This thesis presents a whole-core benchmark of the European Pressurized Reactor (EPR) using multiple transport methods. The core specifications were taken directly from the Final Safety Analysis Report (FSAR) submitted to the Nuclear Regulatory Commission (NRC) and the reactor was modeled in a stylized manner while maintaining full heterogeneity at the pin and assembly level. The geometry and material specifications are given as well as problem-specific cross sections for 2, 4, and 8 energy group calculations. Cross sections were generated using HELIOS, a lattice depletion code based on the Collision Probability Method (CPM). The multi-group cross sections were utilized in the reference calculation, COMET calculation, and response function generation. The reference solution was obtained via an MCNP model identical to the one implemented in COMET. Specific steps towards constructing and running a COMET calculation are outlined. Detailed results including assembly eigenvalues, core eigenvalues, and pin fission densities are presented.

After a radiation dispersion device (RDD) event, there may be internally and/or externally contaminated victims. After the RDD event, victims may require immediate medical assistance prior to decontamination. The dose rates to which a healthcare provider is exposed due to the internal and external contamination of the victim were computed using Monte Carlo simulations and five anthropomorphic phantoms. The dose rates to which the victim is exposed due to his/her own external contamination were also computed. For the external contamination modeling, the contamination is assumed to be distributed over the entire exterior of the victimâ s body. The geometrical models of the human body were based on the MIRD stylized phantom. The specific isotopes considered were 60Co, 137Cs, 131I, 192Ir, and 241Am. The surface contamination was generated by creating a 2-mm thick layer adjacent to the outside of the skin of the victim and uniformly sampling the emissions of the radioactive sources throughout this volume. The attending healthcare provider was assumed to be standing 20 cm from mid-torso of the victim. The organ absorbed doses in both the contaminated individual and a healthcare professional were computed. The effective dose to the victim and the attending healthcare professional were computed using the tissue weighting factors in ICRP Publication 60. For example, the dose rate to a reference male healthcare provider from the victim six hours after the inhalation of one ALI by an adipose male victim will be 0.277 mSv/hr. In addition, the air kerma was computed at different distances from the surfaces of the victim phantom and ratios were generated for the air kerma and the effective dose due to the victim from the surface contamination on the victim.

Submitted by Almir Azevedo (barbio1313@gmail.com) on 2016-05-13T13:22:37Z No. of bitstreams: 0 / Made available in DSpace on 2016-05-13T13:22:37Z (GMT). No. of bitstreams: 0 Previous issue date: 2016-04 / Este trabalho apresenta uma metodologia baseada nos princípios de atenuação de raios gama para a identificação de frações de volume em sistemas bifásicos compostos por óleo-gás e água-gás que são encontrados na indústria petrolífera offshore e onshore. Esta metodologia baseia-se no reconhecimento de contagens por segundo no fotopico da fonte de radiação, utilizando um sistema de detecção composto por um detector de Nal(TI), uma fonte de Cs137 sem colimação posicionada a 180º com relação ao detector em um regime de fluxo estratificado liso. A modelagem matemática para a simulação computacional utilizando o código Monte Carlo N-Particle eXtended (MCNP-X) foi realizada utilizando as medições experimentais das características do detector (resolução energética e eficiência), das características dos materiais água e óleo (densidade e coeficiente de atenuação) e a medição das frações de volume. Para a predição destas frações foram utilizadas redes neurais artificiais (RNAs) e para se obter um treinamento adequado das RNAs para a predição das frações de volume foram simuladas no código MCNP-X um maior número de frações de volume. Dados experimentais foram utilizados no conjunto de padrões necessários para a validação das RNAs e os dados gerados por meio do código computacional MCNP-X foram utilizados nos conjuntos de treinamento e teste das RNAs. Foram utilizadas RNAs do tipo feed-forward multilayer perceptron (MLP) e analisadas duas funções de treinamento, Levenberg-Marquadt (LM) e gradiente descendente com momento (GDM), ambas utilizando o algoritmo de treinamento Backpropagation. As RNAs identificaram corretamente as frações de volume no sistema multifásico, com erros relativos médios inferiores a 1,21%, possibilitando a aplicação desta metodologia para tal propósito

Submitted by Marcele Costal de Castro (costalcastro@gmail.com) on 2017-09-26T18:55:34Z No. of bitstreams: 1 CESAR MARQUES SALGADO M.pdf: 3473675 bytes, checksum: 2b5a708871e86c34a4f43ffa3852b0df (MD5) / Made available in DSpace on 2017-09-26T18:55:34Z (GMT). No. of bitstreams: 1 CESAR MARQUES SALGADO M.pdf: 3473675 bytes, checksum: 2b5a708871e86c34a4f43ffa3852b0df (MD5) Previous issue date: 2003-08 / Este trabalho visa estudar possíveis procedimentos para a determinação de espectros de fótons, gerado por um tubo raios X, utilizados em diagnóstico médico (RXD) que opera na faixa de m20 a 150 Kv, permitindo, assim, o estabelecimento mais preciso das qualidades dos feixes de RXD, contribuindo para diminuir as incertezas nos processos de calibração de câmaras de ionização. Com esta finalidade, foram selecionados dois tipos de detectores, um detector de telureto de cádmio e zinco (CZT) e outro de germânio (HPGe planar). A interação do feixe de raios X com esses detectores fornece uma distribuição de altura de pulsos (DAP) que não representa o espectro verdadeiro de fótons incidentes, devido à presença de fótons escapes K, espelhamento Compton e ao fato da eficiência de detecção diminuir abruptamente com o aumento da energia dos fótons. Uma análise detalhada destes efeitos espúrios envolvidos na detecção foi realizada com a utilização do código MCNP 4B (código computacional para transporte de nêutrons e fótons) na modelagem dos detectores. Um procedimento de desmembramento (stripping) é descrito para a implementação em um computador pessoal para o detector HPGe, obtendo assim, o real espectro de energia de fótons do aparelho de raios X. As curvas de resposta do detector, obtidos por modelagem foram comparadas com dados obtidos experimentalmente, utilizando-se fontes pontuais. A validade deste método é testada por comparação com os espectros teóricos para as condições do tubo de raios X e, também, comparando-se os valores de kerma no ar determinados para este detector e medidos por uma câmara de ionização padrão secundário.

Submitted by Almir Azevedo (barbio1313@gmail.com) on 2018-04-27T12:15:53Z No. of bitstreams: 1 dissertação mestrado ien 2018 Roos Sophia de Freitas Dam.pdf: 1402693 bytes, checksum: 6a8bc7c98b0b3e3e3d55619f51dd2f72 (MD5) / Made available in DSpace on 2018-04-27T12:15:53Z (GMT). No. of bitstreams: 1 dissertação mestrado ien 2018 Roos Sophia de Freitas Dam.pdf: 1402693 bytes, checksum: 6a8bc7c98b0b3e3e3d55619f51dd2f72 (MD5) Previous issue date: 2018-02 / Agitadores ou misturadores são amplamente utilizados nas indústrias química, farmacêutica e de cosméticos quando processos como dispersão e homogeneização são desejados. Estes equipamentos são utilizados para misturar líquidos, promover reações de substâncias químicas, manter líquidos homogêneos durante armazenamento Agitadores industriais são construídos com características específicas para cada aplicação, dependendo de parâmetros como densidade, fase e viscosidade dos produtos a serem agitados. Durante a produção de um produto, o equipamento pode falhar e comprometer o procedimento de agitação ou mistura, tornando-se necessário avaliar o desempenho do misturador. Desta forma, é muito importante ter uma ferramenta de diagnóstico e de desempenho para unidades industriais visando garantir a qualidade do produto. O método utilizado neste trabalho baseia-se nos princípios da técnica de Rastreamento de Partícula Radioativa, que correlaciona as contagens obtidas por um arranjo de detectores com a posição instantânea ocupada por uma partícula radioativa. A geometria de detecção desenvolvida utiliza oito detectores cintiladores de NaI(Tl), uma fonte pontual de 137Cs (662 keV) com emissão isotrópica de raios gama e um tubo de policloreto de vinila como seção de teste. O modelo matemático foi desenvolvido utilizando o código MCNP-X, onde inicialmente o tubo é preenchido com ar e a partícula radioativa é posicionada em seu interior. Em um segundo momento, o tubo é preenchido com uma mistura de concreto. Nas duas situações, o algoritmo de localização utilizado pela rede foi capaz de predizer a posição instantânea da partícula radioativa. / Agitators or mixers are highly used in the chemical, pharmaceutical and cosmetic industries when processes such as dispersion and homogenization are desired. This equipment is used to mix liquids, promote reactions of chemical substances, keep homogeneous liquid bulk during storage. Agitators and mixers are designed for each application with specific configurations, depending on the characteristics, such as density, phase and viscosity of the agitated product. During the production process, the equipment may fail and compromise the stirring or mixing procedure, thus it is very important to have a diagnosis tool for these industrial units to assure the quality of the product. The method here presented is based on the principles of the radioactive particle tracking technique, which correlates the counts obtained by an array of detectors with the instantaneous position of the radioactive particle. The detection geometry developed in this work employs eight NaI(Tl) scintillation detectors, a 137Cs (662 keV) point source with isotropic emission of gamma-rays and a polyvinyl chloride tube as a test section. The mathematical model was developed using the MCNP-X code, where the tube is first filled with air and the radioactive particle is positioned inside it. Then, the tube is filled with a concrete mixture. In both situations, the search algorithm given by the network was capable to predict the instantaneous position of the radioactive particle.

Magnesium is an essential element. An adult body contains approximately 21-28 grams of magnesium, with 50-60% present in the bones. Too high or too low levels of magnesium intake can have harmful effects on human body. To study how magnesium intake and storage in the body affect human health, it is important to identify a credible biomarker for the intake and storage. Usually, the amount of magnesium in the body is determined by a blood draw, but blood contains less than 1 percent of the total amount of magnesium in the body. In addition, the concentration of magnesium in blood is not stable. Bone holds the majority of magnesium in the body; therefore, bone is expected to be an ideal biomarker for measuring any surplus or deficiencies in the body. This thesis investigates the feasibility of quantifying magnesium in hand bone <i>in vivo</i> using MCNP simulation models and experiments with magnesium doped phantoms. The fast neutrons, generated by a deuterium-deuterium neutron generator with a flux of 1e9 neutrons/second, were moderated and guided to produce maximum number of thermal neutrons in an irradiation cave with acceptable radiation dose to the hand. The dimensions of the neutron generator along with the current shielding techniques were simulated in MCNP. The data show that the differences between the experimental and simulated calibration lines resulted in a percent difference of 9.40%. The experimental detection limit for bone magnesium was found to be 334 µg magnesium/g dry bone with a total body dose of 11 µSv.

Master’s thesis focuses on the history and current progress in research of molten salt reactors around the world, with an emphasis placed on the properties of molten salts and the problems associated with their use. In relation to the practical part, one chapter is devoted to the creation of input file in the MCNP software. The practical part deals with neutron activation analysis of graphite prism experiment, which is filled with powder NaCl salt. This experiment is focused on the effect of salt on neutron flux with different energy levels. The whole problem was also simulated in the MCNP environment along with the experiment. At the end of the thesis, the individual methods are compared and evaluated.

Radiation is defined as ionizing if it has enough energy to remove electrons from atoms or molecules when it passes through or collides with matter. This ability implies potentially detrimental effects on living tissue. Ionizing radiation shielding is therefore a discipline of great practical importance. The thesis builds upon the author's previous work on the topic and widens the scope of discussion with theoretical and practical issues of advanced shielding calculations. The theoretical part of the thesis describes several approaches to calculating fluence or absorbed dose at an arbitrary point in space. Point-kernel methods provide sufficiently accurate results for simpler shielding problems. In many practical cases, however, calculations based on the transport theory are necessary. There are two basic types of transport calculations: deterministic transport calculations in which the linear Boltzmann equation is solved numerically, and Monte Carlo calculations in which a simulation is made of how particles migrate stochastically through the problem geometry. Advantages and disadvantages of both methods are discussed. In the practical part are the results of radiation shielding calculations performed with a major Monte Carlo code - MCNP6, compared with those obtained in the experiments, which were carried out at the Ionizing Radiation Laboratory at Department of Electrical Power Engeneering, FEEC BUT. The experiments consisted of placing a cobalt-60 radioisotope source at three different positions inside a lead collimator, and counting pulses with two different scintillation detectors positioned in front of the opening of the collimator, alternately with or without lead shield located between the source and the used detector. Agreement of the calculations and the data from the measurements is reasonable, given the inherent uncertainties of the experimental set-up. Performed sensitivity analysis shows relative importances of different parameters used as inputs in simulations, such as densities of materials, or dimensions of the scintillation crystals. Annotated MCNP input files used for simulation are also part of the thesis.

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