Elastic (stress-strain) halo associated to ion-induced nano-tracks in lithium niobate: role of crystal anisotropy

The elastic strain/stress fields (halo) around a compressed amorphous nano-track (core) caused by a single high-energy ion impact on LiNbO3 are calculated. A method is developed to approximately account for the effects of crystal anisotropy of LiNbO3 (symmetry 3m) on the stress fields for tracks oriented along the crystal axes (X, Y or Z). It only considers the zero-order (axial) harmonic contribution to the displacement field in the perpendicular plane and uses effective Poisson moduli for each particular orientation. The anisotropy is relatively small; however, it accounts for some differential features obtained for irradiations along the crystallographic axes X, Y and Z. In particular, the irradiation-induced disorder (including halo) and the associated surface swelling appear to be higher for irradiations along the X- or Y-axis in comparison with those along the Z-axis. Other irradiation effects can be explained by the model, e.g. fracture patterns or the morphology of pores after chemical etching of tracks. Moreover, it offers interesting predictions on the effect of irradiation on lattice parameters

Recrystallization of amorphous nano-tracks and uniform layers generated by swift-ion-beam irradiation in lithium niobate.

The thermal annealing of amorphous tracks of nanometer-size diameter generated in lithium niobate (LiNbO3) by Bromine ions at 45 MeV, i.e., in the electronic stopping regime, has been investigated by RBS/C spectrometry in the temperature range from 250°C to 350°C. Relatively low fluences have been used (<1012 cm−2) to produce isolated tracks. However, the possible effect of track overlapping has been investigated by varying the fluence between 3×1011 cm−2 and 1012 cm−2. The annealing process follows a two-step kinetics. In a first stage (I) the track radius decreases linearly with the annealing time. It obeys an Arrhenius-type dependence on annealing temperature with activation energy around 1.5 eV. The second stage (II) operates after the track radius has decreased down to around 2.5 nm and shows a much lower radial velocity. The data for stage I appear consistent with a solid-phase epitaxial process that yields a constant recrystallization rate at the amorphous-crystalline boundary. HRTEM has been used to monitor the existence and the size of the annealed isolated tracks in the second stage. On the other hand, the thermal annealing of homogeneous (buried) amorphous layers has been investigated within the same temperature range, on samples irradiated with Fluorine at 20 MeV and fluences of ∼1014 cm−2. Optical techniques are very suitable for this case and have been used to monitor the recrystallization of the layers. The annealing process induces a displacement of the crystalline-amorphous boundary that is also linear with annealing time, and the recrystallization rates are consistent with those measured for tracks. The comparison of these data with those previously obtained for the heavily damaged (amorphous) layers produced by elastic nuclear collisions is summarily discussed.

Molecular Dynamics Simulations of Lead and Lithium in Liquid Phase

Pb17Li is today a reference breeder material in diverse fusion R&D programs worldwide. Extracting dynamic and structural properties of liquid LiPb mixtures via molecular dynamics simulations, represent a crucial step for multiscale modeling efforts in order to understand the suitability of this compound for future Nuclear Fusion technologies. At present a Li-Pb cross potential is not available in the literature. Here we present our first results on the validation of two semi-empirical potentials for Li and Pb in liquid phase. Our results represent the establishment of a solid base as a previous but crucial step to implement a LiPb cross potential. Structural and thermodynamical analyses confirm that the implemented potentials for Li and Pb are realistic to simulate both elements in the liquid phase.

Uncertainty assessment methodologies applied to Tritium production in fusion lithium breeding blankets

- Need of Tritium production - Neutronic objectives - The Frascati experiment - Measurements of Tritium activity

Liquid lead and Lithium. A Comprehensive molecular dynamics study

Pb17Li is today a reference breeder material in diverse fusion R&D programs worldwide. One of the main issues in these programs is the problem of liquid metals breeder blanket behavior. Structural material of the blanket should meet high requirements because of extreme operating conditions. Therefore the knowledge of eutectic properties like optimal composition, physical and thermodynamic behavior or diffusion coefficients of Tritium are extremely necessary for current designs. In particular, the knowledge of the function linking the tritium concentration dissolved in liquid materials with the tritium partial pressure at a liquid/gas interface in equilibrium, CT=f(PT), is of basic importance because it directly impacts all functional properties of a blanket determining: tritium inventory, tritium permeation rate and tritium extraction efficiency. Nowadays, understanding the structure and behavior of this compound is a real goal in fusion engineering and materials science. Simulations of liquids can provide much information to the community; not only supplementing experimental data, but providing new tests of theories and ideas, making specific predictions that require experimental tests, and ultimately helping to lead to the deeper understanding and better predictive behavior.

Molecular dynamics simulations of lead and lithium in liquid phase

Pb17Li is today a reference breeder material in diverse fusion R&D programs worldwide. Extracting dynamic and structural properties of liquid LiPb mixtures via molecular dynamics simulations, represent a crucial step for multiscale modeling efforts in order to understand the suitability of this compound for future Nuclear Fusion technologies. At present a Li-Pb cross potential is not available in the literature. Here we present our first results on the validation of two semi-empirical potentials for Li and Pb in liquid phase. Our results represent the establishment of a solid base as a previous but crucial step to implement a LiPb cross potential. Structural and thermodynamical analyses confirm that the implemented potentials for Li and Pb are realistic to simulate both elements in the liquid phase.

New simulations to qualify eutectic lithium-lead as breeder material

Pb17Li is today a reference breeder material in diverse fusion R&D programs worldwide. One of the main issues is the problem of liquid metals breeder blanket behavior. The knowledge of eutectic properties like optimal composition, physical and thermodynamic behavior or diffusion coefficients of Tritium are extremely necessary for current designs. In particular, the knowledge of the function linking the tritium concentration dissolved in liquid materials with the tritium partial pressure at a liquid/gas interface in equilibrium, CT =f(PT ), is of basic importance because it directly impacts all functional properties of a blanket determining: tritium inventory, tritium permeation rate and tritium extraction efficiency. Nowadays, understanding the structure and behavior of this compound is a real goal in fusion engineering and materials science. Atomistic simulations of liquids can provide much information; not only supplementing experimental data, but providing new tests of theories and ideas, making specific predictions that require experimental tests, and ultimately helping to a deeper understanding

Studies of a Self Cooled Lead Lithium blanket for HiPER Reactor

Within the frame of the HiPER reactor, we propose and study a Self Cooled Lead Lithium blanket with two different cooling arrangements of the system First Wall – Blanket for the HiPER reactor: Integrated First Wall Blanket and Separated First Wall Blanket. We compare the two arrangements in terms of power cycle efficiency, operation flexibility in out-off-normal situations and proper cooling and acceptable corrosion. The Separated First Wall Blanket arrangement is superior in all of them, and it is selected as the advantageous proposal for the HiPER reactor blanket. However, it still has to be improved from the standpoint of proper cooling and corrosion rates

Crack mechanical failure in lithium niobate crystal under ion irradiation; novel simulation by extended finite elements

Swift heavy ion irradiation (ions with mass heavier than 15 and energy exceeding MeV/amu) transfer their energy mainly to the electronic system with small momentum transfer per collision. Therefore, they produce linear regions (columnar nano-tracks) around the straight ion trajectory, with marked modifications with respect to the virgin material, e.g., phase transition, amorphization, compaction, changes in physical or chemical properties. In the case of crystalline materials the most distinctive feature of swift heavy ion irradiation is the production of amorphous tracks embedded in the crystal. Lithium niobate is a relevant optical material that presents birefringence due to its anysotropic trigonal structure. The amorphous phase is certainly isotropic. In addition, its refractive index exhibits high contrast with those of the crystalline phase. This allows one to fabricate waveguides by swift ion irradiation with important technological relevance. From the mechanical point of view, the inclusion of an amorphous nano-track (with a density 15% lower than that of the crystal) leads to the generation of important stress/strain fields around the track. Eventually these fields are the origin of crack formation with fatal consequences for the integrity of the samples and the viability of the method for nano-track formation. For certain crystal cuts (X and Y), these fields are clearly anisotropic due to the crystal anisotropy. We have used finite element methods to calculate the stress/strain fields that appear around the ion-generated amorphous nano-tracks for a variety of ion energies and doses. A very remarkable feature for X cut-samples is that the maximum shear stress appears on preferential planes that form +/-45º with respect to the crystallographic planes. This leads to the generation of oriented surface cracks when the dose increases. The growth of the cracks along the anisotropic crystal has been studied by means of novel extended finite element methods, which include cracks as discontinuities. In this way we can study how the length and depth of a crack evolves as function of the ion dose. In this work we will show how the simulations compare with experiments and their application in materials modification by ion irradiation.

Crack mechanical failure in ceramic materials under ion irradiation: case of lithium niobate crystal

Swift heavy ion irradiation (ions with mass heavier than 15 and energy exceeding MeV/amu) transfer their energy mainly to the electronic system with small momentum transfer per collision. Therefore, they produce linear regions (columnar nano-tracks) around the straight ion trajectory, with marked modifications with respect to the virgin material, e.g., phase transition, amorphization, compaction, changes in physical or chemical properties. In the case of crystalline materials the most distinctive feature of swift heavy ion irradiation is the production of amorphous tracks embedded in the crystal. Lithium niobate is a relevant optical material that presents birefringence due to its anysotropic trigonal structure. The amorphous phase is certainly isotropic. In addition, its refractive index exhibits high contrast with those of the crystalline phase. This allows one to fabricate waveguides by swift ion irradiation with important technological relevance. From the mechanical point of view, the inclusion of an amorphous nano-track (with a density 15% lower than that of the crystal) leads to the generation of important stress/strain fields around the track. Eventually these fields are the origin of crack formation with fatal consequences for the integrity of the samples and the viability of the method for nano-track formation. For certain crystal cuts (X and Y), these fields are clearly anisotropic due to the crystal anisotropy. We have used finite element methods to calculate the stress/strain fields that appear around the ion- generated amorphous nano-tracks for a variety of ion energies and doses. A very remarkable feature for X cut-samples is that the maximum shear stress appears on preferential planes that form +/-45º with respect to the crystallographic planes. This leads to the generation of oriented surface cracks when the dose increases. The growth of the cracks along the anisotropic crystal has been studied by means of novel extended finite element methods, which include cracks as discontinuities. In this way we can study how the length and depth of a crack evolves as function of the ion dose. In this work we will show how the simulations compare with experiments and their application in materials modification by ion irradiation.

Conceptual design of liquid lithium laboratory in the TECHNOFUSION project

The Technofusion project involves the construction of a relevant set of scientific technical facilities in Madrid, providing new tools to fusion energy community.

Self Cooled Lead Lithium blanket and reactor for HiPER

Within the frame of the HiPER reactor, we propose and study a Self Cooled Lead Lithium blanket with two different cooling arrangements of the system First Wall – Blanket for the HiPER reactor: Integrated First Wall Blanket and Separated First Wall Blanket. We compare the two arrangements in terms of power cycle efficiency, operation flexibility in out-off-normal situations and proper cooling and acceptable corrosion. The Separated First Wall Blanket arrangement is superior in all of them, and it is selected as the advantageous proposal for the HiPER reactor blanket. However, it still has to be improved from the standpoint of proper cooling and corrosion rates

An Assessment of Lithium Resources

Se hace un estudio de los recursos de una sustancia estratégica para la UE, a fin de orientar las investigaciones para un incremento de reservas que garanticen el abastecimiento de esta materia prima de gran demanda

Particle Patterning on Lithium Niobate waveguides via photovoltaic tweezers

Successful micro and nano-particle patterning on iron doped lithium niobate waveguides using photovoltaic fields is reported. This technique previously used in bulk crystals is here applied to waveguide configuration. Well defined particle patterns are obtained using two types of planar waveguides (by proton exchanged and swift heavy ion irradiation) and metallic and dielectric neutral particles. The use of waveguide configuration has allowed a reduction of the light exposure time until 3 s, two orders of magnitude smaller than typical values used in bulk.

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.