ELITA FONTENELE URANO DE CARVALHO

ELITA FONTENELE URANO DE CARVALHO

(Fonte: Lattes)

ORCID

ScopusID

ResearcherID

Resumo

Graduation at Química Industrial from Universidade Federal do Ceará (1978), master's at Nuclear Engineering from Universidade de São Paulo (1992) and doctorate at Nuclear Engineering from Universidade de São Paulo (2004). Has experience in Nuclear Engineering, focusing on Conversion, Enrichment and Manufacture of Nuclear Fuel, acting on the following subjects: combustivel nuclear, fluoreto, tratamento de efluentes, veneno queimavel and meio ambiente. (Text obtained from the Currículo Lattes on October 8th 2021)

Possui graduação em Química Industrial pela Universidade Federal do Ceará (1978), mestrado em Tecnologia Nuclear pela Universidade de São Paulo (1992) e doutorado em Tecnologia Nuclear pela Universidade de São Paulo (2004). Pesquisador do Instituto de Pesquisas Energéticas e Nucleares da Comissão Nacional de Energia Nuclear. Experiência na área de Engenharia de Materiais e Química com ênfase em Conversão, Enriquecimento, Fabricação de Combustível Nuclear, Tratamento de efluentes radioativos e convencionais e reaproveitamento de resíduos industriais e técnicas de caracterização química de materiais. Membro do Instituto Nacional de Tecnologia- INCT para Reatores Nucleares Inovadores. Autor de capítulo de livro intitulado "Radioisotopes: Applications in Physical Sciences, 2011 ISBN: 9789533075105. Título do capítulo: Research Reactor Fuel Fabrication to Produce Radioisotopes. Professor de pós- graduação da Universidade São Paulo nas áreas de caracterização de materiais e de combustível nuclear. Professor visitante na Escola Politécnica da Universidade de São Paulo - modulo I e II de processamento de combustivel nuclear. Bolsista de Produtividade Desen. Tec. e Extensão Inovadora 2 (Texto extraído do Currículo Lattes em 08 out. 2021)

Página do item completo

Resultados de Busca

Agora exibindo 1 - 10 de 161

Manufacturing high-uranium-loaded dispersion fuel plates in Brazil
2024 - DURAZZO, MICHELANGELO; SOUZA, JOSE A.B.; CARVALHO, ELITA F.U. de; RESTIVO, THOMAZ A.G.; GENEZINI, FREDERICO A.; LEAL NETO, RICARDO M.
The Nuclear and Energy Research Institute (IPEN-CNEN/SP) has developed and made available for routine production the technology for manufacturing dispersion-type fuel elements for research reactors. However, the fuel produced is limited to a uranium loading of 2.3 gU/cm3 (U3O8) or 3.0 gU/cm3 (U3Si2). To reduce Brazil’s dependence on foreign sources of Mo-99, the Brazilian government plans to construct a new research reactor, the 30 MW open pool Brazilian Multipurpose Reactor (RMB), which will mainly produce domestic Mo-99. Low-enriched uranium fuel will be used in the RMB, and increasing uranium loading will be important to increase the reactor core’s reactivity and fuel life. Uranium loadings of 3.2 gU/cm3 for the U3O8-Al and 4.8 gU/cm3 for the U3Si2-Al are considered the technological limit and have been well demonstrated worldwide. This work aimed to study the manufacturing process of these two highly uranium-loaded dispersion fuels and redefine current procedures. Additionally, UMo-Al dispersion fuel has been extensively studied globally and is likely to be the next commercially available technology. This new fuel utilizes a dispersion of UMo alloy with 7–10 wt% Mo, resulting in a uranium loading between 6 and 8 gU/cm3. We also studied this fuel type for potential use in the RMB research reactor. This work outlines the primary procedures for manufacturing these three types of fuels and the necessary adjustments to IPEN-CNEN/SP current technology. The manufacturing process proved to be well adapted to these new fuels, requiring only minor modifications. A comparison was made of the microstructures of fuel plate meat using three types of uranium compounds. The microstructures of U3Si2-Al and U10Mo-Al dispersions were found to be adequate, while that of U3O8-Al meat deviated significantly from the concept of an ideal dispersion.
Emprego de rotas eletroquímicas na separação seletiva e de alto rendimento de iodo-131 para aplicação como radiofármaco
2022 - MARINHO, T.C.; CARVALHO, E.F.U.; FERNANDES, V.C.; SANTIAGO, E.I.
O iodo-131 é um radiofármaco emissor de partícula β- , utilizado no tratamento do câncer de tireóide. Os processos de produção do iodo-131 via fissão de urânio-235 ou irradiação de alvos de telúrio, resultam em soluções compostas por diferentes metais, como molibdênio (Mo), telúrio (Te) e rutênio (Ru). O objetivo deste trabalho foi empregar rotas eletroquímicas baseadas na mudança de temperatura e aplicação de potencial para separação de iodo. Inicialmente Mo, Te e Ru foram analisados quanto à sua interferência no processo de separação e o iodo quanto ao potencial em que a reação se processa com maior velocidade. Em seguida, o iodo foi separado, capturado e testado qualitativamente. Os testes indicaram boa captura em 25° C e 40 °C e pouca captura em 60 °C.
Effects of aluminum distearate addition on UO2 sintering and microstructure
2022 - FREITAS, ARTUR C. de; COSTA, DIOGO R.; JARDIM, PAULA M.; LEAL NETO, RICARDO M.; CARVALHO, ELITA F.U. de; DURAZZO, MICHELANGELO
Uranium dioxide (UO2) is widely used as a fuel in commercial nuclear light-water reactors (LWRs). Rigorous control of density, pore, and grain size of UO2 pellets are important prerequisites for fuel performance. Solid lubricants, frequently used in pellets manufacturing, minimize structural defects on compaction such as cracks and end-capping, promoting grain growth during sintering. This work presents and discusses the effects of the aluminum distearate (ADS) addition on the sintering behavior and microstructure of UO2 fuel pellets. UO2 and UO2-0.2wt% ADS pellets were sintered at 1760 °C for 5.7 h for comparison purposes. The results show that the densification rate increases using the solid lubricant, but the shrinkage is lowered by 0.7% due to low homogenization. The average grain size was increased by about 35% during sintering. Based on our results and a literature review, a mechanism for grain growth by aluminum addition is proposed.
Nickel electrodeposition in LEU metal foil annular targets to produce Mo-99
2022 - IANELLI, RICARDO F.; SALIBA-SILVA, ADONIS M.; TAKARA, ERIKI M.; GARCIA NETO, JOSE; SOUZA, JOSE A.B.; CARVALHO, ELITA F.U. de; DURAZZO, MICHELANGELO
The most used production route of Mo-99 is through the fission of U-235 in irradiation targets that are irradiated in research reactors. The annular target is a promisor design since it can incorporate high U-235 quantities, thus increasing the production yield of Mo-99. This target type uses a foil of uranium metal enveloped by a thin nickel foil that acts as a diffusion barrier. The process of uranium enveloping with nickel foil is today done manually. This operation risks the nickel foil breaking up during target assembling. In the present work, we studied the nickel electrodeposition over uranium metal foil surfaces to replace nickel foils. A pre-forming procedure of the uranium metal foil by calendering was developed to facilitate the assembling operation. The electrodeposition was done over the uranium foil pre-conformed in a tubular shape. An automated apparatus for electrodeposition of nickel in uranium tubular-shaped foil was developed. The results showed that the high nickel adherence to uranium metal depends on the proper activation of the uranium surface. Among the activation processes studied, the mechanical activation showed good adhesion of the nickel layer, with a loss of only 0.16% of uranium mass. Homogeneous and regular 12 μm thickness electrodeposited layers over the uranium metal were obtained. This work showed that the process could be used in continuous production technology, such as the production of irradiation targets.
Manufacturing LEU-foil annular target in Brazil
2022 - DURAZZO, MICHELANGELO; SOUZA, JOSE A.B.; IANELLI, RICARDO F.; TAKARA, ERIKI M.; GARCIA NETO, JOSE S.; SALIBA-SILVA, ADONIS M.; CARVALHO, ELITA F.U. de
Molybdenum-99 is the most important isotope because its daughter isotope, technetium-99m, has been the most used medical radioisotope. The primary method used to produce Mo-99 derives from the fission of U-235 incorporated in so-called irradiation targets. Two routes are being developed to make Mo-99 by fissioning with low enriched uranium (LEU) fuel. The first adopts UAlx-Al dispersion plate targets. The second uses uranium metal foil annular targets. The significant advantage of uranium foil targets over UAlx-Al dispersion targets is the high density of uranium metal. This work presents the experience obtained in the development of the uranium metal annular target manufacturing steps. An innovative method to improve the procedure for assembling the uranium foil on the tubular target was presented. The experience attained will help the future production of Mo-99 in Brazil through the target irradiation in the Brazilian Multipurpose Reactor (RMB).
Increasing productivity in the manufacture of UAl2–Al dispersion-plate targets for Mo-99 production
2021 - DURAZZO, MICHELANGELO; CONTURBIA, GIOVANNI L.C.R.; CARVALHO, ELITA F.U. de
Molybdenum-99 is the most important isotope because its daughter isotope, technetium-99 m, has been the most widely used medical radioisotope. The primary method employed to produce Mo-99 derives from the fission of U-235 incorporated in so-called irradiation targets. Pushed by the international Mo-99 crisis that occurred in 2009/2010, Brazil has decided to construct a new research reactor, the Brazilian Multipurpose Reactor (RMB), to produce this vital radioisotope to meet the Brazilian domestic demand. As part of this effort, it has been developed the process for manufacturing the target to be used in the production of Mo-99 by nuclear fission. The low enriched uranium (LEU) aluminide with the predominant phase UAl2 was the starting material. The picture-frame technique was used to clad UAl2–Al briquette with aluminum to obtain plate-type targets. It was developed an innovative method that allows increasing the productivity of this type of target based on multi-core rolling. A thermomechanical treatment was designed to get targets composed basically of a mixture of UAl3/UAl4 that are the required phases for a proper radiochemical dissolution after irradiation. The manufacturing process proved to be suitable for domestic production of targets, fulfilling the specification to produce Mo-99 in the Brazilian Multipurpose Reactor.
Quantification of effluents in the production of nuclear fuel
2021 - SAKAI, M.C.C.B.; RIELLA, H.G.; CARVALHO, E.F.U.
The Brazil with the purpose of becoming self-sufficient in the production of radioisotopes and radioactive sources used in nuclear medicine, agriculture and the environment developed the project of a multipurpose reactor of 30 megawatts of power to attend the national demand. In the Nuclear and Energy Research Institute (IPEN), Nuclear Fuel Center (CCN) is responsible for the manufacture of fuel for the reactor IEA-R1 and, possibly, the multipurpose reactor fuels. In order to meet the demand for the reactors was designed a new manufacturing plant with a maximum capacity of 60 fuel elements, which is 10 nowadays. The increase in production as a consequence will increase the volume of effluents generated. The current concern with the environment it is necessary to draw up a management plan to make the process sustainable, which will result in environmental, economic and social benefits. The fuel production process generates various types of effluent containing uranium or not – being solid, gaseous and liquid with different physical and chemical characteristics. The aim of this work is to follow the process of nuclear fuel production and to identify, quantify and characterize the effluents, especially liquids, to subsequently draw up a management plan and eventually dispose of responsibly in the environment.
A simulation model for capacity planning of nuclear fuel plants for research reactors
2021 - NEGRO, M.L.N.; DURAZZO, M.; MESQUITA, M.A.; SCURO, N.L.; CARVALHO, E.F.U.; ANDRADE, D.A.
The demand for nuclear fuel for research reactors is increasing worldwide. However, some nuclear fuel factories have low production volumes. Literature regarding how to expand the capacity of those facilities in a safe and reliable way is scarce. Thus, the purpose of this work is to propose and validate a conceptual model for increasing the production capacity of such factories. The facilities addressed here are those that produce plate-type fuel elements loaded with LEU U3Si2-Al, which are typically used in nuclear research reactors. Data from a real nuclear fuel plant was collected and applied to the model, thus setting up a case study. Two different strategies, as well as several production scenarios, were conceived for the use of the model. Each scenario experiments with the different possibilities of enlarging capacity. Discrete events simulation was used in order to cover all production scenarios. The tests indicated significant increases in productive capacity, thus showing that the model fully achieved its proposed objective. One of the main conclusions to be highlighted is the model’s effectiveness, which was demonstrated by using the model in two different strategies and obtaining increases in capacity with both of them.
Kinects and factors on chemical dissolution of aluminum alloy AA6061 in NaOH alkaline media
2020 - TAKARA, E.M.; SOUZA, J.B. de; CARVALHO, E.F.U.; SILVA, A.S.
Nuclear Medicine is the Field of science that uses radioactive materials in order to diagnose and treat human body deceases. One of the most used radioisotopes for images diagnose purpose is the metastable technetium-99 (99mTc) because of its low decay half life (6 hours) and energy emission of 140keV that ensures low exposition time with the capacity of generating high quality images. The 99mTc is generated by the molibdenum-99(99Mo) radioactive decay during about 66 hours. The 99Mo is fabricated via nuclear fission of low encriched uranium (LEU) through plate irradiation targets (UAlx). The irradiation target cladding is made of Aluminum alloy AA6061 and its substrate is composed by 235U powder scattered in an AA1050 matrix. In general, studies are made targeting the prevention of corrosion mechanisms but the chemical dissolution in alkaline media, under hot cells, are one of the steps required for the post-processing methods of irradiation targets The time spent after irradiation is an important factor because the half life radioactive decay of the produced radioisotopes is relative short, then the procedures of dissolution, extraction, purify and distribution must be optimized in order to increase efficiency. This work presents a study of the factors impact involved on the chemical dissolution of the cladding aluminum alloys (temperature, NaOH solution concentration and dissolution time) as well as the kinects of the process associating it with the formation and destruction of oxides using electrochemical impedance spectroscopy (EIS) and scanning electron microscopy (SEM). It was found that the involved parameters contribute individually more effective and that there is no relevant association between the factors. Solution temperature showed to be the most influent factor following by exposition time. It was presented a equivalent circuit model which demonstrates the reaction kinects and the growing of passive layers that slow down the process before it turns up into a soluble phase.
The RMB project
2020 - PERROTTA, J.A.; CARVALHO, E.F.U.; DURAZZO, M.; SANTOS, L.R. dos; BAPTISTA, J.A.; SILVA, J.E.R. da; JUNQUEIRA, F.C.; SANTOS, ADIMIR dos; ARAUJO, A.M.V. de; TOMAZELLI, I.
The Brazilian Nuclear Energy Commission (CNEN) decided to construct a new research reactor, named RMB (Brazilian Multipurpose Reactor). It is a 30 MW open pool-type research reactor using low enriched uranium fuel, and several associated facilities and laboratories. To establish an infrastructure for producing fuel assemblies for RMB operation and uranium targets for Mo-99 production, the RMB technical secretary has developed a coordinated project for the fuel cycle management system, putting together the fuel technology actors in Brazil. The goals of this coordinated project were: (i) to have a centrifuge cascade enriching uranium up to 20 wt% with the capacity to supply RMB yearly needs; (ii) to upgraded the CNEN existing infrastructure to produce nuclear fuel assemblies and uranium targets for the RMB yearly needs; (iii) to produce a set of fuel assemblies for a real RMB mockup core at the IPEN/MB-01 Critical Facility of CNEN. The RMB project design incorporates structures, systems and components (SSC) for interim storage of spent fuels for the hole plant lifetime. This paper presents details of the coordinated project that gives support and sustainability to the RMB fuel cycle supply and the spent fuel SSC designed.