SABINE NEUSATZ GUILHEN

SABINE NEUSATZ GUILHEN

(Fonte: Lattes)

Resumo

Possui graduação em Química com atribuições Tecnológicas e Biotecnológicas pelo Instituto de Química da Universidade de São Paulo (2005), mestrado (2009) e doutorado (2018) em Tecnologia Nuclear (Materiais) pelo Instituto de Pesquisas Energéticas e Nucleares (IPEN), Universidade de São Paulo. Tem experiência em Química Analítica com ênfase em Análise de Traços, atuando principalmente no desenvolvimento de métodos analíticos empregando técnicas espectrofotométricas (AAS, ICP OES e ICP-MS) para caracterização de amostras ambientais, arqueológicas, biológicas, forenses e nucleares. Atualmente, ocupa o cargo de Tecnologista em "Caracterização Química" no Centro de Química e Meio Ambiente (CQMA) do IPEN (CNEN/SP), onde desempenha atividades de pesquisa e desenvolvimento tecnológico em atendimento às demandas institucionais ligadas ao Ciclo do Combustível Nuclear e aos Programas de Pesquisa de caráter multidisciplinar, em apoio a projetos de Inovação Tecnológica e ao Programa de Pós-Graduação do IPEN/USP. Além disso, atua na geração de produtos tecnológicos e no desenvolvimento de materiais adsorventes de baixo custo e alto valor agregado visando o aproveitamento de materiais e resíduos naturais e/ou renováveis no tratamento de efluentes e rejeitos. (Texto extraído do Currículo Lattes em 4 maio 2023)

Lattes

Página do item completo

Resultados de Busca

Agora exibindo 1 - 5 de 5

X-ray fluorescence spectrometry
2023 - SCAPIN, M.A.; TESSARI-ZAMPIERIA, M.C.; GUILHENA, S.N.; COTRIM, M.E.B.
This study aims to develop reliable analytical methodology that is, cost-effective, and requires minimal sample quantity to quantify uranium content in nuclear waste and others. The Energy Dispersive X-ray Fluorescence Spectrometry (EDXRF) technique was used, and a rigorous comparison was made between the fundamental parameters (FP) method and the empirical (EMP) method. Statistical evaluation of results demonstrated that the FP method showed a satisfactory level of confidence for precision and limit of quantification.
Brazilian clays for environmental solutions applied to radioactive waste management
2021 - MACHADO, G.G.; KRUPSKAYA, V.V.; ZAKUSIN, S.V.; HARADA, J.; VICENTE, R.; SOUZA, R.P.; ARAUJO, L.G.; MONTALVAN, E.T.; ESPINOSA, D.C.R.; KAHN, H.; GUILHEN, S.N.
Clayey materials have been adopted in most nuclear waste producing countries as a key constituent in engineered barrier systems for final disposal facilities at all levels of radioactive wastes (LILW-SL, LILW-LL, and HLW). The following study presents a thorough characterization upon five Brazilian clay-rich deposits, mostly smectite bearing clays, aiming to evaluate their expected performance as barrier under the conditions associated to a Low and Intermediate Level Waste Repository; being the former a matter of national strategic interest. Samples coming from the Brazilian states of Paraná, Bahia, Paraíba, and Maranhão were treated and analyzed by means of X-Ray diffraction as main technique. Other techniques such as FTIR, LALLS, XRF, and SEM-EDS, were performed in order to establish the mineralogical composition, particle size distribution, and chemical composition. Moreover, several standard clay treatments over the <1 μm size fraction were carried out to reveal information regarding layer charge, major interlayer cations, unit formula and other crystal features of smectite species present in a mineralogical assembly, aiming to provide information for the construction of a molecular model over which would be realistic to simulate the diffusion of radionuclides. Results obtained on 133Cs adsorption experiments indicate that mineralogical composition would probably be the single most influential factor controlling transport capacity of positively charged radionuclides in the current setup. The composition is especially expressed in terms of smectite contents, favoring montmorillonite rich materials containing majorly Na+ as compensating cation in interlayer position. All tested samples can be considered as suitable candidates to be used in the design of final destination storage for nuclear waste. Thus, efficiency on 133Cs adsorption trials also indicate that these materials could have potential uses as sorptive matrices (Sorbents) for water treatment of radionuclide polluted waters such as TENORM waste waters. However, these trends are yet to be contrasted against hydraulic conductivity measurements and swelling pressure in order to have a more comprehensive perspective of this clayey prospects as barrier enhanced layer; aligned to the multilayer barrier system approach for nuclear waste management.
Direct determination of aluminum in low-enriched UAlx targets (UAlx-Al) by ICP OES
2021 - GUILHEN, S.N.; SOUZA, A.L.; COTRIM, M.E.B.; PIRES, M.A.F.
The production of molybdenum-99 (99Mo) using low-enriched uranium targets (< 20% 235U) dispersed in aluminum (UAlx) is a very promising strategy towards the independence in 99Mo local production. A thorough control must be performed to ensure that these targets meet the regulatory requirements to achieve the expected efficiency in the reactor. The determination of the targets’ composition is of high interest, because the distribution of Al in different phases may have an impact on the U concentration. Among the techniques used for this purpose, inductively coupled plasma optical emission spectrometry (ICP OES) stands out because of its high sensitivity and precision, allowing for simultaneous determination of several elements in a variety of samples and matrices. However, because U exhibits a complex emission spectrum, spectral interferences are prone to affect the analysis of Al, calling for time consuming preparation steps to remove the U from the matrix. This study proposes a method of direct determination of Al in UAlx targets through the selection of specific emission lines enabled by the evaluation of the associated interferences on the recovery values.
Influence of adsorption parameters on uranium adsorption capability by biochar derived from macauba coconut residue
2019 - GUILHEN, S.N.; COLETI, J.; TENORIO, J.A.S.; FUNGARO, D.A.
Biochar (BC) is a carbon-rich product obtained when biomass is thermally decomposed at relatively low temperatures (under 700ºC) and limited supply of oxygen in a process called pyrolysis. Because of its porous structure, charged sur-face and surface functional groups, BC exhibits a great potential as an adsorbent. Its characteristics strongly depend on the feedstock and the pyrolysis conditions. The aim of this study was to evaluate the adsorption potential for the remov-al of uranium, U(VI) from aqueous solutions using BC obtained through slow pyrolysis of the macauba coconut endo-carp. The influence of parameters such as pH, sorbent dose and initial concentration on the adsorption of U(VI) was investigated. The BC obtained at 350 °C (BC350) presented a removal percentage of approx. 80 %, demonstrating its applicability for the treatment of uranium contaminated aqueous solutions.
Application of bias correction methods to improve U3Si2 sample preparation for quantitative analysis by WDXRF
2019 - SCAPIN, M.A.; GUILHEN, S.N.; AZEVEDO, L.C.; COTRIN, M.E.B.; PIRES, M.A.F.
The determination of silicon (Si), total uranium (U) and impurities in uranium-silicide (U3Si2) samples by wavelength dispersion X-ray fluorescence technique (WDXRF) has been already validated and is currently implemented at IPEN’s X-Ray Fluorescence Laboratory (IPEN-CNEN/SP) in São Paulo, Brazil. Sample preparation requires the use of approx-imately 3 g of H3BO3 as sample holder and 1.8 g of U3Si2. However, because boron is a neutron absorber, this proce-dure precludes the recovery of U3Si2 from the samples, preventing its use as nuclear fuel. Consequently, a significant amount of uranium is wasted in this process. An estimated average of 15 samples per month is expected to be analyzed by WDXRF, resulting in approx. 320 g of U3Si2 that wouldn’t return to the nuclear fuel cycle. The purpose of this paper is to present a new preparation method, replacing H3BO3 by cellulose acetate {[C6H7O2(OH)3-m(OOCCH3)m], m = 0~3}, thus enabling the recovery of the boron-free U3Si2 from the samples. The results demonstrate that the suggested sample preparation approach is statistically satisfactory, allowing the optimization of the procedure.