JOSE ANTONIO SENEDA

JOSE ANTONIO SENEDA

(Fonte: Lattes)

Resumo

Possui Doutorado em Tecnologia Nuclear pela Universidade de São Paulo (2006). Atualmente é Professor Colaborador do Instituto de Pesquisas Energéticas e Nucleares-CNEN/SP. Tem experiência na área de Engenharia Química e Nuclear, atuando principalmente em P&D&E de processos de troca iônica para separação, recuperação e descontaminação de Urânio, Tório e Terras Raras, além de Energias Renováveis com alinhamento em Gestão de Risco destas áreas.(Texto extraído do Currículo Lattes em 14 out. 2021)

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Use of the ion exchange technique for purification of lithium carbonate for nuclear industry
2021 - ANDRADE, MARIANA N.; OLIVEIRA, GLAUCIA C.; CONTRIM, MARYCEL E.B.; SENEDA, JOSE A.; BUSTILLOS, OSCAR V.
Preparation of neodymium acetate for use in nuclear area and nanotechnology
2017 - QUEIROZ, C.A.S.; PEDREIRA FILHO, W.R.; SENEDA, J.A.
Neodymium and its compounds are being increasingly applied in the manufacture of new materials. In nuclear area neodymium isotopes are used in a variety of scientific applications. Nd-142 has been used to produce short-lived Tm and Yb isotopes. Nd-146 has been suggested to produce Pm-147 and Nd-150 has been used to study double beta decay. Due to the several modern applications using nanomaterials, more and more highly rare earth compounds have been demanded. The researches at IPEN uses the experience gained in rare earth separation for the preparation of some pure acetates, purity > 99.9% for application in nanotechnology research. A simple and economical chemical process to obtaining neodymium acetate of high purity is studied. The raw material in the form of mixed rare earths carbonate comes from Brazilian monazite. It is used the technique of strong cationic exchange resin, proper to water treatment, to the neodymium's fractionation and it is achieved a purity of 99.9% in Nd2O3 and yield greater than or equal 80%, with the elution of rare earths by EDTA solution in pH controlled. The complex of EDTA-neodymium is transformed in neodymium oxide, subsequently the oxide is dissolved in acetic acid to obtain the neodymium acetate. The solid salt was characterized via molecular absorption spectrophotometry, mass spectrometry, thermal analysis, chemical analysis and X ray diffraction. In summary the analytical data collected allowed to conclude that the stoichiometric formula for the neodymium acetate prepared is Nd(CH 3COOH)3.1.5H2O.
Applications of lithium in nuclear energy
2017 - OLIVEIRA, GLAUCIA A.C. de; BUSTILLOS, JOSE O.V.; FERREIRA, JOAO C.; BERGAMASCHI, VANDERLEI S.; MORAES, RAFAELI M. de; GIMENEZ, MAISE P.; MIYAMOTO, FLAVIA K.; SENEDA, JOSE A.
Lithium is a material of great interest in the world, it is found in different minerals on Earth's crust (spodumene, lepidolite, amblygonite and petalite) also in salt pans. This element belongs to alkaline group and has two natural isotopes: Li-6 and Li-7. In the nuclear field, lithium isotopes are used for different purposes. The Li-6 is applied in the production of energy, because its section of shock is larger than the other isotope. The Li-7 regulates the pH in refrigerant material in the primary circuits of the Pressurized Water Nuclear Reactor (PWR). In nuclear reactor, lithium is used as a heat transfer due its boiling temperature (1342°C), making it an excellent thermal conductor. However, to reach all these applications, lithium must have high purity (> 99%). The main processes to reach a high purity level of lithium employee a combination of solvent extraction and ion exchange process, to obtain its salts or ending with chemical electrolysis of its chlorides to obtain its pure metal. This work presents a review of new applications of Lithium in Nuclear Energy and its purification and enrichment processes.
Study on radiogenic lead recovery from residues in thorium facilities using ion exhange anb electrochemical process
2007 - SENEDA, JOSE A.; FORBICINI, CHRISTINA A.L.G. de O.; QUEIROZ, CARLOS A. da S.; VASCONCELLOS, MARI E. de; FORBICINI, SERGIO; RIZZO, SORAYA M. da R.; ABRAO, ALCIDIO
Recovery of uranium from the filtrate of 'ammonium diuranate' prepared from uranium hexafluoride
2000 - SENEDA, J.A.; FIGUEIREDO, F.F.; ABRAO, A.; CARVALHO, F.M.S.; FRAJNDLICH, E.U.C.
Recovery of uranium from the filtrate of 'ammonium diuranate' prepared from uranium hexafluoride
2001 - SENEDA, J.A.; FIGUEIREDO, F.F.; ABRAO, A.; CARVALHO, F.M.S.; FRAJNDLICH, E.U.C.
The hydrolysis of uranium hexafluoride and its conversion to 'ammonium diuranate' yields an alkaline solution containing ammonium fluoride and low concentrations of uranium. The recovery of the uranium has the advantage of saving this valuable metal and the avoidance of unacceptable discarding the above mentioned solution to the environment. The recovery of uranium(VI) is based on its complex with the excess of fluoride in the solution and its adsorption on to an anionic ion-exchange resin. The 'ammonium diuranate' filtrate has an approximate concentration of 130 mg l-1 of uranium and 20 g l-1 of ammonium fluoride. An effective separation and recovery of the uranyl fluoride was achieved by choosing a suitable pH and flow rate of the uranium-bearing solution on to the resin column. The effluent ammonium fluoride will be recovered as well. Uranium fluoride adsorbed by the ion-exchanger is then transformed into the corresponding uranyl tricarbonate complex by percolation of a dilute ammonium carbonate solution. Finally, the free fluoride uranium carbonate is eluted from the resin with a more concentrate ammonium carbonate solution. The eluate now can be storage to be precipitated as ammonium uranyl carbonate (AUC). The loading and elution of uranium(VI) on to the ion-exchange column was followed up by a fluorescence test. A flux prepared using a mixture of sodium and potassium carbonate and ammonium fluoride was used for the uranium fluorescence test (ultraviolet lamp). Based on the successful recovery of the uranium on a laboratory scale this process is being considered for use with enriched uranium.
Study on radiogenic lead recovery from residues in thorium facilities using ion exchange and electrochemical process
2010 - SENEDA, J.A.; FORBICINI, C.A.L.G. de O.; QUEIROZ, C.A. da S.; VASCONCELLOS, M.E. de; FORBICINI, S.; RIZZO, S.M. da R.; SALVADOR, VERA L.R.; ABRÃO, A.
Brazil has one of the biggest mineral thorium reserves, enabling the use of this material in its nuclear reactors. Consequently, this cycle of the fuel would need an initial purification stage of the natural thorium, generating residues from chemical treatment. This work provides operational parameters for the recovery of existing values in these residues, especially the radiogenic lead, that is a radioisotope of thorium decay chain, using ionic exchange technique associated to the electrochemical one. The treatment by ionic exchange in anionic resin and hydrochloric acid medium, provides about 33.4% of radiogenic lead. At the electrochemical process, lead was reduced to a metal in nitric acid medium, presenting a recovery of 98%. The electrochemical process presents an increase in the cost, nevertheless the technological importance of the radiogenic lead in the production of new elements, besides being a strategic material, justifies its use.