R. Carrera

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.

Passive plasma vertical stabilization in a single-turn tokamak configuration

The plasma vertical stability in a single-turn tokamak configuration is analyzed. The stabilization effects of the vacuum vessel and poloidal field magnet are studied numerically with rigid and magnetohydrodynamic models. An analytic model dispersion relation is derived to estimate the effect of the single-turn toroidal field magnet on the plasma vertical stability. The typical growth time of the mode is found to be >1 s. The stability advantages of the single-turn configuration for a high-current tokamak plasma and the differences among the three models used are discussed. A single-turn tokamak configuration seems to be appropriate for a fusion ignition experiment in that it simplifies plasma control and makes feasible the control of high-current, elongated tokamak ignition plasma. 14 refs., 9 figs. 2 tabs.

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

Assessment of structural activation in the operation of the fusion ignition experiment IGNITEX

Summary form only given. The layout of the IGNITEX facility was modeled, and the activation responses of the various candidate materials were determined. The ONEDANT code was used to simulate the neutron transport, and the REAC2 code was employed to evaluate neutron activation. Neutron cross sections for the transport calculation were derived from the VITAMIN-E library. Steel, aluminum, copper, titanium, and Inconel were analyzed as vacuum vessel materials. Both GlidCop and BeCu copper alloys were studied as magnet materials. Steel, aluminum, and fiberglass were considered as cryostat materials. Different materials for cryostat covers were investigated. It is concluded that, in selecting component materials, it is important to account for activation levels and for additional shielding requirements for hands-on maintenance

Neutron streaming analysis in the fusion ignition experiment IGNITEX

Summary form only given. The effect of neutron streaming in the out-vessel maintenance inside of the primary shielding in IGNITEX has been analyzed by means of Monte Carlo techniques. The transport of neutrons and photons through diagnostic and vacuum penetrations during the IGNITEX pulse was simulated with the 3-D, MCNP numerical code. Various types of penetrations were considered, including maintenance ports (horizontal outboard access), horizontal diagnostic ports (horizontal outboard sight), and vertical diagnostic ports (vertical top/bottom sight). Shielding and design modifications for access piping to the IGNITEX machine were established on the basis of a multidimensional radiation analysis

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

Out-vessel maintenance for the fusion ignition experiment IGNITEX

Summary form only given. Maintenance requirements and procedures for the regions exterior to the vacuum vessel of the proposed ignition experiment IGNITEX have been developed. The IGNITEX device is a single-turn-coil. ohmically heated tokamak. The single-turn coil almost completely covers the vacuum vessel and serves as a good radiation shield, limiting activation of structures surrounding the plasma and reducing radiation from the activated interior of the machine. This results in the possibility of hands-on maintenance in the test cell. Due to the presence of diagnostic penetrations in the magnet system, there will be some direct radiation streaming from the vessel. Lead plugs which can be moved into place during maintenance periods are considered to provide shielding against photon emission from the active interior of the machine. Procedures to minimize and handle the activation of exterior structures in the path of the direct streaming have been developed

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