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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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