La compréhension du transport d'impuretés dans le coeur des plasmas de tokamaks est un enjeu principal de la fusion par confinement magnétique. En effet les impuretés sont omni-présentes dans les tokamaks et leur présence dans le coeur a des effets négatifs sur le confinement du plasma (dilution, rayonnement). Récemment une attention particulière s'est portée sur le flux convectif turbulent dû au gradient de rotation toroïdale pour expliquer les profils plat/creux d'impuretés observés expérimentalement dans le coeur du plasma. Dans cette thèse une approche numérique a été adoptée avec l'utilisation entre autres de codes tels que NEO pour le transport néoclassique et GKW pour le transport turbulent, tout les deux incluant l'effet de la rotation toroïdale. Une comparaison du facteur de piquage du carbone (R/LnC) mesuré expérimentalement (dans le tokamak européen JET) et obtenu numériquement est faite pour un grand nombre de plasma en mode H (mode de confinement amélioré). La comparaison entre les mesures expérimentales de R/LnC et les résultats numériques donne lieu à deux constats. Premièrement la partie convective du flux correspondant au gradient de rotation toroïdale a un impact important sur R/LnC et principalement à valeurs élevées de ce gradient. Deuxièmement les simulations surestiment ce piquage dans le coeur du plasma où les profils expérimentaux sont creux. Ce désaccord, observé à haute collisionalité uniquement, est également obtenu pour le transport de moment ce qui pourrait être la signature d'un méchanisme de brisure de symmétrie (important pour le transport d'impureté et de moment) manquant. / Understanding impurity transport in the core of tokamak plasmas is central to achieving controlled fusion. Indeed impurities are ubiquitous in these devices and their presence in the core are detrimental to plasma confinement (fuel dilution, Bremsstrahlung). Recently, specific attention was given to the convective mechanism related to the gradient of the toroidal rotation to explain experimental flat/hollow impurity profiles in the plasma core. In this thesis, up-to-date modelling tools (NEO for neoclassical transport and GKW for turbulent transport) including the impact of toroidal rotation are used to study both the neoclassical and turbulent contributions to impurity fluxes. A comparison of the experimental and modelled carbon density peaking factor (R/LnC) is performed for a large number of baseline and hybrid H-mode plasmas (increased confinement regimes) with modest to high toroidal rotation from the European tokamak JET. Confrontation of experimental and modelled carbon peaking factor yields two main results. First roto-diffusion is found to have a nonnegligible impact on the carbon peaking factor at high values of the toroidal rotation frequency gradient. Second, there is a tendency to overpredict the experimental R/LnC in the core inner region where the carbon density profiles are hollow. This disagreement between experimental and modelled R/LnC, closely related to the collisionality, is also observed for the momentum transport channel which hints at a common parallel symmetry breaking mechanism lacking in the simulations.

Orientador: Munemasa Machida / Dissertação (mestrado) - Universidade Estadual de Campinas, Instituto de Fisica, Gleb Wataghin / Made available in DSpace on 2018-08-14T10:48:49Z (GMT). No. of bitstreams: 1 Nascimento_Fellypedo_M.pdf: 4493365 bytes, checksum: 50c2d66d1ba6fb6004585fe057e964e5 (MD5) Previous issue date: 2009 / Resumo: Neste trabalho, foram feitas medidas simultâneas e três linhas e emissão e hidrogênio no tokamak NOVA-UNICAMP. A partir das medidas e brilho as emissões das linhas Ha , H b e Hg e fazendo uso de coeficientes que constam nas tabelas de Johnson e Hinnov, foi possível determinar temperaturas e densidades eletrônicas no plasma ao longo de descargas o tokamak. Para isto, foi utilizada, e aperfeiçoada, uma técnica desenvolvida num trabalho e doutoramento recente do nosso grupo, a qual faz uso do conceito de unicidade do tempo de confinamento de partículas. Os principais aprimoramentos realizados neste diagnóstico foram: utilização de três espectrômetros para medidas simultâneas das emissões e hidrogênio, instalação e fibras ópticas para coletar a luz emitida pelo plasma, adoção de um sistema de colimação para obter um certo grau e definição espacial nas medidas, uso de um maior número e valores e temperaturas na análise dos dados e desenvolvimento de um novo método (algorítimo) para obter os valores de temperaturas e densidades dos elétrons no plasma. As temperaturas e densidades eletrônicas médias obtidas ficaram em torno e 7,5 eV e 7,0 ·10 12cm-3, respectivamente. Estes valores estão entro do espera o para tais parâmetros na borda do tokamak NOVA-UNICAMP. Isto indica que este diagnóstico pode ser usado para monitorar ensidades e temperaturas e elétrons em plasmas gerados por tokamaks. Além isso, foram efetuados alguns experimentos com detectores multicanal e o gás hidrogênio foi trocado pelo hélio, na tentativa de mostrar a versatilidade do diagnóstico proposto. / Abstract: In this work, we have made simultaneous measurements of three hydrogen emission lines on our tokamak. From the measurements of absolute brightness of the Ha , H b e Hg lines an using data from Johnson an Hinnov table, was possible to determine electronic ensities an temperatures during the tokamak ischarges. For this,we have used, an refined, a technique developed in a recent PhD thesis in our work group. This technique uses the concept of particle confinement time uniqueness. The main upgrades made in this diagnostic were: the use of three spectrometers for simultaneous measurements of the hydrogen emissions, installation of optical fibers to collect the light emitte by the plasma, adoption of a collimation system for having some spatial definition of the measurements, use of a greater range of temperature values uring the data analysis and development of a new method (algorithm) for obtaining the electronic densities and temperatures in the plasma. The average temperature and density obtained was about 7.5 eV and 7.0 ·1012cm-3, respectively. The results obtained are in accordance with the expected values for these parameters at the edge of the NOVA-UNICAMP tokamak plasma. This indicates that this diagnostic can be used to monitor the electronic densities and temperatures in tokamak plasmas. Additionally, we have made experiments with multichannel detectors, and the hydrogen gas was replaced by helium, in an attempt to show the versatility of the proposed diagnostic. / Mestrado / Física / Mestre em Física

Thermonuclear fusion has potential to offer an economically, environmentally and socially acceptable supply of energy. A promising reactor design to execute thermonuclear fusion is the toroidal magnetic confinement device, tokamak. The tokamak still faces challenges in the major areas which can be categorised into confinement, heating and fusion technology. This thesis addresses the problem of confinement, in particular the role of transport along magnetic field lines perturbed by diverse MHD instabilities.<p><p>Unstable modes such as ideal ballooning-peeling, tearing etc. break closed magnetic surfaces and destroy the axisymmetry of the magnetic configuration in a tokamak, providing deviation of magnetic field lines from unperturbed magnetic surfaces. Radial gradients of plasma parameters have nonzero projections along such lines and drive parallel particle and heat flows which contribute to the radial transport. Such transport can significantly affect confinement as this takes place by the development of neoclassical tearing modes (NTMs) in the core and edge localised modes (ELMs) at the plasma periphery.<p><p>In this thesis, transport of heat through non-overlapped magnetic island chains is first investigated using the 'Optimal path' approach, which is based on the principal of minimum entropy production. This model shows how the effective heat conduction through islands increases with parallel heat conduction and with the perturbation level. A more standard analytical approach for the limit cases of "small" and "large" islands is also presented. Transport of heat through internally heated magnetic islands is next investigated by further development of the 'Optimal path' method. In addition the approach by R. Fitzpatrick, has been extended for this investigation. By application of these approaches to experimental observations made at TEXTOR tokamak, heat flux limit, limiting parallel heat conduction in low collisional plasmas, is elucidated.<p><p>Models to study transport of heat and particles due to ELMs have also been developed. Energy losses during ELMs have been estimated considering contribution from parallel conduction due to electrons and parallel convection of ions, with constant level of the magnetic field perturbation, steady profiles for density and temperature, and by accounting for the heat flux limit. The estimate shows good agreement with experimental observations. The model is developed further by accounting for the time evolution of the perturbation level due to ballooning mode, and of density and temperature profiles. / Doctorat en Sciences / info:eu-repo/semantics/nonPublished

This thesis describes an experimental study of high intensity, pulsed, optically pumped, far-infrared (FIR) lasers. The work was motivated by the need for a radiation source for the measurement of the ion temperature in magnetically confined, high temperature plasmas (e.g. tokamak plasmas), using Thomson scattering. Constraints imposed by the plasma parameters, the scattering geometry and available detector sensitivities lead to the requirement of a radiation source wavelength between 30μm and 1mm and a source power . 1 MW in a bandwidth 60 MHz. Results are presented for a 496μm, 500 watt, methyl fluoride (CH₃F) cavity laser, with a bandwidth of and < 30MHz, which was optically pumped by a 9.55μm CO₂ laser. Results are also presented for an optically excited mirrorless, super-radiant, CH₃F laser, which generated over 0.6MW of FIR radiation within a bandwidth of about 300MHz. The performance of this laser has also been simulated by a computer model, which allows the optimum operating parameters to be predicted. An assembly constructed on the principle of the injection laser, in which low power narrow-band oscillator radiation is used to control the output of a super-radiant system, has been used to generate 250 kW of 496 andmu;m radiation, with a bandwidth of < 60 MHz. Investigations of the FIR output from heavy water vapour (D₂O) in a super-radiant laser assembly, optically excited by several different CO₂ laser wavelengths, have resulted in the generation of 60 ns (FWHM) pulses of FIR radiation with average powers of 1.3, 9.2 and 15.8MW, at wavelengths of 385, 119 and 66μm, respectively. All these lasers were found to have a higher CO₂ to FIR photon conversion efficiency than the 496μm CH₃F laser. In addition, the energy level spacing in D₂O is such that the molecule can generate narrow bandwidth radiation more readily than the CH₃F molecule. From this work it is concluded that an injection laser assembly, similar to that used with CH₃F, but containing D₂O vapour, optically pumped by a 9.26μm CO₂ laser and generating several megawatts of 385μm radiation, would satisfy the source requirements mentioned above.

It has been recognized for some time that the very high fluence of fast (14.1MeV) neutrons produced by deuterium-tritium fusion will represent a major materials challenge for the development of next-generation fusion energy projects such as a fusion component test facility and demonstration fusion power reactor. The best-understood and most promising solutions presently available are a family of low-activation steels originally developed for use in fission reactors, but the ferromagnetic properties of these steels represent a danger to plasma confinement through enhancement of magnetohydrodynamic instabilities and increased susceptibility to error fields. At present, experimental research into the effects of ferromagnetic materials on MHD stability in toroidal geometry has been confined to demonstrating that it is still possible to operate an advanced tokamak in the presence of ferromagnetic components. In order to better quantify the effects of ferromagnetic materials on tokamak plasma stability, a new ferritic wall has been installated in the High Beta Tokamak—Extended Pulse (HBT-EP) device. The development, assembly, installation, and testing of this wall as a modular upgrade is described, and the effect of the wall on machine performance is characterized. Comparative studies of plasma dynamics with the ferritic wall close-fitting against similar plasmas with the ferritic wall retracted demonstrate substantial effects on plasma stability. Resonant magnetic perturbations (RMPs) are applied, demonstrating a 50% increase in n = 1 plasma response amplitude when the ferritic wall is near the plasma. Susceptibility of plasmas to disruption events increases by a factor of 2 or more with the ferritic wall inserted, as disruptions are observed earlier with greater frequency. Growth rates of external kink instabilities are observed to be twice as large in the presence of a close-fitting ferritic wall. Initial studies are made of the influence of mode rotation frequency on the ferritic effect, as well as observations of the effect of the ferritic wall on disruption halo currents.

This thesis is organized into three body chapters: (1) the first use of naturally rotating tearing modes to diagnose intrinsic error fields is presented with experimental results from the EXTRAP T2R reversed field pinch, (2) a large scale study of locked modes (LMs) with rotating precursors in the DIII-D tokamak is reported, and (3) an in depth study of LM induced thermal collapses on a few DIII-D discharges is presented. The amplitude of naturally rotating tearing modes (TMs) in EXTRAP T2R is modulated in the presence of a resonant field (given by the superposition of the resonant intrinsic error field, and, possibly, an applied, resonant magnetic perturbation (RMP)). By scanning the amplitude and phase of the RMP and observing the phase-dependent amplitude modulation of the resonant, naturally rotating TM, the corresponding resonant error field is diagnosed. A rotating TM can decelerate and lock in the laboratory frame, under the effect of an electromagnetic torque due to eddy currents induced in the wall. These locked modes often lead to a disruption, where energy and particles are lost from the equilibrium configuration on a timescale of a few to tens of milliseconds in the DIII-D tokamak. In fusion reactors, disruptions pose a problem for the longevity of the reactor. Thus, learning to predict and avoid them is important. A database was developed consisting of 2000 DIII-D discharges exhibiting TMs that lock. The database was used to study the evolution, the nonlinear effects on equilibria, and the disruptivity of locked and quasi-stationary modes with poloidal and toroidal mode numbers m=2 and n=1 at DIII-D. The analysis of 22,500 discharges shows that more than 18% of disruptions present signs of locked or quasi-stationary modes with rotating precursors. A parameter formulated by the plasma internal inductance l_i divided by the safety factor at 95% of the toroidal flux, q_95, is found to exhibit predictive capability over whether a locked mode will cause a disruption or not, and does so up to hundreds of milliseconds before the disruption. Within 20 ms of the disruption, the shortest distance between the island separatrix and the unperturbed last closed flux surface, referred to as d_edge, performs comparably to l_i/q_95 in its ability to discriminate disruptive locked modes, and it also correlates well with the duration of the locked mode. On average, and within errors, the n=1 perturbed field grows exponentially in the final 50 ms before a disruption, however, the island width cannot discern whether a LM will disrupt or not up to 20 ms before the disruption. A few discharges are selected to analyze the evolution of the electron temperature profile in the presence of multiple coexisting locked modes during partial and full thermal quenches. Partial thermal quenches are often an initial, distinct stage in the full thermal quench caused by radiation, conduction, or convection losses. Here we explore the fundamental mechanism that causes the partial quench. Near the onset of partial thermal quenches, locked islands are observed to align in a unique way, or island widths are observed to grow above a threshold. Energy analysis on one discharge suggests that about half of the energy is lost in the divertor region. In discharges with minimum values of the safety factor above 1.2, and with current profiles expected to be classically stable, locked modes are observed to self-stabilize by inducing a full thermal quench, possibly by double tearing modes that remove the pressure gradient across the island, thus removing the neoclassical drive.

An extended neoclassical rotation theory (poloidal and toroidal) is developed from the fluid moment equations, using the Braginskii decomposition of the viscosity tensor extended to generalized curvilinear geometry and a neoclassical calculation of the parallel viscosity coefficient interpolated over collision regimes. Important poloidal dependences of density and velocity are calculated using the Miller equilibrium flux surface geometry representation, which takes into account elongation, triangularity, flux surface compression/expansion and the Shafranov shift. The resulting set of eight (for a two-ion-species plasma model) coupled nonlinear equations for the flux surface averaged poloidal and toroidal rotation velocities and for the up-down and in-out density asymmetries for both ion species are solved numerically. The numerical solution methodology, a combination of nonlinear Successive Over-Relaxation(SOR) and Simulated Annealing(SA), is also discussed. Comparison of prediction with measured carbon poloidal and toroidal rotation velocities in a co-injected and a counter-injected H-mode discharges in DIII-D [J. Luxon, Nucl. Fusion 42, 614 (2002)] indicates agreement to within <10% except in the very edge in the co-injected discharge.

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.

Thèse de doctorat : Sciences. Génie électrique : Ecole centrale de Lyon : 2001. / Bibliogr. 19 réf.

Thèse de doctorat : Sciences. Génie électrique : Ecole centrale de Lyon : 2001. / Titre provenant de l'écran-titre. Bibliogr. 19 réf.