Fuel inventory and generation of carbon and metal dust in a tokamak are perceived to be serious safety and economy issues for the steady-state operation of a fusion reactor, e.g. ITER. These topics have been explored in this thesis in order to contribute to a better understanding and the development of methods for controlling and curtailing fuel accumulation and dust formation in controlled fusion devices. The work was carried out with material facing fusion plasmas in three tokamaks: TEXTOR in Forschungszentrum Jülich (Germany), Tore Supra in the Nuclear Research Center Cadarache (France) and JET in Culham Centre for Fusion Energy (United Kingdom). Following issues were addressed: (a) properties of material migration products, i.e. co-deposited layers and dust particles; (b) impact of fuel removal methods on dust generation and on modification of plasma-facing components; (c) efficiency of fuel and deposit removal techniques; (d) degradation mechanism of diagnostic components - mirrors - and methods of their regeneration. / <p>QC 20121116</p>

Material erosion, melting, splashing, bubbling, and ejection during disruption events in future large tokamak reactors are of serious concern to component longevity. The majority of the heat flux during disruptions will be incident on the divertor, which will be made from tungsten in the future large tokamak ITER. Electrothermal plasma sources operating in the confined controlled arc discharge regime produce heat fluxes in the range expected for hard disruptions in future large tokamaks. The radiative heat flux produced inside of the capillary discharge channel is from the formed high density (10^23 - 10^27/m^3) plasma with heat fluxes of up to 125 GW/m^2 over a period of 100s of microseconds, making such sources excellent simulators for ablation studies of plasma-facing materials in tokamaks during hard disruptions. Experiments have been carried out with the PIPE device exposing tungsten to these high heat flux plasmas. SEM images have been taken of the tungsten surfaces, cross sections of tungsten surfaces, and ejected material. Melting and bubble/void formation has been observed on the tungsten surface. The tungsten surface shows evidence of melt-layer flow and the existence of voids and cracks in the exposed material. The ejected material does not show direct evidence of liquid material ejection which would lead to splashing. EDS analysis has been performed on the ejected material which demonstrates a lack of deposited solid tungsten particulates greater than micron size. / Master of Science

Experimental and numerical studies simulating the gas-cooled divertor plate design concept have been carried out. While thermo-fluid and thermo-mechanical analyses have been previously performed to show the feasibility of the divertor plate design and its ability to accommodate a maximum heat flux of up to 10 MW/m2, no experimental data have heretofore been published to support or validate such analyses. To that end, this investigation has been undertaken. A test module with prototypical cross-sectional geometry has been designed, constructed, and instrumented. Experiments spanning the prototypical Reynolds numbers of the helium-cooled divertor have been conducted using pressurized air as the coolant. A second test module where the planar jet exiting the inlet manifold is replaced by a two-dimensional hexagonal array of circular jets over the entire top surface of the inlet manifold has also been tested. The thermal performance of both test modules with and without a porous metallic foam layer in the gap between the outer surface of the inlet manifold and the cooled surfaces of the pressure boundary were directly compared. For a given mass flow rate, the slot design with the metallic foam insert showed the highest heat transfer coefficient, with a pressure drop lower than that of the array of circular jets without foam. Additionally, numerical simulations matching the experimental operating conditions for the two cases without foam were performed using the computational fluid dynamics software package, FLUENT® v6.2. Comparisons of the experimental and numerical pressure drop, temperature, and heat transfer coefficient were made.

L'objectif de ce travail était d'élaborer un matériau à gradient de propriétés fonctionnelles (MGF) W/Cu afin de remplacer la couche de compliance (Cu-OFHC) dans les composants face au plasma des machines de fusion thermonucléaire de type ITER. La particularité de ce travail étant de réaliser ces matériaux sans dépasser la température de fusion du cuivre dans le but de contrôler la microstructure des matériaux. Le cofrittage est la solution la plus attractive pour les réaliser. La première étape du travail a donc été de diminuer la température de frittage du tungstène afin de réaliser ce cofrittage. La mise en forme d'un MGF continus étant délicat, des calculs thermomécaniques ont été réalisés afin de déterminer le nombre et la composition chimique des couches W-Cu pour augmenter la durée de vie des CFPs. Les conditions de frittage par Spark Plasma Sintering ont été optimisées afin d'avoir une densité maximale des monomatériaux WxCu1-x. L'influence de la teneur en cuivre et de la densité des monomatériaux sur les propriétés thermiques et mécaniques a été étudiée. Les conditions de frittage SPS des monomatériaux ont été appliquées sur des assemblages W/CuCrZr composés de plusieurs couches intercalaires. L'importance du temps d'assemblages pour l'intégrité de ceux-ci a été mise en évidence. L'étude du temps de palier lors des assemblages W/CuCrZr a permis d'identifier un paramètre permettant de qualifier l'intégrité de l'assemblage quelle que soit la composition et la nature de la couche de compliance. De plus, les phénomènes associés à la formation des interfaces de l'assemblage ont été identifiés. L'interface W/WxCu1-x est formée par l'extrusion du cuivre de la couche WxCu1-x dans les porosités du tungstène. L'interface WyCu1-y/CuCrZr est formée par la migration du cuivre de la couche CuCrZr dans la couche WyCu1-y. Enfin l'optimisation des conditions d'assemblage a montré que les contraintes mécaniques dues à la densification du Matériau à gradient de Propriétés Fonctionnelles pouvaient être limitées en frittant préalablement ce matériau. / The objective of this study was to develop a Functionally Graded Material (FGM) W / Cu to replace the compliance layer (Cu-OFHC) in the plasma facing components of thermonuclear fusion reactor like ITER. The peculiarity of this work is to elaborate these materials without exceeding the melting temperature of copper in order to control its microstructure. The co-sintering is the most attractive solution to achieve this goal.The first phase of this study has been to decrease the sintering temperature of the tungsten to achieve this co-sintering. The elaboration of a Functionally Graded Materials being delicate, thermo-mechanical calculations were performed in order to determine the number and chemical composition in order to increase the lifespan of Plasma Facing Components. Spark Plasma Sintering conditions were optimized in order to achieve maximum density of WxCu1-x composites. The effect of copper content and density of the WxCu1-x composites on thermal and mechanical properties was investigated. The SPS conditions were applied for W/CuCrZr assemblies with a compliance layer composed of several interlayers. The importance of time for the integrity of assemblies thereof has been highlighted.The study of the dwell time during W/CuCrZr assembly leads to identify a parameter to characterize the integrity of the interface regardless of the composition and the nature of the layer of compliance. Moreover, the phenomena associated with the formation of the interface assembly have been identified. The interface W/WxCu1-x is formed by the extrusion of the copper layer of the WxCu1-x inside the tungsten porosities. The WyCu1-y/CuCrZr interface is formed by copper migration of CuCrZr layer inside the WyCu1-y layer. Finally optimization assembly conditions showed that the mechanical stresses due to the densification of the Functionally Graded Materials can be limited by sintering the FGM before the assembly.

Metal surface cleaning is highly required in different fields of modern industry. Nuclear industry seeks for new methods for oxidized surface decontamination, and thermonuclear installations require the cleaning of plasma facing components from tritium-containing deposited layer. The laser ablation is proposed as an effective and safe method for metal surface cleaning and decontamination. The important factor influencing the laser heating and ablation is the in-depth distribution of laser radiation. The model of light propagation in a scattering layer on a metal substrate is developed and applied to analyse the features of light distribution. To simulate the contaminated surfaces, the stainless steel AISI 304L was oxidized by laser and in a furnace. Radioactive contamination of the oxide layer was simulated by introducing europium and/or sodium. The decontamination factor of more than 300 was demonstrated with found optimal cleaning regime. The decreasing of the corrosion resistance was found after laser cleaning. The ablation thresholds of ITER-like surfaces were measured. The cleaning productivity of 0.07 m2/hour∙W was found. For mirror surfaces, the damage thresholds were determined to avoid damage during laser cleaning. The possibility to restore reflectivity after thin carbon layer deposition was demonstrated. The perspectives of further development of laser cleaning are discussed.

Divertors are in-vessel, plasma-facing, components in magnetic-confinement fusion reactors. Their main function is to remove the fusion reaction ash (α-particles), unburned fuel, and eroded particles from the reactor, which adversely affect the quality of the plasma. A significant fraction (~15 %) of the total fusion thermal power is removed by the divertor coolant and must, therefore, be recovered at elevated temperature in order to enhance the overall thermal efficiency. Helium is the leading coolant because of its high thermal conductivity, material compatibility, and suitability as a working fluid for power conversion systems using a closed high temperature Brayton cycle. Peak surface heat fluxes on the order of 10 MW/m^2 are anticipated with surface temperatures in the region of 1,200°C to 1,500°C. Recently, several helium-cooled divertor designs have been proposed, including a modular T-tube design and a modular finger configuration with jet impingement cooling from perforated end caps. Design calculations performed using the FLUENT® CFD software package have shown that these designs can accommodate a peak heat load of 10 MW/m^2. Extremely high heat transfer coefficients (~50,000 W/(m^2 K)) were predicted by these calculations. Since these values of heat transfer coefficient are considered to be outside of the experience base for gas-cooled systems, an experimental investigation has been undertaken to validate the results of the numerical simulations. Attention has been focused on the thermal performance of the T-tube and the finger divertor designs. Experimental and numerical investigations have been performed to support both divertor geometries. Excellent agreement has been obtained between the experimental data and model predictions, thereby confirming the predicted performance of the leading helium-cooled divertor designs for near- and long-term magnetic fusion reactor designs. The results of this investigation provide confidence in the ability of state-of-the-art CFD codes to model gas-cooled high heat flux plasma-facing components such as divertors.

Le nettoyage des surfaces métalliques est nécessaire dans différents domaines de l'industrie moderne. L'industrie nucléaire cherche de nouvelles méthodes de décontamination des surfaces oxydées, et les installations thermonucléaires nécessitent le nettoyage des composants face au plasma pour enlever la couche déposée contenant tritium. L'ablation laser est proposée comme une méthode efficace et sûre pour le nettoyage des surfaces métalliques et leur décontamination. Le facteur important influençant le chauffage et l'ablation laser est la distribution en profondeur de l’intensité laser. Le modèle de propagation de la lumière dans une couche diffusant sur un substrat métallique est développé et appliqué pour analyser les caractéristiques de distribution de lumière. Pour simuler les surfaces contaminées, l'inox AISI 304L a été oxydé par laser et chauffé dans un four. La contamination radioactive de la couche d'oxyde a été simulée par l'introduction d’europium et / ou de sodium. Un facteur de décontamination de plus de 300 a été démontré avec le régime de nettoyage optimal trouvé. Une diminution de la résistance à la corrosion a aussi été montrée après un nettoyage laser. Les seuils d'ablation des surfaces ITER-like (représentatives d’ITER) ont été mesurés. Une vitesse de nettoyage de 0,07 m2/W∙h a été trouvée. Pour les surfaces miroir, les seuils de dommages étaient déterminés pour éviter les dommages lors du nettoyage au laser. La possibilité de restaurer la réflectivité après le dépôt d’une couche mince de carbone a été démontrée. Les perspectives de développement ultérieur de nettoyage laser sont discutées. / Metal surface cleaning is highly required in different fields of modern industry. Nuclear industry seeks for new methods for oxidized surface decontamination, and thermonuclear installations require the cleaning of plasma facing components from tritium-containing deposited layer. The laser ablation is proposed as an effective and safe method for metal surface cleaning and decontamination. The important factor influencing the laser heating and ablation is the in-depth distribution of laser radiation. The model of light propagation in a scattering layer on a metal substrate is developed and applied to analyse the features of light distribution. To simulate the contaminated surfaces, the stainless steel AISI 304L was oxidized by laser and in a furnace. Radioactive contamination of the oxide layer was simulated by introducing europium and/or sodium. The decontamination factor of more than 300 was demonstrated with found optimal cleaning regime. The decreasing of the corrosion resistance was found after laser cleaning. The ablation thresholds of ITER-like surfaces were measured. The cleaning productivity of 0.07 m2/hour∙W was found. For mirror surfaces, the damage thresholds were determined to avoid damage during laser cleaning. The possibility to restore reflectivity after thin carbon layer deposition was demonstrated. The perspectives of further development of laser cleaning are discussed.

Thermonuclear fusion may become an attractive future power source. The most promising of all fusion machine concepts is the tokamak. Despite decades of active research, still huge tasks remain before a fusion power plant can go online. One of these important tasks deals with the interaction between the fusion plasma and the reactor wall. This work focuses on how eroded wall materials of different origin and mass are transported in a tokamak device. Element transport can be examined by injection of certain species of unique and predetermined origin, so called tracers. Tracer experiments were conducted at the TEXTOR tokamak before its final shutdown. This offered an unique opportunity for studies of the wall and other internal components: For the first time it was possible to completely dismantle such a machine and analyse every single part of reactor wall, obtaining a detailed pattern of material migration. Main focus of this work is on the high-Z metals tungsten and molybdenum, which were introduced by WF6 and MoF6 injection into the TEXTOR tokamak in several material migration experiments. It is shown that Mo and W migrate in a similar way around the tokamak and that Mo can be used as tracer for W transport. It is further shown how other materials - medium-Z (Ni), low-Z (N-15 and F), fuel species (D) - migrate and get deposited. Finally, the outcome of dust sampling studies is discussed. It is shown that dust appearance and composition depends on origin, formation conditions and that it can originate even from remote systems like the NBI system. Furthermore, metal splashes and droplets have been found, some of them clearly indicating boiling processes. / <p>QC 20151203</p>

This thesis describes the design, manufacture and characterisation of thick vacuum plasma sprayed tungsten (W) coatings on steel substrates. Fusion is a potentially clean, sustainable, energy source in which nuclear energy is generated via the release of internal energy from nuclei. In order to fuse nuclei the Coulomb barrier must be breached - requiring extreme temperatures or pressures – akin to creating a ‘star in a box’. Tungsten is a promising candidate material for future fusion reactors due to a high sputtering threshold and melting temperature. However, the large coefficient of thermal expansion mismatch with reactor structural steels such as the low activation steel Eurofer’97 is a major manufacturing and in-service problem. A vacuum plasma spraying approach for the manufacture of tungsten and tungsten/steel graded coatings has been developed successfully. The use of graded coatings and highly textured 3D interface surfi-sculpt substrates has been investigated to allow the deposition of thick plasma sprayed tungsten coatings on steel substrates. Finite element models have been developed to understand the residual stresses that develop in W/steel systems and made use of experimental measurements of coating thermal history during manufacture and elastic moduli measured by nano-indentation. For both the graded and surfi-sculpt coating, the models have been used to understand the mechanism of residual stress redistribution and relief in comparison with simple W on steel coatings, particularly by consideration of stored strain energy. In the case of surfi-sculpt W coatings, the patterned substrate gave rise to regular stress concentrating features, and allowed 2mm thick W coatings to be produced reproducibly without delamination. Preliminary through thickness residual stress measurements were compared to model predictions and provided tentative evidence of significant W coating stress relief by regulated coating segmentation.