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Made available in DSpace on 2014-10-09T12:42:25Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:02:08Z (GMT). No. of bitstreams: 0 / 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. / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Made available in DSpace on 2014-10-09T12:50:03Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:02:47Z (GMT). No. of bitstreams: 0 / Neste trabalho são apresentados os estudos das condições de resposta de um diodo de Si, com estrutura de múltiplos anéis de guarda, na detecção e espectrometria de partículas alfa. Este diodo foi fabricado por meio do processo de implantação iônica (Al/p+/n/n+/Al) em um substrato de Si do tipo n com resistividade de 3 kohm•cm, 300 mícrons de espessura e área útil de 4 mm2. Para usar este diodo como detector, a face n+ deste dispositivo foi polarizada reversamente, o primeiro anel de guarda aterrado e os sinais elétricos extraídos da face p+. Estes sinais eram enviados diretamente a um pré-amplificador desenvolvido em nosso laboratório, baseado no emprego do circuito híbrido A250 da Amptek, seguido da eletrônica nuclear convencional. Os resultados obtidos com este sistema na detecção direta de partículas alfa do Am-241evidenciaram excelente estabilidade de resposta com uma elevada eficiência de detecção (= 100 %). O desempenho deste diodo na espectrometria de partículas alfa foi estudado priorizando-se a influência da tensão de polarização, do ruído eletrônico, da temperatura e da distância fonte-detector na resolução em energia. Os resultados mostraram que a maior contribuição para a deterioração deste parâmetro é devida à espessura da camada morta do diodo (1 mícron). No entanto, mesmo em temperatura ambiente, a resolução medida (FWHM = 18,8 keV) para as partículas alfa de 5485,6 keV (Am-241) é comparável àquelas obtidas com detectores convencionais de barreira de superfície freqüentemente utilizados em espectrometria destas partículas. / Dissertacao (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

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Les isotopes d’américium représentent une part importante des déchets nucléaires à haute activité et à durée de vie longue dans le combustible usé. Parmi les options de retraitement envisagées, sa transmutation en réacteurs à neutrons rapides au sein de pastilles d’oxydes mixtes d’uranium-américium de composition U1-xAmxO2±δ est une option prometteuse qui permettrait de diminuer l’empreinte écologique des sites d’entreposage des déchets ultimes. Dans ce contexte, cette thèse est consacrée à l’étude de ces composés sur une large gamme de compositions (7,5 %mol ≤ Am/(U+Am) ≤ 70 %mol) focalisée sur leur fabrication à partir d’oxydes simples et l’évaluation de leur stabilités structurales, thermodynamique et sous auto-irradiation. Les résultats mettent en évidence l’influence majeure de la réduction de l’américium en Am+III, aussi bien dans les mécanismes de formation de la solution solide U1-xAmxO2±δ par voie solide que dans la stabilisation de cations d’uranium oxydés, accompagnés de la formation de défauts dans le sous-réseau d’oxygène tels que des lacunes et des clusters cuboctaédriques. Par ailleurs, les données acquises concernant la stabilité en température des composés U1-xAmxO2±δ (existence d’une lacune de miscibilité, comportement en vaporisation) ont été comparées à des calculs basés sur un nouveau modèle thermodynamique décrivant le système ternaire U-Am-O. Enfin, les effets structuraux de l’auto-irradiation α dans les composés U1-xAmxO2±δ ont été analysés par DRX, XAS et MET, permettant d’étudier l’influence de la teneur en américium sur le gonflement structural et de décrire l’évolution des défauts structuraux radio-induits. / Americium isotopes represent a significant part of high-level and long-lived nuclear waste in spent fuels. Among the envisaged reprocessing scenarios, their transmutation in fast neutron reactors using uranium-americium mixed-oxide pellets (U1-xAmxO2±δ) is a promising option which would help decrease the ecological footprint of ultimate waste repository sites. In this context, this thesis is dedicated to the study of such compounds over a wide range of americium contents (7.5 at.% ≤ Am/(U+Am) ≤ 70 at.%), with an emphasis on their fabrication from single-oxide precursors and the assessment of their structural and thermodynamic stabilities, also taking self-irradiation effects into account. Results highlight the main influence of americium reduction to Am+III, not only on the mechanisms of solid-state formation of the U1-xAmxO2±δ solid solution, but also on the stabilization of oxidized uranium cations and the formation of defects in the oxygen sublattice such as vacancies and cuboctahedral clusters. In addition, the data acquired concerning the stability of U1-xAmxO2±δ compounds (existence of a miscibility gap, vaporization behavior) were compared to calculations based on new thermodynamic modelling of the U-Am-O ternary system. Finally, α-self-irradiation-induced structural effects on U1-xAmxO2±δ compounds were analyzed using XRD, XAS and TEM, allowing the influence of americium content on the structural swelling to be studied as well as the description of the evolution of radiation-induced structural defects.

Le combustible nucléaire usé, après avoir été débarrassé de l'uranium, du plutonium et potentiellement du neptunium par le procédé PUREX, est encore constitué d'environ la moitié des éléments du tableau périodique. Au sein de ceux-ci, l'américium est majoritairement responsable des émissions thermiques à long terme des colis de déchets. En le recyclant, la compacité des sites de stockage pourrait être significativement améliorée. Le procédé d'extraction liquide-liquide EXAm, dont l'étape clé est la séparation Am/Cm, a donc été développé afin de séparer l'Am seul. Pour ce faire, un mélange d'extractants est utilisé conjointement au TEDGA (N,N',N,N'-tétraéthyl-diglycolamide), un complexant aqueux. Ce dernier permet en effet d'améliorer la séparation Am/Cm en maintenant préférentiellement le curium en phase aqueuse, mais sa relation structure sélectivité est encore mal comprise. Ces travaux de thèse concernent donc la synthèse et l'évaluation d'analogues structuraux du TEDGA afin de mieux comprendre l'influence de sa structure sur la sélectivité Am/Cm dans le procédé EXAm. Durant cette étude, 14 analogues du TEDGA ont été synthétisés et 17 molécules ont été évaluées en extraction liquide-liquide. Plusieurs modifications structurales ont été étudiées : la longueur et l'encombrement stérique des chaines alkyle portées par les azotes, la taille de l'espaceur, ainsi que l'utilisation d'amides secondaires. Ainsi, ces travaux ont montré qu'à partir des chaines tétrabutyle, il n'est plus possible de maintenir les molécules en phase aqueuse et que l'ajout d'encombrement stérique sur les chaines alkyles ou la modification de la balance hydrophilie/lipophilie diminue systématiquement la sélectivité des ligands. L'introduction d'amides secondaires (-CONHR) donne des ligands extractibles par le solvant (formation de liaison hydrogène avec les extractants) ce qui les rend inutilisables dans le procédé EXAm. Quant à l'espaceur, il joue un rôle prépondérant sur la sélectivité : le raccourcir (malonamides) entraine la disparition du caractère complexant de la molécule à forte acidité, et l'allonger entraine une préférence du ligand pour l'Am au lieu du Cm (inversion de sélectivité). L'ensemble de l'étude met en lumière la particularité de la sélectivité apportée par le TEDGA face aux autres diglycolamides ainsi que la difficulté d'améliorer le procédé en utilisant cette famille de molécules. Néanmoins la meilleure compréhension de sa chimie a permis d'affiner sa modélisation dans le procédé et l'étude de sa relation structure sélectivité montre qu'une amélioration de la sélectivité des ligands pourrait encore être envisagée en rigidifiant l'espaceur. / After the reprocessing of uranium, plutonium and eventually neptunium by the PUREX process, the spent fuel is still composed of half of the periodic table. Among these elements, the main responsible for the heat of the wastes is americium. Its reprocessing could allow improving the compactness of deep geological storage of the wastes. Thus the liquid-liquid extraction process called EXAm was developed in order to recover the americium alone. The key step of the process is the Am/Cm separation. An extractant mixture is used with an aqueous complexing agent: TEDGA (N,N',N,N'-tetraethyl-diglycolamide). It allows to enhance the Am/Cm separation by keeping preferentially curium in the aqueous phase, but its structure selectivity relationship is not well known yet. Therefore, the purpose of this thesis is to synthesize and evaluate some structural analogues of TEDGA, in order to better understand the impact of its structure on the Am/Cm selectivity in the EXAm process.During this study, 14 analogues of TEDGA were synthesized and 17 molecules were evaluated in liquid-liquid extraction. Several structural modifications were studied: length and steric hindrance of the N-alkyl chains, size of the spacer, and the introduction of secondary amide functions. This work shows that it is not possible to maintain the ligand in the aqueous phase from tetrabutyl derivatives, and the addition of steric hindrance, or modification of hydrophilicity/lipophilicity balance, systematically decreases the selectivity of ligands. The addition of secondary amide functions (-CONHR) makes the molecules extractible by the solvent (formation of hydrogen bonds with the extractants), therefore they are unsuitable to be used in the EXAm process. The spacer has the main impact on the selectivity: the complexation capacity in high acid medium disappears when it is shortened (malonamide), whereas the ligand has a preference for Am instead of Cm (inversion of selectivity) when it is lengthen (TEDOODA and TETOUDA derivatives).This whole study shows the peculiar selectivity given by the TEDGA in comparison with other diglycolamides, and the difficulties to enhance the process using this family of ligands. Nevertheless, a better understanding of its chemistry has allowed to define more clearly its modeling in the process and the studying of its structure/selectivity relationship has shown that the enhancing of the ligands selectivity probably requires spacer preorganization.

In this thesis, the feasibility to use a medium sized sodium cooled fast reactor fully loaded with MOX fuel for efficient transmutation of americium is investigated by simulating the safety performance of a BN600-type fast reactor loaded with different fractions of americium in the fuel, using the safety parameters obtained with the SERPENT Monte Carlo code. The focus is on americium mainly due to its long-term contribution to the radiotoxicity of spent nuclear fuel and its deterioration on core's safety parameters. Applying the SAS4A/SASSYS transient analysis code, it is demonstrated that the power rating needs to be reduced by 6% for each percent additional americium introduction into the reference MOX fuel, maintaining 100 K margin to fuel melting, which is the most limiting failure mechanism.Safety analysis of a new Accelerator Driven System design with a smaller pin pitch-to-diameter ratio comparing to the reference EFIT-400 design, aiming at improving neutron source efficiency, was also performed by simulating performance for unprotected loss of flow, unprotected transient overpower, and protected loss-of-heat-sink transients, using neutronic parameters obatined from MCNP calculations. Thanks to the introduction of the austenitic 15/15Ti stainless steel with enhanced creep rupture resistance and acceptable irradiation swelling rate, the suggested ADS design loaded with nitride fuel and cooled by lead-bismuth eutectic could survive the full set of transients, preserving a margin of 130 K to cladding rupture during the most limiting transient. The thesis concludes that efficient transmutation of americium in a medium sized sodium cooled fast reactor loaded with MOX fuel is possible but leads to a severe power penalty. Instead, preserving transmutation rates of minor actinides up to 42 kg/TWhth, the suggested ADS design with enhanced proton source efficiency appears like a better option for americium transmutation. / QC 20110318

The objective of this project was the study of basic interaction processes in the systems actinide - clay organics - aquifer and actinide - natural clay - clay organics - aquifer. Thus, complexation, redox, sorption and diffusion studies were performed. To evaluate the influence of nitrogen, phosphorus and sulfur containing functional groups of humic acid (HA) on the complexation of actinides in comparison to carboxylic groups, the Am(III) and U(VI) complexation by model ligands was studied by UV-Vis spectroscopy and TRLFS. The results show that Am(III) is mainly coordinated via carboxylic groups, however, probably stabilized by nitrogen groups. The U(VI) complexation is dominated by carboxylic groups, whereas nitrogen and sulfur containing groups play a minor role. Phosphorus containing groups may contribute to the U(VI) complexation by HA, however, due to their low concentration in HA they play only a subordinate role compared to carboxylic groups. Applying synthetic HA with varying sulfur contents (0 to 6.9 wt.%), the role of sulfur functionalities of HA for the U(VI) complexation and Np(V) reduction was studied. The results have shown that sulfur functionalities can be involved in U(VI) humate complexation and act as redox-active sites in HA for the Np(V) reduction. However, due to the low content of sulfur in natural HA, its influence is less pronounced. In the presence of carbonate, the U(VI) complexation by HA was studied in the alkaline pH range by means of cryo-TRLFS (-120°C) and ATR FT-IR spectroscopy. The formation of the ternary UO2(CO3)2HA(II)4− complex was detected. The complex formation constant was determined with log β0.1 M = 24.57 ± 0.17. For aqueous U(VI) citrate and oxalate species, luminescence emission properties were determined by cryo-TRLFS and used to determine stability constants. The existing data base could be validated. The U(VI) complexation by lactate, studied in the temperature range 7 to 65°C, was found to be endothermic and entropy-driven. In contrast, the complex stability constants determined for U(VI) humate complexation at 20 and 40°C are comparable, however, decrease at 60°C. For aqueous U(IV) citrate, succinate, mandelate and glycolate species stability constants were determined. These ligands, especially citrate, increase solubility and mobility of U(IV) in solution due to complexation. The U(VI) sorption onto crushed Opalinus Clay (OPA, Mont Terri, Switzerland) was studied in the absence and presence of HA or low molecular weight organic acids, in dependence on temperature and CO2 presence using OPA pore water as background electrolyte. Distribution coefficients (Kd) were determined for the sorption of U(VI) and HA onto OPA with (0.0222 ± 0.0004) m3/kg and (0.129 ± 0.006) m3/kg, respectively. The U(VI) sorption is not influenced by HA (50 mg/L), however, decreased by low molecular weight organic acids (> 1×10-5 M), especially by citrate and tartrate. With increasing temperature, the U(VI) sorption increases both in the absence and in the presence of clay organics. The U(VI) diffusion in compacted OPA is not influenced by HA at 25 and 60°C. Predictions of the U(VI) diffusion show that an increase of the temperature to 60°C does not accelerate the migration of U(VI). With regard to uranium-containing waste, it is concluded that OPA is suitable as host rock for a future nuclear waste repository since OPA has a good retardation potential for U(VI).

The accumulation of spent nuclear fuel may be hindering the expansion of nuclear electricity production. However, the reprocessing and recycling of spent fuel may reduce its volume and environmental burden. Although fast spectrum reactors are the preferred modality for transuranic element transmutation, such fast spectrum systems are in very short supply. It is therefore legitimate to investigate the recycling potential of thermal spectrum systems, which constitute the overwhelming majority of nuclear power plants worldwide. To do so efficiently, several new fuel assembly designs are proposed in this Thesis: these include (1) Mixed Oxide Fuel (MOX), (2) MOX fuel with Americium coating, (3) Inert-Matrix Fuel (IMF) with UOX as inner zone, and (4) IMF with MOX as inner zone. All these designs are investigated in a multi-recycling strategy, whereby the spent fuel from a given generation is re-used for the next generation. The accumulation of spent nuclear fuel may be hindering the expansion of nuclear electricity production. However, the reprocessing and recycling of spent fuel may reduce its volume and environmental burden. Although fast spectrum reactors are the preferred modality for transuranic element transmutation, such fast spectrum systems are in very short supply. It is therefore legitimate to investigate the recycling potential of thermal spectrum systems, which constitute the overwhelming majority of nuclear power plants worldwide. To do so efficiently, several new fuel assembly designs are proposed in this Thesis: these include (1) Mixed Oxide Fuel (MOX), (2) MOX fuel with Americium coating, (3) Inert-Matrix Fuel (IMF) with UOX as inner zone, and (4) IMF with MOX as inner zone. All these designs are investigated in a multi-recycling strategy, whereby the spent fuel from a given generation is re-used for the next generation.

Made available in DSpace on 2014-10-09T12:46:41Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T13:56:33Z (GMT). No. of bitstreams: 1 07611.pdf: 2728562 bytes, checksum: f4e2c166198e6ed56d8ad3f09429fc60 (MD5) / Dissertacao (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP