Viabilidad técnica para la explotación y separación isotópica de Li y prospectiva en el mercado de baterías de ION-7Li y de 6Li como material base de diseño en reactores de fusión nuclear

En claro alineamiento con estrategias de sostenibilidad en el uso de recursos naturales en un escenario constante de aumento de la demanda energética mundial, el desarrollo de la tecnología energética en la Historia de la Especie Humana muestra un vector de evolución permanente desde su origen en el sentido del desarrollo y uso de nuevas fuentes energéticas con la explotación de recursos naturales de manera más eficiente: soluciones energéticas con aumento de la densidad energética (exoenergía de proceso por unidad de masa de recurso natural). Así el cambio de escala en la demanda de explotación del Litio como recurso natural se viene presentando en la última década ligada al desarrollo del mercado de las baterías "ion-Litio" y los requisitos de combustible (Deuterio y Litio) en el camino de la fusión nuclear como opción energética próxima. El análisis anticipado de las demandas sinérgicas a escala de ambos mercados aparece de enorme interés prospectivo en sus aspectos técnicos: (1) tecnologías de base para la extracción mineral y de agua marina y (2) su enriquecimiento isotópico (de interés sinérgico; 7Li para baterías eficientes ion-litio; 6Li como regenerador de tritio en ciclo de combustible en fusión nuclear) a la vez que en sus aspectos económicos. Este Proyecto realiza: (1) un ejercicio de análisis prospectivo de la demanda y de mercado para el enriquecimiento 6Li/7Li para las próximas décadas, (2) se califican los desarrollos tecnológicos específicos que van a poder permitir la producción a escala conforme a la demanda; (3) se selecciona y califica una técnica [de centrifugación / termo-difusión/ destilación combinada] como opción tecnológicamente viable para la producción a escala de formas litiadas; (4) se propone un diseño conceptual de planta de producción y finalmente (5) propone un estudio de viabilidad para la demostración de proceso y construcción de dicha planta de demostración de la nueva capacidad tecnológica. ABSTRACT Clearly aligned with sustainability strategies under growing world energy demand in the use of natural resources the development of energy technology in the history of the human species shows a vector of ongoing evolution from its origin in the sense of the development and use of new energy sources with the exploitation of natural resources in a more efficient manner. The change of scale in the demand for exploitation of Lithium as a natural resource appears during the last decade as bound to the deployment of "lithium-ion" batteries market and to the Nuclear Fusion fuels (deuterium and lithium) supply scaled demands. The prospective analysis of demands to scale in both markets appears in scene with huge prospective interest in its technical aspects: (1) base technologies for mineral and water marine extraction (2) its isotopic enrichment (synergistic interests; 7Li efficient battery Li-ion; 6Li as fusion nuclear fuel breeder (tritium) as well as in its economic aspects. This Project: (1) propose a prospective analysis exercise of the synergistic supply demand for coming decades for the enrichment of 6Li and 7Li, (2) qualifies specific technological developments ongoing to respond to supply demand; (3) select and qualifies an appropriate technique [combined centrifugation/thermo-diffusion/distillation] as technologically viable option for lithiated forms scaled-production; (4) proposes a conceptual design of production plant based on the technique and finally (5) proposes a feasibility study for the process demonstration and construction of this new technological capability Demonstration Plant.

Neutronics and activation of the preliminary reaction chamber of HiPER reactor based in a SCLL blanket

The HiPER reactor design is exploring different reaction chambers. In this study, we tackle the neutronicsand activation studies of a preliminary reaction chamber based in the following technologies: unpro-tected dry wall for the First Wall, self-cooled lead lithium blanket, and independent low activation steelVacuum Vessel. The most critical free parameter in this stage is the blanket thickness, as a function ofthe6Li enrichment. After a parametric study, we select for study both a ?thin? and ?thick? blanket, with?high? and ?low?6Li enrichment respectively, to reach a TBR = 1.1. To help to make a choice, we com-pute, for both blanket options, in addition to the TBR, the energy amplification factor, the tritium partialpressure, the203Hg and210Po total activity in the LiPb loop, and the Vacuum Vessel thickness requiredto guarantee the reweldability during its lifetime. The thin blanket shows a superior performance in thesafety related issues and structural viability, but it operates at higher6Li enrichment. It is selected forfurther improvements. The Vacuum Vessel shows to be unviable in both cases, with the thickness varyingbetween 39 and 52 cm. Further chamber modifications, such as the introduction of a neutron reflector,are required to exploit the benefits of the thin blanket with a reasonable Vacuum Vessel.

Symmetry issues in Directly Irradiated Targets

In direct drive Inertial Confinement Fusion (ICF), the typical laser beam to laser beam angle is around 30o. This fact makes the study of the irradiation symmetry agenuine 3D problem. In this paper we use the three dimensional version of the MULTI hydrocode to assess the symmetry of such ICF implosions. More specifically, we study a shock-ignition proposal for the Laser-M´egajoule facility (LMJ) in which two of the equatorial beam cones are used to implode and pre compress a spherical capsule (the “reference” capsule of HiPER project) made of 0.59 mg of pure Deuterium-Tritium mixture. The symmetry of this scheme is analysed and optimized to get a design inside the operating limits of LMJ. The studied configuration has been found essentially axial-symmetric, so that the use of 2D hydrocodes would be appropriate for this specific situation

Modelización y validación experimental del transporte de tritio en metal líquido y otros materiales para reactores de fusión nuclear magnética e inercial

La presente tesis se centra en el estudio de los fenómenos de transporte de los isótopos de hidrógeno, y más concretamente del tritio, en materiales de interés para los reactores de fusión nuclear. Los futuros reactores de fusión nuclear necesitarán una Planta de Tritio, con una envoltura regeneradora (breeding blanket) y unos sistemas auxiliares claves para su diseño. Por lo tanto su desarrollo y cualificación son cruciales para demostrar que los reactores de fusión son una opción viable como futura fuente de energía. Se han resaltado los diferentes retos de la difusión y retención de estas especies ligeras para cada sistema de la Planta de Tritio, y se han identificado las necesidades experimentales y paramétricas para abordar las simulaciones de difusión, como factores de transporte como la difusividad, absorción/desorción, solubilidad y atrapamiento. Se han estudiado los fenómenos de transporte y parámetros del T en el metal líquido LiPb, componente del breeding blanket tanto para una planta de fusión magnética como inercial. Para ello se han utilizado dos experimentos con características diversas, uno de ellos se ha llevado a cabo en un reactor de alto flujo, y por lo tanto, en condiciones de irradiación, y el otro sin irradiación. Los métodos de simulación numérica aplicados se han adaptado a los experimentos para las mediciones y para estudiar el régimen de transporte. En el estudio de estos experimentos se ha obtenido un valor para algunos de los parámetros claves en el transporte y gestión del tritio en el reactor. Finalmente se realiza un cálculo de la acumulación y difusión de tritio en una primera pared de tungsteno para un reactor de fusión inercial. En concreto para el proyecto de fusión por láser europeo, HiPER (para sus fases 4a y 4b). Se ha estudiado: la implantación de los isótopos de H y He en la pared de W tras una reacción de fusión por iluminación directa con un láser de 48MJ; el efecto en el transporte de T de los picos de temperatura en el W debido a la frecuencia de los eventos de fusión; el régimen de transporte en la primera pared. Se han identificado la naturaleza de las trampas más importantes para el T y se ha propuesto un modelo avanzado para la difusión con trampas. ABSTRACT The present thesis focuses into study the transport phenomenons of hydrogen isotopes, more specifically tritium, in materials of interest for nuclear fusion reactors. The future nuclear reactors will be provided of a Tritium Plant, with its breeding blanket and its auxiliary systems, all of them essential components for the plant. Therefore a reliable development and coalification are key issues to prove the viability of the nuclear fusion reactors as an energy source. The currently challenges for the diffusion and accumulation of these light species for each system of the TP has been studied. Experimental and theoretical needs have been identified and analyzed, specially from the viewpoint of the parameters. To achieve reliable simulations of tritium transport, parameters as diffusivity, absorption/desorption, solubility and trapping must be reliables. Transport phenomenon and parameters of T in liquid metal have been studied. Lead lithium is a key component of the breeding blanket, either in magnetic or inertial fusion confinement. Having this aim in mind, two experiments with different characteristics have been used; one of them has been realized in a high flux reactor, and hence, under irradiation conditions. The other one has been realized without radiation. The mathematical methods for the simulation have been adapted to the experiments, for the measures and also to study the transport behavior. A value for some key parameters for tritium management has been obtained in these studies. Finally, tritium accumulation and diffusion in a W first wall of an inertial nuclear fusion reactor has been assessed. A diffusion model of the implanted H, D, T and He species for the two initial phases of the proposed European laser fusion Project HiPER (namely, phase 4a and phase 4b) has been implemented using Tritium Migration Analysis Program, TMAP7. The effect of the prompt and working temperatures and the operational pulsing modes on the diffusion are studied. The nature of tritium traps in W and their performance has been analyzed and discussed.

Fast ignition driven by quasi-monoenergetic ions: Optimal ion type and reduction of ignition energies with an ion beam array

Fast ignition of inertial fusion targets driven by quasi-monoenergetic ion beams is investigated by means of numerical simulations. Light and intermediate ions such as lithium, carbon, aluminum and vanadium have been considered. Simulations show that the minimum ignition energies of an ideal configuration of compressed Deuterium-Tritium are almost independent on the ion atomic number. However, they are obtained for increasing ion energies, which scale, approximately, as Z2, where Z is the ion atomic number. Assuming that the ion beam can be focused into 10 ?m spots, a new irradiation scheme is proposed to reduce the ignition energies. The combination of intermediate Z ions, such as 5.5 GeV vanadium, and the new irradiation scheme allows a reduction of the number of ions required for ignition by, roughly, three orders of magnitude when compared with the standard proton fast ignition scheme.

Dust tracking techniques applied to the STARDUST facility: First results

An important issue related to future nuclear fusion reactors fueled with deuterium and tritium is the creation of large amounts of dust due to several mechanisms (disruptions, ELMs and VDEs). The dust size expected in nuclear fusion experiments (such as ITER) is in the order of microns (between 0.1 and 1000 μm). Almost the total amount of this dust remains in the vacuum vessel (VV). This radiological dust can re-suspend in case of LOVA (loss of vacuum accident) and these phenomena can cause explosions and serious damages to the health of the operators and to the integrity of the device. The authors have developed a facility, STARDUST, in order to reproduce the thermo fluid-dynamic conditions comparable to those expected inside the VV of the next generation of experiments such as ITER in case of LOVA. The dust used inside the STARDUST facility presents particle sizes and physical characteristics comparable with those that created inside the VV of nuclear fusion experiments. In this facility an experimental campaign has been conducted with the purpose of tracking the dust re-suspended at low pressurization rates (comparable to those expected in case of LOVA in ITER and suggested by the General Safety and Security Report ITER-GSSR) using a fast camera with a frame rate from 1000 to 10,000 images per second. The velocity fields of the mobilized dust are derived from the imaging of a two-dimensional slice of the flow illuminated by optically adapted laser beam. The aim of this work is to demonstrate the possibility of dust tracking by means of image processing with the objective of determining the velocity field values of dust re-suspended during a LOVA.

Diseño de un módulo de IFMIF para irradiación de envolturas regeneradoras basadas en metales líquidos para reactores de fusión

La fusión nuclear es, hoy en día, una alternativa energética a la que la comunidad internacional dedica mucho esfuerzo. El objetivo es el de generar entre diez y cincuenta veces más energía que la que consume mediante reacciones de fusión que se producirán en una mezcla de deuterio (D) y tritio (T) en forma de plasma a doscientos millones de grados centígrados. En los futuros reactores nucleares de fusión será necesario producir el tritio utilizado como combustible en el propio reactor termonuclear. Este hecho supone dar un paso más que las actuales máquinas experimentales dedicadas fundamentalmente al estudio de la física del plasma. Así pues, el tritio, en un reactor de fusión, se produce en sus envolturas regeneradoras cuya misión fundamental es la de blindaje neutrónico, producir y recuperar tritio (fuel para la reacción DT del plasma) y por último convertir la energía de los neutrones en calor. Existen diferentes conceptos de envolturas que pueden ser sólidas o líquidas. Las primeras se basan en cerámicas de litio (Li2O, Li4SiO4, Li2TiO3, Li2ZrO3) y multiplicadores neutrónicos de Be, necesarios para conseguir la cantidad adecuada de tritio. Los segundos se basan en el uso de metales líquidos o sales fundidas (Li, LiPb, FLIBE, FLINABE) con multiplicadores neutrónicos de Be o el propio Pb en el caso de LiPb. Los materiales estructurales pasan por aceros ferrítico-martensíticos de baja activación, aleaciones de vanadio o incluso SiCf/SiC. Cada uno de los diferentes conceptos de envoltura tendrá una problemática asociada que se estudiará en el reactor experimental ITER (del inglés, “International Thermonuclear Experimental Reactor”). Sin embargo, ITER no puede responder las cuestiones asociadas al daño de materiales y el efecto de la radiación neutrónica en las diferentes funciones de las envolturas regeneradoras. Como referencia, la primera pared de un reactor de fusión de 4000MW recibiría 30 dpa/año (valores para Fe-56) mientras que en ITER se conseguirían <10 dpa en toda su vida útil. Esta tesis se encuadra en el acuerdo bilateral entre Europa y Japón denominado “Broader Approach Agreement “(BA) (2007-2017) en el cual España juega un papel destacable. Estos proyectos, complementarios con ITER, son el acelerador para pruebas de materiales IFMIF (del inglés, “International Fusion Materials Irradiation Facility”) y el dispositivo de fusión JT-60SA. Así, los efectos de la irradiación de materiales en materiales candidatos para reactores de fusión se estudiarán en IFMIF. El objetivo de esta tesis es el diseño de un módulo de IFMIF para irradiación de envolturas regeneradoras basadas en metales líquidos para reactores de fusión. El módulo se llamará LBVM (del inglés, “Liquid Breeder Validation Module”). La propuesta surge de la necesidad de irradiar materiales funcionales para envolturas regeneradoras líquidas para reactores de fusión debido a que el diseño conceptual de IFMIF no contaba con esta utilidad. Con objeto de analizar la viabilidad de la presente propuesta, se han realizado cálculos neutrónicos para evaluar la idoneidad de llevar a cabo experimentos relacionados con envolturas líquidas en IFMIF. Así, se han considerado diferentes candidatos a materiales funcionales de envolturas regeneradoras: Fe (base de los materiales estructurales), SiC (material candidato para los FCI´s (del inglés, “Flow Channel Inserts”) en una envoltura regeneradora líquida, SiO2 (candidato para recubrimientos antipermeación), CaO (candidato para recubrimientos aislantes), Al2O3 (candidato para recubrimientos antipermeación y aislantes) y AlN (material candidato para recubrimientos aislantes). En cada uno de estos materiales se han calculado los parámetros de irradiación más significativos (dpa, H/dpa y He/dpa) en diferentes posiciones de IFMIF. Estos valores se han comparado con los esperados en la primera pared y en la zona regeneradora de tritio de un reactor de fusión. Para ello se ha elegido un reactor tipo HCLL (del inglés, “Helium Cooled Lithium Lead”) por tratarse de uno de los más prometedores. Además, los valores también se han comparado con los que se obtendrían en un reactor rápido de fisión puesto que la mayoría de las irradiaciones actuales se hacen en reactores de este tipo. Como conclusión al análisis de viabilidad, se puede decir que los materiales funcionales para mantos regeneradores líquidos podrían probarse en la zona de medio flujo de IFMIF donde se obtendrían ratios de H/dpa y He/dpa muy parecidos a los esperados en las zonas más irradiadas de un reactor de fusión. Además, con el objetivo de ajustar todavía más los valores, se propone el uso de un moderador de W (a considerar en algunas campañas de irradiación solamente debido a que su uso hace que los valores de dpa totales disminuyan). Los valores obtenidos para un reactor de fisión refuerzan la idea de la necesidad del LBVM, ya que los valores obtenidos de H/dpa y He/dpa son muy inferiores a los esperados en fusión y, por lo tanto, no representativos. Una vez demostrada la idoneidad de IFMIF para irradiar envolturas regeneradoras líquidas, y del estudio de la problemática asociada a las envolturas líquidas, también incluida en esta tesis, se proponen tres tipos de experimentos diferentes como base de diseño del LBVM. Éstos se orientan en las necesidades de un reactor tipo HCLL aunque a lo largo de la tesis se discute la aplicabilidad para otros reactores e incluso se proponen experimentos adicionales. Así, la capacidad experimental del módulo estaría centrada en el estudio del comportamiento de litio plomo, permeación de tritio, corrosión y compatibilidad de materiales. Para cada uno de los experimentos se propone un esquema experimental, se definen las condiciones necesarias en el módulo y la instrumentación requerida para controlar y diagnosticar las cápsulas experimentales. Para llevar a cabo los experimentos propuestos se propone el LBVM, ubicado en la zona de medio flujo de IFMIF, en su celda caliente, y con capacidad para 16 cápsulas experimentales. Cada cápsula (24-22 mm de diámetro y 80 mm de altura) contendrá la aleación eutéctica LiPb (hasta 50 mm de la altura de la cápsula) en contacto con diferentes muestras de materiales. Ésta irá soportada en el interior de tubos de acero por los que circulará un gas de purga (He), necesario para arrastrar el tritio generado en el eutéctico y permeado a través de las paredes de las cápsulas (continuamente, durante irradiación). Estos tubos, a su vez, se instalarán en una carcasa también de acero que proporcionará soporte y refrigeración tanto a los tubos como a sus cápsulas experimentales interiores. El módulo, en su conjunto, permitirá la extracción de las señales experimentales y el gas de purga. Así, a través de la estación de medida de tritio y el sistema de control, se obtendrán los datos experimentales para su análisis y extracción de conclusiones experimentales. Además del análisis de datos experimentales, algunas de estas señales tendrán una función de seguridad y por tanto jugarán un papel primordial en la operación del módulo. Para el correcto funcionamiento de las cápsulas y poder controlar su temperatura, cada cápsula se equipará con un calentador eléctrico y por tanto el módulo requerirá también ser conectado a la alimentación eléctrica. El diseño del módulo y su lógica de operación se describe en detalle en esta tesis. La justificación técnica de cada una de las partes que componen el módulo se ha realizado con soporte de cálculos de transporte de tritio, termohidráulicos y mecánicos. Una de las principales conclusiones de los cálculos de transporte de tritio es que es perfectamente viable medir el tritio permeado en las cápsulas mediante cámaras de ionización y contadores proporcionales comerciales, con sensibilidades en el orden de 10-9 Bq/m3. Los resultados son aplicables a todos los experimentos, incluso si son cápsulas a bajas temperaturas o si llevan recubrimientos antipermeación. Desde un punto de vista de seguridad, el conocimiento de la cantidad de tritio que está siendo transportada con el gas de purga puede ser usado para detectar de ciertos problemas que puedan estar sucediendo en el módulo como por ejemplo, la rotura de una cápsula. Además, es necesario conocer el balance de tritio de la instalación. Las pérdidas esperadas el refrigerante y la celda caliente de IFMIF se pueden considerar despreciables para condiciones normales de funcionamiento. Los cálculos termohidráulicos se han realizado con el objetivo de optimizar el diseño de las cápsulas experimentales y el LBVM de manera que se pueda cumplir el principal requisito del módulo que es llevar a cabo los experimentos a temperaturas comprendidas entre 300-550ºC. Para ello, se ha dimensionado la refrigeración necesaria del módulo y evaluado la geometría de las cápsulas, tubos experimentales y la zona experimental del contenedor. Como consecuencia de los análisis realizados, se han elegido cápsulas y tubos cilíndricos instalados en compartimentos cilíndricos debido a su buen comportamiento mecánico (las tensiones debidas a la presión de los fluidos se ven reducidas significativamente con una geometría cilíndrica en lugar de prismática) y térmico (uniformidad de temperatura en las paredes de los tubos y cápsulas). Se han obtenido campos de presión, temperatura y velocidad en diferentes zonas críticas del módulo concluyendo que la presente propuesta es factible. Cabe destacar que el uso de códigos fluidodinámicos (e.g. ANSYS-CFX, utilizado en esta tesis) para el diseño de cápsulas experimentales de IFMIF no es directo. La razón de ello es que los modelos de turbulencia tienden a subestimar la temperatura de pared en mini canales de helio sometidos a altos flujos de calor debido al cambio de las propiedades del fluido cerca de la pared. Los diferentes modelos de turbulencia presentes en dicho código han tenido que ser estudiados con detalle y validados con resultados experimentales. El modelo SST (del inglés, “Shear Stress Transport Model”) para turbulencia en transición ha sido identificado como adecuado para simular el comportamiento del helio de refrigeración y la temperatura en las paredes de las cápsulas experimentales. Con la geometría propuesta y los valores principales de refrigeración y purga definidos, se ha analizado el comportamiento mecánico de cada uno de los tubos experimentales que contendrá el módulo. Los resultados de tensiones obtenidos, han sido comparados con los valores máximos recomendados en códigos de diseño estructural como el SDC-IC (del inglés, “Structural Design Criteria for ITER Components”) para así evaluar el grado de protección contra el colapso plástico. La conclusión del estudio muestra que la propuesta es mecánicamente robusta. El LBVM implica el uso de metales líquidos y la generación de tritio además del riesgo asociado a la activación neutrónica. Por ello, se han estudiado los riesgos asociados al uso de metales líquidos y el tritio. Además, se ha incluido una evaluación preliminar de los riesgos radiológicos asociados a la activación de materiales y el calor residual en el módulo después de la irradiación así como un escenario de pérdida de refrigerante. Los riesgos asociados al módulo de naturaleza convencional están asociados al manejo de metales líquidos cuyas reacciones con aire o agua se asocian con emisión de aerosoles y probabilidad de fuego. De entre los riesgos nucleares destacan la generación de gases radiactivos como el tritio u otros radioisótopos volátiles como el Po-210. No se espera que el módulo suponga un impacto medioambiental asociado a posibles escapes. Sin embargo, es necesario un manejo adecuado tanto de las cápsulas experimentales como del módulo contenedor así como de las líneas de purga durante operación. Después de un día de después de la parada, tras un año de irradiación, tendremos una dosis de contacto de 7000 Sv/h en la zona experimental del contenedor, 2300 Sv/h en la cápsula y 25 Sv/h en el LiPb. El uso por lo tanto de manipulación remota está previsto para el manejo del módulo irradiado. Por último, en esta tesis se ha estudiado también las posibilidades existentes para la fabricación del módulo. De entre las técnicas propuestas, destacan la electroerosión, soldaduras por haz de electrones o por soldadura láser. Las bases para el diseño final del LBVM han sido pues establecidas en el marco de este trabajo y han sido incluidas en el diseño intermedio de IFMIF, que será desarrollado en el futuro, como parte del diseño final de la instalación IFMIF. ABSTRACT Nuclear fusion is, today, an alternative energy source to which the international community devotes a great effort. The goal is to generate 10 to 50 times more energy than the input power by means of fusion reactions that occur in deuterium (D) and tritium (T) plasma at two hundred million degrees Celsius. In the future commercial reactors it will be necessary to breed the tritium used as fuel in situ, by the reactor itself. This constitutes a step further from current experimental machines dedicated mainly to the study of the plasma physics. Therefore, tritium, in fusion reactors, will be produced in the so-called breeder blankets whose primary mission is to provide neutron shielding, produce and recover tritium and convert the neutron energy into heat. There are different concepts of breeding blankets that can be separated into two main categories: solids or liquids. The former are based on ceramics containing lithium as Li2O , Li4SiO4 , Li2TiO3 , Li2ZrO3 and Be, used as a neutron multiplier, required to achieve the required amount of tritium. The liquid concepts are based on molten salts or liquid metals as pure Li, LiPb, FLIBE or FLINABE. These blankets use, as neutron multipliers, Be or Pb (in the case of the concepts based on LiPb). Proposed structural materials comprise various options, always with low activation characteristics, as low activation ferritic-martensitic steels, vanadium alloys or even SiCf/SiC. Each concept of breeding blanket has specific challenges that will be studied in the experimental reactor ITER (International Thermonuclear Experimental Reactor). However, ITER cannot answer questions associated to material damage and the effect of neutron radiation in the different breeding blankets functions and performance. As a reference, the first wall of a fusion reactor of 4000 MW will receive about 30 dpa / year (values for Fe-56) , while values expected in ITER would be <10 dpa in its entire lifetime. Consequently, the irradiation effects on candidate materials for fusion reactors will be studied in IFMIF (International Fusion Material Irradiation Facility). This thesis fits in the framework of the bilateral agreement among Europe and Japan which is called “Broader Approach Agreement “(BA) (2007-2017) where Spain plays a key role. These projects, complementary to ITER, are mainly IFMIF and the fusion facility JT-60SA. The purpose of this thesis is the design of an irradiation module to test candidate materials for breeding blankets in IFMIF, the so-called Liquid Breeder Validation Module (LBVM). This proposal is born from the fact that this option was not considered in the conceptual design of the facility. As a first step, in order to study the feasibility of this proposal, neutronic calculations have been performed to estimate irradiation parameters in different materials foreseen for liquid breeding blankets. Various functional materials were considered: Fe (base of structural materials), SiC (candidate material for flow channel inserts, SiO2 (candidate for antipermeation coatings), CaO (candidate for insulating coatings), Al2O3 (candidate for antipermeation and insulating coatings) and AlN (candidate for insulation coating material). For each material, the most significant irradiation parameters have been calculated (dpa, H/dpa and He/dpa) in different positions of IFMIF. These values were compared to those expected in the first wall and breeding zone of a fusion reactor. For this exercise, a HCLL (Helium Cooled Lithium Lead) type was selected as it is one of the most promising options. In addition, estimated values were also compared with those obtained in a fast fission reactor since most of existing irradiations have been made in these installations. The main conclusion of this study is that the medium flux area of IFMIF offers a good irradiation environment to irradiate functional materials for liquid breeding blankets. The obtained ratios of H/dpa and He/dpa are very similar to those expected in the most irradiated areas of a fusion reactor. Moreover, with the aim of bringing the values further close, the use of a W moderator is proposed to be used only in some experimental campaigns (as obviously, the total amount of dpa decreases). The values of ratios obtained for a fission reactor, much lower than in a fusion reactor, reinforce the need of LBVM for IFMIF. Having demonstrated the suitability of IFMIF to irradiate functional materials for liquid breeding blankets, and an analysis of the main problems associated to each type of liquid breeding blanket, also presented in this thesis, three different experiments are proposed as basis for the design of the LBVM. These experiments are dedicated to the needs of a blanket HCLL type although the applicability of the module for other blankets is also discussed. Therefore, the experimental capability of the module is focused on the study of the behavior of the eutectic alloy LiPb, tritium permeation, corrosion and material compatibility. For each of the experiments proposed an experimental scheme is given explaining the different module conditions and defining the required instrumentation to control and monitor the experimental capsules. In order to carry out the proposed experiments, the LBVM is proposed, located in the medium flux area of the IFMIF hot cell, with capability of up to 16 experimental capsules. Each capsule (24-22 mm of diameter, 80 mm high) will contain the eutectic allow LiPb (up to 50 mm of capsule high) in contact with different material specimens. They will be supported inside rigs or steel pipes. Helium will be used as purge gas, to sweep the tritium generated in the eutectic and permeated through the capsule walls (continuously, during irradiation). These tubes, will be installed in a steel container providing support and cooling for the tubes and hence the inner experimental capsules. The experimental data will consist of on line monitoring signals and the analysis of purge gas by the tritium measurement station. In addition to the experimental signals, the module will produce signals having a safety function and therefore playing a major role in the operation of the module. For an adequate operation of the capsules and to control its temperature, each capsule will be equipped with an electrical heater so the module will to be connected to an electrical power supply. The technical justification behind the dimensioning of each of these parts forming the module is presented supported by tritium transport calculations, thermalhydraulic and structural analysis. One of the main conclusions of the tritium transport calculations is that the measure of the permeated tritium is perfectly achievable by commercial ionization chambers and proportional counters with sensitivity of 10-9 Bq/m3. The results are applicable to all experiments, even to low temperature capsules or to the ones using antipermeation coatings. From a safety point of view, the knowledge of the amount of tritium being swept by the purge gas is a clear indicator of certain problems that may be occurring in the module such a capsule rupture. In addition, the tritium balance in the installation should be known. Losses of purge gas permeated into the refrigerant and the hot cell itself through the container have been assessed concluding that they are negligible for normal operation. Thermal hydraulic calculations were performed in order to optimize the design of experimental capsules and LBVM to fulfill one of the main requirements of the module: to perform experiments at uniform temperatures between 300-550ºC. The necessary cooling of the module and the geometry of the capsules, rigs and testing area of the container were dimensioned. As a result of the analyses, cylindrical capsules and rigs in cylindrical compartments were selected because of their good mechanical behavior (stresses due to fluid pressure are reduced significantly with a cylindrical shape rather than prismatic) and thermal (temperature uniformity in the walls of the tubes and capsules). Fields of pressure, temperature and velocity in different critical areas of the module were obtained concluding that the proposal is feasible. It is important to mention that the use of fluid dynamic codes as ANSYS-CFX (used in this thesis) for designing experimental capsules for IFMIF is not direct. The reason for this is that, under strongly heated helium mini channels, turbulence models tend to underestimate the wall temperature because of the change of helium properties near the wall. Therefore, the different code turbulence models had to be studied in detail and validated against experimental results. ANSYS-CFX SST (Shear Stress Transport Model) for transitional turbulence model has been identified among many others as the suitable one for modeling the cooling helium and the temperature on the walls of experimental capsules. Once the geometry and the main purge and cooling parameters have been defined, the mechanical behavior of each experimental tube or rig including capsules is analyzed. Resulting stresses are compared with the maximum values recommended by applicable structural design codes such as the SDC- IC (Structural Design Criteria for ITER Components) in order to assess the degree of protection against plastic collapse. The conclusion shows that the proposal is mechanically robust. The LBVM involves the use of liquid metals, tritium and the risk associated with neutron activation. The risks related with the handling of liquid metals and tritium are studied in this thesis. In addition, the radiological risks associated with the activation of materials in the module and the residual heat after irradiation are evaluated, including a scenario of loss of coolant. Among the identified conventional risks associated with the module highlights the handling of liquid metals which reactions with water or air are accompanied by the emission of aerosols and fire probability. Regarding the nuclear risks, the generation of radioactive gases such as tritium or volatile radioisotopes such as Po-210 is the main hazard to be considered. An environmental impact associated to possible releases is not expected. Nevertheless, an appropriate handling of capsules, experimental tubes, and container including purge lines is required. After one day after shutdown and one year of irradiation, the experimental area of the module will present a contact dose rate of about 7000 Sv/h, 2300 Sv/h in the experimental capsules and 25 Sv/h in the LiPb. Therefore, the use of remote handling is envisaged for the irradiated module. Finally, the different possibilities for the module manufacturing have been studied. Among the proposed techniques highlights the electro discharge machining, brazing, electron beam welding or laser welding. The bases for the final design of the LBVM have been included in the framework of the this work and included in the intermediate design report of IFMIF which will be developed in future, as part of the IFMIF facility final design.

NADH-quinone oxidoreductase: PSST subunit couples electron transfer from iron–sulfur cluster N2 to quinone

The proton-translocating NADH-quinone oxidoreductase (EC 1.6.99.3) is the largest and least understood enzyme complex of the respiratory chain. The mammalian mitochondrial enzyme (also called complex I) contains more than 40 subunits, whereas its structurally simpler bacterial counterpart (NDH-1) in Paracoccus denitrificans and Thermus thermophilus HB-8 consists of 14 subunits. A major unsolved question is the location and mechanism of the terminal electron transfer step from iron–sulfur cluster N2 to quinone. Potent inhibitors acting at this key region are candidate photoaffinity probes to dissect NADH-quinone oxidoreductases. Complex I and NDH-1 are very sensitive to inhibition by a variety of structurally diverse toxicants, including rotenone, piericidin A, bullatacin, and pyridaben. We designed (trifluoromethyl)diazirinyl[3H]pyridaben ([3H]TDP) as our photoaffinity ligand because it combines outstanding inhibitor potency, a suitable photoreactive group, and tritium at high specific activity. Photoaffinity labeling of mitochondrial electron transport particles was specific and saturable. Isolation, protein sequencing, and immunoprecipitation identified the high-affinity specifically labeled 23-kDa subunit as PSST of complex I. Immunoprecipitation of labeled membranes of P. denitrificans and T. thermophilus established photoaffinity labeling of the equivalent bacterial NQO6. Competitive binding and enzyme inhibition studies showed that photoaffinity labeling of the specific high-affinity binding site of PSST is exceptionally sensitive to each of the high-potency inhibitors mentioned above. These findings establish that the homologous PSST of mitochondria and NQO6 of bacteria have a conserved inhibitor-binding site and that this subunit plays a key role in electron transfer by functionally coupling iron–sulfur cluster N2 to quinone.

A protein residing at the subunit interface of the bacterial ribosome

Surface labeling of Escherichia coli ribosomes with the use of the tritium bombardment technique has revealed a minor unidentified ribosome-bound protein (spot Y) that is hidden in the 70S ribosome and becomes highly labeled on dissociation of the 70S ribosome into subunits. In the present work, the N-terminal sequence of the protein Y was determined and its gene was identified as yfia, an ORF located upstream the phe operon of E. coli. This 12.7-kDa protein was isolated and characterized. An affinity of the purified protein Y for the 30S subunit, but not for the 50S ribosomal subunit, was shown. The protein proved to be exposed on the surface of the 30S subunit. The attachment of the 50S subunit resulted in hiding the protein Y, thus suggesting the protein location at the subunit interface in the 70S ribosome. The protein was shown to stabilize ribosomes against dissociation. The possible role of the protein Y as ribosome association factor in translation is discussed.

Cyp7b, a novel brain cytochrome P450, catalyzes the synthesis of neurosteroids 7α-hydroxy dehydroepiandrosterone and 7α-hydroxy pregnenolone

Steroids produced locally in brain (neurosteroids), including dehydroepiandrosterone (DHEA), influence cognition and behavior. We previously described a novel cytochrome P450, Cyp7b, strongly expressed in rat and mouse brain, particularly in hippocampus. Cyp7b is most similar to steroidogenic P450s and potentially could play a role in neurosteroid metabolism. To examine the catalytic activity of the enzyme mouse Cyp7b cDNA was introduced into a vaccinia virus vector. Extracts from cells infected with the recombinant showed NADPH-dependent conversion of DHEA (Km, 13.6 μM) and pregnenolone (Km, 4.0 μM) to slower migrating forms on thin layer chromatography. The expressed enzyme was less active against 25-hydroxycholesterol, 17β-estradiol and 5α-androstane-3β,17β-diol, with low to undetectable activity against progesterone, corticosterone, and testosterone. On gas chromatography and mass spectrometry of the Cyp7b metabolite of DHEA the retention time and fragmentation patterns were identical to those obtained with authentic 7α-hydroxy DHEA. The reaction product also comigrated on thin layer chromatography with 7α-hydroxy DHEA but not with 7β-hydroxy DHEA; when [7α-3H]pregnenolone was incubated with Cyp7b extracts the extent of release of radioactivity into the medium suggested that hydroxylation was preferentially at the 7α position. Brain extracts also efficiently liberated tritium from [7α-3H]pregnenolone and converted DHEA to a product with a chromatographic mobility indistinguishable from 7α-hydroxy DHEA. We conclude that Cyp7b is a 7α-hydroxylase participating in the synthesis, in brain, of neurosteroids 7α-hydroxy DHEA, and 7α-hydroxy pregnenolone.

Locust adipokinetic hormones: carrier-independent transport and differential inactivation at physiological concentrations during rest and flight.

Since concomitant release of structurally related peptide hormones with apparently similar functions seems to be a general concept in endocrinology, we have studied the dynamics of the lifetime of the three known adipokinetic hormones (AKHs) of the migratory locust, which control flight-directed mobilization of carbohydrate and lipid from fat body stores. Although the structure of the first member of the AKHs has been known for 20 years, until now, reliable data on their inactivation and removal from the hemolymph are lacking, because measurement requires AKHs with high specific radioactivity. Employing tritiated AKHs with high specific radioactivity, obtained by catalytic reduction with tritium gas of the dehydroLeu2 analogues of the AKHs synthesized by the solid-phase procedure, studies with physiological doses of as low as 1.0 pmol per locust could be conducted. The AKHs appear to be transported in the hemolymph in their free forms and not associated with a carrier protein, despite their strong hydrophobicity. Application of AKHs in their free form in in vivo and in vitro studies therefore now has been justified. We have studied the degradation of the three AKHs during rest and flight. The first cleavage step by an endopeptidase is crucial, since the resulting degradation products lack any adipokinetic activity. Half-lives for AKH-I, -II and -III were 51, 40, and 5 min, respectively, for rest conditions and 35, 37, and 3 min, respectively, during flight. The rapid and differential degradation of structurally related hormones leads to changes in the ratio in which they are released and therefore will have important consequences for concerted hormone action at the level of the target organ or organs, suggesting that each of the known AKHs may play its own biological role in the overall syndrome of insect flight.

An analyzer for radioactive gas /

An instrument that can continuously monitor air for tritium has been developed. It utilizes a flow type ionization chamber and a vibrating reed electrometer. A response of double normal background is produced by 9.4 x 10⁻⁵ [mu]c H³/cc air; the presently accepted MPC is 2.0 x 10⁻⁵ [mu]c H³/cc air. Its operation is completely automatic.

A dynamic mathematical model for sequential leach bed anaerobic digestion of organic fraction of municipal solid waste

A mathematical model that describes the operation of a sequential leach bed process for anaerobic digestion of organic fraction of municipal solid waste (MSW) is developed and validated. This model assumes that ultimate mineralisation of the organic component of the waste occurs in three steps, namely solubilisation of particulate matter, fermentation to volatile organic acids (modelled as acetic acid) along with liberation of carbon dioxide and hydrogen, and methanogenesis from acetate and hydrogen. The model incorporates the ionic equilibrium equations arising due to dissolution of carbon dioxide, generation of alkalinity from breakdown of solids and dissociation of acetic acid. Rather than a charge balance, a mass balance on the hydronium and hydroxide ions is used to calculate pH. The flow of liquid through the bed is modelled as occurring through two zones-a permeable zone with high flushing rates and the other more stagnant. Some of the kinetic parameters for the biological processes were obtained from batch MSW digestion experiments. The parameters for flow model were obtained from residence time distribution studies conducted using tritium as a tracer. The model was validated using data from leach bed digestion experiments in which a leachate volume equal to 10% of the fresh waste bed volume was sequenced. The model was then tested, without altering any kinetic or flow parameters, by varying volume of leachate that is sequenced between the beds. Simulations for sequencing/recirculating 5 and 30% of the bed volume are presented and compared with experimental results. (C) 2002 Elsevier Science B.V. All rights reserved.

The sustained delivery of antisense oligodeoxynucleotides using biodegradable polymer microspheres

Antisense technology is a novel drug discovery method, which provides an essential tool for directly using gene sequence information to rationally design specific inhibitions of mRNA, to treat a wide range of diseases. The efficacy of naked oligodeoxynucleotides (ODNs) is relatively short lived due to rapid degradation in vivo. The entrapment of ODNs within biodegradable sustained-release delivery systems may improve ODN stability and reduce dose required for efficacy. Biodegradable polymer microspheres were evaluated as delivery devices for ODNs and ribozymes. Poly(lactide-co-glycolide) polymers were used due to their biocompatibility and non toxic degradation products. Microspheres were prepared using a double emulsion-deposition method and the formulations characterised. In vitro release profiles were characterised by an initial burst effect during the first 48 hours of release followed by a more sustained release. The release profiles were influenced by microsphere size, copolymer molecular weight, copolymer ratio, ODN loading, ODN length, and ODN chemistry. The serum stability of ODNs was significantly improved when entrapped within polymer microspheres. The cellular association of ODNs entrapped within small spheres (1-2μm) was improved by approximately 20-fold in A431 carcinoma cells compared with free ODNs. Fluorescence microscopy studies showed a more diffuse subcellular distribution when delivered as a microsphere formulation compared with free ODNs, which exhibited the characteristic punctate periplasmic distribution. For in vivo evaluation, polymer microspheres containing fluorescently-labelled ODNs were stereo-taxically administered to the neostriatum of the rat brain. Free ODN resulted in a punctate cellular distribution after 24 hours. In comparison ODN delivered using polymer microspheres were intensely visible in cells 48 hours post administration, and fluorescence appeared to be diffuse covering both cytosolic and nuclear regions. Whole-body autoradiography was also used to evaluate the biodistribution of free tritium labelled ODN and ODN entrapped microspheres, following subcutaneous administration to Balb-C mice. Polymer entrapped ODN gave a similar biodistribution to free ODN. Free ODN was distributed within 24 hours, whereas polymer released ODN was observed still presented in organs and at the site of administration seven days post administration.

Evidence for the removal of CFC-11, CFC-12, and CFC-113 at the groundwater–surface water interface in the Everglades

Poor agreement between 3H/3He ages and CFC-11 and CFC-12 ages suggests that CFCs may not be conservative tracers in the Everglades National Park. 3H/3He ages were used to calculate the expected concentration of CFC-11 and CFC-12 in groundwater from wells 2 to 73 m deep. The expected concentrations of CFCs were compared to the measured concentrations and plots of the % CFC-12 and CFC-11 remaining offered no evidence that significant CFC removal was occurring in the groundwater at depths ≥2 m, suggesting that CFC removal occurs at shallower depths. Except where CFC contamination was suspected, CFC-11, CFC-12 and CFC-113 concentrations in fresh surface water were nearly always below solubility equilibrium with the atmosphere. Measurements of CFC-11, CFC-12 and CFC-113 in pore water indicate a 50–90% decrease in concentration 5 cm below the groundwater–surface water (GW–SW) interface. In the same 5 cm interval CH4 concentrations increased by 300–1000%. This suggested that CFCs were removed at the GW–SW interface, possibly by methane-producing bacteria. CFC derived recharge ages should therefore be viewed with caution when recharging water percolates through anoxic methanogenic sediments.