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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.