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


Session FK-10.1 - Waste Form Development
FK-10.1:IL01  Recent Advances in the Immobilization of Low- or Intermediate-level Radioactive Waste in Cementitious Materials

Cementitious materials designed for low- or intermediate-level radioactive waste solidification and stabilization usually include substantial amounts of Portland cement in their formulation, possibly blended with supplementary cementitious materials such as fly ash or blast furnace slag. Portland cement benefits from technology transfer from civil engineering research, and from 200 years of experience on its durability under various service environment. However, wastes produced by nuclear activities are very diverse and some of their components may chemically react with cement phases, aggregates or mixing water, thus reducing the quality of the final product. Over the last ten years, increasing attention has thus been paid to alternative inorganic binders. They may indeed offer new solutions for the conditioning of such deleterious wastes with a view to simplify the conditioning process, obtain cement-waste forms as inert as possible from the physical and chemical points of view and/or minimize the amount of packages to be produced. In this presentation, the focus is placed on alkali-activated binders, calcium sulfoaluminate and magnesium phosphate cements. These binders are briefly presented and their potential for waste conditioning is then discussed.

FK-10.1:IL02  Phosphate-based Glasses and Glass Ceramics for Immobilization of Lanthanides and Actinides
S.V. STEFANOVSKY, O.I. STEFANOVSKY, Frumkin Institute of Physical Chemistry and Electrochemistry RAS, Moscow, Russia; S.E. VINOKUROV, Vernadsky Institute of Geochemistry and Analytical Chemistry RAS, Russia

Russian HLW vitrification plant at “Mayak” plant uses phosphate-based glasses. Initially the reason of this selection was specific chemical composition of Russian HLW with high sodium and aluminum contents and the absence of calciner capable to treat inhomogeneous sludges. Moreover phosphate-based glasses have lower melting point and incorporate higher troublesome components (sulfates, chlorides, molybdates, and multivalent elements) than borosilicate glasses. However, they are “shorter” and have higher tendency to devitrification as compared to borosilicate glasses. At first Russian HLW glasses were on sodium-aluminophosphate basis with composition approximately (mol.%) 40 Na2O, 20 Al2O3, 40 P2O5. Recently we have modified this composition by replacing of up to 50% Al2O3 with Fe2O3. Such a replacement increases chemical durability and resistance to devitrification and radiation of the glasses. The structure of the glasses and speciation of Fe, La, Ce, Nd, Eu, Gd, U, Np, and Pu were studied in details. An increase of Fe oxide content in the glasses provides increasing of Fe-bearing waste loading in the final waste form. The glasses and glass ceramics on sodium-aluminum-iron-phosphate basis are promising waste forms for lanthanide fraction of HLW and minort actinides.

FK-10.1:L03  Synthesis and Thermal Behavior of Thorium-incorporated Rhabdophane
DANWEN QIN, A. MESBAH, S. SZENKNECT, N. CLAVIER, N. DACHEUX, Institut de Chimie Séparative de Marcoule, Bagnols sur Cèze Cedex, France

Monazite minerals (LnPO4) are considered as potential ceramic wasteforms for the specific conditioning of actinides (III, IV). In this frame, thorium incorporation as CaxThxNd1-2xPO4 solid solution was successfully achieved through the precipitation of the hydrated rhabdophane (CaxThxNd1-2xPO4·nH2O). A multiparametric study was thus undertaken to determine the optimal conditions such as: starting stoichiometry, temperature, heating time, pH leading to single phase CaxThxNd1-2xPO4·nH2O powdered samples. The excess of calcium appeared to be a prevailing factor with a suggested initial Ca:Th ratio of 10:1. Similarly, the recommended heating time should exceed 4 days while the optimal temperature of synthesis was 110 °C. The conversion at 1100°C under air during 6 hours of the rhabdophane samples conducted to the formation of pure CaxThxNd1-2xPO4 compounds. Coupled TGA, dilatometry and PXRD study allowed the identification of the different dehydration and the transformation steps leading to the anhydrous monazite-cheralite compounds showing a clear impact of thorium insertion in the structure.

FK-10.1:IL04  Immobilization of Fission Products in Glass and Glass Ceramic Matrices

Radioactive waste vitrification has been carried out industrially in several countries for nearly 40 years. Research is still continuing in order to safely condition new types of wastes and develop new technologies. Waste are integrated in the vitreous network by reactions occurring in vitrification furnaces set to operate at high temperature. The state (solid, super-cooled liquid, liquid) of the precursor and waste microstructures evolve with time and temperature until a high-temperature homogeneous liquid is formed. During the different steps of the process, elements are integrated into the glass structure up to their incorporation limit. The waste loading can be limited in certain cases in order to obtain a homogeneous glass at the microscopic scale or exceed by using macroscopically homogeneous vitro-ceramic matrices in order to contain larger quantities of waste. This talk will be focus on academic research conducted to understand the phenomenology behind the formation of the glass and glass-ceramic and its evolution after cooling from atomic to macroscopic scale. Following a general introduction on French nuclear glass waste forms, new results help to better explain and control the mechanisms of crystallization, phase separation and diffusion during nuclear glass synthesis will be present.

FK-10.1:L06  Thorium Incorporation in the Xenotime Based Ceramic
A. MESBAH1, N. CLAVIER1, S. SZENKNECT1, J. LOZANO-RODRIGUEZ2, N. DACHEUX1, 1ICSM, UMR 5257 CNRS - CEA - ENSCM - Université de Montpellier, Site de Marcoule - Bat 426, Bagnols/Cèze, France; 2HZDR, Institute of Resource Ecology, the Rossendorf Beamline at ESRF, Grenoble, France

Phosphate cceramics are considered as a promising candidates for the specific conditioning of actinides (III, IV). This fact comes from their easy way of preparation and high chemical durability. The incorporation of actinides in such phases has been extensively studied mostly over solid-state routes, however, in the recent years the use of wet chemistry methods was developed leading to the formation of homogeneous compounds. Therefore, solid solutions of Er-xThx(PO4)1-x(SiO4)x were obtained in application of the method developed to prepare pure coffinite (USiO4), i.e. hydrothermal conditions at 250°C for 7 days. The structure crystallizes in the zircon-type structure (I41/amd group, tetragonal system) as observed for the end-members ThSiO4 and ErPO4. Er-xUx(PO4)1-x(SiO4)x analogues were also synthesized in comparable conditions; Both systems were thoroughly characterized by PXRD, EXAFS, FT-IT and µ-Raman and showed an exciting structural properties.

Session FK-10.2 -  Challenging Waste Constituents
FK-10.2:IL01  X-Ray Diffraction and Adsorption Spectra Reveal Zr and Ti Coordination Environment in Actinides Immobilization by Glass-Ceramics
CHANGZHONG LIAO, KAIMIN SHIH, Department of Civil Engineering, The University of Hong Kong, Hong Kong SAR, China

Management of high-level radioactive waste, especially minor actinides and Pu, is a worldwide problem that challenges the development of nuclear energy. Zirconolite-based glass-ceramic has been considered as a promising candidate for minor actinides and Pu immobilization. To obtain zirconolite-based glass-ceramics with satisfactory properties, the crystallization mechanisms of hosting phases during thermal treatment require detailed investigation. In this study, a glass-ceramic matrix with zirconolite as a single crystalline phase has been successfully synthesized with the SiO2-Al2O3-CaO-TiO2-ZrO2-Na2O-Nd2O3 system via a two-step thermal treatment scheme. X-ray diffraction results show that cubic-zirconia was precipitated at the first stage and then transformed to the zirconolite phase during the later heat treatment. Results from X-ray absorption spectra (XAS) reveal that there is no considerable change in the local environments of Zr and Ti atoms after nucleation. The XAS results also show that Zr existed as ZrO6, ZrO7 and ZrO8, while TiO4, TiO5 and TiO6 co-exist in the parent glass and nucleated sample. Crystallization process can lead to the change of Zr- and Ti-species, such as converting ZrO6 to ZrO7 and ZrO8 and converting TiO4 to TiO5 and TiO6.

FK-10.2:IL02  Recovery of Actinides from Nuclear Waste Using Pyro-electrochemical Process
WEIQUN SHI, Laboratory of Nuclear Energy Chemistry, Institute of High Energy Physics,Chinese Academy of Sciences, Beijing, China

Pyrometallurgical process is one of the most promising options for the reprocessing of advanced nuclear fuels and transmutation blankets which are notable to possess high burn-up and high content of Pu and minor actinides. In a typical pyrometallurgical process, actinides(An) are anodically dissolved in LiCl-KCl eutectic along with the active fission products. At the same time, predominant uranium is recovered onto a solid stainless steel cathode, whilst plutonium and minor actinides are deposited together at the liquid cadmium cathode (LCC). As significant amount (~6% wt) of Ln remains in the LCC, according to previous investigations, the deposition potential disparity of An and Ln on the solid Al cathode are much larger than those on other active solid or liquid cathodes, and therefore the separation of An from Ln by using a solid Al electrode should be more efficient. Keeping this in mind, the separation of An from Ln by the co-reduction of An cations and Al3+ was studied in the present work. The results showed that all Ln3+, Th4+and U4+ could be co-reduced with Al3+ through forming various Ln-Al intermetallic compounds. Efficient separation of An over Ln can be achieved with acceptable extraction efficiency. Conceptual process design based on this approach is undergoing.

FK-10.2:IL04  Removal of Noble Metals from High Level Liquid Waste by Silica-based Anion Exchangers
YUEZHOU WEI, X. WANG, S. NING, Guangxi University, Nanning, China; Y. WU, Q. ZOU, Shanghai Jiao Tong University, Shanghai, China

High level liquid waste (HLLW) generated from reprocessing of spent nuclear fuel by the PUREX process contains significant amount of fission product noble metals (Pd, Ru, Rh, Tc). They tend to form separate phases as metallic or alloy state during the vitrification process of HLLW and cause deterioration in the stability of the glasses. On the other hand, these noble metals may be potentially important resources as industrial materials in the future. Therefore, separation of these noble metals from HLLW is of great significance. To selectively remove these noble metals from HLLW, we have developed an advanced ion exchange process by using novel silica-based anion exchangers, which were synthesized by immobilizing functional organic resins in porous silica particles with a mean diameter of 60 μm and pore size of around 50-600nm. This new type of anion exchangers has fast diffusion kinetics, improved chemical stability and low pressure drop in a packed column. Adsorption and separation behavior of Pd, Ru, Rh and Tc with different oxidation state in nitric acid solution was studied experimentally and theoretically. Small scale separation tests using simulated and actual HLLW solutions were carried out to verify the feasibility of the proposed process.

Session FK-10.3 - Waste Form Performance Testing, and Characterization

FK-10.3:IL01  Modern Irradiation Testing Techniques to Simulate the Irradiation Performance of Waste Forms
S. PEUGET1, A.H. MIR1, S. MIRO2, Y. SERRUYS2, I. MONNET3, C. JEGOU1, 1CEA, DEN, DE2D, SEVT, LMPA, Laboratoire d’Étude des Matériaux et Procédés Actif, Bagnols-sur-Cèze, France; 2CIMAP-GANIL (CEA-CNRS-ENSICAEN-Univ. Caen), Caen Cedex, France; 3CEA, DEN, Service de Recherches de Métallurgie Physique, Laboratoire JANNUS, Gif-sur-Yvette, France

Many materials used in nuclear industry in the space applications are subjected to simultaneous irradiation with many particles like, photons, electrons, ions and neutrons. In order to ensure and guarantee their long term structural integrity, it is important to experimentally simulate such complex irradiation ageing scenario. For Borosilicate glasses that are subjected to the simultaneous irradiation from beta and alpha decays, we have to understand if multi particle irradiation scenarios can alter the mechanisms of the damage formation and subsequent damage evolution from the one inferred from the single particle irradiations. To that purpose, mono, sequential and simultaneous beam irradiations involving alpha particles, heavy ions and electrons were carried out to understand the impact of the inelastic and elastic energy loss interaction on the damage formation in nuclear glasses. It is shown that under mutli-beam irradiations the final damage state of the glass is the result of the competiting effects of damage generation form heavy ions and partial damage recovery from alpha particle. Consequently, the accurate simulation by external irradiations of the complex irradiation ageing scenario of a nuclear glass must be performed by multi-beam irradiation experiments.

FK-10.3:IL02  Impact of Near Field Evolution on the Stability of Vitrified Waste and Spent Nuclear Fuel
K. LEMMENS, C. CACHOIR, K. FERRAND, T. MENNECART, S. CAES, E. VALCKE, Belgian Nuclear Research Centre, Mol, Belgium

In designs for geological disposal of vitrified nuclear waste and spent nuclear fuel, the waste form is in general separated from the host rock by engineered barriers. These engineered barriers and the disturbed/damaged host rock zone is called the near field. This near field will be affected both by the far field (the host rock) and by the waste form. Hence, its physicochemical properties will change with time. These changes can again have an impact on the durability of the waste form. The impact will be different for vitrified waste, whose borosilicate matrix is relatively soluble, and for spent nuclear fuel, whose UO2 matrix is very insoluble under the reducing conditions expected in situ. We give an overview of possible interactions and the effects they may have for the life time of the waste matrices, focusing on a near field that is dominated by concrete with an initially high pH. An important evolution in such design is the expected pH decrease at the interface with the waste form, depending on the porosity and crack evolution of the concrete. The pH evolution is also important for the life time of any carbon steel barriers, aiming at delaying waste form-water contact.

FK-10.3:IL03  Synchrotron-based Three-dimensional X-ray Imaging of Crystalline Ceramic Waste Form Materials
WILSON K.S. CHIU, Department of Mechanical Engineering, University of Connecticut, Storrs, CT, USA

Advances in materials increasingly rely on the 3-D characteristics of its microstructure and chemistry at the nanoscale. This invited lecture will review 3-D imaging techniques that measure microstructural and chemical properties, and discuss its influence on material function and reliability. The presentation will then focus on three-dimensional microstructural imaging methods for crystalline ceramic waste form materials. Two 3-D imaging modalities using x-ray absorption and fluorescence-based synchrotron-based transmission x-ray microscopy, will be demonstrated by characterizing the 3-D structural and chemical distribution. Challenges and opportunities for future work will then be discussed. The goal of this research will be to obtain a scientific and engineering understanding into how microstructure-induced transport mechanisms govern performance, with a long-term goal to improve current materials and create new materials that will enable improved device performance and increased long-term reliability.

Session FK-10.4 - Waste Immobilization Facilities and Repository Design
FK-10.4:IL01  International Experience in Radioactive Waste Vitrification
M.I. OJOVAN, R.A. ROBBINS, International Atomic Energy Agency (IAEA), Vienna, Austria

Vitreous materials are the overwhelming world-wide choice for the immobilisation of HLW resulting from nuclear fuel reprocessing due to glass tolerance for the chemical elements found in the waste as well as its inherent stability and durability. Vitrification is a mature technology and has been used for high-level nuclear waste immobilization for more than 50 years. Borosilicate glass is the formulation of choice in most applications although other formulations are also used e.g. phosphate glasses are used to immobilize high level wastes in Russia. The excellent durability of vitrified radioactive waste ensures a high degree of environment protection. Waste vitrification gives high waste volume reduction along with simple and cheap disposal facilities. Although vitrification requires a high initial investment and then operational costs, the overall cost of vitrified radioactive waste is usually lower than alternative options when account is taken of transportation and disposal expenses. Glass has proven to be also a suitable matrix for intermediate and low-level radioactive wastes and is currently used to treat legacy waste in USA, and NPP operational waste in Russia and South Korea. This report is also outlining IAEA activities aiming to support utilisation of vitreous material for nuclear waste immobilization.

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