Focused Session FR-10
Materials Issues in Radioactive Nuclear Waste Treatment and Disposal


FR-10.1:IL01  Geopolymer Foam as Inorganic Monolithic Sorbent for the Decontamination of Liquid Radioactive Waste
S. Petlitckaia, CEA, DES, ISEC, DE2D, SEAD, LCBC, Univ Montpellier, Marcoule, Bagnols sur Cèze, France; Y. Barré, CEA, DES, ISEC, DMRC, STDC, LPSD, Univ Montpellier, Marcoule, Bagnols sur Cèze, France; J. Vicente, Aix Marseille Univ, IUSTI, CNRS, UMR 7343, Marseille, France; A. Poulesquen, CEA, DES, ISEC, DE2D, SEAD, LCBC, Univ Montpellier, Marcoule, Bagnols sur Cèze, France

In nuclear industry, liquid radioactive waste coming from reprocessing plant or nuclear accident has to be treated in order to decontaminate these effluents. One way to lowering the radioactivity of the effluent is to synthesize inorganic monolithic sorbent that are less sensitive to radiolysis phenomena than organic ones. Geopolymer binders are good candidates to fulfill these specifications. This work aims to synthesize monolithic geopolymer foams who can act both as sorbent for decontamination of liquid radioactive waste and containment matrix. The macroporosity has to be connected in order to facilitate the transport of contaminated fluid without pressure drop, and the control of the chemical parameters of the geopolymers allows to tailor the mesoporous network. After describing the method of synthesis, I will discuss the advantage to functionnalize geopolymer foams by copper hexacyanoferrate to selectively decontaminate some radioactive effluents containing cesium. In a second part, the structure of the generated macroporous network will be assess using X-ray tomography. From these 3D observations and characterizations, the permeability of the geopolymers was computed according to the topology of the network that depend of the chemical formulation.

FR-10.1:IL02  Development of Zirconolite Ceramic Wasteforms for Pu Immobilisation
L.R. BLACKBURN1, L.J. Gardner1, A.R. Mason1, Shi-Kuan Sun2, E.R. Maddrell3, M.C. Stennett1, C.L. Corkhill1, N.C. Hyatt1, 1University of Sheffield, UK; 2Guangdong University of Technology, China; 3National Nuclear Laboratory, UK

More than 340 metric tonnes of separated Pu are held internationally, stored largely as unirradiated PuO2, presenting a need to develop advanced ceramic and glass materials capable of acting as a disposal matrix. Zirconolite, ideally CaZrTi2O7, is an accessory mineral found in a variety of localities that has demonstrated affinity for the incorporation of a wide variety of elements, including actinides, and is therefore a candidate ceramic matrix for the immobilisation and sequestration of Pu. The zirconolite structure can accommodate moderate Pu loadings through homovalent solid solution i.e. CaZr1-xPuxTi2O7 or heterovalent substitution with suitable charge compensation i.e. Ca1-xPuxZrTi2-2xFe2xO7. The safety case for Pu disposal in a zirconolite matrix is underpinned through the fabrication of radioactive and inactive surrogate trials, developing confidence in the immobilisation approach. In the current work, we summarise an extensive wasteform development campaign, aiming to systematically characterise the incorporation of Ce/U/Th within the zirconolite system, targeting a large array of compositions using both conventional sintering and a batch hot isostatic pressing (HIP) process.

FR-10.1:IL03  High-entropy A2B2O7-type Oxide Ceramics for the Immobilization of High-level Radioactive Waste
GUO-JUN ZHANG, L. ZHOU, F. LI, J.-X. LIU, State Key Laboratory for Modification of Chemical Fibers and Polymer Materials, College of Materials Science and Engineering, College of Science, Institute of Functional Materials, Donghua University, Shanghai, China

Immobilization of radioactive nuclear waste is a technical means to fix radionuclides in stable substrates by solidification, embedding or capsule treatment of radioactive waste, which is convenient for later safety treatment, transportation, storage and disposal. Developing a new types of immobilization materials that can coexist stably with long-lived radionuclides remains a central challenge, which involves the design of the composition and microstructure of materials. In this work, we report a series of high-entropy A2B2O7-type oxide ceramics (Eu1-xGdx)2(Ti0.2Zr0.2Hf0.2Nb0.2Ce0.2)2O7 (x=0, 0.5 and 1.0) successfully prepared by solid state method from oxides raw materials, and the characteristics of the high-entropy ceramics were characterized. At the same time, we demonstrated that high-entropy ceramics are the potential candidate as immobilizing hosts for high-level radioactive waste. The static aqueous leaching test indicates that the normalized leaching rates for the simulated radionuclides Ce (LRCe) and Gd (LRGd) in as-prepared high-entropy ceramics are approximately 10−8~10−6 g·m−2·d−1 after 42 days testing, much lower than those reported values in doped-RE2Zr2O7 (10−6~10−3 g·m−2·d−1).

FR-10.1:IL04  Thermodynamic Stability, Radiation Damage and Leaching Effects in Tunnel Structured Hollandite Materials
K. BRINKMAN, Materials Science and Engineering, Clemson University, Clemson, SC, USA

Crystalline ceramic waste form material systems are designed using existing minerology and knowledge of durable crystalline materials found in nature. The hollandite mineral group is one of the most studied as a host for Cs immobilization. In this study the A site species consisting of Ba and Cs with a wide range of B site dopants including Zn, Ga, Fe, Al and Cr were incorporated into the hollandite structure. The enthalpies of formation of the hollandite phases measured using high temperature oxide melt solution calorimetry were found to be negative, indicating these hollandites are thermodynamically stable with respect to their constituent oxides. Furthermore, the formation enthalpies were more negative and hence more favorable with increased Cs content across a wide range of B site dopants. In addition to formation energy measurements, select monolithic a powder-based aqueous leaching tests indicated that the hollandite phase with higher Cs-loading enhanced Cs retention. In parallel, three Cs end member compositions of the tunnel structured hollandite samples (Cs1.33Ga1.33Ti6.67O16, Cs1.33Fe1.33Ti6.67O16, and Cs1.33Zn0.67Ti7.33O16) were synthesized and irradiated by a 1 GeV Au ion beam in order to study effects of B-site dopants on radiation stability.

FR-10.1:IL05  Recent Advances in the Immobilization of Low- or Intermediate-level Radioactive Waste in Cementitious Materials: Potential of Magnesium Potassium Phosphate Cements
C. CAU DIT COUMES1, G. Poras1, D. Chartier1, D. Lambertin1, C. Cannes2, S. Delpech2, S. Perrin1, 1CEA, DES, ISEC, DE2D, SEAD, Univ Montpellier, Marcoule, France; 2Institut de Physique Nucléaire, CNRS, Univ. Paris-Sud, Orsay, France

Stabilization and solidification of low- and intermediate-level radioactive waste using Portland cement (PC), possibly blended with fly ash or blastfurnace slag, is a well-established practice. This solution corresponds to a technical and economical optimum for a wide range of wastes. However, its use can be restricted by deleterious interactions occurring between the waste and the cement matrix. Over the last decade, increasing attention has been paid to alternative inorganic binders showing a better compatibility with problematic waste streams. In this presentation, the focus is placed on magnesium phosphate cements (MPC) which may offer new prospects for the conditioning of reactive metals or highly irradiating waste. The hydration process of MPC is presented and the interest of using thermodynamic modelling to predict the phase evolution is highlighted. Then, the electrochemical behavior of aluminum and aluminum-magnesium alloys in a MPC matrix is investigated using impedance spectroscopy. Besides, the radiolytic H2 production of MPC materials is determined under external gamma or internal alpha irradiation. Finally, it is shown that the H2 release, whether it arises from corrosion or irradiation, can be further reduced by adding an inorganic getter to the MPC matrix.

FR-10.1:L09  Incorporation of Tetravalent Actinides in Monazite Structures
A. ROCHE, S. Szenknect, A. Mesbah, N. Clavier, R. Podor, N. Dacheux, ICSM, UnivMontpellier, CEA, CNRS, ENSCM, Site de Marcoule, Bagnols-sur-Cèze, France

Due to various properties of interest, several phosphate-based ceramics have been considered for long as potential matrices for the specific immobilization of radionuclide. Among them, monazite-type materials (Ln,An)PO4, have been widely studied due to high chemical flexibility and durability. While the incorporation of trivalent actinides is easy by both wet and dry chemical methods, that of tetravalent actinides is limited to dry chemical routes due to persistent difficulties in the precipitation processes. However, wet chemistry routes can improve the chemical homogeneity and sinterability and thus of the chemical durability of the materials in weathering conditions. This work aims at the preparation of materials crystallizing with the monazite structure such as cheralite. The incorporation of uranium (IV) in CaxAnxNd1-2xPO4 (0 ≤ x ≤ 0.1) solid solutions was undertaken by direct precipitation of rhabdophane-type precursors. A multiparametric study first led to specify the operating conditions (starting stoichiometry, temperature, heating time) leading to single-phase powdered CaxAnxLn1-2xPO4·nH2O samples. The conversion then sintering from rhabdophane-type precursors into highly durable cheralites was further studied.

FR-10.1:L11  Understanding the Evolution of an Interface during the Dissolution of Nd-doped UO2 by Macro-/Microscopic Dual Approach
T. BARRAL, L. ClaparEde, R. Podor, N. Dacheux, ICSM, UnivMontpellier, CEA, CNRS, ENSCM, Site de Marcoule, Bagnols-sur-Cèze, France

Recycling of actinides is a key issue to consider in order to preserve natural fissile resources and to reduce the long-term radiotoxicity of HLW in deep geological disposal. Due to the high radioactivity of SNF, the use of model compounds is often required to understand their behavior during reprocessing. Thus, this work aimed at monitoring the behavior of a large panel of Ln-doped UO2 sintered pellets (Ln = Nd, Gd) during dissolution using a macro-/microscopic dual approach. First, precursors of oxide materials were prepared by hydroxide precipitation leading to nano-sized powders. Well densified pellets were obtained by sintering these powders through RT shaping then firing at 1600°C under Ar or Ar/H2 atmosphere. Multiparametric dissolution tests of the pellets were performed under static conditions varying temperature, nitric acid concentration and chemical composition of the solid, allowing the determination of normalized dissolution rates. At the microscopic scale, operando monitoring of the solid/liquid interface was implemented by ESEM to follow the consequences of dissolution on the ceramics microstructure. It showed different dissolution pathways depending on the sintering atmosphere under reprocessing conditions.

FR-10.2:IL01  Challenges in the Management of Spent Nuclear Fuel
G. LEINDERS, Belgian Nuclear Research Centre (SCK CEN), Fuel Materials, Institute for Nuclear Materials Science, Mol, Belgium; C. SCHREINEMACHERS, Forschungszentrum Julich, Julich, Germany

Nuclear technology is being used in many civil application fields, ranging from commercial energy generation, to the production of radioisotopes for industry and pharma, material tests for fission and fusion technology, and for analytical purposes such as neutron scattering analyses. Evidently, it has required the development and exploitation of various types of nuclear power plants, suited for these specific needs. The most widely used nuclear fuel configuration is that of uraniumdioxide (UO2) and U-Pu mixed-oxide (MOX) ceramic fuel pellets confined in metallic fuel pins. However, various aspects related to nuclear fuel management remain challenging and may pose problems to the future sustainable use of nuclear technology. A primary concern has been the safe handling and ultimate disposal of nuclear fuel discharged from a reactor (i.e. spent fuel), which presents elevated levels of radioactivity and heat emission during a geological timescale (up to 10E6 years). Additionally, in all stages of the nuclear fuel cycle uncontrolled oxidation and corrosion reactions may cause ruptures in confinement barriers, leading to contamination of the surroundings. In this contribution current insights and future developments in the management of nuclear fuel will be presented and discussed.

FR-10.2:L05  Solid Fixation on Grafted Mesoporous Silica for Actinide Uptake
C. REY, X. Deschanels, J. Causse, ICSM, CEA, CNRS, Université de Montpellier, ENSCM, Bagnols-sur-Cèze, France; G. Zante, S. Le Caër, Université Paris-Saclay, CEA, CNRS, NIMBE, UMR 3685, Gif-sur-Yvette, France; V. Bouniol, S. Sene, Y. Guari, ICGM, Université de Montpellier, CNRS, ENSCM, Montpellier, France

In recent years, there has been growing interest in the focus on research to develop new hybrid materials capable of selectively extracting actinides [1]. Mesoporous silicas functionalized with organic ligands have been studied for this purpose. In this study, several ligands: Acetamide phosphonate (AcPhos), Propionamide phosphonate (PropPhos), Hydroxypyridine-N-oxide (Hopo), Tributyl phosphate (TBP) were compared to assess their sorption capacity towards U and Th under various acidic conditions. The impact of electron radiolysis of 10 MeV up to 4MGy on the sorption properties was also determined. All materials were characterized by SAXS, BET, TGA, NMR and microscopy techniques. This preliminary study shows that the AcPhos and ProPhos ligands exhibit interesting sorption capacities, even after radiolysis, which led to their selection for carrying out sorption tests on plutonium.
[1] G. Fryxell, and al.; Environ. Sci. Technol. 2005, 39, 1324-1331

FR-10.2:L06  Fabrication and Characterization of Americium Transmutation Target Microspheres
G. COLAK1, 2, G. Leinders1, a.R. Delville1, F. Jutier1, M. Verwerft1, J. Vleugels2, 1Belgian Nuclear Research Centre (SCK CEN), Institute for Nuclear Materials Science, Mol, Belgium; 2KU Leuven, Department of Materials Engineering, Leuven, Belgium

One major challenge in the utilization of nuclear energy is the management of spent nuclear fuel, even after uranium and plutonium recycling. The radiotoxic minor actinides (MAs), particularly americium (less than 0.1 wt.% of spent fuel), could be extracted and recycled via partitioning and transmutation. In the frame of the ASOF-project at SCK CEN, the conversion of aqueous feeds into solid oxide target in form of microspheres is investigated via internal gelation method. While these conversion methods show benefits in terms of automation and reduction of the contamination risk, major challenges such as the production of MA contaminated liquid wastes remain. Another method to mitigate or solve this problem is under investigation; in particular, the fabrication of doped microspheres via infiltration method. In this process highly porous uranium oxide-based microspheres, using a sol-gel method, are fabricated and later infiltrated via an americium nitrate solution. Internal gelation may affect the surface topography and internal pore structure and hence affect infiltration efficiency. In this contribution, latest insights on the fabrication of uranium oxide-based microspheres via internal gelation (Nd3+ or Am3+), as well as their analytical characterization will be presented.

FR-10.3:IL01  Characterization of Radiation Effects in Ceramics with Spallation Neutron Probes
M. LANG, Department of Nuclear Engineering, University of Tennessee, Knoxville, TN, USA

The development of durable materials for radionuclide immobilization has been central to efforts to dispose of wastes generated by the nuclear fuel cycle. There still exist, however, large gaps in the understanding of fundamental modes of waste form degradation under long-term self-irradiation. We have shown neutron total scattering measurements with pair distribution function (PDF) analysis can be utilized to uniquely characterize radiation effects in a wide range of fluorite-derived wasteform materials including pyrochlore, spinel, and actinide oxides. These measurements enable detailed analysis of both cation and anion defect behavior, and short-range order, which is particularly important for the investigation of amorphous materials. Recent results for several complex oxides demonstrate that radiation effects are more complex than previously thought with distinct processes occurring over different length scales. For example, disordered pyrochlore and spinel are composed of local structural units that maintain atomic order and exist in configurations that are different than the expected average structure determined using traditional techniques (XRD). Here we will highlight the importance of short- and medium-range analysis for a comprehensive description of radiation behavior.

FR-10.3:IL02  Recent Experimental Evidence for the Suppression of the dissolution of spent nuclear fuel by H2 Gas
Th. MENNECART1, L. Iglesias Pérez2, M. Herm2, C. Cachoir1, K. Lemmens1, V. Metz2, K. Meert3, 1SCK CEN – Belgian Nuclear Research Centre, Mol, Belgium; 2KIT-INE - Karlruhe Institute of Technology, Institute for Nuclear Waste Disposal, Eggenstein-Leopoldshafen, Germany; 3ONDRAF/NIRAS - Belgian Agency for Radioactive Waste and Enriched Fissile Material, Bruxelles, Belgium

In many countries, the long term management strategy for at least part of the spent fuel (SF) inventory, is direct geological disposal where the safety is assured by a multi-barrier system comprised of a geological barrier and an engineered barrier system. The latter should prevent the release of contaminants from the waste at least a few thousand years (thermal phase). On the longer term, corrosion will lead to perforation of the metallic engineered barriers, and the ground water will come in contact with the SF, which will then slowly dissolve. Two types of dissolution are possible in anoxic conditions: one governed by the matrix (UO2) solubility and a faster oxidative dissolution due to the presence of oxidizing species produced by the water radiolysis. Also, in contact with the iron-based constituents surrounding the SF, the ground water will produce a strong reducing species H2. The oxidative species can react with H2 instead of the SF, impeding or even suppressing the fuel dissolution. This presentation will give a general overview of the experimental evidence for this mechanism, will show how the behavior of SF in presence of H2 is investigated and will address the similarities and differences with leaching experiments performed with depleted and alpha doped UO2 materials.

FR-10.3:IL03  In-operando Raman Spectroscopy applied to the Understanding of MOX Fuel Alteration Mechanisms
S. MIRO, C. Jégou, L. Sarrasin, M. Tribet, V. Broudic, C. Marques, S. Peuget, CEA, DES, ISEC, DE2D, Université de Montpellier, Marcoule, France

Micro-Raman spectroscopy is a powerful technique for the characterization of nuclear materials (fuels, claddings, simulated fuel debris, glass and glass-ceramic). This technique remains in particular a very suitable tool for studying the aging of a defective spent-fuel fuel rod stored underwater for several decades. Under these conditions of intense gamma irradiation, the water radiolysis induces an oxidative dissolution of the fuel and the precipitation of secondary phases like uranium peroxides. This process is well described for UOX fuels but much less studied for MOX fuels. An original experimental approach was therefore implemented, combining MOX fuel leaching experiments in pure labeled water (H218O) under gamma irradiation field, with the characterization of the leached surfaces in hot cell. Thanks to Raman mapping, we evidence the preferential alteration of the uranium-rich zones during leaching experiments and the local precipitaion of studtite phase on these areas. No secondary plutonium-based phase could be observed. Isotopic analysis of the Raman bands of the studtite explained the respective roles of radiolytic species in alteration mechanisms. The radicals would be mainly involved during the fuel surface oxidation and H2O2 during the studtite precipitation.

FR-10.3:IL05  Radiation Damage and Corrosion of Single and Multi-phase Amorphous and Crystalline Solids
ANAMUL HAQ MIR, MIAMI Irradiation Facility, School of Computing and Engineering, University of Huddersfield, UK

Several different types of materials ranging from simple single phase amorphous to complex multi-phase amorphous and crystalline solids have been proposed as potential nuclear waste conditioning matrices. Such materials would need to maintain their physical, chemical, and mechanical integrity for hundreds of thousands of years in a geological disposal facility (GDF). Such materials will experience radiation damage due to the presence of radioactive elements and corrosion from the underground fluids over time. When compared to the original as-synthesised material, the coupled effects of radiation damage, gas generation and underground corrosion introduce uncertainties in their performance. Understanding and quantifying the impact of such changes is an important step in developing the safety case for the GDF and research is underway around the world on the subject. Over the years, we have studied several different types of materials and evaluated their radiation and corrosion resistance using ex-situ and in-situ ion irradiation facilities and controlled corrosion experiments. This talk will give an overview of this work highlighting our current understanding and some scientific issues regarding the performance of these materials when subjected to extreme conditions of a GDF.

FR-10.4:IL02  Production of Mixed Actinide Oxide Microparticle Reference Materials – Materials Science Aspects and Challenges
S. NEUMEIER, P. Kegler, S.K. Potts, M. Klinkenberg, D. Bosbach, I. Niemeyer, Forschungszentrum Jülich GmbH, Institute of Energy and Climate Research - Nuclear Waste Management and Reactor Safety (IEK-6), Juelich, Germany; S. Hammerich, Heidelberg University, Institute of Earth Sciences, Heidelberg, Germany

The analysis of individual micrometre-sized particles collected by IAEA safeguards inspectors on swipe samples during in-field verification activities requires the implementation of a sustainable quality control system incl. the development of suitable reference materials. In this context, pure U-oxide as well as doped reference microparticles, e.g. with lanthanides (Ln), Th and Pu are of special interest. To this end, the production of pure and Ln-doped uranium-oxide based microparticles utilising an aerosol-based particle production process will be presented with a focus on the discussion of material science aspects. The process allows for the production of pure and doped microparticles with monodisperse particle size distribution according to the IAEA’s requirements. µ-Raman and XAS measurements reveal the formation of U3O8 microparticles with a minor fraction of schoepite phase indicating ageing of these particles during storage under laboratory environment. Results from systematic shelf-life and structural investigations will be discussed allowing for a refined understanding of the structural and chemical stability and therefore, of the storage and performance conditions of these particles to assure their long-term applicability as nuclear reference materials.

FR-10.4:L03  Synthesis of Silicotitanates for Sr Sorption – Influence of the Ti/Si Ratio
T. Tratnjek1, J. Causse1, A. Hertz2, X. Deschanels1, 1Univ Montpellier, ICSM, CEA, CNRS, ENSCM, Marcoule, France; 2CEA Marcoule, DES/ISEC/DE2D/SEAD/LPSD, Marcoule, France

In this work, crystalline silicotitanates (CST) nanomaterials of sitinakite, Na2Ti2O3SiO4•2H2O, which can be used for Sr2+ removal from aqueous media, were synthesized and characterized. XRD analyses demonstrated that with the increase of the Ti/Si ratio in the starting gel, an increase of the crystallinity was observed, up to a point where sodium nonatitanate (Na4Ti9O20) formation was observed. N2 adsorption/desorption analyses displayed specific surface areas ranging between 56 m2.g-1 and 290 m2.g-1. Extraction kinetics experiments showed that above an estimated value ranging between 50 and 100 m2.g-1, the specific surface area does not influence the sorption rate of the CST anymore. The main conclusions from the sorption experiments are that, without Ca2+, both CST and nonatitanate present high sorption capacities of about 200 mg.g-1. When Ca2+ is added, a drastic drop of the sorption capacities was observed for the nonatitanate, which was also observed for all the CST in a much less pronounced way. Hence, by varying the Ti/Si ratio in the precursor gel, it is possible to tailor the CST’s properties. Furthermore, the CST showed a high affinity towards Sr2+, even when a competitor cation (Ca2+) was added in excess, which makes it a promising candidate for 90Sr decontamination.

FR-10.4:IL04  Advanced Nuclear Waste Form Research at UCI
S.C. Finkeldei, University of California, Irvine, Department of Chemistry, Irvine, CA, USA

A summary of recent progress in materials chemistry involved in the synthesis and characterization of ceramic nuclear waste forms will be presented. Utilization of wet-chemical, innovative synthesis approaches in combination with a wide range of characterization tools, facilitate a better understanding of the structure-property relationships of advanced nuclear waste form candidates. This knowledge is applied to tailor and fine-tune materials with targeted functionality that relies on a well-defined set of physical properties. Several classes of materials will be discussed, including complex oxides as potential nuclear waste forms. Insights into disorder, stress and strain of radiation damaged pyrochlores as potential waste forms for, e.g. Plutonium will be presented. Moreover, research on the influence of cladding and waste container material towards the UO2 redox chemistry under repository relevant conditions will be discussed. Ongoing efforts at UCI are supported by the recently established cluster of nuclear chemistry laboratories in combination with the TRIGA reactor and characterization facilities and will be briefly introduced.

FR-10.4:IL05  Nanoporosity and Irradiation - New Perspectives for the Treatment of Radioactive Effluents?
J. LIN1, G. TOQUER1, C. GRYGIEL2, S. DOURDAIN1, Y. GUARI3, C. REY1, J. CAUSSE1, X. DESCHANELS1, 1ICSM, CEA, CNRS, ENSCM, Univ Montpellier, Marcoule, France; 2CIMAP, CEA-CNRS-ENSICAEN-UNICAEN, Caen, France; 3ICGM, Univ Montpellier, CNRS, ENSCM, Montpellier, France

Considering their large interfacial surface, nanoporous materials offer interesting perspectives for the study of the evolution of damage induced by irradiation [1]. In order to study this phenomenon, thin films and powders of mesoporous silica (SBA15, MCM41) produced by sol-gel process were respectively irradiated with ions (Au, Xe…) and electrons (0.5 - 2 MeV). Different techniques have been implemented (BET / BJH, SAXS, RRX, microscopies, IR, NMR, etc.) to characterize the porous network as well as the silica walls of these materials according to the irradiation conditions (fluence, particle energy). In all cases, significant compaction of the porous network was observed, inducing a collapse of the mesoporosity [2-3]. The presentation aims to discuss these different observations, and clarified the role of interfaces on the evolution of defects created by irradiation. From a technological point of view, mesoporous silica would allow at the same time the separation of the RadioNuclide using a selective organic function, and their encapsulation after collapse of the porosity by irradiation effect. This new concept envisaged for the management of contaminated effluents, sometimes called "separation - conditioning", would result in obtaining a primary wasteform matrix.
[1] P. Makowski, X. Deschanels, A. Grandjean, D. Meyer, G. Toquer and F. Goettmann, New J. Chem., 36 (2012) 531.
[2] Y. Lou, S. Dourdain, C. Rey, Y. Serruys, D. Siméone, N. Mollard, X. Deschanels, Micropor. Mesopor. Mater., 251 (2017) 146.
[3] J. Lin, G. Toquer, C. Grygiel, S. Dourdain, Y. Guari, C. Rey, J. Causse, X. Deschanels, « Behavior of mesoporous silica under 2 MeV electron beam irradiation » Microporous Mesoporous Mater. 328 (2021) 111454.


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