Andrey Bukaemskiy

Synthesis and Properties of Reaction-Bonded SiC Ceramic with Embedded UO2 - TRISO Coated Particles

Ceramics based on SiC are characterized by extreme hardness, high thermal conductivity, relatively low thermal expansion and chemical durability. In principle, SiC ceramics can be considered as a long-term stable matrix for final disposal of radioactive waste, such as coated fuel particles (CP) separated from the graphite matrix from spent HTR (high-temperature gas-cooled reactor) fuel pebbles. In the present work, SiC-based ceramic with the embedded UO2 - TRISO (tristructural-isotropic) coated particles was synthesized by the reaction-bonding process. The synthesis was performed in standard SiC crucible. Several physico-mechanical properties of the synthesized samples were investigated. It was shown that the coated particles in the reaction-bonded silicon carbide (RBSiC) matrix are distributed homogeneously. The amount of pores is insignificant and the crippling of the coated particles is not observed. Besides, the junction between CP and RBSiC matrix and between RBSiC matrix and the SiC crucible is very good. For all ceramic components of the synthesized samples, namely, for the UO2-kernels, SiC-layers, SiC crucible wall and for the synthesized RBSiC ceramics, the values of microhardness and fracture toughness were measured and compared with the reference data. The strength properties, such as tensile strength of the synthesized samples, failure mechanism of the reaction-bonded SiC ceramic with embedded UO2 - TRISO coated particles, microstructure of the fracture surfaces, the peculiarities of the coated particles fracture were investigated in detail. Moreover, the diffusion of radioactive tracers (137Cs, 241Am, 36Cl, 3H) across synthesized ceramic was studied and the high safety characteristics of the synthesized ceramic were demonstrated.

Reprocessability of molybdenum and magnesia based inert matrix fuels

Abstract This work focuses on the reprocessability of metallic 92Mo and ceramic MgO, which is under investigation for (Pu,MA)-oxide (MA = minor actinide) fuel within a metallic 92Mo matrix (CERMET) and a ceramic MgO matrix (CERCER). Magnesium oxide and molybdenum reference samples have been fabricated by powder metallurgy. The dissolution of the matrices was studied as a function of HNO3 concentration (1-7 mol/L) and temperature (25-90°C). The rate of dissolution of magnesium oxide and metallic molybdenum increased with temperature. While the MgO rate was independent of the acid concentration (1-7 mol/L), the rate of dissolution of Mo increased with acid concentration. However, the dissolution of Mo at high temperatures and nitric acid concentrations was accompanied by precipitation of MoO3. The extraction of uranium, americium, and europium in the presence of macro amounts of Mo and Mg was studied by three different extraction agents: tri-n-butylphosphate (TBP), N,Nʹ-dimethyl-N,Nʹ-dioctylhexylethoxymalonamide (DMDOHEMA), and N,N,N’,N’- -tetraoctyldiglycolamide (TODGA). With TBP no extraction of Mo and Mg occurred. Both matrix materials are partly extracted by DMDOHEMA. Magnesium is not extracted by TODGA (D < 0.1), but a weak extraction of Mo is observed at low Mo concentration.

Assessment of the release behaviour of 14C from irradiated nuclear graphite from a German research reactor

The behaviour of irradiated graphite (i-graphite) from a German research reactor was investigated under different boundary conditions for assessment of 14C-speciation and the release kinetics into the aqueous and gas phase. The results showed that most of released 14C remains in the solution. Under near neutral conditions, volatile 14C is predominantly in form of CO2 (i.e. 90%) with some minor fraction of 14COu2009+u200914Corg. Almost negligible 14C release into the gas phase was detected under cementitious conditions, although the total 14C release increases. Evaluated release rates of 14C are discussed in the context of i-graphite disposal in the GDF Schacht Konrad.

Fixation of Actinides in Zirkonium Based Ceramics

In recent years, various fuel types or targets, free from U-238, have been proposed for the transmutation of plutonium and the minor actinides. Yttrium-stabilised zirconia is currently under investigation because of its small neutron cross section and is high solubility with PuO2 and AmO2 . The aim of the present work is to investigate the behaviour of Yttrium Stabilised Zirconium Oxide in presence of Cerium, as simulate of minor actinides. Several oxide materials composed of Zr, Y, Ce (5–25%), have been synthesised by coprecipitation method at different calcination temperature and the products have been characterised by Optical Microscopy (OM), Thermogravimetry (TG) coupled with Thermal Differential Analysis (DTA), X-ray Line Diffraction (XRD), Scanning Electron Microscopy (SEM) Energy Dispersive X-Ray analyser (EDX) etc. The investigations show that, the structure of the powder, characterised by the phase condition, the lattice parameter, the crystalline size and the lattice distorsion, varies with the CeO2 content, and has an influence on the sinterability. “Mild” methods of powder preparation (drying, low-temperature calcination, wet grinding by attrition) were used before sintering. For all investigated Cerium concentrations, the optimal parameter of pressing and sintering were determined. For a wide Cerium concentration region, from 5 to 20 wt%, the density of compact materials is 92–95% of TD, the microhardness 13.5 GPa and the pellets have a good morphology, grains and pores being homogeneous distributed in material.Copyright