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Made available in DSpace on 2014-10-09T12:34:36Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:01:01Z (GMT). No. of bitstreams: 0 / Em instalações nucleares o controle do material nuclear é uma das atividades da maior importância. A Comissão Nacional de Energia Nuclear (CNEN) e a Agencia Internacional de Energia Atomica (AIEA) quando de suas inspeções rotineiras tem os dados fornecidos como um fator de segurança. Ter um sistema de controle de material nuclear que permita a qualquer momento reportar a quantidade e a localização dos diversos itens a serem inspecionados é um fator de primordial importância nos dias de hoje. Neste trabalho objetivou-se aprimorar um sistema já existente utilizando para seu desenvolvimento uma plataforma mais amigável através da linguagem de programação VisualBasic (Microsoft Corporation) para facilitar a equipe de operação do Reator IEA-R1 o fornecimento de dados que possibilitem o melhor controle dos materiais nucleares do Reator IEA-R1. Esses dados tem permitido o desenvolvimento de trabalhos a serem apresentados em congressos nacionais ou internacionais bem como em dissertações de mestrado ou teses de doutorado. O programa foi desenvolvido para atender as exigências das normas de salvaguarda da CNEN e da AIEA, mas suas funções podem ser ampliadas conforme as necessidades futuras. Este sistema poderá ser utilizado em outros reatores que por ventura sejam contruidos no pais, pois é bem pratico e sua utilização permite um um controle efetivo sobre o material nuclear da instalação. / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

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Carbon materials play a fundamental role in the development of fusion reactors, for both the generation of electric power and the production of nuclear materials. It is possible to synthesise graphite and carbon materials from coal. Coal is available in large quantities and could be used for the production of high-purity carbon graphite. However, it contains large quantities of impurities that need to be removed prior to graphitisation/carbonisation. The impurity levels of certain elements in this graphite must be kept at very low levels. Boron, which absorbs neutrons strongly, should be below 500 ppb. Europium and gadolinium, which absorb neutrons and are activated to highly radioactive products, as is cobalt, should be as low as 50 ppb. Lithium transforms to tritium, which leads to the circulating helium becoming radioactive. Other elements, such as calcium, sodium, silicon, thorium and uranium, should not be ignored. The purpose of this study was to lower or remove completely the impurities and trace elements in coal that affect the quality of nuclear-grade graphite. The organic part of Tshikondeni coal was dissolved in a solvent, dimethylformamide (DMF), on addition of sodium hydroxide. The first stage of purification is centrifugation and filtration, which removes most of the impurities. The recovered organic material, known as ‘Refcoal’, may be converted to graphitisable coke. Some elements, significantly boron and cobalt, associate with the organic material in solution and are not sufficiently separated by centrifugation and filtration. Further purification was employed during each process step in the conversion of coal solution into graphite. Different methods of purification were employed in this study. They included chlorination, acid treatment and the ion-exchange or complexation method. Chlorine gas and hexachlorocyclohexane (benzene hexachloride) were used in the chlorination method. Acids such as hydrochloric, hydrofluoric and ascorbic were used in acid treatment. In the ion-exchange method, reagents such as methane, starch, potassium cyanide, ethylene-diaminetetraacetic acid, sodium fluoride, sodium sulphate, ice, glycerol and sodium nitrate were used. All the treated Refcoal was coked at 1 000º C. Pyrolysis was applied in other methods with the aim of volatilising elements that form volatile halides at higher temperatures. Analysis was done for elements such as calcium, cobalt, europium, gadolinium, lithium, sodium, silicon, thorium and uranium, and other elements in the periodic table. Inductively coupled plasma mass spectroscopy and inductively coupled plasma optical emission spectroscopy were used to analyse the concentrations of the trace elements in the coal (treated and untreated) and the coked Refcoal. In inductively coupled plasma mass spectroscopy, microwave digestion and fusion were applied as methods of preparation. However, the instrumentation gave different results for the same sample. The results showed that specific methods work for specific elements. The chlorination method and the acid-treatment method (especially using hydrofluoric acid and hydrochloric acid) gave better purification for most of the trace elements and other elements. Better purification was achieved with elements such as, boron, calcium, europium, gadolinium, lithium, sodium and silicon. All the treatments failed to lower uranium and thorium to the level required for nuclear-grade graphite. However, uranium has a low boron equivalent and does not pose serious problems with respect to nuclear usage. All the methods failed to remove cobalt and this remains a problem. / Dissertation (MSc (Chemistry))--University of Pretoria, 2008. / Chemistry / MSc / unrestricted

Un des plus grands défis de la Physique aujourd’hui est de créer une source d’énergie propre, durable et efficace qui puisse satisfaire les besoins de la société actuelle et future avec le minimum d’impact sur l’environnement. Dans ce cadre, un grand effort de recherche internationale est dévoué à l’étude de nouveaux systèmes de production d’énergie ; réacteurs de fission de Génération IV et réacteurs de fusion nucléaire sont en particulier en train d’être développés. Les matériaux utilisés dans ces réacteurs seront soumis à des hauts niveaux de radiation, ce qui rend nécessaire l’étude de leur comportement sous irradiation pour permette le succès du développement de ces nouvelles technologies. Dans cette thèse, deux matériaux ont été étudiés : la silice amorphe (a-SiO₂) et l’oxyde de magnésium (MgO). Ces deux matériaux sont des oxydes isolants avec des applications dans l’industrie de l’énergie nucléaire. Des irradiations avec des ions de haute énergie ont été réalisées sur différentes plateformes d’accélérateurs d’ions pour induire l’endommagement de ces deux matériaux par irradiation ; ensuite, les mécanismes d’endommagement ont été caractérisés en utilisant, principalement, des techniques d’analyse par faisceau d’ions (techniques IBA).Un des objectifs de cette thèse était de développer la technique d’ionoluminescence (qui est une technique IBA très peu connue) et de l’appliquer à l’étude des mécanismes d’endommagement par irradiation des matériaux, démontrant alors le potentiel de cette technique. L’ionoluminescence de trois types différents de silice (avec des différentes teneurs en OH) a ainsi été étudiée en détail et utilisée pour décrire la création et l’évolution des défauts ponctuels sous irradiation. Dans le cas de MgO, l’endommagement produit par irradiation avec des ions Au⁺ à 1.2 MeV a été caractérisé en utilisant la technique de spectrométrie de rétrodiffusion Rutherford en configuration de canalisation et la diffraction des rayons X. Finalement, l’ionoluminescence de MgO sous différentes conditions d’irradiation a aussi été étudiée. Les résultats obtenus dans cette thèse aident à comprendre les processus d’endommagement par irradiation dans les matériaux, ce qui est indispensable pour le développement de nouvelles sources d’énergie nucléaire. / One of the most important challenges in Physics today is the development of a clean, sustainable, and efficient energy source that can satisfy the needs of the actual and future society producing the minimum impact on the environment. For this purpose, a huge international research effort is being devoted to the study of new systems of energy production; in particular, Generation IV fission reactors and nuclear fusion reactors are being developed. The materials used in these reactors will be subjected to high levels of radiation, making necessary the study of their behavior under irradiation to achieve a successful development of these new technologies. In this thesis two materials have been studied: amorphous silica (a-SiO₂) and magnesium oxide (MgO). Both materials are insulating oxides with applications in the nuclear energy industry. High-energy ion irradiations have been carried out at different accelerator facilities to induce the irradiation damage in these two materials; then, the mechanisms of damage have been characterized using principally Ion Beam Analysis (IBA) techniques. One of the challenges of this thesis was to develop the Ion Beam Induced Luminescence or ionoluminescence (which is not a widely known IBA technique) and to apply it to the study of the mechanisms of irradiation damage in materials, proving the power of this technique. For this purpose, the ionoluminescence of three different types of silica (containing different amounts of OH groups) has been studied in detail and used to describe the creation and evolution of point defects under irradiation. In the case of MgO, the damage produced under 1.2 MeV Au⁺ irradiation has been characterized using Rutherford backscattering spectrometry in channeling configuration and X-ray diffraction. Finally, the ionoluminescence of MgO under different irradiation conditions has also been studied.The results obtained in this thesis help to understand the irradiation-damage processes in materials, which is essential for the development of new nuclear energy sources.

No description available.

Radiation safety is the biggest concern of the nuclear industry, and co-robots are a crucial component to insuring that safety. Currently, radiation mapping data is typically gathered using hand held detectors or other detection systems requiring constant human interaction. This results in direct exposure to radiation of the individual performing the survey. Co-robots can coordinate computer algorithms and human input to determine the most efficient and accurate methods of surveying these same regions while eliminating health hazards. These surveying methods can then be adapted for multiple uses in the industry including nonproliferation, maintenance, and accident response scenarios. This work describes the process by which two vehicles were modified to detect radiation with minimal human interaction. An algorithm was developed to control the robot and to navigate the area of interest while ensuring that all sources are found. A compact detector system was used to keep the vehicles as small and light as possible. The vehicles were constructed to satisfy the requirements of the detector system and relay the necessary information back to the control station. The process, which is nearly fully autonomous, can map an area of interest and proceed to characterize the radiation materials that are found using neutron and gamma spectroscopy. The vehicles were tested in several scenarios which included obstacles, multiple sources, and shielding of the sources to determine the practicality of these co-robots. The evaluation of these co-robots was critical, as the future of radiation safety lies in the research and construction of small autonomous radiation detection systems to minimize the risk that radiation exposure poses to humans.

Este trabalho consiste em uma pesquisa experimental comparativa entre o zircônio puro e a liga comercial de aplicação nuclear Zircaloy-4, com ênfase nas características de encruamento, recristalização e textura cristalográfica. Foram utilizadas várias técnicas complementares de análise microestrutural tais como microscopia óptica, microscopia eletrônica de varredura com análise química de microrregiões por dispersão de energia de raios X característicos, difração de raios X, calorimetria exploratória diferencial, medidas de dureza e de condutividade elétrica. Para as determinações de macrotextura foi utilizado um goniômetro dedicado de raios X. No estado como recebido, enquanto o zircônio puro apresentava grãos recristalizados com diâmetro médio de aproximadamente 50μm, a liga apresentava granulação alfa em plaquetas grosseiras com diâmetro médio do pré-grão beta de aproximadamente 1,1mm. Experiências de laminação e a determinação de curvas de limite de redução sem a presença de trincas em função da temperatura mostraram que enquanto o zircônio puro apresentou níveis altos de plasticidade na temperatura ambiente, a liga Zircaloy-4 apresentou baixa ductilidade e muitas trincas. As ductilidades dos dois materiais, especialmente da liga Zircaloy-4, aumentaram significativamente a partir de 300°C. A 500°C as ductilidades de ambos são idênticas. Utilizando-se deformações e recozimentos diferenciados foram obtidas tiras de mesma espessura, com grãos equiaxiais e diâmetros médios de grão de aproximadamente 9µm para os dois materiais. Os estudos de recristalização revelaram que, enquanto para o zircônio puro a recuperação contribui significativamente para o amolecimento, no caso da liga Zircaloy-4, o amolecimento ocorre quase que exclusivamente por recristalização. As temperaturas de recristalização do zircônio puro foram mais baixas que as da liga. Os átomos de soluto em solução sólida foram responsáveis pelos dois efeitos concorrentes; aumento da energia armazenada na deformação e aumento da resistência à recristalização. Além da caracterização microestrutural mencionada, foram realizadas determinações de textura cristalográfica para os dois materiais em diferentes condições. Com relação às texturas de laminação do zircônio puro, para uma mesma temperatura, em cerca de 50% de redução a textura de laminado a frio {1 1 2 2} já estava plenamente formada e se alterou muito pouco a partir desta redução, até cerca de 90%. Com o aumento de temperatura de deformação para a mesma redução, a textura de laminado a frio se manteve estável até 300°C. A amostra de Zircaloy-4 preparada para possuir um tamanho de grão de 9 m tinha uma textura próxima de {0 0 0 2} , demonstrando que os tratamentos térmicos e mecânicos utilizados para obtenção dessa amostra foram eficientes na redução da textura de laminado a frio {1 1 2 2} . Recozimentos com duração de uma hora a 550 e 575°C, tanto em zircônio puro como na liga Zircaloy-4, foram suficientes para provocar recristalização estática. A 600°C, uma mudança na orientação cristalográfica foi verificada em Zircaloy-4, tendendo a {0 0 0 2} , enquanto em zircônio puro os planos basais continuam estáveis. O uso de funções de distribuição de orientação cristalográfica (FDOC) auxiliaram na detecção de um segundo grupo orientado, que tende à orientação {1 0 1 1} , além do grupo que reforça as fibras D0 e Rf . A mudança de textura ocorreu durante o crescimento de grão em ambos os materiais. De um modo geral, os resultados mostraram que o zircônio puro tende a ser mais suscetível à recristalização e ao crescimento de grão do que a liga Zircaloy-4. Entretanto, tanto zircônio como a liga são resistentes à modificação de textura, sendo que esta ocorreu principalmente com o crescimento de grão, em temperaturas após a completa recristalização primária. / This work shows a comparative experimental research between pure zirconium and the nuclear-grade zirconium alloy Zircaloy-4. This work emphasizes the characteristics of strain hardening, recrystallization, and crystallographic texture. Was used several complementary techniques for microstructural analysis such as optical microscopy, scanning electron microscopy with chemical analysis (EDS), X-ray diffraction, differential scanning calorimetry, indentation hardness and electrical conductivity. For measurements of macrotexture was used a dedicated X-ray goniometer. In the as received state, while pure zirconium showed grains recrystallized with an average diameter of about 50µm, the alloy had rough alpha plates with average diameter of beta pregrain of about 1,1mm. Rolling experiments and determination of reduction limit curves without cracks as a function of temperature showed that while zirconium pure showed high levels of plasticity at room temperature, the alloy zircaloy-4 showed low ductility and many cracks. The ductilities of the two materials, mainly zircaloy-4, significantly increased from 300°C. At 500°C, the ductilities were identical. Using different strains and annealing were obtained strips of equal thickness, with equiaxed grains and grain average diameters of about 9µm for both materials. Recrystallization studies revealed that recovery contributes significantly to softening of pure zirconium. In the case of the alloy zircaloy-4, the softening occurs almost exclusively by recrystallization. The temperature of recrystallization of the pure zirconium were lower than the alloy. The solute atoms in the solid solution were responsible for the two competing effects, the increase of the strain energy stored and the increasing of recrystallization resistance. Crystallographic texture measurements were made for both materials under different conditions. With respect to the rolling textures of pure zirconium, in about 50% reduction of the cold-rolled texture {1 1 2 2} was already fully formed and changed very little from this reduction to about 90%. With the increase of temperature strain to the same reduction, texture cold rolled remained stable up to 300°C. The sample of zircaloy-4 prepared to have a grain size of 9m had a texture close to {0 0 0 2} , demonstrating that the thermal and mechanical treatments used to obtain this sample were effective in reducing texture of cold-rolled {1 1 2 2} . One hour annealings at 550 and 575°C, in pure zirconium and Zircaloy-4, were suffcient to cause static recrystallization. At 600 °C a change in crystallographic orientation was seen in zircaloy-4, tends to {0 0 0 2} , while in pure zirconium the basal planes remains stable. The use of orientation distribution functions (ODF) aided in the detection of a second oriented group, which tends to orientation {1 0 1 1} , besides the group that reinforced D0 and Rf fibers. The change in texture occurred during the grain growth in both materials. In general, the results showed that pure zirconium tends to be more susceptible to recrystallization and grain growth than Zircaloy-4. Nevertheless, Both zinconium and Zircaloy-4 are resistant to texture changes. The texture changes occurred mainly in grain growth, at temperatures after complete recrystallization.

Este trabalho consiste em uma pesquisa sobre aços inoxidáveis ferríticomartensíticos de ativação reduzida (RAFM): EUROFER (9Cr-1W) e ODS-EUROFER (9Cr-1W-0,3Y2O3), envolvendo o encruamento e a recristalização destas duas ligas com o objetivo de estudar a influência de uma dispersão de partículas nanométricas na recristalização de aços inoxidáveis. O conceito de materiais de ativação reduzida é discutido e é apresentada a aplicação destes aços tanto na estrutura de diversor do ITER quanto na primeira parede no módulo de câmara fértil do reator DEMO. As placas, no estado revenido, foram laminadas a frio em um laminador de pequeno porte. As curvas de encruamento de ambos os materiais mostram um comportamento quase linear. Os tratamentos isócronos de uma hora, entre 300 e 750 °C, resultaram curvas de amolecimento que indicam uma forte resistência à recristalização da liga ODS-EUROFER, em concordância com os modelos teóricos. A liga EUROFER-97 apresentou recristalização muito similar a liga comercial 430, mas com maior dureza inicial, devido a maior quantidade de elementos intersticiais. / This work studies reduced activation ferritic-martensitic (RAFM) stainless steels: EUROFER (9Cr-1W) and ODS-EUROFER (9Cr-1W-0,3Y2O3), and their work hardening and recrystallization behaviour to better understand the influence of a dispersion of nanometric particles on the recrystallization process of stainless steels. The concept of reduced activation materials is discussed and the application of these steel alloys, such as in the divertor structure of ITER (International Thermonuclear Experimental Reactor) and as in the DEMO reactor breeding blanket first wall is shown. The plates, in the as-tempered condition, were cold rolled in a laboratory rolling mill. The work hardness curves of both materials presented an approximately linear behavior with strain increase. One hour isochronal treatments, in the temperature range from 300 to 750 °C, resulted in softening curves that indicated a strong resistance to recrystallization of the alloy ODS-EUROFER, supporting the theorical models. The EUROFER-97 recrystallization showed a similar behaviour to the commercial 430 alloy, however with higher initial hardness, due to the larger amount of interstitial elements.

Hydrogen embrittlement in ferrous and non-ferrous alloys is, and has been for over a century, a prominent issue within many sectors of industry. Despite this, the mechanisms by which hydrogen embrittlement occurs and the suitable means for its prevention are yet to be fully established. As hydrogen fuel becomes a prominent feature in modern concepts of a sustainable global energy infrastructure and nuclear power enters its renaissance, with commercially viable fusion plants on the horizon, hydrogen embrittlement is becoming an ever more pertinent issue. This has led to a considerable demand for novel alloys resistant to hydrogen embrittlement, notably within the bearings industry, where the commonly conflicting properties of high strength and hydrogen embrittlement resistance are required. This work investigates the mechanisms through which hydrogen embrittlement and irradiation damage occur in steels and nickel-based alloys respectively, with novel alloys designed for improved resistance. Through the engineering of secondary phases, optimised for helium and/or hydrogen trapping capacity, the novel alloys present the benefits of such trapping species with respect to embrittlement resistance. Such species have been studied in depth with respect to their interactions with hydrogen, establishing a novel mechanism of hydrogen embrittlement - the hydrogen enhanced dissolution and shearability of precipitates, leading to enhanced localised plasticity.