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Study of salinity in aqueous medium using x-ray beam with mcnp-x code
Get PDFIn offshore production, it is possible that the produced water presents geochemical characteristics that correspond to the mixture of formation water (connate water) and the sea water (injection water), and the
physical-chemical behavior of the injected water allows a considerable variation in the index of salinity altering the water/oil ratio during transportation and/or extraction. Injection water is generally used to raise the reservoir pressure, increasing the percentage of extracted oil. This water has a significant amount of salts that generate some difficulties, such as measuring fractions of volume in multiphase systems. One way to check the effects of salinity would be to regularly measure the amount of salt present in the water. In this way, this work presents a methodology to measure the concentration and the types of salts using nuclear techniques through the MCNP-X computational code. The measurement geometry uses an X-ray beam (40-100 keV) and NaI(Tl) scintillation detector positioned diametrically opposed to the source. The studied samples were the NaCl, KCl and MgCl2 salts in aqueous solution. The results present the possibility of differentiating the formation and injection waters due to differences in the salt concentrations
Analytical methods for analysis of neutron cross-sections of amino acids and proteins
Get PDFTwo unpublished analytical processes were developed at IEN-CNEN-RJ for the analysis of neutron cross-
sections of chemical compounds and complex molecules, the method of data parceling and grouping (P&G) and the method of data equivalence and similarity (E&S) of cross-sections. The former allows the division of a
complex compound or molecule so that the parts can be manipulated to construct a value of neutron cross-
section for the compound or the entire molecule. The second method allows by comparison obtain values of
neutron cross-sections of specific parts of the compound or molecule, as the amino acid radicals or its parts. The processes were tested for the determination of neutron cross-sections of the 20 human amino acids and a small database was built for future use in the construction of neutron cross-sections of proteins and other components of the human being cells, also in other industrial applications
Development of a mobile game based on virtual reality tools to sensitize the population about the nuclear power plant’s emergency plan
Get PDFThe Angra dos Reis Nuclear Power Plant’s emergency plan, it´s the bunch of instructions that every citizen must adopt in case of an emergency situation. It’s highly important, that all the people living in the power plant´s surroundings truly understand every single step of the plan, because only in this way people will know how to react in case of a necessity. To hit this goals, the Brazil`s Electronuclear, made educational booklets, in the shape of comic books, trying to guide the population about the plan. On the other hand, we have an increasingly connected world, making possible that digital games, be very well accepted by the population. So this project has as an objective, developing a digital tool, in form of a mobile game that shows in a playful and interactive way for the user, the emergency plan, complementing the educational process and social actions made by many institutions. With the information taken from the booklets, objects and buildings were modeled in Autodesk 3ds Max, allied with the Unity 3D Game Engine, to make a city, inspired in Angra do Reis (RJ). The player has to follow all the security protocols giving by the Eletronuclear according with the rules provided by the National Nuclear Energy Commission. Is expected, with this game that will be available for the Eletronuclear that more people have the chance to know and believe in the efficiency of the emergency plan already established
Análise do levantamento radiométrico durante operação do Reator Argonauta
Get PDFO reator Argonauta do Instituto de Engenharia Nuclear - IEN/CNEN, opera, normalmente, a potências entre 1,7 e 340 W, em procedimentos de neutrongrafia, produção de radionuclídeos e aulas experimentais de física de reator para cursos de pós-graduação. As doses provenientes nêutrons e radiação gama são medidas quando o reator encontra-se crítico, dentro do salão do reator e nas regiões circunvizinhas. Um estudo dos dados obtidos foi realizado para avaliar a necessidade diária desse levantamento no salão do reator. Tendo em vista o princípio ALARA, que visa a otimizar e minizar a dose recebida pelo indivíduo, propomos, neste trabalho, mediante uma análise dos dados adquiridos em levantamentos radiométricos ocupacionais, uma reformulação da rotina de monitoração de área praticada pela equipe de proteção radiológica do Instituto de Engenharia Nucelar, visto que outras rotinas de acompanhamento quanto à proteção radiológica também são aplicadas na rotina do reator. As operações em análise ocorreram com o reator operando a potência de 340 W, em intervalos de 60, 120 e 180 minutos, nos pontos de monitoração em áreas controlada, supervisionada e livre. Os resultados mostraram valores de dose significativos na saída do canal J-9 quando a operação ocorre com este aberto. Com 180 min de funcionamento, os valores medidos de taxa de dose foram menores do que os valores em operações de 60 min e 120 min. No ponto na área supervisionada, em local externo ao salão do reator, situado na direção do canal J-9, o valor reduz mais que de 14% em qualquer tempo de operação em relação à taxa de dose medida no ponto em frente ao canal. Há uma redução 50% nas taxas de dose para as operações com e J-9 fechado. Os resultados sugerem uma nova periodicidade do levantamento radiométrico cujo modo de operação se mantenha em condições semelhantes, desde que associada a outras práticas relevantes de radioproteção
Metodologia modificada para aplicação em eventos não planejados de desarmes automáticos e manuais em usinas nucleares
No full textO presente trabalho apresenta as metodologias existentes para redução de desarmes de reatores não planejados. Adicionalmente, propõe uma metodologia modificada para redução de desarmes automáticos e manuais não planejados de reatores de usinas nucleares. Esta metodologia modificada tem como principal diferença em relação as metodologias existentes o universo amostral utilizado para as análises das causas de cada desarme de reator nuclear ocorrido. Enquanto as metodologias existentes analisam apenas a causa raiz de cada desarme e se aplica quando se está analisando um grande universo de usinas nucleares, a metodologia proposta analisa a causa raiz e todos os fatores causais. No desenvolvimento desta metodologia para análise dos desarmes de reator ocorridos na usina estudada foi verificado que o universo amostral considerando apenas a causa raiz de cada desarme seria muito pequeno quando comparado com a metodologia utilizada pela indústria nuclear. Conforme está descrito nesta dissertação, o INPo analisou a causa raiz de todos os desarmes de reatores que ocorreram nos Estados Unidos nos anos de 2009 e 2010. Nesta análise mais de 140 desarmes foram analisados . A aplicação da metodologia INPO para a usina estudada se torna imprecisa pois o universo amostral do número de desarmes analisados fica muito reduzido. Por isso este trabalho propõe uma metodologia modificada para o estudo de caso da usina nuclear estudada. Nesta metodologia modificada foram considerados a causa raiz e os fatores contribuintes que de alguma forma possam ter contribuído para a ocorrência de desarmes do reator, desta forma aumentando significativamente o universo amostral de falhas apuradas. O trabalho também apresenta os principais processos existentes que ajudam a evitar desarmes do reator, e faz uma relação entre um tipo de falha identificada e os processos que devem ser melhorados para se evitar esta falha. Por fim, faz um estudo de caso de uma usina nuclear empregando esta nova metodologia. Neste estudo de caso são identificadas as principais falhas que provocaram desarmes do reator nuclear e os principais processos a serem melhorados, e apresenta os bons resultados utilizando-se esta nova metodologia propost
Cálculo de célula com simulação de queima usando o código WIMSD-5B considerando diferentes bibliotecas de dados nucleares
Get PDFThis work proposes to implement the cell calculation considering the fuel burning using the WIMSD-5B code. The cell calculation procedure allows determining the nuclear parameters present in the multi-group neutron diffusion equation and for this purpose the neutron transport theory is used in a problem with dimensional reduction, but in contrast is considered a large number of groups associated with the neutron spectrum. There are a variety of reactor physics codes that determine the nuclear parameters by solving the neutron transport equation applied to an equivalent cell representing a fuel element. The WIMSD-5B code is a deterministic code that solves the transport equation using collision probability method. The simulation of fuel burning in the cell calculation took into account different nuclear data libraries. The WIMSD-5B code supports several nuclear data libraries and in the present work the following libraries were used: IAEA, ENDFB-VII.1, JENDL3.2, JEFF3.1 and JEF2.2, all formatted for 69 energy groups.Este trabalho propõe programar o cálculo de célula considerando a queima de combustível utilizando o código WIMSD-5B. O procedimento de cálculo de célula permite determinar os parâmetros nucleares presentes na equação de difusão de nêutrons multi-grupo e para este fim a teoria do transporte de nêutrons é utilizada em um problema com redução dimensional, mas, em contrapartida é considerado um grande número de grupos de energia associados ao espectro de nêutrons. Há uma variedade de códigos de física do reator que determinam os parâmetros nucleares resolvendo a equação de transporte de nêutrons aplicada em uma célula equivalente representando um elemento combustível. O código WIMSD-5B é um código determinístico e que resolve a equação de transporte utilizando método de probabilidade de colisão. A simulação de queima de combustível no cálculo de célula levou em conta diferentes bibliotecas de dados nucleares. O código WIMSD-5B suporta várias bibliotecas de dados nucleares e no presente trabalho foram utilizadas as seguintes bibliotecas: IAEA, ENDFB-VII.1, JENDL3.2, JEFF3.1 e JEF2.2, todas formatadas para 69 grupos de energia
Modelo para mapeamento das percepções dos usuários a respeito dos serviços prestados por uma biblioteca especializada
Get PDFNowadays a question arises: how to evaluate the users’ satisfaction regarding the services provided by a library? This paper aims to propose and apply a model for evaluating the satisfaction of the users of a library, regarding the library services. The proposal should be divided into three steps: a systematized literature review that has supported the construction of a questionnaire; a data collection though applying the questionnaire to a sample; and, the analysis of the data collected. In the first step, it was performed an extensive and systematized literature review on the bases Scientific Electronic Library Online - SciELO, Scopus and ISI Web of Science was developed to support the construction this model. The results from the literature review were the input for defining the set of evaluation criteria and to prepare a questionnaire for evaluating the users’ satisfaction. In the second step the questionnaire was applied to a sample by users of a library specialized in the field of nuclear technology, in Rio de Janeiro Brazil. In the third step, the data collected were complied and analysed though statistical frequency analysis. The sample was composed of 117 respondents of whom 94 returned forms. The results indicated that the criteria on which the library best performed were the conservation and adequacy of the collection. In the other hand, the results also have shown that the update criteria and international coverage of the collection are the ones that have should receive more attention, once they were perceived as having low performance and high importance for the users’ satisfaction. The main contributions of this research are: the criteria set and the frame for performing the survey and the way data were analysed.No ambiente da gestão de bibliotecas, vem à tona a seguinte questão: como avaliar a satisfação dos usuários com os serviços prestados pelas bibliotecas? Este artigo propõe e aplica um modelo para a avaliação da satisfação dos usuários quanto aos serviços oferecidos por uma biblioteca e, também, identificar o grau de importância dos diferentes aspectos deste tipo de serviço. Para isto uma revisão sistematizada da literatura foi desenvolvida a fim de dar suporte à construção do modelo. Esta revisão foi especialmente útil para definir o conjunto de critérios de avaliação e para a elaboração do instrumento de coleta, o qual foi aplicado a uma amostra composta por 117 usuários de uma biblioteca especializada na área de tecnologia nuclear, dos quais retornaram 94 formulários. Os dados coletados indicaram a importância e o desempenho de cada um dos aspectos considerados no questionário, apontando os critérios nos quais os usuários ficaram mais satisfeitos foram a conservação e a adequação do acervo. Por outro lado, os resultados também mostraram que atualização e cobertura internacional do acervo são os aspectos prioritários para ações de melhoria, visto que são aqueles em que os usuários estão menos satisfeitos e também são aqueles que os usuários consideram mais importantes. Destacam-se como contribuições da pesquisa: o conjunto de critérios; o instrumento de coleta; e, a proposta para tratamento dos dados
RELIABILITY STUDY OF THE AUXILIARY FEED-WATER SYSTEM OF A PRESSURIZED WATER REACTOR BY FAULTS TREE AND BAYESIAN NETWORK
Get PDFThis paper aims to present a study of the reliability of the Auxiliary Feed-water System (AFWS) through the methods of Fault Tree and Bayesian Network. Therefore, the paper consists of a literature review of the history of nuclear energy and the methodologies used. The AFWS is responsible for providing water system to cool the secondary circuit of nuclear reactors of the PWR type when normal feeding water system failure. How this system operates only when the primary system fails, it is expected that the AFWS failure probability is very low. The AFWS failure probability is divided into two cases: the first is the probability of failure in the first eight hours of operation and the second is the probability of failure after eight hours of operation, considering that the system has not failed within the first eight hours. The calculation of the probability of failure of the second case was made through the use of Fault Tree and Bayesian Network, that it was constructed from the Fault Tree. The results of the failure probability obtained were very close, on the order of 10-3
TRANSIENT MODELLING OF A NATURAL CIRCULATION LOOP UNDER VARIABLE PRESSURE
Get PDFThe objective of the present work is to model the transient operation of a natural circulation loop, which is one-tenth scale in height to a typical Passive Residual Heat Removal system (PRHR) of an Advanced Pressurized Water Nuclear Reactor and was designed to meet the single and two-phase ow similarity criteria to it. The loop consists of a core barrel with electrically heated rods, upper and lower plena inter- connected by hot and cold pipe legs to a seven-tube shell heat exchanger of countercurrent design, and an expansion tank with a descending tube. A long transient characterized the loop operation, during which a phenomenon of self-pressurization, without self-regulation of the pressure, was experimentally observed. This represented a unique situation, named natural circulation under variable pressure (NCVP). The self-pressurization was originated in the air trapped in the expansion tank and compressed by the loop
water dilatation, as it heated up during each experiment. The mathematical model, initially oriented to the single-phase ow, included the heat capacity of the structure and employed a cubic polynomial approximation for the density, in the buoyancy term calculation. The heater was modelled taking into account the di erent heat capacities of the heating elements and the heater walls. The heat exchanger was
modelled considering the coolant heating, during the heat exchanging process. The self-pressurization was modelled as an isentropic compression of a perfect gas. The whole model was computationally im-plemented via a set of nite di erence equations. The corresponding computational algorithm of solution was of the explicit, marching type, as for the time discretization, in an upwind scheme, regarding the space discretization. The computational program was implemented in MATLAB. Several experiments
were carried out in the natural circulation loop, having the coolant ow rate and the heating power as control parameters. The variables used in the comparison between experimental and calculated data were some relevant loop temperatures and pressures. The results obtained from the computational model agree qualitatively well with the experimental NCVP data
A STUDY USING VIRTUAL REALITY AS A SOURCE OF COMPLEMENTARY INFORMATION FOR NUCLEAR MEDICINE PATIENTS AND ITS RELATIVES
Get PDFThis work consists in the research, analysis and unification of the guidelines transmitted to the patients and their relatives in the radioiodine therapy procedures. The goal is to provide greater understanding of the use of nuclear radiation and better understanding of treatment, to help patients better adapt to therapy, to demystify misconceptions about radiation use, and to improve care for their protection and for people close to them. Based on written and verbal information, collected in the scientific literature and in loco accompanying the routine of the therapeutic rooms of Nuclear Medicine Services in Rio de Janeiro, the set of actions that define scenarios experienced by radioiodine therapy patients and their helpers is being generated. Based on this information, a virtual environment is being developed in the Virtual Reality Laboratory of the Institute of Nuclear Engineering (IEN / CNEN), a virtual environment that will allow the visualization of the procedures and instructions passed to the patients by the SMN1 teams. With this virtual environment, the patient will be able to perform an immersive visualization and to experience the different phases of the treatment, increasing the chances of efficiency of their participation in the process