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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.

Extraction of large-scale coherent structure from plasma turbulence using rake probe and wavelet analysis in a tokamak Rev. Sci. Instrum. 77, 063505 (2006); Vortex-like coherent structures are observed in the edge plasma of ohmically heated ADITYA tokamak ͓ Phys. Rev. Lett. 69, 1375 ͑1992͔͒. The structures are observed on statistical basis when the floating potential fluctuations are analyzed using conditional averaging technique. The structures, which have dipole nature, experience stretching until their radial isolation across the limiter is destroyed. The potential fluctuation also shows non-Gaussian statistics indicating intermittency in broadband turbulence of the edge plasma.

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.