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,

The 2-D radial vs. poloidal structure and motion of edge turbulence in NSTX were measured by using high-speed imaging of the visible light emission from a localized neutral gas puff. Edge turbulence images are shown and analyzed for Ohmic, L-mode and H-mode plasma conditions. Typical edge turbulence poloidal correlation lengths as measured using this technique are ≈ 4±1 cm and autocorrelation times are 40±20 µsec in all three regimes. The relative fluctuation level is typically smaller in H-mode than in Lmode, and transitions from H-to L-mode and can occur remarkably quickly (≈ 30 µsec). The 2-D images often show localized regions of strong light emission which move both poloidally and radially through the observed region at a typical speed of ≈ 10 5 cm/sec, and sometimes show spatially coherent modes.2

A method for determining plasma shape and global properties of high-beta non-circular tokamak equilibria (βp, β, q and ℓi) from magnetic sensor data is described. The technique uses a least-squares fitting procedure to find the plasma boundary and a global force balance analysis (as opposed to a complete solution of the MHD equilibrium equation) to determine plasma pressure. Estimates of the uncertainties in the computed quantities are also obtained, and the method is fast enough that it can be used to provide a complete time history (up to 100 time points) of ISX-B high-beta discharges within two to three minutes of each shot. Values obtained using this method are in excellent agreement with more detailed measurements and with free-boundary MHD equilibrium computations.

The ITER Ion Cyclotron Heating and Current Drive system will deliver 20MW of radio frequency power to the plasma in quasi continuous operation during the different phases of the experimental programme. The system also has to perform conditioning of the tokamak first wall at low power between main plasma discharges. This broad range of reqiurements imposes a high flexibility and a high availabiUty. The paper highlights the physics and design reqiurements on the IC system, the main features of its subsystems, the predicted performance, and the current procurement and installation schedide.

Neutral-beam injection of up to 2.5 MW into plasmas in the ISX-B tokamak (R0 = 0.93 m, a = 0.27 m, BT = 0.9–1.5 T, Ip = 70–210 kA, n̄e = 2.5–10×1013 cm−3) has created plasmas with volume-averaged beta of up to ∼ 2.5%, peak beta values of up to ∼ 9%, and root-mean-square beta values of up to ∼ 3.5%. Energy confinement time is observed to decrease by about a factor of two as beam power goes from 0 to 2.5 MW; the decrease is caused predominantly by the electron confinement time falling below the predictions of ‘Alcator scaling’ by a factor of 3–4 at high beam power. An empirical relationship of the form fits our measurements over a wide range of plasma parameters. The function f(Pb), where Pb is the beam power, is linear for Pb ≤ 1.2 MW but tends to saturate for 1.2 MW ≤ Pb ≤ 2.5 MW. Although the equilibria attained in ISX-B are predicted to be above the threshold for the ideal magnetohydrodynamic (MHD) ballooning instability, no evidence of these modes is observed.