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Made available in DSpace on 2014-10-09T12:52:19Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:00:31Z (GMT). No. of bitstreams: 0 / 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. / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

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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).

Les isotopes de l’Américium sont les principaux contributeurs à la radioactivité des déchets nucléaires. Parmi les scénarios pour diminuer la toxicité des déchets, la transmutation dans les réacteurs à neutrons rapides utilisant des pastilles d’oxyde mixte (U,Am)O2 est une voie prometteuse. Dans ce cadre, la connaissance des propriétés thermodynamiques du système U-Am-O est essentielle pour prédire le comportement des pastilles (U,Am)O2 en conditions nominale et accidentelle. Cette thèse est dédiée à l’étude expérimentale d’oxydes mixtes (U,Am)O2 dans une large gamme de composition (7,5 % at. ≤ Am/(Am+U) ≤ 70 % at.). L’objectif est d’acquérir des données pour développer un modèle thermodynamique avec la méthode semi-empirique CALPHAD. Les résultats peuvent être classés en trois catégories : données structurales, données de diagramme de phase et données thermodynamiques. Pour la modélisation thermodynamique d’un système ternaire, l’optimisation des sous-systèmes binaires est nécessaire. Comme des questions restaient en suspens sur le système Am-O, le diagramme de phase Am-O a tout d’abord été étudié par diffraction des rayons X à haute température. L’existence d’un domaine de composition de la phase bcc AmO1.61 a été mis en évidence et la lacune de miscibilité dans la phase fluorite, proposée dans la littérature, n’a pas été confirmée. Grâce à ces nouveaux résultats, le modèle CALPHAD de Gotcu et al a été modifié. Dans une deuxième étape, des analyses structurales des dioxydes (U,Am)O2±x ont été effectuées par DRX, XAS et spectroscopie RAMAN. La DRX a permis de confirmer que tous les échantillons sont constitués d’une seule phase de structure fluorite. Le rapport O/M (avec M=U+Am) mesuré à température ambiante est inférieur à 2 ; la stabilité de l’Américium trivalent Am3+ a été mise en évidence. Celle-ci induit l’oxydation partielle de l’U4+ en U5+. Cette distribution de charge s’accompagne par la formation de défauts de l’oxygène complexes dans la structure fluorite. Lors de l’étude par DRX HT des oxydes mixtes sous air, il a été montré que la présence d’Am3+ stabilise la phase fluorite par rapport aux oxydes plus riches en oxygène (U4O9, U3O8). De nouvelles données de diagramme de phase ont été obtenues : des conodes dans les domaines biphasés M4O9-M3O8 and MO2+x-M3O8 et la solubilité de l’Américium dans les oxydes M4O9 et M3O8. L’étude du diagramme de phase U-Am-O a été poursuivie par la détermination des températures de solidus/liquidus des oxydes mixtes par une technique de chauffage laser, sous argon et sous air, et par la caractérisation des échantillons après fusion par SEM et XAS. La température de fusion des oxydes mixtes diminue avec une teneur croissante d’Américium (Am/(Am+U)) et d’oxygène (O/(Am+U)). Finalement, les propriétés thermodynamiques des oxydes (U,Am)O2±x ont été mesurées : les incréments enthalpiques par calorimétrie de chute et les pressions partielles des espèces gazeuses par Spectrométrie de Masse couplée à une cellule de Knudsen (KEMS). Une contribution d’excès de la capacité calorifique a été observée à haute température, attribuée à la réduction des oxydes (avec formation de lacunes d’oxygène). Les résultats de KEMS ont permis de déterminer une composition congruente de vaporisation à 2300 K, pour un rapport Am/(Am+U) de 0,6 et un rapport O/(U+Am) inférieur à 1,9. Finalement, la modélisation thermodynamique du système U-Am-O par la méthode CALPHAD a été abordée par la description de la phase fluorite. Un bon accord est obtenu entre le modèle et les données de potentiel d’oxygène pour l’oxyde (U0.5Am0.5)O2±x et de distribution des cations. De plus, le modèle permet de reproduire de façon satisfaisante les données de KEMS. En perspective de ce travail, la modélisation thermodynamique du ternaire sera étendue à la description des équilibres de phase mettant en jeu les oxydes M4O9, M3O8 et la phase liquide. / Americium isotopes are the main contributors to the long-term radiotoxicity of the nuclear wastes, after the plutonium extraction. Among the reprocessing scenarios, the transmutation in fast neutron reactors using uranium-americium mixed oxide (U,Am)O2±x pellets seems promising. In this frame, the knowledge of the thermodynamics of the U-Am-O ternary system is of essential for the prediction of the behavior of (U,Am)O2 pellets and their possible interaction with the cladding, under normal and accidental conditions. This thesis is dedicated to the experimental investigation of U-Am mixed oxides on a wide range of Am contents (7.5 at.% ≤ Am/(Am+U) ≤ 70 at.%), with the aim to collect data for developing a thermodynamic model based on the semi-empirical CALPHAD method. The obtained results can be classified in three categories: structural, phase diagram and thermodynamic data. For the thermodynamic modeling of the ternary system, the assessment of the binary sub-systems is first required. As open questions still existed on the Am-O system, a first part of the work was dedicated to the study of the Am-O phase diagram by high-temperature (HT) XRD. The existence of a composition range of the bcc AmO1.61 phase was highlighted and the miscibility gap in the fluorite phase, proposed in the literature, was not found. Thanks to the new experimental data, the existing CALPHAD model of Gotcu et al. was modified. In a second step, structural investigations were performed on synthesized (U,Am)O2±x dioxides by coupling XRD, XAS and Raman spectroscopy. For all the compositions, the XRD confirmed the formation of a single fluorite structure. The O/M ratio (with M=U+Am) at room temperature was determined to be lower than 2; the stability of trivalent americium Am3+ in the dioxide solid solution was highlighted, which induces a partial oxidation of uranium from U4+ to U5+. This charge distribution, peculiar for a dioxide, is accompanied by the formation of complex oxygen defects in the fluorite structure. By a HT-XRD investigation of the mixed oxides under air combined with XAS characterization of the oxidized samples, it was shown that the presence of Am3+ leads to a stabilization of the dioxide fluorite phase toward the formation of oxides richer in oxygen, in comparison to the U-O system. New phase diagram data were obtained in the oxygen rich region at 1470 K: tie-lines in the M4O9-M3O8 and MO2+x-M3O8 domains were determined and the solubility of americium in the M4O9 and M3O8 oxides was estimated. The investigation of the U-Am-O phase diagram continued at higher temperature with the study of the solidus/liquidus transitions using a laser-heating technique, under argon and air, and post-melting characterizations conducted by SEM and XAS. The melting temperature of Am-U dioxides decreases with the increase of both the Am/(Am+U) and O/M ratios. Finally, thermodynamic properties of the U1-yAmyO2±x oxides were measured: enthalpy increments using drop calorimetry, partial vapor pressures by Knudsen cell effusion mass spectrometry (KEMS). An excess contribution to the heat capacity at high temperature was observed and this was attributed to the reduction of the dioxides at high temperature (formation of oxygen vacancies). The KEMS results lead to determine the congruent vaporization composition at 2300 K, for a Am/(Am+U) ratio of 0.6 and an O/M ratio lower than 1.9. Finally, the CALPHAD thermodynamic assessment of the U-Am-O system was started, by focusing the attention on the modelling of the fluorite phase. A good agreement between the model and the oxygen potential data for (U0.5Am0.5O2±x) and the cation distribution was achieved. Furthermore, the model is able to satisfactorily reproduce the KEMS data and hence the equilibrium between the dioxide and gas phase. For the perspectives of this work, the optimization of the thermodynamic model should be extended to describe the phase equilibria involving the M4O9, M3O8 oxides and the liquid phase.

The MURE code is based on the coupling of a Monte Carlo static code and the calculation of the evolution of the fuel during irradiation and cooling periods. The MURE code has been used to analyse two different questions, concerning the mono-recycling of Am in present French Pressurized Water Reactor, and the conversion of high enriched uranium (HEU) used in the Miniature Neutron Source Reactor in Ghana into low enriched uranium (LEU) due to proliferation resistance issues. In both cases, a detailed comparison is made on burnup and the induced radiotoxicity of waste or spent fuel. The UOX fuel assembly, as in the open cycle system, was designed to reach a burn-up of 46GWd/T and 68GWd/T. The spent UOX was reprocessed to fabricate MOX assemblies, by the extraction of Plutonium and addition of depleted Uranium to reach burn-ups of 46GWd/T and 68GWd/T, taking into account various cooling times of the spent UOX assembly in the repository. The effect of cooling time on burnup and radiotoxicity was then ascertained. Spent UOX fuel, after 30 years of cooling in the repository required higher concentration of Pu to be reprocessed into a MOX fuel due to the decay of Pu-241. Americium, with a mean half-life of 432 years, has high radiotoxic level, high mid-term residual heat and a precursor for other long lived isotope. An innovative strategy consists of reprocessing not only the plutonium from the UOX spent fuel but also the americium isotopes which dominate the radiotoxicity of present waste. The mono-recycling of Am is not a definitive solution because the once-through MOX cycle transmutation of Am in a PWR is not enough to destroy all the Am. The main objective is to propose a "waiting strategy" for both Am and Pu in the spent fuel so that they can be made available for further transmutation strategies. The MOXAm (MOX and Americium isotopes) fuel was fabricated to see the effect of americium in MOX fuel on the burn-up, neutronic behavior and on radiotoxicity. The MOXAm fuel showed relatively good indicators both on burnup and on radiotoxicity. A 68GWd/T MOX assembly produced from a reprocessed spent 46GWd/T UOX assembly showed a decrease in radiotoxicity as compared to the open cycle. All fuel types understudy in the PWR cycle showed good safety inherent feature with the exception of the some MOXAm assemblies which have a positive void coefficient in specific configurations, which could not be consistent with safety features. The core lifetimes of the current operating 90.2% HEU UAl fuel and the proposed 12.5% LEU UOX fuel of the MNSR were investigated using MURE code. Even though LEU core has a longer core life due to its higher core loading and low rate of uranium consumption, the LEU core will have it first beryllium top up to compensate for reactivity at earlier time than the HEU core. The HEU and LEU cores of the MNSR exhibited similar neutron fluxes in irradiation channels, negative feedback of temperature and void coefficients, but the LEU is more radiotoxic after fission product decay due to higher actinides presence at the end of its core lifetime.

In der Umwelt vorkommende Radionuklide wurden als Tracer für Migrationsverhalten verwendet. Low-level-γ-Spektrometrie und zusätzliche Plutoniumanalysen ermöglichten den Nachweis sehr geringer Konzentrationen radioaktiver Nuklide. Der atmosphärische Eintrag wurde über Regenwasser- und Staubfilterproben gemessen. Untersuchungen an Filterstäuben ermöglichten eine nachträgliche Charakterisierung der Eintragssituation vor und nach dem Tschernobyl-Unfall im Freiberger Raum. Die Radionuklidtiefenverteilungen in Fichtenwald- und Hochmoorböden und deren weiterführende Analyse erlaubten Aussagen zum Stofftransport und zu bodenbildenden Prozessen. Die Migrationsdynamik in weitgehend ungestörten Hochmooren unterscheidet sich von der anthropogen überprägter Moore. Die Eignung verschiedener Radionuklide zur Datierung bzw. zeitlichen Markierung von Moorbodenschichten und damit zur Verwendung in der Moorstratigraphie wurde geprüft. Eine gute Übereinstimmung der Ergebnisse verschiedener Datierungs- und Markierungsmethoden zeigte sich v. a. für die Profile der ungestörten, rezent wachsenden ombrogenen Hochmoore.