A. Brooks

Principal physics developments evaluated in the ITER design review

As part of the ITER Design Review and in response to the issues identified by the Science and Technology Advisory Committee, the ITER physics requirements were reviewed and as appropriate updated. The focus of this paper will be on recent work affecting the ITER design with special emphasis on topics affecting near-term procurement arrangements. This paper will describe results on: design sensitivity studies, poloidal field coil requirements, vertical stability, effect of toroidal field ripple on thermal confinement, material choice and heat load requirements for plasma-facing components, edge localized modes control, resistive wall mode control, disruptions and disruption mitigation.

Physics issues in the design of high-beta, low-aspect-ratio stellarator experiments

High-beta, low-aspect-ratio ~‘‘compact’’ ! stellarators are promising solutions to the problem of developing a magnetic plasma configuration for magnetic fusion power plants that can be sustained in steady state without disrupting. These concepts combine features of stellarators and advanced tokamaks and have aspect ratios similar to those of tokamaks ~2‐4!. They are based on computed plasma configurations that are shaped in three dimensions to provide desired stability and transport properties. Experiments are planned as part of a program to develop this concept. A b54% quasi-axisymmetric plasma configuration has been evaluated for the National Compact Stellarator Experiment ~NCSX!. It has a substantial bootstrap current and is shaped to stabilize ballooning, external kink, vertical, and neoclassical tearing modes without feedback or close-fitting conductors. Quasi-omnigeneous plasma configurations stable to ballooning modes at b54% have been evaluated for the Quasi-Omnigeneous Stellarator ~QOS! experiment. These equilibria have relatively low bootstrap currents and are insensitive to changes in beta. Coil configurations have been calculated that reconstruct these plasma configurations, preserving their important physics properties. Theory- and experiment-based confinement analyses are used to evaluate the technical capabilities needed to reach target plasma conditions. The physics basis for these complementary experiments is described.

Physics Design of a High-beta Quasi-axisymmetric Stellarator

Note: 8th Toki 11th International Stellarator Conference, Toki-City, Japan, September/October 1997, Proc. published in J. Plasma and Fusion Res., SERIES, Vol. 1, 429 - 432 (1998) Reference CRPP-CONF-1998-055 Record created on 2008-05-13, modified on 2016-08-08

Physics of compact stellarators

Recent progress in the theoretical understanding and design of compact stellarators is described. Hybrid devices, which depart from canonical stellarators by deriving benefits from the bootstrap current which flows at finite beta, comprise a class of low aspect ratio A<4 stellarators. They possess external kink stability (at moderate beta) in the absence of a conducting wall, possible immunity to disruptions through external control of the transform and magnetic shear, and they achieve volume-averaged ballooning beta limits (4%–6%) similar to those in tokamaks. In addition, bootstrap currents can reduce the effects of magnetic islands (self-healing effect) and lead to simpler stellarator coils by reducing the required external transform. Powerful physics and coil optimization codes have been developed and integrated to design experiments aimed at exploring compact stellarators. The physics basis for designing the national compact stellarator will be discussed.

NCSX MAGNETIC CONFIGURATION FLEXIBILITY AND ROBUSTNESS

Abstract The National Compact Stellarator Experiment (NCSX) will study the physics of low-aspect ratio, high-β, quasi-axisymmetric stellarators. To achieve the scientific goals of the NCSX mission, the device must be capable of supporting a wide range of variations in plasma configuration about a reference baseline equilibrium. We demonstrate the flexibility of NCSX coils to support such configuration variations and demonstrate the robustness of performance of NCSX plasmas about reference design values of the plasma current Ip, β, and profile shapes. The robustness and flexibility calculations make use of free-boundary plasma equilibrium constructions using a combination of nonaxisymmetric modular coils and axisymmetric toroidal and poloidal field coils. The primary computational tool for the studies is STELLOPT, a free-boundary optimization code that varies coil currents to target configurations with specific physics properties.

Use of a genetic algorithm for compact stellarator coil design

A new global optimization technique for designing stellarator coils has been developed and applied to the design of coils for the National Compact Stellarator Experiment. Using this technique coil sets were found with fewer coils and lower current densities than those obtained with traditional methods. A new coil design procedure which uses a genetic algorithm as the core optimization method is described and the resulting advanced coil designs presented.

Equilibrium and Flux Surface Issues in the Design of the NCSX

Abstract Equilibrium issues encountered in the design of the National Compact Stellarator Experiment (NCSX) are discussed, focusing particularly on equilibrium magnetic islands. Significant improvements have been made to the VMEC equilibrium code to deal with numerical challenges at the low aspect ratios characterizing the NCSX design. Modifications to the PIES code have increased its speed, allowing routine evaluation of flux surfaces for candidate configurations. An optimizer has been built around the PIES code for healing magnetic islands, modifying the coil shapes to suppress resonant components of the magnetic field while preserving desired physics and engineering properties. The modified coils produce improved flux surface quality for a range of configurations. Neoclassical effects, which are not included in the PIES calculations, are estimated using a cylindrical model and are found to further reduce island widths significantly.

Constructing Integrable High-pressure Full-current Free-boundary Stellarator Magnetohydrodynamic Equilibrium Solutions

For the (non-axisymmetric) stellarator class of plasma confinement devices to be feasible candidates for fusion power stations it is essential that, to a good approximation, the magnetic field lines lie on nested flux surfaces; however, the inherent lack of a continuous symmetry implies that magnetic islands responsible for breaking the smooth topology of the flux surfaces are guaranteed to exist. Thus, the suppression of magnetic islands is a critical issue for stellarator design, particularly for small aspect ratio devices. Pfirsch-Schluter currents, diamagnetic currents, and resonant coil fields contribute to the formation of magnetic islands, and the challenge is to design the plasma and coils such that these effects cancel. Magnetic islands in free-boundary high-pressure full-current stellarator magnetohydrodynamic equilibria are suppressed using a procedure based on the Princeton Iterative Equilibrium Solver [Reiman and Greenside, Comp. Phys. Comm. 43 (1986) 157] which iterate s the equilibrium equations to obtain the plasma equilibrium. At each iteration, changes to a Fourier representation of the coil geometry are made to cancel resonant fields produced by the plasma. The changes are constrained to preserve certain measures of engineering acceptability and to preserve the stability of ideal kink modes. As the iterations continue, the coil geometry and the plasma simultaneously converge to an equilibrium in which the island content is negligible, the plasma is stable to ideal kink modes, and the coils satisfy engineering constraints. The method is applied to a candidate plasma and coil design for the National Compact Stellarator Experiment [Reiman, et al., Phys. Plasmas 8 (May 2001) 2083].