The National Spherical Torus Experiment (NSTX) is being built at PPPL to test the fusion physics principles for the ST concept at the MA level. The NSTX nominal plasma parameters are R 0 = 85 cm, a = 67 cm, R/a ³ 1.26, B T = 3 kG, I p = 1 MA, q 95 = 14, elongation k £ 2.2, triangularity d £ 0.5, and plasma pulse length of up to 5 sec. The plasma heating / current drive (CD) tools are High Harmonic Fast Wave (HHFW) (6 MW, 5 sec), Neutral Beam Injection (NBI) (5 MW, 80 keV, 5 sec), and Coaxial Helicity Injection (CHI). Theoretical calculations predict that NSTX should provide exciting possibilities for exploring a number of important new physics regimes including very high plasma beta, naturally high plasma elongation, high bootstrap current fraction, absolute magnetic well, and high pressure driven sheared flow. In addition, the NSTX program plans to explore fully noninductive plasma start-up as well as a dispersive scrape-off layer for heat and particle flux handling. MotivationA broad range of encouraging advances has been made in the exploration of the Spherical Torus (ST) concept. 1 Such advances include promising experimental data from pioneering experiments, theoretical predictions, near-term fusion energy development projections such as the Volume Neutron Source 2 , and future applications such as power plant studies 3 . Recently, the START device has achieved a very high toroidal beta b T » 40% regime with b N » 5.0 at low q 95 » 3. 4 The National Spherical Torus Experiment (NSTX) is being built at PPPL to test the fusion physics principles for the ST concept at the MA level. 5 The NSTX device/plasma configuration allows the plasma shaping factor, I p q 95 / a B , to reach as high as 80 an order of magnitude greater than that achieved in conventional high aspect ratio tokamaks. The key physics objective of NSTX is to attain an advanced ST regime; i.e., simultaneous ultra high beta (b), high confinement, and high bootstrap current fraction (f bs ). 6 This regime is considered to be essential for the development of an economical ST power-plant because it minimizes the recirculating power and power plant core size. Other NSTX mission elements crucial for ST power plant development are the demonstration at the MA level of fully noninductive operation and the development of acceptable power and particle handling concepts. NSTX Facility Design Capability and Technology ChallengesThe NSTX facility is designed to achieve the NSTX mission with the following capabilities: ¥ I p = 1 MA for low collisionality at relevant densities, ¥ R/a ³ 1.26, including OH solenoid and coaxial helicity injection 7 (CHI) for startup,

S p h e m d research from 1979 to the present is reviewed including over 160 references. Emphasis is on undersbnding and interpretation of results. lo addition to summarizing results some new interpretations are presented. An introduction and brief history is followed by a discussion of generalized helicity and its time derivative. Formation and sustainment are discussed including five different methods. flux core, &pinch r-pinch, coaxial source, conical @-pinch, and kinked z-pinch. All methods use helicity injections. Sten~ly-state methods and rules for designing spheromak experiments are covered, followed by equilibrium and stability. Methods of stabilizing the tilt and shift modes are discussed as well a their impact on the reactor designs. Current-driven and pressure-driven instabilities as well 3s relaxation in general w covered. Energy confinement is discussed in terms of helicity decay time and and ,fs limits. The confinement in high and IOW open-flux geometries are mmpnred and the reactor implications discussed.

Experimental spheromak magnetic equilibria are measured which differ significantly from the minimum-energy state, and are well described by a numerical model where jw/Bhas a linear dependence on the poloidal flux function. Equilibria are determined in a nonperturbing manner by the combination of measurements of flux-conserver image currents with calculations from this model. These equilibria are corroborated by the observation of nondisruptive rotating internal kink distortions (with toroidal mode numbers n = 1, 2, and 3), coupled with theoretical MHD thresholds for the onset of these modes. PACS numbers: 52.55.Hc, 52.35.Py, 52.70.Ds In a spheromak, the magnetic fields are generated primarily by internal currents rather than by external coils. Once established, these fields are conjectured 1 to relax towards a state of minimum energy subject to the constraint that the magnetic helicity 2 is conserved. In a closed system the minimum-energy equilibrium satisfies VxB = XB with X^/xoJ 11/5 = const. Since competing effects are certainly present in any experiment, small deviations from a uniform, constant X can be expected. However, these departures from the minimum-energy or "Taylor" state are expected 1 to relax towards this lowest-energy configuration on a time scale shorter than the resistive diffusion time.We report results from the compact toroid experiment 3 ' 4 (CTX) which give spheromak equilibria (determined in a nonperturbing manner) not with X = const, but with X = X(iJ/), where t// is the normalized poloidal flux function [t|/ = (poloidal flux value)/ (total poloidal flux)]. The departures in magnetic energy of these equilibria with respect to the minimumenergy state is small. Coherent oscillations are seen, generated by rotating kink modes within the equilibria. The onset of the modes is shown to be consistent with the slope of X(i|/) from the equilibrium measurement.Other spheromak experiments 5,6 have measured the magnetic fields with internal probes, and have found experimental agreement with a zero-pressure, constants model. Hart et al? obtained for their data better agreement by including a finite-plasma-pressure correction to a constant-X model,In CTX, the X(i/i) profile is inferred from external measurements of induced image currents flowing in a mesh flux conserver 8 (MFC) surrounding the plasma, combined with results from numerical calculations of the equilibrium. This general technique is in principle similar to that used before to establish the MHD equilibrium in noncircular-cross-section tokamaks [see, e.g., Luxon and Brown, 9 and references therein]. Application of the technique to the CTX spheromak benefits from increased sensitivity to equilibrium changes since the spheromak is a small-aspect-ratio system; changes in the equilibrium affect the position of the magnetic axis and have a large effect on toroidal MFC image currents near the symmetry axis. Arrays of small Rogowski loops (5% relative calibration) measure the MFC currents, with the ratios of the currents (filtered to remove osci...

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