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

Thermal conductivity is an important thermophysical property needed for effectively predicting fuel performance. As part of the Next Generation Nuclear Plant (NGNP) program, the thermal conductivity of tri-isotropic (TRISO) fuel needs to be measured over a temperature range characteristic of its usage. The composite nature of TRISO fuel requires that measurement be performed over the entire length of the compact in a non-destructive manner. No existing measurement system is capable of performing such a measurement. A measurement system has been designed based on the steady-state, guarded-comparative-longitudinal heat flow technique. The system as currently designed is capable of measuring cylindrical samples with diameters ~12.3-mm (~0.5″) with lengths ~25-mm (~1″). The system is currently operable in a temperature range of 400 K to 1100 K for materials with thermal conductivities on the order of 10 W/m/K to 70 W/m/K. The system has been designed, built, and tested. An uncertainty analysis for the determinate errors of the system has been performed finding a result of 5.5%. Finite element modeling of the system measurement method has also been accomplished demonstrating optimal design, operating conditions, and associated bias error. Measurements have been performed on three calibration/validation materials: SS304, 99.95% pure iron, and inconel 625. In addition, NGNP graphite with ZrO2 particles and NGNP AGR-2 graphite matrix only, both in compact form, have been measured. Results from the SS304 sample show agreement of better than 3% for a 300–600°C temperature range. For iron between 100–600°C, the difference with published values is < 8% for all temperatures. The maximum difference from published data for inconel 625 is 5.8%, near 600°C. Both NGNP samples were measured from 100–800°C. All results are presented and discussed. Finally, a discussion of ongoing work is included as well as a brief discussion of implementation under other operating conditions, including higher temperatures and adaptation for use in a glovebox or hot cell.

The high temperature reactor (HTR) requires a completely new fuel design as it operates at around 1000°C in normal conditions and can reach up to 1600°C in case of an accident. The fuel and its cladding consist fully of ceramic materials, which precludes the possibility of a core meltdown and thus ensures inherent safety. The integral part of all HTR core designs is the tristructural-isotropic (TRISO) particle, which encapsulates the fissionable materials in succeeding coatings of pyrolytic carbon and silicon carbide (SiC). An exceptional mechanical integrity of the silicon carbide layer in all conditions is required to ensure full fission product retention. Within this work simulated TRISO fuel has been fabricated by fluidized bed chemical vapour deposition and was annealed in protective atmosphere up to 2200°C for short durations. Subsequent mechanical tests showed only minor reductions in the fracture strength of the SiC up to 2000°C. Substantial weight loss and crystal growth were observed after annealing at 2100°C and above. Raman spectroscopy identified the formation of a multi-layered graphene film covering the SiC grains after annealing and scanning electron microscopy revealed significant porosity formation within the coating from 1800°C onwards. These observations were attributed towards an evaporation-precipitation mechanism of SiC at very elevated temperatures that only slightly diminishes the hardness, elastic modulus or fracture strength, but might still be problematic in respect to fission product retention of the SiC layer. The new technique of high temperature nanoindentation was applied to measure the elastic modulus and hardness of SiC in-situ up to 500°C in argon atmosphere. The elastic modulus was found to be only slightly reduced over the measurement range, while the hardness showed a significant drop. Investigations of the deformation zone beneath the indenter tip executed by transmission electron microscopy showed slip and deformation twinning. On specimens that had been subject to neutron irradiation an irradiation hardening effect was noted. The elastic modulus showed only a minor increase compared with the non-irradiated samples. Oxidation experiments were carried out in air up to 1500°C. Analysis of the oxidation layer showed the formation of amorphous silica and cristobalite for the highest temperatures.

An advanced concept high temperature reactor (HTR) design has been proposed - The ‘U-Battery’, which utilises a unique sealed coolant loop, and is intended to operate with minimal human oversight. In order to reduce the need for moving parts within the design, CO2 has been selected as a candidate coolant, potentially allowing a naturally circulated system. HTR fuel is held within a semi-graphitic fuel-matrix material, and this has not previously been tested within a CO2 environment. Graphite in CO2 is subject to two oxidation reactions, one thermally driven and one radiolytically. As such, the oxidation performance of fuel-matrix material has been tested within CO2 at both high temperatures and under ionising radiation within a sealed-system. Performance has been compared to that of the Gilsocarbon and NBG-18 nuclear graphite grades. Gilsocarbon is the primary graphite grade used within the currently operating AGR fleet within the UK, and as such is known to have acceptable oxidation performance under reactor conditions. NBG-18 is a modern graphite grade, and is a candidate material for use within the U-Battery. Virgin characterisation of all materials was performed, including measurements of bulk mass and volume, skeletal volumes and surface areas. High-resolution optical microscopy has also been performed and pore size distributions inferred from digital image analysis. All results were seen to agree well with literature values, and the variation between samples has been quanti- fied and found to be < 10% between samples of Gilsocarbon, and < 4% for samples of fuel-matrix and NBG-18. Thermal performance of fuel-matrix material was observed between 600 °C – 1200 °C and seen to be broadly comparable to that of the nuclear graphite grades tested. NBG-18 showed surprisingly poor performance at 600°C, with an oxidation rate of 3×10−4%/min, approximately ten times faster than Gilsocarbon in similar conditions, and three times faster than fuel-matrix material. The radiolytic oxidation performance of fuel-matrix material and NBG-18 has been observed by irradiating sealed quartz ampoules. Ampoules were pressurised with CO2 prior to irradiation, and the pressure after 30 days of irradiation was measured and seen to fall by 50%. Radiolytic oxidation, and the subsequent radiolysis of the reaction product, CO, was seen to cause significant carbonaceous deposition on the internal surfaces of the ampoule and throughout the samples. Due to the short irradiation times available in the present study, an investigation of the microporosity within irradiated samples has been carried out, using nitrogen adsorption and small-angle neutron scattering (SANS). Pore size distributions produced from SANS show the closure of microporosity within NBG-18, most likely as a result of low-temperature neutron irradiation.As a result of this work, CO2 is no longer a candidate coolant for use with the U-Battery design, due to the rapid deposition observed following irradiation.

The Liquid Salt Cooled Reactor (LSCR) provides several potential benefits compared to pressurized water-cooled reactor systems. These include low operating pressure of the liquid salt coolant, the high burnup tolerance of the fuel, and the high operating temperatures which leads to increases in efficiency. However, due to inherently low heavy metal loading, the fuel cycle design presents specific challenges. In order to study options for optimizing the fuel design and fuel cycle, SCALE6.1 was used to create simplified models of the reactor and look at various parameters. The primary parameters of interest included packing factor and fuel enrichment. An economic analysis was performed on these results by developing a simple fuel cycle cost (FCC) model that could be used to compare the different options from an economic standpoint. The lithium enrichment of the FLiBe coolant was also investigated. The main focus was to understand the practical limitations associated with the Li-7 enrichment and whether it could be used for beneficial purposes. The main idea was to determine whether a lower-than-equilibrium enrichment could be used at reactor start up so that the Li-6 isotope acts as a burnable absorber. The results for the lithium enrichment study showed that the enrichment converges over time, but the amount of time required to reach steady state is much too long and the FLiBe coolant could not be utilized for reactivity control as a burnable absorber. The results found through this research provide reasonable guidelines for expected costs and narrow down the types of configurations that should be considered as fuel design options for the LSCR. Additionally, knowledge was gained on methods for modeling the system not only accurately but also efficiently to reduce the required computing power and time.

Thermal conductivity is an important thermophysical property needed for effectively predicting nuclear fuel performance. As part of the Next Generation Nuclear Plant (NGNP) program, the thermal conductivity of tri-isotropic (TRISO) fuel needs to be measured over a temperature range characteristic of its usage. The composite nature of TRISO fuel requires that measurement be performed over the entire length of the compact in a non-destructive manner. No existing measurement system is capable of performing such a measurement. A measurement system has been designed based on the steady-state, guarded comparative-longitudinal heat flow technique. The system is capable of measuring cylindrical samples with diameters ∼12.3 mm (∼0.5 in.) with lengths ∼25 mm (∼1 in.). The system is currently operable in a temperature range of 100-700°C for materials with thermal conductivities on the order of 10-70 W*m-1*K-1. The system has been designed, built, and tested. An uncertainty analysis for the determinate errors of the system has been performed finding a result of 6%. Measurements have been performed on three calibration/validation materials: a certified glass ceramic reference material, 99.95% pure iron, and Inconel 625. The deviation of the validation samples is < 6-8% from the literature values. In addition, surrogate NGNP compacts and NGNP graphite matrix-only compacts have been measured. The results give an estimation of the thermal conductivity values that can be expected. All the results are presented and discussed. A Finite Element Analysis was done to compare the accuracy of multiple effective conductivity models. The study investigated the effects of packing structure, packing fraction, matrix thermal conductivity, and particle heat generation. The results show that the Maxwell and the Chiew & Glandt models provide the most accurate prediction of the effective thermal conductivity of the TRISO fuel compacts. Finally, a discussion of ongoing work is included as well as the possibility of correlating effective thermal properties of fuel compacts to their constituents with measurements of well-defined samples.

Dans un souci de réduction du volume de déchets nucléaires et de revalorisation des matières combustibles, une stratégie de gestion du combustible des réacteurs à haute température (HTR) est développée dans cette étude. La réduction de volume passe par la séparation des particules TRISO hautement radioactives et du graphite faiblement radioactif (les deux étant réunis dans un assemblage de combustible appelé "compact") tandis que le recyclage total nécessite la séparation du coeur de la particule, valorisable, et de sa gangue, déchet ultime. Les méthodes de séparation doivent préserver l'intégrité des TRISO afin d'empêcher la fuite des radioéléments. Ainsi, le traitement de choc thermique entre l'azote liquide et l'eau chaude permet une division partielle des compacts mais ne permet de récupérer que peu de particules. L'érosion du graphite par jet d'eau à haute pression présente le risque de fracturer les particules. La combustion totale du carbone libère toutes les billes. Le traitement des compacts par les ultrasons dans l'eau érode le graphite en fonction de l'intensité de travail, des direction et distance d'attaque, de la température et du gaz de saturation, nettoyant les particules. L'attaque acide des compacts par un mélange H2O2 + H2SO4 provoque l'intercalation du graphite par l'acide, faisant gonfler la structure et libérant ainsi les billes intactes. Les TRISO d'une part et leurs gangues d'autre part ont ensuite été vitrifiées par frittage de manière à obtenir une forte densité, jusqu'à un taux de 25% vol. Enfin, la lixiviation des composites dans l'eau ultrapure à 90°C montre de fortes propriétés de confinement.

This document gives an overview of the proposed MSc study. The main goal of the study is to model the cases listed in the code benchmark study of the International Atomic Energy Agency CRP–6 fuel performance study (Verfondern & Lee, 2005). The platform that will be employed is the GETTER code (Keshaw & van der Merwe, 2006). GETTER was used at PBMR for the release calculations of metallic and some non–metallic long–lived fission products. GETTER calculates the transport of fission products from their point of fission to release from the fuel surface taking into account gas precursors and activation products. Results show that for certain experiments the codes correspond very well with the experimental data whilst in others there are orders of magnitude differences. It can be seen that very similar behaviour is observed in all codes. Improvements are needed in updating the strontium diffusion coefficient and in understanding, on a deeper level, the transport of silver in TRISO particles and how it deviates from simple diffusion models. / Thesis (M.Sc. Engineering Sciences (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2012.

This document gives an overview of the proposed MSc study. The main goal of the study is to model the cases listed in the code benchmark study of the International Atomic Energy Agency CRP–6 fuel performance study (Verfondern & Lee, 2005). The platform that will be employed is the GETTER code (Keshaw & van der Merwe, 2006). GETTER was used at PBMR for the release calculations of metallic and some non–metallic long–lived fission products. GETTER calculates the transport of fission products from their point of fission to release from the fuel surface taking into account gas precursors and activation products. Results show that for certain experiments the codes correspond very well with the experimental data whilst in others there are orders of magnitude differences. It can be seen that very similar behaviour is observed in all codes. Improvements are needed in updating the strontium diffusion coefficient and in understanding, on a deeper level, the transport of silver in TRISO particles and how it deviates from simple diffusion models. / Thesis (M.Sc. Engineering Sciences (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2012.