GIOVANNI LARANJO DE STEFANI

GIOVANNI LARANJO DE STEFANI

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STC-MOX-Th
2020 - SANTOS, THIAGO A. dos; STEFANI, GIOVANNI L. de
O trabalho trata da criação de um programa elaborado em ambiente MATLAB que calcula os limites térmicos de projeto de um típico reator a água pressurizada (PWR), que é a temperatura central da pastilha combustível e a taxa de ebulição nucleada (DNBR). Outras distribuições de temperatura e grandezas hidrodinâmicas do líquido refrigerante, como a entalpia e a queda de pressão também são calculadas. O código possui peculiaridades, como o fato de permitir cálculos com combustíveis de UO2 puro e proporções do óxido misto de Urânio/Tório - MOX (U,Th). Estas, além da sua interface amigável com o usuário provam que o código pode ser utilizado em trabalhos de pesquisa , bem como em disciplinas de graduação e pós graduação voltadas ao estudo de termo-hidráulica de reatores nucleares em cursos de graduação e pós graduação de engenharia (nuclear e/ou da energia) espalhados pelos país, como no caso do curso de graduação de Engenharia de Energia da Universidade Federal do ABC, onde é uma disciplina optativa. Para a validação do código foram utilizados dados do reator AP-1000 da Westinghouse. O programa se apresentou com comportamento físico dentro do esperado para o modelo, gerando resultados confiáveis para eventuais projetos de reatores (validado com dados experimentais e outros programas), bem como propicia a alunos uma experiência diferenciada dentro da aprendizagem dos conceitos empregados na área, uma vez que o programa permite uma análise mais profunda de determinados conceitos na área de termo-hidráulica que dentro da aula expositiva e com exercícios convencionais não poderiam ser explorados.
Thermohydraulic analysis of a fuel element of the AP1000 reactor with the use of mixed oxides of U / Th using the computational fluid dynamic code (CFX)
2019 - CUNHA, CAIO J.C.M.R.; RODRÍGUEZ, DANIEL G.; LIRA, CARLOS A.B.O.; STEFANI, GIOVANNI L.; LIMA, FERNANDO R.A.
The present work carried out a thermohydraulic analysis of a typical fuel assembly of the reactor AP1000 changing the type of fuel, of UO2 conventionally used for a mixture of oxides of (U,Th)O2 realizing some simplifications in the original design, with the objective to develop of an initial methodology capable of predicting the thermohydraulic behavior of the reactor within the limits established by the manufacturer. An expression for the power density was determined using a coupled neutronic thermohydraulic calculation; once the final expression for power density was determined, the axial and radial temperature profiles in the assembly, as well as the pressure drop and the distribution of the coolant density, were evaluated. Due to the increase in research done on thorium, such as the work of [1], [2], [3], [4] and [5], as well as the mass diffusion of the AP1000, as is the case with [6] and [7]. The present study developed a simplified model, where burnable poisons and spacer grids were not considered, however, it is a consistent model, but with the insertion of these, a more accurate representation of the reactor is expected, providing operational transient analyzes. This tends to strengthen the lines of research that have been carrying out work on the AP1000, as well as in the general sphere of nuclear power plants.
Consolidation of the new nuclear calculation methodology of the IEA-R1 reactor
2017 - STEFANI, GIOVANNI L. de; CONTI, THADEU das N.
The IEA-R1 neutron and thermo-hydraulic calculation methodology is composed of 5 computational codes from the area of reactor physics, which have a symbiotic dependence on each other. Since the outputs of each code will be used to generate the input of the next code. The programs involved in this methodology are LEOPARD, HAMMER-TECHNION, TWODB, CITATION and COBRA. Each of these codes is responsible for a specific type of calculation. In a first two-year study, between the years 2008 and 2010, these IEA-R1 nuclear reactor codes were integrated into a single management code. This management code had as main objective the reduction of the time spent by the calculation team of the reactor and to prevent against errors in the manipulation of data and data output. In this study the calculation time was reduced by 99%. The present article presents the closing of this work, being a document with the consolidation of 7 years of the use of the new calculation methodology implemented in the IEA-R1 reactor, demonstrating its efficiency and reliability, besides the proper registration of this project that had great importance Within the IPEN research reactor.
Feasibility to convert an advanced PWR from UO2 to a mixed (U,Th)O2 core
2017 - STEFANI, GIOVANNI L. de; MAIORINO, JOSE R.; MOREIRA, JOAO M. de L.; SANTOS, THIAGO A. dos; ROSSI, PEDRO C.R.
This work presents the neutronics and thermal hydraulics feasibility to convert the UO2 core of the Westinghouse AP1000 in a (U-Th)O2 core, rather than the traditional uranium dioxide, for the purpose of reducing long-lived actinides, especially plutonium, and generates a stock pile of 233U, which could in the future be used in advanced fuel cycles, in a more sustainable process and taking advantage of the large stock of thorium available on the planet and especially in Brazil. The reactor chosen as reference was the AP1000, which is considered to be one of the most reliable and modern reactor of the current Generation III, and its similarity to the reactors already consolidated and used in Brazil for electric power generation. The results show the feasibility and potentiality of the concept, without the necessity of changes in the core of the AP1000, and even with advantages over this. The neutron calculations were made by the SERPENT code. The results provided a maximum linear power density lower than the AP1000, favoring safety. In addition, the delayed neutron fraction and the reactivity coefficients proved to be adequate to ensure the safety of the concept. The results show that a production of about 260 Kg of 233U per cycle is possible, with a minimum production of fissile plutonium that favors the use of the concept in U-Th cycles.
A thermal hydraulic analisys in PWR reactors with UO2 or (U-Th)O2 fuel rods employing a simplified code
2017 - SANTOS, THIAGO A. dos; MAIORINO, JOSE R.; STEFANNI, GIOVANNI L. de
In order to project a nuclear reactor, the neutronic calculus must be validated, so that its thermal limits and safety parameters are respected. Considering this issue, this research aims to evaluate the APTh thermal limits. This PWR is a project develope composed of Uranium and Thorium oxide mixed (U,Th)O2. For this purpose, a simplified, although conservative, code was developed in a MATLAB environment named hydraulics Code-Mixed Oxide Thorium”. This code provides axial and radial temperature distribution, as well as DNBR distribution over the hottest channel of the reactor core. Moreover, it brings other hydraulic quantities, such as pressure drop over the fuel rod, considering any fuel proportion of (U,Th)O2.The software uses basic laws of conservation of mass, momentum and energy, it also calculates the thermal conduction equation, considering the thermal conductive coefficie finite elements method was used. Furthermore, the proportion of 36% of UO2 was used to evaluate the temperature over the fuel rod and DNBR minimum in three burn conditions: beginning, middl program has proven to be efficient in every condition and the results evidenced that the APTh an initial analysis, has its thermal limits within the recommended security parameters.
Automatização da metodologia de cálculo neutrônico e termo-hidráulico do reator nuclear IEA-R1
2008 - STEFANI, GIOVANNI L. de; CONTI, THADEU das N.
Finalização da automatização dos cálculos neutrônicos e termo hidráulicos e parâmetros de segurança do reator IEA-R1
2011 - STEFANI, GIOVANNI L. de; CONTI, THADEU das N.
Automatizaçao dos programas e codigos computacionais utilazados na metologia de calculo neutronico e termohidraulico do reator nuclear IEA-R1
2009 - SETEFANI, GIOVANNI L.; CONTI, THADEU das N.
Implementation of the optimization for the methodology of the neutronic calculation and thermo-hydraulic in IEA-R1 reactor
2011 - STEFANI, GIOVANNI L. de; CONTI, THADEU das N.; SANTOS, THIAGO A. dos; FEDORENKO, GIULIANA G.; CASTRO, VINICIUS A.; MAIO, MIREIA F.