A técnica de captura de nêutrons em Boro é uma técnica promissora de tratamento de câncer, ela usa do alto LET das partículas provenientes da reação 10B(n,α)7Li para destruir as células cancerígenas. O desenvolvimento desta técnica começou em meados da década de 50 e até hoje ela é alvo de estudos e pesquisas em diversos centros espalhados pelo mundo, no Brasil construiu-se uma instalação que tem como objetivo realizar pesquisas em BNCT, esta instalação está localizada junto ao canal de irradiação número três do reator nuclear de pesquisa IEA-R1 e possui uma blindagem biológica projetada para atender as normas de radioproteção. Esta blindagem biológica foi desenvolvida para permitir que se realizem experimentos com o reator ligado a potência máxima, fazendo com que não seja necessário ligar e desligar o reator para se irradiar amostras. Entretanto quando se abre o canal de irradiação o background do salão de experimentos do salão de experimentos aumenta e esta variação de background inviabiliza a realização das medidas do grupo de pesquisa em difração de nêutrons que utiliza o canal de irradiação número seis. Este trabalho tem como objetivo acrescentar melhorias na blindagem a fim de reduzir ao máximo essa variação de background fazendo com que seja possível realizar medidas na instalação de pesquisas em BNCT sem interferir nas medidas do grupo de pesquisa do canal de irradiação seis. Para isto, utilizou o código MCNP5, dosímetros termoluminescentes e detectores de ativação tipo folha para planejar melhorias na blindagem biológica. Calculou-se com o auxílio do código uma melhoria que consegue reduzir em média o fluxo térmico em 71,2 ± 13 % e verificou-se experimentalmente uma redução média de 70 ± 9 % na dose devido aos nêutrons térmicos. / The technique of neutron capture in boron is a promising technique in cancer treatment, it uses the high LET particles from the reaction 10B (n, α) 7Li to destroy cancer cells.The development of this technique began in the mid-\'50s and even today it is the object of study and research in various centers around the world, Brazil has built a facility that aims to conduct research in BNCT, this facility is located next to irradiation channel number three at the research nuclear reactor IEA-R1 and has a biological shielding designed to meet the radiation protection standards. This biological shielding was developed to allow them to conduct experiments with the reactor at maximum power, so it is not necessary to turn on and off the reactor to irradiate samples. However, when the channel is opened for experiments the background radiation in the experiments salon increases and this background variation makes it impossible to perform measurements in a neutron diffraction research that utilizes the irradiation channel number six. This study aims to further improve the shielding in order to minimize the variation of background making it possible to perform the research facility in BNCT without interfering with the action of the research group of the irradiation channel number six. To reach this purpose, the code MCNP5, dosimeters and activation detectors were used to plan improvements in the biological shielding. It was calculated with the help of the code an improvement that can reduce the average heat flow in 71.2% ± 13 and verified experimentally a mean reduce of 70 ± 9% in dose due to thermal neutrons.

A técnica de captura de nêutrons em Boro é uma técnica promissora de tratamento de câncer, ela usa do alto LET das partículas provenientes da reação 10B(n,α)7Li para destruir as células cancerígenas. O desenvolvimento desta técnica começou em meados da década de 50 e até hoje ela é alvo de estudos e pesquisas em diversos centros espalhados pelo mundo, no Brasil construiu-se uma instalação que tem como objetivo realizar pesquisas em BNCT, esta instalação está localizada junto ao canal de irradiação número três do reator nuclear de pesquisa IEA-R1 e possui uma blindagem biológica projetada para atender as normas de radioproteção. Esta blindagem biológica foi desenvolvida para permitir que se realizem experimentos com o reator ligado a potência máxima, fazendo com que não seja necessário ligar e desligar o reator para se irradiar amostras. Entretanto quando se abre o canal de irradiação o background do salão de experimentos do salão de experimentos aumenta e esta variação de background inviabiliza a realização das medidas do grupo de pesquisa em difração de nêutrons que utiliza o canal de irradiação número seis. Este trabalho tem como objetivo acrescentar melhorias na blindagem a fim de reduzir ao máximo essa variação de background fazendo com que seja possível realizar medidas na instalação de pesquisas em BNCT sem interferir nas medidas do grupo de pesquisa do canal de irradiação seis. Para isto, utilizou o código MCNP5, dosímetros termoluminescentes e detectores de ativação tipo folha para planejar melhorias na blindagem biológica. Calculou-se com o auxílio do código uma melhoria que consegue reduzir em média o fluxo térmico em 71,2 ± 13 % e verificou-se experimentalmente uma redução média de 70 ± 9 % na dose devido aos nêutrons térmicos. / The technique of neutron capture in boron is a promising technique in cancer treatment, it uses the high LET particles from the reaction 10B (n, α) 7Li to destroy cancer cells.The development of this technique began in the mid-\'50s and even today it is the object of study and research in various centers around the world, Brazil has built a facility that aims to conduct research in BNCT, this facility is located next to irradiation channel number three at the research nuclear reactor IEA-R1 and has a biological shielding designed to meet the radiation protection standards. This biological shielding was developed to allow them to conduct experiments with the reactor at maximum power, so it is not necessary to turn on and off the reactor to irradiate samples. However, when the channel is opened for experiments the background radiation in the experiments salon increases and this background variation makes it impossible to perform measurements in a neutron diffraction research that utilizes the irradiation channel number six. This study aims to further improve the shielding in order to minimize the variation of background making it possible to perform the research facility in BNCT without interfering with the action of the research group of the irradiation channel number six. To reach this purpose, the code MCNP5, dosimeters and activation detectors were used to plan improvements in the biological shielding. It was calculated with the help of the code an improvement that can reduce the average heat flow in 71.2% ± 13 and verified experimentally a mean reduce of 70 ± 9% in dose due to thermal neutrons.

The objective of this research was to develop a complex MCNPX model of the human head to predict absorbed dose distributions during proton therapy of ocular tumors. Absorbed dose distributions using the complex geometry were compared to a simple MCNPX model of the human eye developed by Oertli. The proton therapy beam used at Laboratori Nazionali del Sud-INFN was chosen for comparison. Dose calculations included dose due to proton and secondary interactions, multiple coulombic energy scattering, elastic and inelastic scattering, and non-elastic nuclear reactions. Benchmarking MCNPX was accomplished using the proton simulations outlined by Oertli. Once MCNPX was properly benchmarked, the proton beam and MCNPX models were combined to predict dose distributions for three treatment scenarios. First, an ideal treatment scenario was modeled where the dose was maximized to the tumor volume and minimized elsewhere. The second situation, a worst case scenario, mimicked a patient starring directly into the treatment beam during therapy. During the third simulation, the treatment beam was aimed into the bone surrounding the eye socket to estimate the dose to the vital regions of the eye due to scattering. Dose distributions observed for all three cases were as expected. Superior dose distributions were observed with the complex geometry for all tissues of the phantom and the tumor volume. This study concluded that complex MCNPX geometries, although initially difficult to implement, produced superior dose distributions when compared to simple models.

The objective of this research was to develop a complex MCNPX model of the human head to predict absorbed dose distributions during proton therapy of ocular tumors. Absorbed dose distributions using the complex geometry were compared to a simple MCNPX model of the human eye developed by Oertli. The proton therapy beam used at Laboratori Nazionali del Sud-INFN was chosen for comparison. Dose calculations included dose due to proton and secondary interactions, multiple coulombic energy scattering, elastic and inelastic scattering, and non-elastic nuclear reactions. Benchmarking MCNPX was accomplished using the proton simulations outlined by Oertli. Once MCNPX was properly benchmarked, the proton beam and MCNPX models were combined to predict dose distributions for three treatment scenarios. First, an ideal treatment scenario was modeled where the dose was maximized to the tumor volume and minimized elsewhere. The second situation, a worst case scenario, mimicked a patient starring directly into the treatment beam during therapy. During the third simulation, the treatment beam was aimed into the bone surrounding the eye socket to estimate the dose to the vital regions of the eye due to scattering. Dose distributions observed for all three cases were as expected. Superior dose distributions were observed with the complex geometry for all tissues of the phantom and the tumor volume. This study concluded that complex MCNPX geometries, although initially difficult to implement, produced superior dose distributions when compared to simple models.

This thesis describes a tool called TXSAMC (Transport Cross Sections from Applied Monte Carlo) that produces shielded and homogenized multigroup cross sections for small fast reactor systems. The motivation for this tool comes from a desire to investigate reactor systems that are not characterized well by existing tools. Proper investigation usually requires the use of deterministic codes to characterize the timedependent reactor behavior and to link reactor neutronics codes with thermal-hydraulics and/or other physics codes. Deterministic codes require an accurate set of multigroup cross section libraries. The current process for generating these libraries is time consuming. TXSAMC offers a shorter route for generating these libraries. TXSAMC links three external codes together to create these libraries. The code creates an MCNP (Monte Carlo N-Particle) model of the reactor and calculates the zoneaveraged scalar flux in various tally regions and a core-averaged scalar flux tallied by energy bin. The core-averaged scalar flux provides a weighting function for NJOY. The zone-averaged scalar flux data is used in TRANSX for homogenization and shielding. The code runs NJOY to produce multigroup cross sections that are tabulated by nuclide, temperature and background cross section in MATXS (Material-wise cross section) format. This library is read by TRANSX which, in conjunction with the RZFLUX (Regular Zone-averaged Flux) files, shields the cross sections and homogenizes them. The result is a macroscopic cross section for the cell within the reactor from which the RZFLUX file was written. The cross sections produced by this process have been tested in five different sample problems and have been shown to be reasonably accurate. For reactor cells containing fuel pins, the typical error in the overall fission, nusigf, (n,2n), absorption and total RRD is only a few percent and is often less than one percent. It appears that the error is less for hexagonal lattices than for square lattices. A significant amount of error is associated with threshold reactions like (n,2n) in the sodium coolant. For the square lattice test problems, a reduction in error occurs when smaller tally regions are selected. This reduction was not observed for hexagonal lattice reactors. Overall, the cross sections produced by TXSAMC performed very well.

In this report an MCNP (Monte Carlo N-Particle) model is described for the reactor vessel at Ringhals 2. The model is validated against the specific activity in neutron dosimeters, extracted in 1977, 1984 and 1994. The validation showed that the calculations of the model are within the requirements of a maximum of 20 percent uncertainty for every neutron dosimeter except one, extracted after the first cycle. The uncertainty of this cycle was mostly due to the operation data rather than to the MCNP model. The model has been used to investigate various questions concerning radiation damage. The reliability of the traditional measure of radiation damage, fast neutron flux (En > 1MeV) has been evaluated.  This has been done by taking the ratio for this and another measure of radiation damage, DPA (Displacement Per Atom), for various positions and layers. The results show good reliability, except for at the outer layers of the vessel wall, where the traditional measure underestimates the radiation damage. Inspections are carried out in connection with the change of fuel to investigate any possible cracking on the internal structures of the reactor vessel.  New data on local differences in the radiation of these have therefore been calculated for future evaluations.  This is in order to be able to focus the inspections mainly on those internal parts that are exposed to the highest dose of radiation. An estimation of the neutron dose after 40, 50 and 60 years of operation has been calculated for the surface of the reactor vessel that is being exposed to the highest neutron flux. The result confirms earlier appreciation that the radiation damage to the reactor vessel is not a limiting factor for the future operation of Ringhals 2. The report also presents which surface of the vessel wall that has been exposed to a neutron dose of 1017 n/cm2 for neutrons with En > 1 MeV.

No description available.

Este trabalho apresenta uma metodologia de cálculo para a obtenção de doses em estruturas do olho humano, como: esclera, coróide, retina, nervo óptico, corpo vítreo, câmara anterior, lente, além do tumor devido ao tratamento com placas oftálmicas. Construiu-se um modelo de olho humano com suas principais estruturas e dimensões fieis, além de um modelo matemático para uma placa de Co-60 e uma placa de sementes de I-125, levando-se em conta tamanho e disposição geométrica das fontes reais, com o código de Monte Carlo MCNP-4C. Esse modelo é capaz de calcular as distribuições de dose axiais e radiais para qualquer ponto do olho e para cada uma de suas estruturas. Construiu-se, também, um simulador de acrílico para o olho. Esse simulador é formado por uma esfera de acrílico fatiada em lâminas de 1 mm de espessura para simular as mesmas condições de simulação realizada pelos código MCNP-4C, fornecendo as doses axiais e radiais em filmes radiográficos. O simulador foi utilizado para validar os cálculos realizados com o código MCNP-4C. Os dados obtidos desse modelo matemático servirão para montar um banco de dados de doses para todas as estruturas do olho, posições e tamanhos de tumores e quaisquer placas oftálmicas utilizadas para tratamento. Esse banco de dados será a parte principal para a construção de um “software" nacional para cálculos de dose, que poderá fazer parte de um sistema de planejamento confiável para ser utilizado em radioterapia/braquiterapia. / This work presents a calculational methodology for dose determination in human eye structures, such as: sclera, choroid, retina, lens, vitreous body, optic nerve and disc, and cornea, as well as tumor due to treatment to the eye plaques. A human eye model was constructed taking into consideration its main structural and dimension characteristics. Beyond that a mathematical model for the Co-60 and I-125 plaques with all geometric details were built employing the MCNP-4C code. This model is able to calculate the axial and radial doses in any point of the eye and for each of its structures. An acrylic eye simulator was also built with the aim to obtain experimental results for the both model validations. This simulator is made of an acrylic sphere split into foils of 1 mm thickness which allow the introduction a radiographic film to measure the axial and radial doses. The experimental data were used to validate the MCNP-4C results. The data from the mathematical model will serve as the basis to build a data bank for all the eye structures allowing different position and sizes of tumor as well as the replacement of all ophthalmic plaques used in the treatment. This data bank will be the principal part for the construction of a national software for the dose calculation and can be of great help for a reliable treatment system planning in radiotherapy/brachytherapy.

Este trabalho apresenta uma metodologia de cálculo para a obtenção de doses em estruturas do olho humano, como: esclera, coróide, retina, nervo óptico, corpo vítreo, câmara anterior, lente, além do tumor devido ao tratamento com placas oftálmicas. Construiu-se um modelo de olho humano com suas principais estruturas e dimensões fieis, além de um modelo matemático para uma placa de Co-60 e uma placa de sementes de I-125, levando-se em conta tamanho e disposição geométrica das fontes reais, com o código de Monte Carlo MCNP-4C. Esse modelo é capaz de calcular as distribuições de dose axiais e radiais para qualquer ponto do olho e para cada uma de suas estruturas. Construiu-se, também, um simulador de acrílico para o olho. Esse simulador é formado por uma esfera de acrílico fatiada em lâminas de 1 mm de espessura para simular as mesmas condições de simulação realizada pelos código MCNP-4C, fornecendo as doses axiais e radiais em filmes radiográficos. O simulador foi utilizado para validar os cálculos realizados com o código MCNP-4C. Os dados obtidos desse modelo matemático servirão para montar um banco de dados de doses para todas as estruturas do olho, posições e tamanhos de tumores e quaisquer placas oftálmicas utilizadas para tratamento. Esse banco de dados será a parte principal para a construção de um “software” nacional para cálculos de dose, que poderá fazer parte de um sistema de planejamento confiável para ser utilizado em radioterapia/braquiterapia. / This work presents a calculational methodology for dose determination in human eye structures, such as: sclera, choroid, retina, lens, vitreous body, optic nerve and disc, and cornea, as well as tumor due to treatment to the eye plaques. A human eye model was constructed taking into consideration its main structural and dimension characteristics. Beyond that a mathematical model for the Co-60 and I-125 plaques with all geometric details were built employing the MCNP-4C code. This model is able to calculate the axial and radial doses in any point of the eye and for each of its structures. An acrylic eye simulator was also built with the aim to obtain experimental results for the both model validations. This simulator is made of an acrylic sphere split into foils of 1 mm thickness which allow the introduction a radiographic film to measure the axial and radial doses. The experimental data were used to validate the MCNP-4C results. The data from the mathematical model will serve as the basis to build a data bank for all the eye structures allowing different position and sizes of tumor as well as the replacement of all ophthalmic plaques used in the treatment. This data bank will be the principal part for the construction of a national software for the dose calculation and can be of great help for a reliable treatment system planning in radiotherapy/brachytherapy.

"The MCNP6 radiation transport code was used to quantify changes in the absorbed dose tissue conversion factors for LiF, Al2O3, and silicon-based electronic dosimeters. While normally calibrated in-air and applied to all general geometric measurements, tissue conversion factors for each dosimeter were obtained at various depths in a simulated water phantom and compared against the standard in-air calibration method. In these experiments, a mono-energetic photon source was modeled at energies between 30 keV and 300 keV for a point-source placed at the center of a water phantom, a point-source placed at the surface of the phantom, and for a 10-cm radial field geometry. Again, mono-energetic photon source was modeled up to 1300 keV for a disk-source placed at the surface of the phantom and dosimetric calculations were obtained for water, LiF, Al2O3, and silicon at depths of 1 mm to 35 cm from the source. The dosimeter’s absorbed dose conversion factor was calculated as a ratio of the absorbed dose to water to that of the dosimeter measured at a specified phantom depth. The dosimeter’s calibration value also was obtained for both mono and polyenergetic source and the calibration value from poly-energetic source was compared with the intensity weighted average calibration value from mono-energetic photon. The calculated changes in the tissue conversion factors are significant because the American Association of Physicists in Medicine (AAPM) recommend that measurements of a brachytherapy or diagnostic source be made with an overall uncertainity of 5% or better. Yet, based on results, the absorbed dose tissue conversion factor for a LiF dosimeter was found to deviate from its calibration value by up to 9%, an Al2O3 dosimeter by 43%, and a silicon dosimeter by 61%. These uncertainties are in addition to the normal measurement uncertainties. By applying these tissue correction factors, these data may be used to meet the AAPM measurement requirements for mono-energetic and poly-energetic sources at measurement depths up to 35 cm under the irradiation geometries investigated herein. "