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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Pure cerium dioxide preparation for use as spectrochemical standard and analysed by inductively coupled plasma mass spectrometry (SF ICP-MS)
2005 - QUEIROZ, C.A.S.; PEDREIRA, W.R.; ABRAO, A.; ROCHA, S.M.R.; VASCONCELLOS, M.E.; SENEDA, J.A.; FORBICINI, C.A.L.G.O.; BOAVENTURA, G.R.; PIMENTEL, M.M.
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.
Primary standardization of sup(72)Ga
2006 - KOSKINAS, MARINA F.; MOREIRA, DENISE S.; TAKEDA, MAURO N.; SENEDA, JOSE A.; DIAS, MAURO da S.
The activity of 72Ga sources produced by irradiation at the IEA-R1 reactor have been performed in a 4 π β - γ coincidence system by using the extrapolation technique. The measurements were undertaken selecting two windows in the γ-channel, in order to check the consistency of the results. A Monte Carlo calculation was performed in order to predict the behavior of the observed activity as a function of the 4πβ detector efficiency and the results were compared to experimental values. © 2006 Elsevier Ltd. All rights reserved.
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.