Spherical Tokamak Configuration Research Articles

Multi-harmonic electron cyclotron heating and current drive scenarios for non-inductive start-up and ramp-up in high field ST-40 spherical tokamak

Non-inductive start-up and ramp-up is an important topic for spherical tokamak reactor design as the central solenoid implementation is highly restrictive particularly for the low-aspect-ratio tokamak configuration. In the high field spherical tokamak (ST), ST-40 with B T0 ⩽ 3 T, a preparation is underway for high power ECH and ECCD current start-up/ramp-up experiments utilizing two MW-class 140/105 GHz gyrotrons. Here, we explored various ECH/ECCD scenarios for a low-field-side (LFS) launch-angle steerable waveguide launcher placed near the mid-plane region. Due to the large toroidal field variation of ST configuration, multiple cyclotron harmonic resonance layers could exist within the plasma. In this start-up and ramp-up regime, both fundamental and second harmonic ECH resonances must be considered. We find that even with the presence of X-II resonance layer in the plasma, an efficient X-I ECH and ECCD regime can be accessed for the low electron temperature T e0 as low as 200 eV which is a typical starting temperature of ECH heated plasmas in an open-field-line configuration. The presence of X-II resonance could become significant at higher T e0 as X-II absorption increases with T e0 which could reduce the current ramp-up efficiency as the power reaching X-I is reduced. Finally for the pure X-I regime where the 2Ωe resonance is moved outside the plasma with B T0 ∼ 3.4 T, we find that it is possible to reach the full current of I p ∼ 1 MA fully non-inductively with the ECH power of ∼1 MW at n e0 ∼ 1.0 × 1019 m−3 using 105 GHz frequency gyrotron. By reducing the outer limiter position R L ∼ 78 cm to 70 cm, the pure X-I regime is recovered at the rated ST-40 magnetic field of B T0 ∼ 3.0 T. This X-I regime is accessible with a relatively broad range of launched n II or the launching angles. A survey of X-mode X-II ECH and ECCD at higher density regimes is also shown for completeness.

Recent progress on spherical torus research

The spherical torus or spherical tokamak (ST) is a member of the tokamak family with its aspect ratio (A = R0/a) reduced to A ∼ 1.5, well below the normal tokamak operating range of A ≥ 2.5. As the aspect ratio is reduced, the ideal tokamak beta β (radio of plasma to magnetic pressure) stability limit increases rapidly, approximately as β ∼ 1/A. The plasma current it can sustain for a given edge safety factor q-95 also increases rapidly. Because of the above, as well as the natural elongation κ, which makes its plasma shape appear spherical, the ST configuration can yield exceptionally high tokamak performance in a compact geometry. Due to its compactness and high performance, the ST configuration has various near term applications, including a compact fusion neutron source with low tritium consumption, in addition to its longer term goal of an attractive fusion energy power source. Since the start of the two mega-ampere class ST facilities in 2000, the National Spherical Torus Experiment in the United States and Mega Ampere Spherical Tokamak in UK, active ST research has been conducted worldwide. More than 16 ST research facilities operating during this period have achieved remarkable advances in all fusion science areas, involving fundamental fusion energy science as well as innovation. These results suggest exciting future prospects for ST research both near term and longer term. The present paper reviews the scientific progress made by the worldwide ST research community during this new mega-ampere-ST era.

Non-inductive plasma initiation and plasma current ramp-up on the TST-2 spherical tokamak

Plasma current (Ip) start-up in a spherical tokamak (ST) by waves in the lower-hybrid (LH) frequency range was investigated on TST-2. A low current (∼1 kA) ST configuration can be formed by waves over a broad frequency range (21 MHz–8.2 GHz in TST-2), but further Ip ramp-up (to ∼10 kA) is most efficient with waves in the LH frequency range. Ip ramp-up to 15 kA was achieved with 60 kW of net RF power PRF in the fast wave (FW) polarization at 200 MHz excited by the inductively coupled combline antenna. X-ray measurements showed that the photon flux and temperature are higher in the direction opposite to Ip, consistent with acceleration of electrons by a uni-directional RF wave. There is evidence that the LH wave is excited nonlinearly by the FW, based on the frequency spectra measured by magnetic probes. Similar efficiencies of Ip ramp-up were obtained with the inductive combline antenna and the dielectric-loaded waveguide array (‘grill’) antenna, and tendencies for the current drive efficiency to increase with plasma current and toroidal field were observed. During operation of the grill antenna, wavevector components were measured by an array of magnetic probes. Results were qualitatively consistent with expectations based on dispersion relations for the FW and the LH wave. A capacitively coupled combline antenna has been developed to improve coupling to the plasma and the wavenumber spectrum of the excited LH wave, and will be tested in 2013.

Comparison of rotamak plasma discharges in cylindrical and spherical devices

A new cylindrical chamber rotamak with one magnetic coil added in the midplane was constructed, and a series of experiments were performed at this device for comparison with the experiments previously conducted at a spherical chamber device. The experiments have been conducted in two regimes: the field-reversed configuration (FRC) and the spherical tokamak (ST) configuration when a steady toroidal magnetic field is added. The results from the new device further confirm that within a certain range of the equilibrium magnetic field Bv, the plasma current driven by a rotating magnetic field (RMF) grows linearly with Bv in both FRC and ST rotamak discharges. The ST configuration is more stable and allows one to achieve higher plasma current Ip with a proper choice of the toroidal field (TF); plasma current is found to reach its maximum at an optimum value of the applied TF. The measured plasma current, electron density and temperature in the cylindrical device are lower than those measured in the spherical device. The quadrupole structure of the self-generated toroidal magnetic field in the FRC regime is observed in both devices but with different magnitudes. In the cylindrical device, when a current is applied in the middle coil, plasma current can be enhanced up to 250% depending on the initial conditions.

Comparison of rotamak plasmas in FRC and ST configurations

The results of experiments in a newly rebuilt rotamak are presented. A comparison is made for two types of operations: the common field-reversed configuration (FRC) produced by driving plasma current with a rotating magnetic field (RMF), and the spherical tokamak (ST) configuration when a steady toroidal magnetic field is added. In both cases the driven plasma current develops two current rings, but in the ST configuration the inner ring current density is about three times larger than that in the FRC case. The addition of the toroidal field enables the ST to overcome the current limit for a given radio-frequency power. The total plasma current is found to have a peak at an optimum value of the applied toroidal field; the optimum toroidal field depends on the value of the equilibrium magnetic field. With proper choice of the toroidal field magnitude, the plasma current can be enhanced two- or three-fold in comparison with the limits in the cases of too a small or too large toroidal field. In the rotamak–ST, the temperature of electrons is increased by at least 50%. The density profile is triangular-shaped in the ST case and almost uniform in the FRC. The measurements of the oscillating fields indicate that the penetration of RMF into plasma is greatly enhanced in the ST case. The analysis of RMF profiles in the plasma supports the hypothesis that the improved penetration is due to the excitation of a whistler wave mode.

The basics of spherical tokamaks and progress in European research

When the aspect ratio of a tokamak (A = R/a) decreases significantly, there is a transformation of the well studied tokamak toroidal magnetic configuration into the spherical tokamak (ST) configuration. This configuration has high natural plasma elongation and triangularity and other unique equilibrium and stability properties of ST configuration, which are discussed in this paper.European research into ST physics is well advanced in spite of the young age of this branch of fusion science. An overview of selected experimental and theoretical results obtained at Ioffe, Culham and Frascati is given with the emphasis on their complementarity and links to the main stream of tokamak research, such as ITER. An outline of the basic ST advantages and the potential of ST research for new insights into magnetic confinement is also given. More detailed descriptions of recent advances in ST theory and experiment may be found in the invited papers by Akers and Ono in the proceedings of this conference.

The spherical tokamak programme at Culham

The spherical tokamak (ST) is the low aspect ratio limit of the conventional tokamak and appears to offer attractive physics properties in a simpler device. The START (Small Tight Aspect Ratio Tokamak) experiment provided the world's first demonstration of the properties of hot plasmas in an ST configuration and was operational at Culham from January 1991 to March 1998, obtaining plasma currents of up to 300 kA and pulse durations of ∼50 ms. Its successor, MAST, is nearing completion and is a purpose built, high vacuum machine designed to have a tenfold increase in plasma volume with plasma currents of up to 2 MA. Current drive and heating will be by a combination of induction-compression as on START, a high performance central solenoid, with 1.5 MW ECRH and 5 MW of NBI. The promising results from START are reviewed, and the many challenges posed for the next generation of purpose built STs (such as MAST) are described.