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,

levels which might have a significant role in the light shift of the 22p level due to the 1.06-/im laser field are 6s, 7s, Ad, and 5d. These are far from being resonantly coupled to the 22p level, at least 1700 cm" 1 away. Their relative positions are such that their combined effects are partially cancelled* A rough evaluation showed that under these conditions the 5d level, which is expected to be responsible for the largest effect, contributes to the shift of the 22p level an amount of approximately 3xl0" 3 MHz/ MW-cm' 2 . This is at least 4 orders of magnitude less than the measured shift, and is thus completely negligible, With respect to the shift Lv g of the ground state, since it cannot be measured alone the best procedure is to calculate it as carefully and precisely possible. A calculation based on Fig. 1 has been carried out. 6 The result is &v g = -26.3 MHz/MW-cm" 2 . The dashed line in Fig. 3 corresponds to the sum of the two calculated shifts Ai/ e + Ay g , whereas the straight line corresponds to a least-squares fit on the measured shifts. Agreement between experimental and theoretical results is satisfactory.To conclude, this experiment provides clear evidence for the shift of a Rydberg level, due to an intense and strongly nonresonant em field. It is of interest to note that in a pure quantum treat-PACS numbers: 52.55.Gb, 52.35.Py On the PDX tokamak, large-amplitude magnetohydrodynamic (MHD) fluctuations have been observed during plasma heating by injection of high-ment, radiative corrections can be interpreted as the sum of spontaneous and stimulated radiative corrections. The net effect of spontaneous radiative corrections due to vacuum fluctuations is well known to be responsible for the Lamb shift. In the same spirit, the light shifts which have been studied in our experiment can perhaps be viewed as resulting from the stimulated radiative corrections induced by an intense and nonresonant em field.We thank Professor CI. Cohen-Tannoudji for many helpful discussions concerning both the experiment and its interpretation. We are indebted to Dr. M. Aymar and Dr. M. Crance for their calculation of the shift of the ground state.Strong magnetohydrodynamic activity has been observed in PDX neutral-be am-heated discharges. It occurs for fi T q^ 0.045 and is associated with a significant loss of fast ions and a drop in neutron emission. As much as 20%~-40% of the beam heating power may be lost. The instability occurs in repetitive bursts of oscillations of ^ 1 msec duration at 1-6-msec intervals. The magnetohydrodynamic activity has been dubbed the "fishbone instability" from its characteristic signature on the Mirnov coils.

Polarimetry measurements of the Doppler-shifted H fl emission from a neutral hydrogen beam on the PBX-M tokamak have been employed in a novel technique for obtaining q{r) and magnetic field pitchangle profiles using the Stark effect. The resulting q(r) profile is very broad and its central value, #(0), is significantly below 1, which has important implications for theoretical models of sawteeth. PACS numbers: 52.70.Ds, 52.70.Kz The current-density profile and the safety factor, q(r), are important in the theoretical modeling of plasma equilibrium, stability, and confinement. Many observed phenomena, however, are not well understood due to the lack of consistent and detailed experimental data. For example, current-driven instabilities have been observed or predicted under various conditions, and are known to affect plasma confinement and transport. Since such discharges generally involve auxiliary heating, q(r) measurements in the past have not been available in the resulting high-density plasmas. In addition, when #(0) < 1, models predict the plasma to be unstable to resistive kink modes, which are believed to be responsible for sawtooth oscillations in the central temperature. Several theoretical models have been proposed to explain these observations, but to distinguish between them, a detailed knowledge of the safety-factor profile evolution is needed, including evidence beyond experimental uncertainty for a drop in the central #(0) below unity. There are many other instability issues related to the current or q(r) profile, such as the understanding of ballooning or fishbone modes, that could be resolved with detailed profile information.Techniques for active control of the current density, and the q(r) profile by using rf, neutral beams, or other means are presently underway or being planned. To understand the effects of modifying current-density profiles on stability, confinement, and transport, a credible means of measuring the poloidal field distribution is necessary. Here again, such techniques must be operable in high-density, auxiliary-heated plasmas that are at or near the reactorlike parameters of the present generation of fusion devices.To date, methods for obtaining current-density and safety-factor profiles have suffered from several problems. Frequently, the measurement is line integrated, which requires a numerical inversion process to obtain spatial information. 1,2 This introduces some uncertainty, particularly for noncircular plasma shapes. Neutralbeam probe techniques have shown good spatial resolution, but so far suffered from beam attenuation for densities above -lxl0 13 cm~3. 3 ' 4 In this Letter we report the first measurements of the central rotational transform, #(0), using the motional Stark effect (MSE) to polarize the spectral emission from a 55-keV neutral-hydrogen beam. One of the principal advantages of this technique is that it can provide a local (~l-2-cm resolution) and highly accurate measurement of the pitch angle of the magnetic field, Y p (r) =tan~1[5 p (/')/ J 5r(r)],...

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.