High-harmonic Fast-wave Power Research Articles

The role of rectified currents in far-field RF sheaths and in SOL losses of HHFW power on NSTX

Radio-frequency (RF) rectification is an important sheath phenomenon for wave heating of plasma in fusion devices and is proposed to be the mechanism responsible for converting high-harmonic fast-wave (HHFW) power in the National Spherical Torus eXperiment (NSTX) into a heat flux to the divertor. RF rectification has two aspects: current rectification and voltage rectification, and, while the latter is emphasized in many application, we demonstrate the importance of current rectification in analysis of the NSTX divertor during HHFW heating. When rectified currents are accounted for in first-principle models for the heat flux to the tiles, we predict a sizeable enhancement for the heat flux in the presence of an RF field: for one case studied, the predicted heat flux increases from 0.103 MW/m2 to 0.209 MW/m2. We also demonstrate how this rectification scales with injected HHFW power by tracking probe characteristics during a HHFW power ramp; the rectified current may be clamped at a certain level. This work is important for minimizing SOL losses of HHFW power in NSTX-U but may also have implications for near-field studies of ICRF antennae: ignoring rectified current may lead to underestimated heat fluxes and overestimated rectified voltages.

The contribution of radio-frequency rectification to field-aligned losses of high-harmonic fast wave power to the divertor in the National Spherical Torus eXperiment

The National Spherical Torus eXperiment (NSTX) can exhibit a major loss of high-harmonic fast wave (HHFW) power along scrape-off layer (SOL) field lines passing in front of the antenna, resulting in bright and hot spirals on both the upper and lower divertor regions. One possible mechanism for this loss is RF sheaths forming at the divertors. Here, we demonstrate that swept-voltage Langmuir probe characteristics for probes under the spiral are shifted relative to those not under the spiral in a manner consistent with RF rectification. We estimate both the magnitude of the RF voltage across the sheath and the sheath heat flux transmission coefficient in the presence of the RF field. Although precise comparison between the computed heat flux and infrared (IR) thermography cannot yet be made, the computed heat deposition compares favorably with the projections from IR camera measurements. The RF sheath losses are significant and contribute substantially to the total SOL losses of HHFW power to the divertor for the cases studied. This work will guide future experimentation on NSTX-U, where a wide-angle IR camera and a dedicated set of coaxial Langmuir probes for measuring the RF sheath voltage directly will quantify the contribution of RF sheath rectification to the heat deposition from the SOL to the divertor.

High-Harmonic Fast-Wave Power Flow along Magnetic Field Lines in the Scrape-Off Layer of NSTX

A significant fraction of high-harmonic fast-wave (HHFW) power applied to NSTX can be lost to the scrape-off layer (SOL) and deposited in bright and hot spirals on the divertor rather than in the core plasma. We show that the HHFW power flows to these spirals along magnetic field lines passing through the SOL in front of the antenna, implying that the HHFW power couples across the entire width of the SOL rather than mostly at the antenna face. This result will help guide future efforts to understand and minimize these edge losses in order to maximize fast-wave heating and current drive.

High non-inductive fraction H-mode discharges generated by high-harmonic fast wave heating and current drive in the National Spherical Torus Experiment

A deuterium H-mode discharge with a plasma current of 300 kA, an axial toroidal magnetic field of 0.55 T, and a calculated non-inductive plasma current fraction of 0.7–1 has been generated in the National Spherical Torus Experiment by 1.4 MW of 30 MHz high-harmonic fast wave (HHFW) heating and current drive. Seventy-five percent of the non-inductive current was generated inside an internal transport barrier that formed at a normalized minor radius ∼0.4. Three quarters of the non-inductive current was bootstrap current, and the remaining non-inductive current was generated directly by HHFW power inside a normalized minor radius ∼0.2.

Profiles of fast ions that are accelerated by high harmonic fast waves in the National Spherical Torus Experiment

Combined neutral beam injection and high-harmonic fast-wave (HHFW) heating accelerate deuterium fast ions in the National Spherical Torus Experiment (NSTX). With 1.1 MW of HHFW power, the neutron emission rate is about three times larger than in the comparison discharge without HHFW heating. Acceleration of fast ions above the beam injection energy is evident on an E||B type neutral particle analyzer (NPA), a 4-chord solid state neutral particle analyzer (SSNPA) array and a 16-channel fast-ion D-alpha (FIDA) diagnostic. The accelerated fast ions observed by the NPA and SSNPA diagnostics mainly come from passive charge exchange reactions at the edge due to the NPA/SSNPA localization in phase space. The spatial profile of accelerated fast ions that is measured by the FIDA diagnostic is much broader than in conventional tokamaks because of the multiple resonance layers and large orbits in NSTX. The fast-ion distribution function calculated by the CQL3D Fokker–Planck code differs from the measured spatial profile, presumably because the current version of CQL3D uses a zero-banana-width model. In addition, compressional Alfven eigenmode activity is stronger during the HHFW heating and it may affect the fast-ion spatial profile.

ECH and HHFW Start-Up Experiments on the TST-2 Spherical Tokamak

Plasma start-up experiments without using inductive field have been performed on the TST-2 spherical tokamak device. A low power ECH (2.45GHz/4kW) and a medium power High Harmonic Fast Wave (HHFW) (21MHz/up to 50kW) were used. Comparisons of various operational parameters were carried out to find the optimum conditions for the toroidal and the poloidal field strengths, poloidal field configurations, ECH power and HHFW power. Using only ECH, currents up to about 0.4 kA were generated. Current increments by up to 0.4 kA were observed when we apply HHFW to ECH plasmas.

Overview of the initial NSTX experimental results

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 theplasma 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 walllimited, 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 of27 kA produced up to 260 kA toroidal current without using an ohmic solenoid. With theinjection 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 acentral density of 3.5 × 1013 cm-3, increasing the plasma energy to 59 kJ and the toroidal β, βT, to 10%. The NBIsystem 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.