[en] The main aim of the National Spherical Torus Experiment (NSTX) is to establish the fusion physics principles of the spherical torus (ST) concept. The NSTX device began plasma operations in February 1999 and the plasma current Ip was successfully brought up to the design value of 1 million amperes on December 14, 1999. The planned plasma shaping parameters, k = 1.6 ± 2.2 and d = 0.2 ± 0.4, were achieved in inner limited, single null and double null configurations. The CHI (Coaxial Helicity Injection) and HHFW (High Harmonic Fast Wave) experiments were also initiated. A CHI injected current of 27 kA produced up to 260 kA of toroidal current without using an ohmic solenoid. With an injection of 2.3 MW of HHFW power, using twelve antennas connected to six transmitters, electrons were heated from a central temperature of 400 eV to 900 eV at a central density of 3.5 x 1013 cm-3 increasing the plasma energy to 59 kJ and the toroidal beta, bT to 10 %. Finally, the NBI system commenced operation in Sept. 2000. The initial results with two ion sources (PNBI = 2.8 MW) shows good heating, producing a total plasma stored energy of 90 kJ corresponding to bT = 18 % at a plasma current of 1.1 MA

[en] An overview of the research capabilities and the future plans on the MA-class National Spherical Torus Experiment (NSTX) at Princeton is presented. NSTX research is exploring the scientific benefits of modifying the field line structure from that in more conventional aspect ratio devices, such as the tokamak. The relevant scientific issues pursued on NSTX include energy confinement, MHD stability at high beta, non-inductive sustainment, solenoid-free start-up, and power and particle handling. In support of the NSTX research goal, research tools are being developed by the NSTX team. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a high beta Demo device based on the ST, are being considered. For these, it is essential to develop high performance (high beta and high confinement), steady-state (non-inductively driven) ST operational scenarios and an efficient solenoid-free start-up concept. We will also briefly describe the Next-Step-ST (NSST) device being designed to address these issues in fusion-relevant plasma conditions

[en] Power and particle balance studies on the Advanced Toroidal Facility (ATF) torsatron are carried out using a rail limiter system. Both top and bottom limiters are made of graphite tile arrays, and these tiles are instrumented with thermocouples and Langmuir probes for calorimetric and particle flux measurements. Initial experimental results indicate that the limiter power loss accounts for about 12% of the total and the radiation loss for about 30% of the total; the rest of the plasma heating power appears to be going to the vessel wall. The particle flux to the limiters is also about 18%. The fractions of power and particle flux to the limiters are relatively lower than in Tokamaks because of the low edge safety factor, q ∼ 1 rather than q ∼ 3 as in a typical Tokamak, at the natural boundary of the ATF plasma (which results from the magnetic stellarator configuration of this currentless device). Therefore, for limiters of the same size, these fractions are about a factor of q lower in ATF than in a comparable Tokamak device. (Author)

[en] Mode-conversion between the ordinary, extraordinary and electron Bernstein modes near the plasma edge may allow signals generated by electrons in an over-dense plasma to be detected. Alternatively, high frequency power may gain accessibility to the core plasma through this mode conversion process. Many of the tools used for ion cyclotron antenna design can also be applied near the electron cyclotron frequency. In this paper, we investigate the the possibilities for an antenna that may couple to electron Bernstein modes inside an over-dense plasma. The optimum values for wavelengths that undergo mode-conversion are found by scanning the poloidal and toroidal response of the plasma using a warm plasma slab approximation with a sheared magnetic field. Only a very narrow region of the edge can be examined in this manner; however, ray tracing may be used to follow the mode converted power in a more general geometry. It is eventually hoped that the methods can be extended to a hot plasma representation. Using antenna design codes, some basic antenna shapes will be considered to see what types of antennas might be used to detect or launch modes that penetrate the cutoff layer in the edge plasma

[en] Recent upgrades to the NSTX facility have led to improved plasma performance. Using 5MW of neutral beam injection, plasmas with toroidal β_T (= 2(micro)₀< p>/B_T² where B_T is the vacuum toroidal field at the plasma geometric center) > 30% have been achieved with normalized β_N (= β_TaB_I/I_p) ∼ 6% · m · T/MA.. The highest β discharge exceeded the calculated no-wall β limit for several wall times. The stored energy has reached 390kJ at higher toroidal field (0.55T) corresponding to β_T ∼ 20% and β_N = 5.4. Long pulse (∼1s) high β_p (∼1.5) discharges have also been obtained at higher β_φ (0.5T) with up to 6MW NBI power. The highest energy confinement times, up to 120ms, were observed during H-mode operation which is now routine. Confinement times of ∼1.5 times ITER98pby2 for several τ_E are observed during both H-Mode and non-H-Mode discharges. Calculations indicate that many NSTX discharges have very good ion confinement, approaching neoclassical levels. High Harmonic Fast Wave current drive has been demonstrated by comparing discharges with waves launched parallel and anti-parallel to the plasma current

[en] Recent upgrades to the NSTX facility have led to improved plasma performance. Using 5MW of neutral beam injection, plasmas with toroidal β_T (= 2μ₀< p>/B_T² where B_T is the vacuum toroidal field at the plasma geometric center) > 30% have been achieved with normalized β_N (=β_TaB_I/I_p) ∼ 6% · m · T/MA. The highest β discharge exceeded the calculated no-wall β limit for several wall times. The stored energy has reached 390kJ at higher toroidal field (0.55T) corresponding to β_T ∼ 20% and β_N = 5.4. Long pulse (∼1s) high β_p (∼1.5) discharges have also been obtained at higher B_φ (0.5T) with up to 6MW NBI power. The highest energy confinement times, up to 120ms, were observed during H-mode operation which is now routine. Confinement times of ∼ 1.5 times ITER98pby2 for several τ_E are observed during both H-Mode and non-H-Mode discharges. Calculations indicate that many NSTX discharges have very good ion confinement, approaching neoclassical levels. High Harmonic Fast Wave current drive has been demonstrated by comparing discharges with waves launched parallel and anti-parallel to the plasma current.

[en] An effort is underway to improve the voltage standoff capabilities of ion cyclotron range of frequencies (ICRF) heating and current drive systems. One approach is to develop techniques for determining the location of an electrical breakdown (arc) when it occurs. A technique is described which uses a measurement of the reflection coefficient of a swept frequency signal to determine the arc location. The technique has several advantages including a requirement for only a small number of sensors and very simple data interpretation. In addition a test stand is described which will be used for studies of rf arc behavior. The device uses a quarter-wave resonator to produce voltages to 90 kV in the frequency range of 55-80 MHz

[en] Full text: Linear plasma generators are cost effective facilities to simulate divertor plasma conditions of present and future fusion reactors. They are used to address important R&D gaps in the science of plasma material interactions and towards viable plasma facing components for fusion reactors. Next generation plasma generators have to be able to access the plasma conditions expected on the divertor targets in ITER and future devices. The steady-state linear plasma device MPEX will address this regime with electron temperatures of 1–10 eV and electron densities of 10²¹–10²⁰/m³. The resulting heat fluxes are about 10 MW/m². MPEX is designed to deliver those plasma conditions with a novel radio frequency plasma source able to produce high density plasmas and heat electron and ions separately with electron Bernstein wave (EBW) heating and ion cyclotron resonance heating (ICRH) with a total installed power of 800 kW. The linear device Proto-MPEX, forerunner of MPEX consisting of 12 water-cooled copper coils, is operational since May 2014. Its helicon antenna (100 kW, 13.56 MHz) and EC heating systems (200 kW, 28 GHz) have been commissioned. The operational space was expanded considerably in the last year. 12 MW/m² was delivered on target. Electron temperatures of about 20 eV have been achieved in combined helicon and ECH/EBW heating schemes at low electron densities. Overdense heating with electron Bernstein waves was achieved at low heating powers. The operational space of the density production by the helicon antenna was pushed to 2 x 10¹⁹/m³ at relatively high magnetic fields of 0.7 T, which would allow ECH absorption for 2nd harmonic X-mode overdense heating schemes. Proto-MPEX has been prepared to allow for first material sample exposures. The experimental results from Proto-MPEX will be used for code validation (B2-EIRENE, COMSOL, VORPAL, AORSA, GENRAY) to enable predictions of the source and heating performance for MPEX. MPEX, in its last phase, will be capable to expose neutron-irradiated samples. In this concept, targets will be irradiated in ORNL’s High Flux Isotope Reactor and then subsequently exposed to fusion reactor relevant plasmas in MPEX. The current state of the MPEX preconceptual design and unique technologies already developed, including the concept of handling irradiated samples, will be presented. (author)

[en] 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 over the last two years, including neutral-beam (up to 7 MW) and high-harmonic fast-wave 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 _T> up to 35%. Normalized beta values often exceed the no wall limit, and studies suggest that passive wall mode stabilization is enabling this for broad pressure profiles characteristic of H-mode plasmas. 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 fueling 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. High-harmonic fast-wave (HHFW) effectively heats electrons, and its acceleration of fast beam ions has been observed. Evidence for HHFW current drive is by comparing of the loop voltage evolution in plasmas with matched density and temperature profiles but varying phases of launched HHFW waves. A peak heat flux of 10 MW/m superscript ''2'' has been measured in the H-mode, with large asymmetries in the power deposition being observed between the inner and outer strike points. Noninductive plasma start-up 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

[en] Coaxial Helicity Injection (CHI) has been successfully used in the National Spherical Torus Experiment (NSTX) for a demonstration of closed flux current generation without the use of the central solenoid. The favorable properties of the Spherical Torus (ST) arise from its very small aspect ratio. However, small aspect ratio devices have very restricted space for a substantial central solenoid. Thus methods for initiating the plasma current without relying on induction from a central solenoid are essential for the viability of the ST concept. CHI is a promising candidate for solenoid-free plasma startup in a ST. The method has now produced closed flux current up to 160 kA verifying the high current capability of this method in a large ST built with conventional tokamak components.