Abstract.A multi-device database of disruption characteristics has been developed under the auspices of the International Tokamak Physics Activity magneto-hydrodynamics topical group. The purpose of this ITPA Disruption Database (IDDB) is to find the commonalities between the disruption and disruption mitigation characteristics in a wide variety of tokamaks in order to elucidate the physics underlying tokamak disruptions and to extrapolate toward much larger devices, such as ITER and future burning plasma devices. In contrast to previous smaller disruption data collation efforts, the IDDB aims 2 to provide significant context for each shot provided, allowing exploration of a wide array of relationships between pre-disruption and disruption parameters. The IDDB presently includes contributions from nine tokamaks, including both conventional aspect ratio and spherical tokamaks. An initial parametric analysis of the available data is presented. This analysis includes current quench rates, halo current fraction and peaking, and the effectiveness of massive impurity injection. The IDDB is publicly available, with instruction for access provided herein.3

An experiment aimed at investigating the cause of fluctuation suppression by gas puffing in the edge and scrape-off layer plasma of the ADITYA tokamak is reported. The fluctuation suppression is accompanied by a flattening of the radial profiles of the floating potential, plasma density and electrostatic Reynolds stress. Both the fluctuation suppression and the simultaneous reduction in particle flux occur throughout the observed region of the plasma edge. The bolometer data also show an increase in the edge radiation from 20% to 40% of the input power during the gas puff. These results indicate a significant role of edge cooling in the fluctuation suppression mechanism and are shown to be well explained in terms of a concomitant reduction in the growth rate of the interchange instability associated with the electron temperature gradient.

Several experiments, related to controlled thermonuclear fusion research and highly relevant for large size tokamaks, including ITER, have been carried out in ADITYA, an ohmically heated circular limiter tokamak. Repeatable plasma discharges of a maximum plasma current of ~160 kA and discharge duration beyond ~250 ms with a plasma current flattop duration of ~140 ms have been obtained for the first time in ADITYA. The reproducibility of the discharge reproducibility has been improved considerably with lithium wall conditioning, and improved plasma discharges are obtained by precisely controlling the position of the plasma. In these discharges, chord-averaged electron density ~3.0–4.0  ×  1019 m−3 using multiple hydrogen gas puffs, with a temperature of the order of ~500–700 eV, have been achieved. Novel experiments related to disruption control are carried out and disruptions, induced by hydrogen gas puffing, are successfully mitigated using the biased electrode and ion cyclotron resonance pulse techniques. Runaway electrons are successfully mitigated by applying a short local vertical field (LVF) pulse. A thorough disruption database has been generated by identifying the different categories of disruption. Detailed analysis of several hundred disrupted discharges showed that the current quench time is inversely proportional to the q edge. Apart from this, for volt–sec recovery during the plasma formation phase, low loop voltage start-up and current ramp-up experiments have been carried out using electron cyclotron resonance heating (ECRH). Successful recovery of volt–sec leads to the achievement of longer plasma discharge durations. In addition, the neon gas puff assisted radiative improved confinement mode has also been achieved in ADITYA. All of the above mentioned experiments will be discussed in this paper.

First indigenously built tokamak ADITYA, operated over 2 decades with circular poloidal limiter has been upgraded to a tokamak named ADITYA Upgrade for the purpose having shape plasma operation with open divertor geometry. Experiment research in ADITYA-U has made significant progress, since last FEC 2016. After installation of PFC and standard tokamak diagnostics, the Phase-I plasma operations were conducted from December 2016 with graphite toroidal belt limiter. Purely Ohmic discharges in circular plasmas supported by Filament pre-ionization was obtained. The plasma parameters, Ip ~ 80-95 kA, duration ~ 80-180 ms with toroidal field (max.) ~ 1T and chord-averaged electron density ~ 2.5 x 10^19 m^-3 has been achieved. Being a medium sized tokamak, runaway electron (RE) generation, transport and mitigation experiments have always been one of the prime focus of ADITYA-U. MHD activities and density enhancement with H2 gas puffing studied. The Phase-I operation was completed in March 2017. The Phase-II operation preparation in ADITYA-U includes calibration of magnetic diagnostics followed by commissioning of major diagnostics and installation of baking system. After repeated cycles of baking the vacuum vessel up to ~ 130°C, the Phase-II operations resumed from February 2018 and are continuing to achieve plasma parameters close to the design parameters of circular limiter plasmas using real time plasma position control. Hydrogen gas breakdown was observed in more than ~2000 discharge including Phase-I and Phase-II operation without a single failure. Several experiments, including the primary RE control with lower E/P operation and secondary RE control with fuelling of Supersonic Molecular Beam Injection as well as sonic H2 gas puffing during current flat-top and Neon gas puffing for better plasma confinement are undergoing. The dismantling of ADITYA and reassembling of ADITYA-U along with experimental results of Phase-I and Phase-II operations from ADITYA-U will be discussed.

Disruptions, induced in Aditya tokamak by hydrogen gas puffing, are successfully mitigated through stabilization of magnetohydrodynamic (MHD) modes by applying a bias voltage to an electrode placed inside the last-closed flux surface prior to the gas injection. Above a threshold voltage sheared E r × B φ rotation of the plasma generated by the edge biasing leads to substantial reduction in the growth of MHD modes (m/n = 3/1, 2/1), which causes avoidance of disruptions through prevention of mode overlapping and subsequent ergodization of magnetic field lines.

Intense visible lines from Be-like oxygen impurity are routinely observed in the Aditya tokamak. The spatial profile of brightness of a Be-like oxygen spectral line (2p3p 3D3–2p3d 3F4) at 650.024 nm is used to investigate oxygen impurity transport in typical discharges of the Aditya tokamak. A 1.0 m multi-track spectrometer (Czerny–Turner) capable of simultaneous measurements from eight lines of sight is used to obtain the radial profile of brightness of O4+ spectral emission. The emissivity profile of O4+ spectral emission is obtained from the spatial profile of brightness using an Abel-like matrix inversion. The oxygen transport coefficients are determined by reproducing the experimentally measured emissivity profiles of O4+, using a one-dimensional empirical impurity transport code, STRAHL. Much higher values of the diffusion coefficient compared with the neo-classical values are observed in both the high magnetic field edge region and the low magnetic field edge region of typical Aditya ohmic plasmas, which seems to be due to fluctuation-induced transport. The diffusion coefficient at the limiter radius in the low-field (outboard) region is typically ∼ twice as high as that at the limiter radius in the high-field (inboard) region.

Lithiumization of the vacuum vessel wall of the Aditya tokamak using a lithium rod exposed to glow discharge cleaning plasma has been done to understand its effect on plasma performance. After the Li-coating, an increment of ∼100 eV in plasma electron temperature has been observed in most of the discharges compared to discharges without Li coating, and the shot reproducibility is considerably improved. Detailed studies of impurity behaviour and hydrogen recycling are made in the Li coated discharges by observing spectral lines of hydrogen, carbon, and oxygen in the visible region using optical fiber, an interference filter, and PMT based systems. A large reduction in O I signal (up to ∼ 40% to 50%) and a 20% to 30% decrease of Hα signal indicate significant reduction of wall recycling. Furthermore, VUV emissions from O V and Fe XV monitored by a grazing incidence monochromator also show the reduction. Lower Fe XV emission indicates the declined impurity penetration to the core plasma in the Li coated discharges. Significant increase of the particle and energy confinement times and the reduction of Z eff of the plasma certainly indicate the improved plasma parameters in the Aditya tokamak after lithium wall conditioning.

Negative spikes followed by positive ones in the loop voltage signal during the discharge are observed in the Aditya Tokamak [S. B. Bhatt et al., Indian J. Pure Appl. Phys. 27, 710 (1989)]. These spikes are always accompanied by hard x-ray bursts caused by sudden loss of runaway electrons. The observed growth of m=3 mode seemed responsible for the losses of localized beams of runaway electrons (Eγ∼1–5 MeV) from the plasma region around q=3 magnetic surface. The movement of these runaway electrons during their extraction from inside the plasma induces both positive and negative electric fields at those locations. In this paper, a one-dimensional toroidal electric field diffusion model is used to estimate the induced electric field at the plasma boundary, which matches quite well with the observed spikes in loop voltage in both magnitude as well as its temporal evolution.