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Submitted by Claudinei Pracidelli (cpracide@ipen.br) on 2015-06-09T18:28:55Z No. of bitstreams: 0 / Made available in DSpace on 2015-06-09T18:28:55Z (GMT). No. of bitstreams: 0 / A Terapia por Captura de Nêutrons pelo Boro (BNCT) é uma técnica radioterapêutica, que visa o tratamento de alguns tipos de câncer, em que sua energia útil é proveniente da reação nuclear promovida pela incidência de nêutrons térmicos no isótopo de 10B. No Brasil existe uma instalação, localizada junto ao canal de irradiação número 3 do Reator de Pesquisas IEA-R1 do IPEN, que foi projetada para o desenvolvimento de pesquisas em BNCT. Para uma aplicação adequada da técnica é necessário que o feixe de irradiação na posição de amostra, seja composto predominantemente por nêutrons térmicos com reduzida contaminação dos componentes do feixe, correspondente aos nêutrons epitérmicos e rápidos e à radiação gama. Este trabalho tem como objetivo monitorar e avaliar o feixe de irradiação na posição de irradiação de amostras, através do uso de detectores de ativação (folhas de ativação), e a partir de simulações utilizando o código de transporte de radiação, MCNP, avaliar mudanças na instalação, mais especificamente no conjunto de filtros e moderadores, para que se aprimore as condições de irradiação na instalação. O trabalho propos uma nova metodologia de cálculo para estudos de otimização do feixe a partir do recurso de redução de variância presente no MCNP, o wwg (weight window generation). Com os resultados obtidos através da adoção de um conjunto maior de folhas de ativação, foi possível a discriminação experimental do feixe de nêutrons em 5 faixas de energia e concluir que a instalação para estudos em BNCT do IPEN possui fluxo de nêutrons térmicos de 108 n/cm2.s, intensidade suficiente para que os estudos na área possam ser realizados com grande potencial de alteração de suas componentes conforme demanda experimental. / Dissertação (Mestrado em Tecnologia Nuclear) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Materiais radiologicamente equivalentes ao tecido humano (Tissue Equivalent Material - TEM) têm a finalidade de evitar exposição injustificável à radiação de pacientes e são amplamente utilizados no controle de qualidade de equipamentos de diagnóstico por imagem. Esses materiais devem ser caracterizados para que se possa confiar em sua semelhança, em termos de suas propriedades de interação com a matéria, aos tecidos que substituem. Uma das maneiras de caracterizar os materiais é verificando se o espectro transmitido através deles se assemelha ao transmitido através do tecido que substituem. O método Monte Carlo (MC) é uma ferramenta útil no processo de caracterização dos TEM pois pode evitar o processo custoso de realizar experimentos de transmissão de raios-X. Esse trabalho investigou a aplicabilidade do método MC à caracterização de TEM de tecido mamário (bTEM) utilizados no controle de qualidade de equipamentos de mamografia. Para verificar a aplicabilidade do método MC, uma série de resultados de simulações foi comparada a resultados experimentais. Espectros de raios X transmitidos foram comparados diretamente através da média de seus resíduos reduzidos (Mean Weighted Squared Residuals - MWSR). Comparações foram feitas através de grandezas derivadas dos espectros. Essas grandezas foram: as camadas semi-redutoras (primeira e segunda), a energia média e a energia efetiva. Foi realizada uma discussão acerca da eficiência de cada uma dessas comparações, através da estimativa do poder de cada teste de hipótese. Os experimentos de transmissão de radiação foram realizados em duas instalações, no Laboratório de Dosimetria das Radiações e Física Medida da Universidade de São Paulo, onde foi utilizado um tubo de raios X com anodo de tungstênio adaptado para qualidades de feixe utilizadas em aplicações mamográficas, e no Centro de Desenvolvimento de Tecnologia Nuclear da Comissão Nacional de Energia Nuclear, equipado com um mamógrafo clínico com anodos de tungstênio e molibdênio. Diversas condições experimentais foram variadas para assegurar a robustez das conclusões, tais como as combinações anodo/filtro, os materiais constituintes dos bTEM, suas glandularidades, espessuras e as tensões de pico. Os espectros sem nenhuma atenuação (0 mm) também foram medidos e utilizados nas comparações. Os espectros foram medidos com um detector comercial de CdTe. Dosímetros termoluminescentes foram utilizados para estimar a dose depositada em diversas regiões do bTEM, e esses resultados foram também comparados às simulações. Além da estimativa do nível de exatidão alcançado pelo código de MC nas referidas condições, também se concluiu que o teste de hipótese do MWSR teve o maior poder estatístico, de 0,996. O MWSR foi o teste que demonstrou a compatibilidade dos espectros medidos o maior número de vezes. Esse teste aceitou 48% dos pares de espectros contra 40% de aceitação do teste da primeira camada semi-redutora, que foi o segundo teste com maior aceitação. / man radiologically tissue equivalent materials (TEM) have the purpose of avoiding unjustifiable irradiation of patients; they are largely used in the quality control of image diagnostic equipment. These materials must be characterized so that their similarity to the tissues they simulate can be relied upon, regarding their properties of interaction with radiation. One way of characterizing the materials is by checking the resemblance between their transmitted spectrum to the one of tissue they simulate. The Monte Carlo (MC) method is a useful tool in the TEM characterization process, since it may avoid the realization of costly experiments of transmitted X-ray spectrometry. MC may even dismiss preliminary experiments. This work investigated the applicability of the MC method to the characterization of breast tissue TEM (bTEM) used in the quality control of mammography equipment. To evaluate the applicability of the MC method, a series of simulation results was compared to experimental data. Transmitted spectra were directly compared through their mean weighted squared residues (MWSR), and by the comparisons of spectra derived quantities, as it is commonly done in the literature. These quantities were: the half value layers (first and second), the mean energy and the effective energy. A discussion about the efficiency of each one of these comparisons was made by estimating the statistical power of each hypothesis test. The radiation transmission experiments were carried out in two facilities: at the Laboratory of Radiation Dosimetry and Medical Physics of the University of São Paulo, where a tungsten anode X-ray tube adapted to mammographic applications was used. The rest of the experiments was done at the Nuclear Technology Development Center of the National Commission of Nuclear Energy, equipped with a clinical mammographic equipment with anodes composed by tungsten and molybdenum. Several experimental conditions were varied to ensure the robustness of the conclusions, such as the anode/filter combination, the bTEM constituent materials, their glandularities, thicknesses and peak voltages. Spectra with no attenuation (0 mm) were also measured and used in the comparisons. The spectral measurements were done with a commercial CdTe detector. Thermo-luminescent dosimeters were used to estimate the dose deposited at several regions inside the bTEM, and these results were also compared to simulations. In addition to estimating the level of accuracy achieved by the MC code in the mentioned conditions, it was also concluded that the highest statistical power was scored by the MWSR and it was of 0.996. The MWSR was also the test which attested this compatibility of the measured spectra the most. It accepted 48% of the spectra pair against 40% acceptance of the first half value layer test, which was the second test with most acceptance.

No description available.

Neutron radiation detectors are an integral part of the Department of Homeland Security (DHS) efforts to detect the illicit trafficking of radioactive or special nuclear materials into the U.S. In the past decade, the DHS has deployed a vast network of radiation detection systems at various key positions to prevent or to minimize the risk associated with the malevolent use of these materials. The greatest portion of this detection burden has been borne by systems equipped with 3He because of its highly desirable physical and nuclear properties. However, a dramatic increase in demand and dwindling supply, combined with a lack of oversight for the existing 3He stockpile has produced a critical shortage of this gas which has virtually eliminated its viability for detector applications. A number of research efforts have been undertaken to develop suitable 3He replacements; however, these studies have been solely targeted toward simple detection cases where the overall detection efficiency is the only concern. For these cases, an insertion of additional detectors or materials can produce reaction rates that are sufficient, because the neutron spectral response is essentially irrelevant. However, in applications such as safeguards, non-proliferation efforts, and material control and accountability programs (MC&A), a failure to use detectors that are spectrally matched to 3He can potentially produce dire consequences. This is because these more difficult detection scenarios are associated with fissile material assessments for 239Pu and other actinides and these analyses have almost universally been calibrated to an equivalent 3He response. In these instances, a “simple” detector or material addition approach is neither appropriate nor possible, due to influences resulting from the complex nature of neutron scattering in moderators, cross sections, gas pressure variations, geometries, and surrounding structural interference. These more challenging detection cases require a detailed computational transport analysis be performed for each specific application. A leveraged approach using adjoint transport computations that are validated by forward transport and Monte Carlo computations and laboratory measurements can address these more complex detection cases and this methodology was utilized in the execution of the research. The initial task was to establish the fidelity of a computational approach by executing radiation transport models for existing BF3 and 3He tubes and then comparing the modeling results to laboratory measurements made using these identical devices. Both tubes were 19.6 cm in height, 1-inch in diameter, and operated at 1 and 4 atm pressure respectively. The models were processed using a combination of forward Monte Carlo and forward and adjoint 3-D discrete ordinates (SN) transport methods. The computer codes MCNP5 and PENTRAN were used for all calculations of a nickel-shielded plutonium-beryllium (PuBe) source term that provided a neutron output spectra equivalent to that of weapons-grade plutonium (WGPu). Once the computational design approach was validated, the adjoint SN method was used to iteratively identify six distinct plug-in models that matched the neutron spectral response and reaction rate of a 1-inch diameter 3He tube with a length of 10 cm and operating at 4 atm pressure. The equivalent designs consist of large singular tubes and dual tubes containing BF3 gas, 10B linings, and/or 10B-loaded polyvinyl toluene (PVT). The reaction rate for each plug-in design was also verified using forward PENTRAN and MCNP5 calculations. In addition to the equivalent designs, the adjoint method also yielded various insights into neutron detector design that can lead to additional designs using a combination of different detector materials such as BF3/10B-loaded PVT, 10B-lined tubes/10B-loaded PVT, etc.

Materiais radiologicamente equivalentes ao tecido humano (Tissue Equivalent Material - TEM) têm a finalidade de evitar exposição injustificável à radiação de pacientes e são amplamente utilizados no controle de qualidade de equipamentos de diagnóstico por imagem. Esses materiais devem ser caracterizados para que se possa confiar em sua semelhança, em termos de suas propriedades de interação com a matéria, aos tecidos que substituem. Uma das maneiras de caracterizar os materiais é verificando se o espectro transmitido através deles se assemelha ao transmitido através do tecido que substituem. O método Monte Carlo (MC) é uma ferramenta útil no processo de caracterização dos TEM pois pode evitar o processo custoso de realizar experimentos de transmissão de raios-X. Esse trabalho investigou a aplicabilidade do método MC à caracterização de TEM de tecido mamário (bTEM) utilizados no controle de qualidade de equipamentos de mamografia. Para verificar a aplicabilidade do método MC, uma série de resultados de simulações foi comparada a resultados experimentais. Espectros de raios X transmitidos foram comparados diretamente através da média de seus resíduos reduzidos (Mean Weighted Squared Residuals - MWSR). Comparações foram feitas através de grandezas derivadas dos espectros. Essas grandezas foram: as camadas semi-redutoras (primeira e segunda), a energia média e a energia efetiva. Foi realizada uma discussão acerca da eficiência de cada uma dessas comparações, através da estimativa do poder de cada teste de hipótese. Os experimentos de transmissão de radiação foram realizados em duas instalações, no Laboratório de Dosimetria das Radiações e Física Medida da Universidade de São Paulo, onde foi utilizado um tubo de raios X com anodo de tungstênio adaptado para qualidades de feixe utilizadas em aplicações mamográficas, e no Centro de Desenvolvimento de Tecnologia Nuclear da Comissão Nacional de Energia Nuclear, equipado com um mamógrafo clínico com anodos de tungstênio e molibdênio. Diversas condições experimentais foram variadas para assegurar a robustez das conclusões, tais como as combinações anodo/filtro, os materiais constituintes dos bTEM, suas glandularidades, espessuras e as tensões de pico. Os espectros sem nenhuma atenuação (0 mm) também foram medidos e utilizados nas comparações. Os espectros foram medidos com um detector comercial de CdTe. Dosímetros termoluminescentes foram utilizados para estimar a dose depositada em diversas regiões do bTEM, e esses resultados foram também comparados às simulações. Além da estimativa do nível de exatidão alcançado pelo código de MC nas referidas condições, também se concluiu que o teste de hipótese do MWSR teve o maior poder estatístico, de 0,996. O MWSR foi o teste que demonstrou a compatibilidade dos espectros medidos o maior número de vezes. Esse teste aceitou 48% dos pares de espectros contra 40% de aceitação do teste da primeira camada semi-redutora, que foi o segundo teste com maior aceitação. / man radiologically tissue equivalent materials (TEM) have the purpose of avoiding unjustifiable irradiation of patients; they are largely used in the quality control of image diagnostic equipment. These materials must be characterized so that their similarity to the tissues they simulate can be relied upon, regarding their properties of interaction with radiation. One way of characterizing the materials is by checking the resemblance between their transmitted spectrum to the one of tissue they simulate. The Monte Carlo (MC) method is a useful tool in the TEM characterization process, since it may avoid the realization of costly experiments of transmitted X-ray spectrometry. MC may even dismiss preliminary experiments. This work investigated the applicability of the MC method to the characterization of breast tissue TEM (bTEM) used in the quality control of mammography equipment. To evaluate the applicability of the MC method, a series of simulation results was compared to experimental data. Transmitted spectra were directly compared through their mean weighted squared residues (MWSR), and by the comparisons of spectra derived quantities, as it is commonly done in the literature. These quantities were: the half value layers (first and second), the mean energy and the effective energy. A discussion about the efficiency of each one of these comparisons was made by estimating the statistical power of each hypothesis test. The radiation transmission experiments were carried out in two facilities: at the Laboratory of Radiation Dosimetry and Medical Physics of the University of São Paulo, where a tungsten anode X-ray tube adapted to mammographic applications was used. The rest of the experiments was done at the Nuclear Technology Development Center of the National Commission of Nuclear Energy, equipped with a clinical mammographic equipment with anodes composed by tungsten and molybdenum. Several experimental conditions were varied to ensure the robustness of the conclusions, such as the anode/filter combination, the bTEM constituent materials, their glandularities, thicknesses and peak voltages. Spectra with no attenuation (0 mm) were also measured and used in the comparisons. The spectral measurements were done with a commercial CdTe detector. Thermo-luminescent dosimeters were used to estimate the dose deposited at several regions inside the bTEM, and these results were also compared to simulations. In addition to estimating the level of accuracy achieved by the MC code in the mentioned conditions, it was also concluded that the highest statistical power was scored by the MWSR and it was of 0.996. The MWSR was also the test which attested this compatibility of the measured spectra the most. It accepted 48% of the spectra pair against 40% acceptance of the first half value layer test, which was the second test with most acceptance.

京都大学 / 0048 / 新制・課程博士 / 博士(エネルギー科学) / 甲第18664号 / エネ博第309号 / 新制||エネ||63(附属図書館) / 31578 / 京都大学大学院エネルギー科学研究科エネルギー応用科学専攻 / (主査)教授 大垣 英明, 教授 白井 康之, 教授 松田 一成 / 学位規則第4条第1項該当 / Doctor of Energy Science / Kyoto University / DFAM

The department of Radiation Sciences at Umeå University has obtained an old linear accelerator, intended for educational purposes. The goal of this thesis was to find proper reinforced radiation protection in an intended bunker (a room with thick concrete walls), to ensure that the radiation outside the bunker falls within acceptable levels. The main method was with the use of Monte Carlo-simulations. To properly simulate the accelerator, knowledge of the energy distribution of emitted radiation was needed. For this, a novel method for spectra determination, using several depth dose measurements including off-axis, was developed. A method that shows promising results in finding the spectra when measurements outside the primary beam are included. The found energy spectrum was then used to simulate the accelerator in the intended bunker. The resulting dose distribution was visualized together with 3D CAD-images of the bunker, to easily see in which locations outside the bunker where the dose was high. An important finding was that some changes are required to ensure that the public does not receive too high doses of radiation on a public outdoor-area that is located above the bunker. Otherwise, the accelerator is only allowed to be run 1.8 hours per year. A workaround to this problem could be to just plant a thorn bush, covering the dangerous area of radius 3m. After such a measure has been taken, which is assumed in the following results, the focus moves to the radiation that leaks into the accelerator’s intended control room, which is located right outside the bunker’s entrance door. The results show that the accelerator is only allowed to be run for a maximum of 6.1 or 3.3 hours per year (depending on the placement of the accelerator in the room), without a specific extra reinforced radiation protection consisting mainly of lead bricks. With the specific extra protection added, the accelerator is allowed to be run 44 or 54 hours per year instead, showing a distinct improvement. However, the dose rate to the control room was still quite high, 13.7 μGy/h or 11.2 μGy/h, compared to the average dose received by someone living in Sweden, which is 0.27 μGy/h. Therefore, further measures are recommended. This is however a worst case scenario, since the leakage spectrum from the accelerator itself was simulated as having the same energy spectrum as the primarybeam at 0.1 % of the intensity, which is the maximum leakage dose according to the specifications for the accelerator. This is probably an overestimation of the intensity. Also, the energy spectra of the leakage is probably of lower energy than the primary beam in at least some directions. Implementing more knowledge of the leak spectra in future work, should therefore result in more allowed run hours for the accelerator.

Effizienzsteigernde Methoden haben die Aufgabe, die Rechenzeit von Monte Carlo Simulationen zur Lösung von Strahlungstransportproblemen zu verringern. Dazu gehören weitergehende Quell- oder Geometrievereinfachungen und die Gewichtsfenstertechnik als wichtigstes varianzreduzierendes Verfahren, entwickelt in den 1950er Jahren. Die Schwierigkeit besteht bis heute in der Berechnung geeigneter Gewichtsfenster. In dieser Arbeit wird ein orts- und energieabhängiger Gewichtsfenstergenerator basierend auf dem vorwärts-adjungierten Generator von T.E. BOOTH und J.S. HENDRICKS für das Strahlungstransportprogramm AMOS entwickelt und implementiert. Dieser ist in der Lage, die Gewichtsfenster sowohl iterativ zu berechnen und automatisch zu setzen als auch, deren Energieeinteilung selbstständig anzupassen. Die Arbeitsweise wird anhand des Problems der tiefen Durchdringung von Photonenstrahlung demonstriert, wobei die Effizienz um mehrere Größenordnungen gesteigert werden kann. Energieabhängige Gewichtsfenster sorgen günstigstenfalls für eine weitere Verringerung der Rechenzeit um etwa eine Größenordnung. Für eine praxisbezogene Problemstellung, die Bestrahlung eines Personendosimeters, kann die Effizienz hingegen bestenfalls vervierfacht werden. Quell- und Geometrieveränderungen sind gleichwertig. Energieabhängige Fenster zeigen keine praxisrelevante Effizienzsteigerung. / The purpose of efficiency increasing methods is the reduction of the computing time required to solve radiation transport problems using Monte Carlo techniques. Besides additional geometry manipulation and source biasing this includes in particular the weight windows technique as the most important variance reduction method developed in the 1950s. To date the difficulty of this technique is the calculation of appropriate weight windows. In this work a generator for spatial and energy dependent weight windows based on the forward-adjoint generator by T.E. BOOTH and J.S. HENDRICKS is developed and implemented in the radiation transport program AMOS. With this generator the weight windows are calculated iteratively and set automatically. Furthermore the generator is able to autonomously adapt the energy segmentation. The functioning is demonstrated by means of the deep penetration problem of photon radiation. In this case the efficiency can be increased by several orders of magnitude. With energy dependent weight windows the computing time is decreased additionally by approximately one order of magnitude. For a practice-oriented problem, the irradiation of a dosimeter for individual monitoring, the efficiency is only improved by a factor of four at best. Source biasing and geometry manipulation result in an equivalent improvement. The use of energy dependent weight windows proved to be of no practical relevance.