques.

Although spent fuel reprocessing remains the reference scenario in France, direct disposal in deep geological repository is also studied as an option within the framework of the French national plan for the radioactive waste management. Following UO₂ spent fuel that have been intensively studied, the alteration mechanisms of U_1-xPu_xO₂ fuels and especially, Mimas MOX fuels are under investigation to establish long-term evolution models. From the literature, the key mechanism controlling UO₂ dissolution and the associated radionuclides release is an oxidizing dissolution induced by H₂O₂ produced by water radiolysis. Such mechanism can be affected by the MOX microstructure and plutonium content. Moreover, the radionuclides release from MOX spent fuel is affected by the groundwater chemistry. Especially, the presence of cementitious backfill material should create alkaline chemical environment likely to affect the water radiolysis yield and the nature of secondary phases formed at the interface. Furthermore, the use of a non-radioactive surrogate material with comparable properties to MOX fuel has relevant practical advantages. As such, finding a suitable surrogate material allowing multi-parametric studies is a major challenge to improve our knowledge of the MOX fuels alteration in various chemical environments. This work aims to investigate the analogy between U_1-xCe_xO₂ surrogate materials and unirradiated Mimas MOX fuel in the presence of H₂O₂. Once this analogy established, the behavior of Mimas MOX fuel and heterogeneous U_1-xCe_xO₂ surrogate in alkaline solution was compared to evaluate the impact of alpha radiation on the alteration mechanisms.

First, both homogeneous and heterogeneous U_1-xCe_xO₂ dense pellets with x ranging from 0 to 0.25 were prepared through wet and dry chemistry routes, respectively. Surrogate materials were then submitted to dynamic leaching experiments at pH = 7.2 and room temperature. The feeding solution containing 0.20 mmol.L^-1 H₂O₂, was kept under air and renewed every 48 to 72 h to guarantee the H₂O₂ stability during the whole experiment. Normalized alteration rates were determined from uranium concentration measured in the leachates after reaching the steady state. Post-alteration characterizations by Raman spectroscopy, environmental SEM and XPS were achieved. The secondary phase precipitation did not occur at the homogeneous (U,Ce)O₂ materials surface and the dissolution rate was divided by a factor 3 when increasing the Ce molar content from 0.08 to 0.25. However, studtite precipitation was observed all over UO₂ surface, leading to a continuous uranium concentration decreases in the outflow. The same results were obtained with heterogeneous U_0.92Ce_0.08O₂. However, studtite was found to precipitate on UO₂ grains only. This result was consistent with that observed for heterogeneous (U,Pu)O₂ in the same conditions, which confirmed the reliability of cerium as a valuable plutonium analogue.

Then, heterogeneous Mimas MOX fuel and surrogate material with an average composition of U_0.93Pu_0.07O₂ and U_0.92Ce_0.08O₂, respectively were altered over several months under static conditions in alkaline solutions containing 2 mmol.L^-1 of silicates at pH 12, room temperature and under anoxic conditions. The a-activity of the MOX pellet was 1.34 GBq.g_MOX^-1. No matter the presence of alpha-radiation, the uranium concentration in solution reached the same stable value of (5.5 ± 0,6)×10^-6 mmol.L^-1 after 120 days of alteration. Geochemical calculations showed that the uranium concentration measured at equilibrium was compatible with the solubility of the U(VI)-phase, clarkeite (Na(UO₂)O(OH)), or U(IV)-phases coffinite (USiO₄) and UO₂·2H₂O. However, altered surfaces characterizations by SEM and Raman spectroscopy did not reveal the presence of secondary phases. The Raman spectra of the altered MIMAS MOX fuel were characteristic of non-oxidized UO₂ and PuO₂ surfaces. These results rather indicated that oxidative dissolution was inhibited by the presence of silicate ions, as already observed for UO₂ SIMFUEL in the literature.

In safety assessments for disposal of spent nuclear fuel (SNF) in a deep geological repository (DGR), water access, consecutive failure of canisters and loss of integrity of fuel cladding is considered in the long-term. Since most radionuclides produced during irradiation of nuclear fuel in a reactor are trapped within the SNF matrix, it is indispensable to understand the processes leading to the dissolution of the matrix and its dissolution rates to evaluate the performance of SNF in the near-field of such a DGR.

In the present study, irradiated UO₂ and mixed oxide fuel (MOX) specimens were sampled from fuel rod segments with an average burn-up of 50.4 and 38.0 MWd/kg_HM, respectively, and used in the experiments.

Static leaching experiments under anoxic/reducing conditions were performed with cladded pellets and decladded fragments of the sampled SNFs in salt brines, granitic/bentonitic or cementitious groundwaters. The experiments were periodically sampled and solution aliquots and gas phases were analyzed.

As observed in other published studies on SNF dissolution under reducing conditions, a large scatter of the uranium concentration is seen in the initial stages of our experiments. After about one year of leaching, the aqueous concentration of uranium approaches slowly towards the solubility limit of U(IV) independent of the type of studied SNF samples (various irradiated UO₂ and MOX fuels). A similar behaviour is observed for other redox-sensitive actinides. The release rate of uranium decreases significantly in all experiments within first 400 days of leaching. Although the rate is very low, it is > 0 and a continuous release is observed. Comparison of results from SNF experiments under various redox conditions demonstrate that oxidative matrix dissolution is inhibited due to the presence of hydrogen.

In the framework of interim storage of spent MOX fuel assemblies in pools during several decades, it is necessary to take into account an incidental scenario. In the event of cladding failure, corrosion processes can lead to a deterioration/damage of the failed rod and to a radionuclide release into water. In order to study spent MIMAS MOX fuel behaviour in pools, a leaching experiment with chemical and radiological conditions as realistic as possible have been performed. In a MOX fuel rod irradiated up to 47GWj/t, a cladded segment was cut and polished on one side (opened section). It was immersed in a boric acid solution (2g/l), at 27°C, under air and under ⁶⁰Co-source, providing a gamma-irradiation field of about 70Gy/h during the whole leaching time (4 months). Solution has been regularly sampled and analysed in order to measure radionuclides releases and the concentration of hydrogen peroxide (H₂O₂) formed by water radiolysis. The spent fuel segment was regularly taken out from the solution to characterize its surface evolution (by optical microscopy coupled with Raman spectroscopy).

Results showed that uranium releases stabilized after 3 months (total released fraction of about 1.6x10^-5 of the inventory), and the final analysis of the leachate showed an important solid fraction (>0.45µm) containing uranium. Solution analyses, surface observations and Raman spectra acquired on it enabled to follow the massive precipitation of studtite, an uranium peroxide formed through the reaction of UO₂²⁺ with the H₂O₂ generated by the radiolysis of water. The studtite precipitation seems to control the uranium concentrations in solution for these conditions. Uranium is not a good tracer of spent fuel matrix alteration, conversely to ⁹⁰Sr and ¹³⁷Cs which are linearly released in the solution during the leaching. These experimental results are compared to the modelling data obtained from radiolysis and solution chemistry simulations.

The disposal of spent nuclear fuels (SNF) in a deep geological repository (DGR) is regarded as the best practical waste management option in many countries. The long-term safety of a DGR over an assessment time frame of up to one million years necessitates a comprehensive understanding of the corrosion behaviour of SNF once the waste canister is breached and groundwater comes into contact. Although various studies have addressed this topic in the last decades, some of the processes contributing to the (radiolytic) matrix corrosion of SNF in the generally reducing repository environment are not fully understood. Furthermore, only limited efforts were deployed to study the corrosion behaviour of irradiated MOX fuels. To examine the effects of environmental conditions on SNF corrosion, the SF-ALE project (Spent Fuel Autoclave Leaching Experiments) was started as a collaboration between the Belgian Nuclear Research Centre SCK CEN and the Forschungszentrum Jülich GmbH.

Within SF-ALE, irradiated MOX fuel rod segments (burn-up between 29 GWd/t_HM and 52 GWd/t_HM) were leached in bicarbonate water at neutral pH and in synthetic cementitious water at pH 13.5 under reducing atmosphere (4 vol% H₂ in Ar at 40 bars pressure) in order to assess the release of various radionuclides and fission gases over a timeframe of 3.5 years. Following the leaching phase, a post-leaching characterisation of the fuel rod segments was initiated. Scanning electron microscopy analyses revealed that the structure of the MOX fuel matrix was affected differently by the exposure to the varied environmental conditions. Furthermore, secondary phases and alteration products were observed on the fuel surfaces. This contribution introduces initial results of the post-leaching characterisation and their implications to the general understanding of the corrosion behaviour of SNF under repository relevant conditions.

For the safety assessment of a future deep repository for spent nuclear fuel, the rate and mechanism for dissolution of fission products and actinides from the fuel is a key parameter. For the vast majority of the fuel rods, the cladding is expected to be intact at the time of disposal. There is however a small fraction of rods where the cladding has failed and the fuel has been exposed to water and/or air prior to encapsulation. In these rods an alteration of the fuel matrix can be expected. In order to predict the mechanism and dissolution rate of radionuclides from failed fuel, the effects of this matrix alteration need to be investigated. Although the leaching behaviour of spent nuclear fuel under deep repository conditions has been studied extensively, studies of failed fuel under these conditions are relatively scarce. However, at the Studsvik Hot Cell laboratory, aerated leaching tests of spent fuel started in the 1980s and two of these were finalized only recently. This has provided an opportunity to investigate the effects of long-term contact with water under aerated conditions similar to wet interim storage conditions.

As reported previously, when finalizing the leaching studies after 37 years leaching time, the fuel samples were subject to visual inspection, X-ray diffraction investigations and leaching in carbonate solution in order to study the formation and properties of secondary phases (mainly studtite and metashoepite). In the present work we have performed scanning electron microscopy including WDS-analysis as well as Laser Ablation-ICP-MS studies on one of the fuel samples. The results show that the secondary phases formed upon long term exposure to aerated deionized water are depleted of most radionuclides, with exception of Cs, and to some extent Rb and Sr, which seem to be retained in the secondary phase.

During interim storage of fuel assemblies in pools the prospect of a through-wall cladding defect must be taken into consideration. The pool water is subjected to a strong radiation field producing oxidizing species, the main molecular species of which is hydrogen peroxide. When brought into contact with the fuel, hydrogen peroxide generated by water radiolysis leads to the precipitation of studtite (UO₂)(O₂)(H₂O)₂ at the surface of the fuel. This phase having a density lower than that of the fuel can induce a worsening of the defect and thus impact the mechanical strength of the rods. It is therefore important to understand their formation mechanisms at the surface of the spent fuel.

In order to assess this problem, an experimental approach aiming to understand its formation mechanism in the presence of alpha radiolysis of water was conducted. Thus, an ¹⁸O-labeled solution was irradiated with He²⁺ ions through a UOx target, itself leached by the radiolysed solution. An in-situ Raman monitoring of the studtite precipitation kinetics was carried out. The use of ¹⁸O-labeled water was chosen, in order to distinguish the air oxygen dissolved in the solution from that contained in the water.

The absence of the presence of ¹⁶O in the peroxide bond added to the results of the Chemsimul software, reveals that contrary to gamma radiolysis, the oxygen dissolved in the water does not intervene in the mechanism of production of studtite in alpha radiolysis. We deduce that the process of H₂O₂ formation is the result of the primary radiolytic yield during the heterogeneous chemistry step by the OH°+ OH° recombination. This work teaches us that the mechanisms of H₂O₂ production in aerated water, leading to the peroxide bonds of studtites, are dependent on TEL effects.

Experimental research on active materials often goes along with the generation of compositionally complex waste streams for which a suitable route towards safe (interim) storage is lacking. The complexity of the streams invokes the need for tailored solutions for the individual components. Research on possible reuse but mainly final disposal of the spent fuel is an important part of the new nuclear fuel concepts. A prerequisite for any route is that the waste form can be accepted by the national organizations for waste disposal.

At NRG, research on molten salt reactor (MSR) fuels, both fluorides and chlorides, is ongoing. A part of the research is dedicated to waste handling following irradiation experiments in the HFR Petten. The created spent fuel waste will be finally disposed by the national organization for waste disposal (COVRA). To get this new waste accepted, it must be first fully characterized and the fluoride and chloride waste must be transformed to chemically stable and acceptable waste streams. It is foreseen that if new forms of waste are offered for disposal, additional tests related to final disposal are required; the chemical stability of the immobilized waste forms, most notably cemented waste, must be tested by specific leaching experiments to meet the waste acceptance criteria.

A comprehensive literature survey has been done to summarize the possible ways of handling the MSR waste including molten salt specific challenges (such as radiolysis leading to halide gas formation). Based on the review, different routes have been identified, and have been applied to experimental cold and ‘semi-hot’ tests. Two of the tested methods show the most promising results; i) direct defluorination/dechlorination method and ii) vitrification using ironphosphate glass.

The proposed and tested route for waste handling will be finally applied at NRG to irradiated and fully characterized MSR fuels from the NRG MSR irradiation programme.

Separation of volatile radionuclides including Iodine and noble gases from the off-gas streams of a used nuclear fuel reprocessing facility or advanced reactors has been a topic of significant research. The current technology uses energy intensive cryogenic distillation, which is expensive. Another downside of this approach is the accumulation of ozone due to radiolysis of oxygen. Therefore, alternate technologies, and associated materials, are needed for separation of noble gases selectively over other gases including CO2, N2, O2 and Ar. Pacific Northwest National Laboratory is exploring a new class of materials called metal organic frameworks for separation of noble gases selectively at near room temperature. Our laboratory results demonstrate the removal of these gases with high adsorption capacity and selectivity compared to benchmark materials, such as zeolites and activated carbons. The high selectivity towards noble gases over other gases at low concentration indicates the perfect match between the pore size and the kinetic diameter of the gas species. In this talk I will focus on recent results from our laboratory on separation of noble gases at near room temperature using porous metal organic frameworks.

A nuclear waste form is a stable, solid matrix for the immobilization of radioactive and hazardous constituents present in nuclear waste. There are a variety of waste forms currently in use and many more being studied for potential use. Or center is developing new materials as potential waste forms. To achieve this goal we are preparing and testing numerous actinide containing materials. I will present some of our efforts focussing on the crystal growth of uranium and transuranium containing phases via two different crystal growth routes, mild hydrothermal and high temperature solution flux growth and their evaluation as potential waste forms. The mild hydrothermal route works extremely well for crystallizing complex fluoride phases, such as Na₃GaU^IV₆F₃₀, Na₃AlNp^IV₆F₃₀, and Na₃FePu^IV₆F₃₀, while the high temperature flux route works well for crystallizing oxide phases, such as Cs₂Pu^IVSi₆O₁₅ and Na₂Pu^VO₂(BO₃). The synthesis and structures of these phases as well as a series of new chalcogenides will be discussed, along with our appraoch of identifying potential compositions that we can pursue synthetically.

Ion exchange and adsorption methods are effective ways for the removal of Cs⁺ ions from radioactive effluents. In the practical application process, most of adsorbents used as powdered materials show difficulties in term of separation from the contaminated liquid, or pressure challenges through a high flow resistance in fix bed processes. A suitable grain size preparation of adsorbents can overcome these drawbacks by tailoring desired size particle and facilitating column operations with a low flow resistance. However, the particle size of the adsorbents can be critical on the breakthrough exhausted points and the efficiency of the adsorbent in a continuous treatment process.

This presentation will focus on the influence of the grain size of geopolymer based adsorbent on their Cs⁺ adsorption performances both in batch and fixed-bed process. For that purpose, synthesized geopolymer was used as prepared and as binder to support NaY zeolite particle in 20 wt% charged composite material. These adsorbents were prepared with three various grain sizes (50/100/500 µm) to remove Cs⁺ in batch and column operations. The efficiency and adsorption characteristics were investigated through kinetics, adsorption isotherms and breakthrough curves experimental data. We characterized the porosity and microstructure of the adsorbents and compared their adsorption properties in the exchange process. Comparison of batch and column adsorption experiments coupled with modelling for column study is used for a detailed explanation of various process parameters. The results of these experiments show some challenges for bed fixed column utilization by the choice of the grain size and the importance to more accurately optimize the design of column adsorption system to assess the transport of Cs in geopolymer derived system.

A chief characteristic of hollandite is a high tolerance for large cations (Ba²⁺, Cs⁺), which are traditionally problematic to immobilize. Prior work demonstrated a positive correlation between tunnel A-site Cs content, thermodynamic stability, and a corresponding decrease in elemental release. This work investigated the stability relationship among two suites of samples, which varied in their tunnel A-site occupancy of small cations, namely, potassium (K) produced by two different synthetic routes. The K-hollandite samples, K_xMg_x/2Ti_8-x/2O₁₆, (0<𝑥<2), were synthesized by solid-state as well as sol-gel methods. High-temperature oxide melt solution calorimetry was applied to measure their formation enthalpies to identify stability trends. It was found that phase stability tracks with A-site cation content (K⁺), which correlates well with our previous studies on Cs-hollandite. Thermochemistry, structural features, and electrical conductivity measurements will be discussed in light of density functional theory models and current structure-property relations for these materials systems. Tunnel-structured materials are of interest for a wide range of applications from nuclear waste immobilization to electrochemical energy storage.

Significant success has been observed in loading hollandite-structured phases with the fission product elements cesium and barium. While some understanding of the limits to the content and stability of the phases has been obtained via experimental scoping studies, a detailed understanding that would allow efficient design of these waste forms is still lacking. That issue is addressed in this effort where these systems have been extensively modeled, and where those models are being extended to the simultaneous incorporation of actinide elements. First principles calculations were thus performed on actinide-bearing aluminum-substituted hollandite phases to examine the potential use of the structures for also effectively immobilize U, Np, and Pu. The DFT-calculated formation enthalpies suggest the relative stabilities of these structures, providing likely targets for synthesis studies. These are used together with CALPHAD modeling of the phases using the compound energy formalism to determine their ultimate phase stability. The results provide an emerging picture of solid solubilities and potential ability to design highly loaded waste forms.

The Center for Hierarchical Waste Form Materials (CHWM) is composed of different international teams and aims to develop hierarchical materials for an efficient immobilization of radioactive elements. In this frame, aluminosilicated-based materials have been considered for the selective trapping and conditioning of Cs.

First, geopolymers, which are alkali-activated materials composed of tetrahedra of aluminate and silicate obtained at ambient temperature and pressure, were studied as adsorbent for Cs. Their ability to entrap Cs by ionic exchange is strongly depending on their Si/Al ratio. Indeed, an adapted Si/Al ratio is needed to create mesoporosity and allow the access of the geopolymer grain center for a high adsorption capacity with a fast kinetic. However, their selectivity for Cs is very limited because geopolymers are amorphous. Moreover, their leaching resistance is not as good as those of crystalline materials such as zeolites.

The material synthesis has been modified by curing the same precursor solutions hydrothermally. This leads to the formation of crystalline zeolitic structures instead of amorphous geopolymers. Depending on the Si/Al ratio and the curing time, different zeolite phases can be obtained (Faujasite, NaP1, Analcime…), which impact the Cs adsorption properties. Indeed, the crystallographic parameters of the zeolite have to present an adapted cage size to selectively host Cs by ionic exchange. While the formation of NaP1 does not significantly modify the Cs trapping properties, the synthesis of Analcime instead of geopolymer strongly reduces the Cs adsorption properties because the size of the hydrated Cs⁺ ion is larger than the micropore channels of the zeolitic structure. However, it has been shown that, for a specific Si/Al ratio, a mixed of NaP1-ANA is obtained with larger micropores (or specific defects) particularly adapted for the Cs adsorption. Therefore, this material presents a high capacity as well as an important selectivity for Cs toward Na.