A incorporação do gadolínio diretamente no combustível de reatores nucleares para geração de eletricidade é importante para compensação da reatividade e para o ajuste da distribuição da densidade de potência, permitindo ciclos de queima mais longos, com intervalo de recarga de 18 meses, otimizando-se a utilização do combustível. A incorporação do Gd2O3 sob a forma de pó homogeneizado a seco diretamente com o pó de UO2 é o método comercialmente mais atraente devido à sua simplicidade . Contudo, este método de incorporação conduz a dificuldades na obtenção de corpos sinterizados com a densidade niínima especificada, devido a um bloqueio no processo de sinterização. Pouca informação existe na literatura específica sobre o possível mecanismo deste bloqueio, restrita principalmente à hipótese da formação de uma fase (U,Gd)O2 rica em gadolínio com baixa difusividade. Este trabalho tem como objetivo a investigação do mecanismo de bloqueio da sinterização neste sistema, contribuindo para o esclarecimento da causa deste bloqueio e na elaboração de possíveis soluções tecnológicas. Foi comprovado experimentalmente que o mecanismo responsável pelo bloqueio é baseado na formação de poros estáveis devido ao efeito Kirkendall, originados por ocasião da formação da solução sólida durante a etapa intermediária da sinterização, sendo difícil a sua eliminação posterior, nas etapas finais do processo de sinterização. Com base no conhecimento deste mecanismo, possíveis propostas são apresentadas na direção da solução tecnológica do problema de densificação característico do sistema UO2-Gd203. / The incorporation of gadolinium directly into nuclear power reactor fuel is important from the point of reactivity compensation and adjustment of power distribution enabling thus longer fliel cycles and optimized fuel utilization. The incorporation of Gd2O3 powder directly into the UO2 powder by dry mechanical blending is the most attractive process because of its simplicity. Nevertheless, processing by this method leads to difficulties while obtaining sintered pellets with the minimum required density. This is due to blockages during the sintering process. There is little information in published literature about the possible mechanism for this blockage and this is restricted to the hypothesis based on formation of a low difiiisivity Gd rich phase (U,Gd)O2. The objective of this investigation has been to study the blockage mechanism in this system during the sintering process, contributing thus, to clarify the cause for the blockage and to propose feasible technological solutions. Experimentally it has been shown that the blocking mechanism is based on pore formation because of the Kirkendall effect. Formation of a solid solution during the intermediate stage of sintering leads to formation of large pores, which are difficult to remove in the final stage of sintering. Based on this mechanism, technical solutions have been proposed to resolve densification problems in the UO2-Gd2O3 system.

Le dioxyde d'uranium (UO2) est le combustible nucléaire le plus largement utilisé dans les réacteurs nucléaires à travers le monde. En conditions d’exploitation, UO2 est soumis au flux de neutrons et subit des réactions en chaîne de fission nucléaire, ce qui crée un grand nombre de produits de fission et des défauts ponctuels. L'étude du comportement des produits de fission et des défauts ponctuels est importante pour comprendre les propriétés du combustible sous irradiation. Nous effectuons des calculs de structure électronique basés sur la théorie de la fonctionnelle de la densité (DFT) pour modéliser les dégâts d’irradiation à l'échelle atomique. La méthode DFT+U est utilisé pour décrire les fortes corrélations des électron 4f du cérium et des électrons 5f de l’uranium dans les matériaux étudiés (UO2, CeO2 et (U, Ce)O2). (U, Ce)O2 est étudié car il est considéré comme un matériau modèle peu radioactif d'oxydes d’actinides mixtes comme (U, Pu)O2 qui est le combustible d'oxydes mixtes (MOX) utilisé dans les réacteurs à eau légère et les réacteurs à neutrons rapides. Le dioxyde de cérium (CeO2) est étudié pour des données de référence de (U, Ce)O2. Nous effectuons une étude DFT+U des défauts ponctuels et des produits de fission gazeux (Xe et Kr) dans CeO2 et comparons nos résultats à ceux déjà existants pour l’UO2. Nous étudions les propriétés en volume, ainsi que le comportement des défauts pour (U, Ce)O2, et comparons nos résultats à ceux de (U, Pu)O2. En outre, pour l'étude des défauts dans UO2, des améliorations méthodologiques sont explorées considérant l'effet de couplage spin-orbite et l’effet de taille finie de la supercellule de modélisation. / Uranium dioxide (UO2) is the most widely used nuclear fuel in existing nuclear reactors around the world. While in service for energy supply, UO2 is submitted to the neutron flux and undergoes nuclear fission chain reactions, which create large number of fission products and point defects. The study of the behavior of the fission products and point defects is important to understand the fuel properties under irradiation. We conduct electronic structure calculations based on the density functional theory (DFT) to model this radiation damage at the atomic scale. The DFT+U method is used to describe the strong correlation of the 4f electrons of cerium and 5f electrons of uranium in the materials studied (UO2, CeO2 and (U, Ce)O2). (U, Ce)O2 is studied because it is considered as a low radioactive model material of mixed actinide oxides such as the MOX fuel (U, Pu)O2 used in light water reactors and fast neutron reactors. Cerium dioxide (CeO2) is studied to provide reference data of (U, Ce)O2. We perform a DFT+U study of point defects and gaseous fission products (Xe and Kr) in CeO2 and compare our results to the existing ones of UO2We study the bulk properties as well as the behavior of defects for (U, Ce)O2, and compare our results to the ones of (U, Pu)O2. Furthermore, for the study of defects in UO2, methodological improvements are explored considering the spin-orbit coupling effect and the finite-size effect of the simulation supercell.

Made available in DSpace on 2014-10-09T12:26:35Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:06:28Z (GMT). No. of bitstreams: 0 / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Made available in DSpace on 2014-10-09T12:54:58Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:07:31Z (GMT). No. of bitstreams: 0 / Dissertacao (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

Dans les nouvelles centrales nucléaires dites 4ème génération, comme d’ailleurs les anciennes, le dioxyde d’uranium devra opérer dans des milieux hostiles de températures et d’irradiation avec la présence des produits de fission (PF) et des particules alpha (α). Le fonctionnement dans ces conditions extrêmes induira des déplacements d’atomes et dégradera les propriétés thermiques et mécaniques du combustible UO2. La compréhension du comportement des défauts lacunaires, des PF et de l’hélium est cruciale pour prévoir le comportement du dioxyde d’uranium au sein de ces futures installations nucléaires. La première partie de cette thèse est consacrée à l’étude des défauts lacunaires induits par l’implantation de krypton et d’iode (quelques MeV) dans l’UO2 polycristallin et leurs stades de recuits. L’analyse par spectroscopie d’annihilation de positons (PAS) a permis de mettre en évidence la création de défauts de Schottky VU-2VO dans le cas des implantations iode et la formation de clusters lacunaires contenant du gaz pour les implantations krypton. L’évolution en température de ces défauts générés dépend des paramètres d’implantation (nature des ions, énergie, fluence). Cette étude a montré les rôles importants que peuvent jouer les défauts lacunaires et la présence des gaz de fission dans l’évolution du matériau UO2. Ensuite, nous nous sommes intéressés à l’étude et à la caractérisation, par PAS et les techniques d’analyse par faisceau d’ions (NRA/C et RBS/C), du comportement de l’hélium dans l’UO2. Les mesures de NRA/C et RBS/C révèlent une localisation d’une grande fraction d’hélium dans les sites interstitiels octaédriques de la matrice UO2. La localisation de l’hélium reste stable dans ces sites pour T< 600°C, évoluent légèrement entre 600 et 700°C et devient aléatoire à 800°C. Les mesures PAS mettent en évidence trois stades d’évolution des défauts lacunaires : la recombinaison par migration des interstitiels d’oxygène, l’agglomération des défauts entre 600 et 800°C et leur dissociation et élimination lorsque la température augmente. Ces résultats suggèrent que le transport d'hélium est assisté par les défauts lacunaires. / In the new fourth generation nuclear plants, as in the old ones, uranium dioxide must operate in hostile environments of temperature and irradiation with the presence of fission products (FP) and alpha particles (α). Operation in these extreme conditions will induce atoms displacements and degrade the thermal and mechanical properties of UO2 fuel. Understanding the behavior of induced vacancy defects, FP and helium is crucial to predict the uranium dioxide behavior in the future nuclear reactors. The first part of this thesis is dedicated to the study of vacancy defects induced by krypton and iodine implantation (a few MeV) in the UO2 polycrystalline and of their evolution under annealing. Analysis by positron annihilation spectroscopy (PAS) has highlighted the creation of Schottky defects VU-2VO in the case of iodine implantations and formation of vacancy clusters containing the gas for krypton implantation. The temperature evolution of these defects depends on the implantation parameters (nature of the ion energy, fluence). This study showed the important roles that can play vacancy defects and the presence of fission gases in the evolution of UO2 material. Then we were interested in the study of the helium behavior in UO2 its location and migration, agglomeration and interaction with vacancy defects by using PAS and ion beam analysis (NRA/C and RBS/C). The NRA/C and RBS/C characterizations showed a localization of a large helium fraction in the octahedral interstitial sites of the UO2 matrix. The helium location in these sites remains stable for T <600°C, changing slightly between 600 and 700°C and becomes random at 800°C. Positron annihilation spectroscopy reveals three stages of vacancy defects evolution : The recombination with oxygen interstitial migration, defects agglomeration between 600 and 800°C and their dissociation and elimination when the temperature increases. These results suggest that the He transport is assisted by the vacancy defects.

Made available in DSpace on 2014-10-09T12:26:35Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:06:28Z (GMT). No. of bitstreams: 0 / Neste trabalho foram determinados experimentalmente os índices espectrais 28r* e 25d* e o fator de eficiência de contagem gama através da técnica de varredura gama de varetas combustíveis no reator nuclear IPEN/MB-01. A vantagem deste método experimental consiste no fato de terem sido eliminados a maioria dos fatores de correção advindos dos cálculos, permanecendo somente os fatores de rendimento médio de fissão e a fração de fissão no 235U na determinação do 25d*. Os experimentos foram efetuados com luvas de cádmio de diferentes espessuras: 0,55 mm, 1,10 mm e 2,20 mm. As incertezas experimentais inferiores a 1% e a excelente caracterização dos dados geométricos e materiais do reator IPEN/MB-01 permitem utilizar os resultados obtidos como benchmark para a validação de bibliotecas de dados nucleares. Sendo assim, foi utilizado o programa MCNP-5 com as bibliotecas de dados nucleares: ENDF/B-VI.8, ENDF/B-VII.0, JENDL-3.3 e JEFF-3.1. A comparação entre os valores advindos dos cálculos e os resultados experimentais mostrou que houve progressos sensíveis com as bibliotecas de dados nucleares atuais. Os desvios entre a comparação dos valores calculados e os resultados experimentais são inferiores a 2 %, sendo que o melhor desempenho foi obtido com a biblioteca de dados nucleares ENDF/B-VII.0 e a incerteza máxima na comparação dos resultados foi de -1,4 %, para as bibliotecas de dados nucleares JEFF-3.1 e JENDL-3.3. / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Made available in DSpace on 2014-10-09T12:54:58Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:07:31Z (GMT). No. of bitstreams: 0 / O processamento químico a partir do hexafluoreto de urânio (UF6), permite uma flexibilidade na produção de combustíveis à base de siliceto de urânio (U3Si2) e octóxido de urânio (U3O8). Atualmente no IPEN-CNEN/SP desenvolvem-se trabalhos visando o processamento de combustíveis com alta concentração de urânio, por meio da substituição do U3O8 por U3Si2. Para a obtenção de U3Si2, duas possibilidades podem ser consideradas na preparação da matéria-prima utilizada, que é o tetrafluoreto de urânio (UF4), são elas: a redução do urânio presente na solução hidrolisada do UF6 utilizando-se cloreto estanhoso (SnCl2) e a hidrofluoretação do dióxido de urânio (UO2) proveniente do tricarbonato de amônio e uranilo (TCAU). Descreve-se neste trabalho um procedimento para obtenção de tetrafluoreto de urânio (UF4), utilizando-se como matéria-prima os filtrados gerados na preparação de determinados compostos nos processos de reconversão do hexafluoreto de urânio (UF6), mais especificamente o amonioperóxidofluoranato (APOFU). Os filtrados consistem principalmente de uma solução contendo altas concentrações dos íons amônio (NH4 +), fluoreto (F-) e baixa concentração de urânio. O processo descrito visa principalmente a recuperação do NH4F e do urânio, como UF4, por meio da cristalização do bifluoreto de amônio (NH4HF2) e em uma etapa posterior, a adição deste ao UO2, ocorrendo a fluoração e decomposição. O UF4 obtido foi caracterizado química e fisicamente e será reciclado para ser usado na unidade de produção de urânio metálico para a obtenção de U3Si2, utilizado como combustível para o reator IEA-R1m. / Dissertacao (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

A incorporação do gadolínio diretamente no combustível de reatores nucleares para geração de eletricidade é importante para compensação da reatividade e para o ajuste da distribuição da densidade de potência, permitindo ciclos de queima mais longos, com intervalo de recarga de 18 meses, otimizando-se a utilização do combustível. A incorporação do Gd2O3 sob a forma de pó homogeneizado a seco diretamente com o pó de UO2 é o método comercialmente mais atraente devido à sua simplicidade . Contudo, este método de incorporação conduz a dificuldades na obtenção de corpos sinterizados com a densidade niínima especificada, devido a um bloqueio no processo de sinterização. Pouca informação existe na literatura específica sobre o possível mecanismo deste bloqueio, restrita principalmente à hipótese da formação de uma fase (U,Gd)O2 rica em gadolínio com baixa difusividade. Este trabalho tem como objetivo a investigação do mecanismo de bloqueio da sinterização neste sistema, contribuindo para o esclarecimento da causa deste bloqueio e na elaboração de possíveis soluções tecnológicas. Foi comprovado experimentalmente que o mecanismo responsável pelo bloqueio é baseado na formação de poros estáveis devido ao efeito Kirkendall, originados por ocasião da formação da solução sólida durante a etapa intermediária da sinterização, sendo difícil a sua eliminação posterior, nas etapas finais do processo de sinterização. Com base no conhecimento deste mecanismo, possíveis propostas são apresentadas na direção da solução tecnológica do problema de densificação característico do sistema UO2-Gd203. / The incorporation of gadolinium directly into nuclear power reactor fuel is important from the point of reactivity compensation and adjustment of power distribution enabling thus longer fliel cycles and optimized fuel utilization. The incorporation of Gd2O3 powder directly into the UO2 powder by dry mechanical blending is the most attractive process because of its simplicity. Nevertheless, processing by this method leads to difficulties while obtaining sintered pellets with the minimum required density. This is due to blockages during the sintering process. There is little information in published literature about the possible mechanism for this blockage and this is restricted to the hypothesis based on formation of a low difiiisivity Gd rich phase (U,Gd)O2. The objective of this investigation has been to study the blockage mechanism in this system during the sintering process, contributing thus, to clarify the cause for the blockage and to propose feasible technological solutions. Experimentally it has been shown that the blocking mechanism is based on pore formation because of the Kirkendall effect. Formation of a solid solution during the intermediate stage of sintering leads to formation of large pores, which are difficult to remove in the final stage of sintering. Based on this mechanism, technical solutions have been proposed to resolve densification problems in the UO2-Gd2O3 system.

Dans le combustible nucléaire UO2, utilisé dans les réacteurs à eau pressurisée (REP), le Cs, élément volatil compte parmi les produits de fission (PF) les plus abondamment produits. De plus, l’isotope 137Cs est connu pour être particulièrement radiotoxique. En cas d’accident, le relâchement de cet isotope est donc problématique et son étude est cruciale pour la sûreté nucléaire. En France, l’IRSN (Institut de Radioprotection et de sureté nucléaire) développe des codes de prédictions du relâchement des PF depuis le combustible, tels que MFPR (Module for Fission Product Release). Ces codes nécessitent d’être alimentés par des données fondamentales sur le comportement des PF. Ainsi, la connaissance des coefficients de diffusion de ces éléments dans la matrice combustible en fonction de la température et de l’atmosphère (pouvant oxyder le combustible en UO2+x) est primordiale. Dans ce contexte, l’objectif de cette thèse, menée en collaboration avec l’IRSN, est d’étudier la migration du Cs dans le dioxyde d’uranium stœchiométrique et sur-stœchiométrique, en conditions représentatives d’un fonctionnement normal et accidentel d’un REP, avec et sans la présence de Mo. Ce dernier est un PF abondamment produit qui agit comme tampon d’oxydation du combustible et est capable d’avoir des interactions chimiques avec le césium. De telles interactions pourraient affecter le comportement du Cs, et donc son relâchement depuis le combustible. Il a donc été nécessaire d’envisager les éventuelles interactions entre le Cs et le Mo dans le cadre de notre étude. La démarche expérimentale a consisté à simuler la présence de Cs et/ou Mo dans des pastilles d’UO2 ou d’UO2+x. par implantations ioniques des isotopes stables 133Cs et/ou 95Mo. Des recuits à haute température (950-1600°C) sous atmosphère contrôlée ou des irradiations en régime électronique couplées en température ont ensuite été réalisés, permettant d’induire la migration du Cs et du Mo. La spectrométrie de masse à ionisation secondaire (SIMS) a été utilisée pour suivre l’évolution des profils de concentration des éléments implantés, permettant d’extraire les coefficients de diffusion apparents du Cs dans UO2 et UO2+x en fonction des différents traitements. Une étude complémentaire de la microstructure a été réalisée par spectroscopie Raman et microscopie électronique en transmission (MET). Le Cs est très mobile dans UO2 sous atmosphère réductrice même si une partie et piégée sous forme de bulles à faible profondeur. Nous avons mis en évidence que la présence de Mo diminuait fortement cette mobilité. La même tendance est observée dans UO2+x sous atmosphère oxydante. Néanmoins les mécanismes d’immobilisation du Cs par le Mo diffèrent selon les conditions redox de recuit. En atmosphère réductrice, les expériences MET ont montré la formation de paires bulles de Cs-précipités métalliques de Mo dans les échantillons co-implantés. En atmosphère oxydante, l’absence de mobilité du Cs pourrait être liée à l’oxydation du Mo rendant possible des interactions chimiques Cs-Mo. Pour la première fois, des potentiels semi-empiriques ont été utilisés pour réaliser des calculs de dynamique moléculaire sur la diffusion du Cs et du Mo dans UO2 et UO2+x. Ces calculs nous ont aussi permis de caractériser les mécanismes de diffusion de l’oxygène dans ces matériaux en présence de ces deux PF / In the nuclear fuel UO2, which is widely used in Pressurized Water Reactor (PWR), Cs is a volatile element and is one of the most abundant fission product (FP). Furthermore, 137Cs is known to be highly radiotoxic. During a hypothetical accident, release of Cs would be particularly problematic for the environment. Hence, study of this element is of major concern for nuclear safety. To assess this issue, the French nuclear safety institute (IRSN) develops codes to predict FP release from nuclear fuel in normal and accidental conditions. This code requires fundamental data on FP behavior such as diffusion coefficient of these elements in UO2 as a function of temperature and atmosphere conditions (leading to UO2+x formation in oxidative conditions). The aim of this PhD, supported by the IRSN, is to study Cs migration in stoichiometric and hyper-stoichiometric uranium dioxide with and without the presence of Mo, in normal and accidental conditions of a PWR. This latter element is also an abundant FP, which is important to consider because it acts as an oxygen buffer in the fuel and may interact chemically with Cs. Such interactions may affect Cs behavior, hence its release from the fuel. Therefore, Cs-Mo interactions are considered in our study. The experimental procedure consists in simulating the Cs and/or Mo presence in UO2 and UO2+x pellets by ion implantation of stable isotopes 133Cs and/or 95Mo. Then, high temperature annealing (950 °C - 1600 °C) under controlled atmosphere or electronic excitations induced by irradiation coupled with temperature are performed to induce Cs and Mo migration. Secondary Ion Mass Spectrometry (SIMS) is used to follow the concentration profile evolution of these elements, allowing extracting effective diffusion coefficients of Cs in UO2 and UO2+x as a function of irradiation or thermal treatment. Microstructure characterizations were made by Raman spectroscopy and transmission electron microscopy (TEM). We show that Cs is mobile in UO2 under reducing atmosphere, even though some of the Cs is trapped in Cs-bubbles located near the surface. We evidence that Mo presence prevents Cs to be mobile. The same tendency is observed in UO2+x under oxidizing atmosphere. Nevertheless, Cs immobilization mechanisms in presence of Mo vary upon redox conditions used during annealing. In reducing conditions, TEM experiments showed formation of Cs bubbles associated with Mo metallic precipitates in co-implanted samples. In oxidative conditions, absence of Cs mobility could be explained by Mo oxidation leading to possible Cs-Mo chemical interactions. For the first time, semi-empirical potentials were used to perform molecular dynamic (MD) calculations on Cs and Mo diffusion in UO2 and UO2+x. These simulations also allowed characterizing oxygen diffusion mechanisms in these matrixes in presence of Cs and Mo

Uranoxid (UOx) är ett energitätt material som ofta används i kärnbränsle. UOx-pulver pressas och sintras för att tillverka urandioxidkutsar som förs in i bränslestavar. Stavarna monteras slutligen ihop till ett bränsleknippe. Tillverkningsprocessens stabilitet och förutsägbarhet är viktiga. För att åstadkomma önskvärda egenskaper hos UO2-kutsarna är karaktärisering av UOx-pulvret centralt. Sintringsaktivitet är den viktigaste egenskapen när det kommer till att beskriva hur UOx-pulvret beter sig vid reduktion i högtemperatursintring. Återcyklat UO2 oxideras till U3O8 och kan användas till att styra sintringsaktiviteten tack vare dess porbildande egenskaper. Denna rapport beskriver karaktäriseringen av UOx-pulver och kuts med avseende på fysiokemiska egenskaper relaterade till sintringsaktivitet. Statistiska analyser av historiska data utfördes även och visade på en komplex relation mellan pulveregenskaper och sintringsaktivitet. Effekten av U3O8-pulver i blandningar av UO2-pulver med hög och låg sintringsaktivitet undersöktes. Att variera U3O8-batch hade ingen inverkan på diameterkrympning efter sintring utom i ett fall. Blandningar av UO2-pulver visade på avvikande egenskaper jämfört med det jungfruliga pulvret. UO2-pulvrets kemiska aktivitet undersöktes via oxidering med H2O2. Förbrukningshastigheten av H2O2 var densamma för hög- och lågaktiva UO2-pulver vid samma förhållande mellan specifik yta och lösningsvolym. / Uranium oxide (UOx) is an energy dense material commonly used in nuclear fuel. UOx powder is pressed and sintered to produce uranium dioxide (UO2) pellets which are loaded into fuel rods. The rods are then mounted together in a final nuclear fuel assembly. Stability and predictability of the manufacturing processes during UO2 pellet production is of high importance. To achieve desired properties and quality of the UO2 pellets, the ability to assess the characteristics of the UOx powder is crucial. Sinterability is the most important characteristic which describes the behavior of the UOx powder during reduction in high temperatures. Recycled uranium dioxide is oxidized into U3O8 powder which can be used to modify the sinterability due to its pore forming ability. This study describes the characterization of uranium oxide powders and pellets regarding physicochemical properties relating to sintering behavior. Statistical analyses of historical data were also performed and showed a complexity of the relation between powder properties and  sinterability. The effect of U3O8 powder in different blends of UO2 powders of high and low sinterability were analyzed. Varying U3O8 powder batch did not influence the diameter shrinkage after sintering except for one case. UO2 powder blends showed deviating behavior from their virgin powder constituents. Chemical activity of UO2 was analyzed by oxidation with H2O2. The consumption rate of H2O2 was shown to be equal for active and incative UO2 powders under equal specific surface area/solution volume ratio.