SiC is used as the main diffusion barrier in the fuel spheres of the pebble bed modular reactor (PBMR). The PBMR is a modern high temperature nuclear reactor. However, the release of silver from the fuel spheres has raised some doubts about the effectiveness of this barrier, which has led to many studies on the possible migration paths of silver. The reported results of these studies have shown largely differing results concerning the magnitude and temperature dependence of silver being transported through the fuel particle coatings. Results from earlier investigations could be interpreted as a diffusion process governed by an Arrhenius type temperature dependence. In this study, the silver diffusion in 6H-SiC was investigated using two methods. In the first method a thin silver layer was deposited on 6H-SiC by vapour deposition while in the second method silver was implanted in 6H-SiC at room temperature, 350°C and 600°C to a fluence of 2×1016 silver ions cm-2. Finally the effect of neutron irradiation on the diffusion of silver was investigated for the samples implanted at 350°C and 600°C. Silver depth profiles before and after annealing were determined by Rutherford backscattering (RBS). Both isothermal and isochronal annealing were used in this study. Diffusion coefficients as well as detection limits were extracted by comparing the silver depth profiles before and after annealing. The radiation damage after implantation and their recovery after isothermal and isochronal annealing were analysed by Rutherford backscattering spectroscopy combined with channelling. The results of in-diffusion of silver into 6H-SiC at temperatures below the melting point (960°C) using un-encapsulated 6H-SiC samples with 100 nm deposited silver indicated no in-diffusion of silver; however, disappearance of silver occurred at these temperatures. For the encapsulated samples, no in-diffusion of silver was observed at 800°C, 900°C and 1000°C but silver disappeared from the samples’ surface and was found on the walls of the quartz glass ampoule. This disappearance of silver was established to be due to the wetting problem that existed between silver and SiC. The room temperature implantation resulted in a completely amorphous surface layer of approximately 270 nm thick. Epitaxial re-growth from the bulk was already taking place during annealing at 700°C and the crystalline structure seemed to be fully recovered at 1600°C, for samples that were sequentially isochronally annealed from 700°C in steps of 100°C up to 1600°C. However, no silver signal was detected at this temperature, which left certain doubts regarding the crystalline structure of the samples at this temperature. This was speculated to be due to thermal etching of the top original amorphous layer while the deeper amorphous layer was epitaxial re-growth from the bulk. The decomposition of SiC, giving rise to a carbon peak in the RBS spectra due to evaporation of Si, was clearly observed on the same samples at 1600°C. Isothermal annealing at 1300°C for 10 h cycles up to 80h caused epixatial re-growth from the bulk during the first annealing cycle (10h). No further epitaxial re-growth from the bulk was observed up to 80h. This was believed to be due to the amorphous layer re-crystallising into crystals that were randomly oriented to the 6H-SiC substrate. No diffusion of silver was observed at temperatures below 1300°C but silver seemed to form precipitates at these temperatures. Diffusion of silver towards the surface accompanied by silver loss from the surface began at 1300°C and was very high at 1400°C, with silver profiles becoming asymmetric and closer to the surface. The loss of silver was already taking place at 1100°C. This loss was found to be due to the following: diffusion of silver towards the surface; the mass flow of silver via holes that were observed to be becoming larger with higher annealing temperatures on SiC surfaces and thermal etching of SiC. Isothermal annealing at 1300°C for 10h up to 80h caused diffusion of silver during the first annealing cycle, while no further diffusion was observed for any further annealing at the same temperature up to 80 h. The diffusion coefficient was not calculated due to the lack of information on the structural evolution of SiC during the first annealing cycle. Isothermal annealing at 1300°C and 1350°C for 30 minute cycles up to 120 minutes caused high diffusion during the first cycle and reduced diffusion during the second cycle, while no diffusion was observed for any further annealing longer than the second cycle. The higher diffusion during the first 30 minutes was due to ion induced amorphization. The diffusion of silver in amorphised SiC was measured at different temperatures in the range 1300°C to 1385°C and yielded to Do ~ 1.4 × 10-12 m2s-1 and Ea ~ 3.3 × 10-19 J. These values were found to be approximately the same as the values of silver diffusion in polycrystalline CVD-grown SiC found by our group which were due to grain boundary diffusion: Do ~ 4×10-12 m2 s-1 and Ea ~ 4×10-19 J. Implantation of silver at 600°C retained crystallinity although distortions occurred in the implanted region while implantation at 350°C also retained crystallinity but more distortions occurred as compared to silver implanted at 600°C. This was caused by the fact that at 600°C, the displaced atoms were more mobile because of their higher thermal energy than at 350°C. The higher thermal energy increased the probability of the displaced atoms combining with their original lattice sites. Annealing of these samples at 1300°C, 1350°C and 1500°C caused the annihilation of some defects but certain others were retained. No diffusion of silver was observed during annealing of the samples (implanted at 350°C and at 600°C) at 1300°C, 1350°C and 1500°C but silver moved towards the surface at 1500°C. The upper limit of the diffusion coefficient of D < 10-21 m2s-1 was obtained at 1300°C. The movement of silver towards the surface was found to be due to thermal etching at 1500°C. Neutron irradiation of these samples caused no silver diffusion but silver -110mAg, due to -109Ag capturing a neutron during neutron irradiation, was detected in the samples. / Thesis (PhD)--University of Pretoria, 2010. / Physics / unrestricted

One of two project reports: The other part is designated as Part A / <p> A fundamental and important part of nuclear reactor development and analysis today is the study of neutronics following a breach in the primary heat transport circuit. In the past, much of this analysis has concentrated on the calculation of the thermalhydraulic changes which occur following a loss of coolant accident and the effects these subsequently have on neutron kinetics. The purpose of this present study is to examine the influence of neutronic parameters on the size and shape of power pulses which result from loss of coolant accidents. The parameters studied are shutdown system delay times, shutoff rod drop curves, and fuel burnup distribution. </p> / Thesis / Master of Engineering (MEngr)

This work investigated material corrosion by nuclear reactor coolants, including pressurized water reactor (PWR) coolant, boiling water reactor (BWR) coolant, high-temperature steam, lead-bismuth eutectic (LBE), and molten salt. Novel cladding materials for accident tolerant fuel (ATF) and additive manufacture (AM) Ni-based alloy were studied in water coolants. Similarly, the ATF material and Ni-based alloys were also examined under high-temperature steam to understand the corrosion behavior in beyond design basis accident (BDBA) scenarios. In addition to isothermal corrosion, stress corrosion cracking (SCC) and oxide layer in situ measurements were also conducted. Unlike conventional studies in liquid LBE that focused on Fe-based alloys, the present studies also investigated Ni-based alloys to explore the Ni content effects on the corrosion by LBE at high temperatures under saturated oxygen conditions. In molten salt environments, the corrosion behaviors of both Ni-based and Fe-based alloys were investigated. This study developed a redox potential range for mitigating corrosion by using a redox couple of UF4 /UF3 and a novel approach of potential measurements against F2/ F- potential experimentally. / Doctor of Philosophy / This work focuses on material degradation in harsh and extreme nuclear environments, including light water reactors and advanced reactors such as molten salt and liquid metal coolant reactors. Given the renewed interest in advanced nuclear reactors as a resource of clean energy, advanced material development, including structural, fuel, and coolant materials, has become a significant and trending research area. Based on our past experiences, we have seen the detrimental effects of material failure due to corrosion. Systems are inherently safe in the absence of material degradation. Nevertheless, this is an idealistic thought, as corrosion is inevitable. Therefore, this research focuses on corrosion mitigation, as absolute material preservation is impossible. This work includes corrosion studies in aqueous environments in light water reactors and advanced nuclear reactors under normal operation and extreme conditions such as accident environments. Much of this work provides insights into material corrosion behavior and mitigation that helps nuclear reactor operators ensure safe operations. Commercially available alloys such as SS316, Hast. X and Hast. N were examined in primary water, molten salt, high-temperature steam, and liquid bismuth environment and model Fe-Cr-Si alloys were investigated in water and steam to compare the corrosion mechanisms.

Thesis: Sc. D., Massachusetts Institute of Technology, Department of Nuclear Engineering, 1977 / Vita. / Includes bibliographical references. / Sc. D. / Sc. D. Massachusetts Institute of Technology, Department of Nuclear Engineering

An accidental criticality may occur in a solution of fissile material. Since the processing of nuclear materials in solution is prevalent throughout the fuel cycle, it would be judicious to have the capability to predict a possible hazard. In view of this concern, a computer simulation was performed of the Los Alamos accident of December 30, 1958, in which the actuation of an electric stirrer produced a sudden criticality. A complete equation of state for a liquid containing gas bubbles was coupled with the equations of energy, momentum, and space-independent point kinetics. Multiplication calculations, implemented with the Monte Carlo Code for Neutron and Photon Transport (MCNP), were performed on thermally expanding solution geometries, to generate a reactivity feedback representation. With the knowledge of the total energy produced in the accident, the maximum reciprocal period on which the power rose was computed.

The development of magnox fuel cans at the Heston Laboratories of A.P.C. Ltd. required a non-contact method of measuring fuel can temperature during heat transfer tests. Infra-red techniques were examined and a two colour ratio pyrometer was selected as offering the greatest prospect of success. A design specification was compiled for an instrument having a lower temperature limit of 250°C with a target area of approximately 1mm². During heat transfer tests a pressurised rig is used and an observation window is necessary. This leads to some deterioration in performance which it is impracticable to remove by internal chopping and the pyrometer is designed for outside use. Operational wavebands around 1.8 and 2.3 microns are chosen and a lead sulphide infra-red detector operating at room temperature is selected. The properties of the detector lead to a design where the signal is chopped at 900 cps and modulated at 30 cps by the energy in each waveband. Conventional optical systems are considered and rejected in view of the large size of chopper wheel necessary. An alternative system is devised in which a commercially available pencil galvanometer is used to combine the function of chopping at both 900 cps and 30 cps. This enables an optical system to be designed which, because it employs reflecting optics, is compact and for the most part axially symmetric. The signal ration is extracted using an automatic gain control technique and a high signal to noise ratio is achieved with a phase sensitive rectifier at the output. The pulses which activate the galvanometer chopper and phase sensitive circuits are derived from a master oscillator operating at 900 cps. The performance of the system is assessed and it meets the specification in all respects except for a small surface geometry effect. The underlying cause for this is examined and a means for overcoming it is suggested. The performance when scanning a magnox can during bench tests is illustrated and it can be seen that although agreement to within 2°C is obtained with thermocouples placed at the fin roots, the geometry effect prevents the isolation of fin tip temperature. During the development of the instrument the possibility of wider application was an important consideration. Some of the applications which have arisen are discussed briefly and an indication of the performance of the pyrometer is given. Furthermore the system is capable of easy modification for use with a cooled detector and other regions of the infra-red spectrum.

M.Phil. (Energy Studies) / Nodal diffusion methods are often used to calculate the distribution of neutrons in a nuclear reactor core. They require few-group homogenized neutron cross sections for every heterogeneous sub-region of the core. The homogenized cross sections are pre-calculated at various reactor states and represented in a way that facilitates the reconstruction of cross sections at other possible states. In this study a number of such representations were built for the homogenized cross sections of a MOX (mixed oxide) fuel assembly via hierarchical Lagrange interpolation on Clenshaw-Curtis sparse grids. These cross sections were represented as a function of various thermal hydraulic and material composition parameters of a pressurized water reactor core (i.e. burnup, soluble boron concentration, fuel temperature, moderator temperature and moderator density), which are generally referred to as state parameters. Representations were produced for the homogenized cross sections of a number of individual isotopes, as well as the e ective (lumped) cross section of all the materials in the assembly. This was done for both two and six energy groups. Additionally, two sets of state parameter intervals were considered for each of the group structures. The first set of intervals was chosen to correspond to conditions that may be encountered during day-to-day reactor operations. The second set of intervals was chosen to be applicable to the simulation of accident scenarios and therefore have wider ranges for fuel temperature, moderator temperature and moderator density.

Este trabalho tem como objetivo a avaliação dosimétrica de componentes semicondutores (detectores Barreira de Superfície e fotodiodos PIN) para aplicação em medições de dose equivalente em campos de baixo fluxo de nêutrons (rápidos e térmicos), utilizando uma fonte de AmBe de alto fluxo, a instalação de Neutrongrafia do reator IEA-R1 (fluxos térmicos/epitérmicos) e fluxo de nêutrons rápidos do núcleo do reator IPEN/MB-01 (UCRI Unidade Crítica). Para a detecção de nêutrons (térmicos, epitérmicos e rápidos) foram usados componentes moderadores e conversores (parafina, boro e polietileno). Os fluxos resultantes da moderação e conversão foram utilizados para a irradiação de componentes semicondutores (SSB - Barreira de Superfície e fotodiodos). Foi utilizado também um conversor misto constituído de uma folha de polietileno borado (marca Kodak). O método de simulação por Monte Carlo foi utilizado para avaliar de forma analítica a espessura ótima da parafina. O resultado obtido foi similar ao verificado experimentalmente e serviu para avaliar o fluxo de nêutrons emergentes do moderador (parafina). Da mesma forma, através de simulação, foi avaliado também o fluxo de nêutrons rápidos que atinge o conversor de polietileno que cobre a face sensível dos semicondutores. O nível de radiação gama foi avaliado cobrindo o detector por inteiro com uma folha de cádmio de 1 mm de espessura. O reator IPEN/MB-01 foi usado para avaliar a resposta dos detectores para nêutrons rápidos de alto fluxo. Os resultados, de uma forma geral, mostraram concordância e similaridade com os trabalhos desenvolvidos por outros grupos de pesquisas. Foi também estabelecida uma abordagem para o cálculo de dose equivalente utilizando os espectros obtidos nas experiências. / The main objetive of this research is the dosimetric evaluation of semiconductor componentes (surface barrier detectors and PIN photodiodes) for applications in dose equivalent measurements on low dose fields (fast and thermal fluxes) using an AmBe neutron source, the IEA-R1 reactor neutrongraphy facility (epithermal and thermal fluxes) and the Critical Unit facility IPEN/MB-01 (fast fluxes). As moderator compound to fast neutrons flux from the AmBe source was used paraffin and boron and polyethylene as converter for thermal and fast neutrons measurements. The resulting fluxes were used to the irradiation of semiconductor components (SSB Surface Barrier Detector and PIN photodiodes). A mixed converter made of a borated polyethylene foil (Kodak) was also used. Monte Carlo simulation metodology was employed to evaluate analytically the optimal paraffin thickness. The obtained results were similar to the experimental data and allowed the evaluation of emerging neutron flux from moderator, as well as the fast neutron flux reaching the polyethylene covering the semiconductor sensitive surface. Gamma radiation levels were evaluated covering the whole detector with cadmium foil 1 mm thick, allowing thermal neutrons blockage and gamma radiation measurements. The IPEN/MB-01 facility was employed to evaluate the detector response for high neutron flux. The results were in good agreement with other studies published. Using the obtained spectra an approach to dose equivalent calculation was established.

Após os acidentes nucleares ocorridos no mundo, critérios e requisitos extremamente rígidos para a operação das instalações nucleares foram determinados pelos órgãos internacionais que regulam essas instalações. A partir da ocorrência destes eventos, as operadoras de plantas nucleares necessitam simular alguns acidentes e transientes, por meio de programas computacionais específicos, para obter a licença de operação de uma planta nuclear. Com base neste cenário, algumas ferramentas computacionais sofisticadas têm sido utilizadas como o Reactor Excursion and Leak Analysis Program (RELAP5), que é o código mais utilizado para a análise de acidentes e transientes termo-hidráulicos em reatores nucleares no Brasil e no mundo. Uma das maiores dificuldades na simulação usando o código RELAP5 é a quantidade de informações geométricas da planta necessárias para a análise de acidentes e transientes termo-hidráulicos. Para a preparação de seus dados de entrada é necessário um grande número de operações matemáticas para calcular a geometria dos componentes. Assim, a fim de realizar estes cálculos e preparar dados de entrada para o RELAP5, um pré-processador matemático amigável foi desenvolvido, neste trabalho. O Visual Basic for Applications (VBA), combinado com o Microsoft Excel, foi utilizado e demonstrou ser um instrumento eficiente para executar uma série de tarefas no desenvolvimento desse pré-processador. A fim de atender as necessidades dos usuários do RELAP5, foi desenvolvido o Programa de Cálculo do RELAP5 PCRELAP5 onde foram codificados todos os componentes que constituem o código, neste caso, todos os cartões de entrada inclusive os opcionais de cada um deles foram programados. Adicionalmente, uma versão em inglês foi criada para PCRELAP5. Também um design amigável do PCRELAP5 foi desenvolvido com a finalidade de minimizar o tempo de preparação dos dados de entrada e diminuir os erros cometidos pelos usuários do código RELAP5. Nesse trabalho, a versão final desse pré-processador foi aplicada com sucesso para o Sistema de Injeção de Emergência (SIE) da usina Angra 2. / Nuclear accidents in the world led to the establishment of rigorous criteria and requirements for nuclear power plant operations by the international regulatory bodies. By using specific computer programs, simulations of various accidents and transients likely to occur at any nuclear power plant are required for certifying and licensing a nuclear power plant. Based on this scenario, some sophisticated computational tools have been used such as the Reactor Excursion and Leak Analysis Program (RELAP5), which is the most widely used code for the thermo-hydraulic analysis of accidents and transients in nuclear reactors in Brazil and worldwide. A major difficulty in the simulation by using RELAP5 code is the amount of information required for the simulation of thermal-hydraulic accidents or transients. The preparation of the input data requires a great number of mathematical operations to calculate the geometry of the components. Thus, for those calculations performance and preparation of RELAP5 input data, a friendly mathematical preprocessor was designed. The Visual Basic for Application (VBA) for Microsoft Excel demonstrated to be an effective tool to perform a number of tasks in the development of the program. In order to meet the needs of RELAP5 users, the RELAP5 Calculation Program (Programa de Cálculo do RELAP5 PCRELAP5) was designed. The components of the code were codified; all entry cards including the optional cards of each one have been programmed. In addition, an English version for PCRELAP5 was provided. Furthermore, a friendly design was developed in order to minimize the time of preparation of input data and errors committed by users. In this work, the final version of this preprocessor was successfully applied for Safety Injection System (SIS) of Angra 2.

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. / In nuclear facilities, the nuclear material control is one of the most important activities. The National Commission of Nuclear Energy (CNEN) and the International Atomic Energy Agency (IAEA), when inspecting routinely, regards the data provided as a major safety factor. Having a control system of nuclear material that allows the amount and location of the various items to be inspected, at any time, is a key factor today. The objective of this work was to enhance the existing system using a more friendly platform of development, through the VisualBasic programming language (Microsoft Corporation), to facilitate the operation team of the reactor IEA-R1 Reactor tasks, providing data that enable a better and prompter control of the IEAR1 nuclear material. These data have allowed the development of papers presented at national and international conferences and the development of master´s dissertations and doctorate theses. The software object of this study was designed to meet the requirements of the CNEN and the IAEA safeguard rules, but its functions may be expanded in accordance with future needs. The program developed can be used in other reactors to be built in the country, since it is very practical and allows an effective control of the nuclear material in the facilities.