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.

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High harmonic fast wave heating and current drive (CD) are being developed on the National Spherical Torus Experiment [M. Ono et al., Nucl. Fusion 41, 1435 (2001)] for supporting startup and sustainment of the spherical torus plasma. Considerable enhancement of the core heating efficiency (η) from 44% to 65% has been obtained for CD phasing of the antenna (strap-to-strap ϕ=−90°, kϕ=−8m−1) by increasing the magnetic field from 4.5to5.5kG. This increase in efficiency is strongly correlated to moving the location of the onset density for perpendicular fast wave propagation (nonset∝B×k∥2∕ω) away from the antenna face and wall, and hence reducing the propagating surface wave fields. Radio frequency (RF) waves propagating close to the wall at lower B and k∥ can enhance power losses from both the parametric decay instability (PDI) and wave dissipation in sheaths and structures around the machine. The improved efficiency found here is attributed to a reduction in the latter, as PDI losses are little changed at the higher magnetic field. Under these conditions of higher coupling efficiency, initial measurements of localized CD effects have been made and compared with advanced RF code simulations.

The time scales for sawtooth repetition and heat pulse propagation are much longer (10's of msec) in the large tokamak TFTR than in previous, smaller tokamaks. This extended time scale coupled with more detailed diagnostics has led us to revisit the analysis of the heat pulse propagation as a method to determine the electron heat diffusivity, x e) in the plasma. A combination of analytic and computer solutions of the electron heat diffusion equation are used to clarify previous work and develop new methods for determining x e-Direct comparison of the predicted heat pulses with soft X-ray and ECE data indicates that the space-time evolution is diffusive. However, the x e determined from heat pulse propagation usually exceeds that determined from background plasma power balance considerations by a factor ranging from 2 to 10, Some hypotheses for resolving this discrepancy are discussed.

Experiments in the National Spherical Torus Experiment (NSTX) have shown beneficial effects on the performance of divertor plasmas as a result of applying lithium coatings on the graphite and carbonfiber-composite plasma-facing components. These coatings have mostly been applied by a pair of lithium evaporators mounted at the top of the vacuum vessel which inject collimated streams of lithium vapor towards the lower divertor. In NBI-heated, deuterium H-mode plasmas run immediately after the application of lithium, performance modifications included decreases in the plasma density, particularly in the edge, and inductive flux consumption, and increases in the electron and ion temperatures and the energy confinement time. Reductions in the number and amplitude of ELMs were observed, including complete ELM suppression for periods up to 1.2 s, apparently as a result of altering the stability of the edge. However, in the plasmas where ELMs were suppressed, there was a significant secular increase in the effective ion charge Z eff and the radiated power as a result of increases in the carbon and medium-Z metallic impurities, although not of lithium itself which remained at a very low level in the plasma core, <0.1%. The impurity buildup could be inhibited by repetitively triggering ELMs with the application of brief pulses of an n = 3 radial field perturbation. The reduction in the edge density by lithium also inhibited parasitic losses through the scrape-off layer of ICRF power coupled to the plasma, enabling the waves to heat electrons in the core of H-mode plasmas produced by NBI. Lithium has also been introduced by injecting a stream of chemically stabilized, fine lithium powder directly into the scrape-off layer of NBI-heated plasmas. The lithium was ionized in the SOL and appeared to flow along the magnetic field to the divertor plates. This method of coating produced similar effects to the evaporated lithium but at lower amounts.