M. C. Zarnstorff

Measuring Δ′ from electron temperature fluctuations in the Tokamak Fusion Test Reactor

A method is developed for determining directly from experimental data the classical tearing mode stability parameter Δ′. Specifically, an analytical fit function is derived for the electron temperature fluctuations (Te) in the vicinity of a magnetic island. Values of Δ′ determined from the fit function parameters for m/n=2/1, 3/2 and 4/3 modes (m and n are poloidal and toroidal mode numbers) are obtained using the high resolution Te profile data from major radius shift (“jog”) experiments on the Tokamak Fusion Test Reactor (TFTR) [D. Meade et al., Proceedings of the International Conference on Plasma Physics and Controlled Nuclear Fusion. Washington, District of Columbia, 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. I, pp. 9–24]. It is found that the n⩾2 modes have Δ′<0.

Neoclassical tearing modes in Tokamak Fusion Test Reactor experiments. I. Measurements of magnetic islands and Δ

Tearing-type modes are observed in most high confinement operation regimes in the Tokamak Fusion Test Reactor (TFTR) [Nucl. Fusion 35, 1429 (1995)]. Three different methods are used to measure the magnetic island widths: external magnetic coils, internal temperature fluctuation from electron cyclotron emission (ECE) diagnostics and an experiment where the plasma major radius is rapidly shifted (“Jog” experiments). A good agreement between the three methods is observed. Numerical and analytic calculations of Δ′ (the tearing instability index) are compared with an experimental measurement of Δ′ using the tearing mode eigenfunction mapped from the jog data. The obtained negative Δ′ indicates that the observed tearing modes cannot be explained by the classical current-gradient-driven tearing theory.

A magnetohydrodynamic stability study of reverse shear equilibria in the Tokamak Fusion Test Reactor

A study is presented of the low‐n (n=1,2,3) magnetohydrodynamic stability of equilibria with reverse shear safety factor profiles. The low‐n stability boundaries are found to be characterized by resonance structures due to internal so‐called ‘‘infernal’’ mode types of instabilities. The parametric dependence of shear reversal width and depth, current, and pressure gradient on the beta limit are determined by using profile models that allow each parameter to be varied independently. Reverse magnetic shear is found to have a stabilizing influence for modes with toroidal mode numbers n≥2 leading to the possibility of improved β limits in the Tokamak Fusion Test Reactor (TFTR) [Plasma Phys. Controlled Nucl. Fusion Res. 26, 11 (1984)].

Novel design methods for magnetic flux loops in the National Compact Stellarator Experiment

Magnetic pickup loops on the vacuum vessel (VV) can provide an abundance of equilibrium information for stellarators. A substantial effort has gone into designing flux loops for the National Compact Stellarator Experiment (NCSX) [Zarnstorff et al., Plasma Phys. Controlled Fusion 43, A237 (2001)], a three-field period quasi-axisymmetric stellarator under construction at the Princeton Plasma Physics Laboratory. The design philosophy, to measure all of the magnetic field distributions normal to the VV that can be measured, has necessitated the development of singular value decomposition algorithms for identifying efficient loop locations. Fields are expected to be predominantly stellarator symmetric (SS)—the symmetry of the machine design—with toroidal mode numbers per torus (n) equal to a multiple of 3 and possessing reflection symmetry in a period. However, plasma instabilities and coil imperfections will generate non-SS fields that must also be diagnosed. The measured symmetric fields will yield important...

Transitionless enhanced confinement and the role of radial electric field shear

Evidence is presented for the role of radial electric field shear in enhanced confinement regimes attained without sharp bifurcations or transitions. Temperature scans at constant density, created in the reheat phase following deuterium pellet injection into supershot plasmas in the Tokamak Fusion Test Reactor [J. D. Strachan, et al., Phys. Rev. Lett. 58, 1004 (1987)] are simulated using a physics-based transport model. The slow reheat of the ion temperature profile, during which the temperature nearly doubles, is not explained by relatively comprehensive models of transport due to Ion Temperature Gradient Driven Turbulence (ITGDT), which depends primarily on the (unchanging) electron density gradient. An extended model, including the suppression of toroidal ITGDT by self-consistent radial electric field shear, does reproduce the reheat phase. The extended reheat at constant density is observed in supershot but not L-Mode plasmas.

Effects of orbit squeezing on ion transport processes close to magnetic axis

It is shown that ion thermal conductivity close to the magnetic axis in tokamaks is reduced by a factor of |S|5/3 if (Mi/Me)2/3(Te/Ti)4/3/|S|5/3≫1. Here, S is the orbit squeezing factor, Mi(Me) is the ion (electron) mass, and Ti(Te) is the ion (electron) temperature. The reduction reflects both the increase of the fraction of trapped particles by a factor of |S|1/3, and the decrease of the orbit size in units of the poloidal flux ψ by a factor of |S|2/3.

Energetic particle transport in compact quasi-axisymmetric stellarators

Hamiltonian coordinate, guiding center code calculations of the confinement of suprathermal ions in quasi-axisymmetric stellarator (QAS) designs have been carried out to evaluate the attractiveness of compact configurations which are optimized for ballooning stability. A new stellarator particle following code is used to predict the confinement of thermal and neutral beam ions in a small experiment with R=145u200acm, B=1–2u200aT and for alpha particles in a reactor size device. As for tokamaks, collisional pitch angle scattering drives ions into ripple wells and stochastic field regions, where they are quickly lost. In contrast, however, such losses are enhanced in QAS so that high edge poloidal flux has limited value in improving ion confinement. The necessity for reduced stellarator ripple fields is emphasized.

Progress in NCSX Construction

The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge National Laboratory (ORNL). Its mission is to develop the physics understanding of the compact stellarator and evaluate its potential for future fusion energy systems. Compact stellarators use 3D plasma shaping to produce a magnetic configuration that can be steady state without current drive or feedback control of instabilities. The NCSX has major radius 1.4 m, aspect ratio 4.4, 3 field periods, and a quasi-axisymmetric magnetic field. It is predicted to be stable and have good magnetic surfaces at beta > 4% and to have tokamak-like confinement properties. The device will provide the plasma configuration flexibility and the heating and diagnostic access needed to test physics predictions. Component production has advanced substantially since the first contracts were placed in 2004. Manufacture of the vacuum vessel was completed in 2006. All eighteen modular coil winding forms have been delivered, and twelve modular coils have been wound and epoxy impregnated. A contract for the (planar) toroidal field coils was placed in 2006 and manufacture is in progress. Assembly activities have begun and will be the projects main focus in the next few years. The engineering challenge of NCSX is to meet the requirements for complex geometries and tight tolerances within the cost and schedule constraints of a construction project. This paper will focus on how the engineering challenges of component production have been resolved, and how the assembly challenges are being met.

Next Steps in Quasi-Axisymmetric Stellarator Research

The quasi-axisymmetric (QA) stellarator, a 3-D magnetic configuration with close connections to tokamaks, offers solutions for a steady state, disruption-free fusion system. A new experimental facility, QUASAR, provides a rapid approach to the next step in QA development, an integrated experimental test of its physics properties, taking advantage of the designs, fabricated components, and detailed assembly plans developed for the NCSX project. A scenario is presented for constructing the QUASAR facility for physics research operations starting in 2019. Operating in deuterium, such a facility would investigate the scale-up in size and pulse length from QUASAR, while a suitably equipped version operating in deuterium-tritium (DT) could address fusion nuclear missions. New QA optimization strategies, aimed at improved engineering attractiveness, would also be tested.

Facilities for quasi-axisymmetric stellarator research

The quasi-axisymmetric (QA) stellarator, a three-dimensional magnetic configuration with close connections to tokamaks, offers solutions for a steady-state, disruption-free fusion system. A new experimental facility, QUASAR, provides a rapid approach to the next step in QA development, an integrated experimental test of its physics properties, taking advantage of the designs, fabricated components, and detailed assembly plans developed for the NCSX project. A scenario is presented for constructing the QUASAR facility for physics research operations starting in 2019. A facility for the step beyond QUASAR, performance extension to high temperature, high pressure sustained plasmas is described. Operating in DD, such a facility would investigate the scale-up in size and pulse length from QUASAR, while a suitably equipped version operating in DT could address fusion nuclear missions, with operation starting in 2027.