FABIO BRANCO VAZ DE OLIVEIRA

FABIO BRANCO VAZ DE OLIVEIRA

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Efeito do tratamento térmico nas propriedades microestruturais e eletroquímicas na liga La0,7Mg0,3Al0,3Mn0,4Co0,5Ni3,8 do tipo de AB5
2022 - SOARES, E.P.; CASINI, J.S.; LIMA, N.B.; LEAL NETO, R.M.; OLIVEIRA, F.B.; FARIA, R.N.
Neste artigo, foi investigada detalhadamente as curvas de isotermas de PCT Sievert dinâmico e microestruturas liga La0,7Mg0,3Al0,3Mn0,4Co0,5Ni3,8 do tipo de AB5, no estado bruto de fusão e tratada termicamente a 750, 850 e 1000°C por 16 horas e resfriada dentro do forno. Os resultados de DR-X, utilizando o método refinamento de Rietveld mostraram que houve uma sequencia de alterações das fases presentes na microestrutura estrutura da liga, modificadas pelo efeito dos tratamentos térmicos aplicados em todas as amostras, foi identificada a estrutura hexagonal do tipo CaCu5. Tendo a microestrutura da liga no estado bruto de fusão as fases LaNi5, MgNi2 e LaMg2Ni9 presentes na estrutura. Após a liga ter sido tratada termicamente as fases foram modificadas foram identificadas pelo refinamento utilizando o método de Rietveld, com o aparecimento da fase Al6Mn e LaCo13 e o total desaparecimento da fase LaMg2Ni9 e diminuição da fase MgNi2 alterando as suas estruturas cristalinas para de romboédrica para ortorrômbica e hexagonal para cubica respectivamente, nos tratamentos térmicos de 750° e 850°C por 16 horas promoveram melhor capacidade de descarga nas propriedades eletroquímicas da liga. As analises de isotermas de PCT da liga no estado bruto de fusão é comparada após ser tratada termicamente, onde a relação hidrogênio liga H/M diminui, e também para uma diminuição da pressão de formação platô. Essas alterações promovidas pelos tratamentos térmicos influenciaram nos resultados obtidos no desempenho eletroquímico, mostrando que a liga La0,7Mg0,3Al0,3Mn0,4Co0,5Ni3,8 tratada a 850°C por 16 horas obteve 420 mA.h que foi melhor capacidade máxima de descarga (Cmax.) e sua capacidade de estabilidade cíclica (Sn) se manteve em 90%.
Avaliação de combustível cerâmicos compostos baseados em urânio, tório e plutônio para reatores nucleares
2022 - GOMES, D.S.; OLIVEIRA, F.B.V.
The energy generated using thorium as nuclear fuel is an attractive way to preserve the uranium reserves and reduce the radiotoxicity wastes. Today, global thorium reserves are around four times that of uranium reserves, and ThO2 is cheaper than UO2. The next generation of reactor shows fast reactor types using (U-Pu)O2. It contains combinations of ceramic fuels based on UO2, PuO2, and ThO2. Using a new version of the FRAPCON code, get the ability to predict thorium fuels. The FRAPCON fuel codes permit the addition of capacity to simulate thorium fuel and compare fuel performance with standard UO2. The proposed strategy uses ThO2 75 wt% composed of UO2 25 wt% with enrichment equal to 19.5 wt%. In comparison, the second strategy uses a balanced composition of ThO2 at 93 wt.% combined with PuO2 at 7.0 wt.%.
Combustível composto avançado UO2-UN com condutividade térmica melhorada e alta densidade
2022 - GOMES, D.S.; OLIVEIRA, F.B.V.
Today, many efforts focus on ceramic fuels that can replace standard uranium dioxide (UO2). In this context, uranium nitride (UN) could replace UO2 used in light water reactors (LWRs). The UN fuel has a higher uranium density and better thermal conductivity than UO2. The drawback of UN is a lower oxidation resistance in contact with the water. Regardless, adding a compound that acts as a protective barrier against nitride oxidation, such as ZrN, U3Si2, and UO2, could reduce water oxidation. Early, UO2–UN composites were fabricated by hot pressing UO2–UN powder mixtures within 1300 °C–1590 °C. The manufacturing process adopted for the fuel pellet fabrication changed to the spark plasma sintering (SPS) method. Using SPS can avoid the exceptional resistance to producing a perfect microstructure and the high costs associated with 15N enrichment. The composite fuel suggested is UN-10% U3Si5 verified with FRAPCON code.
Avaliação da liberação de gás de fissão para UO2 dopado por Cr2O3
2022 - GOMES, D.S.; OLIVEIRA, F.B.V.
Uranium dioxide has been the most common ceramic fuel used to generate electric power in the last sixty years. The lower addition of chromic oxide (Cr2O3) shows the benefits of large grain size. Using (Cr2O3-Al2O3)-doped UO2 comprises improved mechanical response and fission gas retention properties. The addition of Cr2O3 to UO2 slightly affected its thermal properties as proposed for pressurized water reactors and boiling water reactor designs. Advanced doped pellet technology (ADOPT) can improve fuel cycle economics and accident tolerance. Today, doped fuel pellets exhibit several favorable characteristics: large grain size, increased density, minor fission gas release (FGR), and reduced pellet cladding interactions. In simulation are used FRAPCON to simulate doped fuel and ferritic alloy as cladding. Most of the material properties of the Cr2O3-doped UO2 were identical to those of the classic UO2.
Thermal analysis of nuclear fuel using silicon carbide nanocomposite dispersion in UO2
2020 - GOMES, D.S.; OLIVEIRA, F.B.
After the Fukushima Daiichi disaster happened in Japan in 2011, it started a global effort to get more tolerant fuels. In 2019, the fleet of power reactors designated for electricity suppliers made up 451 power units, generating around 402 GWe. The nuclear power represents 11.2% of the electricity generated, avoiding about 1.2 GT of CO2. The civilian reactors are operating using the uranium dioxide (UO2) as the fuel, which shows poor thermal conductivity of 7.8 W/m-K at room temperature. The fuel temperatures can reach up until 1500 °C at regular operation. Silicon Carbide Nanotube (SiC-CNT) dispersed in the UO2 matrix containing 5 to 20% vol of SiC-CNTs permits to increases the thermal conductivity. The novel fuel concept improves the thermal conductivity of 30% with the addition of 5% of silicon carbide. The fuel pellet UO2-SiC/CNTs are sintered using Spark Plasma Sintering (SPS) with a hold time of 5 minutes, at 1300 °C, and a pressure of 40 MPa. The fuel mixture shows a better density, low porosity, and acceptable grain size distribution compared to traditional sintering routes. It simulated the fuel mixtures using fuel performance code FRAPCON adapted to the thermals and mechanic properties of compounds. This study showed the possibility of increasing the safety margins of nuclear fuel using the addition of a small fraction of nanocomposite.
Analysis of a pressurized power reactor using thorium mixed fuel under regular operation
2019 - GOMES, DANIEL de S.; STEFANI, GIOVANNI L. de; OLIVEIRA, FABIO B.V. de
This work discusses a parametric study applied to nuclear power generation based on a mixed fuel formed by the composition of thorium-uranium oxide (Th-U)O2. Also, approached in this study the physical neutrons models of a fuel system composed of ThO2 75 wt% and UO2 25 wt%, with 19.5% enrichment of U-235. The thermodynamic features of the thorium-uranium fuel system compared with the properties of uranium dioxide. Thorium-based fuel operating extended fuel cycles reach of over 80 GWd/MTU in a pressurized water reactor (PWR). Homogenous distribution of thorium-based fuel, used on the reactor core, could reduce Pu-239, once U-233 production capacity dependent on Th-232 replacing U-238 in the fuel matrix. The mixed oxide fuel has a lower buildup of Pu-239, causing the linear heat rate distribution slope to flatten and lowering fuel porosity. The release of gaseous fission products models for (Th-U)O2 could have different diffusion coefficients when compared to uranium oxide models. Besides, resulting in lower thermal gradients than UO2 and a reduction in fuel swelling. This parametric study reviews the aspects of radioactive decay chains of uranium and thorium. It founded the simulation using approved nuclear codes, such as SERPENT for neutron physics calculations and the FRAPCON code, which defines the licensing process. The results show that thoria based fuel has a higher performance than UO2 fuel in regular operation and can improve safety margins.
Behavior of thorium plutonium fuel on light water reactors
2019 - GOMES, DANIEL S.; SILVA, ANTONIO T. e; OLIVEIRA, FABIO B.V. de; LARANJO, GIOVANNI S.
Designs using thorium-based fuel are preferred when used in compliance with sustainable energy programs, which should preserve uranium deposits and avoid the buildup of transuranic waste products. This study evaluates a method of converting uranium dioxide (UO2) to thorium-based fuel, with a focus on Th-Pu mixed oxide (Th-MOX). Applications of Th-MOX for light water reactors are possible due to inherent benefits over commercial fuels in terms of neutronic properties. The fuel proposed, (Th-Pu)O2, can be helpful because it would consume a significant fraction of existing plutonium. Aside from the reactor core, the proposed fuel could be useful in existing technology, such as in a pressurized water reactor (PWR). However, licensing codes cannot support Th-MOX fuel without implementing adaptations capable of simulating fuel behavior using the FRAPCON code. The (Th-Pu)O2 fuel should show a plutonium content that produces the same total energy release per fuel rod when using UO2 fuel. Thorium is a fertile material and demands a slightly higher plutonium content when used in Th-MOX. Mixed ceramic oxides show thermodynamic responses that depend on the comprising chemical fractions, and there is little information in databases on irradiation effects. The neutronic analysis is carried out using the SERPENT code to quantify transuranic production and compare this production with the original UO2 fuel assembly. Parameters such as delayed neutron fraction and temperature reactivity coefficient are also determined. Through these analytical methods, the viability and sustainability of the proposed new fuel assembly can be demonstrated in a closed fuel cycle.
Experimental study of radiation influence on thermophysical properties of Al2O3 and ZrO2 nanofluids
2017 - PINHO, PRISCILA G.M.; OLIVEIRA, FABIO B.V.; ANDRADE, DELVONEI A.; ROCHA, MARCELO S.
Nanofluids are a promising technology for application in nuclear reactor systems for high heat flux transport. As demonstrated by the recent researches, nanofluids have very interesting physical properties with respect to its ability to remove and transport of heat. There is, currently, research groups in the world conducting investigations on the influence of ionizing radiation on nanofluids and the possibility of its use as working fluid or cooling of the core of nuclear reactors core in cases of accidents. Among the countless applications presently proposed for the nanofluids, the applications in energy have special attention by academic and industrial interest. Studies demonstrate that nanofluids based on metal oxide nanoparticles have physical properties that characterize them as promising working fluids, mainly, in industrial systems in which high heat flux want to be removed. The nuclear reactors for power production are examples of industry where such an application has been proposed. However, there are no concrete results about the ionizing radiation effects on nanofluids properties. This work aims to present the initial results of the current study carried out with the objective to check the effects caused by that ionizing radiation on nanofluids based on Al2O3 and ZrO2 nanoparticles. Results from thermophysical analyses demonstrate that particular behavior on thermal conductivity, and density of such nanofluids can be observed as a function of temperature under no ionizing radiation effect. New investigations will analyze the application potentiality of some nanofluids in nuclear systems for heat transfer enhancement under ionizing radiation influence.
A CFD numerical model for the flow distribution in a MTR fuel element
2017 - ANDRADE, D.A.; ANGELOA, G.; ANGELO, E.; SANTOS, P.H.G.; OLIVEIRA, F.B.V.; TORRES, W.M.; UMBEHAUN, P.E.; SOUZA, J.A.B.; BELCHIOR JUNIOR, A.; SABUNDJIAN, G.; PRADO, A.C.
Previously, an instrumented dummy fuel element (DMPV-01), with the same geometric characteristics of a MTR fuel element, was designed and constructed for pressure drop and flow distribution measurement experiments at the IEA-R1 reactor core. This dummy element was also used to measure the flow distribution among the rectangular flow channels formed by element fuel plates. A CFD numerical model was developed to complement the studies. This work presents the proposed CFD model as well as a comparison between numerical and experimental results of flow rate distribution among the internal flow channels. Numerical results show that the model reproduces the experiments very well and can be used for the studies as a more convenient and complementary tool.
Numerical simulation of isothermal upward two-phase flow in a vertical tube of annular section
2017 - CERAVOLO, FLAVIO E.; ROCHA, MARCELO da S.; OLIVEIRA, FABIO B.V. de; ANDRADE, DELVONEI A. de
A numerical simulation of a vertical, upward, isothermal two-phase flow of air bubbles and water in an annular channel applying Computational Fluid Dynamics code (CFD) was carried out. The simulation considers an Eulerian frame, with two-fluid model, specific correlations for turbulence model considering the dispersion and bubble induction turbulence. The work intends to assess whether the code represents the physical phenomenon accurately by comparing the simulation results with experimental data obtained from literature. The annular channel adopted has equivalent hydraulic diameter of 19.1 mm, where the outer pipe has an internal diameter of 38.1 mm and inner rod 19.1 mm. To represent this geometry, a three-dimensional mesh was generated with 960000 elements, after a mesh independence study. The void fraction distribution, taken radially to the flow section is the main parameter analyzed besides interfacial area concentration, interfacial gas velocity, diameter and distribution of bubbles.