A safe treatment and disposal of spent Tri-Structural Isotropic (TRISO) coated fuel particles is one of the most important issues for developing the next generation of nuclear reactors, such as a Very High Temperature Reactor (VHTR). The project investigates the encapsulation of surrogated TRISO particles in Glass-Graphite Composite (GGC) and in Alumina Borosilicate Glass (ALBG) and compares their geological performance in the repository. The study deals with the assessment and performance of both matrices in a geological repository's conditions, measuring their chemical durability for 28 days at temperatures ranging 25-90°C and using American Standard for Testing Material (ASTM-C1220-98). The leach test revealed that only sintered ALBG with TRISO particles doped in cesium oxide could provide a safe Engineering Barrier System (EBS). The thermal property of the matrices was examined by measuring their thermal diffusivities. The thermal diffusivity of ALBG bearing various proportions of TRISO particles was measured experimentally using Laser Flash Analysis (LFA). The experimental results validated through a numerical method using Image Based Modelling (IBM). The effect of the porosity in decreasing the thermal diffusivity of TRISO particles was also discussed. In addition, the study deals with the immobilisation of ion exchange resins (doped with radioactive and non-radioactive cesium and cobalt) in borosilicate glass. The thermal analysis revealed that a successful immobilisation could be achieved once the sulfur functional group in the resin was decomposed and evaporated in a form of SO2/SO. The minimum required temperature of the heat treatment was 500°C under air environment as a pre-conditioning stage before immobilisation.

The basic fuel unit of the High Temperature Reactor (HTR) of the Pebble Bed Modular Reactor (PBMR) is a uranium dioxide kernel coated with a buffer layer, an inner pyrolytic carbon (IPyC) layer, a silicon carbide (SiC) layer and an outer pyrolytic carbon (OPyC) layer and is commonly referred to as a TRISO particle. Thousands of these micro-spheres are embedded in a graphite matrix and pressed to form a fuel sphere. During the manufacture of the TRISO particles and the fuel spheres there is a production of TRISO particles with cracked/broken layers, especially the SiC layer. Before the irradiation of the fresh fuel in the nuclear reactor it is of the utmost importance to quantify the failed fractions in fresh fuel as this information is very useful in the general understanding of fuel behaviour, calculation of risk and safety margins, and prediction of long term fuel behaviour. For this reason the burn-leach method has been applied for the quality control of the fresh fuel. In this work, several aspects of the burn-leach method that affect the efficiency of the method were studied. Aspects that were investigated are: qualitative aspects, layer properties, quantitative aspects, variants of the burn-leach method and lastly statistical information from the burn-leach data. The results obtained were as follows: Studies in this dissertation suggest that partial leaching of uranium in TRISO particles with a defective SiC layer was a phenomenon that exists. Although UO2 kernel equivalents were successfully determined by burn-leach method for particles with fully broken SiC layers, certain particles leached uranium amounts that did not correspond to single UO2 kernel equivalents; Evidence of occurrences of ‘slow leaching’ in an acidic medium were evident for certain particles. There were remnants of uranium dioxide kernels that had been partially leached after the full 16 hours. This behaviour led to inconclusive results on the absolute number of defective particles in a given population; Investigations suggest that there is at least circumstantial evidence that the BL method combined with X-ray tomography provides information about the integrity of the SiC layer, and why one particle leaches and the other does not. Neither the burn-leach nor the leach-burn-leach analysis is sufficient to be used as a stand-alone method to quantify the number of particles with defective SiC layers in a given TRISO particle population. The two tests need to be coupled to other techniques such as high resolution tomography for an extensive quantification of the layer defects; Burn-leach has to be designed to test for the layer integrity on a microscopic level as opposed to testing for the broken shells only, as was done by the normal burn-leach based on the German program. The leach time was not sufficient in its present form; Burn-leach results indicated that oxidation times of 96 hours at 750 °C under atmospheric pressure did not negatively affect the mechanical strength of the silicon carbide layer of freshly-manufactured TRISO particles, as these particles did not have a high failure fraction. / Dissertation (MSc)--University of Pretoria, 2013. / Materials Science and Metallurgical Engineering / Unrestricted

In modern high temperature nuclear reactors, silicon carbide (SiC) is used as the main diffusion barrier for the fission products in coated fuel spheres called TRISO particles. In the TRISO particle, pyrolytic carbon and SiC layers retain most of the important fission products like xenon, krypton and cesium effectively at temperatures up to 1000 oC. Previous studies have shown that 400 oC to 600 oC implantation of heavy ions into single crystal 6H-SiC causes the SiC to remain crystalline with many point defects and dislocation loops (damage). The release of Xe at annealing temperatures above 1400 oC is governed by the normal volume diffusion without any hindrance of trapping effects. In this study two phenomena in single crystal 6H-SiC implanted by 360 keV Xenon ions were studied using Rutherford Backscattering Spectroscopy (RBS) and channeling. Radiation damage and its annealing behavior at annealing temperatures ranging from 1000 oC to 1500 oC, and the diffusion of xenon in 6H-SiC at these annealing temperatures were investigated. 360keV xenon ions were implanted into a single crystalline wafer (6H-SiC) at 600 oC with a fluence of 1 × 1016 cm-2. The sample was vacuum annealed in a computer control Webb 77 graphite furnace. Depth profiles were obtained by Rutherford backscattering spectrometry (RBS). The same set-up was used to investigate radiation damage of the 6H-SiC sample by channeling spectroscopy. Isochronal annealing was performed at temperatures ranging from 1000 to 1500 °C in steps of 100 oC for 5 hours. Channeling revealed that the 6H-SiC sample retained most of its crystal structure when xenon was implanted at 600 °C. Annealing of the radiation damage took place when the sample was heat treated at temperatures ranging from 1000 oC to 1500 oC. The damage peak almost disappears at 1500 oC but the virgin spectrum was not achieved. This happened because of dechanneling due to extended defects like dislocations remaining in the implanted region. RBS profiles showed that no diffusion of the Xe occurred when the sample was annealed at temperatures from 1000 oC to 1400 oC. A slight shift of the xenon peak position towards the surface after annealing at 1400 °C was observed for 600 oC implantation. After annealing at 1500o C, a shift toward the surface accompanied by a broadening of the Xe peak indicating that diffusion took place. This diffusion was not accompanied by a loss of xenon from the SiC surface. The shift towards the surface is due to thermal etching of the SiC at 1400-1500 °C. Modern high temperature gas-cooled reactors operate at temperatures above 600 oC in the range of 750 oC to 950 oC. Consequently, our results indicate that the volume diffusion of Xenon in SiC is not significant in SiC coated fuel particles. / Dissertation (MSc)--University of Pretoria, 2014. / gm2014 / Physics / unrestricted