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A seven-channel microwave interferometer (100 GHz) is designed, developed, and used to measure the radial profile of the plasma density $({\mathbf {n}}_{\mathbf {e}})$ in Aditya tokamak. The resolution of the density measurement is ${n}_{e}=2 \times {10}^{11}\,\,{\text {cm}}^{-3}$ . The spatial and time resolutions are 7 cm and $10~\mu \text{s}$ , respectively. The chord-averaged measurements from seven channels are Abel inverted to obtain the radial profile of the plasma density. Plasma density measured by the microwave interferometer is well matched with that measured by the Thomson scattering diagnostics.

In fusion devices, various techniques are employed for coating the plasma facing components (PFCs) including the vessel wall with low-Z material like lithium, boron, and silicon in order to enhance the plasma parameters and control. In ADITYA-Upgrade tokamak, different techniques of lithium wall conditioning are developed and implemented to obtain uniform and sustainable coating of Li on PFCs and the vessel wall. In this paper, two techniques used to generate Li from the source are reported. In one of the technique, a heated (fixed temperature of ∼120 °C) Li-rod is placed inside the hydrogen glow discharge cleaning (H-GDC) plasma and the sputtered Li by hydrogen (H) ions and atoms coats the wall and periphery. In the second technique, the Li is vapourized using a high-temperature Li-evaporator and released into the H-GDC plasma for uniform coating of Li on the PFCs and vessel. Significantly enhanced plasma parameters are obtained after Li coating by both techniques, with the evaporated Li performed better than the Li rod case. With the Li coating obtained with evaporated Li at 600 °C (550 mg Li) with H-GDC, the Li wall conditioning has been observed to be sustaining for in a larger number of plasma discharges in comparison to non-H-GDC assisted Li deposition. As the melting temperature of lithium hydride (LiH) is much higher (688.7 °C) than that of lithium (180.5 °C), this enhance the longer Li-coating lifetime relatively due to the formation of Li–H molecules on the vessel wall and PFCs. In ADITYA-U the carbon impurity and hydrogen recycling, due to relatively high surface area of graphite PFCs as well as their proximity to the plasma, limits the plasma performance and effective controls. Hence, H-GDC, H-GDC with Li-rod sputtering or Li evaporation, helium-GDC, argon–hydrogen mixtures-GDC in particular sequence are carried out to obtain better plasma discharges. The Li coating techniques and their effect on tokamak plasma discharges of ADITYA-U are discussed in this paper.

Tearing modes are the resistive magnetohydrodynamic (MHD) modes, which occur in all conventional tokamaks. They are known to degrade plasma confinement by creating magnetic islands which often lead to plasma disruption if their growth is not controlled. They have been extensively studied both theoretically and experimentally, as controlling them is foremost priority for every tokamak, including ITER and future large size tokamaks. In the high diamagnetic-drift frequency regime, the drift mode couple to tearing modes resulting in drift-tearing modes. The growth rate of Drift-Tearing modes decreases as its real frequency increases. In ADITYA as well as in ADITYA Upgrade tokamak, the frequency spectra of Mirnov signal show multiple frequency bands identified as drift tearing modes. Interestingly, the higher frequencies have precisely been found to be the integral multiples of the fundamental frequency. Further analysis reveals that these frequencies do not belong to different modes but harmonics of a single mode. These harmonics are also observed in the density as well as impurity emission. It has been observed that the harmonics generation is strongly correlated with the presence of high energetic electrons, known as runaway electrons, in the plasma. When the magnetic island grows in size, the runaway electrons get trap in these islands. And as the linear growth rate of tearing modes increases in presence of REs, the island size increases further due to the presence of runaway electrons, leading to harmonics generation.

Controlling and mitigating plasma disruptions are a matter of serious concern for tokamak operation since they can cause serious damage to the machine. Dedicated experiments on disruption mitigation have been carried out in ADITYA (R = 75 cm, a = 25 cm), an Ohmically heated circular limiter tokamak. A rapid growth of magneto hydrodynamic (MHD) modes is found to be the major cause of disruptions in ADITYA tokamak. Stimulated disruptions induced by hydrogen gas puffing are successfully mitigated through stabilization of these MHD modes by applying a biased-electrode placed inside the last-closed flux surface prior to the gas injection. However, as biased electrodes cannot be placed inside the plasma in bigger tokamaks, the application of ion-cyclotron waves (ICWs) prior to disruption has been successfully attempted to mitigate disruptions through stabilization of MHD modes in ADITYA tokamak. The amplitude of MHD modes (m/n = 3/1, 2/1) is significantly reduced upon the application of ICW prior to disruption, and the stimulated disruptions are successfully mitigated by preventing the growth and overlapping of these MHD modes. These experimental results demonstrate that MHD driven disruptions may be mitigated due to stabilization of m = 2 modes by ponderomotive force of the IC waves.

The effective charge, Z eff, of the plasmas of the Aditya tokamak has been analyzed to understand its behavior. It has been measured through the monitoring of the visible bremsstruhlang continuum emission around 523.4 nm from the plasma using an optical fiber, interference filter and photo multiplier tube based visible spectroscopic system. It has been found that the values of Z eff fall in the range of 1.7–4.0 and decrease with increasing plasma electron density, n e and the incremental value of Z eff is inversely proportional to n e 2. The value of Z eff reduces in the range of 1.7–2.5 in the discharges produced after the Li coating compare to the values of 2.0–3.5 range in the discharges before the Li coating in the Aditya tokamak. Details analysis on the contribution to Z eff from various impurities suggests that reduction of Z eff after Li coating is not only due to decrease of oxygen concentration, but also other impurities, such as iron, inside the plasma.

Impurity ion toroidal rotation profile has been observed in the ohmic discharges of Aditya-U tokamak from the Doppler-shifted passive charge exchange line emission of C5+ at 529 nm. . The line has been monitored using a space resolved visible spectroscopy diagnostic consisting of 1 m multi-track spectrometer coupled with a charge coupled device (CCD) detector. In typical discharges with central chord-averaged electron density 2.5 x 1019 m-3, the maximum toroidal rotation velocity at the plasma core is found to be -20 km/s in the counter-current direction. Reversal of plasma toroidal rotation is observed for high electron density discharges with central chord-averaged density of ~ 5 x 1019 m-3. During the rotation reversal, direction of toroidal rotation changes to co-current direction in the core. Interestingly, in the discharges with central chord-averaged electron density of 2.5 x 1019 m-3 where neon gas is injected at the plasma current flat-top, the toroidal rotation reversal has been observed after the gas puff.

Iron (Fe) impurity behaviour in ADITYA tokamak plasma has been studied using Fe spectral line emissions in the vacuum ultra violet (VUV) wavelength range. A VUV survey spectrometer has been utilized to record spectral lines at 28.41 nm from Fe14+ , and 33.54 nm and 36.08 nm from Fe15+ . It has been observed that the intensities of the Fe emissions decrease with an increase in plasma electron density. The observed emission and the intensity ratio of Fe14+ and Fe15+ ions from two discharges having relatively low and high plasma density are modelled using an impurity transport code. It is found that the observed data could be modelled using the same ratio of the convective velocity to the diffusion coefficient D profile, but with two different Fe concentrations. The ratio varies from the value of − 0.22 m−1 at the plasma normalized radius to a maximum value of − 0.35 m−1 at . The obtained diffusion coefficient value at the plasma core region is explained in terms of neo-classical transport, indicating that Fe impurity transport follows the same behaviour in the core plasma of the ADITYA tokamak.

The first Indian tokamak ADITYA, operated over 2 decades with circular poloidal limiter has been upgraded to a tokamak named ADITYA Upgrade (ADITYA-U) for attaining shaped-plasma operations with open divertor in single and double-null configurations. The experimental research in ADITYA-U has made significant progress, since last FEC 2016. After installation of PFC and standard tokamak diagnostics in ADITYA-U, the Phase-I plasma operations has been initiated from December 2016 with graphite toroidal belt limiter. Ohmically heated circular plasmas supported by filament pre-ionization with plasma parameters, Ip ~ 80 – 95 kA, duration ~ 80 – 180 ms with maximum toroidal field ~ 1T and chord-averaged electron density ~ 2.5 x 10^19 m^-3 has been successfully obtained. The runaway electron (RE) generation, transport and mitigation experiments along with MHD activities and density enhancement with H2 gas puffing experiments have carried out in Phase-I, which is 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 ~ 135°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. The plasma current has been raised to ~ 135 kA in phase-II with maximum chord averaged electron density of ~ 4 x 10^19 x m^-3. Hydrogen gas breakdown has been observed in more than ~ 2000 discharges including Phase-I and Phase-II operation without a single failure. Several experiments, including the control of REs with fuelling of Supersonic Molecular Beam Injection as well as sonic H2 gas puffing during current flat-top, MHD modes studies using multiple periodic gas puffs and radiative improved modes using Neon gas puff have been carried out. The experimental results from Phase-I and Phase-II operations of ADITYA-U tokamak are discussed in this paper.

Proposed work presents narrow bandpass waveguide filter in Ku band. The Filter design is based on the use of small irises structure, in the transverse planes of the standard (WR-62) rectangular waveguide. The technique of Cavity Resonator Filter coupled with small irises is used to design the filter. The center frequency is 15 GHz. It contains four small irises in this design, out of them two are rectangular and rest is circular. The simulation and optimization of the filter is carried out by using CST microwave studio software and the S-parameters of the same have been optimized for a particular bandwidth around the center frequency using this software.

Self-biased, square-law based Schottky barrier diodes will be used in homodyne microwave interferometer system (MIS) of Steady State Tokamak (SST-1) for the line average plasma density measurement. Proposed signal conditioning electronics (SCE) improves the strength of the weak time varying signal in the range of 5–6 mV by imparting it with a total gain of 1000 (100 × 10) in two steps and a bandwidth of 9.5–10 kHz. In this way, SCE improves the signal strength and also removes the unwanted spurious noise component picked up by the signal from source to its final destination of data analysis. The front end of the SCE comprises of an instrumentation amplifier AD524 having very low input noise level of 0.3 μV and very low input offset voltage of about 50 μV. The remaining portion of the SCE comprises of filter, optocoupler and differential driver for further conditioning of the signal. The final signal is then acquired and stored using PCI extension for instrumentation (PXI) based data acquisition system (DAQ) having a fast sampling rate of 100 kHz. The phase shift undergone in the conditioned signal while passing through plasma, in reference to a known microwave signal operating at a frequency of 130 GHz, is then measured. A detailed description of the designed signal conditioning electronics for the same is presented.

The 42GHz Electron Cyclotron resonance Heating (ECRH) system is used in tokamaks Aditya and SST-1 to carry out ECRH related experiments at fundamental and second harmonic. The Gyrotron capable to deliver 500kW power for 500ms duration is commissioned in SST-1 tokamak hall and ∼75m long transmission line is laid to connect both the tokamaks. The corrugated waveguide (ϕ63.5mm) based transmission line at normal atmospheric pressure is used to transmit power in HE11 mode. The special feature of this line is that it consists of two waveguide switches, which facilitates to transmit the power in either tokamak without changing any hardware. The total loss in the line is less than 1.1dB.

Plasma density control is a prime requirement of tokamak plasma. Various methods are used for fuelling the gas for tokamak plasma. Gas fuelling in tokamak with molecular beam injection is developed for efficient gas injection during the plasma discharge. The penetration depth of the injected molecular beam is more than normal gas puffing. Due to more penetration and very less divergence of beam, (1) wall loading due to fuel gas reduces causing reduction in recycling, and (2) there is an increase in the fuelling efficiency. A supersonic molecular beam injection (MBI) system for Aditya tokamak is developed in house by modifying a piezoelectric gas inlet valve and using 400 μm diameter nozzle at 2–5 kg/cm2 gas pressure. This valve is mounted directly on the Aditya vacuum vessel. Experiments are performed with this MBI system by puffing hydrogen gas of different pulse duration, different number of pulses at various time of tokamak discharge. The change in cord average electron density for different MBI gas pulse sequence is studied. In the present experiment, the plasma density could be raised by 1.5 times by MBI.

Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Zeff. Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values.

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.

Fast wave electron heating experiments are carried out on Aditya tokamak [R = 0.75 m, a = 0.25m,Bt = 0.75T,ne∼1–3E13/cc,Te∼250eV] with the help of indigenously developed 200 kW, 20–40 MHz RF heating system. Significant direct electron heating is observed by fast waves in hydrogen plasma with prompt rise in electron temperature with application of RF power and it increases linearly with RF power. A corresponding increase in plasma beta and hence increase in stored diamagnetic energy is also observed in presence of RF. We observe an improvement of energy confinement time from 2–4msec during ohmic heating phase to 3‐6msec in RF heating phase. This improvement is within the ohmic confinement regime for the present experiments. The impurity radiation and electron density do not escalate significantly with RF power. The direct electron heating by fast wave in Aditya is also predicted by ion cyclotron resonance heating code TORIC.

ADITYA (R0 = 75 cm, a = 25 cm), an ohmically heated circular limiter tokamak is regularly being operated to carry out several experiments related to controlled thermonuclear fusion research. In recent operational campaign, various experiments have been carried out to enhance the discharge performance as well as improve the plasma parameters. A comparative plasma discharges study with SiC and Graphite limiter was carried out to increase the plasma heating and reduce runaways. Excellent plasma heating has been observed in many discharges using Graphite limiter. Good repeatability of low hard X-rays, high temperature discharges was obtained. The control of plasma impurities and hydrogen recycling is very much essential for high performance discharges. The wall conditioning in ADITYA tokamak is carried out by hydrogen glow discharge cleaning (GDC), Pulse discharge cleaning and electron cyclotron resonance (ECR) discharge cleaning techniques with and without lithium wall coating. GDC assisted Lithiumization was found to be the most effective technique for substantial reduction in Ha and low Z (CIII & O-I) impurities. The partial pressure of mass number 18 (H2O) and 28 (N2/C2H4/CO) were regularly monitored before plasma discharge operation. Furthermore, experiment on optimization of pulse gas feed was helped in reducing wall loading and recycling. However, hard X-rays suppression with the application of multiple gas puff has been successfully achieved during negative converter operation, which led to the extension of plasma pulse length up to ~ 250 ms. All the supporting facts and operation aspects are reported.

In Aditya tokamak [S. B. Bhatt et al. Indian J. Pure Appl. Phys. 27, 710 (1989)], an increase in the Hα and C2+ intensity fluctuations from the edge region is observed with an increase in the magnetohydrodynamic (MHD) activity. Very small fluctuation amplitudes of Hα and C2+ intensity are observed in discharges where there is no MHD activity compared to the discharges with MHD activity. These fluctuations in the Hα and C2+, measured by optical filter—photomultiplier tube combination—are modulated by Mirnov oscillations having a dominant peak with a common frequency ∼7–10 kHz. Further investigation reveals the presence of strong coherent fluctuations in density and floating potential at same frequency as well. These observations indicate the existence of a nonelectrostatic instability, which may be based on the coupled mode of the drift mode and the Alfven mode. The coherent density fluctuations give rise to the experimentally observed coherent Hα and C2+ intensity fluctuations.

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.

We describe a simple method of determining the plasma density unambiguously with the help of a microwave interferometer. The frequency of the Klystron source is modulated with the help of a staircase voltage pulse. The height of each staircase is adjusted in such a way that the corresponding phase shift is 90°. Signals proportional to the cosine and sine of the phase shift due to plasma are generated, and the plasma density is unambiguously measured. The advantages of the method are highlighted in the text.