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,

As part of the ITER Design Review, the physics requirements were reviewed and as appropriate updated. The focus of this paper will be on recent work affecting the ITER design with special emphasis on topics affecting near-term procurement arrangements. This paper will describe results on: design sensitivity studies, poloidal field coil requirements, vertical stability, effect of toroidal field ripple on thermal confinement, heat load requirements for plasma-facing components, edge localized modes control, resistive wall mode control, disruptions and disruption mitigation.

Abstract-The spherical tokamak (ST) is a leading candidate for a fusion nuclear science facility (FNSF) due to its compact size and modular configuration. The National Spherical Torus eXperiment (NSTX) is a MA-class ST facility in the U.S. actively developing the physics basis for an ST-based FNSF. In plasma transport research, ST experiments exhibit a strong (nearly inverse) scaling of normalized confinement with collisionality, and if this trend holds at low collisionality, high fusion neutron fluences could be achievable in very compact ST devices. A major motivation for the NSTX Upgrade (NSTX-U) is to span the next factor of 3-6 reduction in collisionality. To achieve this collisionality reduction with equilibrated profiles, NSTX-U will double the toroidal field, plasma current, and NBI heating power and increase the pulse length from 1-1.5s to 5s. In the area of stability and advanced scenarios, plasmas with higher aspect ratio and elongation, high βN , and broad current profiles approaching those of an ST-based FNSF have been produced in NSTX using active control of the plasma β and advanced resistive wall mode control. High non-inductive current fractions of 70% have been sustained for many current diffusion times, and the more tangential injection of the 2nd NBI of the Upgrade is projected to increase the NBI current drive by up to a factor of 2 and support 100% non-inductive operation. More tangential NBI injection is also projected to provide non-solenoidal current ramp-up (from IP = 0.4MA up to 0.8-1MA) as needed for an ST-based FNSF. In boundary physics, NSTX and higher-A tokamaks measure an inverse relationship between the scrape-off layer heat-flux width and plasma current that could unfavorably impact nextstep devices. Recently, NSTX has successfully demonstrated very high flux expansion and substantial heat-flux reduction using a snowflake divertor configuration, and this type of divertor is incorporated in the NSTX-U design. The physics and engineering design supporting NSTX Upgrade are described.

The Helicity Injected Torus [T. R. Jarboe, Fusion Technol. 15, 7 (1989)] is a low aspect ratio tokamak that is formed and sustained by coaxial helicity injection with no transformer. Toroidal plasma currents of over 200 kA have been achieved with electron temperatures in the 100 eV range and electron density between 1019 and 1020 m−3. The major radius is 0.3 m and the minor radius is 0.2 m. New results from equilibrium and stability analysis of the external magnetic diagnostics and new results from the Transient Internal Probe (TIP), an internal magnetic field diagnostic, are presented. A mechanism for the transfer of current drive on the open to the closed flux regions is presented.

Theoretical studies have predicted that the Z-pinch can be stabilized with a sufficiently sheared axial flow [U. Shumlak and C. W. Hartman, Phys. Rev. Lett. 75, 3285 (1995)]. A Z-pinch experiment is designed to generate a plasma which contains a large axial flow. Magnetic fluctuations and velocity profiles in the plasma pinch are measured. Experimental results show a stable period which is over 700 times the expected instability growth time in a static Z-pinch. The experimentally measured axial velocity shear is greater than the theoretical threshold during the stable period and approximately zero afterwards when the magnetic mode fluctuations are high.