Symposium ES6 : Scientific Basis for Nuclear Waste Management

Synroc, mineral-titanate based ceramics for immobilisation of high-level radioactive wastes were invented by Ringwood in the late 1970s. A demonstration plant utilising inactive materials was set up at ANSTO (then the AAEC) was soon set up, with the process route based on sol-gel precursors, drying + calcination in a rotary calciner, followed by graphite-die hot-pressing at ~1200oC. The materials had ~100-1000 times more aqueous durability than borosilicate glass in short-term laboratory tests, although borosilicate glass had been actively demonstrated at full scale at that stage. However in the late 1990s, a synroc derivative was chosen ahead of glass for immobilising impure US/Russian Pu. Although this was not significantly pursued, synroc gained considerable credibility. Since the early 2000s, synroc has morphed from a titanate-based wasteform for reprocessing waste to a hot isostatic pressing platform for high- and intermediate level wastes which are problematic for glass in terms of waste loading and processing temperatures. While at ANSTO considerable basic as well as commercial research is being undertaken on a wide range of ceramics, glass-ceramics and more latterly glasses, ANSTO is currently in the detailed design stages for dealing with waste from its own 99Mo radiopharmaceutical production.

A possible method for the reduction of civil Pu stockpiles is the reuse of Pu in mixed oxide fuel (MOX). During MOX fuel production, residues unsuitable for further recycle will be produced. Due to their high actinide content MOX residues require immobilization within a robust host matrix. Although it is possible to immobilize actinides in vitreous wasteforms; ceramics, such as brannerite (UTi₂O₆), are attractive due to their high waste loading capacity and relative insolubility. A range of uranium brannerites, formulated M_x(U_0.9Ce_0.1)_1-xTi₂O₆ (M = Gd³⁺, Ca²⁺), were prepared using a mixed oxide route. Charge compensation of divalent and trivalent cations was expected to occur via the oxidation of U⁴⁺ to higher valence states (U⁵⁺ or U⁶⁺). Ce⁴⁺ was added as an analogue for the Pu⁴⁺ fraction in mixed oxide fuel. Gd³⁺ was added to act as a neutron absorber in the final Pu bearing wasteform. X-ray powder diffraction of synthesized specimens found that phase distribution was strongly affected by processing atmosphere (air, Ar or H₂/N₂). In all cases prototypical brannerite was formed accompanied by different secondary phases dependent on processing atmosphere. Microstructural analysis (SEM) of the sintered samples confirmed the results of the X-ray powder diffraction. Analysis of Ce L-III edge and Ti K edge XANES found that irrespective of processing conditions, Ce⁴⁺ had been reduced to Ce³⁺ and Ti was present in the tetravalent oxidation state. Analysis of U L-III edge XANES confirmed that charge compensation was achieved by oxidation of U⁴⁺. The preliminary results presented here indicate that brannerite is a promising host matrix for mixed oxide fuel residues.

The UK has over 100 tonnes of separated PuO₂ stored at the Sellafield site. The UK’s policy for dealing with this plutonium material is to fabricate all usable material into MOx fuel.¹ Unfortunately, due to contamination by elements including; Cl, Fe, Cr and Am, not all of the material is suitable for reuse as fuel and has been classified as higher activity waste. These Pu-residues require immobilisation into stable wasteforms for long-term storage and eventual geological disposal. One proposed treatment plan for these wastes is to process them into glass-ceramic of full ceramic wasteforms by hot isostatically pressing the waste and precursor materials inside stainless steel canisters. Glass-ceramic materials are proposed for the low purity streams of these highly variable wastes, whereby the glass phase provides wasteform flexibility to accommodate impurities and variations in the waste feed composition. The plutonium partitions into the more durable ceramic phase, zirconolite (CaZrTi₂O₇). Zirconolite has excellent wasteform properties including durability and radiation tolerance, and readily accepts actinides and rare earths into its crystal structure.

In this work, the formation of zirconolite is shown to vary as a function of the glass fraction and composition, such that an Al rich glass promotes a higher yield of zirconolite.² The thermodynamic activity of Si in the system drives the crystalline phase assemblage, by determining whether it is consumed in the amorphous glass phase or unwanted crystalline phases sphene (CaTiSiO₅) and zircon (ZrSiO₄). After defining an optimised formulation that minimises the presence of unwanted phases, cerium was utilised as an actinide surrogate in waste incorporation experiments. The digestion of CeO₂ and Ce partitioning into the ceramic phase is studied by SEM, EDX and XRD whilst the oxidation state of the Ce is identified from Ce L3 edge XANES data. All samples were processed by hot isostatic pressing at 1250 °C, 103 MPa (15,000 psi) for 4 hr in 30 ml stainless steel canisters.

References
¹ Nuclear Decommissioning Authority (NDA), Progress on approaches to the management of separated plutonium - Position paper - v1.0. Nuclear Decomissioning Authority, 2014.
² E. Maddrell, S. Thornber, and N. Hyatt, “The influence of glass composition on crystalline phase stability in glass-ceramic wasteforms,” J. Nucl. Mater., (2014).

Microporous stannosilicates consisting of heteropolyhedral structure where the simplest units are SnO₆ octahedra and SiO₄ tetrahedra have been raising considerable interest.^{1, 2} Sn-umbite (K₂SnSi₃O_9^●H₂O) crystallising in orthorhombic system with the space group P2₁2₁2₁ (a = 10.101, b = 13.136, c = 7.157 Å) has shown its ion exchange to both Cs and Sr.³ It is investigated that the ion exchange can be significantly improved by modifying the crystallographic and electrostatic environment via framework doping. Partial (25%) incorporation of pentavalent elements for Sn⁴⁺ on the octahedral site was achieved via hydrothermal synthesis. The substitutions were confirmed by XRD and XRF and unit cell parameters calculated. The structural incorporations lead to a slight change of the unit cell volume, suggesting an isomorphous substitution can be achieved. As compared to Sn-umbite, the substituted Sn-umbites show remarkable increases in both Cs and Sr capacity. An increase in ion exchange properties can be explained in terms of their inherent tunnel sizes to accommodate counterions due to partial substitution and bond strengths associated with the charge-neutralising cations and framework oxygens.^{4, 5} The Cs- and Sr-exchanged umbites were thermally converted by hot isostatic pressing for evaluation of ceramic wasteforms for Cs and Sr immobilisation.

References

1. Lin, Z.; Rocha, J.; Valente, A., Chemical Communications 1999, 2489-2490.
2. Peixoto, M. A. R.; Ferdov, S., Journal of Porous Materials 2013, 20, 1171-1178.
3. Pertierra, P.; Salvado, M. A.; Garcia-Granda, S.; Khainakov, S. A.; Garcia, J. R., Thermochimica Acta 2004, 423, 113-119.
4. Cherry, B. R.; Nyman, M.; Alam, T. M., Journal of Solid State Chemistry 2004, 177, 2079-2093.
5. Chitra, S.; Sudha, R.; Kalavathi, S.; Mani, A.; Rao, S.; Sinha, P., Journal of Radioanalytical and Nuclear Chemistry 2012, 1-7.

Layered materials, like titanate and vanadate nanostructures, manganese oxides or metal sulphides, were recently found to be efficient in the selective capture of cesium from multi-component aqueous solutions. Even though the retention mechanism was not fully clarified, there were strong indications that the retention performance of such materials is mainly related to their lamellar structure.
In the present study, various vanadate nanostructures with different morphologies (nanotubes and nanosheets) were elaborated and characterized so as to determine their structural and textural properties. Structural characterizations were performed using such different techniques as XRD, TEM, SEM, XPS, elemental analysis, ICP. Special attention was paid to improve the wettability and dispersion ability of nanoparticles in aqueous media. Cesium and strontium adsorption onto such materials was investigated in ultrapure water multi-component aqueous solutions. The individual adsorption isotherms were measured with the aid of HPLC by varying the solution composition in order to quantify the rate and extent of the underlying sorption phenomena. In particular, vanadate nanosheets were found to present a very high selectivity toward cesium irrespective of the aqueous medium used.
The computer-assisted structure determination was combined with Monte Carlo simulations to elucidate the adsorption mechanism depending on the nature of the compensating cations.
The selectivity and reversibility of cesium sorption will be considered as the main criteria for the selection of materials for further shaping and sizing.

Environmental contamination with radionuclides has resulted from accidental releases of radionuclides from nuclear facilities. Approximately 28 million m³ of soil, for example, was exposed and contaminated with radionuclides from the Fukushima Daiichi nuclear disaster. Decommissioning of nuclear facilities also produces a large amount of contaminated soil waste. Unfortunately, the very strong and specific adsorption of cesium into clay interlayers hampers the remediation of soils with general treatment techniques. Although marked progress has been made in understanding the mechanism of the sorption of cesium by clay since the Fukushima accident, the desorption of cesium from clay has not been successfully investigated. For example, attempts at cesium desorption using cation exchange agents including potassium, ammonium, magnesium and hydrogen ions have yielded the removal of only a very small amount of cesium owing to the stong adsorption of cesium in clay. For this reason, an efficient cesium desorption technique must be developed for the treatment of cesium-contaminated soil waste.
In this presentation, we report the desorption behavior of cesium from clay minerals (i.e. montmorillonite, vermiculite, and illite) by various ion exchange agents and a decontamination process for cesium-contaminated clay. We hypothesized that polycations having a high charge density will enhance the ion exchange with cesium ions owing to the extremely high local concentration of cations resulting on their adsorption in clay. We demonstrated significantly improved cesium desorption using a polymer-type cation exchanger, and the cesium desorption behavior by polycation treatment was compared with the results using single cations and cationic surfactants under various reaction conditions. For example, cationic polyethyleneimine successfully removed most cesium ions from montmorillonite (~97%) and vermiculite (~91%) under acidic reaction conditions, even though a limited amount of cesium (~60%) could be desorbed from illite by the polyethyleneimine treatment. Nevertheless, polyethyleneimine desorbed a significantly larger amount of cesium from illite than did strong acids and surfactants. After the cesium desorption step, the polymeric cation-exchange agent was readily separated from the aqueous waste containing desorbed cesium ions by an ultrafiltration membrane, and the cesium ions could then be concentrated by cesium-adsorbents for a reduction of the waste volume.

Layered materials, like titanates and hydroxyapatites, manganese oxides or metal sulphides, were recently found to present a high selectivity for the capture of strontium from multi-component aqueous solutions. The lamellar structure of such materials is considered to strongly influence the retention performance.
In the present study, nanoflower-like manganate nanostructures were synthesized and their structural and textural properties were evaluated to understand the retention behavior of these innovative solids. Structural characterization was based on the use of such different techniques as XRD, TEM, SEM, XPS, elemental analysis, ICP. In view of the application envisaged, cesium and strontium adsorption isotherms were measured both from ultrapure water and multi-component aqueous solutions by means of HPLC. The impact of the solution composition on the individual adsorption of cesium and strontium was quantified. High selectivity performance of such materials towards strontium was clearly demonstrated in comparison with previously reported structures.
Molecular simulations were performed to rationalize the retention process. The selectivity and reversibility of strontium sorption will be considered as the main criteria for the selection of materials for further shaping and sizing.

Recently, some crystalline hybrid porous solids known as Metal Organic Frameworks (MOFs) emerged as promising systems for gas adsorption and drug encapsulation.¹ Indeed these hybrid solids, constituted by inorganic nodes (metal chains or clusters) linked each other by organic linkers, present a large specific surface area and pore volumes as well as a high chemical versatility allowing to modulate the chemical and physical properties indefinitely.
Even if few results are available in the literature, the adsorption by such solids possessing extra-framework ions could be evidenced as a plausible, economic and simple solution to eliminate toxic contaminants in water.² Various parameters can influence the ability for sorption of porous solids such as the charge of the framework, the adsorption interaction and the diffusion in the pores.³
For this study, we have followed a strategy combining experimental techniques (adsorption calorimetry and isotherm, X-ray Diffraction) with computational approach (Molecular Dynamics) to elucidate both the microscopic mechanisms in parallel of the macroscopic behaviors allowing us to rationalize the ion exchange process for both anion (I^-) and cation (Sr²⁺) in MOFs containing extra-framework ions. Further, one anionic and one cationic MOFs have been chosen for this study, in which the impact of the topology and the framework charge can be discussed on the ionic exchange properties, i.e. the adsorbed ion quantity as well as the adsorption energy.

[1] G. Férey, Chem. Soc. Rev., 2008, 37, 191
[2] (a) A. Sachse, A. Merceille, Y. Barrès, A. Grandjean, F. Fajula, A. Galarneau, Micro Meso Mater., 2012, 164, 251, (b) C. Delchet, A. Tokarev, X. Dumail et al., RSC Adv., 2012, 2, 5707, (c) S. Sen Gupta, K.G. Bhattacharyya, Phys. Chem. Chem. Phys., 2012, 14, 6698

The separation of Krypton (Kr) from Xenon (Xe) is an industrially relevant problem. Kr and Xe are widely used in fluorescent light bulbs. High-purity Xe, has been used in commercial lighting, medical imaging, anaesthesia and neuroprotection. During the reprocessing of used nuclear fuel, two of the gases of concern is radioactive 133Xe and 85Kr. By the time fuel is processed, Xe would decay down to stable isotope however Kr has long half-life as a result can not be released into atmosphere freely. Effectively separating Kr from Xe in nuclear reprocessing plants, would lead to a considerable reduction in storage costs, and in potential revenue generated from the sale of pure Xe. The conventional method to separate these two gases is fractional distillation at cryogenic temperatures, which is an energy intensive process. Furthermore, even after cryogenic distillation, trace levels of radioactive Kr in the Xe-rich phase are too high to permit further use. Alternative environmental friendly separation technologies therefore could save energy. In this respect, membrane technology could play a key role in making this separation less energy intensive and therefore economically feasible. Membrane separation processes have several advantages over conventional fractional distillation; for instance, it is a viable energy-saving method, since it does not involve any phase transformation, furthermore, the required membrane process equipment is simple, easy to operate, control and scale-up. In particular, if prepared in membrane form, metal organic frameworks combine highly desirable properties, such as uniform micropores, high surface areas, and exceptional thermal and chemical stability, making them ideal candidates for challenging molecular gas separations, such as Kr/Xe separation.

Herein, we demonstrate the feasibility of preparing continuous and reproducible ZIF-8 (a type of MOF) membranes for Kr/Xe separation. It has been demonstrated that the effective aperture size of ZIF-8 is in the range of 0.40 nm to 0.42 nm, which makes this particular MOF composition ideal candidate for molecular sieve Kr over Xe. In the ideal case scenario, Kr molecules would diffuse rapidly through the pores, while Xe at most will diffuse slowly meaning that high Kr selectivities could be potentially achieved based on molecular diffusion differences. The synthesized ZIF-8 membranes displayed separation selectivities for Kr/Xe gas mixtures as high as 16.2, and Kr permeances as high as 37.9 GPU at transmembrane pressures of 138 KPa. To our best knowledge this work represents the first example of any MOF membrane composition displaying effective separation for Kr/Xe gas mixtures.

The projected UK plutonium contaminated material (PCM) waste volume is >30000 m³ with 70% arising at Sellafield. The current baseline treatment is supercompaction with cement encapsulation. Thermal treatment, i.e. in-container or plasma vitrification has been identified as the main alternative waste treatment method. Key drivers for the application of thermal treatment processes include the reduced volume, improved passive safety, and superior long term stability of the vitrified wasteform products. These advantages have led to a renewed interest in thermally treating various UK ILW streams, including PCM waste. To support the increased investment in thermal treatment technologies, a fundamental understanding of the processes and the impact of waste inventory needs to be established. The research aims to provide the evidence necessary to support a major investment in the thermal treatment of plutonium contaminated materials.
Laboratory scale experiments using PCM waste simulants (using Ce as a Pu surrogate) and glass forming additives have been performed in order to understand the reactions and processes of waste digestion and incorporation during thermal treatment.

Characterisation of the vitrified product phase assemblage have been performed using techniques that include XRD, SEM/EDX and EXANES. Mossbauer spectroscopy was used to investigate the REDOX conditions of the melt. It was found that PuO₂ (CeO₂ surrogate) from the PCM is physically and chemically immobilised in the resulting materials, i.e no residual PuO₂ (CeO₂) remains after processing. All of the analysis indicated that Ce was incorporated into the vitreous phase in all samples. Estimated volume reductions of ca. 80–95% were demonstrated, against a baseline of un-compacted 200 L PCM waste drums.
The research also aims to gain a greater understanding of the vitrified product stability with respect to generic ILW disposal concepts, through accelerated dissolution experiments. The most likely disposal option is for the resulting vitrified ILW product to be placed in a geological disposal facility in a high – pH environment with cemented ILW and a cement-based backfill. Therefore, the potential effects of such a high pH (12.5), calcium rich cement-based environment on the dissolution behaviour of simulant ILW glasses have been studied using a modified version of the product consistency test (PCT).
The research will contribute to accelerating the acquisition of knowledge and experience required to support the NDA in deploying thermal technologies as a national asset for ILW treatment.

The US Department of Energy is evaluating the “modified-open” nuclear fuel cycle to increase the efficiency of nuclear power production and reduce the amount of high level waste (HLW). In the nuclear fuel cycle, part of the fission products generated during burn-up in a nuclear reactor are non-fissionable, and once separated from the fissionable material must be immobilized in stable waste forms. The majority of these products are in the following three waste streams generated by the projected transuranic extraction (TRUEX^plus) process: alkali/alkaline-earths (¹³⁷Cs and ⁹⁰Sr), lanthanides (Ln), and transition metals. A glass-ceramic, with targeted crystalline phase assemblage comprising: powellite AEMoO₄, oxyapatite, (A,AE)_xLn_(10-x)Si₆O₂₆ (where A is alkali and AE is alkaline-earth) and lanthanide borosilicate (e.g., Ln₅BSi₂O₁₃), is being developed to immobilize these non-fissionable products. The proposed multi-phase borosilicate glass-ceramic waste forms are expected to exhibit significantly higher chemical durability in comparison to the reference borosilicate glass along with higher waste loading (~50%) and higher thermal stability.
The major hurdle in the development of this glass-ceramic is the high MoO₃ (~14 mass%) and alkali (Rb₂O, Cs₂O ~12 mass%) content of the waste stream. The presence of these species can lead to liquid-liquid phase separation and the uncontrolled crystallization of alkali/alkaline-earth molybdates. Similar challenges are also being faced during vitrification of French and UK HLW.
A key barrier to maturation and exploitation of glass-ceramic technology is the gap in our fundamental understanding of the mechanisms of phase separation and crystallization which lead to the development of the desired phase assemblage and microstructure determining long-term product performance. The challenge is in predictably achieving the targeted phase assemblage and microstructure, requiring a detailed understanding of the transformation process as a function of both cooling rate and melt chemistry.
Accordingly, the present study aims at understanding the fundamental science controlling the solubility of molybdenum in nuclear waste glasses. The compositional dependence of MoO₃ solubility in four–to–six components simplified nuclear waste glass compositions in the system Na₂O-CaO-B₂O₃-Al₂O₃-Nd₂O₃-SiO₂ has been studied. The solubility limit of molybdenum in these glasses has been determined by inductively coupled plasma – optical emission spectroscopy. The molecular structure of glasses has been studied by various spectroscopic techniques, while phase separation and crystalline phase evolution in glasses and glass-ceramics has been followed by electron microscopy, and X-ray diffraction, respectively. The obtained results pertaining to solubility of molybdenum in glasses along with the discussion about structural mechanisms controlling the same will form the gist of the presentation.

Glassy materials for HLW immobilization were produced from HLW surrogate, quartz sand, datolite (CaBSiO₄OH), and bentonite clay at a temperature of up to 1200 °C. Waste loading ranged between 20 and 40 wt.%. The glasses were characterized by X-ray diffraction, scanning electron microscopy and Fourier-Transform infrared spectroscopy. Glasses with waste loading of up to 35 wt.% obtained by melt pouring onto a metal plate were found to be rather homogeneous but contained minor noble metal oxides and britholite (at high waste loadings) while those annealed in turned-off furnace were partly devitrified. Average chemical composition of britholite corresponded to formula Na_1.00Ca_4.02Y_0.33Ce_0.05Nd_3.64Gd_0.17Si_6.79O_24.39. The glass network is built from SiO₄ units with one or two bridging oxygens and complex borate groups with primarily ternary coordinated boron. Increase of waste loading resulted in shift of band’s maxima to lower wavenumbers exhibiting increasing the fraction of SiO₄ unit with lower number of bridging oxygen ions and thus reduction of glass network connectedness. Glasses with up to 30 wt.% waste loading kept their high hydrolytic durability making them suitable for HLW immobilization.

Spent Fuel Management at Sellafield includes reprocessing of spent nuclear fuel from stations across the UK and also from overseas. At Sellafield, methods have been developed for the processing of high level wastes, including highly active liquor (HAL), which is a by-product of reprocessing irradiated nuclear fuel
This Highly Active (HA) raffinate is concentrated in evaporators in the Highly Active Liquor Evaporation & Storage (HALES) facility before feeding to the Waste Vitrification Plant (WVP). Here, the resultant HAL feed is calcined and combined with glass before pouring into containers to produce an immobilised HA wasteform
Ruthenium is a fission product possessed of two relatively long lived isotopes: Ru-103 (t_1/2 = 39.8 days) and Ru-106 (t_1/2 = 1 year). Both isotopes form part of the inventory of HA waste during reprocessing of spent fuel. Volatilisation of fission products in nuclear waste generally occurs at high temperature – apart from ruthenium where volatilisation occurs at the lower temperature stages of the vitrification process
Given its volatile nature and high specific radioactivity, ruthenium presents a strong challenge to the nuclear industry in effectively managing its abatement. Part of the challenge is to fully understand the highly complex solution chemistry under conditions relevant to HA waste streams and associated abatement systems
Experimental work within the National Nuclear Laboratory (NNL), UK has demonstrated the presence of oxidising metal ions in HA waste (e.g. Ce(IV)) enhancing volatility of ruthenium through a chemical conversion of Ru(III) species to what is assumed to be RuO₄. A better understanding of these species, their electrochemical processes and reaction kinetics is required to underpin the empirical evidence gathered to date, in particular to develop gravimetric, electrochemical and spectroscopic analytical methods that will improve the understanding of ruthenium speciation in high nitric acid environments, establish the kinetics of inter-conversion between ruthenium species and establish the mechanism by which metal ions such as Ce(IV) may oxidise ruthenium.
We have studied the electrochemical behaviour of ruthenium and present here the thermodynamics of complexed and uncomplexed ruthenium. Electrochemical and spectroscopic methods have been used to determine as bought RuCl₃ to be a mixture of Ru(III) and Ru(IV). Subsequently, a method to electroreduce the mixture to a pure Ru(III) solution was developed. Complexed RuNO³⁺solutions show no sign of any Ru(IV) present, indicating NO stabilises against Ru(III) to (IV) oxidation. Once Ru(IV) has formed, tetroxide generation occurs, in both complexed and uncomplexed systems at 1.2V vs. Ag/AgCl. These results suggest the Ru(III) to (IV) transition is the key precursor process for volatilisation, implying nitrate complexation plays no role in promoting volatilisation and volatility of ruthenium is an intrinsic ruthenium problem coupled with nitric acid chemistry.
Sukhraaj K. Johal^a, Colin Boxall^a*, Colin Gregson^b,Carl J. Steele^c
aThe Lloyd’s Register Foundation Centre for Nuclear Engineering, Engineering Department, Lancaster University, Bailrigg, Lancashire, LA1 4YR, U.K.
^bNational Nuclear Laboratory, Sellafield, Seascale, Cumbria, CA20 1PG, U.K.
^cSellafield Ltd., Sellafield, Seascale, Cumbria, CA20 1PG, U.K.

The dissolution rate of spent nuclear fuel depends on intrinsic factors such as fuel structure and burn-up, as well as environmental factors, including groundwater composition. The burn-up of future spent fuel to be disposed of is expected to increase, causing actinide accumulation in the rim zone and an increase of the content of lanthanides and other fission products. The high burn-up structure at the fuel rim is characterised by much smaller fuel grains and a large number of submicron fission gas bubbles, which both increase the surface area. The increased actinide content in spent fuel at higher burn-ups leads to a higher a-dose rate in the surrounding water and the higher content of fission products will also contribute to a higher b- and g-dose rate initially. The higher dose rates together with the increased surface are expected to increase the dissolution rate. All available experimental results show that the presence of fission products like lanthanides and other dopants in the UO₂ matrix has an inhibiting effect on UO₂ dissolution. The increase of non-uranium cation concentration at high burn-up seems to counteract effectively the influence of higher surface area and higher dose rates.
The anoxic corrosion of massive iron containers considered in most deep disposal concepts produces large amounts of dissolved hydrogen in the groundwater. At the relatively low repository temperatures, hydrogen is expected to be inert in bulk solution. During the last decade, a large impact of dissolved hydrogen on the dissolution of the LWR or MOX fuel and UO₂(s) doped with ²³³U or ²³⁸Pu has been observed. For hydrogen concentrations above a certain limit, the dissolution rates of these highly radioactive materials drop to very low values. A discussion of the results obtained with spent fuel or α-doped UO₂ in the presence of a range of hydrogen concentrations is presented. Typical for all measurements under such conditions are the very low long term concentrations of uranium and other redox-sensitive radionuclides, such as Tc and the minor actinides. Measurements of the radiolytic oxygen after long leaching periods result in values below detection limit. The investigation of the surface of spent fuel or UO₂(s) pellets doped with ²³³U by XPS after long periods of testing shows absence of oxidation. The kinetics of the release of non-redox sensitive elements such as Sr and Cs, used to estimate fuel matrix dissolution rates, is also discussed. Potential mechanisms responsible for the observed behaviour are based mainly on data from studies on the interaction of water adsorbed on the surfaces of metal oxides or actinide oxides with radiation. A discussion of the relative importance for a repository of the oxidative versus the non-oxidative dissolution of the fuel matrix will also be given.

Recently, cement has been considered as a possible material present in the Deep Geological Disposal (DGD) (ENRESA, 2014), for instance as a sealing material. Therefore, it is considered necessary to determine the effect of cementitius water in the Spent Fuel (SF) near field. With this objective, a series of electrochemical experiments were performed to ascertain the influence of two important components of cementitious water: calcium and silicate. The electrode was prepared by using 3% at. SIMFUEL, as a chemical analogue of SF with 41 GWd/T_U of burn-up.
Test solutions were prepared at pH 12 with NaCl 0.1 mol.dm^-3, and various concentrations of both Na₂SiO₃ and/or CaCl₂. The corrosion process was studied by performing cyclo-voltammograms from -1200 mV to 400 mV at 10 mV.s^-1, potentiostatic experiments at 200 mV for 1 hour and, finally, corrosion potential experiments for 24 hours. After performing both potentiostatic and potential corrosion experiments, the SIMFUEL suface was analysed by means of X-Ray Photoelecton Spectroscopy (XPS).
The results showed that the experimentally used silicate concentrations had no significant effect on the cyclic-voltammograms obtained, although its presence decreased the SIMFUEL oxidation at potential values above -100 mV. When calcium was added to the dissolution, the whole oxidation process was shifted to higher potentials. The XPS results obtained after performing the potentiostatic experiments at 200 mV showed that in the absence of both silicate and calcium, the surface was highly oxidized, with 75% of uranium as U(VI). When silicate was added to the electrolyte, the XPS spectrum showed a decrease of the U(VI) amount on the surface and both U(VI) and U(V) were approximately 38%. After calcium was added to the electrolyte solution, the predominant component on the surface was identified as U(V). Finally, after the corrosion potential experiments, the electrode surface presented a similar composition with 45% of U(V) as the main oxidized state, either with or without silicate in solution. However, when calcium was added to the electrolyte, the SIMFUEL surface showed that the predominant oxidized state was U(IV).

References:
ENRESA, 2014, 7th Plan Nacional de I+D. 2014-2018.

The oxidation of uranium dioxide (UO₂) has been widely studied due to the potential risks that this process may cause in the event of shielding failure during the storage of such a nuclear fuel. In case of failure under dry interim storage conditions, the UO₂ matrix of the spent nuclear fuel (SNF) might be oxidized owing to its contact with the atmospheric oxygen and the high temperatures present due to the decay heat of the SNF. The transformation of UO₂ into U₃O₈ via the two-step reaction UO₂→U₄O₉/U₃O₇→U₃O₈ entails an increase in volume of around 36% and, consequently, it might cause the loss of integrity of the UO₂ matrix. Since this fuel matrix is responsible for retaining the fission products and transuranium elements formed by the irradiation process, release of radionuclides into the biosphere might occur.
In spite of the large number of studies that have been carried out on this matter, a more specific characterization of the different uranium oxides involved in the conversion of UO₂ into U₃O₈ needs to be done for a better understanding of the structural and chemical evolution of the system. On this basis, the present study is focused on the first oxidation stage, from UO₂ to U₄O₉, with the aim of characterizing the UO_2+x (x < 0.25) hyperstoichiometric oxides in detail, as well as assessing the structural evolution taking place as oxidation proceeds. For this purpose, different UO_2+x powder samples, with controlled degree of non-stoichiometry, have been identified by thermogravimetric analysis and characterized by X-ray diffraction (XRD) and Raman spectroscopy. XRD analysis reflects that the commonly assumed Vegard’s law is not applicable over the whole hyperstoichiometry range, since a slight increase of the lattice constant is observed for 0.13 < x < 0.20. A quantitative Raman analysis of the UO_2+x spectra as a function of the oxidation degree is also shown. A new method to characterize any UO_2+x sample (for x < 0.20), based on the shift of the 630 cm^-1 band observed in the Raman spectrum, is proposed here for the first time. Moreover, three structure transitions have been detected at x = 0.05, 0.11 and 0.20, giving rise to four distinct regions associated with consecutive structural rearrangements over the hyperstoichiometry range: x < 0.05, 0.05 < x < 0.11, 0.11 < x < 0.20 and 0.20 < x < 0.25.

The fraction of fission products that dissolve faster than the Spent Nuclear Fuel (SNF) matrix and were segregated during irradiation are called Instant Release Fraction (IRF), and can be considered as the most important source of radiological risk in the performance assessment of a deep geologic repository. After the irradiation, IRF radionuclides (RN) like Cs and I are accumulated in the surrounding of the fuel pellet. Previous experiments with a BWR SNF with a Burn-Up (BU) of 42 GWd/t_HM and 215 W/cm of Linear Power Density (LPD) (42BWR), have shown that the RN trapped in the dishing (space between pellets) are the major contribution to the I and Cs IRF. Similar experiments were performed in this work with a PWR SNF with a BU of 60 GWd/t_HM and a LPD of 255 W/cm (60PWR). The contribution of the dishing in this case may be higher because of the higher LPD and BU of the PWR SNF but the morphological changes of the pellet during irradiation can also have an impact on the IRF RN accumulation in the dishing.
Static leaching experiments were performed using simulated granitic groundwater and three different cladded segments (CS). The CS were cut taking into account the position of the dishing, one in the top of the CS (TOP), another in the middle of the CS (MID) and the last one without dishing (FULL). Leaching solution was stirred continuously and completely replaced at each sampling time.
After the first analysis it is possible to observe that the dishing has a minor effect on the IRF corrosion experiments with the 60PWR SNF, especially when compared with the dishing effect on the 42BWR. In the 60PWR the fractures are much less abundant and narrower than in the 42BWR which makes more difficult for the water to reach the dishing in the MID CS, retarding the RN release from this region. Moreover, the dishing of the 42BWR TOP CS is completely open and slightly concave but the 60PWR one has still part of the following pellet attached to its surface and it is almost flat, decreasing the amount of IRF RN accumulated in it.
Since these two fuels had a different irradiation history, their pellets have also suffered from different morphological changes and the effect of the dishing on the IRF corrosion tests has also different magnitudes. Therefore, based on the first analysis, it is possible to consider that the dishing effect on the IRF corrosion tests depends on the irradiation history and the morphological characteristics of the SNF.

Nuclear power is of great importance to the future of low carbon energy production and the ability to separate and recover the actinide elements from spent fuel is a key requirement for a sustainable nuclear fuel cycle. While the extraction of U and Pu for the fabrication of new fuel is well established with the PUREX process, recovery of the actinides, and their separation from the chemically similar lanthanides, remains challenging.

A range of new organic extractant molecules, such as N,N,N’,N’’ tetraoctyl diglycolamide (TODGA), have been developed for the recovery of trivalent actinides through solvent extraction processes and it is important that they be well characterised with new understanding required for the associated chemical extraction mechanisms and kinetics.

Consequently, a study of the interfacial and mass transport kinetics of cerium extraction by TODGA has been conducted using a rotating diffusion cell (RDC) apparatus. The RDC comprises two solution phases which are separated by a defined area membrane interface and subjected constant rotation. This rotation establishes controlled hydrodynamic flow and well characterised boundary / diffusion layer conditions within each solution phase, facilitating the study of both diffusion and kinetic contributions to the rate of mass-transfer and the interrogation of the mechanism of extraction.

Studies to date have revealed significant insights into the Ce(III) / TODGA extraction system, indicating an interesting dependency on local hydrodynamics at the solution phase boundary with the key complexation reaction occurring in the aqueous phase. The extraction rate of Ce(III) has been shown to correlate with aqueous [Ce(III)] while the simultaneous extraction of HNO₃ by TODGA is also demonstrated. The use of HNO₃-pre-contacted TODGA indicates that the extraction of the acid may be inhibitive towards the continued extraction of metal ions and warrants further investigation.

A theoretical description of the Ce(III) / TODGA RDC system has been developed and combined with spectrometric quantification of the interfacial flux allowing for the determination of several key rate parameters including both the forward / complexation and back / decomplexation reaction rates, the aqueous decomplexation length and the interfacial rate constant.

The Swedish spent nuclear fuel canister design KBS-3 consists of a copper cylinder surrounding an iron insert that holds the spent fuel. Like in most other canister designs the mass of iron constitutes the majority of the canister weight. In order for groundwater to access the spent fuel in a future repository the outer canister must fail and iron corrosion occur. Spent nuclear fuel dissolution will therefor likely proceed under conditions of simultaneous anoxic iron corrosion. The iron corrosion can likely supress the spent fuel release by creation of strongly reducing conditions from Fe(II) formation and the generation of large quantities of hydrogen. Redox sensitive radionuclides may either be reductively precipitated by dissolved Fe(II) or from interaction with iron corrosion products such a magnetite or green rusts. The generated hydrogen (up to several MPa) may also inhibit the spent nuclear fuel dissolution at the surface of the fuel via the so called hydrogen effect. In order to probe these effects an autoclave experiment was performed in which a basket with PWR spent nuclear fuel (burnup ~43 MWd/kgU) was suspended in an autoclave containing a simplified groundwater (10 mM NaCl, 2 mM NaHCO3) together with iron powder. The autoclave was sparged and pressurised with argon. Following the expected initial rise in radionuclide concentrations from dissolution of pre-oxidised phases and the so called instant release fraction the U concentration dropped to 3x10^-9 M within 76 days, in-line with the expected solubility of amorphous UO₂, expected to form under reducing conditions. Any measurable Cs and Sr release also ceased within 223 days indicating a complete transition from dissolution of instant release fractions to conditions with inhibition of the dissolution of the fuel matrix. Gas phase and pressure monitoring showed a steady build-up of hydrogen at a rate higher than what could be attributed to radiolysis, reaching hydrogen partial pressures of serval hundred kPa. The results indicate no passivation of the iron corrosion, with magnetite as the likely major iron corrosion product.

Silver-functionalized silica aerogel (Ag⁰-aerogel) is being developed for the removal and sequestration of iodine compounds from the off-gas of a nuclear fuel reprocessing plant. This material shows promise as a potential replacement for silver mordenite because of its high selectivity and sorption capacity for iodine, and its feasible sequestration to a durable SiO₂-based waste form. The iodine-loaded Ag⁰-silica aerogel can be rapidly consolidated with hot isostatic pressing (HIP) and spark plasma sintering (SPS) at moderate temperatures and pressures into a waste form consisting of AgI particles encapsulated in the fused silica matrix. Highly iodine-loaded Ag⁰-aerogel was successfully consolidated with HIP at 1200°C with a 30-min hold and under 207 MPa. The fully densified sample had a bulk density of 3300 kg/m³ and contained ~39 mass% of iodine, The promising preliminary results were also obtained for samples consolidated with SPS, which offers the advantage of high densification rates at a lower processing temperature. The presentation will summarize the results from a series of consolidation studies.

The local chemistry of Tc-99 in oxide glasses is important for understanding the incorporation and long-term release of Tc from nuclear waste glasses, both those for legacy defense wastes and fuel reprocessing wastes. It is known that Tc preferentially forms Tc(VII), Tc(IV), or Tc(0) in glass, depending on the level of reduction of the melt. Tc(VII) in oxide glasses is normally assumed to be isolated pertechnetate TcO₄^- anions surrounded by alkali, but can occasionally precipitate alkali pertechnetate salts such as KTcO₄ and NaTcO₄ when Tc concentration is high. In all these cases of Tc(VII), Tc is 4-coordinated with oxygen. A reinvestigation of the chemistry of alkali-technetium-oxides formed under oxidizing conditions and at temperatures similar to those used in the melting of nuclear waste glasses showed that higher coordinated alkali Tc(VII) oxide species have been reported, including those with the TcO₅^- and TcO₆^- anions. The chemistry of alkali Tc(VII) and other alkali-Tc-oxides is reviewed, along with relevant synthesis conditions.
Additionally, we report the attempts to make alkali compounds of K, Na, and Li technetates as TcO₅^- and TcO₆^-. It was found that higher coordinated species are very sensitive to water, and easily decompose into their respective pertechnetates. It was difficult to get pure compounds, but mixtures of the pertechnetate and another phase(s) were frequently found, as evidenced by x-ray absorption spectroscopy (XAS), neutron diffraction (ND), and Raman spectroscopy. In addition, low temperature electron paramagnetic resonance (EPR) measurements showed the possibility of Tc(IV) and Tc(VI) in Na₃TcO₅ and Na₅TcO₆ compounds.
It was suspected that smaller counter cations would result in more stable technetates. To confirm the synthesis method, LiReO₄ and Li₅ReO₆ were created, and their Raman spectra match those in the literature. Subsequently, the Tc versions LiTcO₄ and Li₅TcO₆ were synthesized and characterized by ND, Raman spectroscopy, XANES, EXAFS, and nuclear magnetic resonance (NMR). The Li₅TcO₆ was a stable compound which appears to have the same structure as that known for Li₅ReO₆. Analysis of LiTcO₄ is still underway.
Some implications of the experimental work on stability of alkali technetate compounds and possible role in the volatilization of Tc are discussed.

The capture of radioactive iodine (I2) remains a important concern for safe nuclear wastes, since a more efficient technology related to the retention of radioelements is still needed. While various solids have been already tested such as zeolites, clays, MOFs,... it is still required to rationalize the impact of chemical composition of solids on the adsorption of this vapor. To reach this aim, we propose to study by Monte Carlo simulations using classical force fields various Metal Organic Frameworks (or MOFs). These solids are porous frameworks containing both organic and inorganic parts which can be modulated as will. Here, we focus on a series of UiO-66 (a MOF stable to the water) with various functions (Br, Cl, CF3, COOH,...) carried by the phenyl rings to investigate the influence of chemical functions on the adsorption mechanism. Furthermore ionic MOFs containing anions (such as Cl-, NO3-,...) or alkali/alkali-earth cations as compensating ions such as MIL-141, MIL-127 and Zn-BTeC have been also studied to determine the impact of the electrostatic charges on the I2 adsorption.

Using molecular simulations, it is possible to calculate the impact of such chemical modifications both on the affinity of the solids at low loading and the saturation at high loading. Furthermore, microscopic models allow us to elucidate the adsorption sites as well as to determine the adsorption isotherms, which can lead to propose some recommendations for the design of new MOFs in view of the capture of iodine vapor.

The immobilisation of radioiodine produced in the nuclear fuel cycle is a growing priority for nuclear wasteform research and development. In particular, 129I is of concern for used nuclear fuel reprocessing facilities due to its very long half-life (1.6 x 10⁷ years) and its high mobility in most geological environments. CuI in water has a very low solubility product and unlike AgI does not decompose when exposed to water containing Fe metal. We have investigated various methods for its production and have investigated its consolidation by sintering in argon and/or hot isostatic pressing in stainless steel cans. Ag4Al3Si3O12I has been formed by sintering in air or hot isostatic pressing in Ni or Cu cans at temperatures in the 750-900^oC range. HiPed samples were investigated by SEM to check the stoichiometry of the AgI sodalite phase, the general microstructure and the reaction between the sodalite and the can material. PCT and MCC-1 leach data will be reported on material HIPed in Cu cans.

A vital aspect of any sustainable nuclear fuel cycle is the utilization of a viable waste form for the immobilization of radioisotopes and fission products. One isotope of particular concern within the radioactive waste community is Iodine-129. The current proposed technology for the removal of I-129 from the radioactive waste stream typically involves the use of caustic scrubbing and Ag solid sorbents to capture iodine off-gas. Following the capture of iodine gas, the Ag solid sorbents are either dissolved and stored in waste tanks as a liquid (as is done with caustic solution) or stored in waste tanks as a solid. This storage of capture medium, however, is not a viable long-term solution of immobilization and a waste form remains to be thoroughly developed for various reasons. Conventional borosilicate vitrification, for instance, is not a viable option due to the low solubility of iodine in many glass chemistries. Most importantly, iodine is highly volatile at typical glass processing temperatures (1000-1100 °C). Alternative waste forms have been explored, but sufficient maturity and satisfactory properties have yet to be achieved to be considered viable.

Apatites have long been considered a possible candidate for the immobilization of radioactive iodine. Due to its ability to accommodate halides into its crystal structure along with its acceptable durability against radiation, apatites make a very viable option. The challenge involving apatite synthesis, however, has always been incorporating the large iodide ion into the crystal structure. The most promising, applicable composition is the iodoapatite, Pb₁₀(VO₄)₆I₂. The synthesis of the iodoapatite, however, has mainly consisted of melting at elevated temperatures (500-800 °C) in controlled, sealed reaction environments due to the volatility of iodine. This method would prove challenging for a large-scale production. Other novel methods have included spark plasma sintering, hot isostatic pressing, microwave heating, and high-energy ball milling, but those methods also continue to have flaws limiting their adoption.

Our research presents the first reported instance of low temperature wet-chemical synthesis of this iodoapatite. By lowering the synthesis parameters to ambient temperature and ambient atmosphere along with the incorporation of a solution-based method, the process can easily be incorporated onto a large scale. Along with the synthesis method, solid solution studies involving the gradual substitution of phosphate for vanadate and calcium for lead, Pb_(10-x)Ca_x(VO₄)_(6-y)(PO₄)_yI_2, were also performed with numerous characterization techniques for the purpose of investigating chemical durability improvement. Sintering studies were performed as well for the purpose of minimization of surface area and, hence, an improvement in chemical durability and leaching resistance. The results pertaining to synthesis and characterization of these minerals will be discussed at the symposium.

The International Simple Glass (ISG) is considered as reference benchmark glass and it has been developed in the frame of an international collaboration for the study of the dissolution mechanisms of high-level vitrified nuclear waste.
In this work the forward dissolution rate of the ISG has been determined in different alkaline solutions, as a simulation of the disposal conditions foreseen by the Belgian concept for geological diposal of vitrified nuclear waste. The determination of the forward dissolution rate has been carried out at 30 °C in four different KOH solutions with pH varying from 9 to 14 and in artificial cementitious water at pH 13.7 ± 0.2.
The values determined in this study have been compared with the rates measured in the same conditions for SON68 glass in a previous work [1]. The results obtained for the two glasses are comparable in both alteration media. However, ISG glass shows a smaller forward dissolution rate with respect to SON68 in KOH at pH 14 (0.204 ± 0.126 g/m²d for ISG and 0.355 ± 0.265 g/m²d for SON68), while in artificial cementitious water the large uncertainties make the comparison of the results more difficult. The forward dissolution rates calculated for the ISG in KOH solutions, moreover, are in good agreement with the initial dissolution rates presented by Inagaki et al. and obtained for a lower pH range [2].

References:
1. Ferrand, K. and K. Lemmens, Determination of the forward rate of dissolution for SON68 and PAMELA glasses in contact with alkaline solutions, in Scientific Basis for Nuclear Waste Management Xxxi, W.E. Lee, et al., Editors. 2008, Materials Research Society: Warrendale. p. 287-294.
2. Inagaki, Y., et al., Initial Dissolution Rate of the International Simple Glass as a Function of pH and Temperature Measured Using Microchannel Flow-Through Test Method. International Journal of Applied Glass Science, 2013. 4(4): p. 317-327.

Within the United Kingdom (UK) , it is proposed that nuclear waste will be disposed in a geological disposal facility, 200 m to 1 km underground¹. This facility will incorporate an engineered barrier system that will be optimised to physically and chemically impede the transport of radionuclides to the biosphere. The facility will house a large volume of cemented Intermediate Level Waste (ILW), in addition to vitrified ILW. A significant volume of concrete will be used in its construction. Interaction of groundwater with the cementitious components of the facility (both the waste and construction materials) will lead to the presence of high pH conditions within a repository. The effect of cement leachates on vitrified wasteforms is not well understood.

We present results from a glass durability study using idealised cement leachates to develop our understanding of glass durability mechanisms in these complex repository like environments. Simulant ILW glasses relevant to the UK disposal program have been utilised. We also investigated a simulant UK high level waste glass (MW-25%) and the International Simple Glass² (ISG), a 6 component borosilicate glass, with components that are common to most borosilicate nuclear glasses. Glass powders were exposed to idealised cement leachates of “intermediate” and “old” ages, approximately representative of GDF conditions at ~1000 and ~10,000 years of operation, according to the product consistency test B³. Analysis of the normalised mass loss and normalised leaching rate of these glasses as a function of cement leachate composition was achieved through analysis of solution concentrations. Simultaneously we present analysis of monolith sample alteration layers by SEM/EDX and GA-XRD. Collectively, these data support a mechanistic understanding of glass dissolution in the context of a complex geological disposal environment for vitrified UK waste.

References
1 Department of Energy & Climate Change, Implementing Geological Disposal, 2014.
2 S. Gin et al, An international initiative on long-term behavior of high-level nuclear waste glass, Materials Today, 2013, vol. 16.
3 Standard Test Methods for Determining Chemical Durability of Nuclear , Hazardous , and Mixed Waste Glasses and Multiphase Glass Ceramics, The Product Consistency Test (PCT) ASTM C1285-14, 2002, vol. 15.

At the Hanford site, USA, low activity tank wastes will be immobilised by vitrification to create 150-350,000m³ of Immobilised Low Activity Waste (ILAW) destined for disposal in a shallow subsurface Integrated Disposal Facility (IDF). During reprocessing, tank wastes have a separable low activity waste fraction that will report to the LAW facility for treatment. Radionuclides of concern in LAW glasses will include ⁶⁰Co, ¹³⁷Cs, ¹⁵⁴Eu, ⁹⁹Tc and ⁹⁰Sr alongside toxic metal contaminants. Conditions in the IDF are expected to differ from those within a deep geological disposal facility for high level waste with temperatures expected to be ~ 15^o C, an arid climate, variable pH and groundwater flow rates.

Project GLAD (Glass Leaching Assessment for Disposability) investigates newly developed leaching technologies for assessing the durability of ILAW glasses. Four new methodologies developed by the U.S. Environmental Protection Agency (EPA) for application to the accelerated ageing of ILAW glass are compared to established leaching tests accepted for evaluation of high level waste glasses, including PCT and MCC-1 protocols. The GLAD project studies the process of glass dissolution and constituent leaching as a function of temperature, pH, groundwater composition and flow rate.

Three candidate glasses, LAW A44, ORP LB2 and LAW A23 were analysed by US EPA leaching methods ‘1313’ and ‘1315’ in both deionised water and synthetic groundwater and results were compared to those obtained by standard PCT-B and MCC-1 tests. In addition, longer term leaching tests were performed to determine the effects of alteration layer formation on the rate of contaminant leaching.

Furthermore, simplified glasses, representative of candidate glasses, were designed and analysed in parallel to improve understanding of how glass composition affects dissolution rates.

Funding for this work was provided by William F. Hamel, Jr., Assistant Manager, of the U.S. Department of Energy Office of River Protection Waste Treatment & Immobilization Plant Project.

Project collaborators: Pacific Northwest National Laboratory (PNNL) and the Consortium for Risk Evaluation with Stakeholder Participation (CRESP).

HLW waste glasses are to be stored in a deep geologic repository. Some potential repository geologies have oxidizing ground waters while some have reducing or anoxic ground waters. The differences in the oxidizing potential of the groundwater, expressed as groundwater Eh, causes different rates of dissolution of the major glass components such as B, alkali, and silica. Moreover, the groundwater Eh has a profound impact on the release of multivalent species such as iron and uranium from the glass. ASTM-C1220 experiments at 90°C were performed in an Ar glovebox under anoxic conditions with iron present in both deionized water and in simulated basaltic groundwater. The deionized water and basalt groundwater were sparged of oxygen by bubbling argon gas through the solutions for >48 hours. The basaltic groundwater was then pre-equilibrated with ground basalt under these deoxygenated conditions. After argon sparging, the iron and or ground basalt served as the low Eh buffer. A companion set of experiments was done where the leach vessels were filled in the Ar glovebox but the durability was performed at 90°C in an oven on a bench top. Since Teflon vessels were used for the leach testing, the bench top experiments slowly oxidized. The same uranium doped HLW glass was leached in the anoxic glovebox environment and in the oxic bench top environments.

The leachate solutions were split into two aliquots and one was filtered and one was not. The releases of B, Na, Li, Fe and U were monitored. All oxic tests gave higher releases of U and higher releases of all glass components than the anoxic tests. Since all the tests included an iron bar representative of a potential waste package component, the iron pumping mechanism described in the 1980’s and shown to rapidly deteriorate HLW waste glass did not occur under anoxic conditions as it does under oxic conditions. The filtered samples gave the same releases of soluble B and Li but much lower concentrations of uranium. Geochemical modeling of the measured Eh-pH conditions from the oxic and anoxic experiments at 90°C, using Geochemist’s Workbench software, demonstrated that this is caused by the precipitation of the uranium as UO_2.6667 and/or UO₂(OH)₂.

Microporous inorganic solids composed of networks of octahedrally coordinated zirconium and tetrahedrally coordinated silicon atoms occur extensively in nature, and include petarasite (Na₅Zr₂Si₆O₁₈(Cl,OH)^.2H₂O) and umbite (K₂ZrSi₃O₉^.2H₂O). These are chemically and structurally related to titanosilicates such as CST (Na₂TiSi₂O₇^.2H₂O), the active ingredient in IONSIV, which is known to be an excellent material for the selective removal of radioactive Cs and Sr from aqueous solutions via an ion exchange process. We are investigating these and other related materials for use in the nuclear decommissioning activities in the UK and post-Fukushima clean-up activities in Japan. Results on the parent materials, germanium analogues and transition metal doped analogues will be presented covering synthesis, ion exchange testing and thermal conversion into dense ceramics.

The barium titanate hollandite is a promising crystalline host for Cs-immobilization, particularly for long term disposal of used fuel processed in a combined waste stream. Hollandite has been shown to sequester Cs during the formation of multiphase ceramics; however, Cs incorporation is compositionally dependent. That is, the M-site dopant of hollandite with the form (Ba_xCs_y)(M_zTi_8-z)O₁₆ can impact the formation of the Cs-doped hollandite phase. In this work, we undergo a detailed evaluation of the atomic structure across the Ba-Cs hollandite binary for Zn-, Ga-, and Al-doped hollandite. Quantum-mechanical calculations were used to observe the impact of A-site and B-site ordering on the structural stability of hollandite. The enthalpy of formation was quantified and agrees with calorimetric measurements of related hollandite phases. Ground state geometry optimizations show that, for intermediate compositions (i.e., Cs₂Ba₂Ga₆Ti₁₈O₄₈), mixing on the A-site is not energetically favored. That is, configurations with Cs and Ba mixed within a channel resulted in higher total energy (~ 0.04 eV) as compared with configurations with Cs and Ba segregated into separate channels; however, the energetics of Cs and Ba mixing may be overcome by the decay heat associated with the β-decay of ¹³⁷Cs to ¹³⁷Ba. The B-site dopants (i.e. Zn, Ga, or Al) prefer ordering with high symmetry and tend to arrange within a single tunnel layer; however, the six B-site dopants in the intermediate Ga- and Al-doped systems cannot arrange symmetrically about the eight B-sites that comprise each tunnel layer, and instead align along the tunnel direction.

The Ga-doped hollandite compositions were synthesized and characterized, showing agreement between DFT, XRD, EXAFS, and neutron diffraction measurements of the atomic structure. The lattice parameter associated with the tunnel dimension was found to increase with Cs concentration. A trend of decreasing thermodynamic stability with smaller tunnel cations was ascribed to the increasing structural distortion observed in the system. The interatomic distances and arrangement of tunnel cations reveals that the hollandite structure can strongly stabilize the A-site cations in the tunnel, even at elevated temperatures up to 500K. A direct investigation of cation mobility in tunnels using electrochemical impedance spectroscopy was conducted to demonstrate the ability of the hollandite structure to immobilize cations over a wide compositional range. The pure Cs-hollandite, with the largest tunnel size and longest rod-like microstructural features, exhibited the highest ionic conductivity. Thus, control of grain size and optimized Cs concentration are essential to limit cation motion and propensity for elemental release.

A previous study revealed Cs₂TiNb₆O₁₈ to be the major Cs-containing phase after hot isostatic pressing (HIPing) Cs-loaded IONSIV (a commercial exchanger). This material has demonstrated excellent waste form properties including aqueous durability. Both experimental and theoretical studies have been carried out in order to access if Cs₂TiNb₆O₁₈ is able to retain ¹³⁷Ba²⁺, the transmutation product of ¹³⁷Cs⁺. A series of samples with different charge compensation mechanisms have been synthesised including Cs_2-xBa_xTi^(III)_xTi^(IV)_1-xNb₆O₁₈ (Ti^(IV) reduction to Ti^(III)) and Cs_2-xBa_xTiNb^(IV)_xNb^(V)_6-xO₁₈ (Nb^(V) Reduction to Nb^(IV)). However due to difficult sample preparation, the majority of samples synthesised instead follow the formula Cs_2-xBa_xTi^(IV)_1+xNb^(V)_6-xO_18. Though not what would form in real conditions, these samples could still give good indication whether Ba can be retained in the structure. X-ray diffraction (XRD), X-ray fluorescence (XRF) and microscopy studies have proved inconclusive and therefore a series of calculations have been carried out using the GULP (General Utility Lattice Program) code in order to see if Ba incorporation is energetically favourable.

Phosphates compounds attract attention of specialists for the development of a ceramic-based immobilization product of long-lived actinides and fission products. A successful solution to this problem requires fundamental knowledge in chemistry and crystal chemistry of the chemical systems, containing important components of the waste streams.

Phosphates of the general composition M'_xM''₂(PO₄)₃ with M'=Na/Eu, M''=Zr, and M'=Fe, Ga, Y, In, lanthanides, M''=Hf, belonging to the NaZr₂(PO₄)₃-type structure family (NZP), were synthesized by high temperature treatments of precursors obtained by precipitation. The higher neutron cross section of Hf can potentially be used on demand in some applications in which neutron absorption properties are required. The scientific basis for such «tailoring» of properties of the NZP phosphates is the high isomorphic capacity of the structure. Isomorphism is an important property that must be taken into account for the immobilization of mixed radioactive wastes in ceramic materials.

The structures of selected representatives were refined by the Rietveld method from the X-ray powder diffraction data. New compounds were characterized by several techniques such as SEM-EDX, DTA/TG, Mössbauer and Raman spectroscopy, solid state NMR and high temperature drop-calorimetry.

The solid solution Na_xEu_(1-x)/3Zr₂(PO₄)₃ with Eu mimicking chemical behavior of a 3-valent actinide showed complex structural features, involving a morphotropic transition in the series as a result of ordering of N