A major research goal of the national spherical torus experiment is establishing long-pulse, high beta, high confinement operation and its physics basis. This research has been enabled by facility capabilities developed during 2001 and 2002, including neutral beam (up to 7?MW) and high harmonic fast wave (HHFW) heating (up to 6?MW), toroidal fields up to 6?kG, plasma currents up to 1.5?MA, flexible shape control, and wall preparation techniques. These capabilities have enabled the generation of plasmas with of up to 35%. Normalized beta values often exceed the no-wall limit, and studies suggest that passive wall mode stabilization enables this for H mode plasmas with broad pressure profiles. The viability of long, high bootstrap current fraction operations has been established for ELMing H mode plasmas with toroidal beta values in excess of 15% and sustained for several current relaxation times. Improvements in wall conditioning and fuelling are likely contributing to a reduction in H mode power thresholds. Electron thermal conduction is the dominant thermal loss channel in auxiliary heated plasmas examined thus far. HHFW effectively heats electrons, and its acceleration of fast beam ions has been observed. Evidence for HHFW current drive is obtained by comparision of the loop voltage evolution in plasmas with matched density and temperature profiles but varying phases of launched HHFW waves. Studies of emissions from electron Bernstein waves indicate a density scale length dependence of their transmission across the upper hybrid resonance near the plasma edge that is consistent with theoretical predictions. A peak heat flux to the divertor targets of 10?MW?m?2 has been measured in the H mode, with large asymmetries being observed in the power deposition between the inner and outer strike points. Non-inductive plasma startup studies have focused on coaxial helicity injection. With this technique, toroidal currents up to 400?kA have been driven, and studies to assess flux closure and coupling to other current drive techniques have begun.

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The national spherical torus experiment (NSTX) research programme and progress towards high beta, long pulse operating scenarios

Nuclear Fusion Space Propulsion Research, Experimentation, and Analysis Effort led by the NASA Glenn Research Center (1996-2004)

The National Spherical Torus Experiment (NSTX) is the first step in an investigation of the physics principles of low-aspect-ratio spherical tori (ST) designed to study non-inductive start-up, current sustainability, current profile control, confinement, transport, pressure limits, stability and disruption resilience, as well as unique scrape-off layer (SOL) and divertor characteristics. NSTX started plasma operations in February 1999. During the first extended period of experiments, plasma discharge reproducibility and performance were strongly affected by impurity control and wall conditions. During this time, residual gas analyzer (RGA) data has been used during non-operating periods and between discharges to measure contributions to the vessel base-pressure (2-3E/sup -8/ Torr) that resulted from impurity gases evolving from internal surfaces. The RGA is a quadrupole mass spectrometer (QMS) with a Faraday Cup and electron multiplier. Recently, a second RGA system was installed to monitor fast changes in impurity gas production during and immediately after plasma discharges, which are indicative of changing conditions. In order to achieve the required response time, scan rates were increased by limiting the range of masses scanned. A data acquisition board was used to drive the new RGA system in concurrence with the NSTX shot cycle. Specialized interactive software to put acquired data into MDSplus, control data acquisition and assist in data analysis was written in Visual Basic and IDL.

Residual gas analysis hardware and software data acquisition system at NSTX

A major research goal of the national spherical torus experiment is establishing long-pulse, high beta, high confinement operation and its physics basis. This research has been enabled by facility capabilities developed during 2001 and 2002, including neutral beam (up to 7MW) and high harmonic fast wave (HHFW) heating (up to 6 MW), toroidal fields up to 6 kG, plasma currents up to 1.5 MA, flexible shape control, and wall preparation techniques. These capabilities have enabled the generation of plasmas with {beta}{sub T} {equivalent_to} {l_angle}p{r_angle}/(B{sup 2}{sub T0}/2{mu}{sub 0}) of up to 35%. Normalized beta values often exceed the no-wall limit, and studies suggest that passive wall mode stabilization enables this for H mode plasmas with broad pressure profiles. The viability of long, high bootstrap current fraction operations has been established for ELMing Hmode plasmas with toroidal beta values in excess of 15% and sustained for several current relaxation times. Improvements in wall conditioning and fuelling are likely contributing to a reduction in Hmode power thresholds. Electron thermal conduction is the dominant thermal loss channel in auxiliary heated plasmas examined thus far. HHFW effectively heats electrons, and its acceleration of fast beam ions has been observed. Evidence for HHFW current drive ismore » obtained by comparison of the loop voltage evolution in plasmas with matched density and temperature profiles but varying phases of launched HHFW waves. Studies of emissions from electron Bernstein waves indicate a density scale length dependence of their transmission across the upper hybrid resonance near the plasma edge that is consistent with theoretical predictions. A peak heat flux to the divertor targets of 10MWm{sup -2} has been measured in the H mode, with large asymmetries being observed in the power deposition between the inner and outer strike points. Non-inductive plasma startup studies have focused on coaxial helicity injection. With this technique, toroidal currents up to 400 kA have been driven, and studies to assess flux closure and coupling to other current drive techniques have begun.« less

The NSTX (National Spherical Torus Experiment) facility located at Princeton Plasma Physics Laboratory is the newest national fusion science experimental facility for the restructured US Fusion Energy Science Program. The NSTX project was approved in FY 97 as the first proof-of-principle national fusion facility dedicated to the spherical torus research. On Feb. 15, 1999, the first plasma was achieved 10 weeks ahead of schedule. The project was completed on budget and with an outstanding safety record. This paper gives an overview of the NSTX facility construction and the initial plasma operations.

Making of the NSTX facility

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Making of the NSTX facility