In a deep geological repository for high-level radioactive waste in crystalline rock, the containment of radionuclides can potentially not be guaranteed by the host rock itself. In this case, according to German law (§ 23 (1) and (4) StandAG) in crystalline host rock the safe containment of the nuclear waste for one million years must be guaranteed by the engineered and geo-engineered barriers. Therefore, in such a setting the requirements on the integrity of the waste canisters are specifically high. Hence, it is of great importance to evaluate and understand the processes that might lead to a corrosion of the canister, resulting in a potential loss of their integrity.

The present study compiles the state of knowledge regarding different canister concepts, considered materials and relevant corrosion mechanisms, taking into account the specific premises that are associated with a deep geological disposal site in crystalline host rock in Germany. This includes the evaluation of the hydrogeochemical conditions in the crystalline basement. The compiled hydrogeochemical data are compared with crystalline host rocks in other countries and are interpreted regarding implications for the chemical integrity of the canisters. Additionally, a review of the state of knowledge concerning modelling of canister corrosion is conducted. It assesses whether existing modelling approaches are applicable for simulating the predominant canister corrosion mechanism and which further developments might be necessary.

Altogether, this study evaluates the actual understanding of mechanisms of and controls on container corrosion processes, displaying the scientific basis for the identification of requirements on the design and material properties of canisters for the safe disposal of high-level radioactive waste in crystalline rock in Germany.

The safe and secure management of high-level radioactive waste is a critical issue for any country that generates nuclear power. In many countries including South Korea, direct disposal of high-level radioactive waste has been widely selected as a final option for the radioactive waste management. Because insufficient regulatory requirements can disrupt confusion in research and industry, regulations need to be reviewed and developed. In South Korea, a plan of national policy for the deep geological disposal has been newly suggested. In this study, we describe the multi-ministerial project that implemented by Korean ministries including regulation, research and industry for disposal of high-level radioactive waste. This includes an overview and the development of the regulatory frameworks that support these project. We also discuss the challenges and opportunities associated with the regulatory framework and provide examples in terms of site development and natural barrier. As a result, Draft regulatory requirements of both site development and natural barrier for deep geological disposal are proposed. This includes an overview of the regulatory requirements and the criteria that have been proposed for site development, safety assessment and design of the disposal system. This work highlights the importance of a comprehensive regulatory framework for the safe and secure management of high-level radioactive waste and provides insights for policymakers and practitioners who are interested in developing regulatory frameworks.

In Japan, shallow land disposal sites have already been and may be constructed in the future on marine terraces. It has been reported that deep-seated landslides, which are rapid erosions on terrace cliffs and hillslopes of streams formed on terrace faces, are dominant in the erosion of marine terraces. Therefore, the ability to handle deep-seated landslides is necessary to evaluate the effects of erosion on disposal sites during long-term topographic change. The Landlab code can evaluate deep-seated landslides as well as gradual erosion, mainly on hillslopes in mountainous areas. However, its applicability to marine terraces needs to be confirmed. In order to confirm the applicability of the evaluation model for deep-seated landslide of the Landlab to marine terraces, we first evaluated the occurrence points of deep-seated landslides using the Landlab with 2m DEM for an area with marine terraces. We then compared the points extracted by the Landlab with the hillslopes extracted by the manuals to determine the applicability of the Landlab’s evaluation model. The target area was selected as an area with evidence of deep-seated landslides on marine terraces and streams. The points where deep-seated landslides are likely to occur in the future were evaluated for the field profile using the 2m DEM. The two manuals target the field profile and extract hillslopes where deep-seated landslides are likely to occur in the future based on the characteristics of the slope distribution considering the past landslide and topographic quantities in the same area using 2mDEM. Comparison of the extracted results of deep-seated landslides shows that most of the extracted points of deep-seated landslides by the Landlab were included in the possible deep-seated hillslopes extracted by the two manual methods. The results confirm that the Landlab’s evaluation model is capable of extracting points where deep-seated landslides are likely to occur in the future.

In the safeguards laboratories at Forschungszentrum Jülich an aerosol-based process was established to produce microparticulate reference materials for the International Atomic Energy Agency (IAEA). These reference materials are required for quality control of mass spectrometric verification measurements. One essential part in the particle production process is the drying of the aerosol droplets forming the microparticulate uranium oxide precursor and the subsequent thermal transformation into UO₃ and U₃O₈. Since a water/ethanol (50:50) mixture is used in the process to produce the aerosol, drying experiments of uranyl nitrate with different dopants in either water or ethanol were performed. In particular, ethanol was studied here in order to simulate the impact of the reducing properties of ethanol on the chemistry of the microparticles during the aerosol synthesis process. On the one hand, a series of lanthanides such as La, Gd and Lu were used as dopants to investigate the influence of the ionic radius on the material properties (e.g., crystal structure). On the other hand, Th was employed as dopant to identify material properties that could have an impact on the possible application in age dating measurements. This poster will show first results regarding the investigation on these synthesized materials, which were measured by TG-DSC, subsequently calcined and then structurally analyzed by XRD, Raman and IR spectroscopy.

This contribution provides an overview of a current research network funded by the German Federal Ministry of Education and Research (BMBF), entitled “Fundamental investigations of actinide immobilization by incorporation into solid phases relevant for final disposal” – AcE. The AcE project aims at understanding the incorporation and immobilization of actinides (An) in crystalline, repository-relevant solid phases, such as zirconia (ZrO₂) and UO₂, but also in zircon (ZrSiO₄), pyrochlores (Ln₂Zr₂O₇) and orthophosphates of the monazite type (LnPO₄), which may find use as host matrices for the immobilization and safe disposal of high-level waste streams.

Recent studies by the AcE-project consortium, addressing the structure, properties, and the radiation tolerance of monazites and Zr(IV)-based solid phases containing actinides or their surrogates from the lanthanide series will be presented. Material synthesis strategies in the AcE project have aimed at generating single-phase solid solutions in the form of polycrystalline powders, dense ceramics, and single crystals. Structural studies using powder X-ray diffraction at ambient conditions, but also at high temperatures and pressures have been complemented with a wide range of microscopic and spectroscopic techniques to address differences between the host- and dopant environments in the solid matrices at ambient and extreme conditions. The radiation tolerance of the synthetic solid phases have been investigated by combining external heavy-ion irradiation of inactive Ln-doped materials and in situ self-irradiation of ²⁴¹Am-doped Zr(IV)-phases with monoclinic, cubic defect fluorite and pyrochlore structures. The latter experiments have been conducted in joint efforts with the Joint Research Center in Karlsruhe within the ActUsLab programme.

The United Kingdom holds a substantial inventory of PuO₂, forecast to reach approximately 140 teHM (tonnes equivalent heavy metal) upon completion of reprocessing. This material presents a unique decommissioning prospect for which there is a need to develop a robust management strategy. Prompt immobilisation and disposal within a geological disposal facility (GDF) is a promising route towards ultimate disposition, yet in order to safely underpin the safety case for the geological disposal of Pu, it is necessary to understand the long term evolution of candidate wasteform materials in simulated repository environments. Moreover, there is a need to develop suitable wasteform materials capable of co-accommodating Pu, prescribed quantities of neutron poisoning species, trace processing impurities and transition metal cations capable of providing charge balance for non-stoichiometric compositions. Several baseline wasteform formulations derived from zirconolite, pyrochlore and fluorite-type matrices have been proposed on the basis of high chemical durability, radiation stability and moderate ease of processing. Herein, this talk will provide an overview in recent advances in the formulation refinement and fundamental characterisation of candidate wasteform materials for UK Pu. This includes detailed scoping trials aiming to characterise the incorporation of a representative U, Th and Ce surrogate fraction within zirconolite and pyrochlore phases, fabricated by conventional sintering (CPS), hot isostatic pressing (HIP) and reactive spark plasma sintering (RSPS).

The safe disposal of nuclear waste is one of the intergenerational issues which needs to be solved. A potential route to effectively immobilize radionuclides could be realized by their incorporation into crystalline solid phases in future radioactive waste repositories. In particular, the immobilization of specific waste streams containing minor actinides (Np, Am, Cm) or plutonium in crystalline solid phases may be advantageous compared to glass matrices, which may be less resistant to leaching and disintegration [1-3]. Due to their radiation stability and chemical and structural flexibility, monazite-type compounds are considered suitable matrix materials [4]. To better understand structural changes due to radiation damage, synthetic monazite single crystals with different chemical compositions (La, Nd, Pm, Sm)PO₄ were synthesized by a high-temperature (flux method). Irradiation experiments were performed at the UNILAC beamline of GSI Helmholtz-Centre Darmstadt using 1.7 GeV Au ions and fluences of up to 1e13 ions/cm². The irradiated single crystals were characterized by Raman spectroscopy, secondary electron microscopy and single crystal X-ray diffraction. The irradiation of monazite with 1.7 GeV Au ions results in an embrittlement of the crystals and the formation of a glassy surface layer of about ~48 μm thickness, which correlates well with the projected range of ~44 µm according to SRIM-2013 calculations [5]. The irradiation results in a significant broadening of the Raman modes and further changes in the lattice dynamics. X-ray diffraction experiments revealed the amorphization of the surface layer.

The presentation gives an overview of the structural changes of La monazite single crystals under swift heavy ion irradiation at ion fluences of 5e11 ions/cm² ,1e12 ions/cm² and 2e12 ions/cm². After irradiation, cross sections of the single crystals were prepared and additionally polished with an Ar ion mill to investigate the surface damage along the path of the fast heavy ions using optical light microscopy and Raman spectroscopy. The methods used show a strong surface damage within the projection range of the gold ions due to color change, increase of the FWHM of the Raman band and decrease of crystallinity.

[1] Donald et al. (1997) J. Mater. Sci. 32; [2] Ewing (1999) PNAS, 96; [3] Lumpkin et al. (2006) Elements, 2; [4] Schlenz et al. (2013) Z. Kristallogr. Cryst. Mater. 228; [5] Ziegler et al. (2010) Nucl. Instrum. Methods Phys. Res. B 268

The authors acknowledge the BMBF for financial support in the project No. 02NUK060.

The idea of immobilizing radionuclides in crystalline host materials was put forward 70 years ago. Since then, continuous research has been conducted on a wide variety of crystalline materials that are considered as possible host matrices. However, many challenges remain, owing, e. g., to the complex chemistry of nuclear waste streams and the exceptionally high requirements regarding physical and chemical long-term stability.

Monazite (LnPO_4, Ln = La-Gd) has long been considered as one of the most promising crystalline host materials for long-term storage of radionuclides, especially actinides. The main reasons for this are its chemical flexibility, its excellent aqueous durability and its low recrystallization temperature, which allows for rapid self-healing of radiation induced damages. It has been shown that monazites can accommodate large amounts of trivalent actinides within their crystal structure. However, the incorporation of tetravalent dopants via coupled substitution with divalent cations has proven challenging, even though natural monazite is known to contain significant amounts of Th and U (combined up to 27 wt-%).

To facilitate assessments with respect to selection criteria such as chemical flexibility, radiation resistance and aqueous durability, efforts are made to identify inactive surrogate-models. The use of cerium as a surrogate for tetravalent actinides will be discussed for monazite-type phases based on the solid solution La_1-x(Ca,Ce)_xPO₄ which was extensively studied using powder and single crystal XRD, electron imaging techniques including EPMA, SEM and TEM as well as spectroscopic measurements including Raman, TRLFS, EXAFS and in-situ XAS experiments. Based on these findings the synthesis of active monazites containing up to 50 % Th⁴⁺ was successfully performed as shown by PXRD measurements.

The poster will focus on irradiation studies of monazite-type ceramic pellets from the solid solution La_1‑xCe_xPO₄. Monazite is known for its remarkable ability to recover from radiation damage by a combination of low recrystallisation temperatures (~570 K) and low activation energies for thermal annealing (<3 eV) as well as irradiation-induced recrystallisation which was observed both from external irradiation and self-irradiation.

While various studies have been published investigating the effects of radiation on the monazite structure, the impact of disorder introduced by solid solutions has not yet been studied extensively. For this reason, various compositions of the aforementioned solid solution were irradiated with Au ions at two different fluences and subsequently analysed with SEM, gracing incidence XRD and Raman spectroscopy in order to gain a better understanding of their radiation stability and recrystallisation properties.

The thermal and radiation stability of Zircaloy cladding material that houses spent nuclear fuel (SNF) is an important factor when considering the storage and eventual disposal of SNF in a geological repository. It is known that on the surface of the cladding, oxidised zirconia (ZrO₂) phases are inherently present. Following fuel swelling and rim contact, the zirconia layer on the interior surface can interact with SNF elements, leading to the formation of phases such as pyrochlore and zirconates among others. These phases act as the first intermediate barrier between SNF and the metallic cladding and consequently are important to consider in safety design, particularly for release of radionuclides (RNs). A pertinent RN that contributes significantly to the radiological hazard of SNF, is the isotope Am-241. The chemistry of Am is largely unique, being able to readily dissociate between its tetravalent and trivalent states in oxides, making it difficult to investigate via surrogate studies. Furthermore, Am-241 has a relatively short t_1/2 of 432 years and decays via alpha emission (5.486 MeV), resulting in significant ensuing radiation damage in host materials. Consequentially, understanding the thermal and radiation stability of host material phases incorporating Am-241 is pertinent for safe disposal of SNF. As a part of the national project “AcE” funded by the German Federal Ministry of Education and Research (BMBF) and through the European Commission ActUsLab program, we have investigated several zirconium oxide polymorphs, including but not limited to Nd-pyrochlore and zirconia, doped with 5 mol% Am-241. The particular focus of the investigation is to understand the thermal and radiation stability of the different oxide polymorphs when Am-241 is incorporated. This presentation will highlight some on-going results from this research program, including high-temperature phase transformations, radiation induced lattice swelling, phase separation, and associated apparent redox activity induced by the presence of Am-241.

Recent studies have suggested that crystalline ceramic matrices, such as monazites and zirconia (ZrO₂) have a high potential to be used as immobilization matrices for radioactive waste. At room temperature, zirconia has a monoclinic (m) structure. At higher temperatures, tetragonal (t) and cubic (c) structures can be stabilized. The phase stabilization can also be achieved at ambient conditions by incorporating oversized cations. In addition, several metastable phases (t′, t′′, κ, and t*), can be formed for doped zirconia materials. Out of the several structural polymorphs, especially the cubic structure shows high radiation tolerance, which is important for host matrices containing radioactive elements. In the current study, cerium has been used as an analog for plutonium as these f-elements have identical cation radii and can be stabilized in the trivalent and tetravalent oxidation states. The zirconia samples were co-doped with a small amount of Eu(III) to allow for luminescence spectroscopic analyses of the solid phases. In a first step, the co-precipitation route was applied to synthesize Ce-doped zirconia samples over a wide Ce-concentration range. The phase composition of the samples was investigated with X-ray diffraction, and showed that the radiation tolerant cubic phase was stabilized only for samples with Ce concentrations above 75 mol% . At lower dopant concentrations, a mixture of different phases were present, including monoclinic in a low doping concentration range, tetragonal and tetragonal double prime phases appearing for intermediate Ce-concentrations. The latter phase was detected only by Raman spectroscopy, showing the presence of a defect band at 526.5 cm^-1. In addition, luminescence spectroscopy revealed structural changes in terms of different Eu environments in the t´´ and c samples. To stabilize the cubic phase for low tetravalent doping concentrations, trivalent yttrium (Y) was incorporated as a co-dopant. XRD and Raman analyses show that the cubic phase was stabilized when the concentration of Y was higher than 15 mol%. Finally, using the same co-precipitation route, a series of uranium-doped zirconia samples was synthesized. XRD investigations show a phase transformation from monoclinic to tetragonal and orthorhombic with increasing uranium doping. Identical to the Ce-doped samples, the pure cubic phase was stabilized only in the presence of Y for concentrations higher than 15 mol%. Discerning the crystal structure is crucial to understanding the properties of these phases. Although the binary zirconia systems with only one dopant show different phase compositions for Ce and U, the scenario changes when adding a trivalent co-dopant such as yttrium, which stabilizes the cubic phase both in the presence of uranium and cerium.

Preliminary solubility results for the pure cubic phase of uranium/cerium-doped zirconia co-doped with yttrium will be shown in the poster session.

Inventory of radionuclides is crucial as an input parameter in a safety assessment of radioactive waste disposal. In a geological disposal of irradiated metal wastes, Mo-93 is a remarkable radionuclide due to its long half-life (4,000 y) and low-sorptive property (anionic species). A determination method of Mo-93 by measuring gamma-ray (X-ray) has been previously proposed, however, the method required complex sequential and/or chromatographic separation of Mo-93 from Nb-93m and Zr-93.

In this study, a technique using triple quadrupole inductively coupled plasma mass spectrometry (ICP-QQQ), which is a leading-edge device, is adopted to determine inventory of Mo-93 in irradiated BWR tie plate (i.e., bottom end piece of BWR fuel, burnup of 35.0 GWd/tHM). The greatest advantage using the ICP-QQQ is that isobaric overlap from Nb-93 and Zr-93 can be effectively suppressed, resulting in eliminating the need for complex sequential and/or chromatographic separation before injection. By adopting the technique, Mo-93 inventory was able to be easily determined as 701±82 Bq/g. The result will be discussed with relation to Mo-93 inventory estimated using an activation calculation.

This study was carried out as a part of R&D supporting program titled “Advanced technology development for geological disposal of TRU waste (2022 FY)” under the contract with the Ministry of Economy, Trade and Industry (METI) (Grant Number: JPJ007597).

One of the expected issues in Japan’s radioactive waste management is treatment and disposal for spent Mixed Oxide (MOX) fuel. Spent MOX fuel is to be vitrified after reprocessing, and the heat generation of Am-241 in the spent MOX fuel is a key factor in vitrification and on the geological disposal. It means that if the buffer material temperature exceeds the upper limit temperature of 100℃ due to the heat generated by the vitrified waste, the quality of the buffer material will be degraded, which will affect the nuclide migration. Therefore, waste loading ratio of vitrified waste is estimated by the heat generation, which determines the number of vitrified waste units generated (canister/tHM) and geological repository footprint (m²/tHM). One of the options for reducing the heat generation is blending MOX HLW with UO₂ HLW.

In this study, we evaluated the effect of blending spent MOX fuel with UO₂ fuel on the heat generation considering the cooling period and blending ratio etc. The waste-loading ratio was not a constant value, but the maximum in each blending condition where the buffer material temperature was below 100℃ was calculated, and the number of glass units generated from UO₂–MOX HLW were estimated using the results, being assessed these repository footprints.

It will be also discussed the possibility that these calculation data enable to select optimal conditions (blend ratio and waste loading ratio) when spent UO2 and MOX fuels with various conditions are blended and vitrified.

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).

Nuclear waste glasses often contain many complex phase separations, crystals, and alteration features. Understanding these components is key to predicting and engineering glass durability. This poses a characterization challenge in that heterogeneity may exist over multiple length scales, requiring comprehensive characterization from the arrangement of atoms to mm. The work described herein makes use of advanced, non-synchrotron X-ray characterization methods that span this range: Small Angle X-ray Scattering (SAXS) provides information on nanoscale features, X-ray nano-Computed Tomography (nano-CT) provides 3D representation of features 100 nm – 50 µm, Wide angle X-ray Scattering (WAXS) provides atomic-scale crystallographic arrangements, and large field-of-view X-ray Transmission Imaging maps the materials structure over micron to mm scales. This poster will illustrate structural insight from this characterization scheme on simulated nuclear waste glasses for vitrification of waste streams including Mo and La containing aqueous reprocessing waste and F^- and SO₄^2- rich legacy waste. From the nanoscale to macroscale, a variety of qualitative and quantitative phase information (such as size, morphology, distribution, and interactions) was elucidated. Specifically, hierarchal, segregated, and heterogeneous glass-glass phase separation and crystallization were examined. Further, the non-destructive quality of these methods enabled observation alteration layer formation during vapor hydration testing of the 2^nd International Simple Glass (ISG-2). Reflected-light microscopy, Scanning Electron Microscopy (SEM), Wavelength Dispersive Spectroscopy (WDS), Raman Spectroscopy, and quantitative X-ray Diffraction (XRD) were utilized to verify the results. Combining X-ray imaging, SAXS, and WAXS has the potential to help visualize and evaluate features key to long-term durability of vitrified waste, as well as elucidate kinetic processes in complex glasses. Therefore, this study presents an approach towards more broadly understanding multi-scale structural attributes of nuclear waste glasses.

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.

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.

Treatment of radioactive effluent containing dissolved ¹³⁷Cs and ⁹⁰Sr has a pivotal role in nuclear waste management, however, effective techniques allowing simultaneous removal of these fission products are rather limited and are a topic of interest. Therefore, the specific objective of this study was to investigate the efficacy of composite materials, using collective ion exchange and coagulation, by combining fine clinoptilolite and co-precipitation with barite (BaSO₄), designed to increase selectivity levels towards both Cs⁺ and Sr²⁺ ions. In the batch system, the removal efficiency of BaSO₄ for both ions was examined first, followed by the adsorption kinetic study to confirm the adsorption capacity of natural clinoptilolite. BaSO₄ was found to be very effective in the removal of Sr²⁺ (>99%) while giving very low-level removal in Cs⁺ ions (~14%). Adsorption kinetics of natural clinoptilolite was performed for 50 ppm Cs⁺ and Sr²⁺ simultaneously; they were fitted by the Pseudo-Second Order (PSO) rate model, giving the rapid adsorption equilibrium(~1h). Composite coagulants were then produced using 20 g/L and 40 g/L of natural clinoptilolite combined with BaSO₄ co-precipitation. Higher Sr²⁺ removal was obtained in all cases (>99%), whereas Cs⁺ removal efficiency was not gone beyond ~87%. However, NaCl activation of clinoptilolite was used in the combined system to overcome low Cs⁺ removal efficiency, achieving >95% of removal. Moreover, their physical properties- such as particles size distribution, surface charge, sedimentation rate, and profiles, as well as the compressional yield stress, were also studied to characterize the particle colloidal behavior in terms of whether suspensions are industrially suitable for solid-liquid separation, namely dewatering.

In the previous dose assessment for the public due to radionuclides leaking from waste repositories, dissolved radionuclides were assumed to flow directly into the living environment (ocean, lake, river, etc.) through natural barriers, and nuclide migration was estimated using compartment model stylizing the environmental media. However, it was reported that radionuclides via groundwater could sorb and desorb with soil near spring water points, and that radiocesium was mainly transferred as sorbed to suspended particles in the living environment. In this study, an estimation of Cs-135 migration was performed using newly constructed compartment model for a coastal zone, in order to analytically understand the influence on the migration in the living environment with or without consideration of the Cs-135 sorption on the seabed soil near the spring water point when the Cs-135 flowed into the zone. In addition, the effects were evaluated for the presence or absence of the Cs-135 sorption/desorption on suspended particles and the particle sedimentation. As a result, the Cs-135 concentration in seabed soil was several tens of times higher in the estimate that considered the sorption on seabed soil during inflow than in the estimate that did not. The concentration in seawater immediately above the seabed soil (interface layer) was several hundred times higher when sorption/desorption on suspended particles and the particle sedimentation were considered than when these phenomena were not. These results indicate that it is important to consider the radionuclide sorption on the seabed soil and the migration of radionuclides sorbed on suspended particles in the estimation of radionuclide migration in the living environment because these phenomena could cause the increase of radionuclide concentrations in the interface layer and the seabed soil and the higher exposure due to benthic fish and shellfish ingestion, etc.

A high sorption capability of biotite for radionuclides is based on the experiments using powdered samples. However, considering that biotite exists as a flake form in granite, a representative plutonic rock, radionuclides not only adsorb on its edge parts but also diffuse into its layer structure. In addition, coexisting ions in groundwater may affect such interactions. Especially, saline groundwater contains the high concentration of Na⁺. Therefore, this study examined the sorption of europium (Eu) onto biotite flakes in the coexistence of Na⁺. Eu is a chemical analog of americium (Am) which is a typical α-nuclide in radioactive waste.

The sorption experiments were conducted by mixing 30 mL of 0.5 mM Eu(NO₃)₃ in 0.6 M NaCl or KCl and 3.0 g of biotite flakes (5 mm × 5 mm). The pH value of the solutions was adjusted to 5 to avoid hydrolysis of Eu. The concentration of Eu was monitored daily for 7 days. After that, the distribution of Eu in biotite samples was observed by SIMS.

In the results, the sorption of Eu gradually increased over time due to the diffusion into the layer structure of biotite flakes. Then, the sorption amount of Eu in the coexistence of Na⁺ was less than that in the absence of Na⁺, and Eu hardly adsorbed in the coexistence of K⁺. Moreover, the apparent diffusion coefficients of Eu in biotite flakes were estimated to be on the order of 10^-13 m²/s by applying a two-dimensional diffusion model coupled with the time-change of Eu concentration in the solution to the experimental data. These values were smaller than the effective diffusion coefficient in plutonic rock, which is used for assessing the performance of geological disposal systems. This would mean that the sorption of Eu in plutonic rock will be restricted by the diffusion in biotite flakes.

The deep geological repository concept for radioactive wastes is based on the confinement of the radioactivity over long periods of time by a multiple barrier system. Cementitious materials are used as part of the barriers in most of the repository concepts developed internationally (e.g., as backfill, tunnel lining, or in shaft seals and plugs). Although the behaviour of safety-relevant radionuclides in cementitious environments has been investigated extensively in the last decades, the impact of organic degradation products, originating from organic waste components or from superplasticisers in cementitious materials, on the migration of radionuclides under highly-alkaline, cementitious conditions is not yet fully understood. Therefore, the objective of this work, carried out within the framework of EURAD WP CORI (Cement-Organic-Radionuclide Interaction), was to fill knowledge gaps in the understanding of the impacts of the presence of phthalate (C₈H₄O₄²⁻; degradation product from plasticisers in PVC) and tri-methyl-amine (TMA; degradation product of ion exchange resins) on the migration behaviour of ²⁴¹Am and ¹⁵²Eu in cementitious barriers.

In this context, hardened cement pastes (HCP) were prepared with a water/cement ratio of 0.40 from a composite cement (CEM V/A 42.5N; Calcia, Rombas). The uptake and diffusion of ²⁴¹Am and ¹⁵²Eu in HCP was studied under anoxic conditions in the presence and absence of organics. In the absence of organics, a strong retention of both radionuclides on HPC was observed (R_d values between 10⁵ and 10⁶ dm³ kg^-1). In contrast, at phthalate concentrations exceeding ~10^-3 M, a reduction in the uptake of ²⁴¹Am and ¹⁵²Eu on HCP by several orders of magnitude was observed. This reduction in sorption could be the consequence of the decalcification of calcium silicate hydrates (C-S-H), the main sorbing phase in cementitious materials, due to the increasing formation of Ca-phthalate complexes in solution. These results indicate an increase in the mobility and diffusion of ²⁴¹Am and ¹⁵²Eu in cementitious barriers with increasing phthalate concentrations.

Acknowledgements

The EURAD-CORI project leading to this application has received funding from the European Union’s Horizon 2020 research and innovation programme under grant agreement No 847593.

Buffer material composing engineered barrier in the geological disposal of a high-level radioactive waste develops swelling stress by penetration of groundwater from the surrounding rock mass. In previous studies, we measured the activity of water and the Gibbs free energy of water in Na-montmorillonite which is the major component of Na-bentonite by vapor pressure method, and proposed a model to analyze the swelling stress of bentonite based on thermodynamic theory. However, data for the vapor pressure of water in bentonite are limited. In this study, we determined the activities of water and the Gibbs free energy by measuring relative humidity (RH) for water in Na-bentonite and Na-montmorillonite. We also analyzed the swelling stress of bentonite based on the thermodynamic model and compared to the measured data.

Kunigel-V1 and Kunipia-F (Kunimine Industries Co. Ltd.) were used as a Na-bentonite. The montmorillonite contents of both bentonites are approximately 51% and 99%, respectively. Bentonite powder dried was placed in a polyethylene bottle in an amount of 3.00g each, and adsorbed water vapor in a vacuum chamber. Next, RH and temperature sensors and bottles with bentonite were placed in the vacuum chamber, and the chamber of which inside pressure was reduced to -95kPa or less was submerged in a water bath at 25°C. The RH and temperature in the chamber were measured after 24 hours and the weight of the bentonite was measured. This operation was repeated every about 24 hours. Thus, RH and temperature were measured versus water content (ca. 10-100%).

The activities of water and the Gibbs free energies for both bentonites decreased with decreasing water content below approximately 40%. This trend is the same as in the past studies. The swelling stresses of bentonite calculated using thermodynamic data obtained in this study were generally in good agreement with the measured values.

Metallic beryllium is employed in a wide range of nuclear applications, for example, as moderator, reflector,
or fuel cladding in thermal reactors, due to its low neutron-capture cross section and high potential for
elastic neutron scattering. Moreover, it is also under consideration as interesting material for future fusion
power reactors (e.g., as first wall). Neutron irradiation of beryllium during utilization in fission and fusion
reactors entails the accumulation of significant amounts of short- and long-lived activation products (e.g.,
³H formed from ⁹Be via ⁹Be(n,α)⁶He → ⁶Li(n,α)³H as well as activation products of impurity elements),
requiring its disposal as radioactive waste after decommissioning. One strategy for the management of lowand intermediate-level radioactive metallic wastes is their encapsulation in cementitious matrices.
However, data on the corrosion and reactivity of metallic beryllium in the disposal environment (e.g., with
respect to H₂ evolution) is scarce. Thus, within the frame of the collaborative EC-funded Horizon 2020
project PREDIS (Predisposal management of radioactive waste) the corrosion behaviour of metallic beryllium is investigated in solutions with different pH and composition, simulating potential encapsulation
matrices such as Ordinary Portland Cement (OPC) or magnesium phosphate cement (MPC). Corrosion
rates of metallic beryllium samples are determined experimentally by using gravimetric methods (weight
loss) combined with the determination of Be concentrations in solution via ICP-MS. The obtained results
are compared with results derived by the H₂ release under similar corrosion conditions. Detailed studies of pristine and corroded metal surfaces are carried out using SEM-EDS providing insights into corrosion
mechanisms and the possible initiation places for pit corrosion in dependence of solution composition. In
this presentation, first results regarding the corrosion kinetics and corrosion mechanisms of metallic
beryllium under various conditions are discussed.