Magnetic confinement fusion has the potential to provide a nearly inexhaustible source of energy. Current fusion energy research projects involve conceptual "Tokamak" reactors, inside of which contaminants are "diverted" along magnetic field lines onto collection surfaces called divertor plates. Approximately 15% of the reactor's thermal power is focused on the divertor plates, creating a need for an effective cooling mechanism. Current extrapolations suggest that divertor plates will need to withstand heat fluxes of more than 10 MW/m2. The cooling mechanism will need to use a coolant compatible with the blanket system; currently helium, and use a minimal fraction of the reactor's available pumping power; ie: will need to experience minimal pressure drops. A leading cooling concept is the Helium Cooled Flat Plate Divertor (HCFP). This thesis experimentally examines four variations of the HCFP. The objectives are to: 1. Experimentally determine the thermal performance of the HCFP with a hexagonal pin-fin array in the gap between the impinging jet and the cooled surface over a range of flow rates and incident heat fluxes; 2. Experimentally measure the pressure drop associated with the hexagonal pin-fin array over a range of flow conditions; 3. Determine and compare the thermal performance of and pressure drop associated with the HCFP for two different slot widths, 0.5 mm and 2 mm over a range of flow rates and incident heat fluxes; 4. Compare the performance of the HCFP with a hexagonal pin-fin array with that of the HCFP with a metal-foam insert and the original HCFP; 5. Provide an experimental data set which can be used to validate numerical models of the HCFP design and its variants. 6. Analytically determine the maximum heat flux which the HCFP can be expected to withstand at theoretical operating conditions in the original and pin-fin array configurations.

Techn. Hochsch., Diss., 2004--Aachen.

Techn. Hochsch., Diss., 2004--Aachen.

Les prédictions concernant le fonctionnement des divertors de tokamak reposent généralement sur des codes de transport de bord, consistant en un code de plasma fluide associé à un code de Monte Carlo pour les espèces neutres. Les machines linéaires Magnum-PSI et Pilot-PSI chez DIFFER, produisant des plasmas comparables à ceux d'ITER ($T_e \sim 1$ eV, $n_e\sim10^{20}$ m$^{- 3}$). Dans cette thèse, les décharges de plasma ont été étudiées expérimentalement et par modélisation utilisant le code Soledge2D-Eirene afin de: a) rechercher quels phénomènes doivent être inclus dans la modélisation pour reproduire les tendances expérimentales et b) pour mieux interpréter les expériences . Expérimentalement, l’effet de la pression neutre $P_n$ a été étudié par diffusion Thomson, par une sonde de Langmuir, par spectroscopie visible et par calorimétrie. Nous avons montré qu'un faisceau de plasma peut être efficacement terminé par une couche de gaz neutre. Ensuite, à partir de comparaisons d’expériences et de simulations, nous avons montré qu’il était essentiel d’inclure les collisions élastiques entre le plasma et les molécules pour pouvoir reproduire les expériences. De plus, la $T_e$ proche de la cible est systématiquement surestimé, ce qui sous-estime le taux de recombinaison. Enfin, nous avons montré expérimentalement l’importance de l’inclusion de la recombinaison de surface dans le flux d’énergie de surface dans les plasmas à basse température. Les travaux présentés dans cette thèse contribuent à la compréhension des interactions plasma-neutre, en particulier dans les concepts de divertors plus fermés de nouvelle génération (MAST-upgrade, DIII-D). / Predictions for the operation of tokamak divertors typically rely on edge transport codes, consisting of a fluid plasma code in combination with a Monte Carlo code for neutral species. The linear devices Magnum-PSI and Pilot-PSI at DIFFER, operating with a cascaded arc plasma source that produces plasmas comparable to those expected in the ITER divertor ($T_e \sim 1 $ eV, $n_e \sim 10^{20}$m$^{-3}$). In this thesis, plasma discharges have been studied both experimentally and by modelling using the Soledge2D-Eirene code in order to a) investigate which phenomena need to be included in the modeling to reproduce experimental trends and b) provide new insights to the interpretation of experiments. Experimentally, the effect of neutral pressure $P_n$ was investigated using Thomson scattering, a Langmuir probe, visible spectroscopy and calorimetry. We have shown that a plasma beam can be effectively terminated by a blanket of neutral gas. Next, from comparisons of experiments and simulations, we have found that it is critical to include elastic collisions between the plasma and molecules if experiments are to be reproduced. Furthermore, the near-target $T_e$ is systematically overestimated by the code, underestimating the recombination rate thereby. Lastly, we have experimentally shown the importance of the inclusion of surface recombination to the surface energy flux in low temperature plasmas, an effect that is generally known but difficult to measure in fusion devices. The work presented in this thesis contributes to the understanding of plasma-neutral interactions especially in new generation, closed divertor concepts (i.e. MAST-upgrade, DIII-D).

La fusion nucléaire pourrait offrir une nouvelle source d'énergie stable, non émettrice de CO$_2$ et pérenne. Aujourd’hui, les tokamaks offrent les meilleures performances, en confinant un plasma à haute température au moyen d’un champ magnétique. Deux des enjeux technologiques majeurs pour l'exploitation des tokamaks sont l’extraction de puissance et le confinement du plasma sur des temps longs. Ces enjeux sont associés au transport de particules et de chaleur, déterminé par la turbulence, depuis le plasma centrale vers la zone de bord. Dans cette thèse, nous modélisons la turbulence dans le plasma de bord. Nous étudions en particulier la configuration divertor, dans laquelle le plasma central est isolé des parois au moyen d’un champ magnétique additionnel. Cette géométrie magnétique complexe est simulée avec le code de turbulence fluide TOKAM3X, né de la collaboration de l'IRFM au CEA et du laboratoire M2P2 de l'Université Aix-Marseille.Une comparaison avec des simulations en géométrie simplifiée montre une nature intermittente similaire de la turbulence. Néanmoins, l'amplitude des fluctuations, maximale au plan équatorial, est fortement réduite près du point X, où les lignes de champ deviennent purement toroïdales, en accord avec les données expérimentales récentes. Les simulations en configuration divertor montrent un confinement significativement plus élevé que en géométrie circulaire. Une inhibition partielle du transport radial de matière au niveau du point X contribue à cette amélioration. Ce mécanisme est potentiellement important pour comprendre la transition du mode de confinement faible au mode de confinement élevé, le mode opérationnel prévu pour ITER. / Nuclear fusion could offer a new source of stable, non-CO2 emitting energy. Today, tokamaks offer the best performance by confining a high temperature plasma by means of a magnetic field. Two of the major technological challenges for the operation of tokamaks are the power extraction and the confinement of plasma over long periods. These issues are associated with the transport of particles and heat, which is determined by turbulence, from the central plasma to the edge zone. In this thesis, we model turbulence in the edge plasma. We study in particular the divertor configuration, in which the central plasma is isolated from the walls by means of an additional magnetic field. This complex magnetic geometry is simulated with the fluid turbulence code TOKAM3X, developed in collaboration between the IRFM at CEA and the M2P2 laboratory of the University of Aix-Marseille.A comparison with simulations in simplified geometry shows a similar intermittent nature of turbulence. Nevertheless, the amplitude of the fluctuations, which has a maximum at the equatorial plane, is greatly reduced near the X-point, where the field lines become purely toroidal, in agreement with the recent experimental data. The simulations in divertor configuration show a significantly higher confinement than in circular geometry. A partial inhibition of the radial transport of particles at the X-point contributes to this improvement. This mechanism is potentially important for understanding the transition from low confinement mode to high confinement mode, the intended operational mode for ITER.

Controlling the scrape-off layer (SOL) properties in order to limit divertor erosion and extend component lifetime will be crucial to successful operation of ITER and devices that follow, where intermittent thermal loads on the order of GW/m² are expected. Steady state transport in the edge region is generally turbulent with large, order unity, fluctuations and is convection dominated. Owing to the success of the past fifty years of progress in magnetically confining hot plasmas, in this work we examine convective transport phenomena in the SOL that occur in the relatively "slow", drift-ordered fluid limit, most applicable to plasmas near MHD equilibrium. Diamagnetic charge separation in an inhomogeneous magnetic field is the principal energy transfer mechanism powering turbulence and convective transport examined in this work. Two possibilities are explored for controlling SOL conditions. In chapter 2 we review basic physics underlying the equations used to model interchange turbulence in the SOL and use a subset of equations that includes electron temperature and externally applied potential bias to examine the possibility of suppressing interchange driven turbulence in the Texas Helimak. Simulated scans in E₀×B₀ flow shear, driven by changes in the potential bias on the endplates appears to alter turbulence levels as measured by the mean amplitude of fluctuations. In broad agreement with experiment negative biasing generally decreases the fluctuation amplitude. Interaction between flow shear and interchange instability appears to be important, with the interchange rate forming a natural pivot point for observed shear rates. In chapter 3 we examine the possibility of resonant magnetic perturbations (RMPs) or more generally magnetic field-line chaos to decrease the maximum particle flux incident on the divertor. Naturally occurring error fields as well as RMPs applied for stability control are known to cause magnetic field-line chaos in the SOL region of tokamaks. In chapter 3 2D simulations are used to investigate the effect of the field-line chaos on the SOL and in particular on its width and peak particle flux. The chaos enters the SOL dynamics through the connection length, which is evaluated using a Poincaré map. The variation of experimentally relevant quantities, such as the SOL gradient length scale and the intermittency of the particle flux in the SOL, is described as a function of the strength of the magnetic perturbation. It is found that the effect of the chaos is to broaden the profile of the sheath-loss coefficient, which is proportional to the inverse connection length. That is, the SOL transport in a chaotic field is equivalent to that in a model where the sheathloss coefficient is replaced by its average over the unperturbed flux surfaces. Both fully chaotic and the flux-surface averaged approximation of RMP application significantly lower maximum parallel particle flux incident on the divertor. / text

During several decades of research and development in the field of Magnetically Confined Fusion (MCF) the preferred selection of materials for Plasma Facing Components (PFC) has changed repeatedly. Without doubt, endurance of the first wall will decide research availability and lifespan of the first International Thermonuclear Research Reactor (ITER). Materials erosion, redeposition and mixing in the reactor are the critical processes responsible for modification of materials properties under plasma impact. This thesis presents several diagnostic techniques and their applications for studies of materials transport in fusion devices. The measurements were made at the EXTRAP T2R Reversed Field Pinch operated in Alfvén laboratory at KTH (Sweden), the TEXTOR tokamak, recently shut down at Forschungszentrum Jülich (Germany) and in the JET tokamak at CCFE (UK). The main outcomes of the work are: Development and application of a method for non-destructive capture and characterization of fast dust particles moving in the edge plasma of fusion devices, as well as particles generated upon laser-assisted cleaning of plasma exposed surfaces.  Advancement of conventional broad beam and micro ion beam techniques to include measurement of tritium in the surfaces exposed in future D-T experiments.  Adaption of the micro ion beam method for precision mapping of non uniform elements concentrations on irregular surfaces.  Implementation of an isotopic marker to study the large scale materials migration in a tokamak and development of a method for fast non destructive sampling of the marker on surfaces of PFCs. / <p>QC 20140508</p>

A significant factor in the overall efficiency of the balance of plant for a future magnetic fusion energy (MFE) reactor is the thermal performance of the divertor. A significant fraction of the reactor power is delivered to the divertor as plasma impurities and fusion products are deposited on its surface. For an advanced MFE device, an average divertor heat load of 10 MW/m² is expected at steady-state operating conditions. Helium cooling of the divertors is one of the most effective ways to accommodate such a heat load. Several helium-cooled divertor designs have been proposed and/or studied during the past decade including the T-Tube divertor, the helium-cooled flat plate (HCFP) divertor, the helium-cooled multi-jet (HEMJ) divertor, the helium-cooled modular divertor with integral fin array (HEMP), and the helium-cooled modular divertor with slot array (HEMS). All of these designs rely on some form of heat transfer enhancement via impinging jets or cooling fins to help improve the heat removal capability of the divertor. For all of these designs very large heat transfer coefficients on the order of 50-60 kW/m²-K have been predicted. As the conditions of a fusion reactor and associated helium flow conditions (600 °C and 10 MPa) are difficult to achieve safely in a controlled laboratory environment, the study of these divertors often relies on computer simulations and experimental modeling at non-prototypical, albeit dynamically similar, conditions. Earlier studies were based on the assumption that, for geometrically similar divertor test modules, dynamic similarity can be achieved by matching only the Reynolds number. Experiments conducted in this investigation using different coolants and test module materials have shown this assumption to be false. Modified correlations for the Nusselt number and loss coefficients for the HEMJ and HEMP-like divertor modules have been developed. These have been used to develop generalized performance curves to predict the divertor performance, i.e. the maximum allowable heat flux and corresponding pumping power fraction, at prototypical conditions. Additionally, a numerical study has been performed to optimize the fin array geometry of the HEMP-like divertor module. The generalized correlations and performance curves developed in this investigation can be incorporated into system design codes, thereby allowing system designers to optimize the divertor geometry and operating conditions.

Exhausting the thermal power from a fusion tokamak is a critical engineering challenge. The life of components designed for these conditions has a strong influence on the availability of the machine. For a fusion power plant this dependence becomes increasingly important, as it will influence the cost of electricity. The most extreme thermal loading for a fusion power plant will occur in the divertor region, where components will be expected to survive heat fluxes in excess of 10 MW/m² over a number of years. This research focussed on the development of a heat sink module for operation under such conditions, drawing on advanced cooling strategies from the aerospace industry. A reference concept was developed using conjugate Computational Fluid Dynamics. The results were experimentally validated by matching Reynolds numbers on a scaled model. Heat transfer data was captured using a transient thermochromic liquid crystal technique. The results showed excellent agreement with the corresponding numerical simulations. To facilitate comparison against other divertor heat sink proposals, a nondimensional figure of merit for cooling performance was developed. When plotted against a non-dimensional mass flow rate, the reference heat sink was shown to have superior cooling performance to all other divertor proposals to date. Results from Finite Element Analysis were used in conjunction with the ITER structural design criteria to life the heat sink. The sensitivity of life to both boundary conditions, and local geometric features, were explored. The reference design was shown to be capable of exceeding the life requirements for heat fluxes in excess of 15 MW/m². A number of heat sinks, based on the reference design, were fabricated. These underwent non-destructive testing, before experimentation in a high-heat flux facility developed by the author. The heat transfer performance of the tested modules was found to exceed that predicted by numerical modelling, which was concluded to be caused by the fabrication processes used.

The role of cross field transport and volume recombination are of vital importance for a satisfactory understanding of the plasma edge in magnetically confined devices such as a Tokamak. Plasma fluctuations may travel cross field with significant velocities and play a central role in plasma transport. Cross field transport has been seen to be anomalous in most devices under a very broad range of experimental conditions. In recent years a clear indication of the relation between fluctuation, cross field particle transport and recombination has been reported.The University of Manchester Linear System (the ULS) has been used to observe the Balmer emission of the recombining plasma interacting with a dense neutral Hydrogen gas. The ULS is a device made of a cylindrical vacuum vessel 1.5 m long and 15 cm in radius. The plasma is formed in a separate chamber by a duoplasmatron source in the Demirkhanov configuration; the arc current was limited to 15 A and the potential drop was 100 V. The device is surrounded by a linear solenoid which was used to magnetize the plasma. The highest magnetic field was .1 T. Typical electron temperature in the device spans .1 to 10 eV, and the density 1. E+16 to 5. E+19.Diagnostic includes Langmuir probe and visible spectrometers. In addition, the DivCam imaging system originally designed and built to obtain 2D images of the MAST spherical Tokamak Scrape Off Layer, was used. The DivCam imaging system has enabled to obtain high resolution images of the plasma emission when interacting with the neutral gas. It appears evident that the Electron-Ion Recombination is strongly dependent upon radial transport of plasma particles: light emission attributed to EIR is only observed at a large cross field distance from the plasma source. Moreover, fast imaging of the plasma has also shown the presence of a plasma filament forming and propagating crossfield at the same region of the plasma where the EIR light is observed.To interpret the experimental observations obtained with DivCam, the OSM 1D fluid plasma solver and the EIRENE neutral Monte Carlo solver have been implemented in the linear geometry of the ULS linear system. Both the OSM and the EIRENE solvers were originally intended for tokamak and large magnetic confinement devices. Modelling of the EIR emissivity in the ULS device has demonstrated the importance of the inclusion of turbulent and blob transport in the model to obtain reasonable agreement between the observations and the theoretical predictions. The central density of the plasma filament has been estimated to be approximately .7 E+19 m-3 using EIRENE results.The emission attributed to hydrogenic ions (negative atomic H- and positive molecular ions H2+) and related to Molecular Assisted Recombinations can be estimated within EIRENE using the AMJUEL database. The database provides ion population estimations for three different collisional regimes: in the first regime a large population of vibrational excited hydrogen molecules are assumed to exist within the plasma volume; the second assumes strong Charge Exchange reactions and not vibrational excited molecule; the third assumes electron impact collisions with ground states molecule to be the only ion source. A reasonable agreement between the observations and the EIRENE prediction is only found when using the third estimation suggesting that molecular excitation and charge exchange processes are relatively unimportant under the experimental conditions considered.