Investigations of tokamak dynamics, especially as they relate to the challenge of burn control, require an accurate representation of energy and particle confinement times. While the ITER-98 scaling law represents a correlation of data from a wide range of tokamaks, confinement scaling laws will need to be fine-tuned to specific operational features of specific tokamaks in the future. A methodology for developing, by regression analysis, tokamak- and configuration-specific confinement tuning models is presented and applied to DIII-D as an illustration. It is shown that inclusion of tuning parameters in the confinement models can significantly enhance the agreement between simulated and experimental temperatures relative to simulations in which only the ITER-98 scaling law is used. These confinement tuning parameters can also be used to represent the effects of various heating sources and other plasma operating parameters on overall plasma performance and may be used in future studies to inform the selection of plasma configurations that are more robust against power excursions.

Several methods are presented for improving upon the traditional analytic “circular” method for constructing a flux-surface aligned curvilinear coordinate system representation of equilibrium plasma geometry and magnetic fields, and the most accurate Asymmetric Miller method is applied to calculations of poloidal asymmetries in plasma density, velocity, and electric potential. Techniques for developing an orthogonalized coordinate system from a general curvilinear representation of plasma flux surfaces and for representing the poloidal component of the magnetic field in the orthogonalized curvilinear system are developed generally, in order to be applied to four plasma flux-surface models. The formalism for approximating flux surfaces originally presented by Miller is extended to include poloidal asymmetries between the upper and lower plasma hemispheres, and is subsequently shown to be more accurate at fitting the shapes of flux surfaces calculated using EFIT than both the traditional “circular” model and two alternative curvilinear models of comparable complexity based on Fourier expansions of major radius, vertical position, and minor radius. Applying the coordinate system orthogonalization technique to these four models allows for calculations of the poloidal magnetic field which, upon comparison to a calculation of the poloidal field performed in a Cartesian system using the experimentally based EFIT prediction for the Grad-Shafranov equilibrium, demonstrates that the asymmetric “Miller” model is also superior to other methods at representing the poloidal magnetic field. A system of equations developed by representing the poloidal variations of velocity, density, and electric potential using O(1) Fourier expansions in the flux-surface averaged neoclassical plasma continuity and momentum balances is solved using several variations of both the “Miller” and “circular” curvilinear models to set geometric scale factors, illustrating the effects that these improvements in geometric modeling have on tokamak fluid theory calculations.

Evolution of measured profiles of densities, temperatures and velocities in the edge pedestal region between successive ELM (edge-localized mode) events are analyzed and interpreted in terms of the constraints imposed by particle, momentum and energy balance in order to gain insights regarding the underlying evolution of transport processes in the edge pedestal between ELMs in a series of DIII-D discharges. The data from successive inter-ELM periods during an otherwise steady-state phase of the discharges were combined into a composite inter-ELM period for the purpose of increasing the number of data points in the analysis. These composite periods were partitioned into sequential intervals to examine inter-ELM transport evolution. The GTEDGE integrated modeling code was used to calculate and interpret plasma transport and properties during each interval using particle, momentum, and energy balance. Variation of diffusive and non-diffusive (pinch) particle, momentum, and energy transport over the inter-ELM period are examined for discharges with plasma currents from 0.5 to 1.5 MA and inter-ELM periods from 50 to 220 ms. Diffusive transport is dominant for ρ< 0.925, while non-diffusive and diffusive transport are very large and nearly balancing in the sharp gradient region 0.925 <ρ <1.0. Transport effects of ion orbit loss are significant for ρ > 0.95, and are taken into account. During the inter-ELM period, diffusive transport increases slightly more than non-diffusive transport, increasing total outward transport. Both diffusive and non-diffusive transport have a strong inverse correlation with plasma current. Weakening the electromagnetic pinch may increase outward particle transport, and enable control over the rebuilding of the edge pedestal between ELMs.

Le contrôle des instabilités de bord connues sous le nom d' "Edge Localized Modes" (ELMs) est une question capitale pour le futur tokamak ITER. Ce travail est consacré à l'une des plus prometteuses méthodes de contrôle des ELMs, basée sur un système de bobines produisant des Perturbations Magnétiques Résonantes (PMRs), dont le fonctionnement a été démontré en premier lieu dans le tokamak DIII-D en 2003. Nos objectifs principaux sont, d'une part, d'éclaircir la compréhension physique des mécanismes en jeu, et d'autre part, de proposer un design concret de bobines de contrôle des ELMs pour ITER. Afin de calculer et d'analyser les perturbations magnétiques créées par un ensemble de bobines donné, nous avons développé le code ERGOS. Le premier calcul ERGOS a été consacré aux bobines de contrôle des ELMs de DIII-D, les I-coils. Il montre que celles-ci créent des chaines d'îlots magnétiques se recouvrant au bord du plasma, engendrant ainsi une ergodisation du champ magnétique. Nous avons par la suite utilisé ERGOS pour la modélisation des expériences de contrôle des ELMs à l'aide des bobines de correction de champ d'erreur sur JET et MAST, auxquelles nous participons depuis 2006. Dans le cas de JET, nous avons montré l'existence d'une corrélation entre la mitigation des ELMs et l'ergodisation du champ magnétique au bord, en accord avec le résultat pour DIII-D. Le design des bobines de contrôle des ELMs pour ITER s'est fait principalement dans le cadre d'un contrat EFDA (European Fusion Development Agreement)-CEA, en collaboration avec des ingénieurs et physiciens de l'EFDA et d'ITER. Nous avons utilisé ERGOS intensivement, le cas des I-coils de DIII-D nous servant de référence. Trois designs candidats sont ressortis, que nous avons présentés au cours de la revue de design d'ITER, en 2007. La direction d'ITER a décidé récemment d'attribuer un budget pour les bobines de contrôle des ELMs, dont le design reste à choisir entre deux des trois options que nous avons proposées (ou proches de celles que nous avons proposées). Enfin, dans le but de mieux comprendre les phénomènes de magnétohydrodynamique non-linéaires liés au contrôle des ELMs par PMRs, nous avons recouru à la simulation numérique, notamment avec le code JOREK pour un cas DIII-D. Les simulations révèlent l'existence de cellules de convection induites au bord du plasma par les perturbations magnétiques et le possible "écrantage" des PMRs par le plasma en présence de rotation. La modélisation adéquate de l'écrantage, qui demande la prise en compte de plusieurs phénomènes physiques supplémentaires dans JOREK, a été entamée.

Fast magnetosonic waves in the ion cyclotron range of frequencies have the potential to heat plasma and drive current in a thermonuclear fusion reactor. A code, SELFO-light, has been developed to study the physics of ion cyclotron resonantheating and current drive in thermonuclear fusion reactors. It uses a global full wave solver LION and a new 1D Fokker-Planck solver for the self-consistent calculations of the wave field and the distribution function of ions.In present day tokamak experiments like DIII-D and JET, fast wave damping by ions at higher harmonic cyclotron frequencies is weak compared to future thermonuclear tokamak reactors like DEMO. The strong damping by deuterium, tritium and thermonuclear alpha-particles and the large Doppler width of fast alpha-particles in DEMO makes it difficult to drive the current when harmonic resonance layers of these ionspecies are located at low field side of the magnetic axis. At higher harmonic frequencies the possibility of fast wave current drive diminishes due to the overlapping of alpha-particle harmonic resonance layers. Narrow frequency bands suitable for the fast wave current drive in DEMO have been identified at lower harmonics of the alpha-particles. For these frequencies the effect of formation of high-energy tails in the distribution function of majority and minority ion species on the current drive have been studied. Some of these frequencies are found to provide efficient ion heating in the start up phase of DEMO. The spectrum where efficient current drive can be obtained is restricted due to weak electron damping at lower toroidal mode numbers and strong trapped electron damping at higher toroidal mode numbers. The width of toroidal mode spectra for which efficient current drive can be obtained have been identified, which has important implications for the antenna design. / <p>QC 20130327</p>