Dissertations / Theses on the topic 'Neutron dose'
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Wooten, Hasani Omar. "Time-Dependent Neutron and Photon Dose-Field Analysis." Diss., Georgia Institute of Technology, 2005. http://hdl.handle.net/1853/7153.
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A unique tool is developed that allows the user to model physical representations of complicated glovebox facilities in two dimensions and determine neutral-particle flux and ambient dose-equivalent fields throughout that geometry. The code Pandemonium, originally designed to determine flux and dose rates only, has been improved to include realistic glovebox geometries, time-dependent source and detector positions, time-dependent shielding thickness calculations, time-integrated doses, a representative criticality accident scenario based on time-dependent reactor kinetics, and more rigorous photon treatment. The photon model has been significantly enhanced by expanding the energy range to 10 MeV to include fission photons, and by including a set of new buildup factors, the result of an extensive study into the previously unknown "purely-angular effect" on photon buildup. Purely-angular photon buildup factors are determined using discrete ordinates and coupled electron-photon cross sections to account for coherent and incoherent scattering and secondary photon effects of bremsstrahlung and florescence.
Improvements to Pandemonium result in significant modeling capabilities for processing facilities using intense neutron and photon sources, and the code obtains comparable results to Monte Carlo calculations but within a fraction of the time required to run such codes as MCNPX.
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Veinot, Kenneth Guy. "An angular dependent neutron effective-dose-equivalent dosimeter." Diss., Georgia Institute of Technology, 1999. http://hdl.handle.net/1853/17595.
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Seppälä, Tiina. "FiR epithermal neutron beam model and dose calculation for treatment planning in neutron capture therapy." Helsinki : University of Helsinki, 2002. http://ethesis.helsinki.fi/julkaisut/mat/fysik/vk/seppala/.
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Simpkins, Robert W. "Neutron organ dose and the influence of adipose tissue." Diss., Georgia Institute of Technology, 2003. http://hdl.handle.net/1853/18959.
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Phoenix, Ben. "Synergistic and dose rate effects in Boron Neutron Capture Therapy." Thesis, University of Birmingham, 2013. http://etheses.bham.ac.uk//id/eprint/4084/.
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An investigation of the factors affecting the biological effectiveness of neutron beams suitable for Boron Neutron Capture Therapy (BNCT) has been carried out. The primary experimental work described in this thesis concerns the degree of interaction, if any, between biological damage caused by low LET radiation and that caused by high LET radiation. The second area investigated concerns the biological impact of delivering a BNCT irradiation at differing dose rates. In mixed photon alpha particle irradiations, no synergistic effect was observed above the response from the separate components. Maximum alpha particle doses delivered were 2.54 Gy. In mixed X-ray and alpha particle exposures, no synergy effect was seen with 2.54 Gy of alpha particles delivered to the cells. At the 3.18Gy alpha particle dose level significantly lower cell survival was observed than would be predicted from survival in single fields. Dose rate experiments were carried out in the Massachusetts Institute of Technology (MIT) Fission Converter neutron Beam (FCB). Cells loaded with boric acid were exposed at dose rates differing by a factor of approximately 15. A dose rate effect was observed at both of the irradiation depths used, although this was only clearly significant at 50 mm treatment depth.
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Fortune, Eugene C. IV. "Gamma and neutron dose profiles near a Cf-252 brachytherapy source." Thesis, Georgia Institute of Technology, 2010. http://hdl.handle.net/1853/34781.
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A new generation of medical grade Cf-252 sources was developed in 2002 at the Oak Ridge National Laboratory (ORNL). The combination of small size and large activity of these Cf-252 sources makes them suitable to be used with the conventional high-dose-rate (HDR) remote afterloading systems for interstitial brachytherapy. A recent in-water calibration experiment showed that the measured gamma dose rates near the new source are slightly greater than the neutron dose rates; contradicting the well established neutron-to-gamma dose ratio of approximately 2:1 at locations near a Cf-252 brachytherapy source. Specifically, the MCNP-predicted gamma dose rate is a factor of two higher than the measured gamma dose rate at the distance of 1 cm, and the differences between the two results gradually diminish at distances farther away from the source. To resolve this discrepancy, we updated the source gamma spectrum by including in the ORIGEN-S data library the experimentally measured Cf-252 prompt gamma spectrum as well as the true Cf-252 spontaneous fission yield data to explicitly model delayed gamma emissions from fission products. We also investigated the bremsstrahlung x-rays produced by the beta particles emitted from fission-product decays. The results show that the discrepancy of gamma dose rates is mainly caused by the omission of the bremsstrahlung x-rays in the MCNP runs. By including the bremsstrahlung x-rays, the MCNP results show that the gamma dose rates near a new Cf-252 source agree well with the measured results and that the gamma dose rates are indeed greater than the neutron dose rates.
The calibration experiment also showed discrepancies between the experimental and computational neutron dose profiles obtained. Specifically the MCNP-predicted neutron dose rates were ~25% higher than the measured neutron dose rates at all distances. In attempting to resolve this discrepancy the neutron emission rate was verified by the National Institute of Standards and Technology (NIST) and an experiment was performed to explore the effects of bias voltage on ion chamber charge collection. So far the discrepancies between the computational and experimental neutron dose profiles have not been resolved. Further study is needed to completely resolve this issue and some suggestions on how to move forward are given.
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MUNIZ, RAFAEL O. R. "Desenvolvimento de um simulador antropomorfico para simulacao e medidas de dose e fluxo de neutrons na instalacao para estudos em BNCT." reponame:Repositório Institucional do IPEN, 2010. http://repositorio.ipen.br:8080/xmlui/handle/123456789/9559.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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Ishikawa, Masayori. "Development of New Absorbed Dose Estimation System for Boron Neutron Capture Therapy." Kyoto University, 2002. http://hdl.handle.net/2433/149649.
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Taulbee, Timothy Dale. "Measurement and model prediction of proton-recoil track length distributions in NTA film dosimeters for neutron energy spectroscopy and retrospective dose assessment." Cincinnati, Ohio : University of Cincinnati, 2009. http://www.ohiolink.edu/etd/view.cgi?acc_num=ucin1235764236.
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Thesis (Ph.D.)--University of Cincinnati, 2009.Advisors: Henry Spitz PhD (Committee Chair), Bingjing Su PhD (Committee Member), John Christenson PhD (Committee Member). Title from electronic thesis title page (viewed May 1, 2009). Keywords: NTA; proton-recoil; neutron spectroscopy; dose assessment; track length; Monte Carlo; neutron transport; neutron interactions. Includes abstract. Includes bibliographical references.
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Niemkiewicz, John. "A study on the use of removal-diffusion theory to calculate neutron distributions for dose determination in boron neutron capture therapy /." The Ohio State University, 1996. http://rave.ohiolink.edu/etdc/view?acc_num=osu1487934589976468.
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TARDELLI, TIAGO C. "Avaliação de dados nucleares para dosimetria de nêutrons." reponame:Repositório Institucional do IPEN, 2013. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10587.
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Dissertação (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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Kelm, Robert S. "In-water neutron and gamma dose determination for a new Cf-252 brachytherapy source." Thesis, Atlanta, Ga. : Georgia Institute of Technology, 2009. http://hdl.handle.net/1853/28121.
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Cherniavskiy, I. Y., and V. A. Vinnikov. "The assessment of radiation hazardous areas considering the spectral analysis of the neutron component." Thesis, Національний технічний університет "Харківський політехнічний інститут", 2019. http://repository.kpi.kharkov.ua/handle/KhPI-Press/45079.
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Elazhar, Halima. "Dosimétrie neutron en radiothérapie : étude expérimentale et développement d'un outil personnalisé de calcul de dose Monte Carlo." Thesis, Strasbourg, 2018. http://www.theses.fr/2018STRAE013/document.
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L’optimisation des traitements en radiothérapie vise à améliorer la précision de l’irradiation des cellules cancéreuses pour épargner le plus possible les organes environnants. Or la dose périphérique déposée dans les tissus les plus éloignés de la tumeur n’est actuellement pas calculée par les logiciels de planification de traitement, alors qu’elle peut être responsable de l’induction de cancers secondaires radio-induits. Parmi les différentes composantes, les neutrons produits par processus photo-nucléaires sont les particules secondaires pour lesquelles il y a un manque important de données dosimétriques. Une étude expérimentale et par simulation Monte Carlo de la production des neutrons secondaires en radiothérapie nous a conduit à développer un algorithme qui utilise la précision du calcul Monte Carlo pour l’estimation de la distribution 3D de la dose neutron délivrée au patient. Un tel outil permettra la création de bases de données dosimétriques pouvant être utilisées pour l’amélioration des modèles mathématiques « dose-risque » spécifiques à l’irradiation des organes périphériques à de faibles doses en radiothérapieTreatment optimization in radiotherapy aims at increasing the accuracy of cancer cell irradiation while saving the surrounding healthy organs. However, the peripheral dose deposited in healthy tissues far away from the tumour are currently not calculated by the treatment planning systems even if it can be responsible for radiation induced secondary cancers. Among the different components, neutrons produced through photo-nuclear processes are suffering from an important lack of dosimetric data. An experimental and Monte Carlo simulation study of the secondary neutron production in radiotherapy led us to develop an algorithm using the Monte Carlo calculation precision to estimate the 3D neutron dose delivered to the patient. Such a tool will allow the generation of dosimetric data bases ready to be used for the improvement of “dose-risk” mathematical models specific to the low dose irradiation to peripheral organs occurring in radiotherapy
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Dönsdorf, Esther Miriam [Verfasser]. "Development of a Phoswich Detector for Neutron Dose Rate Measurements in the Earth's Atmosphere / Esther Miriam Dönsdorf." Kiel : Universitätsbibliothek Kiel, 2014. http://d-nb.info/1050977114/34.
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Awotwi-Pratt, Joseph Barton. "Neutron dose equivalent and spectra determination for a medical linear accelerator using dosimetric and Monte Carlo methods." Thesis, University of Surrey, 2003. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.397136.
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Shakshak, Bashir I. O. "The measurements of neutron and gamma dose rates in mixed radiation fields, using a liquid scintillation counter." Thesis, Aston University, 1989. http://publications.aston.ac.uk/8071/.
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Measurements of neutron and gamma dose rates in mixed radiation fields, and gamma dose rates from calibrated gamma sources, were performed using a liquid scintillation counter NE213 with a pulse shape discrimination technique based on the charge comparison method. A computer program was used to analyse the experimental data. The radiation field was obtained from a 241Am-9Be source. There was general agreement between measured and calculated neutron and gamma dose rates in the mixed radiation field, but some disagreement in the measurements of gamma dose rates for gamma sources, due to the dark current of the photomultiplier and the effect of the perturbation of the radiation field by the detector. An optical fibre bundle was used to couple an NE213 scintillator to a photomultiplier, in an attempt to minimise these effects. This produced an improvement in the results for gamma sources. However, the optically coupled detector system could not be used for neutron and gamma dose rate measurements in mixed radiation fields. The pulse shape discrimination system became ineffective as a consequence of the slower time response of the detector system.
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Bonfrate, Anthony. "Développement d'un modèle analytique dédié au calcul des doses secondaires neutroniques aux organes sains des patients en protonthérapie." Thesis, Université Paris-Saclay (ComUE), 2016. http://www.theses.fr/2016SACLS408/document.
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Les doses secondaires neutroniques ne sont actuellement pas estimées lors de la planification de traitement dans les centres de protonthérapie puisque les logiciels de planification de traitement (TPS) ne le proposent pas tandis que les simulations Monte Carlo (MC) et les mesures sont inadaptées pour un environnement clinique. L’objectif de la thèse est de développer un modèle analytique dédié à l’estimation des doses secondaires neutroniques aux organes sains qui reste pratique et simple d’utilisation en routine clinique. Dans un premier temps, la géométrie existante de la gantry installée au Centre de protonthérapie d’Orsay (CPO) de l’institut Curie modélisée avec le code de calcul MCNPX a été étendue à trois configurations de traitement supplémentaires (énergie en entrée de ligne de 162, 192 et 220 MeV). Une approche comparative simulation-mesure a ensuite été entreprise afin de vérifier la capacité de ces modélisations à reproduire les distributions de doses (en profondeur et latérales) des protons primaires ainsi que le champ secondaire neutronique. Des écarts inférieurs à 2 mm ont été observés pour les protons primaires. Pour les neutrons secondaires, les écarts sont plus mitigés avec des rapports simulation sur mesure de ~2 et de ~6, respectivement pour la spectrométrie et les équivalents de dose dans un fantôme physique. L’analyse des résultats a permis d’identifier l’origine de ces écarts et de mettre en perspective la nécessité de conduire de nouvelles études pour améliorer à la fois les mesures expérimentales et les simulations MC. Dans un deuxième temps, une approche purement numérique a été considérée pour calculer les doses neutroniques aux organes sains de fantômes voxélisés représentant des patients d’un an, de dix ans et adulte, traités pour un craniopharyngiome. Une variation de chaque paramètre de traitement a été réalisée afin d’étudier leur influence respective sur les doses neutroniques. Ces paramètres ont pu être ordonnés par ordre décroissant d’influence : incidence de traitement, distance organe-collimateur et organe-champ de traitement, taille/âge des patients, énergie de traitement, largeur de modulation, ouverture du collimateur, etc. Des suggestions ont également été avancées pour réduire les doses neutroniques.Dans un troisième temps, un modèle analytique a été conçu de façon à être utilisable en routine clinique, pour tous les types de tumeur et toutes les installations de protonthérapie. Son entraînement séparé pour trois incidences de traitement a montré des écarts inferieurs à ~30% et ~60 µGy Gy⁻¹ entre les données d’apprentissage (doses neutroniques calculées aux organes sains) et les valeurs prédites par le modèle analytique. La validation a consisté à comparer les doses neutroniques estimées par le modèle analytique à celles calculées avec MCNPX pour des conditions différentes des données d’apprentissage. Globalement, un accord acceptable a été observé avec des écarts moyens de ~30% et ~100 µGy Gy⁻¹. La flexibilité et la fiabilité du modèle analytique ont ainsi été mises en évidence. L’entraînement du modèle analytique à partir d’équivalents de dose neutroniques mesurés dans un fantôme solide au Centre Antoine Lacassagne a confirmé son universalité, bien qu’il requière néanmoins quelques ajustements supplémentaires pour améliorer sa précisionStray neutron doses are currently not evaluated during treatment planning within proton therapy centers since treatment planning systems (TPS) do not allow this feature while Monte Carlo (MC) simulations and measurements are unsuitable for routine practice. The PhD aims at developing an analytical model dedicated to the estimation of stray neutron doses to healthy organs which remains easy-to-use in clinical routine. First, the existing MCNPX model of the gantry installed at the Curie institute - proton therapy center of Orsay (CPO) was extended to three additional treatment configurations (energy at the beam line entrance of 162, 192 and 220 MeV). Then, the comparison of simulations and measurements was carried out to verify the ability of the MC model to reproduce primary proton dose distributions (in depth and lateral) as well as the stray neutron field. Errors within 2 mm were observed for primary protons. For stray neutrons, simulations overestimated measurements by up to a factor of ~2 and ~6 for spectrometry and dose equivalent in a solid phantom, respectively. The result analysis enabled to identify the source of these errors and to put into perspective new studies in order to improve both experimental measurements and MC simulations. Secondly, MC simulations were used to calculate neutron doses to healthy organs of a one-year-old, a ten-year-old and an adult voxelized phantoms, treated for a carniopharyngioma. Treatment parameters were individually varied to study their respective influence on neutron doses. Parameters in decreasing order of influence are: beam incidence, organ-to-collimator and organ-to-treatment field distances, patient’ size/age, treatment energy, modulation width, collimator aperture, etc. Based on these calculations, recommendations were given to reduce neutron doses. Thirdly, an analytical model was developed complying with a use in clinical routine, for all tumor localizations and proton therapy facilities. The model was trained to reproduce calculated neutron doses to healthy organs and showed errors within ~30% and ~60 µGy Gy⁻¹ between learning data and predicted values; this was separately done for each beam incidence. Next, the analytical model was validated against neutron dose calculations not considered during the training step. Overall, satisfactory errors were observed within ~30% and ~100 µGy Gy⁻¹. This highlighted the flexibility and reliability of the analytical model. Finally, the training of the analytical model made using neutron dose equivalent measured in a solid phantom at the center Antoine Lacassagne confirmed its universality while also indicating that additional modifications are required to enhance its accuracy
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Gibson, Christopher R. "Pharmacokinetics, metabolism, and dose optimization simulation studies of Sodium Borocaptate for Boron Neutron capture therapy of Malignant Gliomas /." The Ohio State University, 2001. http://rave.ohiolink.edu/etdc/view?acc_num=osu148639916010674.
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ANGELOCCI, LUCAS V. "Estudo de casos clínicos em radioterapia através do sistema de planejamento AMIGOBrachy." reponame:Repositório Institucional do IPEN, 2016. http://repositorio.ipen.br:8080/xmlui/handle/123456789/26926.
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O sucesso de uma radioterapia depende do correto planejamento da dose a ser entregue ao volume alvo. Na braquiterapia, modalidade da radioterapia onde um radioisótopo selado é implantado intracavitariamente ou intersticialmente no paciente, há menos avanços em sistemas de planejamento de tratamento computacionais do que na teleterapia, amplamente mais utilizada nos serviços típicos. Porém, a braquiterapia, quando aplicável, é preferível por poupar tecidos sadios vizinhos de uma dose desnecessária. O AMIGOBrachy, um sistema de planejamento para braquiterapia de interface amigável, compatibilidade com outros sistemas comerciais em uso e integrado ao código MCNP6 (Monte Carlo N-Particle Transport Code v. 6) foi desenvolvido no Centro de Engenharia Nuclear do Instituto de Pesquisas Energéticas e Nucleares (CEN-IPEN) e atualmente está em processo de validação. Este trabalho contribuiu para este processo, avaliando três diferentes casos clínicos através do AMIGOBrachy com o formalismo do TG43 da AAPM (Associação Americana de Física Médica), protocolo que rege a dosimetria em braquiterapia, e comparando seus resultados com as distribuições de dose calculadas por outros sistemas comerciais consagrados: Varian BrachyVision TM (Varian Medical Systems; Palo Alto, CA, EUA) e Nucletron Oncentra® (Elekta; Estocolmo, Suécia). Os resultados obtidos estão dentro de uma faixa de concordância de ±10%, estando mais discrepantes em regiões muito próximas do aplicador, onde os sistemas de planejamento comerciais e o AMIGOBrachy divergem devido aos diferentes métodos de cálculo. Em pelo menos dois terços da região de interesse, porém, a dose concordou em uma faixa de ±3% para os três casos. Também foram realizadas simulações utilizando o formalismo do TG186 da AAPM, que considera heterogeneidades no tecido, para avaliar o impacto dos mesmos na dose. Em adição ao processo de validação, também foi realizado um estudo em braquiterapia oftálmica para posterior inserção de um módulo adicional ao AMIGOBrachy; para isso, um modelo de olho humano foi desenvolvido utilizando geometria UM (Unstructured Mesh), para validação com o código MCNP6, que apenas nesta versão demonstra um novo recurso capaz de simular uma geometria híbrida: parcialmente analítica, parcialmente UM. O modelo considera dez diferentes estruturas no olho humano: esclera, coroide, retina, corpo vítreo, córnea, câmara anterior, lente, nervo óptico, parede do nervo óptico, e um tumor definido de forma arbitrária crescendo da superfície externa do globo ocular em direção ao seu centro. Os resultados foram comparados com um modelo de olho puramente analítico modelado com o MCNP6 e tomado como referência. Os resultados foram satisfatórios em todas as simulações desenvolvidas, exceto para as estruturas do nervo óptico e sua parede, que devido ao seu pequeno tamanho e distância da fonte, mostraram erros relativos maiores, mas ainda menores que 10%, e não representam problema de preocupação clínica uma vez que recebem doses muito pequenas. Discutiu-se também a eficácia e problemas encontrados nessa nova capacidade do código MCNP de simular geometrias híbridas, uma vez que é recente e ainda apresenta deficiências, que tiveram que ser contornadas no presente trabalho.
Dissertação (Mestrado em Tecnologia Nuclear)
IPEN/D
Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP
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MESSAOUDI, MUSTAPHA. "Elaboration d'un modele de calcul du debit d'equivalent de dose neutron autour des emballages de transport de combustibles irradies." Paris 11, 1993. http://www.theses.fr/1993PA112466.
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Le combustible irradie apres sejour dans le reacteur est emetteur de neutrons et de rayonnement gamma en raison de la formation de noyaux radioactifs classes generalement en deux categories : les produits de fission et les actinides. A des fins de retraitement et (ou) de stockage, le combustible irradie, apres sejour en piscine, est transporte dans des emballages speciaux appeles chateaux de transport, dont les parois sont concues pour attenuer les rayonnements ramenant les debits de dose associes au niveau des valeurs admissibles. L'objet de la these est de faire, d'une part, le bilan des differents problemes poses par les actinides aussi bien du point de vue neutronique et radioactif, afin d'estimer la source de neutron en fonction des parametres d'irradiation et de refroidissement, et d'autre part d'elaborer un modele de calcul de debit de dose neutron associe a chaque assemblage combustible, un formalisme analytique a ete mis au point, ce qui a permis, l'elaboration du systeme de code kahina visant a determiner le debit de dose neutron autour des emballages de transport. L'etude de qualification du systeme kahina a ete realisee avec le code de monte carlo tripoli et des campagnes de mesures faites par ntl. Les ecarts entre les resultats de reference tripoli d'une part, et les mesures experimentales d'autre part et ceux du systeme kahina sur le d. E. D. Neutron au contact, a un metre et a deux metres du cylindre enveloppe de l'emballage montrent que kahina est un outil performant et utile au calcul de routine pour estimer le d. E. D neutron
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CAVALIERI, TASSIO A. "Emprego do NCNP no estudo dos TLDs 600 e 700 visando a implementação da caracterização do feixe de irradiação na instalação de BNCT do IEA-R1." reponame:Repositório Institucional do IPEN, 2013. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10565.
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Dissertação (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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Sadaka, Samir. "Etude theorique et experimentale d'un dosimetre de neutrons rapides." Toulouse 3, 1986. http://www.theses.fr/1986TOU30032.
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Etude theorique de la reponse d'un dosimetre neutronique compose d'un convertisseur fortement hydrogene le(ch::(2))n et d'un detenteur solide visuel de trace le cr39. Etude de la reponse theorique en fonction de l'energie et de l'incidence du flux neutronique. Comparaison avec les resultats experimentaux
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Vasconcellos, Herminiane Luiza de 1987. "Estudo de descritores para distribuição heterogênea de dose." [s.n.], 2015. http://repositorio.unicamp.br/jspui/handle/REPOSIP/276956.
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Orientador: Sandro Guedes de OliveiraDissertação (mestrado) - Universidade Estadual de Campinas, Instituto de Física Gleb Wataghin
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Resumo: Este trabalho baseia-se na analise de descritores de heterogeneidade de dose atraves de programas desenvolvidos em linguagem C++ com base na estatistica de Poisson e probabilidades de ocorrencia de heterogeneidade fundamentadas na teoria de percolacao. A finalidade deste trabalho e obter descritores que possam ser uteis no estudo de efeitos biologicos da radiacao caracteristicos de situacoes em que ha heterogeneidade de dose. Os suportes iniciais deste trabalho se encontram em um relatorio da International Comission on Radiation Units and Measurements, que aborda as questoes de heterogeneidade de dose. Particulas ¿¿ e reacoes envolvendo interacao com neutrons sao as radiacoes que foram focadas na dissertacao e base da aplicacao dos programas desenvolvidos, atraves de resultados obtidos de um experimento em um acelerador linear Elekta Synergy, inter-calibrado com simulacoes de calculo de Monte Carlo. A teoria de percolacao que estuda o comportamento de aglomerados em redes bidimensionais e tridimensionais e baseada em processos randomicos, e pode ser aplicada porque eventos gerados pelas reacoes nucleares ou espalhamentos com neutrons que obedecem a estatistica de Poisson. Os eventos gerados podem ser mapeados a procura de aglomerados, celulas que sao vizinhas nas quais tenham ocorrido eventos. Os aglomerados sao a base da construcao dos descritores. Os resultados encontrados demonstram que os indices de heterogeneidades utilizados fornecem informacoes importantes a respeito da formacao destes aglomerados. Foram comparados os resultados obtidos para os casos 2D e 3D de distribuicao de celulas hipoteticas e foi possivel estudar as relacoes entre os dois casos. Os descritores de heterogeneidade possibilitarao associacoes de dano biologico com a distribuicao de eventos em culturas celulares (caso 2D) e tecidos (caso 3D)
Abstract: The goal of this study is the analysis of dose heterogeneity descriptors through programs developed in C ++ language based on Poisson statistics and probabilities for the occurrence of heterogeneity based on percolation theory. The purpose of this study is to obtain descriptors that may be useful in the study of radiobiological effects characteristic of the situations in which there is dose heterogeneity. The initial support for this work is the report by the International Commission on Radiation Units and Measurements, which describes the dose heterogeneity issues. Álpha particles and reactions involving interaction with neutrons were focused on this thesis are the base of application programs developed from results of an experiment at a linear accelerator Elekta Synergy, inter-calibrated with Monte Carlo simulation. The percolation theory, a theory that studies cluster behavior in two and three-dimensional lattices, is based on random processes, can be applied because the events generated by nuclear reactions with neutrons follow the Poisson statistics. Generated events can be mapped in the search for clusters, neighbor cells in which events occurred. The clusters are the basis for construction of descriptors. The results show that the heterogeneity descriptors provide important information about clusters formation. The results for 2D and 3D cases were compared for distribution of hypothetical cells. and it was possible to study the relations between the two cases. The descriptors of heterogeneity enable biological damage associations with the distribution of events in cell culture (2D case) and tissues (3D case)
Mestrado
Física
Mestra em Física
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Aschan, Carita. "Applicability of thermoluminescent dosimeters in X-ray organ dose determination and in the dosimetry of systemic and boron neutron capture radiotherapy." Helsinki : University of Helsinki, 1999. http://ethesis.helsinki.fi/julkaisut/mat/fysii/vk/aschan/.
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Tardelli, Tiago Cardoso. "Avaliação de dados nucleares para dosimetria de nêutrons." Universidade de São Paulo, 2013. http://www.teses.usp.br/teses/disponiveis/85/85133/tde-20012014-140902/.
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Doses absorvidas e doses efetivas podem ser calculadas utilizando códigos computacionais de transporte de radiação. A qualidade desses cálculos depende dos dados nucleares, no entanto, são raras as informações sobre as diferenças nas doses causadas por diferentes bibliotecas. O objetivo desse estudo é comparar os valores de dose (absorvida e efetiva) obtidos utilizando diferentes bibliotecas de dados nucleares devido a uma fonte externa de nêutrons na faixa de 10-11 a 20 MeV. As bibliotecas de dados nucleares são: JENDL 4.0, JEFF 3.1.1 e ENDF/B-VII.0. Cálculos de doses foram realizados utilizando o código MCNPX considerando o modelo antropomórfico da ICRP-110. As diferenças nos valores das doses absorvidas utilizando as bibliotecas JEFF 3.1.1 e a ENDF/B.VII são pequenas, em torno de 1%, porém os resultados obtidos com a JENDL 4.0 apresentam diferenças de até 85 % compara aos resultados da ENDF/B-VII.0 e JEFF 3.1.1. Diferenças nas doses efetivas são em torno de 1,5% entre ENDF/B-VII.0 e JEFF 3.1.1, e 11 % entre ENDF/B-VII.0 e JENDL 4.0.Absorbed dose and Effective dose are usually calculated using radiation transport computer codes. The quality of the calculations of absorbed dose depends on nuclear data utilized, however, there are rare information about the differences in dose caused by the use of different libraries. The objective of this study is to compare dose values obtained using different nuclear data libraries due to external source of neutrons in the energy range from 10-11 to 20 MeV. The nuclear data libraries used are: JENDL 4.0, JEFF 3.3.1 and ENDF/B.VII. Dose calculations were carried out with the MCNPX code considering the anthropomorphic ICRP 110 model. The differences in the absorbed dose values using JEFF 3.3.1 and ENDF/B.VII libraries are small, around 1%, but the results obtained with JENDL 4.0 presented differences up to 85% compared to ENDF and JEFF results. Differences in effective dose values are around 1.5% between ENDF and JEFF and 11% between ENDF/B.VII and JENDL 4.0.
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Ryckman, Jeffrey M. "Using MCNPX to calculate primary and secondary dose in proton therapy." Thesis, Georgia Institute of Technology, 2011. http://hdl.handle.net/1853/39499.
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Proton therapy is a relatively new treatment modality for cancer, having recently been incorporated into hospitals in the last two decades. Although proton therapy has much higher start up and treatment costs than traditional methods of radiotherapy, it continues to expand in use today. One reason for this is that proton therapy has the advantage of a more precise localization of dose compared to traditional radiotherapy. Other proposed advantages of proton therapy in the treatment of cancer may lead to a faster expanse in its use if proven to be more effective than traditional radiotherapy. Therefore, much research must be done to investigate the possible negative and positive effects of using proton therapy as a treatment modality.
In proton therapy, protons do account for the vast majority of dose. However, when protons travel through matter, secondary particles are created by the interactions of protons and matter en route to and within the patient. It is believed that secondary dose can lead to secondary cancer, especially in pediatric cases. Therefore, the focus of this work is determining both primary and secondary dose.
In order to develop relevant simulations, the specifications of the treatment room and beam were based off of real-world facilities as closely as possible. Using available data from proton accelerators and clinical facilities, an accurate proton therapy nozzle was designed. Dose calculations were performed by MCNPX using a simple water phantom, and then beam characteristics were investigated to ensure the accuracy of the model. After validation of the beam nozzle, primary and secondary dose values were tabulated and discussed. By demonstrating the method of these calculations, the purpose of this work is to serve as a guide into the relatively recent field of Monte Carlo methods in proton therapy.
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Watanabe, Tsubasa. "L-phenylalanine preloading reduces the 10B(n,α)7Li dose to the normal brain by inhibiting the uptake of boronophenylalanine in boron neutron capture therapy for brain tumours." 京都大学 (Kyoto University), 2017. http://hdl.handle.net/2433/225452.
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Hayward, Robert M. "A Coarse Mesh Transport Method with general source treatment for medical physics." Thesis, Atlanta, Ga. : Georgia Institute of Technology, 2009. http://hdl.handle.net/1853/31696.
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Thesis (M. S.)--Nuclear and Radiological Engineering and Medical Physics, Georgia Institute of Technology, 2010.Committee Chair: Rahnema, Farzad; Committee Member: Wang, Chris; Committee Member: Zhang, Dingkang. Part of the SMARTech Electronic Thesis and Dissertation Collection.
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Tinelli, Pascal. "Etude et realisation d'un detecteur microdosimetrique destine a la radioprotection." Toulouse 3, 1986. http://www.theses.fr/1986TOU30154.
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Cette etude a ete consacree a la realisation d'un compteur proportionnel au tissu biologique. Ce detecteur sensible aux neutrons et aux gamma, est destine a mesurer l'equivalent de dose dans le cadre de la radioprotection. L'analyse microdosimetrique des impulsions permet de calculer la dose absorbee et le facteur de qualite, en faisant eventuellement la discrimination entre les deux types de particules
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Petitdidier, Sébastien. "Etude de l'influence de stress électriques et d'irradiations neutroniques sur des HEMTs de la filière GaN." Thesis, Normandie, 2017. http://www.theses.fr/2017NORM2001/document.
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Les transistors HEMTs (High Electron Mobility Transistors) de la filière GaN sont destinés à des applications dans les domaines militaire et spatial. C’est pourquoi nous avons étudié l’influence de trois types de stress électriques : à canal ouvert, à canal pincé et NGB (Negative Gate Bias), ainsi que l’influence de neutrons thermalisés avec une fluence pouvant aller jusqu’à 1,7.1012 neutrons.cm-2, sur leurs performances électriques dc.Dans un premier temps, nous avons étudié des HEMTs AlInN/GaN de laboratoire. Pour les trois stress, nous avons observé une dégradation due à la création de pièges accepteurs et donneurs au cours des différents stress et à la présence de pièges préexistants. Nous avons ensuite irradié ces composants par des neutrons thermalisés et avons observé une légère dégradation des performances électriques des transistors non stressés et stressés à canal ouvert ou pincé. En revanche, nous avons mis en lumière une légère amélioration pour les transistors ayant subi un stress NGB. Nous avons également irradié des MOS-HEMTs AlInN/GaN et conclu que ceux-ci étaient plus sensibles vis à vis des irradiations.Dans un deuxième temps, nous avons stressé de manière analogue des HEMTs AlGaN/GaN du commerce. Dans le cas du stress à canal ouvert, nous avons observé une diminution importante du courant de drain tandis que pour les stress à canal pincé et NGB le courant de drain augmente légèrement à cause d’une libération de pièges préexistants sous l’action du champ électrique vertical. Lors des irradiations avec des neutrons thermalisés, ces transistors, stressés ou non, subissent là encore des dégradationsThe GaN based HEMTs (High Electron Mobility Transistors) are excellent candidates for military and spatial applications. That’s why we have analysed the influence of three different types of bias stress: on-state stress, off-state stress and NGB (Negative Gate Bias), and the influence of thermalized neutrons with a fluence up to 1.7x1012 neutrons.cm-2, on their dc electrical performances.First, we have studied laboratory AlInN/GaN HEMTs. For the three conditions of stress, we have observed a degradation due to pre-existing traps and to the creation of acceptor and donor traps during the stress. Then, we have irradiated these components with thermalized neutrons and we have found a small degradation of the electrical performances of unstressed and on-state stressed and off-state stressed transistors. On the other hand, we have highlighted a slight improvement for NGB stressed components. We have also irradiated AlInN/GaN MOS-HEMTs and we have concluded that they are more sensible to irradiation.In a second time we have stressed in the same way commercial AlGaN/GaN HEMTs. For the on-state stress, we have observed an important increase in the drain current. However, the drain current increases for the on-state and NGB stressed components due to a release of electrons from pre-existing traps under vertical electrical field. During the irradiation with thermalized neutrons, the unstressed and stressed transistors are degraded and a small decrease in the drain current is visible
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Flaspoehler, Timothy Michael. "FW-CADIS variance reduction in MAVRIC shielding analysis of the VHTR." Thesis, Georgia Institute of Technology, 2012. http://hdl.handle.net/1853/45743.
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In the following work, the MAVRIC sequence of the Scale6.1 code package was tested for its efficacy in calculating a wide range of shielding parameters with respect to HTGRs. One of the NGNP designs that has gained large support internationally is the VHTR. The development of the Scale6.1 code package at ORNL has been primarily directed towards supporting the current United States' reactor fleet of LWR technology. Since plans have been made to build a prototype VHTR, it is important to verify that the MAVRIC sequence can adequately meet the simulation needs of a different reactor technology. This was accomplished by creating a detailed model of the VHTR power plant; identifying important, relevant radiation indicators; and implementing methods using MAVRIC to simulate those indicators in the VHTR model.
The graphite moderator used in the design shapes a different flux spectrum than water-moderated reactors. The different flux spectrum could lead to new considerations when quantifying shielding characteristics and possibly a different gamma-ray spectrum escaping the core and surrounding components. One key portion of this study was obtaining personnel dose rates in accessible areas within the power plant from both neutron and gamma sources. Additionally, building from professional and regulatory standards a surveillance capsule monitoring program was designed to mimic those used in the nuclear industry. The high temperatures were designed to supply heat for industrial purposes and not just for power production. Since tritium, a heavier radioactive isotope of hydrogen, is produced in the reactor it is important to know the distribution of tritium production and the subsequent diffusion from the core to secondary systems to prevent contamination outside of the nuclear island.
Accurately modeling indicators using MAVRIC is the main goal. However, it is almost equally as important for simulations to be carried out in a timely manner. MAVRIC uses the discrete ordinates method to solve the fixed-source transport equation for both neutron and gamma rays on a crude geometric representation of the detailed model. This deterministic forward solution is used to solve an adjoint equation with the adjoint source specified by the user. The adjoint solution is then used to create an importance map that can weight particles in a stochastic Monte Carlo simulation. The goal of using this hybrid methodology is to provide complete accuracy with high precision while decreasing overall simulation times by orders of magnitude. The MAVRIC sequence provides a platform to quickly alter inputs so that vastly different shielding studies can be simulated using one model with minimal effort by the user. Each separate shielding study required unique strategies while looking at different regions in the VHTR plant. MAVRIC proved to be effective for each case.
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Saueia, Cátia Heloisa Rosignoli. "Distribuição elementar e de radionuclídeos na produção e uso de fertilizantes fosfatados no Brasil." Universidade de São Paulo, 2006. http://www.teses.usp.br/teses/disponiveis/85/85131/tde-22032012-173112/.
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O fertilizante é considerado um componente essencial para a agricultura, pois sua utilização aumenta e repõe os nutrientes naturais do solo, perdidos por desgaste ou erosão. No processo de obtenção dos fertilizantes fosfatados, o concentrado de rocha reage com ácido sulfúrico concentrado produzindo ácido fosfórico e sulfato de cálcio (fosfogesso), como subproduto. O ácido fosfórico é utilizado para a produção do superfosfato triplo (TSP), superfosfato simples (SSP), monoamônio fosfato (MAP) e diamônio fosfato (DAP). A rocha fosfatada usada como matéria prima apresenta em sua composição radionuclídeos das séries naturais do urânio e tório. Durante o ataque químico do concentrado de rocha, as espécies presentes na reação, estáveis e radioativas, são redistribuídas entre o ácido fosfórico (matéria prima dos fertilizantes), e o fosfogesso, de acordo com sua solubilidade e características químicas. Enquanto os fertilizantes são comercializados, o fosfogesso fica estocado em pilhas podendo impactar o meio ambiente. Com a finalidade de entender a distribuição dos elementos e dos radionuclídeos no processo industrial de produção de fertilizantes fosfatados, foram analisadas amostras de concentrado de rocha, de fertilizantes (SSP, TSP, MAP e DAP) e fosfogesso de três procedências nacionais denominadas indústrias A, B e C. A técnica utilizada para a análise elementar foi a análise por ativação com nêutrons, que permitiu analisar os elementos Ba, Co, Cr, Fe, Hf, Na, Sc, Ta, Th, U, Zn e Zr, e as terras raras, La, Ce, Nd, Sm, Eu, Tb, Yb e Lu. Os resultados obtidos permitiram concluir que em geral, as terras raras se distribuem de forma homogênea em todos os fertilizantes e no fosfogesso, exceto o Lu. Os fertilizantes SSP e TSP apresentaram concentrações de todos os elementos analisados da mesma ordem de grandeza da rocha de origem. O mesmo comportamento foi observado nos fertilizantes MAP e DAP, exceto para os elementos Co, Sc e U. Os elementos pertencentes à série radioativa natural do urânio (238U, 234U, 230Th, 226Ra e 210Pb), do tório (232Th, 228Ra e 228Th) e o K-40, foram determinados por meio da espectrometria gama e alfa. As amostras de fertilizantes MAP e DAP, que são diretamente derivadas do ácido fosfórico, apresentaram baixa concentração para o 226Ra, 228Ra e 210Pb, enquanto que para o U e Th as concentrações encontradas foram da mesma ordem de grandeza da rocha de origem. Os fertilizantes SSP e TSP, que são obtidos pela mistura de ácido fosfórico com concentrado de rocha, apresentaram concentrações mais elevadas para os radionuclídeos das séries naturais. Avaliou-se a exposição devido a sucessivas aplicações de fertilizantes e fosfogesso, calculando-se a dose interna devida à aplicação por 10, 50 e 100 anos. Os valores encontrados estão abaixo do limite de 2,4 mSv a-1, mostrando que esta prática é negligenciável.Fertilizer is considered an essential component for agriculture, because its use increases the natural soil nutrients, which are lost slow waste or erosion. The Brazilian phosphate fertilizer is obtained by wet reaction of igneous phosphate rock with concentrated sulphuric acid, giving as final product, phosphoric acid and dihydrated calcium sulphate (phosphogypsum) as by-product. Phosphoric acid is the starting material for triple superphosphate (TSP), single superphosphate (SSP), monoammonium phosphate (MAP) and diammonium phosphate (DAP). The phosphate rock used as raw material presents in its composition, radionuclides of the U and Th natural series in. During the chemical attack of the phosphate rock, this equilibrium is disrupted and the radionuclides and the elements migrate to intermediate, final products and by-products, according to their solubility and chemical properties. While the fertilizers are commercialized, the phosphogypsum is disposed in stack piles and can cause an impact in the environment. In order to evaluate the radionuclides and the elements distribution in the industrial process of phosphate fertilizer production, samples of concentrated rock, fertilizers (SSP, TSP, MAP and DAP) and phosphogypsum from three national industries (A, B and C), were analyzed. The characterization of the elements Ba, Co, Cr, Fe, Hf, Na, Sc, Ta, Th, U, Zn and Zr, and the rare earths La, Ce, Nd, Sm, Eu, Tb, Yb and Lu, were performed by instrumental neutron activation analysis. The results obtained showed that, in general, the rare earth elements are distributed uniformly in the fertilizers and phosphogypsum, except for Lu. The elemental concentration present in the fertilizers SSP and TSP are of the same order of magnitude of the source rock. The same behavior was observed in the fertilizers MAP and DAP, except for the elements Co, Sc and U. The radionuclides of the U series (238U, 234U, 230Th, 226Ra, 210Pb) and of the Th series (232Th, 228Ra, 228Th) and 40K were determined by gamma and alpha spectrometry. The fertilizers samples, with are derived directly from phosphoric acid, MAP and DAP, presented in their composition low activity concentrations for 226Ra, 228Ra and 210Pb. For U and Th, the concentrations founded in MAP and DAP are more significant, similar to the source rock. SSP and TSP fertilizers, which are obtained by mixing phosphoric acid with different amounts of phosphate rock, presented higher concentrations of all radionuclides of the natural series. Long-term exposure due to successive fertilizer and phosphogypsum application was evaluated. Internal doses due to the application of phosphate fertilizer and phosphogypsum for 10, 50 and 100 years were below 2.4 mSv y-1, showing that the radiological impact of such practice is negligible.
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Strahinja, Ilić. "Анализа функција ефикасних пресека за неутронске реакције на 185Re и 187Re и анализа специфичне константе гама дозе зa 252Cf." Phd thesis, Univerzitet u Novom Sadu, Prirodno-matematički fakultet u Novom Sadu, 2020. https://www.cris.uns.ac.rs/record.jsf?recordId=114888&source=NDLTD&language=en.
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Користећи NAXSUN технику развијену уЈРЦ-Геел,мерени су ефикасни пресеци зареакције изазване неутронима187Re(n, p) 187W и 185Ре (n, 3n) 183Rе мерене у енергетском распону између 13,08 MeV и 19,5 МеV. Ови подаци су прве експериментално добијене вредности за нуклеарне реакције у овом енергетском опсегу неутрона. Добијени резултати упоређени су са постојећим процењеним прорачунимаТАЛИС 1.9 и ЕМПИРЕ 3.2.3 користећи различите доступне моделе. Упоређени су теоријски прорачуни са експерименталним резултатима. У раду је, на основу три снимљена гама спектра калифорнијумовог извора, закључено о утицају акумулације фисионих продуката на укупну специфичну гама константу извора.Koristeći NAXSUN tehniku razvijenu uJRC-Geel,mereni su efikasni preseci zareakcije izazvane neutronima187Re(n, p) 187W i 185Re (n, 3n) 183Re merene u energetskom rasponu između 13,08 MeV i 19,5 MeV. Ovi podaci su prve eksperimentalno dobijene vrednosti za nuklearne reakcije u ovom energetskom opsegu neutrona. Dobijeni rezultati upoređeni su sa postojećim procenjenim proračunimaTALIS 1.9 i EMPIRE 3.2.3 koristeći različite dostupne modele. Upoređeni su teorijski proračuni sa eksperimentalnim rezultatima. U radu je, na osnovu tri snimljena gama spektra kalifornijumovog izvora, zaključeno o uticaju akumulacije fisionih produkata na ukupnu specifičnu gama konstantu izvora.
Using the NAXSUN technique developed at the JRC-Geel, the cross section functions for the neutron induced reactions 187Re(n,p)187W and 185Re(n,3n)183Re have been measured in the energy range between 13.08 MeV and 19.5 MeV. These data are the first experimentally obtained values for those nuclear reactions in this neutron energy range. Obtained results have been compared withexisting evaluated The TALYS 1.9 and EMPIRE 3.2.3 calculations were performed using different available. A comparison between theoretical model calculations and experimental results was made. Based on three recorded gamma ray spectra of a Californiumsource, conclusion is made if there are influences of fission product accumulation on the total specific gamma ray constant of the source.
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MANGUEIRA, THYAGO F. "Avaliacao dosimetrica da solucao fricke gel usando a tecnica de espectrofotometria para aplicacao na dosimetria de eletrons e neutrons." reponame:Repositório Institucional do IPEN, 2009. http://repositorio.ipen.br:8080/xmlui/handle/123456789/9447.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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MURA, LUIZ E. C. "Caracterização dos campos neutrônicos obtidos por meio de armadilhas de nêutrons no interior do núcleo do reator nuclear IPEN/MB-01." reponame:Repositório Institucional do IPEN, 2011. http://repositorio.ipen.br:8080/xmlui/handle/123456789/9999.
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Dissertação (Mestrado)
IPEN/D
Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP
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Nielsen, Adam Derek. "Monte Carlo calculation of fluence-to-ambient dose equivalent conversion coefficients for high-energy neutrons." Thesis, Georgia Institute of Technology, 1998. http://hdl.handle.net/1853/16424.
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SANCHEZ, ANDREA. "Determinacao dos parametros intermediarios de ressonancia no formalismo de multigrupo de energia." reponame:Repositório Institucional do IPEN, 1996. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10481.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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BORGES, ANTONIO A. "Combinacao entre os metodos diferencial e da teoria de pertubacao para calculo dos coeficientes de sensibilidade." reponame:Repositório Institucional do IPEN, 1998. http://repositorio.ipen.br:8080/xmlui/handle/123456789/9267.
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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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BORGES, ANTONIO A. "Uma combinação entre os métodos diferencial e da teoria de perturbação para o cálculo dos coeficientes de sensibilidade." reponame:Repositório Institucional do IPEN, 1998. http://repositorio.ipen.br:8080/xmlui/handle/123456789/9267.
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Desenvolve-se aqui um novo método para calcular coeficientes de sensibilidade. Este novo método é uma combinação entre as duas metodologias usadas para calcular estes coeficientes, que são o método diferencial e o método da teoria da perturbação generalizada. O método consiste em fazer como parâmetro integral o fluxo médio em uma região arbitrária do sistema. Dessa forma, o coeficiente de sensibilidade passa a conter somente o termo correspondente ao fluxo de nêutrons. Para obtenção do novo coeficiente de sensibilidade é feito o cálculo do coeficiente de sensibilidade desse parâmetro integral com relação a σ através do método de perturbação e são obtidas as derivadas funcionais do parâmetro integral genérico com relação a σ e Φ utilizando o método diferencial.
Dissertação (Mestrado em Tecnologia Nuclear)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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Espinosa, Carlos Eduardo. "Modelagem e simulação dos venenos no combustível nuclear em cenários com duas escalas de tempo." reponame:Biblioteca Digital de Teses e Dissertações da UFRGS, 2016. http://hdl.handle.net/10183/150627.
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A presente discussão e uma extensão do modelo de cinética pontual de nêutrons, onde a reatividade e decomposta em termos de contribuição de curtas e longas escalas de tempo. A primeira representa o controle operacional do reator, enquanto a segunda e devido a alteração da composição química do combustível nuclear, como consequência do burn-up. E um primeiro passo em uma nova direção, uma vez que considera os efeitos dos principais venenos na cinética de nêutrons, ou seja, Xenônio-135 e Sam ario-149. O modelo proposto consiste em um sistema de equações não-lineares acoplado para a densidade de nêutrons, para os precursores de nêutrons atrasados e para as cadeias de decaimento dos venenos produtos de fissão. O sistema de equações e resolvido através de um método de decomposição, que expande os termos não-lineares em uma série infinita, obtendo um sistema recursivo, onde a inicialização da recursão e uma equação linear homogênea e os passos de recursão subsequentes consideram contribuições não-lineares como termo fonte constru dos em passos de recursão anteriores. A construção hierárquica do modelo também e realizada, onde graus espaciais de liberdade são considerados. São apresentados casos de estudos com várias estruturas temporais afim de mostrar a robustez da abordagem atual para este tipo de problema.The present discussion is an extension to Neutron point kinetics models, where the reactivity is decomposed in a short and a long term contribution. The rst one represents operational reactor control, whereas the second one is due to the change of the chemical composition of the nuclear fuel as a consequence of burn-up. This is a rst step into a new direction where we consider only the e ects of the principal neutron poisons on neutron kinetics, i.e, Xenon-135 and Samarium-149. The proposed model consists in a system of coupled nonlinear equations for the neutron density, the delayed neutron precursors and the neutron poison decay chains. The equation system is solved using a decomposition method, which expands the non-linear terms in an in nit series, obtaining a recursive system, where the recursion initialization is a homogeneous linear equation and the subsequent recursion steps consider the non-linear contributions as source terms constructed from previous recursion steps. A hierarchical construction of the model is also performed, where spatial degrees of freedom are considered. We present case studies with severe time structure in order to show the robustness of the present approach for this kind of problems.
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Fauth, Anderson Campos 1957. "Observações sobre os momentos transversais dos pions neutros e dos raios gamas do estado intermediário de massa ~3 GeV/c2 (mirim) na produção múltipla de mésons." [s.n.], 1986. http://repositorio.unicamp.br/jspui/handle/REPOSIP/278131.
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Orientador: Kotaro SawayanagiDissertação (mestrado) - Universidade Estadual de Campinas, Instituto de Fisica Gleb Wataghin
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Resumo:É obtida a distribuição de momento transversal dos pions neutros de 32 C-jatos Mirins com SEg > 20 TeV através de dois métodos de acoplamento 2g ® pº. Estes resultados independem de qualquer modelo de produção de partículas. Realiza-se uma simulação, pelo método de Monte Carlo, da produção de pions e avalia-se que os dois métodos conseguem obter em média aproximadamente 50% de acoplamentos corretos. Constata-se que a forma da distribuição dos pions neutros depende fracamente da porcentagem de acoplamentos corretos. A distribuição de momento transversal dos pions neutros dos eventos Mirins é obtida por uma terceira maneira, a qual é completamente independente das duas primeiras, que consiste da composição entre uma solução analítica e o método de Monte Carlo. Os resultados dos três métodos são consistentes entre si
Abstract: Not informed.
Mestrado
Física
Mestre em Física
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Evrard, Nicholas. "When does it pay to be carbon neutral?" Thesis, Stellenbosch : Stellenbosch University, 2012. http://hdl.handle.net/10019.1/80783.
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Thesis (MBA)--Stellenbosch University, 2012.Companies produce carbon and GHG emissions in the course of doing business. Climate change issues and the impact of global warming affect business conditions. Companies need to deal with these issues and to introduce procedures for their mitigation. They can also aim to formulate strategies to enable the company to achieve a sustainable future. This study was designed to evaluate the motivation for South African businesses to voluntarily invest in becoming carbon neutral and to assess the payoff when adopting such strategies. This study has defined the concept of carbon neutrality, the opportunities of pursuing such a strategy and the risks of not doing so for the purpose of understanding the motivational drivers. An adapted framework was developed to assess whether or not such strategies are attractive. The empirical study examined four companies in terms of motivation. The exploratory case studies were compared to the descriptions and the frameworks discussed in the literature review. The study should serve to inform other companies of the possible opportunities and risks of lowcarbon initiatives. Exploring the methods leading to carbon neutrality should also serve as a tool for companies willing to participate in such projects.
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Bezerra, Junior Arandi Ginane. "Teoria cinetica para misturas de gases neutros e ionizados - um metodo alternativo." reponame:Repositório Institucional da UFPR, 1993. http://hdl.handle.net/1884/37014.
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Orientador: Gilberto Medeiros KremerDissertação (mestrado) - Universidade Federal do Parana, Setor de Ciencias Exatas
Resumo: E desenvolvida neste trabalho uma teoria cinética para uma mistura de gases monoatômicos baseada num método alternativo que combina características dos métodos de Chapman-Enskog e Grad, simplificando-os. A partir dele são obtidas as equações constitutivas (leis de Fick, Navier-Stokes e Fourier) para uma teoria linearizada juntamente com aproximações sucessivas para os coeficientes de transporte. O método e aplicado primeiramente para uma mistura de gases neutros e em seguida para uma mistura de gases ionizados. Em ambos os casos são verificadas as relações de reciprocidade de Onsager.
Abstract: Based on an alternative method that combines and simplifies the features of the Chapman-Enskog and the Grad methods, a kinetic theory for monatomic gas mixtures is developed. The constitutive equations (laws of Fick, Navier-Stokes and Fourier) and the sucessive approximations to the transport coefficients are obteined for a linearized theory. Neutral gas mixtures and ionized gas mixtures are analyzed within the framework of this method. In both cases the Onsager reciprocal relations are verified.
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Isolan, Lorenzo. "Verifica radioprotezionistica con tecniche Monte Carlo del bunker per radioterapia dell' ASMN-IRCCS di Reggio Emilia." Master's thesis, Alma Mater Studiorum - Università di Bologna, 2015. http://amslaurea.unibo.it/8589/.
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Conoscere con dettaglio il campo di radiazione che si genera nell'utilizzo di un acceleratore lineare di elettroni durante una seduta di radioterapia è essenziale sia per i pazienti sia per gli operatori.
L'utilizzo del codice Monte Carlo MCNPX 2.7.0 permette di stimare dati dosimetrici dettagliati in zone dove può essere complicato effettuare misurazioni.Lo scopo di questo lavoro è indagare il comportamento del fascio fotonico prodotto nel bunker di radioterapia dell'ASMN-IRCCS di Reggio Emilia, valutando con precisione in particolare
la produzione di fotoneutroni secondari. L'obiettivo è la verifica dell'efficacia delle barriere offerte dalla struttura tenendo in considerazione anche il canale
di penetrazione degli impianti di servizio che costituisce un punto di fuga per le radiazioni.
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KURAMOTO, RENATO Y. R. "Desenvolvimento de uma metodologia baseada no modelo de duas-regiões e em técnicas de análise de ruído microscópico para a medida absoluta dos parâmetros cinéticos Betasub(eff), Lambda e Betasub(eff/Lambda do reator IPEN/MB-01." reponame:Repositório Institucional do IPEN, 2007. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11547.
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Fundação de Amparo à Pesquisa do Estado de São Paulo (FAPESP)
Tese (Doutoramento)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
FAPESP:03/01261-0
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MELO, ADEILSON P. de. "Caracterização do jade e dos silicatos da familia do jade para aplicação em dosimetria das radiações." reponame:Repositório Institucional do IPEN, 2007. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11550.
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Tese (Doutoramento)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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PEREIRA, MARCO A. S. "Emprego dos policarbonatos makrofol-de e CR-30 em radiografia com neutrons." reponame:Repositório Institucional do IPEN, 2000. http://repositorio.ipen.br:8080/xmlui/handle/123456789/9293.
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Dissertacao (Mestrado)
IPEN/D
Intituto de Pesquisas Energeticas e Nucleares, IPEN/CNEN-SP
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Mangueira, Thyago Fressatti. "Avaliação dosimétrica da solução fricke gel usando a técnica de espectrofotometria para aplicação na dosimetria de elétrons e nêutrons." Universidade de São Paulo, 2009. http://www.teses.usp.br/teses/disponiveis/85/85131/tde-29032012-144938/.
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Neste trabalho as principais características dosímetricas da solução Fricke Xilenol Gel (FXG) foram estabelecidas para futura aplicação clínica na dosimetria de elétrons. As curvas de dose resposta para feixes de nêutrons térmicos para pesquisa em Terapia por Captura de Nêutrons (BNCT) e feixes elétrons de aplicação industrial também foram determinadas. A técnica padrão de leitura utilizada foi espectrofotometria. Para o feixe clínico as reprodutibilidades intra e inter-lotes da solução FXG são melhores que 1,4 % e 5,1 % respectivamente, o comportamento da resposta para o intervalo de dose entre 0,2 e 40 Gy é linear e independente da energia e da taxa de dose para o intervalo estudado. Devido aos efeitos da oxidação natural do FXG o tempo ótimo entre o preparo e a irradiação é de 24h e o comportamento da curva de dose resposta não se altera no período estudado para a variação da absorvância líquida do dosímetro. Para o estudo com o campo de nêutrons as curvas de dose resposta do FXG apresentaram comportamento linear em todo intervalo de dose estudado, e para campos industriais de elétrons o comportamento é exponencial decrescente. De acordo com os resultados obtidos para os feixes de radiação estudados, não houve alteração na posição das bandas características do espectro de absorção do FXG. Como testes adicionais, foi determinada a viabilidade do uso do método de leitura do FXG por imagens fotográficas digitais e aplicação do FXG na dosimetria para braquiterapia intracavitária. O bom desempenho do dosímetro FXG nos testes realizados indica que este pode ser utilizado na avaliação tridimensional da dose em tratamento radioterápicos.In this work the main dosimetric characteristics of the Fricke Xylenol Gel (FXG) solution were established for further application in the measurement of dose distribution of clinical electron fields. The dose-response curves of the FXG in a neutron field were also evaluated for the research in Boron Neutron Capture Therapy (BNCT) and industrial electron fields. The standard reading technique was the spectrophotometric. For the clinical field, the intra and inter-batch reproducibility are better than 1.4% and 5.1 %, respectively, the response presents a linear behavior for doses ranging from 0.2 to 40 Gy independently of the energy and the dose rate in the studied ranges. Due to the effects of the FXG natural oxidation, the optimum elapsed time between FXG preparation and irradiation was established as 24h period and the behavior of the dose-response curve of the FXG using the variation in the absorbance relative to the non-irradiated dosimeter as a basis during the whole studied period were not altered. The dose-response to the industrial electron beam presented an exponential decreasing behavior and the neutron beam for research in BNCT presented a linear behavior for the complete studied dose range. According to the obtained results for the different types of radiation studied for the FXG, there was no change in the position of the characteristic bands of the absorption spectrum due to the interaction of these radiation types. Additional tests were performed to determine the digital photographic imaging of FXG analyses viability and the application of FXG dosimetry on intracavitary brachytherapy. The good performance of the FXG dosimeter in the tests that were carried out indicates that this dosimeter may be applied to the tri-dimensional dose evaluation in radiotherapic treatments using electrons and neutron beams.
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SANTOS, DIOGO F. dos. "Caracterização dos campos neutrônicos obtidos por meio de armadilhas de nêutrons a partir da utilização de água pesada (D2O) no interior do núcleo do reator nuclear IPEN/MB-01." reponame:Repositório Institucional do IPEN, 2015. http://repositorio.ipen.br:8080/xmlui/handle/123456789/23825.
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Dissertação (Mestrado em Tecnologia Nuclear)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP