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Modeling of the CTEx subcritical unit using MCNPX code
Get PDFThe present work aims at simulating the subcritical unit of Army Technology Center (CTEx) namely ARGUS pile (subcritical uranium-graphite arrangement) by using the computational code MCNPX. Once such modeling is finished, it could be used in k-effective calculations for systems using natural uranium as fuel, for instance. ARGUS is a subcritical assembly which uses reactor-grade graphite as moderator of fission neutrons and metallic uranium fuel rods with aluminum cladding. The pile is driven by an Am-Be spontaneous neutron source. In order to achieve a higher value for keff, a higher concentration of U235 can be proposed, provided it safely remains below one
Macroscopic multigroup constants for accelerator driven system core calculations
Get PDFThe high-level wastes stored in facilities above ground or shallow repositories, in close connection with its nuclear power plant, can take almost 106 years before the radiotoxity became of the order of the background. While the disposal issue is not urgent from a technical viewpoint, it is recognized that extended storage in the facilities is not acceptable since these ones cannot provide sufficient isolation in the long term and neither is it ethical to leave the waste problem to future generations. A technique to diminish this time is to transmute these
long-lived elements into short-lived elements. The approach is to use an Accelerator Driven System (ADS), a sub-critical arrangement which uses a Spallation Neutron Source (SNS), after separation the minor actinides and the long-lived fission products (LLFP), to convert them to short-lived isotopes. As an advanced reactor fuel, still today, there is a few data around these type of core systems. In this paper we generate macroscopic multigroup constants for use in calculations of a typical ADS fuel, take into consideration, the ENDF/BVI data file. Four energy groups are chosen to collapse the data from ENDF/B-VI data file by PREPRO code. A typical MOX fuel cell is used t
Calibração por comparação do sistema de aquisição de dados de temperatura de um circuito de circulação natural
Get PDFEste trabalho apresenta os procedimentos de calibração por comparação do sistema de aquisição de dados de temperatura do Circuito de Circulação Natural (CCN) do Instituto de Engenharia Nuclear (IEN). A instrumentação do circuito compreende 12 termopares do tipo K, um medidor de vazão eletromagnético, um
sistema de aquecimento e controle, e um sistema de aquisição de dados de temperatura e vazão. A medição de temperaturas no CCN requer o conhecimento dos erros e incertezas dos equipamentos e sistemas envolvidos na medição. Para isso foi feita uma montagem experimental em bancada, formada por um banho termostático, um termômetro de bulbo de referência, doze termopares do tipo K e o sistema de aquisição de dados. A análise foi feita em duas etapas: inicialmente foi feita uma avaliação estatística dos sinais dos termopares, através das diferenças das leituras entre eles. A segunda etapa consistiu na avaliação das incertezas envolvidas no procedimento de calibração utilizado
Alguns passos para o cálculo do núcleo de um reator nuclear com realimentação termohidráulida utilizando o método dos elementos finitos
Get PDFEste trabalho tem por objetivo uma análise do núcleo de um reator nuclear com realimentação termohidráulica. Para isto, alguns passos foram executados nesta análise tais como o emprego do método de cálculo da difusão neutrônica pelo MEF-DIFU, o cálculo da variação de temperatura no elemento combustível (MEF-HEAT) que permite estabelecer a distribuição de temperatura no elemento combustível, o cálculo das de seções de choque, via o programa HAMMER (MEF-HAMM) para diferentes condições de temperatura do núcleo do reator e finalmente uma análise de cinética pontual do reator para analisar o comportamento do mesmo face a estas variações de temperatura, via a densidade dos materiais que compõem o núcleo do reator (principalmente o moderador). Para testar o sistema de cálculo integrado, procurou-se, inicialmente, sem a influência da temperatura, estabelecer uma espessura ótima para o Cermet da placa combustível do reator Argonauta já que esta análise foi realizada anteriormente de uma forma analítica [Aghina, 1990]. Estes passos foram executados com o objetivo de desenvolver a interface MEF: DIFUHAMM- HEAT. Futuramente, o programa de cálculo das seções de choque (HAMMER) poderá ser substituída por um outro que considere a variação de temperatura no cálculo das seções de choque do elemento combustível em função de sua distribuição de temperatura obtida através do programa MEF-HEAT
Virtual evacuation simulation with autonomous avatars
Get PDFThis paper describes the use of virtual reality technology for virtual simulation of crowded evacuation from sites. The approach adopted is the reuse of a game engine, thus taking advantage of all its features for virtual environment design. This work upgrades a previously developed one, in which users played simultaneously in a networked environment, each one controlling his or her own avatar. But for crowded evacuation situations, it would require many users playing simultaneously in networked computers. The more crowded the simulation, the more users needed, what could be difficult a task, depending upon the number of avatars needed. Autonomous avatars can surpass this difficulty, so few users can participate, together with as many autonomous avatars as needed, to simulate the desired crowded scenarios. First results show the viability of such an approach
Sentiment analysis domain based environment
Get PDFIn this article, the Sentiment Analysis, which main task is to understand the polarity of the opinions expressed in the documents, is used to obtain a better understanding of how people express themselves about a particular subject or domain. Such understanding is important because it may provide input in the formulation of policies and actions to be taken in relation to a product or service or to the perceptions related to issues or people. As more people put their thoughts and opinions on a variety of services available on the Web microblogging sites like Twitter, social networks and forums have become a common way for them to express themselves. There, they put in a spontaneous, free and in real time over different views on the issues. However, each domain or Internet service has its own peculiarities. Some jargon is specific to a domain. Certain slang or characteristics of services for people putting their opinions differ significantly, which undermines the use of machine learning systems previously developed for other domains. To validate the proposed methodology a corpus of nuclear texts was assembled, annotated and analyzed. After that, a system was developed to process these data. The results thus obtained provide important information on the identification of key stakeholders and the views expressed on what subjects. With this better understanding of the fears, prejudices and expectations of people regarding the nuclear area, new strategies can be traced to improve the perception of them as the nuclear area
Non-destructive assay employing 2D and 3D digital radiographic imaging acquired with thermal neutrons and reactor-produced radioisotopes
Get PDFThe inner structure of some objects can only be visualized by using suitable techniques, when safety reasons or expensive costs preclude the application of invasive procedures. The kind of agent rendering an object partially transparent, unveiling thus its features, depends upon the object size and composition. As a rough rule of thumb, light materials are transparent to gamma and X-rays while the heavy ones are transparent to neutrons. When, after traversing an object, they hit a proper 2-D detector, a radiograph is produced representing a convoluted cross section, called projection, of that object. Taking a large number of such projections for different object attitudes, it is possible to obtain a 3-D tomography of the object as a map of attenuation coefficients. This procedure however, besides a time-consuming task, requires specially tailored equipment and software, not always available or affordable. Yet, in some circumstances it is feasible to replace the 3-D tomography by a stereoscopy, allowing one to visualize the spatial configuration of the object under analysis. In this work, 2-Dand 3-D radiographic images have been acquired using thermal neutrons and reactor-produced radioisotopes and proper imaging plates as detectors. The stereographic vision has been achieved by taking two radiographs of the same object at different angles, from the detector point of view. After a treatment to render them red-white and green-white they were properly merged to yield a single image capable to be watched with red-green glasses. All the image treatment and rendering has been performed with the software ImageJ
Neutron radiographic testing of samples of special concrete containing recycled pet granules as aggregate
Get PDFThis study aimed at inspecting microcracks in test specimens of special concrete, through neutron radiography tests. The thermal neutron flux used was extracted from the J-9 irradiation channel, placed in the thermal column of Argonauta/IEN/CNEN/RJ reactor, where a neutron radiographic system is installed. The test specimens inspected were molded in a cylindrical shape, with standard concrete and modified concrete where coarse sand was substituted by granules of recycled PET. They were submitted to compression in a SHIMADSU UH F 1000 press, causing microcraks. Then, slices of 50 μm thickness were obtained using an electrical saw. Gadolinium nitrate was used as contrast liquid in order to enhance the visualization of those microcracks. The Neutron Radiography technique proved to be appropriate for this kind of inspection, allowing to clearly visualizing the microcracks. Recycled PET granules met ABNT standards, and may be used in the construction of low income people houses, as structural concrete (25 % PP) or house floors (25% to 50% PEAD). The mechanical properties of compression and elasticity demonstrated for this special concrete, on Civil Engineering conventional tests, and by the neutron radiographic images obtained, showed that its use is viable even for civil construction in areas subject to seismic events
Artificial neural network used in the study of sensitives in iris reactor pressurizer
Get PDFIn general, the technique of sensibility analysis studies the behavior of the ratio between the variation of output
results and the variation of input parameters. This study performed in the reactor pressurizer, which is a
component responsible for control of the pressure inside the vessel, has fundamental importance in designing the security of any concept of advanced reactor. Above all, for its feature of passive action of the pressurizer (there is no spray), this analysis becomes a necessary step for safety and performance of the plant. The direct method through code MODPRESS, which represents the pressurizer model of the International Reactor Innovative and Secure (IRIS), has required large computational effort. Unlike this method, Artificial Neural Networks (ANN), beyond faster, do not require a typical linear behavior of the system. Moreover, they can also use experimental data for their training and learning. If the ANN are satisfactory in this theoretical case may be used for future mapping and forecasting of the behavior of various phenomena in both plant operation and small-scale experiment to be installed in CRCN-NE. Based on the results obtained in this study, one can conclude that the artificial neural networks are presented as an alternative to MODPRESS code, as well as a great tool to calculate the sensitivity coefficient
Alguns passos para o cálculo do núcleo de um reator nuclear com realimentação termohidráulica utilizando o método dos elementos finitos
No full textEste trabalho tem por objetivo uma análise do núcleo de um reator nuclear com realimentação termohidráulica. Para isto, alguns passos foram executados nesta análise tais como o emprego do método de cálculo da difusão neutrônica pelo MEF-DIFU , o cálculo da variação de temperatura no elemento combustível (MEF-HEAT) que permite
estabelecer a distribuição de temperatura no elemento combustível, o cálculo das de seções de choque, via o programa HAMMER (MEF-HAMM) para diferentes condições de temperatura do núcleo do reator e finalmente uma análise de cinética pontual do reator para analisar o comportamento do mesmo face a estas variações de temperatura, via a densidade dos materiais que compõem o núcleo do reator (principalmente o moderador). Para testar o sistema de cálculo integrado, procurou-se, inicialmente, sem a influência da temperatura, estabelecer uma espessura ótima para o Cermet da placa combustível do reator Argonauta já que esta análise foi realizada anteriormente de uma forma
analítica [Aghina, 1990]. Estes passos foram executados com o objetivo de desenvolver a interface MEF: DIFUHAMM-HEAT. Futuramente, o programa de cálculo das seções de choque (HAMMER) poderá ser substituída por
um outro que considere a variação de temperatura no cálculo das seções de choque do elemento combustível em função de sua distribuição de temperatura obtida através do programa MEF-HEAT