Technetium-99 (⁹⁹Tc) is one of radioactive tank waste components contributing the most to the environmental impacts associated with disposal of radioactive wastes currently stored in underground tanks at the Hanford site, WA. Cast Stone has been selected as the preferred waste form for solidification of aqueous secondary liquid effluents from the Hanford Tank Waste Treatment and Immobilization Plant (WTP) process condensates and low-activity waste (LAW) melter off-gas caustic scrubber effluents. Cast Stone is also being evaluated as a supplemental immobilization technology to provide the necessary LAW treatment capacity to complete the Hanford tank waste cleanup mission in a timely and cost effective manner. Research is being conducted to improve the retention of Tc in the Cast Stone waste forms.
One method to improve the performance of the Cast Stone waste forms is addition of “getters” that selectively sequester Tc. Getter materials that remove Tc from solution are expected to reduce Tc(VII) to the less mobile Tc(IV), In order to determine the effectiveness of the various getter materials prior to their solidification in Cast Stone, a series of batch sorption experiments was performed. Seven getter materials were tested for Tc. Testing involved placing getter materials in contact with spiked waste solutions for periods up to 45 days with periodic solution sampling. Two different solution media, 18.2 MOmega; DI H₂O and a 7.8 M Na LAW waste simulant, were used in the batch sorption tests. Each test was conducted at room temperature in an anoxic chamber containing N₂ with a small amount of H₂ (0.7%) to maintain anoxic conditions.
In this paper we present the results of the batch experiments conducted to determine potential Tc getter materials that will undergo continued testing, selection and subsequently incorporation into Cast Stone. Results indicate that most materials perform better in the DI H₂O (18.2 MOmega;) solution than in the 7.8 M Na LAW waste simulant. We have determined that Tc sequestration may be affected by the presence of other redox sensitive elements that are present in the waste simulant, such as Cr(VI). The Tc getter materials have been examined through various solid-state characterization techniques such as SEM/EDS and XANES. The results indicate that the Tc precipitate differs depending on the getter material and that Tc is reduced in most of the getters but at different extent, from Tc(VII) to Tc(IV).

The immobilization of ⁹⁹Tc and ¹³⁷Cs, two problematic nuclear waste isotopes, in an ABO₄-type structure (where A is an alkali metal and B is either Tc or Ru) has been investigated using atomic scale modelling techniques. The structural stability and free energies of several ABO₄ compounds, including that of CsTcO₄ were computed. Most perfect compositions were calculated to be scheelite structured. To understand the potential wasteforms&’ stabilities during and after transmutation of ⁹⁹Tc to ⁹⁹Ru, and ¹³⁷Cs to ¹³⁷Ba, we also computed the structures and energies of a range of defective ABO₄ compounds. Full and partial transmutation of the waste isotopes were considered, i.e., those of compounds such as A(Tc_1-xRu_x)O₄, (Cs_1-xBa_x)TcO₄ and ARuO₄. We present a number of compositions that may prove to be suitable for either ⁹⁹Tc waste as well as the simultaneous encapsulation of ⁹⁹Tc and ¹³⁷Cs.

⁹⁹Tc, a radioactive fission product, is discharged in nuclear fuel reprocessing operations. In its common form in the environment (TcO₄^minus;) it is a major concern for the remediation of contaminated waters. The difficulty with the removal of TcO₄^minus; is a result of its high mobility in solution and the presence of a high concentration of competing anions such as SO₄^2minus;. A functionalized material, developed at PNNL, known as self-assembled monolayers on mesoporous supports (SAMMS) has previously been found to selectively remove contaminant monovalent oxyanions even in the presence of the divalent SO₄^2minus;.
In this work we have constructed an atomistic model of the SAMMS material and validated the model by comparison to experiment. Density functional theory (DFT) calculations were used to parameterize a classical force field that accounts for the specific interactions of the competing oxyanions with the monolayer. Potentials of mean force for oxyanion adsorption were obtained using umbrella sampling and molecular dynamics (MD) simulations in order to study the material&’s preference for binding TcO₄^minus; over SO₄^2minus;. The results show that the pore structure is a key parameter governing the material&’s oxyanion selectivity. Finally, we suggest ways in which the structure of the material can be optimized in order to maximize TcO₄^minus; adsorption.

Among various radioactive nucleotides, ¹³⁷Cs is the most dangerous radioactive nucleotide because of its high fission yield (6.09 %), medium half-life (30.17 years), and very high solubility in water regardless of its counter anion. Once released into the environment, it easily spreads in nature and enters the food chain, causing enormous damage to human and animal health. In this respect, the effective removal of ¹³⁷Cs⁺ ions from contaminated groundwater, seawater and radioactive nuclear waste solutions is crucial for public health and for the continuous operation of nuclear power plants. However, it is an extremely difficult task because ¹³⁷Cs⁺ concentrations are usually incomparably lower than those of the co-existing competing cations (Na⁺, Ca²⁺, Mg²⁺, K⁺, and others).
Herein we report a novel microporous vanadosilicate K-SGU-45 with mixed valences of vanadium (IV and V), which shows an excellent capturing and immobilization of Cs⁺ from ground water, seawater and highly acidic and basic nuclear waste solutions. This material is superior to other known materials in terms of selectivity, capacity, and kinetics, in particular, at very low Cs⁺ concentrations, it was found to be the most effective material for the removal of radioactive Cs⁺.
This work will trigger the syntheses of various vanadium and other transition-metal silicates that capture various radioactive nuclides, such as ⁹⁰Sr²⁺ ions, and other toxic heavy-metal ions. Furthermore, the discovery of unprecedented hexadeca-coordinated Cs⁺ centers, which corresponds to the highest coordination number ever observed in chemistry, has been described.

Strontium-90 and caesium-137 are waste products produced by fission processes; both have long half-lives of 28 and 30 years respectively. Strontium in particular can have a severe biological impact as it has been shown to accumulate in bones after the intake of contaminated food or water.
Ion exchange materials, such as crystalline silicotitanite (CST, Na₂Ti₂O₃SiO₄middot;2H₂O) and commercially available IONSIV^[1], have been implemented in order to target and remove these harmful radioisotopes and have been shown to be somewhat effective. However Strontium and caesium have proven difficult to immobilise selectively in some cases as ion exchange uptake has been shown to be retarded by the presence of competing cations such as calcium or magnesium^[2] .
A range of new materials similar to CST but based on zirconium and tin silicates, such as NaKSnSi₃O₉.H₂O pictured below, have been investigated for their potential ion exchange applications. These materials are robust against thermal, chemical and radioactive conditions which would make them ideal for use in radioactive waste streams.
A range of materials and the ion exchanged heat treated waste forms have been studied using XRD, TGA, XRF, SEM and EDX analysis in order to characterise the structures and probe the ion exchange properties.
References
1.R.G. Anthony, C.V. Philip, R.G. Dosch, Waste Manage, 1993,13, 503
2.T. Möller, R. Harjula, M. Pillinger, A. Dyer, J. Newton, E. Tusa,S. Amin, M. Webb and A. Araya, J. Mater. Chem.,2001, 11, 1526

An extension of the Swedish final repository for short-lived radioactive waste (SFR) is planned and a safety assessment has been performed as part of the licensing process. Within this work, steps have been taken to advance the modelling environment to better integrate its individual parts. It is desirable that an integrating modelling environment provides the framework to set up and solve a consistent hierarchy of models on different scales. As a consequence, the consistent connection between software tools and models needs to be considered, related to the full assessment domain. It should also be possible to include the associated geometry and material descriptions, minimizing simplifications to source data. The usefulness of the analysis software Comsol Multiphysics as component of an integrating modelling environment has been tested and examples of development work are presented.
Geometry handling is an important part of the modelling process and is closely related to modelling assumptions and simplifications. For the SFR, the relevant geometry includes tunnel systems and storage vaults, as well as engineered structures and barriers. CAD geometries developed during planning and design work have been successfully imported into Comsol. The landscape above the repository also constitutes relevant geometrical input for assessment modelling. Development work has allowed the import of geographic information system (ArcGIS) data into Comsol, incorporating digital elevation models as well as soil and sediment domains into model geometries.
The ability to set up and solve a consistent hierarchy of models on different scales is an important capability of an integrating modelling environment. Extracting models for repository scale hydrology from regional hydrogeology models and regional surface hydrology models are two examples. The regional hydrogeology model of the SFR site covers several square kilometres of land and reaches depths of approximately one kilometre. The repository scale model is contained within the regional model and has dimensions one order of magnitude smaller. To calculate the detailed groundwater flow through the repository requires the proper boundary conditions from the regional hydrogeology. A consistent connection was achieved by programming an interface allowing Comsol to extract the near-field boundary conditions and bedrock property fields from the regional model, set up and solved in the DarcyTools software.
The repository scale hydrology models provided a basis for further model developments focused on coupled processes. An interface between Comsol the geochemical simulator PhreeqC has been developed to support reactive transport studies. An important test case involved radionclide transport in a 3D model of a catchment area. The dynamic surface hydrology was simulated with MIKE SHE and coupled to detailed chemical processes occurring in soils and sediments.

Reactive transport modelling entails the integration of hydrogeology and geochemistry. One of the challenges for such integration is the large amount of computational resources needed due to the high non-linearity of the resulting system of equations. Taking advantage of new developments of powerful numerical tools, and based on high performance parallel computing, the solution of large-scale hydro-thermal-geochemical-mechanical models has become possible. A software solution, denoted iDP, has been developed which serves as an interface between 2 standalone simulators: DarcyTools [for groundwater flow in fractured rocks] and PFLOTRAN [for reactive solute transport]. iDP has been applied for the first time to test the new update of Mare Nostrum, the main machine at the Barcelona Supercomputing Centre, the National Supercomputing Centre in Spain. An average of 8,000 processor cores during 15 days were used to solve a large-scale (100 Mcells), long-term (20,000 years) simulation to evaluate the degradation of cement grout that will be injected in the fractures of the granitic rocks during the construction of a deep geological repository for spent nuclear fuel in Forsmark (Sweden). The simulation integrates the complex 3D groundwater flow accounting for the Discrete Fracture Network (DFN) of the site, and the complexity of the geochemical system involved in cement grout dissolution and secondary minerals precipitation within the flowing fractures. Model results allow evaluating the expected durability of the injected cement grout, as well as to evaluate the risk of hyper-alkaline groundwater development and migration towards the depositional area of the repository. This work shows that High Performance Computing of reactive solute transport is a reliable and powerful tool for decision makers involved in the planning and constructions of deep geological repositories for nuclear waste.

A simple and effective model for a safety assessment of a conceptual repository system, in which low-level radioactive wastes that arises from nuclear power plants and other sources has been developed using the commercial GoldSim development tool. The repository system is assumed to be planned for construction on the surface area near the seashore. The computer program based on this model, developed as a GoldSim template, is ready for a total system performance assessment (TSPA), and is able to probabilistically evaluate a nuclide release from a repository and farther transport into the geosphere and biosphere under various normal, disruptive events, and scenarios that can occur after a failure of a waste drum with associated uncertainty. To quantify the nuclide release and transport through the various pathways possible in the near- and far-fields of the repository system under a normal groundwater flow and some alternative scenarios, illustrative evaluations are made and demonstrated through this study. Even though all parameter values associated with the repository system were assumed for the time being, the illustrative results should be informative since the evaluation of such releases is very important not only in view of the safety assessment of the repository, but also for the design feedback of its performance.
The 200L storage drums for low-level waste, which amounts to a total of 125,000 drums, are to be disposed of in concrete containers and then buffered by gravel or grouted with concrete. Impervious materials and multilayered covers for preventing water infiltration and some erosion as well as nuclide release are considered to place on the roof. In GoldSim modeling, a trench and its surrounding are discretized into several compartments ready for run-off, infiltration as well as diffusive and advective transport in and among them. Several principal release pathways from the trenches are set in place: the upper, side, and base pathways, all of which simultaneously reach to the far-field transport. All releases from the trenches are then later transported along with various unsaturated and saturated pathways including surface and subsurface groundwater flow pathways into the natural far-field area.
For trench type repositories at the surface or possibly subsurface depth, normally and commonly, once leakage from a damaged radioactive waste package of a drum, and through tiny holes, happens, the nuclides will spread out to the buffer material surrounding the drum, and then into other possible regions in the trench before farther transporting into the biosphere through various pathways. In the case of transport into the rock medium under the repository, the internal fractures and the major water conducting features (MWCFs) that are assumed to exist in the far-field area of the repository could be one of the main pathways through which the nuclides finally reach the human environment by passing over the geosphere-biosphere interfaces for exposure to human bodies.
The scenario mainly considered here for a probabilistic safety assessment is a normal case, under which nuclides are released by overflow and/or groundwater that normally flows along their own preferential pathways after release from each repository. Through this study, a probabilistic behavior of nuclide releases from a low-level waste trench type repository is illustrated with varying parameters, which were selected among many others in view of their possible consequences and probabilities.

The safe treatment of nuclear waste poses a lasting risk to the environment and has high costs. Buried repositories represent the long term storage which is required to be stable. The stability includes many aspects such as chemical and mechanical stabilities as well as impermeability. Clay minerals are excellent candidates to maintain a long lasting seal of the nuclear waste repository due to their large adsorption capacity and swelling characteristics in aqueous suspensions. However, the interaction and transport of radionuclides in clay minerals, including organic clay minerals, still need to be fully addressed.
Atom level simulations have not yet been fully exploited to investigate these processes not least because of the complexities involved. Here we present our recent work to gain atomistic insights into the factors controlling the interaction of heavy and radioactive ions at clay mineral - water interfaces. Quantum and potential based techniques are used to explore the evolution of systems of different sizes and for different lengths of time enabling us to efficiently evaluate structural and dynamical properties of this class of geosorbents. The interaction of these ions with clay minerals is generally thought to occur on the basal plane, which dominates their morphologies and has been the focus of many investigations. However, the edge surfaces are more reactive and with a greater range of compositions and charge states can indeed provide more efficient interaction sites.

Glass has been widely adopted as the first generation host material for the immobilisation of high level nuclear waste. It is intended that immobilised waste will go for permanent disposal in geological repositories and it is desirable that any wasteform should be durable under these conditions for an extended period. Atomic scale computer simulation can be used to provide a mechanistic basis for the structure and properties of glasses and as a result offers opportunities for the compositional optimisation of nuclear waste glasses.
Experimental studies have reported that zinc oxide improves the durability of nuclear glasses. Through the use of molecular dynamics, in conjunction with a simulated melt-quench procedure, atomic structures of sodium silicate glasses, have been generated with and without zinc. The structure of these glasses was studied through the use of pair distribution functions, ring size distributions and cluster analysis. Using the insights gained from these analyses the structural role of zinc oxide within silicate glass is discussed and consideration is given to reports of its differing roles as a network former and a network modifier. The effects of Zn addition on sodium ion distribution and clustering behaviour within the glasses is also reported. This is used to explain changes to intermediate-range structure and hence provide a possible explanation for the experimentally observed increase in durability obtained with the addition of Zn.

Radioactive iodine-bearing materials, such as spent silver adsorbent, are produced in nuclear reprocessing plants in Japan. According to Japanese disposal plan radioactive wastes that contain a certain quantity of iodine-129 are classified as Transuranic Waste Group 1 (TRU 1) for spent silver adsorbent or as Group 3 for bitumen-solidified waste and they should be disposed of by burial deep underground. Because the half-life of iodine-129 is 15.7 million years, it would be difficult to prevent release of iodine-129 from the wastes into the surrounding environment over such a prolonged time. Moreover, because iodine in its ionic forms is soluble and not readily adsorbed, its migration is not retarded significantly in engineered or natural barriers. Therefore the release of iodine-129 from nuclear wastes needs to be restricted to permit reliable safety assessment; this technique is called “controlled release”. It is desirable that iodine release period will be longer than 100,000 years.
Several techniques for immobilization of iodine have been developed for this purpose. These are narrowed down to three techniques such as synthetic rock, BPI (BiPbO₂I) glass and high performance cement. Iodine will be fixed as AgI in grain boundary of corundum or quartz through hot isostatic pressing (HIP) in the synthetic rock, as BPI in boron-lead based glass, or as some cement minerals such as ettringite in alumina cement. These techniques are assessed by three models such as the leaching model, the distribution equilibrium model, and the solubility-equilibrium model. In this paper current status of these techniques are described.

To support the future expansion of nuclear energy an effective method is needed for the capture and safe storage of radioisotopes released during reprocessing of spent nuclear fuel. The Department of Energy Office of Nuclear Energy (DOE-NE) is currently investigating alternative waste forms for ¹²⁹I. DOE is interested in new waste forms that can provide higher waste loadings, more efficient consolidation routes, lower costs, etc. PNNL has been developing non-oxide aerogels made with metal sulfides, termed chalcogels, for iodine immobilization and thus far, the materials do show promise as a potential replacement avenue for AgZ. These chalcogels are stable in aqueous solutions. Scientists at the university lead on this proposal who area the inventors of the chalcogel class of materials have already demonstrated selective affinity with chalcogels for metal ions in aqueous media such as Cs⁺, Sr²⁺, and Co²⁺. Aerogels have been studied for confinement of radioactive wastes in recent years and are under investigation as waste forms for ¹²⁹I. Aerogels can act as precursors to the final glass matrix that actually immobilizes the wastes. Use of silica aerogels for the purpose has been limited by their brittleness in the presence of water that is commonly present in off-gas treatment and also due to their low permeability to nuclear waste.
Recently, we reported a new type of aerogel made with metal chalcogenides (where chalcogen is S, Se, and/or Te) and is referred as chalcogel. Non-oxide materials such as the chalcogels have different properties than oxide materials and, in this case, some of those differences are actually advantages. For example, the high polarizability of the chalcogens (over oxygen) can be used to capture iodine. We will report the exploration of chalcogels as high affinity materials for capturing iodine and the conversion of the loaded materials to glass forms. The efficiency of a chalcogel-based waste form is expected from strong complex formation based on the high chemical affinity of chalcogen atoms for iodine gas. The strong chemical affinity is due to the soft Lewis acid/soft Lewis base complex formation, according to Pearson&’s Hard/Soft Acid-Base (HSAB) principle. We also report that chalcogels can be chemically tailored to exhibit additional strong I₂ capture mechanisms.

¹²⁹I is a major byproduct generated from nuclear fission of uranium fuel. Due to its adverse health effects in humans, safe removal and storage of ¹²⁹I is of utmost importance across various nuclear energy plants. The sorbents for the absorption of radioiodine has to be stable during the treatment process and also it should be capable of sorbing large amounts of ¹²⁹I. The most commonly used sorbents are silver-loaded zeolites and Ag-loaded silica aerogel. However, due to the poor mechanical stability of silica aerogels in an aqueous environment there is a need to develop new material with better mechanical stability. The chalcogen-based aerogels called “chalcogels” are highly porous and have showed good affinity towards heavy metal ions. Herein we report the use of chalcogels and silver functionalized analogues as host materials for capture and immobilization of ¹²⁹I. Iodine capture was studied with different chalcogels (Sb₄Sn₄S₁₂, Zn₂Sn₂S₆, NiMoS₄ and CoMoS₄), their silver functionalized analogues, and binary metal sulfides. All the chalcogels showed high uptake reaching up to 200 mass% and the iodine chemically reacted with the sorbents to form metal -iodide complexes. We will also report the consolidation of various iodine loaded chalcogels with different glass-forming additives into a final waste form.

Reprocessing of spent nuclear fuel is being considered in the U.S. In that case, the release of volatile ¹²⁹I from reprocessing plants and its safe storage would have to be controlled to meet the Environmental Protection Agency emissions regulations (which require capture of 99.4% of ¹²⁹I) and disposal restrictions. Currently, a silver-loaded zeolite (AgZ) is the baseline material for removing ¹²⁹I. However, recent studies indicate limitations in the sorption performance and long-term stability of AgZ. Also, AgZ requires addition of low-temperature glass to immobilize trapped radioiodine. To avoid these drawbacks, silver-functionalized silica aerogel is being developed for the efficient capture and immobilization of ¹²⁹I. This novel sorbent has a high affinity for iodine at the low concentrations expected in the off-gas and a high sorption capacity, and, after loading with iodine, it can be consolidated into a dense and leach-resistant SiO₂-based waste form. It was demonstrated to have a sorption capacity for I₂ of 480 mg/g, decontamination factors in excess of 10 000, good sorption performance after long-term exposures to dry and humid air, and retention of more than 92% of iodine in the densified product. The presentation will highlight the results from a series of sorption and consolidation studies.

Since 1991, the potential of several specific inorganic host matrices was studied at CEA to ensure a durable immobilization of iodine 129 in the frame of a possible disposal in a deep geological repository.
Due to evidence of retention of xenon 129, decay product of iodine 129, over geological time scales in apatites, these phases were the first materials to be considered. Specifically, a lead-bearing apatite with a good chemical durability was initially developed. Its composition can be written as Pb₁₀(VO₄)_4.8(PO₄)_1.2I₂. At 90°C, in pure water, its leach rate is of 2.28 10^-3 g.m².j^#8209;1 on the basis of iodine release and this rate decreases with time as the progressive replacement of iodide ions by hydroxyl groups along the channels of the crystalline structure occurs. This replacement obeys to a diffusive law and the transformation of iodoapatite grains into hydroxyapatite can be qualified as being pseudomorphic. Current studies are devoted to the shaping of such an iodoapatite in order to get a dense monolith. In so doing, it was found that a reactive sintering by spark plasma sintering at 400°C under a pressure range of 40-70 MPa could offer a clear benefit over sintering techniques in sealed environment (e.g., HIP) of which the use could be seen as more complicated for a reprocessing plant. This allows pellets of more than 92% of the theoretical density to be obtained. However, this process also appears to be very sensitive to scaling effects and it requires a subsequent optimization.
Other apatite compositions were also studied to avoid the use of toxic elements like lead. These apatites were developed on a phospho-calcic basis. They have the noticeable ability of incorporating iodine under its iodate form. Depending on phases constitutive of the geological barrier around the repository site, iodate ions could be less mobile in comparison with iodide which could delays the return of iodine to the biosphere. It was demonstrated that the incorporation mechanism of iodate into such an apatite relies on a substitution of hydroxyls groups. The chemical durability of this apatite is currently evaluated.
Together with ceramics, some glass compositions were also considered. They belong to the AgI-Ag₂O-P₂O₅-Al₂O₃ system. Close compositions were already proposed by Japanese teams for a similar goal. Here, the idea was to improve their properties by addition of a cross-linking reagent of the phosphate network like alumina. These glasses have an intrinsic compatibility with silver iodide which is the form adopted by iodine in most of the iodine capture processes on solid filters. They can incorporate high iodine amounts and their leaching behavior depends on phosphate chain length, iodine amount and alumina content.
Beyond the development of each matrix, the desired goal would be to have a correct opinion on the strengths and drawbacks of these solutions to face with future industrial and regulatory needs.

Apatite structure type, with a typical chemical composition of A₁₀(BO₄)₆C₂ (e.g. , A=Ca, Na, Pb, rare earth, fission product, actinides; B=P or V; C=F, Cl, I.) shows tremendous potentials as advanced waste forms for effective nuclear waste management. A wide range of radionuclides can be incorporated into its crystal structure by coupled substitutions at both cation and anion sublattices. Of particular importance, iodine-bearing apatite with chemical composition of Pb₁₀(VO₄)₆I₂ is proposed to confine extremely mobile and highly volatile I-129, a fission product of uranium fission. However, iodine-bearing apatite are typically synthesized and densified at elevated temperatures, resulting in evitable iodine loss. In this work, Pb₁₀(VO₄)₆I₂ powder samples are synthesized by solid state reaction at room temperature by using High energy ball milling (HEBM) followed by thermal annealing at 200 oC to control the crystallinity. Dense iodine-loaded apatite ceramic pellets were consolidated by state-of-art Spark plasma sintering (SPS) at various temperature (350 ^oC to 700 ^oC) and very short durations (0 ~20 mins). Iodine retention and the microstructure tenability, especially grain size, were investigated as functions of different SPS parameters (temperature, holding time, and pressure). No significant iodine loss was identified by high energy ball milling and during densification during SPS process. The thermal stability, thermal conductivity, and mechanical properties of the densified apatite pellets as durable iodine waste forms were studied. These results highlights that the SPS combining with high energy ball milling is a promising method to consolidate durable ceramic waste forms for confining highly volatile iodine for effective nuclear waste management.

The synthetic rock solidification is a HIPing technique to immobilize radioactive iodine (I-129) in the fuel reprocessing off-gas systems collected by silver nitrate impregnated onto an alumina base sorbent. Although iodine on the sorbent as a form of silver iodide (AgI) is unstable under the reducing condition, the α-alumina matrix of the synthetic rock is expected to fix the AgI physically in the grain boundary to be controlled iodine release after the geological disposal.
In the present study, the MCC-1 type immersion tests have been performed as a function of pH and hydrosulfide ion (HS^-) concentration as a reductant. The synthetic rock sample has been prepared by HIPing at 175MPa and 1473K for 3hours from a simulated spent sorbent saturated with stable iodine. Leached iodine has been under detection limit of an ICP-MS measurement below the HS^- concentration of 10^-5 M due to the stability of AgI itself. As the HS^- concentration of 10^-3 M, the iodine leaching rapidly increases within 100 days due to the AgI dissolution located at the surface and in open pore. The cross-section observation after immersion by EPMA and XRD suggests the following reaction: 2AgI + HS^- = Ag₂S + 2I^- + H⁺. Effect of pH has been clarified after 100 days that the both aluminum and iodine leaching decrease as pH decreases from 12.5 to 8. This indicates that the alumina matrix reasonably controls the iodine release. However the normalized leaching rate of iodine is 10 to 1000 times larger than that of aluminum. The incongruent leaching behavior would be due to the internal pore and grain boundary dissolution.
This research is a part of “Research and development of processing and disposal technique for TRU waste (FY2013)” under a grant from the Japanese Ministry of Economy, Trade and Industry (METI).

For safety assessment analyses of the disposal of spent nuclear fuel (SNF) in deep geological repositories it is indispensable to evaluate the contribution of activation and fission products to the instant release fraction (IRF). This fraction consists of soluble elements with low sorption tendency and contributes significantly to the calculated dose rates.
The IRF is controlled by the segregation of a part of the radionuclide inventory to the gap interface between the cladding and the pellet, to the fractures as well as to grain boundaries. The radionuclides that segregate are the fission gases (Kr and Xe), volatile radionuclides (³⁶Cl, ⁷⁹Se, ¹²⁹I, ¹³⁵Cs and ¹³⁷Cs) and metallic radionuclides (⁹⁹Tc and ¹²⁶Sn). The degree of segregation of the various radionuclides depends highly on in-reactor fuel operating parameters such as linear power rating, fuel temperature, burn-up, ramping processes and interim storage time. In the case of the fission gases, the gas release occurs by diffusion to grain boundaries, grain growth accompanied by grain boundary sweeping, gas bubble interlinkage and intersection of gas bubbles by cracks in the fuel.
During the last three years a EURATOM FP7 Collaborative Project, “Fast / Instant Release of Safety Relevant Radionuclides from Spent Nuclear Fuel (CP FIRST-Nuclides)” was carried out to get a better understanding of the IRF. Within CP FIRST-Nuclides, an irradiated UO₂ fuel pellet with cladding was sampled from a fuel rod segment with an average burn-up of 50.4 MWd/kg_HM. The cladded SNF pellet was leached in 19 mM NaCl + 1 mM NaHCO₃ solution under 40 bar Ar + H₂ atmosphere (p_H2: 4 bar). In the multi-sampling autoclave experiment, gaseous and liquid samples were taken periodically. The gaseous samples were analysed for fission gases by means of gas mass-spectrometry, the liquid samples were analysed for ¹²⁹I, ¹³⁷Cs and other dissolved radionuclides by means of gamma spectrometry, LSC and ICP-MS. In the present communications we focus on the behaviour of fission gases, ¹²⁹I and ¹³⁷Cs as a function of leaching time. After 177 days of leaching experiment, the percentage of fraction of the inventory released into the gaseous and aqueous phases was: 12.8 for the fission gases (Kr + Xe), 9.2 for ¹²⁹I and 2.8 for ¹³⁷Cs, respectively.

MoO₃, which can be found at elevated levels in some high level nuclear waste streams in the UK and France, is one of the most challenging oxides to immobilise in the borosilicate glasses conventionally used for nuclear waste vitrification. MoO₃ usually has a low solubility (le;1 mol%) in silicate glasses and excess MoO₃ in nuclear glass causes the formation of “yellow phase” which is detrimental to the vitrification process. Glass compositions with greater MoO₃ solubility limits are therefore desirable. In this work the solubility and incorporation of MoO₃ in an alkaline earth aluminosilicate glass system (AeAS, Ae = Mg, Ca, Sr, Ba or two of these combined) have been investigated, showing that MoO₃ solubility steadily increases in the order Ba < Sr < Ca < Mg. Up to 5.3 mol% (12.3 wt%) MoO₃ can be retained in magnesium aluminosilicate glass (MAS) without phase separation while only 2.0 mol% (2.5 wt%) MoO₃ can be completely dissolved in barium aluminosilicate glass (BAS). The high MoO₃ solubility in MAS glass provides the possibility of using it as an alternative wasteform for the vitrification of nuclear waste containing high levels of MoO₃. The changes in glass structure and properties caused by MoO₃ incorporation are also assessed. Glass density increases whereas glass transition and crystallisation temperatures decrease with increasing MoO₃ addition. The prepared glasses reveal good thermal stability until glass transition. All visibly homogeneous glasses are X-ray amorphous while the partially crystallised glasses exhibit some small X-ray diffraction peaks which are probably due to corresponding molybdates. In Raman spectra, MoO₃ addition contributes two broad bands which are assigned to vibrations of MoO₄^2#8210; tetrahedra. The intensities of these bands increase along with MoO₃ incorporation until saturation. In the Raman spectra of partly crystallised glasses with combined alkaline earths, the crystalline bands are in accordance with the molybdate with lowest solubility whenever possible, indicating that MoO₃ solubility in glass is controlled by the cation whose molybdate salt has the highest crystallisation tendency. Electron microscopy shows that these separated particles are spherical, with sub-micron diameters and are randomly dispersed within glass. The separated phases are formed through liquid-liquid separation and thereafter crystallisation. Overall AeAS glasses look quite promising for molybdate immobilisation with MAS glasses being particularly attractive.

Currently in Russia some compositions of spent nuclear fuels (SNF) such as molybdenum-bearing SNF of uranium-graphite reactors (AMB) are not reprocessed yet but their reprocessing is under consideration now. High level waste (HLW) from AMB SNF reprocessing is suggested to be incorporate in sodium aluminophosphate (SAP) based glass similarly to different HLW. Mo is one of the troublesome components of HLW causing liquid/liquid phase separation in borosilicate glasses and crystallization of phosphate glasses and reduction of chemical durability of vitrified waste. Therefore the effect of Mo solubility, its valence state and speciation on chemical durability of glasses has to be studied. Incorporation of Mo in SAP glass favors its crystallization and annealing increases the degree of crystallinity. Valence state and local environment of Mo in the materials containing ~2 wt.% MoO₃ were characterized by X-ray absorption fine structure (XAFS). In the quenched samples composed of major vitreous and minor AlPO₄ crystalline phase nearly all Mo is located in the vitreous phase as [Mo⁶⁺#1054;₆] units whereas in the annealed samples Mo is partitioned among vitreous and one or two orthophosphate crystalline phases. The spectra of annealed markedly crystallized samples contain weak response in pre-edge range which can be assigned to the line due to contribution of [#1052;#1086;⁶⁺#1054;₄] units located in the crystalline phase. Mo predominantly exists in a hexavalent state in distorted octahedral environment. Three oxygen ions are positioned at a distance of ~1.70 Å and three - at a distance of ~2.04 Å. In the highly-crystalline annealed samples especially contained 5.4 wt.% MgO the best fit is achieved on the assumption of three different Mo-O distances: 2.5-3 oxygen ions are positioned at a distance of ~1.73 Å, 2.3-3 oxygen ions - at a distance of ~2.05 Å and 1-1.5 oxygen ions - at a distance of ~2.3 Å from Mo⁶⁺ ions. This may be attributed to contribution due to minor Mo in complex orthophosphate. Formation of Mo-bearing phosphates could be the reason of deteriorating of chemical durability of the materials especially after their annealing resulting in increase of the degree of crystallinity.

The Hanford site in southeastern Washington State is the largest repository of nuclear waste in the United States, where 177 underground tanks stored in excess of 50 million gallons of waste. This waste will be mixed with glass-forming additives and vitrified at the Waste Treatment and Immobilization Plant (WTP). A large fraction of the anticipated waste streams are simultaneously high in both Na and Al, leading to frequency crystallization of aluminosilicate phases such as nepheline upon cooling. This aluminosilicate crystallization has been previously shown to be deleterious to chemical durability due to the extraction of alumina and silica from the glass-forming matrix, leaving a residual glass of less durable components. The long-term corrosion resistance is of significance because vitrified waste must tolerate subterranean storage for ge;10⁶ years.
A challenge in the formulation of nuclear waste glasses arises from maintaining a sufficiently low addition of glass-forming additives to maximize waste loading, while ensuring that the addition is sufficiently high to prevent crystallization. In this study a glass composition, designated as A4, was selected due to its particular crystallization behavior. This composition was formulated for a high-alumina waste stream (>25 wt% Al₂O₃) with 45 wt% waste loading. A4 glass was batched from powder precursors and subjected to air quenching, isothermal heat treatments, and canister-centerline cooling (CCC) to observe the crystallization behavior, with the goal of obtaining time-temperature-transformation curves.
Crystal fractions were obtained by x-ray diffraction (XRD) and crystallite structure was observed through scanning electron microscopy (SEM) with composition measured through wavelength dispersive spectroscopy (WDS) and energy dispersive spectroscopy (EDS). Nepheline, a feldspathoid with sodium end-member composition NaAlSiO₄, is the prominent aluminosilicate phase in the CCC high-alumina waste glass, but it forms in unusually large isolated dendritic crystals in the presence of a complex assemblage of crystals of phosphate, spinel, and a residual glass system enriched in Ca, Mg, Zr, and B.
The ultimate goal of this ongoing work is to obtain a kinetic model for crystallization of nepheline and enable compositional design to inhibit rapid crystallization of nepheline in high Na and Al wastes.

To limit foaming and improve the melting rate of simulated high-level waste melter feeds during the vitrification process, the importance of silica grain size was investigated. A high-alumina high-level waste melter feed formulated by Vitreous State Laboratory was prepared using various silica grain sizes. Feed samples were heated at 5°C/min up to 1200°C. To observe volume expansion, feed pellets were created and photographed during heating. Quenched samples from heat treatments were prepared for scanning electron microscopy and crystalline phases were determined with X-ray diffraction. By eliminating fine particles contained in most silica sources, the volume expansion caused by foaming was decreased due to a delay in silica dissolution that improved the overall melting rate.

To efficiently vitrify Hanford waste, the melting process (i.e., melter feed turning into waste glass) must be modeled and optimized. The rate of heat transfer to the melter feed in a waste glass melter, and thus the rate of melting, is strongly affected by the melter feed porosity, especially in the final stages where the glass-forming melt produces foam that insulates the feed from the molten glass. The volume expansion test allows the determination of the melter feed porosity as a function of temperature. This test measures the profile area of the feed pellet as it turns into glass. This contribution presents the calculation of the void fraction (porosity) of the melter feed as a function of temperature, heating rate, and material parameters. The process of finding the void fraction is described as well as results from the application of this process.

Nepheline, (Na,K)AlSiO₄, is an aluminosilicate mineral that can crystallize in waste glass containing a high fraction of Al₂O₃ and Na₂O when it is slow cooled from a glass melt. The formation of nepheline alters the residual glass composition and lowers durability by taking key glass-forming components of alumina and silica out of the glass. Understanding what glass compositions limit or encourage nepheline formation and having a model to predict nepheline formation in waste glasses is critical to achieve the maximum waste loading. Early ternary models only included Na₂O, Al₂O₃, and SiO₂ as predictive variables for nepheline formation and were too conservative. A new model includes Na₂O, Li₂O, CaO, Al₂O₃, B₂O₃, and SiO₂ and appears to show a clear delineation in the data between glasses that do and do not form nepheline upon slow cooling. Details for how this model was constructed and future glass formulation work to evaluate the model will be discussed.

The Department of Energy-Office of River Protection (DOE-ORP) is constructing the Hanford Tank Waste Treatment and Immobilization Plant (WTP) to treat radioactive waste currently stored in underground tanks at the Hanford site in Washington. The WTP that is being designed and constructed by a team led by Bechtel National, Inc. (BNI) will separate the tank waste into High Level Waste (HLW) and Low Activity Waste (LAW) fractions with the majority of the mass (~90%) directed to LAW and most of the activity (>95%) directed to HLW. The pretreatment process, envisioned in the baseline, involves the dissolution of aluminum-bearing solids so as to allow the aluminum salts to be processed through the cesium ion exchange and report to the LAW Facility. There is an oxidative leaching process to affect a similar outcome for chromium-bearing wastes. Both of these unit operations were advanced to accommodate shortcomings in glass formulation for HLW inventories. A by-product of this are a series of technical challenges placed upon materials selected for the processing vessels. There exist additional questions on the adequacy of the design to ensure nuclear safety requirements are met during across the entire spectrum of possible operating conditions.
At the heart of the treatment process is the vitrification of the HLW and LAW waste streams in Joule Heated Ceramic Melters (JHCMs). The JHCM is typically operated at a melt pool temperature of 1150°C. The slurry feed is introduced from the top of the melter and during operation the melt pool is almost entirely covered with unmelted feed termed the cold cap. The Hanford JHCMs are fitted with a patented bubbler system to agitate the melt pool, thus improving heat transfer to the cold cap and, therefore, feed processing rate. The Office of River Protection undertook an extensive investigation focused upon glass formulation improvements and enhancements of operating efficiencies in the vitrification facilities. The outcomes have the potential of profound, but positive, impacts on the baseline flow sheet for the Pretreatment Facility. Not forgetting the impact on more rapid realization of successfully emptying the waste tanks and treating the waste.
The WTP will process and treat approximately 53 million gallons of mixed hazardous wastes (i.e., radioactive and chemical waste). This presentation provides an overview of the project status and technical challenges facing the process, design, and construction of the WTP facilities.

The storage of radioactive waste is an on-going problem around the world. Yucca Mountain with its tuff and granite rocks was declined by the US government. In Germany the salt stocks are again in discussion and the exploration of a suitable repository is starting over again. In discussion are clay deposits - which e.g. Switzerland favors - or granite or other rock material.
Several factors have to be considered for the suitability of such places: geo-mechanical behavior, fluid-tightness, no earthquake region etc. As the nuclear waste might heat up the surroundings, the thermal conductivity of the surrounding materials should be high enough to avoid too much accumulation of the waste heat.
Another question is what happens to the repository material when heated up accidentally. In this contribution, granite, rock salt and clays were investigated by TG-DSC, dilatometry, LFA and evolved gas analysis and results for thermal stability, thermal expansion and thermal diffusivity will be reported.

High-level waste (HLW) from reprocessing of spent nuclear fuel of uranium-graphite reactors (Russian AMB type) is suggested to be vitrified with production of aluminophosphate based glass similarly to current PWR type (Russian WWER). Some of the AMB fuel compositions contain copper and therefore behavior of copper ions in sodium alumonophosphate glasses has to be investigated. Target glass formulations contained ~2.4-2.5 mol.% CuO. The mixtures of chemicals were dried in a dessicator, placed in Pt crucibles, heated to 1000 °C, kept at this temperature for 0.5 hr, and melts were poured onto a stainless steel plate followed by annealing of the materials at a temperature of 500 °C for 14 hrs. The quenched sample was composed of major amorphous phase and minor aluminum orthophosphate (20-30 vol.% of total). The quenched MgO bearing (13.2 mol.%) sample was predominantly amorphous (<5 vol.% AlPO₄). The annealed MgO free sample had higher degree of crystallinity than the annealed MgO-bearing sample but both them contained orthophosphate phases. Cu in the materials was partitioned in favor of the vitreous phase. In all the samples copper is present as major Cu(II) and minor Cu (I) forms. Cu²⁺ ions form planar square complexes (CN=4) with a Cu²⁺-O distance of 1,93-1,95 Å. Two more ions are positioned at a distance of 2,76-2,86 Å from Cu²⁺ ions. So the Cu²⁺ environment looks like a strongly elongated octahedron as it also follows from the absence of the pre-edge peak due to 1s→3d transition in Cu K edge XANES spectra of the materials. Cu⁺ ions form two collinear bonds at Cu⁺-O distances of 1,80-1,85 Å. Thus average Cu coordination number (CN) in the first shell was found to be 2.7-3.0.

Until the early 1970s, borosilicate glass was the reference waste form for immobilizing high-level (reprocessing) nuclear waste. But in the early 1970s, Penn State University (PSU) researchers (Rustum Roy, Will White and Greg McCarthy) introduced the potential use of synthetic minerals which were known to be leach resistant in hot, wet conditions through geological studies and could incorporate key fission product elements in their crystal lattices. These minerals were phosphates and silicates and produced by standard ceramic methods. Ted Ringwood at the Australian National University soon became aware of the PSU work and put forward assemblages of titanate minerals for the same purpose and showed them to be much more durable in water than the phosphates and silicates (and borosilicate glass). Bob Dosch at Sandia also worked on titanate ion exchangers for the same reasons. This presentation will highlight the early ground breaking work which still significantly influences current waste form proposals.

The disposal of high level radioactive waste is one of the most pressing and demanding challenges. With respect to long-term safety aspects of geological disposal, the minor actinides (MA) such as Am, Cm and Np and long-lived fission products such as ³⁵Cl, ¹³⁵Cs, ⁷⁹Se, ⁹⁰Sr and ¹²⁹I may be of particular concern due to their long half-lifes, their high radiotoxicity and mobility in a repository system, respectively. Ceramic waste forms for the immobilisation of these radionuclides have been investigated extensively in the last decades since they seem to exhibit certain advantages compared to other waste forms (incl. borosilicate glasses and spent fuel) such as high loadings and chemical durability. Currently, most on-going nuclear waste management strategies do not include ceramic waste forms. However, it is still important to study this option, e.g. with respect to specific waste streams and certain constraints regarding deep geological disposal.
In the present communication we report on the research program in Jülich regarding ceramic waste forms for the conditioning of MA. It is based on fundamental science and follows an integral approach that covers the separation of elements or elemental groups with similar chemical properties from a waste stream by liquid/liquid extraction as well as the immobilization in ceramic materials as hosts. Various aspects with the focus on single phase waste forms, such as monazite and zirconates with pyrochlore structure will be discussed:
1.) Development and optimisation of synthesis routes suitable for immobilisation of MA into ceramic waste forms and the handling of radionuclides such as sol-gel route, hydrothermal synthesis and co-precipitation,
2.) structural and microstructural characterisation using state of the art spectroscopic (Raman, TRLFS, EXAFS), diffraction (powder and single crystal XRD) and microscopic (SEM, FIB/TEM) techniques,
3.) determination of thermodynamic data (calorimetry) and reactivity under conditions relevant to geological disposal, in particular with respect to dissolution in aqueous environments (static & dynamic dissolution experiments on powders and pellets) as well as
4.) studies on radiation damages (irradiation with α-particles and/or heavy ions, and incorporation of short-lived actinides such as ²³⁸Pu, ²⁴¹Am or ²⁴⁴Cm).
Finally, a fundamental understanding of the long-term behaviour on the atomic scale will help to improve the scientific basis for the safety case of deep geological disposal concepts using ceramic materials.

The complexity and diversity in the chemistry of legacy UK radioactive wastes has necessitated the development and validation of a toolkit of advanced wasteforms. This approach will expand the range of materials to be consigned to a future geological disposal facility, to include new glass, ceramic, and cement wasteforms. This presentation will examine recent advances in the fundamental understanding of these wasteforms and their interaction with conceptual disposal environments, to support the disposal system safety case, including:
* The design, processing and disposability of glass and ceramic products from thermal treatment of plutonium and intermediate level wastes; where we have achieved: control over partitioning of radionuclides between component phases, an understanding of mechanisms of wasteform alteration in the hyperalkaline environment of a cementitious GDF; and the mechanism of the crystalline to amorphous phase transition induced by alpha recoil damage.
* The development of new low pH potassium magnesium phosphate cement systems suitable for encapsulation of reactive metals; where we have achieved a state of the art understanding of the mechanisms of binder phase formation, its impact on the product mechanical properties, radiation stability, and interaction with reactive metals, leading to hydrogen production.
* The application of advanced radio-imaging methodology for determining radionuclide transport in cement backfill, and mechanistic understanding of the immobilisation of problematic radionuclides in new functional cement barrier materials

From X-ray diffraction, scanning electron microscopy, and X-ray near edge structure and diffuse reflectance spectroscopic studies, it was found that approximately 0.03 formula units of mixed hexavalent and pentavalent U, but <0.001 formula units of Pu or Np if any, can be dilutely incorporated into the Ti site of rutile, TiO₂, sintered at 1400^oC in air. The valence of 0.01 formula units (f.u.) of U in Zr_(1-x)Y_xO_1-x/2 (x = 0.23-0.5) fired in air is mainly +6 from X-ray near edge spectroscopy and contradicts earlier X-ray photoelectron spectroscopy data. Some U⁵⁺ is also present from diffuse reflectance work. The valence of 0.01 f.u. of Np in the Zr_(1-x)Y_xO_1-x/2 (x = 0.23-0.5) is found to be +6 from diffuse reflectance study.

Sodalite (Na₈[AlSiO₄]₆Cl₂), a naturally occurring Cl-containing mineral, has long been regarded as a potential immobilisation matrix for the chloride salt wastes arising from pyrochemical reprocessing operations, as it allows for the conditioning of the waste salt as a whole without the need for any pre-treatment. Here the consolidation and densification of Sm-doped sodalite (as an analogue for AnCl₃) has been investigated with the aim of producing fully dense (i.e. > 95 % t.d.) ceramic monoliths via conventional cold-press-and-sinter techniques at temperatures of < 1000 °C. Microstructural analysis of pressed and sintered sodalite powders under these conditions is shown to produce poorly sintered, porous, inhomogeneous pellets. However, by the addition of a sodium aluminophosphate glass sintering aid, fully dense Sm-sodalite ceramic monoliths can successfully be produced by sintering at temperatures as low as 800 °C.

Multiphase ceramic waste forms show promise as a viable alternative to borosilicate glasses for nuclear waste treatment and disposal. These waste forms include hollandite, perovskite, pyrochlore, and zirconolite phases. We synthesized high phase purity pyrochlore (Nd₂Ti₂O₇) via. solid-state reaction. Praseodymium (Pr) and samarium (Sm) were chosen as dopants for our study. These dopants were added from x = 0.1-0.5 according to Nd_2-x(Pr,Sm)_xTi₂O₇. X-ray diffraction (XRD) was used to confirm phases present in the samples and determine lattice parameters. Scanning electron microscopy (SEM) and energy-dispersive spectrometer (EDS) were used to characterize the powders and morphology. SEM micrographs of fractured samples confirm the presence of single, homogenous phase and agree with the XRD and EDS data. XRD data showed the solid solution limit was not yet reached in this system. Lattice volume increased linearly with dopant concentration. Changes in the lattice parameters suggested expansion of the unit cell of the Pr-doped Nd₂Ti₂O₇ in the a and b directions. Selected samples were sintered via spark plasma sintering (SPS) to dense microstructures, which were examined using SEM and EDS. High-density values of about 95-98% of theoretical density were obtained for comparable grain size. Our results show the doping did not impact sinterability of the pyrochlore.

Lanthanum zirconate pyrochlore (La₂Zr₂O₇) have been extensively studied as a model for radioactive waste hosts, primarily due to its ability to recover from heavy ion radiation damage [1]. However, compartively little work has been undertaken examining the stability when substituted by radioactive elements, e.g. U/Pu.
Characterisation of uranium substitution based on the general formulation La₂Zr_2-xU_xO_7+δ , in both air and reducing H₂/N₂ atmospheres has been completed. The stability in both air and H₂/N₂ atmospheres has been shown by X-ray diffraction to be x = 0.8 in H₂/N₂ and 0.6 in air. At these upper limits of the substitution electron diffraction has been used to confirm the existence of the pryochlore superstructure, along with determining the presence of any other ordering. For example, in the H₂/N₂ samples the electron diffraction shows the presence of pyrochlore superstructure at x = 0.8, whereas X-ray diffraction indicates a fluorite structure has been formed. This difference is discussed with respect to both electron and X-ray diffraction, coupled with Raman spectroscopy, and the implications for use of this material as a host matrix for the stable storage of U/Pu.
References:
[1] G.R. Lumpkin et al. J. Phys.: Condens. Matter, 16 (2004) 8557

Pyrochlores (A₂B₂O₇) have been studied extensively for the immobilization of actinides. ^{1,2, 3} This project focuses on studying the effects of helium gas build-up and radiation damage due to alpha decay in ceramic nuclear waste form materials, specifically Gd₂Ti₂O₇ and Gd₂Zr₂O₇. Alpha decay in waste forms produces both helium atoms and atomic displacements that can be replicated using ion implantation and ion-beam irradiation techniques. The accumulation of helium can result in helium platelets and bubbles, and the accumulation of defects can result in the formation of dislocation loops, bubbles and phase transformations. It is well known that Gd₂Ti₂O₇ undergoes a pyrochlore to amorphous transformation at ~0.2 dpa at room temperature and that Gd₂Zr₂O₇ undergoes an order-to-disorder transformation from the pyrochlore to defect fluorite structure at ~0.4 dpa at low temperatures. ¹ Gd₂Ti₂O₇ and Gd₂Zr₂O₇ were synthesized by solid-state synthesis. These materials were irradiated with 7 MeV Au ions at room temperature to create a thick (~1 micron) amorphous state in Gd₂Ti₂O₇ and transform Gd₂Zr₂O₇ to the defect fluorite structure (~1 micron thickness), which are the structures of interest for long-term evaluation. These samples were subsequently implanted with 200 keV He ions to fluences of 2x10¹⁵ and 2x10¹⁶ ions/cm² at room temperature, which at the implantation peak correspond to expected He concentrations at 1000 and 100,000 years, respectively. The helium implanted samples have been irradiated with 7 Mev Au ions slightly evaluated temperatures to radiation doses corresponding to 50,000 years or more. The higher irradiation temperatures are used to accelerate the defect and helium interaction kinetics, simulating irradiation-induced microstructure evolution at longer time scales. After irradiation, the evolution of microstructure (dislocation loops, bubbles, etc.) were characterized using transmission electron microscopoy (TEM), scanning electron microscopy (SEM), and x-ray diffraction (XRD) techniques. The results of this study will be discussed.
1. Wang, S.-X. et al. Radiation stability of gadolinium zirconate: a waste form for plutonium disposition. Journal of materials research14, 4470-4473 (1999).
2. Lian, J. et al. Radiation-induced amorphization of rare-earth titanate pyrochlores. Physical Review B68, 134107 (2003).
3. Ewing, R.C., Weber, W.J. & Lian, J. Nuclear waste disposal—pyrochlore (A2B2O7): Nuclear waste form for the immobilization of plutonium and “minor” actinides. Journal of Applied Physics95, 5949-5971 (2004).

The immobilisation of actinides within the crystalline structure of ceramic waste forms seems to offer certain advantages over other waste forms (incl. borosilicate glasses and spent fuel). Monazite, LnPO₄ (Ln=La-Gd) is a promising ceramic as a waste form for actinides related to long-term safety aspects. For a reliable assessment of their long-term stabilty under conditions relevant to nuclear waste disposal deeper fundamental studies on these materials are necessary.
In the present communication we report on the atomic scale investigation of solid solution formation and on microstructural studies of the synthesis dependent sintering behaviour of monazite-type ceramics. A combined understanding concerning the structural and microstructural properties is of a great importance with regard to key parameters guiding the long-term stability of ceramic materials for safe nuclear disposal. Cluster formation in non ideal solid solutions in the case of minor actinides immobilisation could influence irradiation damages resistance, criticality aspects and dissolution behaviour. Microstructure impacts mechanical properties and corrosion resistance. Certain porosity in accordance with waste loading degree is needed to avoid crack formation due to swelling processes caused by He-evolution from α-decay reaction.
In recent studies we investigated the structural incorporation of Eu(III) in synthetic Eu(III) doped LaPO₄, and GdPO₄, as well as mixtures thereof by site-selective time-resolved laser fluorescence spectroscopy (TRLFS). Eu(III) was taken as an analogue for the long-lived trivalent actinides Pu(III), Am(III) and Cm(III) found in spent nuclear fuel. In the pure LaPO₄ and GdPO₄ monazites, Eu³⁺ substitutes the host cation sites in the highly ordered ceramic materials independent of the ionic radius of the host cation. However, excitation spectra of the mixed Eu(III)-doped (La,Gd)PO₄ monazite phases indicate a slight disordering of the crystal structure.
Additionally the present work was focused on the elaboration of (La,Eu)PO₄ solid solutions through wet chemistry routes then on the study of their densification. In this aim, investigations on in-situ sintering phenomena were carried out by the joint use of dilatometry and high temperature environmental scanning electron microscopy. Particularly, it allowed us to precise the conditions required for the densification of monazite but also to provide new insights on the microstructure development during heat treatment, including grain growth rate.

Coffinite (USiO₄) and associated solid solutions are expected to play an important role in the field of direct storage of spent nuclear fuels in underground repository since they could control the concentration of actinides in groundwaters. However, the thermodynamic properties associated with coffinite, especially the solubility, remain poorly defined. The few thermodynamic data related to coffinite formation or solubility reported in the literature are hardly reliable since none of them were determined experimentally from solubility measurements [1]-[3]. Solubility studies require pure single-phase USiO₄. Most of the natural samples contain coffinite as very small grain crystals [4] and in intimate intergrowths with large amounts of associated minerals. Moreover, for several decades persistent difficulties have been encountered in the preparation of pure single-phase synthetic coffinite. The precipitation of coffinite from a mixture of U(IV)-containing acidic solution and sodium metasilicate appeared as the most promising method to provide USiO₄ samples [5]. In this context, a thorough multiparametric study of the formation of synthetic coffinite was achieved. In this aim, the effect of various parameters such as pH, heating time, U/Si mole ratio and temperature were investigated to point out the optimal operating conditions for the preparation of coffinite.
The optimized protocol allowed the preparation of polyphased samples that contained mainly USiO₄ associated with oxide side products (amorphous SiO₂ and nanoparticles of UO₂). A purification process was developed that conduct to pure synthetic coffinite sample suitable for solubility experiments. The ion activity product in solution equilibrated with USiO₄ was determined by dissolution experiments conducted in 0.1 mol L^-1 HCl under Ar atmosphere at room temperature. The dissolution was congruent and a constant composition of the aqueous solution was reached after 50 day. The solubility product of coffinite was then determined (log*K_S,USiO₄ (298 K) = minus;6.14 ± 0.08). At low temperatures, coffinite appears to be less stable than the mixture of binary oxides, which is consistent with qualitative evidence from petrographic studies of uranium ore deposits. Finally a tentative mechanism was proposed to explain the formation of USiO₄ providing new insights concerning the formation of coffinite in environmental conditions.
[1] Guillaumont, R.; Fanghänel, T.; Fuger, J.; Grenthe, I.; Neck, V.; Palmer, D. A.; Rand, M. H., Chemical Thermodynamics Vol. 5. North Holland Elsevier Science Publishers B.V.: Amsterdam, The Netherlands, 2003, p 919. [2] Langmuir, D., Geochim. Cosmochim. A. 1978, 42, 547-569. [3] Hemingway, B. S., USGS Open file Report 82-619, 1982, p 89. [4] Deditius, A. P.; Utsunomiya, S.; Ewing, R. C., Chem. Geol. 2008, 251, 33-49. [5] Costin, D. T.; Mesbah, A.; Clavier, N.; Dacheux, N.; Poinssot, C.; Szenknect, S.; Ravaux, J., Inorg. Chem. 2011, 50, 11117-11126.