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.

The Ohmically heated circular limiter tokamak ADITYA (R 0 =75 cm, a=25 cm) has been upgraded to a tokamak named the ADITYA Upgrade (ADITYA-U) with an open divertor configuration with divertor plates. The main goal of ADITYA-U is to carry out dedicated experiments relevant for bigger fusion machines including ITER, such as the generation and control of runaway electrons, disruption prediction, and mitigation studies, along with an improvement in confinement with shaped plasma. The ADITYA tokamak was dismantled and the assembly of ADITYA-U was completed in March 2016. Integration of subsystems like data acquisition and remote operation along with plasma production and preliminary plasma characterization of ADITYA-U plasmas are presented in this paper.

Short bursts (∼1 ms) of gas, injecting ∼1017–1018 molecules of hydrogen and/or deuterium, lead to the observation of cold pulse propagation phenomenon in hydrogen plasmas of the ADITYA-U tokamak. After every injection, a sharp increase in the chord-averaged density is observed followed by an increase in the core electron temperature. Simultaneously, the electron density and temperature decrease at the edge. All these observations are characteristics of cold pulse propagation due to the pulsed gas application. The increase in the core temperature is observed to depend on the values of both the chord-averaged plasma density at the instant of gas-injection and the amount of gas injected below a threshold value. Increasing the amount of gas-puff leads to higher increments in the core-density and the core-temperature. Interestingly, the rates of rise of density and temperature remain the same. The gas-puff also leads to a fast decrease in the radially outward electric field together with a rapid increase in the loop-voltage suggesting a reduction in the ion-orbit loss and an increase in Ware-pinch. This may explain the sharp density rise, which remains mostly independent of the toroidal magnetic field and plasma current in the experiment. Application of a subsequent gas-puff before the effect of the previous gas-pulse dies down, leads to an increase in the overall electron density and consequently the energy confinement time.