Mathematical tools for calculating radiation shielding usually have difficult notations that could only be solved by computational methods. The Albedo’s method applied for calculation of shielding proves to be an excellent substitute in determining the incident beam fractions that are reflected, absorbed and transmitted, avoiding the use of the transport equation and diffusion approximation that are extremely important in nuclear reactor designs and irradiation calculations. Based on the simple following of the radiation current path, the method can be characterized as a graphical and analytical solution. This work explores the Albedo’s Method applied to 4-slab shielding with incidence of gamma radiation to an energy group compared to ANISN, computational method consecrated in the area of nuclear calculations.
O setor de pesquisa e desenvolvimento de um país tem suma importância para a disseminação e avanço do conhecimento, implicando diretamente no setor econômico e na visibilidade da nação internacionalmente, sob esta perspectiva a energia nuclear se mostra um assunto crescente e de suma importância para a expansão do programa nuclear brasileiro. O reator Argonauta é um reator de pesquisa localizado na Universidade Federal do Rio de Janeiro (UFRJ) no prédio do Instituto de Engenharia Nuclear (IEN). Por ser um reator de pesquisa, sua potência é baixa e, devido a isso, diversos experimentos podem ser realizados no mesmo. Como em todo reator um dos fatores que estão sempre presentes na determinação da segurança intrínseca é a sua Reatividade. Há diversos fatores que fazem o controle da Reatividade de um reator, dentre eles a temperatura no elemento combustível, no moderador e no refrigerante. Tendo isso em vista, o objetivo deste trabalho é analisar a influência que a temperatura do moderador faz na reatividade do reator onde será possível de determinar o chamado coeficiente global de temperatura (ou coeficiente de temperatura no moderador) do reator Argonauta e compará-lo com os parâmetros nucleares do reator. Com a utilização da metodologia aplicada pelos operadores do reator durante a operação, é possível definir uma relação reatividade vs temperatura para a determinação deste coeficiente e verificação da segurança do reator quanto ao aumento de temperatura.
This research characterizes the temperature profiles inside a cylindrical nuclear fuel pin cooled by water in which the surface temperature of the cladding material is specified. Uniform heat generation by fission is considered. Two configurations were analyzed: 1) a cylindrical fuel rod constituted by uranium; and 2) a cylindrical fuel rod constituted by uranium encased in a zircalloy-4 cladding and a gap filled with helium between them. Heat transfer is constant, and the analysis is one-dimensional because symmetry is observed about the rod centerline. Constant thermal conductivity is assumed for the analytical solutions, while the numerical solution considered variable thermal conductivities. Simulations were performed using the software COMSOL Multiphysics, and the results were plotted using Octave. Results showed agreements between analytical and numerical solutions, which indicates that the methods applied in the present work can be used in order to perform educational and research studies in engineering subjects such as heat transfer and thermal hydraulics in the nuclear area.
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