The ITER Ion Cyclotron Heating and Current Drive system will deliver 20MW of radio frequency power to the plasma in quasi continuous operation during the different phases of the experimental programme. The system also has to perform conditioning of the tokamak first wall at low power between main plasma discharges. This broad range of reqiurements imposes a high flexibility and a high availabiUty. The paper highlights the physics and design reqiurements on the IC system, the main features of its subsystems, the predicted performance, and the current procurement and installation schedide.

An experiment aimed at investigating the antenna properties of different plasma structures of a plasma column as a reconfigurable plasma antenna, is reported. A 30 cm long plasma column is excited by surface wave, which acts as a plasma antenna. By changing the operating parameters, e.g., working pressure, drive frequency, input power, radius of glass tube, length of plasma column, and argon gas, single plasma antenna (plasma column) can be transformed to multiple small antenna elements (plasma blobs). It is also reported that number, length, and separation between two antenna elements can be controlled by operating parameters. Moreover, experiments are also carried out to study current profile, potential profile, conductivity profile, phase relations, radiation power patterns, etc. of the antenna elements. The effect on directivity with the number of antenna elements is also studied. Findings of the study indicate that entire structure of antenna elements can be treated as a phased array broadside vertical plasma antenna, which produces more directive radiation pattern than the single plasma antenna as well as physical properties and directivity of such antenna can be controlled by operating parameters. The study reveals the advantages of a plasma antenna over the conventional antenna in the sense that different antennas can be formed by tuning the operating parameters.

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.

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