Micronucleus test and observation of nuclear alterations in erythrocytes of Nile tilapia exposed to waters affected by refinery effluent

Micronuclei and nuclear alterations tests were performed on erythrocytes of Oreochromis niloticus (Perciformes, Cichlidae) in order to evaluate the water quality from Paraiba do Sul river, in an area affected by effluents from an oil shale processing plant, located in the city of Sao Jose dos Campos, Brazil-SP. Water samples were collected on 2004 May and August (dry season) and on 2004 November and 2005 January (rain season), in three distinct sites, comprising 12 samples. It was possible to detect substances of clastogenic and/or aneugenic potential, as well as cytotoxic substances, chiefly at the point corresponding to the drainage of oil shale plant wastes along the river. The highest incidence of micronuclei and nuclear alterations was detected during May and August, whereas the results obtained in November and January were insignificant. This work shows that the effluent treatment provided by the oil shale plant was not fully efficient to minimize the effect of cytotoxic and mutagenic substances in the test organism surveyed. (c) 2006 Elsevier B.V. All rights reserved.

Phylogeny of Celastraceae Subfamily Salacioideae and Tribe Lophopetaleae Inferred from Morphological Characters and Nuclear and Plastid Genes

The phylogeny of Celastraceae subfamily Salacioideae (ca. 255 species in the Old and New World tropics) and tribe Lophopetaleae (ca. 29 species in southern Asia and the Austral-Pacific) was inferred using morphological characters together with plastid (matK, trnL-F) and nuclear (ITS and 26S rDNA) genes. Brassiantha, a monotypic genus endemic to New Guinea, is inferred to be more closely related to the clade of Dicarpellum (New Caledonia) and Hypsophila (Queensland, Australia) than it is to Hippocrateoideae or Salacioideae. This unambiguously supported resolution indicates that a nectary disk positioned outside the stamens has been convergently derived in these two lineages. The clade of Kokoona and Lophopetalum is resolved as more closely related to Breria and Elaeodendron than it is to Hippocrateoideae or Salacioideae. Sarawakodendron, a monotypic genus endemic to Borneo, is resolved as sister to Salacioideae. Salacioideae are inferred to have an Old World origin that was followed by a single successful radiation within Central and South America. We infer that capsular fruits are primitive within the clade of Hippocrateoideae + Sarawakodendron + Salacioideae, with berries a synapomorphy for Salacioideae. Based on the resolution of Sarawakodendron as sister to Salacioideae, we hypothesize that the filaments of Sarawakodendron arils are homologous to the spiral filaments in the mucilagenous pulp of Salacioideae.

Downregulation of nuclear factor-kappa B (NF-kappa B) pathway by silibinin in human monocytes challenged with Paracoccidioides brasiliensis

Fundação de Amparo à Pesquisa do Estado de São Paulo (FAPESP)

Contaminação radioativa: aspectos fundamentais associados à poluição nuclear

Since its discovery, radioactivity has brought numerous benefits to human societies. It has many applications in medicine, serving as a tool for non-invasive methods for diagnosis and therapies against diseases such as cancer. It also applies to technologies for energy in nuclear power plants with relatively low impacts on terms of perfect security. All applications, however, have risks, requiring maximum caution to drive processes and operations involving radioactive elements because, once released into the environment, they have extremely harmful effects on organisms affected. This paper presents fundamental concepts and principles of nuclear physics in order to understand the effects of radioactive elements released into the environment, culminating on the issue of radioactive contamination. Literature review allowed us to understand the radioactive contamination problem on living beings. Three major nuclear accidents have happened in the last thirty years, two of them in consecutive years. The nuclear accident at Chernobyl, Ukraine, in 1986, polluted large areas, condemning hundreds of thousands of people to live with consequences of the accident and effects of radiation, killing thousands of people throughout the years. In 1987, a major radiological accident occurred in Goiania (GO) when a source of radioactive cesium was violated, leading to the death of those who had direct or indirect contact with cesium. The most recent accident, in March, 2011, was located at the nuclear power plant in Fukushima Prefecture, Japan, after an earthquake and tsunami hit the region. There is no extensive and accurate knowledge about the consequences of the contamination entailed in that accident, although it is possible to verify signals on a global scale. An analysis of reports of contamination of large areas generated by nuclear plants with release of hazardous wastes suggests it is necessary to rethink the energy matrix of the various countries...

Neutronics analyses for fast spectrum nuclear systems and scenario studies for advanced nuclear fuel cycles

The present PhD thesis summarizes the three-years study about the neutronic investigation of a new concept nuclear reactor aiming at the optimization and the sustainable management of nuclear fuel in a possible European scenario. A new generation nuclear reactor for the nuclear reinassance is indeed desired by the actual industrialized world, both for the solution of the energetic question arising from the continuously growing energy demand together with the corresponding reduction of oil availability, and the environment question for a sustainable energy source free from Long Lived Radioisotopes and therefore geological repositories. Among the Generation IV candidate typologies, the Lead Fast Reactor concept has been pursued, being the one top rated in sustainability. The European Lead-cooled SYstem (ELSY) has been at first investigated. The neutronic analysis of the ELSY core has been performed via deterministic analysis by means of the ERANOS code, in order to retrieve a stable configuration for the overall design of the reactor. Further analyses have been carried out by means of the Monte Carlo general purpose transport code MCNP, in order to check the former one and to define an exact model of the system. An innovative system of absorbers has been conceptualized and designed for both the reactivity compensation and regulation of the core due to cycle swing, as well as for safety in order to guarantee the cold shutdown of the system in case of accident. Aiming at the sustainability of nuclear energy, the steady-state nuclear equilibrium has been investigated and generalized into the definition of the ``extended'' equilibrium state. According to this, the Adiabatic Reactor Theory has been developed, together with a New Paradigm for Nuclear Power: in order to design a reactor that does not exchange with the environment anything valuable (thus the term ``adiabatic''), in the sense of both Plutonium and Minor Actinides, it is required indeed to revert the logical design scheme of nuclear cores, starting from the definition of the equilibrium composition of the fuel and submitting to the latter the whole core design. The New Paradigm has been applied then to the core design of an Adiabatic Lead Fast Reactor complying with the ELSY overall system layout. A complete core characterization has been done in order to asses criticality and power flattening; a preliminary evaluation of the main safety parameters has been also done to verify the viability of the system. Burn up calculations have been then performed in order to investigate the operating cycle for the Adiabatic Lead Fast Reactor; the fuel performances have been therefore extracted and inserted in a more general analysis for an European scenario. The present nuclear reactors fleet has been modeled and its evolution simulated by means of the COSI code in order to investigate the materials fluxes to be managed in the European region. Different plausible scenarios have been identified to forecast the evolution of the European nuclear energy production, including the one involving the introduction of Adiabatic Lead Fast Reactors, and compared to better analyze the advantages introduced by the adoption of new concept reactors. At last, since both ELSY and the ALFR represent new concept systems based upon innovative solutions, the neutronic design of a demonstrator reactor has been carried out: such a system is intended to prove the viability of technology to be implemented in the First-of-a-Kind industrial power plant, with the aim at attesting the general strategy to use, to the largest extent. It was chosen then to base the DEMO design upon a compromise between demonstration of developed technology and testing of emerging technology in order to significantly subserve the purpose of reducing uncertainties about construction and licensing, both validating ELSY/ALFR main features and performances, and to qualify numerical codes and tools.

Space charge and dielectric response measurements to assess insulation aging of low-voltage cables used in nuclear power plants

The current design life of nuclear power plant (NPP) could potentially be extended to 80 years. During this extended plant life, all safety and operationally relevant Instrumentation & Control (I&C) systems are required to meet their designed performance requirements to ensure safe and reliable operation of the NPP, both during normal operation and subsequent to design base events. This in turn requires an adequate and documented qualification and aging management program. It is known that electrical insulation of I&C cables used in safety related circuits can degrade during their life, due to the aging effect of environmental stresses, such as temperature, radiation, vibration, etc., particularly if located in the containment area of the NPP. Thus several condition monitoring techniques are required to assess the state of the insulation. Such techniques can be used to establish a residual lifetime, based on the relationship between condition indicators and ageing stresses, hence, to support a preventive and effective maintenance program. The object of this thesis is to investigate potential electrical aging indicators (diagnostic markers) testing various I&C cable insulations subjected to an accelerated multi-stress (thermal and radiation) aging.

On the substrate- and stereospecificity of the plant carotenoid cleavage dioxygenase 7

Strigolactones are phytohormones synthesized from carotenoids via a stereospecific pathway involving the carotenoid cleavage dioxygenases 7 (CCD7) and 8. CCD7 cleaves 9-cis-β-carotene to form a supposedly 9-cis-configured β-apo-10′-carotenal. CCD8 converts this intermediate through a combination of yet undetermined reactions into the strigolactone-like compound carlactone. Here, we investigated the substrate and stereo-specificity of the Arabidopsis and pea CCD7 and determined the stereo-configuration of the β-apo-10′-carotenal intermediate by using Nuclear Magnetic Resonance Spectroscopy. Our data unequivocally demonstrate the 9-cis-configuration of the intermediate. Both CCD7s cleave different 9-cis-carotenoids, yielding hydroxylated 9-cis-apo-10′-carotenals that may lead to hydroxylated carlactones, but show highest affinity for 9-cis-β-carotene.

A design-phase PSA of anuclear-powered hydrogen plant

A probabilistic safety assessment (PSA) is being developed for a steam-methane reforming hydrogenproduction plant linked to a high-temperature gas-cooled nuclear reactor (HTGR). This work is based on the Japan Atomic Energy Research Institute's (JAERI) High Temperature Engineering Test Reactor (HTTR) prototype in Japan. The objective of this paper is to show how the PSA can be used for improving the design of the coupled plants. A simplified HAZOP study was performed to identify initiating events, based on existing studies. The results of the PSA show that the average frequency of an accident at this complex that could affect the population is 7 × 10−8 year−1 which is divided into the various end states. The dominant sequences are those that result in a methane explosion and occur with a frequency of 6.5 × 10−8 year−1, while the other sequences are much less frequent. The health risk presents itself if there are people in the vicinity who could be affected by the explosion. This analysis also demonstrates that an accident in one of the plants has little effect on the other. This is true given the design base distance between the plants, the fact that the reactor is underground, as well as other safety characteristics of the HTGR.

Interactive Graphic Simulation: An Advanced Methodology to Improve the Teaching-Learning Process in Nuclear Engineering Education and Training

Nowadays, computer simulators are becoming basic tools for education and training in many engineering fields. In the nuclear industry, the role of simulation for training of operators of nuclear power plants is also recognized of the utmost relevance. As an example, the International Atomic Energy Agency sponsors the development of nuclear reactor simulators for education, and arranges the supply of such simulation programs. Aware of this, in 2008 Gas Natural Fenosa, a Spanish gas and electric utility that owns and operate nuclear power plants and promotes university education in the nuclear technology field, provided the Department of Nuclear Engineering of Universidad Politécnica de Madrid with the Interactive Graphic Simulator (IGS) of “José Cabrera” (Zorita) nuclear power plant, an industrial facility whose commercial operation ceased definitively in April 2006. It is a state-of-the-art full-scope real-time simulator that was used for training and qualification of the operators of the plant control room, as well as to understand and analyses the plant dynamics, and to develop, qualify and validate its emergency operating procedures.

IFE Plant Technology Overview and contribution to HiPER proposal

HiPER is the European Project for Laser Fusion that has been able to join 26 institutions and signed under formal government agreement by 6 countries inside the ESFRI Program of the European Union (EU). The project is already extended by EU for two years more (until 2013) after its first preparatory phase from 2008. A large work has been developed in different areas to arrive to a design of repetitive operation of Laser Fusion Reactor, and decisions are envisioned in the next phase of Technology Development or Risk Reduction for Engineering or Power Plant facilities (or both). Chamber design has been very much completed for Engineering phase and starting of preliminary options for Reactor Power Plant have been established and review here.

Análisis Integrado de Seguridad de un accidente de SGTR en un reactor nuclear tipo PWR

El accidente de rotura de tubos de un generador de vapor (Steam Generator Tube Rupture, SGTR) en los reactores de agua a presión es uno de los transitorios más exigentes desde el punto de vista de operación. Los transitorios de SGTR son especiales, ya que podría dar lugar a emisiones radiológicas al exterior sin necesidad de daño en el núcleo previo o sin que falle la contención, ya que los SG pueden constituir una vía directa desde el reactor al medio ambiente en este transitorio. En los análisis de seguridad, el SGTR se analiza desde un punto determinista y probabilista, con distintos enfoques con respecto a las acciones del operador y las consecuencias analizadas. Cuando comenzaron los Análisis Deterministas de Seguridad (DSA), la forma de analizar el SGTR fue sin dar crédito a la acción del operador durante los primeros 30 min del transitorio, lo que suponía que el grupo de operación era capaz de detener la fuga por el tubo roto dentro de ese tiempo. Sin embargo, los diferentes casos reales de accidentes de SGTR sucedidos en los EE.UU. y alrededor del mundo demostraron que los operadores pueden emplear más de 30 minutos para detener la fuga en la vida real. Algunas metodologías fueron desarrolladas en los EEUU y en Europa para abordar esa cuestión. En el Análisis Probabilista de Seguridad (PSA), las acciones del operador se tienen en cuenta para diseñar los cabeceros en el árbol de sucesos. Los tiempos disponibles se utilizan para establecer los criterios de éxito para dichos cabeceros. Sin embargo, en una secuencia dinámica como el SGTR, las acciones de un operador son muy dependientes del tiempo disponible por las acciones humanas anteriores. Además, algunas de las secuencias de SGTR puede conducir a la liberación de actividad radiológica al exterior sin daño previo en el núcleo y que no se tienen en cuenta en el APS, ya que desde el punto de vista de la integridad de núcleo son de éxito. Para ello, para analizar todos estos factores, la forma adecuada de analizar este tipo de secuencias pueden ser a través de una metodología que contemple Árboles de Sucesos Dinámicos (Dynamic Event Trees, DET). En esta Tesis Doctoral se compara el impacto en la evolución temporal y la dosis al exterior de la hipótesis más relevantes encontradas en los Análisis Deterministas a nivel mundial. La comparación se realiza con un modelo PWR Westinghouse de tres lazos (CN Almaraz) con el código termohidráulico TRACE, con hipótesis de estimación óptima, pero con hipótesis deterministas como criterio de fallo único o pérdida de energía eléctrica exterior. Las dosis al exterior se calculan con RADTRAD, ya que es uno de los códigos utilizados normalmente para los cálculos de dosis del SGTR. El comportamiento del reactor y las dosis al exterior son muy diversas, según las diferentes hipótesis en cada metodología. Por otra parte, los resultados están bastante lejos de los límites de regulación, pese a los conservadurismos introducidos. En el siguiente paso de la Tesis Doctoral, se ha realizado un análisis de seguridad integrado del SGTR según la metodología ISA, desarrollada por el Consejo de Seguridad Nuclear español (CSN). Para ello, se ha realizado un análisis termo-hidráulico con un modelo de PWR Westinghouse de 3 lazos con el código MAAP. La metodología ISA permite la obtención del árbol de eventos dinámico del SGTR, teniendo en cuenta las incertidumbres en los tiempos de actuación del operador. Las simulaciones se realizaron con SCAIS (sistema de simulación de códigos para la evaluación de la seguridad integrada), que incluye un acoplamiento dinámico con MAAP. Las dosis al exterior se calcularon también con RADTRAD. En los resultados, se han tenido en cuenta, por primera vez en la literatura, las consecuencias de las secuencias en términos no sólo de daños en el núcleo sino de dosis al exterior. Esta tesis doctoral demuestra la necesidad de analizar todas las consecuencias que contribuyen al riesgo en un accidente como el SGTR. Para ello se ha hecho uso de una metodología integrada como ISA-CSN. Con este enfoque, la visión del DSA del SGTR (consecuencias radiológicas) se une con la visión del PSA del SGTR (consecuencias de daño al núcleo) para evaluar el riesgo total del accidente. Abstract Steam Generator Tube Rupture accidents in Pressurized Water Reactors are known to be one of the most demanding transients for the operating crew. SGTR are special transient as they could lead to radiological releases without core damage or containment failure, as they can constitute a direct path to the environment. The SGTR is analyzed from a Deterministic and Probabilistic point of view in the Safety Analysis, although the assumptions of the different approaches regarding the operator actions are quite different. In the beginning of Deterministic Safety Analysis, the way of analyzing the SGTR was not crediting the operator action for the first 30 min of the transient, assuming that the operating crew was able to stop the primary to secondary leakage within that time. However, the different real SGTR accident cases happened in the USA and over the world demonstrated that operators can took more than 30 min to stop the leakage in actual sequences. Some methodologies were raised in the USA and in Europe to cover that issue. In the Probabilistic Safety Analysis, the operator actions are taken into account to set the headers in the event tree. The available times are used to establish the success criteria for the headers. However, in such a dynamic sequence as SGTR, the operator actions are very dependent on the time available left by the other human actions. Moreover, some of the SGTR sequences can lead to offsite doses without previous core damage and they are not taken into account in PSA as from the point of view of core integrity are successful. Therefore, to analyze all this factors, the appropriate way of analyzing that kind of sequences could be through a Dynamic Event Tree methodology. This Thesis compares the impact on transient evolution and the offsite dose of the most relevant hypothesis of the different SGTR analysis included in the Deterministic Safety Analysis. The comparison is done with a PWR Westinghouse three loop model in TRACE code (Almaraz NPP), with best estimate assumptions but including deterministic hypothesis such as single failure criteria or loss of offsite power. The offsite doses are calculated with RADTRAD code, as it is one of the codes normally used for SGTR offsite dose calculations. The behaviour of the reactor and the offsite doses are quite diverse depending on the different assumptions made in each methodology. On the other hand, although the high conservatism, such as the single failure criteria, the results are quite far from the regulatory limits. In the next stage of the Thesis, the Integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermohydraulical analysis of a Westinghouse 3-loop PWR plant with the MAAP code. The ISA methodology allows obtaining the SGTR Dynamic Event Tree taking into account the uncertainties on the operator actuation times. Simulations are performed with SCAIS (Simulation Code system for Integrated Safety Assessment), which includes a dynamic coupling with MAAP thermal hydraulic code. The offsite doses are calculated also with RADTRAD. The results shows the consequences of the sequences in terms not only of core damage but of offsite doses. This Thesis shows the need of analyzing all the consequences in an accident such as SGTR. For that, an it has been used an integral methodology like ISA-CSN. With this approach, the DSA vision of the SGTR (radiological consequences) is joined with the PSA vision of the SGTR (core damage consequences) to measure the total risk of the accident.

Control rod drop surveillance using two friction coefficients

In the case of large burnup, a control rod (CR) guide tube in the pressurized water reactor of a commercial nuclear power plant might bend. As a consequence, a CR drop experiment may indicate an event of a CR partially inserted and whether the CR should be deemed inoperable. Early prevention of such an event can be achieved by measuring two friction coefficients: the hydraulic coefficient and the sliding coefficient. The hydraulic coefficient hardly changes, so that the curvature of the guide tube can only be detected thanks to a variation of the sliding coefficient. A simple model for the CR drop is established and validated with CR drop experiments. If tmx denotes the instant of CR maximum velocity, a linear relationship between (tmx)_2 and the sliding coefficient is found.

Plant regeneration from an endangered valuable cork oak tree by somatic embryogenesis.

Among the several applications of in vitro tissue culture techniques, the conservation of plant germplasm is one of the most widely used. The cork oak is one of the principal tree species in the Western Mediterranean región. Within this área, the Balearic Islands are considered to be a glacial refuge, and therefore a reservoir of genetic resources. A singular tree has been found in the small Minorca Island population. The haplotype of this tree is of Tyrrhenian origin, showing a past link between Minorca and Sardinia. Moreover, this tree do not bear a deletion within an ITS from ribosimic nuclear DNA, which is fairly common in many populations of this species, and indicates that ir may be the descendant of a very ancient population. This tree is currently in a precarious condition, and it has not produced acorns in the last years. Hence there is a clear need of vegetative propagation to conserve this genotype. We have previously developed methods to clone adult cork oak tres by somatic embryogenesis, and therefore the aim of the present work was to clone this singular tree. There braches from the corwn were collected in November 2004, and methods previously described were carried out. By February 2005 somatic embryogenesis was obtained from leaves of the tree with percentages on induction ranging from 17 to 54% depending on the branch, which may show a novel source of variation that requires further study. Spontaneously matured somatic embryos germinated at 46% in average, and the first somatic seedlings from the Alfavaret's cork oak tree were obtained. Therefore, this study shows one of the most relevant applications of somatic embryogenesis: the plant regeneration of valuable genotypes for the in situ and ex situ conservation of forest genetic resources.

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.

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.