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The eight triplets of straps of the ITER ICRF antenna array are fed through 8 matching circuits and 4 hybrids to ensure load resilience. Decouplers are used to mitigate the effects of triplet mutual coupling. They also control the array phasing. The electrical constraints on the decouplers for different layouts with heating (H) or current drive (CD) phasing are compared starting from the TOPICA matrix computed for the last antenna plug design and the reference (most pessimistic) plasma profile “2010low” provided by IO. It is shown that this last profile provides a significant decrease of plasma coupling and increase of mutual coupling with respect to the previous reference profile “Sc2short17”. This results in a larger range of decoupler reactance Xdec and voltage VXdec needed. This range can be reduced when using 10 decouplers instead of the 7 needed for the same forward power PGk+ of the 4 power sources. For H phasing only 4 decouplers could be used but with different PGk+ (PGk+ ratio up to 1.5–2.5). For CD phasing and same plasma profile the power capability Ptot is increased by 25% with a decoupler layout allowing much smaller poloidal phasing than the 90° provided by the hybrids. A decrease of the distance antenna-plasma profile reduces the normalized decoupler voltage VXdec/√Ptot with no significant change of the Xdec range. The recess of the vertical septa between the strap boxes increases the plasma coupling but has the drawback of also increasing the mutual coupling between triplets: the needed range of Xdec and of VXdec/√Ptot is increased in proportion.

Ion absorption of the high harmonic fast wave in a spherical torus [Y.-K. M. Peng et al., Nucl. Fusion 26, 769 (1986)] is of critical importance to assessing the viability of the wave as a means of heating and driving current. Analysis of recent National Spherical Torus Experiment [M. Ono et al., Nucl. Fusion 40, 557 (2000)] shots has revealed that under some conditions when neutral beam and rf power are injected into the plasma simultaneously, a fast ion population with energy above the beam injection energy is sustained by the wave. In agreement with modeling, these experiments find the rf-induced fast ion tail strength and neutron rate at lower B-fields to be less enhanced, likely due to a larger β profile, which promotes greater off-axis absorption where the fast ion population is small. Ion loss codes find the increased loss fraction with decreased B insufficient to account for the changes in tail strength, providing further evidence that this is a rf interaction effect. Though greater ion absorption...

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.

High harmonic fast wave (HHFW) heating has been proposed as a particularly attractive means for plasma heating and current drive in the high beta plasmas that are achievable in spherical torus (ST) devices. The National Spherical Torus Experiment (NSTX) [M. Ono, S. M. Kaye, S. Neumeyer et al., in Proceedings of the 18th IEEE/NPSS Symposium on Fusion Engineering, Albuquerque, 1999 (IEEE, Piscataway, NJ, 1999), p. 53] is such a device. An rf heating system has been installed on the NSTX to explore the physics of HHFW heating, current drive via rf waves and for use as a tool to demonstrate the attractiveness of the ST concept as a fusion device. To date, experiments have demonstrated many of the theoretical predictions for HHFW. In particular, strong wave absorption on electrons over a wide range of plasma parameters and wave parallel phase velocities, wave acceleration of energetic ions, and indications of current drive for directed wave spectra have been observed. In addition HHFW heating has been used to ...

Edge parameters play a critical role in high confinement mode (H-mode) access, which is a key component of discharge optimization in present day toroidal confinement experiments and the design of next generation devices. Because the edge magnetic topology of a spherical torus (ST) differs from a conventional aspect ratio tokamak, H-modes in STs exhibit important differences compared with tokamaks. The dependence of the National Spherical Torus Experiment (NSTX) [C. Neumeyer et al., Fusion Eng. Des. 54, 275 (2001)] edge plasma on heating power, including the low confinement mode (L-mode) to H-mode (L-H) transition requirements and the occurrence of edge-localized modes (ELMs), and on divertor configuration is quantified. Comparisons between good L-modes and H-modes show greater differences in the ion channel than the electron channel. The threshold power for the H-mode transition in NSTX is generally above the predictions of a recent International Tokamak Experimental Reactor (ITER) [ITER Physics Basis Edi...

The National Spherical Torus Experiment (NSTX) at Princeton University, Princeton, NJ, is designed to assess the potential of the low-aspect-ratio spherical torus concept for magnetic plasma confinement. The plasma has been heated by up to 7 MW of neutral beam injection (NBI) at an injection energy of 100 keV and up to 6 MW of high harmonic fast wave (HHFW) at 30 MHz. NSTX has achieved /spl beta//sub T/ of 32%. A variety of MHD phenomena have been observed to limit /spl beta/. NSTX has now begun addressing /spl tau//sub E/ scaling, /spl beta/ limits, and current drive issues. During the NBI heating experiments, a broad T/sub i/ profile with T/sub i/ up to 2 keV, T/sub i/>T/sub e/ and a large toroidal rotation were observed. Transport analysis suggests that the impurity ions have diffusivities approaching neoclassical. For L-Mode plasmas, /spl tau//sub E/ is up to two times the ITER97L L-Mode scaling and exceeds the ITER98pby2 H-Mode scaling in some cases. Transitions to H-Mode have been observed which result in an approximate doubling of /spl tau//sub E/ after the transition in some conditions. During HHFW heating, T/sub e/>T/sub i/ and T/sub e/ up to 3.5 keV were observed. Current drive has been studied using both coaxial helicity injection with up to 390 kA of toroidal current and HHFW. HHFW has produced H-modes with significant bootstrap current fraction at low I/sub p/, high q, and high /spl beta//sub p/.

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 MA on 14 December 1999. The planned plasma shaping parameters, elongation κ = 1.6-2.2 and triangularity δ = 0.2-0.4, were achieved in inner wall limited, and single null and double null diverted configurations. The coaxial helicity injection (CHI) and high harmonic fast wave (HHFW) experiments were also initiated. CHI current of 27 kA produced up to 260 kA toroidal current without using an ohmic solenoid. With the injection of 2.3 MW of HHFW power, using 12 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 × 1013 cm-3, increasing the plasma energy to 59 kJ and the toroidal β, βT, to 10%. The NBI system commenced operation in September 2000. The initial results with two ion sources (PNBI = 2.8 MW) show good heating, producing a total plasma stored energy of 90 kJ corresponding to βT ≈ 18% at a plasma current of 1.1 MA.

The mission of the National Spherical Torus Experiment (NSTX) is to extend the understanding of toroidal physics to low aspect ratio (R/a approximately equal to 1.25) in low collisionality regimes. NSTX is designed to operate with up to 6 MW of High Harmonic Fast Wave (HHFW) heating and current drive, 5 MW of Neutral Beam Injection (NBI) and Co-Axial Helicity Injection (CHI) for non-inductive startup. Initial experiments focused on establishing conditions that will allow NSTX to achieve its aims of simultaneous high-bt and high-bootstrap current fraction, and to develop methods for non-inductive operation, which will be necessary for Spherical Torus power plants. Ohmic discharges with plasma currents up to 1 MA and with a range of shapes and configurations were produced. Density limits in deuterium and helium reached 80% and 120% of the Greenwald limit respectively. Significant electron heating was observed with up to 2.3 MW of HHFW. Up to 270 kA of toroidal current for up to 200 msec was produced noninductively using CHI. Initial NBI experiments were carried out with up to two beam sources (3.2 MW). Plasmas with stored energies of up to 140 kJ and bt =21% were produced.

The ability to measure rf driven waves in the edge of the plasma can help to elucidate the role that surface waves and parametric decay instabilities (PDIs) play in rf power losses on NSTX. A microwave reflectometer has recently been modified to monitor rf plasma waves in the scrape-off layer in front of the 30MHz high harmonic fast wave antenna array on NSTX. In rf heated plasmas, the plasma-reflected microwave signal exhibits 30MHz sidebands, due primarily to the modulation of the cutoff layer by the electrostatic component of the heating wave. Similarly, electrostatic parametric decay waves (when present) are detected at frequencies below the heating frequency, near 28, 26,…MHz, separated from the heating frequency by harmonics of the local ion cyclotron frequency of about 2MHz. In addition, a corresponding frequency matched set of decay waves is also detected near the ion cyclotron harmonics at 2, 4,…MHz. The rf plasma-wave sensing capability is useful for determination of the PDI power threshold as a...

The National Spherical Torus Experiment (NSTX) has explored the effects of shaping on plasma performance as determined by many diverse topics including the stability of global magnetohydrodynamic (MHD) modes (e.g., ideal external kinks and resistive wall modes), edge localized modes (ELMs), bootstrap current drive, divertor flux expansion, and heat transport. Improved shaping capability has been crucial to achieving Βt ∼40%. Precise plasma shape control has been achieved on NSTX using real-time equilibrium reconstruction. NSTX has simultaneously achieved elongation κ∼2.8 and triangularity δ∼0.8. Ideal MHD theory predicts increased stability at high values of shaping factor S≡ q95 Ip (a Bt), which has been observed at large values of the S∼37 [MA (m·T)] on NSTX. The behavior of ELMs is observed to depend on plasma shape. A description of the ELM regimes attained as shape is varied will be presented. Increased shaping is predicted to increase the bootstrap fraction at fixed Ip. The achievement of strong shaping has enabled operation with 1 s pulses with Ip =1 MA, and for 1.6 s for Ip =700 kA. Analysis of the noninductive current fraction as well as empirical analysis of the achievable plasma pulse length as elongation is varied will be presented. Data are presented showing a reduction in peak divertor heat load due to increasing in flux expansion. © 2006 American Institute of Physics.

Plasmas with volume-averaged beta equal to 3% have been produced in beam-heated ISX-B discharges. Previous modelling of such ISX-B plasmas, at somewhat lower beta and with limited detail as regards plasma shape, has shown that high-n ballooning instabilities should be excited, although none is observed. The authors now use data that allow them to specify free-boundary equilibria, which are influenced by the ISX-B poloidal coil system and iron core, and compare these equilibria with criteria for ideal magnetohydrodynamic (MHD) stability to both low- and high-n number ballooning and kink modes. Both analytic estimates and codes are used in the comparison. It is found that the equilibria studied are predicted to be unstable to both low- and high-n modes, in contrast to the experimental results. Several candidates for resolving these differences are discussed.

The ISX-B tokamak at ORNL is designed to have 1.8 MW (and eventually 3 MW) of neutral beam power injected to heat the plasma. This power may raise the s? of the plasna to over 5 percent in less than 50 ms, if the plasma is MHD stable. The results of a numerical simulation of the feedback control system and poloidal coil power supplies necessary to control the resulting noncircular (D-shaped or elliptical) plasma are presented. The resulting feedback control system is shown to be straightforward, although nonlinear voltage-current dependence is assumed in the power supplies. The required power supplied to the poloidal coils in order to contain the plasma under the high heating rates is estimated.