GIOVANNI LARANJO DE STEFANI

GIOVANNI LARANJO DE STEFANI

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Parametric study for enhancing the radioisotope production in the IEA-R1 research reactor
2021 - STEFANI, GIOVANNI L. de; GENEZINI, FREDERICO A.; SANTOS, THIAGO A. dos; MOREIRA, JOAO M. de L.
In this work a parametric study was carried to increase the production of radioisotopes in the IEA-R1 research reactor. The changes proposed to implement in the IEA-R1 reactor core were the substitution of graphite reflectors by beryllium reflectors, the removal of 4 fuel elements to reduce the core size and make available 4 additional locations to be occupied by radioisotope irradiation devices. The key variable analyzed is the thermal neutron flux in the irradiation devices. The proposed configuration with 20 fuel elements in an approximately cylindrical geometry provided higher average neutron flux (average increment of 12.9 %) allowing higher radioisotope production capability. In addition, it provided 4 more positions to install irradiation devices which allow a larger number of simultaneous irradiations practically doubling the capacity of radioisotope production in the IEA-R1 reactor. The insertion of Be reflector elements in the core has to be studied carefully since it tends to promote strong neutron flux redistribution in the core. A verification of design and safety parameters of the proposed core was carried out. The annual fuel consumption will increase about 17 % and more storage space for spent fuel will be required.
The AP-Th 1000
2021 - STEFANI, GIOVANNI L. de; MAIORINO, JOSE R.; MOREIRA, JOAO M. de L.
This work presents the feasibility studies to convert the UO2 core of the Westinghouse AP1000 reactor to a U/Th core aiming at U/Th fuel recycling. The focus of the work is to establish a first core which allows normal operation of the AP1000 reactor and investigate a possible route for generating the 233U for U/Th fuel recycling. The converted core named AP-Th1000 is divided in three homogenous zones with different UO2/ThO2 mass proportions. The reprocessing procedure envisioned is to separate fission products and Pu isotopes, retain Uranium, use this fuel material in subsequent fuel cycles and complement the required fissile material with U with enrichment below 20%. The goal was to gradually reduce the mass proportion of mined Uranium fuel and eventually attain a Th-233U core with similar operation characteristics of current AP1000 core. We perform a detailed three-dimensional full core analysis with the SERPENT code examining core reactivity, power density distribution, and also a preliminary closed cycle study for the first 4 cycles where the production of 233U are evaluated. The goal of converting the AP1000 reactor core to a U/ThO2 fuel cycle was partially accomplished. While the first cycle was thoroughly examined and met all requirements we were not able to find a route to migrate it to a prevalent Th cycle. Basically, two of the set of criteria adopted in the study proved to be too restrictive to attain this goal with homogenous assembly, namely U enrichment below 20% and not recycling Pu. The results indicate that removing these two criteria the conversion factor in the ensuing fuel cycles can be increased and possibly attain a Th cycle without compromising the economics of power generation. The design changes were the elimination of IFBA burnable absorbers and replacement of gray control bundles by black control bundles.
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.
Detailed neutronic calculations of the AP1000 reactor core with the Serpent code
2019 - STEFANI, GIOVANNI L. de; MOREIRA, JOAO M.L.; MAIORINO, JOSE R.; ROSSI, PEDRO C.R.
In this work we present some validation results for reactor core modeling with the Serpent code performed for the first cycle of the AP1000 reactor. The comparison with reported values of the assembly k∞ for cold zero-power condition showed a discrepancy of 0.29%. The kef for full-core static and burnup calculations of the very heterogeneous AP1000 reactor core also presented good agreement with reported values. The kef for states with uniform fuel and moderator temperature distributions showed discrepancies below 0.91%. The boron worth curve obtained from burnup calculations with the Serpent code model results reproduced very well literature results despite using uniform temperature distributions in the modeling. In addition we discuss shadowing effects among burnable absorber rods (IFBA and Pyrex) and control rods which are, together with soluble boron, the control means throughout the first cycle. For instance, the presence of 9 Pyrex rods in an assembly decreased the average reactivity worth of one IFBA rod from 147 pcm to 33 pcm; and the presence of 28 IFBA rods in an assembly decreased the average reactivity worth of one Pyrex rod from 631 pcm to 277 pcm. The reactivity worth of a black control rod reduces about 20% when 28 IFBA rods are inserted in the fuel assembly.
The utilization of thorium in Small Modular Reactors – Part I
2018 - AKBARI-JEYHOUNI, REZA; OCHBELAGH, DARIUSH R.; MAIORINO, JOSE R.; DAURIA, FRANCESCO; STEFANI, GIOVANNI L. de
This work presents a neutronic assessment to convert a Small Modular Reactor (SMR) with uranium core to the thorium mixed oxide core with minimum possible changes in the geometry and main parameters of SMR core. This option is due to most of SMR are designed to be strongly poisoned in the beginning of cycle and to have a long cycle. Thorium can be used as an absorber in the beginning of the cycle and also be used as a fertile material during the cycle, it seems to be a good option to use (Th/U)O2 as SMR’s fuel. The main neutronic objectives of this study is achieving longer cycle length for SMR by using the minimum possible amount of burnable poison and soluble boron in comparison with reference core. The Korean SMART reactor as a certified design SMR has been chosen as the reference core. The calculations have been performed by MCNP code for homogeneous and heterogeneous seed and blanket concept fuel assemblies. The results obtained show that the heterogeneous fuel assembly is the one which gives longer cycle length and used lower amount of burnable poison and soluble boron, and also consumes almost the same amount of 235U.
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.