A scintillation counting system has been constructed with the use of BC-400 and EJ-212 series plastic scintillators along with a subminiature photomultiplier tube to investigate the effect of increasing plastic scintillator thickness on system-integrated counts. Measurements have been carried out using four different gamma sources with different energies ranging from 6keV to 1.332MeV and a Ni-63 beta source of maximum energy of 66keV. A simulation was also carried out in MCNP4a to verify the number of H-3 beta particles with max energy 18.6keV that would reach the plastic scintillator in a vacuum setting as well as in an air medium. Scintillator thicknesses ranged from 10μm to 2500μm. The response of the system was determined by measuring the integrated counts as a function of scintillator thickness. The results of these measurements showed the expected positive linear correlation between scintillator thicknesses and integrated counts for all the gamma sources while the slopes of the correlations of each gamma source was a function of the source energy. The beta particle response showed an initial increase of counts with scintillator thickness followed by a slight decrease. The MCNP simulation confirmed an analytical calculation of the fraction of H-3 beta particles for a given air concentration that would reach the scintillator. These results in conjunction with the experimental findings were used to assess the potential of a plastic scintillator system forming the basis of a tritium monitor for the detection of tritium in high-energy gamma backgrounds for Canadian nuclear power workers. / UOIT

The feasibility of using a thyroid uptake counter, normally used to measure the uptake of radioactive iodine in thyroid treatments, to assay radioactivity deposited in a persons lungs has been investigated. Variations in radioactive material distributions in the lungs, the response of the detector system to radionuclides of interest to homeland security, and the change in detection efficiency due to the varying thicknesses of intervening tissue of the victims have been simulated using the Monte Carlo N-Particle transport code (MCNP5) developed by Los Alamos National Laboratory. Point source and homogenously distributed models were created for Co-60, I-131, Cs-137, Ir-192, and Am-241 sources to simulate radiation transport between the lungs of multiple phantom models representing children and adults and the radiation detection system. To validate the simulations undertaken, the response of the counter to radiation sources in air and behind layers of Lucite have been modeled and compared to measured results.

A Varian 2100C 18 MV photon beam has been modeled in this work using the MCNP5 Monte Carlo particle transport user code. The subsequent beam irradiation was also delivered to a water phantom and benchmarked against experimentally measured depth dose data. The model presented in this work establishes the foundation to which further beam characteristics tuning is required in order to realistically model the beam mentioned above. It has been determined in this work that the initial electron beam energy of this beam model is sufficiently close to the electron beam energy from the linear accelerator used to obtain the benchmark depth dose data.

For the Texas A&M University Nuclear Science Center reactor a fuel depletion code was created to develop real-time fuel management capability. This code package links MCNP8 and ORIGEN26 and is interfaced through a Visual Basic code. Microsoft Visual Basic was used to create a user interface and for pre-and post-processing of MCNP and ORIGEN2 output. MCNP was used to determine the flux for all fuel and control rods within the core while ORIGEN2 used this flux along with the power history to calculate buildup and depletion for tracking the fuel isotopic evolution through time. A comparison of MCNP calculated fluxes and measured flux values were used to confirm the validity of the MCNP model. A comparison to Monteburns was used to add confidence to the correctness of the calculated fuel isotopics. All material isotopics were stored in a Microsoft Access database for integration with the Visual Basic code to allow for isotopics report generation for the Nuclear Science Center staff. This fuel management code performs its function with reasonable accuracy. It gathers minimal information from the user and burns the core over daily operation. After execution it stores all material data to the database for further use within NSCRFM or for isotopic report generation.

A new model for an adult human gastrointestinal tract (GIT) has been developed for use in internal dose estimations to the wall of the GIT and to the other organs and tissues of the body from radionuclides deposited in the lumenal contents of the five sections of the GIT. These sections were the esophagus, stomach, small intestine, upper large intestine, and the lower large intestine. The wall of each section was separated from its lumenal contents. Each wall was divided into many small regions so that the histologic and radiosensitive variations of the tissues across the wall could be distinguished. The characteristic parameters were determined based on the newest information available in the literature. Each of these sections except the stomach was subdivided into multiple subsections to include the spatiotemporal variations in the shape and characteristic parameters. This new GIT was integrated into an anthropomorphic phantom representing both an adult male and a larger-than-average adult female. The current phantom contains 14 different types of tissue. This phantom was coupled with the MCNP 4C Monte Carlo simulation package. The initial design and coding of the phantom and the Monte Carlo treatment employed in this study were validated using the results obtained by Cristy and Eckerman (1987). The code was used for calculating specific absorbed fractions (SAFs) in various organs and radiosensitive tissues from uniformly distributed sources of fifteen monoenergetic photons and electrons, 10 keV - 4 MeV, in the lumenal contents of the five sections of the GIT. The present studies showed that the average photon SAFs to the walls were significantly different from that to the radiosensitive cells (stem cells) for the energies below 50 keV. Above 50 keV, the photon SAFs were found to be almost constant across the walls. The electron SAF at the depth of the stem cells was a small fraction of the SAF routinely estimated at the contents-mucus interface. Electron studies showed that the “self-dose” for the energies below 300 keV and the “cross-dose” below 2 MeV were only from bremsstrahlung and fluorescent radiations at the depth of the stem cells and were not important.

abstract: As photons, electrons, and neutrons traverse a medium, they impart their energy in ways that are analytically difficult to describe. Monte Carlo methods provide valuable insight into understanding this behavior, especially when the radiation source or environment is too complex to simplify. This research investigates simulating various radiation sources using the Monte Carlo N-Particle (MCNP) transport code, characterizing their impact on various materials, and comparing the simulation results to general theory and measurements. A total of five sources were of interest: two photon sources of different incident particle energies (3.83 eV and 1.25 MeV), two electron sources also of different energies (30 keV and 100 keV), and a californium-252 (Cf-252) spontaneous fission neutron source. Lateral and vertical programmable metallization cells (PMCs) were developed by other researchers for exposure to these photon and electron sources, so simplified PMC models were implemented in MCNP to estimate the doses and fluences. Dose rates measured around the neutron source and the predicted maximum activity of activation foils exposed to the neutrons were determined using MCNP and compared to experimental results obtained from gamma-ray spectroscopy. The analytical fluence calculations for the photon and electron cases agreed with MCNP results, and differences are due to MCNP considering particle movements that hand calculations do not. Doses for the photon cases agreed between the analytical and simulated results, while the electron cases differed by a factor of up to 4.8. Physical dose rate measurements taken from the neutron source agreed with MCNP within the 10% tolerance of the measurement device. The activity results had a percent error of up to 50%, which suggests a need to further evaluate the spectroscopy setup. / Dissertation/Thesis / Masters Thesis Electrical Engineering 2014

Submitted by Marcele Costal de Castro (costalcastro@gmail.com) on 2017-12-14T13:55:26Z No. of bitstreams: 0 / Made available in DSpace on 2017-12-14T13:55:26Z (GMT). No. of bitstreams: 0 Previous issue date: 1999-08 / Visando ampliar ainda mais a gama de aplicações, demonstramos neste trabalho as potencialidades da neutrongrafia, aliada à tomografia computadorizada por transmissão, para a detecção de narcóticos e explosivos ocultos por diversos materiais para fins de Segurança Pública Nacional. Analisamos: (1) neutrongrafias térmicas de amostras de cocaína sob diferentes constituições físicas e graus de pureza, submetidas a ocultação por chumbo, fumo, ferro e tecido; (2) imagens tomográficas com feixes de nêutrons térmicos, obtidas a a partir de dados experimentais e simulados de amostras de drogas e explosivos ocultas por materiais leves e pesados. Nos ensaios experimentais, as amostras foram irradiadas com feixes de nêutrons térmicos provenientes do canal de irradiação J-9 do reator Argonauta IEN/CNEN, durante 30 minutos, num fluxo de 2,5x105 n.cm-2 .s-1. O código MCNP foi usado para gerar os dados de projeções para as tomografias simuladas e realizar estudos relativos à moderação, à colimação e à blindagem de um sistema neutrongráfico utilizando o 252Cf como fonte de nêutrons. A partir da análise das imagens das amostras de drogas, concluímos que o sistema neutrongráfico foi capaz de detectá-las, mesmo que ocultas nos materiais citados. Com imagens obtidas foi possível demonstrar o potencial da tomografia utilizando feixes de nêutrons térmicos para a mesma detecção. De acordo com os cálculos neutrônicos, estimamos um fluxo de nêutrons térmicos, no plano de imagem, de até 7x10 elevado à 5 potência n.cm.-2 .s-1 considerando 50 mg de 252Cf, para uma razão de colimação de 7,5.

Made available in DSpace on 2014-06-12T23:15:49Z (GMT). No. of bitstreams: 2 arquivo8634_1.pdf: 1784772 bytes, checksum: 8fd517fc2d3fb09302f697d0aba20ae8 (MD5) license.txt: 1748 bytes, checksum: 8a4605be74aa9ea9d79846c1fba20a33 (MD5) Previous issue date: 2008 / Coordenação de Aperfeiçoamento de Pessoal de Nível Superior / Em Tomografia por Emissão de Pósitrons (PET), os profissionais de saúde estão expostos a fótons de 511 keV, resultante do processo de aniquilação pósitron-elétron. Este valor é cerca de quatro vezes superior à energia média dos fótons com 140 keV, normalmente emitida em ambiente envolvendo Tomografia por Emissão de Fóton Único (SPECT). Apesar disso, aventais equivalentes a 0,5 mm de chumbo que já vem sendo utilizados em tarefas envolvendo a SPECT são empregados, também, na PET, independentemente da energia dos fótons emitidos. Neste contexto, este trabalho teve como objetivo avaliar a eficácia dos aventais equivalentes a 0,5 mm de chumbo na radioproteção individual de profissionais envolvidos em procedimentos para exames por PET. Para tanto, a energia média depositada por partícula foi calculada utilizando o método Monte Carlo, com auxílio do código MCNP4C, nas regiões correspondentes às grandezas operacionais Hp(10) e Hp(0,07), em duas situações de exposição individual: com e sem o uso do avental. Os resultados obtidos indicam que na região Hp(10) a dose absorvida com avental é estatisticamente igual a sem o uso do avental. Em relação à região Hp(0,07), o uso do avental acarreta um aumento de até 26% para a dose absorvida. Com base neste trabalho, aventais equivalentes a 0,5 mm de chumbo não oferecem proteção adequada aos profissionais de saúde envolvidos em procedimentos com Tomografia por Emissão de Pósitrons

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