E. Montalvo

Plasma Simulation and Surface Effects in a High-Field Tokamak Ignition Experiment

Plasma loading of the wall and limiter surfaces will play an important part in ignited tokamak experiments. Fusion plasmas will be characterized by high edge temperatures and large radially outward particle and energy fluxes. Many aspects of plasma-surface interactions presently seen in tokamak experiments will change and altogether new phenomena may appear. In this paper the authors identify possible plasma-surface interactions in a tokamak ignition experiment which can significantly affect its operation. The proposed ignition experiment, IGNITEX, is considered. The IGNITEX experiment uses a single-turn coil to produce a 20 Tesla toroidal field in a compact tokamak. A deuterium-tritium plasma will be gas fueled and ohmically heated to ignition. To obtain detailed edge plasma profiles and fluxes, a 1-D transport code is utilized. Ion and charge-exchange neutral sputtering processes at an all carbon surface system are included. A detailed description of the radial particle and energy flux calculations is given.

Stray field perturbations induced by the poloidal field magnet leads in IGNITEX

Summary form only given. The vacuum stray fields produced by the PF (poloidal field) leads in IGNITEX were calculated in detail. In the proposed design, each PF coil is fed by extended parallel flat plates close together so that they significantly self-cancel the magnetic field produced by the current flowing in opposite directions. The shapes of the terminals are modified based both on design considerations of the toroidal field (TF) magnet and on the perturbed fields produced by the terminals. Emphasis is given to the analysis of the plasma breakdown phase of the IGNITEX discharge. Stray field errors throughout the discharge are analyzed. This provides the basis for the evaluation of boundary conditions to study the magnetic configuration in the plasma as modified by stray field errors

Electromechanical analysis of the toroidal field magnets in the small and large versions of IGNITEX

Summary form only given. The Ignition Technology Demonstration (ITD) program was initiated to design, build, and test the operation of a single turn, 20-T, TF (toroidal field) coil powered by an existing 60-MJ HPG (homopolar generator) power supply system. The ITD TF coil is a 0.06 scale of the 1.5-m major radius IGNITEX baseline design. A finite-element program (TEXCOR) which solves a set of coupled electrical circuit, magnetic diffusion, and thermal diffusion equations with temperature-dependent properties was developed under the ITD program. TEXCOR provides temperatures and magnetic body force densities for a stress analysis of the magnet structure. The stress analysis is performed using a 3-D finite-element code (ABAQUS). The feasibility of the TF magnets of the small and large versions, based on the electromechanical aspect, was evaluated using TEXCOR. The temperature distribution stresses and axial preload specifications of the TF magnets were investigated

Analysis of a plasma disruption in the fusion ignition experiment IGNITEX

Summary form only given. The IGNITEX plasma column is closely surrounded by massive conducting structural material. The conducting structure has proven to have significant effects on plasma equilibrium and stability. Although a minimum number of disruptions are expected in IGNITEX, there is no assurance that disruptions will completely be prevented. A hard disruption will have noticeable effects in the vessel and magnet systems. A possible disruption process in IGNITEX was investigated. The plasma behavior was analyzed with the finite-element MHD code ROTEUS. Both the plasma current effects on the structure and the eddy current effects on the plasma configuration are considered. The vacuum vessel, poloidal field system, toroidal field coil seams, and toroidal field magnet structure were simulated in detail. The effect of initial conditions on the eddy current distribution over the electromagnetic load on the IGNITEX structure was analyzed

Simulation of the ignition pulse in the small and large versions of the IGNITEX experiment

Summary form only given. Simulations of the time evolution of the discharge for the small and large versions of IGNITEX are presented. Fusion ignition requires that the ignition factor, defined as the alpha heating rate divided by the power loss, be larger than one. The ignition factor depends on density and temperature. The plasma should have enough excess power to be able to evolve from the low-temperature region at discharge start-up to the ignited region within the constraint of limited time imposed by the heating of the toroidal field magnet. This time is estimated as ≅6 s for the small version and ≅20 s for the large version. The time evolution of the ion, electron, and thermalized alpha densities and the ion and electron temperatures are described by a system of ordinary differential equations which is integrated over the plasma volume. Electrons are heated by the ohmic power and later in the discharge by the alpha particle energy deposition (considered instantaneous). Ions are heated by collisional interaction with electrons at the beginning of the discharge and directly from alpha particles later in the discharge. Energy losses include those by transport and radiation. The effect of the energy confinement scaling on the plasma conditions at the flat top of the discharge is studied. For neoalcator-type scaling laws, the thermal runaway is controlled by cyclotron radiation emission

Analysis of the poloidal field magnet system leads in the IGNITEX experiment

Summary form only given. Extensive electromagnetic (EM) and magnetohydrodynamics analyses of the plasma discharge in IGNITEX have been carried out with the finite-element code PROTEUS. These results were used in the structural analysis code ABAQUS to investigate the thermomechanical stresses in the PF (poloidal field) leads. Various boundary conditions for the interface between the PF leads and the TF (toroidal field) magnet were considered. Part of the stresses in the leads relies upon the magnet. The time-dependent interaction between the EM and thermomechanical loads in the PF leads was analyzed. The stress distribution in the PF leads configuration and bearing loads was obtained

First wall thermomechanical stress analysis in a fusion ignition experiment

Summary form only given. Erosion calculations of the first wall system during regular operation and during plasma disruptions were made. Estimates were obtained with the DISRUP code. Disruptions that could take place at the peak of the ignited phase of the IGNITEX discharge were considered. Plasma disruptions were simulated by means of detailed equilibrium simulations. Both thermal quench and current quench were considered. The thermomechanical stress levels in the first wall system of IGNITEX bring an ignited plasma disruption were evaluated. A fully three-dimensional finite-element stress analysis was conducted. It was assumed that the first wall material behavior is adequately characterized by a decoupled thermomechanical response. Both melting and evaporation of the coating material were included in the analysis. Computations were performed with the finite-element program ABAQUS

Active plasma control in the fusion ignition experiment IGNITEX

Summary form only given. Active plasma control in IGNITEX was studied numerically with the CMS and PROTEUS codes. The most appropriate PF (poloidal field) coils and sensors for plasma control were identified and active control procedures were defined for the IGNITEX experiment. Possible vertical and horizontal plasma motions were shown to be readily controllable with the proposed internal PF magnet coils. Power supply characteristics required to obtain a stable system were determined