Analysis of single and binary phases in cerium doped sodium bismuth titanate -Na0.5Bi0.5TiO3 materials

The pure and cerium doped sodium bismuth titanate inorganic powders were synthesized by solid state reaction method. The presence of rhombohedral phase was observed in cerium doped NBT compounds. At 1200 ºC, the 5% of cerium doped NBT compound forms single perovskite phase. The samples of x = 0.10 and 0.15 were heat treated to 1350 ºC, the binary phases with cerium and bismuth oxides were observed. The X-ray diffraction, fourier transform infrared spectroscopy, reflectance spectra, differential thermal analysis and thermo gravimetric analysis were used to analyze the various properties of samples. Moreover, the effects of cerium doping and calcining temperature on NBT samples were investigated. In this work we present our recent results on the synthesis and characterization of Ce doped sodium bismuth titanate materials.

Gold Ores Related to Shear Zones, West Santa Comba Fervenza Area (Galicia, NW Spain): A Mineralogical Study

Recent research has discovered high-grade Au ores in NNE-SSW trending shear zones in metamorphic proterozoic and palaeozoic terranes, some 40 km NW of Santiago de Compostela (NW Spain). The orebodies are bound to late-stage Hercynian structures, mainly due to brittle deformation, which are superimposed on earlier ductile shear zones, cutting through various catazonal lithologies, including ortho- and paragneisses, amphibolites, eclogites, and granites. Ore mineralogy, alteration, and ore textures define a frame whose main features are common to all prospects in the area. Main minerals are arsenopyrite and pyrite - accompanied by quartz, adularia, sericite, + (tourmaline, chlorite, carbonates, graphite), as main gangue minerals - with subordinate amounts of boulangerite, bismuthinite, kobellite, jamesonite, chalcopyrite, marcasite, galena, sphalerite, rutile, titanite, scheelite, beryl, fluorite, and minor native gold, electrum, native bismuth, fahlore, pyrrhotite, mackinawite, etc., defining a meso-catathermal paragenesis. Detailed microscopic study allows the author to propose a general descriptive scheme of textural classification for this type of ore. Most of the ores fill open spaces or veins, seal cracks or cement breccias; disseminated ores with replacement features related to alteration (mainly silicification, sericitization, and adularization) are also observed. Intensive and repeated cataclasis is a common feature of many ores, suggesting successive events of brittle deformation, hydrothermal flow, and ore precipitation. Gold may be transported and accumulated in any of these events, but tends to be concentrated in later ones. The origin of the gold ores is explained in terms of hydrotherreal discharge, associated with mainly brittle deformation and possibly related to granitic magmas, in the global tectonic frame of crustal evolution of West Galicia. The mineralogical and textural study suggests some criteria which will be of practical value for exploration and for ore processing. Ore grades can be improved by flotation of arsenopyrite. Non-conventional methods, such as pressure or bacterial leaching, may subsequently obtain a residue enriched in gold.

Obtaining an intermediate band photovoltaic material through the Bi insertion in CdTe

Defect interaction can take place in CdTe under Te and Bi rich conditions. We demonstrate in this work through first principles calculations, that this phenomenon allows a Jahn Teller distortion to form an isolated half-filled intermediate band in the host semiconductor band-gap. This delocalized energy band supports the experimental deep level reported in the host band-gap of CdTe at a low bismuth concentration. Furthermore, the calculated optical absorption of CdTe:Bi in this work shows a significant subband-gap absorption that also supports the enhancement of the optical absorption found in the previous experimental results.

Monte Carlo Neutronics and Thermal-hydraulics coupling applied to Fast Reactors

Un escenario habitualmente considerado para el uso sostenible y prolongado de la energía nuclear contempla un parque de reactores rápidos refrigerados por metales líquidos (LMFR) dedicados al reciclado de Pu y la transmutación de actínidos minoritarios (MA). Otra opción es combinar dichos reactores con algunos sistemas subcríticos asistidos por acelerador (ADS), exclusivamente destinados a la eliminación de MA. El diseño y licenciamiento de estos reactores innovadores requiere herramientas computacionales prácticas y precisas, que incorporen el conocimiento obtenido en la investigación experimental de nuevas configuraciones de reactores, materiales y sistemas. A pesar de que se han construido y operado un cierto número de reactores rápidos a nivel mundial, la experiencia operacional es todavía reducida y no todos los transitorios se han podido entender completamente. Por tanto, los análisis de seguridad de nuevos LMFR están basados fundamentalmente en métodos deterministas, al contrario que las aproximaciones modernas para reactores de agua ligera (LWR), que se benefician también de los métodos probabilistas. La aproximación más usada en los estudios de seguridad de LMFR es utilizar una variedad de códigos, desarrollados a base de distintas teorías, en busca de soluciones integrales para los transitorios e incluyendo incertidumbres. En este marco, los nuevos códigos para cálculos de mejor estimación ("best estimate") que no incluyen aproximaciones conservadoras, son de una importancia primordial para analizar estacionarios y transitorios en reactores rápidos. Esta tesis se centra en el desarrollo de un código acoplado para realizar análisis realistas en reactores rápidos críticos aplicando el método de Monte Carlo. Hoy en día, dado el mayor potencial de recursos computacionales, los códigos de transporte neutrónico por Monte Carlo se pueden usar de manera práctica para realizar cálculos detallados de núcleos completos, incluso de elevada heterogeneidad material. Además, los códigos de Monte Carlo se toman normalmente como referencia para los códigos deterministas de difusión en multigrupos en aplicaciones con reactores rápidos, porque usan secciones eficaces punto a punto, un modelo geométrico exacto y tienen en cuenta intrínsecamente la dependencia angular de flujo. En esta tesis se presenta una metodología de acoplamiento entre el conocido código MCNP, que calcula la generación de potencia en el reactor, y el código de termohidráulica de subcanal COBRA-IV, que obtiene las distribuciones de temperatura y densidad en el sistema. COBRA-IV es un código apropiado para aplicaciones en reactores rápidos ya que ha sido validado con resultados experimentales en haces de barras con sodio, incluyendo las correlaciones más apropiadas para metales líquidos. En una primera fase de la tesis, ambos códigos se han acoplado en estado estacionario utilizando un método iterativo con intercambio de archivos externos. El principal problema en el acoplamiento neutrónico y termohidráulico en estacionario con códigos de Monte Carlo es la manipulación de las secciones eficaces para tener en cuenta el ensanchamiento Doppler cuando la temperatura del combustible aumenta. Entre todas las opciones disponibles, en esta tesis se ha escogido la aproximación de pseudo materiales, y se ha comprobado que proporciona resultados aceptables en su aplicación con reactores rápidos. Por otro lado, los cambios geométricos originados por grandes gradientes de temperatura en el núcleo de reactores rápidos resultan importantes para la neutrónica como consecuencia del elevado recorrido libre medio del neutrón en estos sistemas. Por tanto, se ha desarrollado un módulo adicional que simula la geometría del reactor en caliente y permite estimar la reactividad debido a la expansión del núcleo en un transitorio. éste módulo calcula automáticamente la longitud del combustible, el radio de la vaina, la separación de los elementos de combustible y el radio de la placa soporte en función de la temperatura. éste efecto es muy relevante en transitorios sin inserción de bancos de parada. También relacionado con los cambios geométricos, se ha implementado una herramienta que, automatiza el movimiento de las barras de control en busca d la criticidad del reactor, o bien calcula el valor de inserción axial las barras de control. Una segunda fase en la plataforma de cálculo que se ha desarrollado es la simulació dinámica. Puesto que MCNP sólo realiza cálculos estacionarios para sistemas críticos o supercríticos, la solución más directa que se propone sin modificar el código fuente de MCNP es usar la aproximación de factorización de flujo, que resuelve por separado la forma del flujo y la amplitud. En este caso se han estudiado en profundidad dos aproximaciones: adiabática y quasiestática. El método adiabático usa un esquema de acoplamiento que alterna en el tiempo los cálculos neutrónicos y termohidráulicos. MCNP calcula el modo fundamental de la distribución de neutrones y la reactividad al final de cada paso de tiempo, y COBRA-IV calcula las propiedades térmicas en el punto intermedio de los pasos de tiempo. La evolución de la amplitud de flujo se calcula resolviendo las ecuaciones de cinética puntual. Este método calcula la reactividad estática en cada paso de tiempo que, en general, difiere de la reactividad dinámica que se obtendría con la distribución de flujo exacta y dependiente de tiempo. No obstante, para entornos no excesivamente alejados de la criticidad ambas reactividades son similares y el método conduce a resultados prácticos aceptables. Siguiendo esta línea, se ha desarrollado después un método mejorado para intentar tener en cuenta el efecto de la fuente de neutrones retardados en la evolución de la forma del flujo durante el transitorio. El esquema consiste en realizar un cálculo cuasiestacionario por cada paso de tiempo con MCNP. La simulación cuasiestacionaria se basa EN la aproximación de fuente constante de neutrones retardados, y consiste en dar un determinado peso o importancia a cada ciclo computacial del cálculo de criticidad con MCNP para la estimación del flujo final. Ambos métodos se han verificado tomando como referencia los resultados del código de difusión COBAYA3 frente a un ejercicio común y suficientemente significativo. Finalmente, con objeto de demostrar la posibilidad de uso práctico del código, se ha simulado un transitorio en el concepto de reactor crítico en fase de diseño MYRRHA/FASTEF, de 100 MW de potencia térmica y refrigerado por plomo-bismuto. ABSTRACT Long term sustainable nuclear energy scenarios envisage a fleet of Liquid Metal Fast Reactors (LMFR) for the Pu recycling and minor actinides (MAs) transmutation or combined with some accelerator driven systems (ADS) just for MAs elimination. Design and licensing of these innovative reactor concepts require accurate computational tools, implementing the knowledge obtained in experimental research for new reactor configurations, materials and associated systems. Although a number of fast reactor systems have already been built, the operational experience is still reduced, especially for lead reactors, and not all the transients are fully understood. The safety analysis approach for LMFR is therefore based only on deterministic methods, different from modern approach for Light Water Reactors (LWR) which also benefit from probabilistic methods. Usually, the approach adopted in LMFR safety assessments is to employ a variety of codes, somewhat different for the each other, to analyze transients looking for a comprehensive solution and including uncertainties. In this frame, new best estimate simulation codes are of prime importance in order to analyze fast reactors steady state and transients. This thesis is focused on the development of a coupled code system for best estimate analysis in fast critical reactor. Currently due to the increase in the computational resources, Monte Carlo methods for neutrons transport can be used for detailed full core calculations. Furthermore, Monte Carlo codes are usually taken as reference for deterministic diffusion multigroups codes in fast reactors applications because they employ point-wise cross sections in an exact geometry model and intrinsically account for directional dependence of the ux. The coupling methodology presented here uses MCNP to calculate the power deposition within the reactor. The subchannel code COBRA-IV calculates the temperature and density distribution within the reactor. COBRA-IV is suitable for fast reactors applications because it has been validated against experimental results in sodium rod bundles. The proper correlations for liquid metal applications have been added to the thermal-hydraulics program. Both codes are coupled at steady state using an iterative method and external files exchange. The main issue in the Monte Carlo/thermal-hydraulics steady state coupling is the cross section handling to take into account Doppler broadening when temperature rises. Among every available options, the pseudo materials approach has been chosen in this thesis. This approach obtains reasonable results in fast reactor applications. Furthermore, geometrical changes caused by large temperature gradients in the core, are of major importance in fast reactor due to the large neutron mean free path. An additional module has therefore been included in order to simulate the reactor geometry in hot state or to estimate the reactivity due to core expansion in a transient. The module automatically calculates the fuel length, cladding radius, fuel assembly pitch and diagrid radius with the temperature. This effect will be crucial in some unprotected transients. Also related to geometrical changes, an automatic control rod movement feature has been implemented in order to achieve a just critical reactor or to calculate control rod worth. A step forward in the coupling platform is the dynamic simulation. Since MCNP performs only steady state calculations for critical systems, the more straight forward option without modifying MCNP source code, is to use the flux factorization approach solving separately the flux shape and amplitude. In this thesis two options have been studied to tackle time dependent neutronic simulations using a Monte Carlo code: adiabatic and quasistatic methods. The adiabatic methods uses a staggered time coupling scheme for the time advance of neutronics and the thermal-hydraulics calculations. MCNP computes the fundamental mode of the neutron flux distribution and the reactivity at the end of each time step and COBRA-IV the thermal properties at half of the the time steps. To calculate the flux amplitude evolution a solver of the point kinetics equations is used. This method calculates the static reactivity in each time step that in general is different from the dynamic reactivity calculated with the exact flux distribution. Nevertheless, for close to critical situations, both reactivities are similar and the method leads to acceptable practical results. In this line, an improved method as an attempt to take into account the effect of delayed neutron source in the transient flux shape evolutions is developed. The scheme performs a quasistationary calculation per time step with MCNP. This quasistationary simulations is based con the constant delayed source approach, taking into account the importance of each criticality cycle in the final flux estimation. Both adiabatic and quasistatic methods have been verified against the diffusion code COBAYA3, using a theoretical kinetic exercise. Finally, a transient in a critical 100 MWth lead-bismuth-eutectic reactor concept is analyzed using the adiabatic method as an application example in a real system.

Strategies to unblock the n-GaAs surface when electrodepositing Bi from acidic solutions

Bismuth ultra-thin films grown on n-GaAs electrodes via electrodeposition are porous due to a blockade of the electrode surface caused by adsorbed hydrogen when using acidic electrolytes. In this study, we discuss the existence of two sources of hydrogen adsorption and we propose different routes to unblock the n-GaAs surface in order to improve Bi films compactness. Firstly, we demonstrate that increasing the electrolyte temperature provides compact yet polycrystalline Bi films. Cyclic voltammetry scans indicate that this low crystal quality might be a result of the incorporation of Bi hydroxides within the Bi film as a result of the temperature increase. Secondly, we have illuminated the semiconductor surface to take advantage of photogenerated holes. These photocarriers oxidize the adsorbed hydrogen unblocking the surface, but also create pits at the substrate surface that degrade the Bi/GaAs interface and prevent an epitaxial growth. Finally, we show that performing a cyclic voltammetry scan before electrodeposition enables the growth of compact Bi ultra-thin films of high crystallinity on semiconductor substrates with a doping level low enough to perform transport measurements.