Dennis J Strickler

Features of Spherical Torus Plasmas

The spherical torus is a very small aspect ratio (A 2 are characterized by high toroidal beta (βt > 0.2), low poloidal beta (βp 1.5), and strong magnetic helical pitch (Θ comparable to F). A large near-omnigeneous region is seen in the large major radius, bad curvature region of the plasma in comparison with the conventional tokamaks. These features combine to engender the spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost. Because of its strong paramagnetism and helical pitch, the spherical torus plasma shares some of the desirable features of spheromak and reversed-field pinch (RFP) plasmas, but with tokamak-like confinement and safety factor q. The general class of spherical tori, which includes the spherical tokamak (q>1), the spherical pinch (1>q>0), and the spherical RFP (q<0), have magnetic field configurations unique in comparison with conventional tokamaks and RFPs.

Physics design of the national spherical torus experiment

The mission of the National Spherical Torus Experiment (NSTX) is to prove the principles of spherical torus physics by producing high-beta toroidal plasmas that are non-inductively sustained, and whose current profiles are in steady-state. NSTX will be one of the first ultra low a[P(input) up to 11 MW] in order to produce high-beta toroidal (25 to 40%), low collisionality, high bootstrap fraction (less than or equal to 70%) discharges. Both radio-frequency (RF) and neutral-beam (NB) heating and current drive will be employed. Built into NSTX is sufficient configurational flexibility to study a range of operating space and the resulting dependences of the confinement, micro- and MHD stability, and particle and power handling properties. NSTX research will be carried out by a nationally based science team.

The ARIES-I Tokamak Reactor Study †

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. Three ARIES visions are currently planned for the ARIES program. The ARIES-I design is a DT-burning reactor based on modest extrapolation from the present tokamak physics data base; ARIES-II is a DT-burning reactor which will employ potential advances in physics; and ARIES-III is a conceptual D-3He reactor. The first design to be completed is ARIES-I, a 1000 MWe power reactor. The key features of ARIES-I are: (1) a passively safe and low environmental impact design because of choice of low activation material throughout the fusion power core, (2) an acceptable cost of electricity, (3) a plasma with performance as close as possible to present-day experimental achievements, (4) a high performance, low activation, SiC composite blanket cooled by He, and (5) an advanced Rankine power cycle as planned for near term coal-fired plants. The ARIES-I research has also identified key physics and technology areas with the highest leverage for achieving attractive fusion power system.

Physics issues of compact drift optimized stellarators

Physics issues are discussed for compact stellarator configurations which achieve good confinement by the fact that the magnetic field modulus |B| in magnetic co-ordinates is dominated by poloidally symmetric components. Two distinct configuration types are considered: (1) those which achieve their drift optimization and rotational transform at low β and low bootstrap current by appropriate plasma shaping; and (2) those which have a greater reliance on plasma β and bootstrap currents for supplying the transform and obtaining quasi-poloidal symmetry. Stability analysis of the latter group of devices against ballooning, kink and vertical displacement modes has indicated that stable β values on the order of 15% are possible. The first class of devices is being considered for a low β near term experiment that could explore some of the confinement features of the high β configurations.

Design of the national compact stellarator experiment (NCSX)

Abstract The National Compact Stellarator Experiment (NCSX) [ http://www.pppl.gov/ncsx/Meetings/CDR/CDRFinal/EngineeringOverview_R2.pdf ] is being designed as a proof of principal test of a quasi-axisymmetric compact stellarator. This concept combines the high beta and good confinement features of an advanced tokamak with the low current, disruption-free characteristics of a stellarator. NCSX has a three-field-period plasma configuration with an average major radius of 1.4 m, an average minor radius of 0.33 m and a toroidal magnetic field on axis of up to 2 T. The stellarator core is a complex assembly of four coil systems that surround the highly shaped plasma and vacuum vessel. Heating is provided by up to four, 1.5 MW neutral beam injectors and provision is made to add 6 MW of ICRH. The experiment will be built at the Princeton Plasma Physics Laboratory, with first plasma expected in 2007.

Recent advances in quasi-poloidal stellarator physics issues

The quasi-poloidal stellarator (QPS) hybrid has been developed using a stellarator optimization approach that has proven to be compatible with both low aspect ratio and significantly reduced neoclassical transport relative to anomalous levels. A unique characteristic of this type of quasi-symmetry is a reduced viscous damping level for poloidal plasma flows. Since the plasma-generated E × B and diamagnetic flows are nearly poloidal, minimal parallel flows (and viscous stress) are required to achieve parallel pressure balance in comparison with configurations such as the tokamak, in which the plasma induced flows are nearly perpendicular to the direction of minimum viscosity and relatively larger parallel flows are required. In addition to this impact on neoclassical flows it is also anticipated that quasi-poloidal symmetry will minimize resistance to self-organized plasma-turbulence-driven shear flows and ease access to enhanced confinement states. In order to test these and other transport issues, the QPS device has been designed with a high degree of flexibility by allowing variable current capability not only in its vertical and toroidal coilsets but also in each separate modular coil group. Numerical optimizations have demonstrated that this flexibility can be used not only to modify transport properties, such as the poloidal viscosity, but also to directly suppress magnetic islands.

Designing Coils for Compact Stellarators

A method is presented for designing coils for compact stellarators. In contrast to methods that select a finite number of coils from an optimal continuous surface current distribution, the COILOPT code solves for the optimal parameters in an explicit representation of modular coils on a toroidal winding surface that is well separated from the plasma boundary, together with the coefficients of the winding surface. The problem is posed as a balance between approximating a prescribed magnetic configuration and satisfying certain critical engineering requirements. Results are presented for quasi-axisymmetric and quasi-poloidal compact stellarator designs.

Physics aspects of the Compact Ignition Tokamak

The Compact Ignition Tokamak (CIT) is a proposed modest-size ignition experiment designed to study the physics of alpha particle heating. The basic concept is to achieve ignition in a modest-size minimum cost experiment by using a high plasma density to achieve nT{sub E} {approx} 2 x 10{sup 20}s/m{sup 3} required for ignition. The high density requires a high toroidal field (10 T). The high toroidal field allows a large plasma current (10 MA) which provides a high level of ohmic heating, improves the energy confinement, and allows a relatively high beta ({approx} 6%). The present CIT design also has a high degree of elongation (k {approx} 1.8) to aid in producing the large plasma current. A double null poloidal divertor and pellet injection are part of the design to provide impurity and particle control, improve the confinement, and provide flexibility for improving the plasma profiles. Auxiliary heating is expected to be necessary to achieve ignition, and 10-20 MW of ICRF is to be provided.

The ARIES-III D-3He tokamak-reactor study

A description of the ARIES-III research effort is presented, and the general features of the ARIES-III reactor are described. The plasma engineering and fusion-power-core design are summarized, including the major results, the key technical issues, and the central conclusions. Analyses have shown that the plasma power-balance window for D-/sup 3/He tokamak reactors is small and requires a first wall (or coating) that is highly reflective to synchrotron radiation and small values of tau /sub ash// epsilon /sub e/ (the ratio of ash-particle to energy confinement times in the core plasma). Both first and second stability regimes of operation have been considered. The second stability regime is chosen for the ARIES-III design point because the reactor can operate at a higher value of tau /sub ash// tau /sub E// tau /sub E/ approximately=2 (twice that of a first stability version), and because it has a reduced plasma current (30 MA), magnetic field at the coil (14 T), mass, and cost (also compared to a first-stability D-/sup 3/He reactor). The major and minor radii are, respectively 7.5 and 2.5 m.<<ETX>>

DEVELOPMENT OF A ROBUST QUASI-POLOIDAL COMPACT STELLARATOR

Abstract A compact quasi-poloidally symmetric stellarator (QPS) plasma and coil configuration is described that has desirable physics properties and engineering feasibility with a very low aspect ratio plasma bounded by good magnetic flux surfaces both in vacuum and at β = 2%. The plasma is robust with respect to variations of pressure and the resulting bootstrap current, which leave the bounding flux surface approximately unchanged and thus reduce active positional control requirements. This configuration was developed by reconfiguring the QPS modular coils and applying a new computational method that maximizes the volume of good (integrable) vacuum flux surfaces as a measure of robustness. The stellarator plasma and coil design code STELLOPT is used to vary the coil geometry to determine the plasma geometry and profiles that optimize plasma performance with respect to neoclassical transport, infinite-n ballooning stability up to β = 2%, and coil engineering parameters. The normal component of the vacuum magnetic field is simultaneously minimized at the full-beta plasma boundary.