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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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