A fibra de coco tem sido usada como um adsorvente alternativo e de baixo custo na remoção de diversos metais pesados. A biossorção é um processo que tem alcançado grande importância nas últimas décadas no tratamento de efluentes e de rejeitos radioativos. Este estudo apresenta a eficiência de remoção dos íons 133Cs e 241Am de soluções aquosas utilizando-se a biomassa bruta e ativada. Os estudos foram realizados em batelada e os parâmetros analisados foram: os efeitos do pH e da concentração da solução, tamanho de partícula do biomassa e tempo de contato. Os modelos de isotermas de Langmuir e Freundlich foram aplicados, bem como os modelos cinéticos de ordem de reação. A cinética que melhor representa o processo de adsorção dos íons estudados foi o modelo de pseudo-segunda ordem. O modelo de isotermas que se ajusta ao processo de adsorção do 133Cs e do 241Am é o de Freundlich. Verificou-se também que a melhor condição de remoção para o 241Am foi de cerca de 94% a partir de 30 minutos tanto para a biomassa bruta quanto para a ativada ao passo que o 133Cs foi de 75% a partir de 40 minutos com a biomassa ativada. Os resultados indicaram que a fibra de coco pode ser uma alternativa de tratamento de rejeitos radioativos líquidos que contenham, em sua composição, estes radionuclídeos. / Sorption is one of the most studied methods to reduce the volume of radioactive waste streams. The coconut fibers have been used as an inexpensive alternative adsorbent to removal of several heavy metals from liquid wastes. The aim of this work is to evaluate the potential of coconut fiber, raw and chemically modified form, to remove 133Cs and 241Am from aqueous solutions. The studies were performed at batch conditions and the following parameters were analyzed: contact time, pH and concentration of ions in aqueous solution and the particle size. The Langmuir and Freundlich models were studied to describe the isotherms and the kinetic process was studied using the pseudofirst and pseudo-second orders. For 241Am and 133Cs, the Freundlich isotherm described better the adsorption process. For 241Am, the best removal condition was achieved at 30 minutes with both biomasses. For 133Cs the best condition was achieved from at 40 minutes with modified biomass. Removal kinetics for both ions was fitted by pseudo-second order model. The results indicate that coconut fiber can be used as an alternative sorbent in the treatment of liquid radioactive wastes containing these radionuclides.

Os pára-raios radioativos foram fabricados no Brasil até 1989, quando a Comissão Nacional de Energia Nuclear (CNEN) suspendeu a concessão de uso de material radioativo nesses artefatos. Desde então, o pára-raios radioativo tem sido substituído por outro, do tipo Franklin, e recolhido como rejeito radioativo. Entretanto, apenas 23 % do total fabricado no país foram entregues à CNEN. Esta situação é preocupante, pois a chance, desses artefatos serem descartados como resíduo comum e chegarem a lixões, é grande, uma vez que, segundo dados do Instituto Brasileiro de Geografia e Estatística (IBGE), em 2000, 63,6 % dos municípios brasileiros dispunham o resíduo nesses locais. Além disso, o amerício, o radionuclídeo mais empregado, é classificado como sendo um elemento de alta toxicidade, quando ingerido ou inalado. No presente trabalho, foram realizados experimentos de migração de Am-241 em lisímetros, com o objetivo de se avaliar o risco de contaminação provocada por pára-raios radioativos descartados como resíduo comum. Fontes radioativas removidas de pára-raios foram inseridas em lisímetros preenchidos com resíduo orgânico, coletado no restaurante do Instituto de Pesquisas Energéticas e Nucleares, IPEN-CNEN/SP, e chorume gerado foi periodicamente analisado para determinar suas características como pH, potencial redox, teor de sólidos e a concentração do material radioativo. O crescimento microbiano também foi avaliado, pelo método de contagem direta do número de unidades formadoras de colônia. A estimativa de risco foi baseada no cálculo de dose para membros do público, sendo a ingestão de água a via mais provável de exposição. O valor obtido foi cerca de 1000 vezes inferior ao limite de dose anual estabelecido, pela Comissão Internacional de Proteção Radiológica (ICRP), demonstrando que o risco de contaminação provocado pelo descarte de pára-raios em lixões é baixo. / Radioactive lightning rods were manufactured in Brazil until 1989, when the licenses for using radioactive sources in these products were lifted by the national nuclear authority. Since then, radioactive devices have been replaced by Franklin type one and collected as radioactive waste. However, only 23 percent of the estimated total number of installed rods was delivered to Brazilian Nuclear Commission (Comissão Nacional de Energia Nuclear – CNEN). This situation is of concern to us as there is a possibility of the rods being discarded as domestic waste, considering that in Brazil, 63.6 percent of the municipal solid waste is disposed at uncontrolled garbage dump, according to Instituto Brasileiro de Geografia e Estatística (IBGE) in 2000. In addition, americium, the most common employed radionuclide, is classified as a high toxicity element, when ingested or inhaled. In the present study, it was performed migration experiments of Am-241 by lysimeter system in order to evaluate the risk of contamination caused by radioactive lightning rods disposed as a common solid waste. Sources removed from lightning rods were placed inside lysimeters filled with organic waste, collected at the restaurant of Instituto de Pesquisas Energéticas e Nucleares, IPEN-CNEN/SP, and the generated leachate was periodically analyzed to determine its characteristics such as pH, redox potential, solid content and concentration of the radioactive material. Microbial growth was also evaluated by counting the number of colony forming units. The equivalent dose to members of the public has been calculated considering the ingestion of drinking water, the most probable mode of exposure. The final result was about 145 times below the effective dose limit of 1 mSv.year-1 for members of the public, established by the International Commission on Radiological Protection (ICRP), demonstrating that the risk caused by lightning rods disposed at uncontrolled garbage dump is low.

A fibra de coco tem sido usada como um adsorvente alternativo e de baixo custo na remoção de diversos metais pesados. A biossorção é um processo que tem alcançado grande importância nas últimas décadas no tratamento de efluentes e de rejeitos radioativos. Este estudo apresenta a eficiência de remoção dos íons 133Cs e 241Am de soluções aquosas utilizando-se a biomassa bruta e ativada. Os estudos foram realizados em batelada e os parâmetros analisados foram: os efeitos do pH e da concentração da solução, tamanho de partícula do biomassa e tempo de contato. Os modelos de isotermas de Langmuir e Freundlich foram aplicados, bem como os modelos cinéticos de ordem de reação. A cinética que melhor representa o processo de adsorção dos íons estudados foi o modelo de pseudo-segunda ordem. O modelo de isotermas que se ajusta ao processo de adsorção do 133Cs e do 241Am é o de Freundlich. Verificou-se também que a melhor condição de remoção para o 241Am foi de cerca de 94% a partir de 30 minutos tanto para a biomassa bruta quanto para a ativada ao passo que o 133Cs foi de 75% a partir de 40 minutos com a biomassa ativada. Os resultados indicaram que a fibra de coco pode ser uma alternativa de tratamento de rejeitos radioativos líquidos que contenham, em sua composição, estes radionuclídeos. / Sorption is one of the most studied methods to reduce the volume of radioactive waste streams. The coconut fibers have been used as an inexpensive alternative adsorbent to removal of several heavy metals from liquid wastes. The aim of this work is to evaluate the potential of coconut fiber, raw and chemically modified form, to remove 133Cs and 241Am from aqueous solutions. The studies were performed at batch conditions and the following parameters were analyzed: contact time, pH and concentration of ions in aqueous solution and the particle size. The Langmuir and Freundlich models were studied to describe the isotherms and the kinetic process was studied using the pseudofirst and pseudo-second orders. For 241Am and 133Cs, the Freundlich isotherm described better the adsorption process. For 241Am, the best removal condition was achieved at 30 minutes with both biomasses. For 133Cs the best condition was achieved from at 40 minutes with modified biomass. Removal kinetics for both ions was fitted by pseudo-second order model. The results indicate that coconut fiber can be used as an alternative sorbent in the treatment of liquid radioactive wastes containing these radionuclides.

Os pára-raios radioativos foram fabricados no Brasil até 1989, quando a Comissão Nacional de Energia Nuclear (CNEN) suspendeu a concessão de uso de material radioativo nesses artefatos. Desde então, o pára-raios radioativo tem sido substituído por outro, do tipo Franklin, e recolhido como rejeito radioativo. Entretanto, apenas 23 % do total fabricado no país foram entregues à CNEN. Esta situação é preocupante, pois a chance, desses artefatos serem descartados como resíduo comum e chegarem a lixões, é grande, uma vez que, segundo dados do Instituto Brasileiro de Geografia e Estatística (IBGE), em 2000, 63,6 % dos municípios brasileiros dispunham o resíduo nesses locais. Além disso, o amerício, o radionuclídeo mais empregado, é classificado como sendo um elemento de alta toxicidade, quando ingerido ou inalado. No presente trabalho, foram realizados experimentos de migração de Am-241 em lisímetros, com o objetivo de se avaliar o risco de contaminação provocada por pára-raios radioativos descartados como resíduo comum. Fontes radioativas removidas de pára-raios foram inseridas em lisímetros preenchidos com resíduo orgânico, coletado no restaurante do Instituto de Pesquisas Energéticas e Nucleares, IPEN-CNEN/SP, e chorume gerado foi periodicamente analisado para determinar suas características como pH, potencial redox, teor de sólidos e a concentração do material radioativo. O crescimento microbiano também foi avaliado, pelo método de contagem direta do número de unidades formadoras de colônia. A estimativa de risco foi baseada no cálculo de dose para membros do público, sendo a ingestão de água a via mais provável de exposição. O valor obtido foi cerca de 1000 vezes inferior ao limite de dose anual estabelecido, pela Comissão Internacional de Proteção Radiológica (ICRP), demonstrando que o risco de contaminação provocado pelo descarte de pára-raios em lixões é baixo. / Radioactive lightning rods were manufactured in Brazil until 1989, when the licenses for using radioactive sources in these products were lifted by the national nuclear authority. Since then, radioactive devices have been replaced by Franklin type one and collected as radioactive waste. However, only 23 percent of the estimated total number of installed rods was delivered to Brazilian Nuclear Commission (Comissão Nacional de Energia Nuclear – CNEN). This situation is of concern to us as there is a possibility of the rods being discarded as domestic waste, considering that in Brazil, 63.6 percent of the municipal solid waste is disposed at uncontrolled garbage dump, according to Instituto Brasileiro de Geografia e Estatística (IBGE) in 2000. In addition, americium, the most common employed radionuclide, is classified as a high toxicity element, when ingested or inhaled. In the present study, it was performed migration experiments of Am-241 by lysimeter system in order to evaluate the risk of contamination caused by radioactive lightning rods disposed as a common solid waste. Sources removed from lightning rods were placed inside lysimeters filled with organic waste, collected at the restaurant of Instituto de Pesquisas Energéticas e Nucleares, IPEN-CNEN/SP, and the generated leachate was periodically analyzed to determine its characteristics such as pH, redox potential, solid content and concentration of the radioactive material. Microbial growth was also evaluated by counting the number of colony forming units. The equivalent dose to members of the public has been calculated considering the ingestion of drinking water, the most probable mode of exposure. The final result was about 145 times below the effective dose limit of 1 mSv.year-1 for members of the public, established by the International Commission on Radiological Protection (ICRP), demonstrating that the risk caused by lightning rods disposed at uncontrolled garbage dump is low.

Plutonium Uranium Reduction EXtraction (PUREX) technology is a solvent extraction process used to recover plutonium and uranium from spent nuclear fuel. The solvent system is composed of an aqueous nitric acid phase in contact with an organic phase made up of tributyl phosphate in an organic diluent. During the separation process, the PUREX solvent system is subject to an intense multi-component radiation field (gamma rays, alpha particles, beta particles, neutrons, and fission fragments) rendering it susceptible to radiolytic degradation, which reduces its performance. Despite the PUREX process being used for over sixty years, a complete quantitative mechanistic understanding of the radiolytic degradation processes is not available. Nitrous acid is the most significant radiolytic degradation product of nitric acid, especially as its chemical and physical properties alter the formulation of the PUREX solvent system. Furthermore, nitrous acid exhibits complex redox relationships with a number of actinides, with plutonium being of greatest concern to the performance of the PUREX process. A combination of experimental and computational (stochastic and deterministic) techniques have been used to investigate the radiolysis of the PUREX solvent system's aqueous phase, specifically the radiolytic formation of nitrous acid, and its conjugate base nitrite, as a function of solvent system formulation, absorbed dose (up to 1.7 kGy), and radiation quality (cobalt-60 gamma rays and alpha particles from plutonium and americium alpha decay). The research presented in this thesis focuses on: (i) the experimental radiation chemistry of solutions of nitric acid and sodium nitrate over the range of concentrations 1 × 10-3 to 6 mol dm-3, and (ii) the development of a multi-scale modelling approach for evaluating the radiolysis of aqueous systems in terms of reaction mechanisms. The experimental and modelling studies provide insight into the radiation chemistry of the PUREX solvent system's aqueous phase, mechanistically demonstrating how the radiation chemical yield of nitrous acid and nitrite is dependent upon the interplay between non-homogeneous radiation track chemistry and secondary bulk homogeneous chemistry. This interplay is influenced by low pH, the presence of chemical scavengers and redox active metal ions, and radiation quality. These findings will act as a benchmark for the development of advanced reprocessing schemes, which must seriously consider how modifications in solvent system formulation and fuel composition may affect this dynamic interplay, and ultimately the generation of secondary highly active liquid waste.

In this thesis, the feasibility to use sodium cooled fast reactors loaded with MOX, metallic and nitride fuels for efficient transmutation of americium is investigated by performing transient analysis for cases with different americium contents in fuels, using safety parameters obtained with the SERPENT Monte Carlo code. It was then demonstrated that there is no solid limit for the Am introduction into oxide, metallic and nitride fuels that were loaded into sodium fast reactors. Instead, higher Am contents could be permitted if specific levels of power penalty were accepted. Transient analysis of a new Accelerator Driven System design with higher neutron source efficiency than the reference EFIT-400 design, was also performed. Based on simulation results, the suggested ADS design was proved to survive the full set of transients, preserving 130 K margin to cladding rupture during the most limiting transient. After comparing Am transmutation performances in SFRs and the suggested ADS, it can be concluded that: 1. Nitride fuel could provide the highest Am transmutation efficiency, when loaded into SFRs; 2. One SFR loaded with nitride fuel is sufficient to transmute Am inventory produced by more than 15 commercial LWRs within the same time period, which is three times higher than the supporting ratio reported for the suggested ADS; 3. The total fraction of ADS power in the power park is half of cases for critical reactors. / QC 20120201

This dissertation focuses on key intergroup and intragroup separations found in the back end of the nuclear fuel cycle, specifically americium from lanthanides and americium from other actinides, most importantly americium from curium. Our goal is to implement a liquid-solid separation process to reduce waste and risk of contamination by the development of metal(IV) phosphate phosphonate inorganic organic hybrid ion exchange materials with the ideal formula of M(O6P2C6H4)0.5(O3POA) * nH2O, where M = Zr or Sn, A = H or Na. These materials have previously shown to have high affinity for Ln, this work will expand on the previous studies and provide methods for the above target separation, exploiting oxidation state and ion charge to drive the separation process. The optimum hydrothermal reaction conditions were determined by adjusting parameters such as reaction temperature and time, as well as the phosphonate to phosphate (pillarto-spacer) ligands ratio. Following these results four bulk syntheses were performed and their ion exchange properties were thoroughly examined. Techniques such as inductively coupled mass spectrometry and liquid scintillation counting were used to determine the affinity of the materials towards Na+, Cs+, Ca2+, Sr2+, Ni2+, Nd3+, Sm3+, Ho3+, Yb3+, NpO2+, Pu4+, PuO22+, Am3+, AmO2+, and Cm3+. Separation factors in the thousands have been observed for intergroup separations of the Ln from the alkali, alkaline earth, and low valent transition metals. A new method for Am oxidation was developed, which employed Na2S2O8 as the oxidizing agent and Ca(OCl)2 as the stabilizing agent for AmO2+ synthesis. Separation factors of 30-60 for Nd3+ and Eu3+ from AmO2+, as well as 20 for Cm3+ from AmO2+ were observed at pH 2. The work herein shows that a liquid-solid separation can be carried out for these difficult separations by means of oxidation and ion exchange.

There are a variety of uses for reference neutron fields including detector response and dosimeter studies. The Georgia Institute of Technology has a 252Cf spontaneous fission source and an AmBe (α, n) source available for use in its research programs. In addition, it has iron, lead, beryllium, tantalum, heavy water, and polyethylene spheres to modify the neutron energy distributions from these neutron sources. This research characterized the neutron leakage spectra from the source inside spherical shells using a Bonner sphere spectrometer. All the neutron fields measured were also computed with a Monte Carlo code to determine the neutron fluence rate and ambient dose equivalent rate. The comparison of experimental data and calculations are used to provide further insight into the neutron spectra as modified by the spheres. The characterization of these modified sources will provide data to assist in using the resulting neutron fields in other research activities. To measure each neutron field combination, one of the two sources was placed in the center of an attenuating sphere. The neutron field was first measured at a variety of source-to-detector distances with a Bonner Sphere System. The spectrometer measurements, specifically the count rates of the different Bonner spheres, as a function of distance from the source is fitted to obtain corrections for room-scatter and air-scatter of neutrons using the Eisenhauer, Schwartz, and Johnson method. Using these corrections, the count rates free of room return is obtained at 1 m from the source and unfolded using the BUMS software to obtain the reported fluence and dose equivalent rates. These results are compared to those generated by the Monte Carlo Neutral Particle (MCNP) code. Models were made in MCNP for each of the source and moderating sphere combinations. The neutron fluence and dose rates were tallied during the MCNP simulation. The unfolded experimental data and the MCNP calculations showed good agreement for most of source-attenuating sphere combinations, thereby reinforcing the experimental results.

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