Submitted by (ana.araujo@ufrpe.br) on 2016-07-05T19:39:13Z No. of bitstreams: 1 Frank Sinatra Gomes da Silva.pdf: 1131089 bytes, checksum: 2c4bf5cf9af313b266e2630e4726c0c9 (MD5) / Made available in DSpace on 2016-07-05T19:39:13Z (GMT). No. of bitstreams: 1 Frank Sinatra Gomes da Silva.pdf: 1131089 bytes, checksum: 2c4bf5cf9af313b266e2630e4726c0c9 (MD5) Previous issue date: 2008-02-25 / The Monte Carlo method is an important tool to simulate radioactive particles interaction with biologic medium. The principal advantage of the method when compared with deterministic methods is the ability to simulate a complex geometry. Several computational codes use the Monte Carlo method to simulate the particles transport and they have the capacity to simulate energy deposition in models of organs and/or tissues, as well models of cells of human body. Thus, the calculation of the absorbed dose to thyroid’s follicles (compound of colloid and follicles’ cells) have a fundamental importance to dosimetry, because these cells are radiosensitive due to ionizing radiation exposition, in particular, exposition due to radioisotopes of iodine, because a great amount of radioiodine may be released into the environment in case of a nuclear accidents. In this case, the goal of this work was use the code of particles transport MNCP4C to calculate absorbed doses in models of thyroid’s follicles, for Auger electrons, internal conversion electrons and beta particles, by iodine-131 and short-lived iodines (131, 132, 133, 134 e 135), with diameters varying from 30 to 500 μm. The results obtained from simulation with the MCNP4C code shown an average percentage of the 25% of total absorbed dose by colloid to iodine- 131 and 75% to short-lived iodine’s. For follicular cells, this percentage was of 13% toiodine-131 and 87% to short-lived iodine’s. The contributions from particles with low energies, like Auger and internal conversion electrons should not be neglected, to assessment the absorbed dose in cellular level. Agglomerative hierarchical clustering was used to compare doses obtained by codes MCNP4C, EPOTRAN, EGS4 and by deterministic methods. / O método Monte Carlo é uma poderosa ferramenta para simular a interação de partículas radioativas com a matéria biológica. A principal vantagem do método, quando comparado com métodos determinísticos, tem sido a habilidade de adequarse de forma precisa a qualquer geometria complexa. Vários códigos computacionais simulam o transporte de partículas via método Monte Carlo, com capacidade para simular o depósito de energia em modelos geométricos que vão desde órgãos e/ou tecidos do corpo, como em modelos de células pertencentes a órgãos do corpo humano. Nesse sentido, o cálculo da dose absorvida pelos folículos tiroideanos (composto de colóide e células foliculares) tem sido de fundamental importância na dosimetria, uma vez que essas células são bastante radiosensíveis à exposição pela radiação ionizante, em particular exposição essa devido aos radioisótopos de iodo, que são resultados de produtos de fissão em casos de acidentes nucleares. Dessa forma, o objetivo desse trabalho foi o de utilizar o código para transporte de partículas MCNP4C para calcular doses absorvidas em modelos de folículos tiroideanos, devido aos elétrons Auger, elétrons de conversão interna e partículas beta, do iodo-131 e dos isótopos de meia-vida curta (iodos 132, 133, 134 e 135),para folículos com diâmetros que variaram de 30 até 500 μm. Os resultados obtidos pela simulação com o MCNP4C apresentaram um percentual médio de 25% da dose total absorvida pelo colóide para o iodo-131 e de 75% para os iodos de meia-vida curta. Para as células foliculares, esse percentual foi em média de 13% para o iodo- 131 e de 87% para os iodos de meia-vida curta, ressaltando assim a importância de simular partículas de baixa energia, como os elétrons Auger e elétrons de conversão interna, para a avaliação da dose absorvida a nível celular. Técnicas hierárquicas de análise de agrupamento foram usadas para comparações entre doses obtidas pelos códigos MCNP4C, EPOTRAN, EGS4 e doses calculadas por métodos determinísticos.

Les études de scénarios d'un parc électronucléaire, ainsi que les études de sûreté, sont essentielles pour explorer les différentes stratégies du nucléaire du futur. Pour mener à bien ces études, il est nécessaire d'estimer le temps d'irradiation d'un combustible donné, ainsi que sa composition isotopique pendant la campagne de production d'électricité. Ces estimations reposent sur des simulations de réacteurs nucléaires, dont les calculs d'évolution doivent être les plus représentatifs possible. Les schémas de calcul classiquement utilisés s'effectuent en deux étapes : un calcul cellule pour résoudre l'équation du transport des neutrons (de type déterministe ou Monte Carlo) suivi d'un calcul cœur (déterministe). Le calcul cellule est une simulation d'évolution d'un assemblage dans des conditions infinies. À partir de ce calcul, des sections efficaces homogénéisées et condensées, ainsi que des grandeurs de diffusion, sont calculées comme données d'entrée pour l'étape suivante, celle du calcul cœur. Le calcul cellule est donc une étape fondamentale et celui-ci doit être le plus représentatif possible d'un assemblage du cœur. Or, les approximations à la base de ce modèle sont nombreuses, plus particulièrement les fuites neutroniques sont négligées. L'objectif de ce travail est d'étudier les effets physiques de fuites neutroniques et de quantifier les biais associés par rapport à une simulation infinie. Dans une première partie, la problématique des fuites neutroniques axiales est étudiée. Dans ce cas, les fuites de neutrons provoquent une variation forte du spectre neutronique localisée dans les derniers centimètres de l'assemblage ainsi qu'une variation plus faible mais globale sur l'ensemble de l'assemblage. Une deuxième partie est dédiée aux fuites radiales de neutrons. L'effet des assemblages voisins, ainsi que le comportement particulier des assemblages en position périphérique sont étudiés et les biais de composition en fin d'irradiation sont quantifiés. Un calcul d'évolution d'un réacteur très simplifié permet de visualiser, dans une dernière partie, l'ensemble des effets physiques observés et qui impactent l'évolution de l'irradiation. De nombreuses approximations du calcul cellule restent à explorer, comme le suivi de réactivité, par l'intermédiaire de la concentration du poison de neutrons thermiques solubilisé dans le modérateur ou présent dans le combustible. Cependant, la détermination des phénomènes physiques à prendre en compte pour le calcul cellule représente une première étape indispensable vers une amélioration de la représentativité du calcul cellule, voire conduire à des nouvelles méthodes de simulation d'un cœur du réacteur. À terme, les quantifications des biais liés aux fuites neutroniques serviront à estimer l'incertitude sur les compositions isotopiques du combustible en fin d'irradiation. Ces incertitudes, propagées dans les études de scénarios, permettront de quantifier le degré de validité des résultats obtenus. / Scenario studies of an electronuclear fleet, as well as safety studies, are essential to explore the different nuclear strategies of the future. To carry out these studies, it is necessary to estimate the irradiation time of a given fuel and its composition during the electricity production campaign. These estimates are based on the simulations of nuclear reactors, for which the calculations of the evolution must be as representative as possible. The calculation schemes usually used are divided into two stages: a cell calculation to solve the neutron transport equation (deterministic or Monte Carlo simulation) followed by a core calculation (deterministic code) The cell calculation is a simulation of the evolution of an assembly under infinite conditions. Based upon this calculation, homogenized and condensed cross-sections along with scattering quantities are calculated as input data for the next stage, the core calculation. The cell calculation is therefore a fundamental step and must be representative of a core assembly evolution as much as possible. However, the approximations used for this model are numerous, especially the neutron leakages are neglected. The objectives of this work is to study the physical effects of neutron leakage and to compute the associated biases compared to an infinite assembly simulation. In the first part, the problem of axial neutron leakage will be broached. In this case, neutron leakage causes a strong variation of the neutron spectrum in the last centimeters of the assembly as well as a smaller variation but over the entire assembly. The second part deals with the radial leakage. The effect of the neighboring assemblies and the particular behavior of the assemblies in the peripheral position are studied. Moreover, the isotopic composition biases at the end of the cycle are quantified. In the third and last part, a simplified calculation of the evolution of a reactor enables to visualize all the observed physical effects impacting the evolution of the irradiation. Several other approximations of the cell calculation are still to be investigated, such as the reactivity monitoring through the concentration of thermal neutron poison dissolved in the moderator or present in the fuel. Nonetheless, establishing of the physical phenomena taken into account for the cell calculation represents an essential first step towards an improvement of the cell calculation and may lead to new simulation methods for reactor cores. In the future, the quantification of the biases related to neutron leakage will be used to estimate the uncertainties on the isotopic composition of the fuel at the end of the cycle. These uncertainties, propagated into the scenarios studies, will assess the validity of the obtained results.

X-ray fluorescence is a well-understood phenomenon in which irradiation of certain materials, such as gold, with x-rays causes the emission of secondary x-rays with characteristic energies. By performing computed tomography using these fluorescence x-rays, the material of interest can be imaged inside an object. Our research group has already demonstrated that x-ray fluorescence computed tomography (XFCT) imaging using a typical 110 kVp microfocus x-ray tube is feasible for a small animal-sized object containing a distribution of a solution of low concentration gold nanoparticles. The primary goal of this thesis work was to study the effect of source x-ray energy spectra on gold fluorescence detection using the XFCT system. A computational approach using the Monte Carlo method was used. First, a computational model was created using the Monte Carlo N-Particle (MCNP) transport code based on the experimental setup of the pre-existing XFCT system. Simulations were run to verify the validity of the MCNP model as an accurate representation of the actual system by means of comparison with experimentally-obtained data. Finally, the model was used for further purely computational work. In the MCNP model, the source spectrum was changed to reflect several theoretical and experimentally obtained options. The effect of these changes on gold fluorescence production was documented and quantified using the signal-to-background ratio and other qualitative measures. The results from this work provided clues on how to improve the detection of fluorescence photons from gold nanoparticle-loaded objects using the XFCT system. This will benefit future research on the development of the XFCT system in the context of making it more feasible for gold nanoparticle-based preclinical molecular imaging applications.

CANDU (CANada Deuterium Uranium) reactor utilizes Pressure Tube (PT) fuel channel design and heavy water as a coolant. Fuel channel annulus gas acts as an insulator to minimize heat losses from the coolant to the moderator. Since fuel bundles are continuously under high neutron fluxes, annulus gas nuclides undergo different nuclear transformations generating new composition of the gas that might have different physical properties which are undesirable for the annulus system. In addition, gas nuclides become radioactive and lead to an increase of the radioactive material inventory in the reactor and consequently to an increase of radiation levels. Pressure Tube Reactor (PTR) and Pressure Tube Supercritical Water Reactor (PT SCWR) fuel channel models have been developed in Monte Carlo N-Particle (MCNP) code. Neutron fluxes in the fuel channel annulus gas have been obtained by simulating different types of neutron sources in both PTR and PT SCWR fuel channels. Transmutation rates of annulus gases have been calculated for different gases (CO2, N2, Ar and Kr) at different pressures and temperatures in both fuel channels. The variation of the transmutation rates, neutron fluxes and gas densities in the annulus gas have been investigated in PTR and PT SCWR fuel channels at constant pressures and different temperatures. MCNP code along with NIST REFPROP [14] and other software tools have been used to conduct the calculations. / UOIT

Equispaced beam arrangements are typically used for IMRT plans. This beam arrangement provides adequate dose coverage to the target while sparing dose to other structures. However, an equispaced beam arrangement may not provide the best dose coverage to the target while sparing dose to the other structures. Beam angle optimization attempts to optimize the beam directions to produce a better IMRT plan; this is achieved by increasing dose to the target while minimizing dose to the remaining structures. Most methods of beam angle optimization attempt to optimize the beam angles and the beam intensity profiles to find an optimal set of beam angles. This thesis attempts to optimize the beam angles without determining the beam intensity profiles. An MCNP simulation is run to score the beam directions; the simulation is run as an adjoint problem to reduce simulation time, with the target as the source and the detectors scoring the dose for the gantry angles of the beam. Then, an optimization algorithm is run to select a set of beam angles for an optimized IMRT plan. The optimized IMRT plan is compared to an equispaced IMRT plan on a commercial treatment planning system to determine if this method of beam angle optimization is better than using an equispaced beam arrangement. The results of this thesis indicate that the coupling of an MCNP simulation for scoring with an optimization algorithm to select beam angles will produce a better IMRT plan than an equispaced IMRT plan. Three different geometries were used and for all geometries, the optimized IMRT plan had a higher average dose to the target while maintaining or increasing dose sparing to the critical structure and normal tissue.

Reprocessing nuclear fuel is becoming more viable in the United States due to the anticipated increase in construction of nuclear power plants, the growing stockpile of existing used nuclear fuel, and a public desire to reduce the amount of this fuel. However, a new reprocessing facility in non-weapon states must be safeguarded and new reprocessing facilities in weapon states will likely have safeguards due to political and material accountancy reasons. These facilities will have state of the art controls and monitoring methods to safeguard special nuclear materials, as well as to provide real-time monitoring. The focus of this project is to enable the development of a safeguards strategy that uses well established photon measurement methods to characterize samples from the UREX+3a reprocessing method using a variety of detector types and measurement times. It was determined that the errors from quantitative measurements were too large for traditional safeguards methods; however, a safeguards strategy based on qualitative gamma ray and neutron measurements is proposed. The gamma ray detection equipment used in the safeguard strategy could also be used to improve the real-time process monitoring in a yet-to-be built facility. A facility that had real-time gamma detection equipment could improve product quality control and provide additional benefits, such as waste volume reduction. In addition to the spectral analyses, it was determined by Monte Carlo N Particle (MCNP) simulations that there is no noticeable self shielding for internal pipe diameters less than 2 inches, indicating that no self shielding correction factors are needed. Further, it was determined that HPGe N-type detectors would be suitable for a neutron radiation environment. Finally, the gamma ray spectra for the measured samples were simulated using MCNP and then the model was extended to predict the responses from an actual reprocessing scenario from UREX+3a applied to fuel that had a decay time of three years. The 3-year decayed fuel was more representative of commercially reprocessed fuel than the acquired UREX+3a samples. This research found that the safeguards approach proposed in this paper would be best suited as an addition to existing safeguard strategies. Real-time gamma ray detection for process monitoring would be beneficial to a reprocessing facility and could be done with commercially available detectors.

An x-ray fluorescence detection system has been designed by our research group for quantifying the amount of gold nanoparticles presented within the phantom and animals during gold nanoparticle-aided cancer detection and therapy procedures. The primary components of the system consist of a microfocus x-ray source, a Pb beam collimator, and a CdTe photodiode detector. In order to optimize and facilitate future experimental tasks, a Monte Carlo model of the detection system has been created by using the MCNP5 code. Specifically, the model included an x-ray source, a Pb collimator, a CdTe detector, and an acrylic plastic phantom with four cylindrical columns where various materials such as gold nanoparticles, aluminum, etc. can be inserted during the experiments. In this model, 110 kVp x-rays emitted into a 60o cone from the focal spot of the x-ray source were collimated to a circular beam with a diameter of 5 mm. The collimated beam was then delivered to the plastic phantom with and without a gold nanoparticle-containing column. The fluence of scattered and gold fluorescence x-rays from the phantom was scored within the detector's sensitive volume resulting in various photon spectra and compared with the spectra acquired experimentally under the same geometry. The results show that the current Monte Carlo model can produce the results comparable to those from actual experiments and therefore it would serve as a useful tool to optimize and troubleshoot experimental tasks necessary for the development of gold nanoparticle-aided cancer detection and therapy procedures.

There are a variety of uses for reference neutron fields including detector response and dosimeter studies. The Georgia Institute of Technology has a 252Cf spontaneous fission source and an AmBe (α, n) source available for use in its research programs. In addition, it has iron, lead, beryllium, tantalum, heavy water, and polyethylene spheres to modify the neutron energy distributions from these neutron sources. This research characterized the neutron leakage spectra from the source inside spherical shells using a Bonner sphere spectrometer. All the neutron fields measured were also computed with a Monte Carlo code to determine the neutron fluence rate and ambient dose equivalent rate. The comparison of experimental data and calculations are used to provide further insight into the neutron spectra as modified by the spheres. The characterization of these modified sources will provide data to assist in using the resulting neutron fields in other research activities. To measure each neutron field combination, one of the two sources was placed in the center of an attenuating sphere. The neutron field was first measured at a variety of source-to-detector distances with a Bonner Sphere System. The spectrometer measurements, specifically the count rates of the different Bonner spheres, as a function of distance from the source is fitted to obtain corrections for room-scatter and air-scatter of neutrons using the Eisenhauer, Schwartz, and Johnson method. Using these corrections, the count rates free of room return is obtained at 1 m from the source and unfolded using the BUMS software to obtain the reported fluence and dose equivalent rates. These results are compared to those generated by the Monte Carlo Neutral Particle (MCNP) code. Models were made in MCNP for each of the source and moderating sphere combinations. The neutron fluence and dose rates were tallied during the MCNP simulation. The unfolded experimental data and the MCNP calculations showed good agreement for most of source-attenuating sphere combinations, thereby reinforcing the experimental results.

Commercial sterilization of medical devices may be performed using electron beam irradiators, which operate at various electron energies. The potential for activating components of the devices has been discussed, with current standards stating that an electron energy greater than 10 MeV requires assessment of potential induced radioactivity. There does not appear to be a literature citation for this energy limit, but it is the accepted default assumption within the industry. This research was directed at evaluating potential activation in medical products sterilized in electron beam as a function of the electron maximum energy. Monte Carlo simulation of a surrogate medical device was used to calculate photon and neutron fields resulting from electron irradiation, which were used to calculate concentrations for several radionuclides. The predominant mechanism for inducing radioactivity is photoneutron production in metal elements. Other mechanisms, including photoneutron production in deuterium with subsequent neutron capture, neutron capture of the photoneutrons produced in metal elements, and isomeric excitation, are all possible means of inducing radioactivity in similar conditions, but none made a perceptible contribution to activation in these experiments. The experiments confirmed that 10 MeV is a conservative assumption that any lower energy does not create significant activation. However, in the absence of a limited number of elements, the amount of induced radioactivity at 11 MeV and 12 MeV could also be considered insignificant. When based on an estimate of the amount of metal present in a medical device, the sum-of-fractions comparison to the US Nuclear Regulatory Commission exempt concentration limits is less than unity for all energies below 12.1 MeV, which suggests that there is minimal probability of significant induced activity at energies above the generally-accepted standard 10 MeV upper energy limit. / text

The focus in this thesis is to improve knowledge of the BWR related uncertainties void, channel bow, and control rods. The presence of void determines the moderation of neutrons in BWRs. A high void fraction is less efficient in moderating neutrons than a low one. As a consequence, the ratio of thermal to fast neutrons is dependent on the surrounding void fraction. In this thesis, calculations with 2D/3D codes corroborate this dependence, the void correlation, to be linear and very robust to changes in different reactor parameters. The void fraction could be predicted from the ratio of simultaneously measured reaction rates from thermal and fast neutron detectors over the whole core with an uncertainty of ±1.5%. The only parameter found disturbing the void correlation significantly is channel bow. However, since channel bow is the only phenomenon found biasing the void correlation, it is found that the void prediction methodology can be used to indicate channel bow with a sensitivity of 4% per mm bow. Consequently, large channel bows could easily be detected. Increased knowledge of void fractions and channel bow could increase both safety and economy of nuclear power production. This thesis also investigates how 2D/3D codes used in production perform in calculating detailed impact of control rods on pin powers and their ability to perform control rod depletion calculations in the reflector region. It is found that the axial resolution used in 3D nodal codes has very large impact on pin power gradients, i.e., using a standard nodal size of ~15 cm can cause underestimations of 50% in pin power gradients, which could lead to fuel damages. In addition, two methods for determining the neutron flux in the control rod when it is withdrawn from the core are presented. Both methods can be used in a 3D nodal code to reproduce the neutron flux in the reflector region with an uncertainty of ±3%. / Felaktigt tryckt som Digital Comprehensive Summaries of Uppsala Dissertations from the Faculty of Science and Technology 715