Exploration of spherical torus physics in the NSTX device

Abstract

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,

E × B Plasma Rotation and n = 1 Oscillation Observed in the NSTX-CHI Experiments

“…This paper shows that the E × B plasma rotation along with the n = 1 distortion is verified by ion Doppler spectroscopic measurements for the long pulse high current discharge demonstrated on NSTX. This result obtained on the NSTX is consistent with observations on the HIT-II.The NSTX device parameters are: major/minor radii of 0.85/0.65 m, elongation ≤ 2.5, plasma volume 12.5 m 3author's e-mail: nagata@eng.u-hyogo.ac.jp [5]. The NSTX vacuum vessel is split in two using toroidal ceramic breaks at the top and bottom to electrically insulate the central column and the inner divertor plates from the outer wall and the outer divertor plates as shown in Fig.…”

“…author's e-mail: nagata@eng.u-hyogo.ac.jp [5]. The NSTX vacuum vessel is split in two using toroidal ceramic breaks at the top and bottom to electrically insulate the central column and the inner divertor plates from the outer wall and the outer divertor plates as shown in Fig.…”

Intermittency in the scrape-off layer of the National Spherical Torus Experiment during H-mode confinement

“…The measurements described in this paper were obtained in the National Spherical Torus Experiment (NSTX) [15]. This experiment is a low aspect ratio tokamak with a major radius R = 0.86 m and a minor radius a = 0.67 m. Discharges with plasma current I p = 0.8-1.2 MA and toroidal field on axis B t = 0.45-0.50 T were used with auxiliary heating in the form of neutral beam injection (NBI) raging from 0 MW (i.e., only Ohmic heating) to 7 MW.…”

“…In this paper we present results from the National Spherical Torus Experiment (NSTX) [15]. In this experiment both the turbulence power and frequency of intermittent blobs are reduced in H-mode compared to L-mode.…”

Time-Scales for Non-Inductive Current Buildup in Low-Aspect-Ratio Toroidal Geometry

“…This interest stems from the fundamental geometry of a low-aspect-ratio torus, which precludes the presence of a large internal solenoid or "OH" coil. While present laboratory experiments [2][3][4][5] are equipped with relatively small OH coils for inductive current buildup, it is expected that the next generation of experiments and reactor concepts will rely totally on non-inductive techniques [6] .…”