DANIEL DE SOUZA GOMES

DANIEL DE SOUZA GOMES

(Fonte: Lattes)

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Graduating from Fundação Educacional Inaciana Padre Sabóia de Medeiros, FEI (1987), Master in Electrical Engineering from Escola Politécnica of the University of São Paulo (2002), Ph.D. in Nuclear Technology from the University of São Paulo, USP (2014). Post Doctorate by the Energy and Nuclear Research Institute, IPEN (2018). He is currently a technologist at the National Nuclear Energy Commission IPEN-SP, at the nuclear engineering center (CEN). (Text obtained from the Currículo Lattes on May 4th 2023)

Possui graduação em Engenharia Elétrica pela Fundação Educacional Inaciana Padre Sabóia de Medeiros FEI (1987), mestrado em Engenharia Elétrica pela Escola Politécnica da Universidade de São Paulo (2002), doutorado em Tecnologia Nuclear pela Universidade de São Paulo (2014). Pós Doutorado pelo Instituto de Pesquisas Energéticas e Nucleares, (2018). Atualmente é tecnologista da Comissão Nacional de Energia Nuclear IPEN-SP, centro de engenharia nuclear. (Texto extraído do Currículo Lattes em 04 maio 2023)

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Overview of the physical properties of molten salt reactor using FLiBe
2023 - GOMES, DANIEL de S.
Currently, there are six Generation IV nuclear reactor designs in development. Four are fast neutron reactors, and all designs operate at higher temperatures that permit hydrogen production. Thus, the interest in fluoride salts has grown due to their hightemperature application in fission and fusion reactor designs. The aircraft propulsion project was the initial plan, which used molten salt as a coolant and was started by Bettis and Briant in the 1940s. The molten salt reactor has been designed to operate at temperatures of 700 to 800°C with fissile material dissolved in a molten fluoride salt composition. Molten fluoride salts are stable at high temperatures, show good thermodynamic properties, and can also dissolve actinides and fission products easily. It creates a candidate for a thorium reactor with more than 45% efficiency. The purpose of this work was to investigate the physical characteristics of two systems of fluoride salt combinations, namely LiF-BeF2 (FliBe) and LiF-NaF-KF (FliNaK), including melting temperature, density, and heat capacity. The aim is to characterize the advantages of the various designs proposed for Generation IV by reviewing properties evidenced by safety improvements and limitations.
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.
Computational assessment using sensitivity analysis of more tolerant fuels
2020 - GOMES, DANIEL de S.
The secure operation of nuclear technology, including the avoidance of nuclear accidents and the optimal use of fuel cycles, poses a significant challenge. As of 2020, a total of 442 nuclear reactors are in operation in 30 countries across the world, accounting for over 13% of the global electricity demand. Over the last 50 years, light water reactors have operated using uranium dioxide as fuel and zirconium alloys as cladding. However, extensive research on more tolerant fuels has been conducted with high priority since the Fukushima disaster in 2011. New tolerance ideals seek the replacement of the standard fuel system with state-of-the-art options soon. Such investigations have predominantly focused on the responses and physical properties of new candidates for tolerant fuels, which are superior to those of standard fuel systems, based on close time-advanced alloys manufactured by substituting zirconium alloys with FeCrAl for enhanced corrosion resistance. Further, more tolerant fuels exhibit high uranium densities, with better thermal conductivities than uranium dioxide. This study proposes a fuel-licensing code as safety criteria collaborating with stochastic models and conservative rules. It also incorporates a sensitivity analysis using U3Si2 as fuel and FeCrAl alloys as cladding.
Progress of enhanced conductivity fuels using UO2-graphene
2021 - GOMES, DANIEL de S.; STEFANI, GIOVANNI L. de
Uranium dioxide (UO2) is the most used fuel in light water reactors. At present, a sizable cumulative experience exists regarding the use of UO2 as a fuel. However, UO2 has reduced thermal conductivity Experiments shown that adding a second phase with higher thermal conductivity will improve the thermal conductivity of mixed fuel. Materials such as MO and BeO dispersed in the UO2 matrix attract the most attention. Graphene has excellent thermal conductivity and a low absorption cross-section. Metallurgic routes used in UO2–carbon composites use the spark plasma sintering method. Thus, analyze the behavior of graphene nanoparticles dispersed in a uranium dioxide matrix, simulated with FRAPCON code. Early experiments revealed that using UO2–10 vol.% silicon carbide improved the thermal conductivity by 30%. Graphene properties have a substantial impact on the thermal response of the fuel. In extension, carbon allotropic forms sintered with UO2 are potential options like UO2-diamond and UO2-nanotubes.
Assessment of advanced ferritic alloys used as cladding materials in nuclear power reactors
2021 - GOMES, DANIEL de S.; GIOVEDI, CLAUDIA
The fuel performance code, Fuel Analysis under Steady-state and Transients (FAST), permits cladding options, such as zirconium alloys and iron-chromium-aluminum (FeCrAl). FAST code support as cladding Kanthal, CM35, and CM36 alloys. We implemented a comparative analysis between ferritic alloys, steel, and zircaloy. Many features of ferritic alloys classify as more tolerant materials, such as high resistance to steam oxidation, reduced hydrogen release, and longer coping time. But the neutron penalty must reduce cladding thickness to let a greater fuel volume. Both ferritic alloys and austenitic steel show higher corrosion resistance, also avoiding hydrogen releases. FeCrAl provides more resistant corrosion cracking than stainless steel. The properties of steel 348 are comparable to those of FeCrAl alloys. Steel exhibits superior thermal conductivity, linear thermal expansion, and mechanical strength. Both offer similar specific heat, melting points, and densities. The chemical composition of the steel has 66% iron and 19% chromium, compared with Kanthal APMT™, which uses 68.8% iron and 22% chromium. The results found real advantages related to safety risks using ferritic cladding materials.
Assessment of the UO2-graphene composite fuel proposed for nuclear reactors
2021 - GOMES, DANIEL de S.
Analysis of chromium oxide doped pellet for LWRs using FRAPCON code
2021 - GOMES, DANIEL de S.
Advanced nuclear fuels based on thorium mixed oxides
2021 - GOMES, DANIEL de S.