Prevention of ECCD triggering explosive bursts in reversed magnetic shear tokamak plasmas for disruption avoidance

The explosive burst excited by neoclassical tearing mode (NTM) is one of the possible candidates of disruptive terminations in reversed magnetic shear (RMS) tokamak plasmas. For the purpose of disruption avoidance, numerical investigations have been implemented on the prevention of explosive burst triggered by the ill-advised application of electron cyclotron current drive (ECCD) in RMS configuration. Under the situation of controlling NTMs by ECCD in RMS tokamak plasmas, a threshold in EC driven current has been found. Below the threshold, not only are the NTM islands not effectively suppressed, but also a deleterious explosive burst could be triggered, which might contribute to the major disruption of tokamak plasmas. In order to prevent this ECCD triggering explosive burst, three control strategies have been attempted in this work and two of them have been recognized to be effective. One is to apply differential poloidal plasma rotation in the proximity of outer rational surface during the ECCD control process; The other is to apply two ECCDs to control NTM islands on both rational surfaces at the same time. In the former strategy, the threshold is diminished due to the modification of classical TM index. In the latter strategy, the prevention is accomplished as a consequence of the reduction of the coupling strength between the two rational surfaces via the stabilization of inner islands. Moreover, the physical mechanism behind the excitation of the explosive burst and the control processes by different control strategies have all been discussed in detail.

EAST steady-state long pulse H-mode with core-edge integration for CFETR

Recent EAST experiment has successfully demonstrated long pulse steady-state high plasma performance scenario and core-edge integration since the last IAEA in 2018. A discharge with a duration over 60s with βP ~2.0, βN ~1.6, H98y2~1.3 and internal transport barrier on electron temperature channel is obtained with multi-RF power heating and current drive. A higher βN (βN~1.8, βp~2.0, H98y2~1.3, ne/nGW~0.75) with a duration of 20s is achieved by using the modulated neutral beam and multi-RF power, where several normalized parameters are close or even higher than the phase III 1GW scenario of CFETR steady-state. High-Z impurity accumulation in the plasma core is well controlled in a low level by using the on-axis ECH. Modelling shows that the strong diffusion of TEM turbulence in the central region prevents tungsten impurity to accumulate. More recently, EAST has demonstrated compatible core-edge integration discharges in the high βp scenario: high confinement H98y2>1.2 with high βP~2.5/βN~2.0 and fbs~50% is sustained with reduced divertor heat flux at high density ne/nGW~0.7 and moderate q95~6.7. By combining active impurity seeding through radiative divertor feedback control and strike point splitting induced by resonant perturbation coil, the peak heat flux is reduced by 20-30% on the ITER-like tungsten divertor, here a mixture of 50% neon and 50% D2 is applied.

Simulating the impact of charge exchange on beam ions in MAST-U

A model for simulating charge exchange (CX) of fast ions with background atoms in magnetically confined fusion plasmas has been implemented in the ASCOT orbit-following code. The model was verified by comparing simulated reaction mean free paths to analytical values across a range of fusion-relevant parameters. ASCOT was used to simulate beam ions slowing down in the presence of CX reactions in a MAST-U target scenario. ASCOT predicts the CX-induced loss of beam power to be 22%, which agrees to within 15% with the TRANSP prediction. Because of CX, plasma heating and current drive by beam ions are strongly reduced towards the edge. However, an overall lower but noticeable increase of up to 20% in current drive is predicted closer to the core. The simulated deposition of fast CX atoms on the wall is concentrated around the outer midplane, with estimated peak power loads of 70–80 kWm-2 on the central poloidal field coils (P5) and the vacuum vessel wall between them. This analysis demonstrates that ASCOT can be used to simulate fast ions in fusion plasmas where CX reactions play a significant role, e.g., in spherical tokamaks and stellarators.

Development of high-power, long-pulse, multi-frequency ECH/CD system for JT-60SA

A gyrotron and a matching optics unit (MOU) for the multi-frequency electron cyclotron heating and current drive (ECH/CD) system in JT-60SA have been developed successfully. The gyrotron demonstrated stable operation at high powers of 1.5 MW for 5 s and 1.9 MW for 1 s at 110 GHz. To obtain high HE11 mode purity (> 90%) at the outlet of the waveguide in the ECH/CD launcher, a MOU for operating at three frequencies of 82 GHz, 110 GHz, and 138 GHz that includes three pairs of water-cooled phase correcting mirrors has been developed, which allows the mirrors to be changed without opening the evacuated MOU. The mode purity at the inlet of a dummy load in the transmission line was evaluated at high-power and an HE11 mode purity of > 90% was obtained at three frequencies.

Current drive experiments in SST1 tokamak with lower hybrid waves

The steadystate superconducting tokamak (SST1) is aimed to demonstrate long pulse plasma discharges employing non-inductive current drive by means of lower hybrid current drive (LHCD) system. The major and minor radius of the machine is 1.1m and 0.2m respectively. The LHCD system for SST1 comprises of klystrons, each rated for 0.5MW-CW rf power at a frequency of 3.7 GHz. The grill antenna comprises of two rows, each row accommodating 32 waveguide elements. Electron cyclotron resonance (ECR) breakdown assisted Ohmic plasma is formed in SST1 to overcome the issues associated with low loop voltage start-ups. With recent modifications in the poloidal coils configuration, even with narrow EC pulse (~50ms), good repeatable and consistent Ohmic plasmas could be produced which helped in carrying out LHCD current drive experiments on SST1. These experiments demonstrated both fully as well as partially driven non-inductive plasma current in SST1 tokamak. Discharges with zero loop voltages were obtained. The interaction of lower hybrid waves with plasma and generation of suprathermal electrons could be established using energy spectra measured by CdTe detectors. Various other signatures like drop in loop voltages, negative loop voltages, spikes in hard x-rays and increase in 2nd harmonic ECE signal, further confirmed the current drive by LHW’s. The beneficial effect of LHW’s in suppressing hard x-rays was also demonstrated in these experiments. The non-inductive current drive in SST1 could also be established by modulating LH power. The longest discharge of ~650ms could be obtained in SST1 with the help of LHW’s. In this paper, the experimental results obtained with LHCD experiments on SST1 is reported and discussed in more details.

A Validation Roadmap of Multi-Physics Simulators of the Resonator of MW-Class CW Gyrotrons for Fusion Applications

For a few years the multi-physics modelling of the resonance cavity (resonator) of MW-class continuous-wave gyrotrons, to be employed for electron cyclotron heating and current drive in magnetic confinement fusion machines, has gained increasing interest. The rising target power of the gyrotrons, which drives progressively higher Ohmic losses to be removed from the resonator, together with the need for limiting the resonator deformation as much as possible, has put more emphasis on the thermal-hydraulic and thermo-mechanic modeling of the cavity. To cope with that, a multi-physics simulator has been developed in recent years in a shared effort between several European institutions (the Karlsruher Institut für Technologie and Politecnico di Torino, supported by Fusion for Energy). In this paper the current status of the tool calibration and validation is addressed, aiming at highlighting where any direct or indirect comparisons with experimental data are missing and suggesting a possible roadmap to fill that gap, taking advantage of forthcoming tests in Europe.

Suppression of toroidal Alfvén eigenmodes by the electron cyclotron current drive in KSTAR plasmas

Advanced operation scenarios such as high poloidal beta (βP) or high q min are promising concepts to achieve the steady-state high-performance fusion plasmas. However, those scenarios are prone to substantial Alfvénic activity, causing fast-ion transport and losses. Recent experiments with the advanced operation scenario on KSTAR tokamak have shown that the electron cyclotron current drive (ECCD) is able to mitigate and suppress the beam-ion driven toroidal Alfvén eigenmodes (TAEs) for over several tens of global energy confinement time. Co-current directional intermediate off-axis ECCD lowers the central safety factor slightly and tilts the central q-profile shape so that the continuum damping in the core region increases. Besides, the rise of central plasma pressure and increased thermal-ion Landau damping contribute to TAE stabilization. While the TAEs are suppressed, neutron emission rate and total stored energy increase by approximately 45% and 25%, respectively. Fast-ion transport estimated by TRANSP calculations approaches the classical level during the TAE suppression period. Substantial reduction in fast-ion loss and neutron deficit is also observed. Enhancement of fast-ion confinement by suppressing the TAEs leads to an increase of non-inductive current fraction and will benefit the sustainment of the long-pulse high-performance discharges.