Electron cyclotron resonance heating (ECRH) two-pulse experiments are carried out on the tokamaks SST-1 and Aditya-U using single 42-GHz gyrotron. Initially, the system was used to carry out either breakdown or heating. A new anode modulator power supply with fast rise time and fall time (1 ms) has been integrated with 42-GHz gyrotron system, which facilitate to switch the gyrotron for more than one pulse within plasma shot. The first pulse is used for the plasma breakdown at low-loop voltage and second pulse is used for plasma heating. The power in the first pulse is maintained low (less than 150 kW) for the breakdown at fundamental harmonic, while power in the second pulse is more than 200 kW for plasma heating. In both the tokamaks, SST-1 and Aditya-U, two pulse experiments have been carried out and heating effect is observed clearly in Aditya-U tokamak during second ECRH pulse. This article discusses about the two ECRH pulse experiments on both the tokamaks SST-1 and Aditya-U.

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.

Runaway Electrons (REs) generated during plasma disruptions in fusion grade tokamaks have the potential to severely damage the plasma-facing components. Designing optimal plasma discharge scenarios for RE suppression in future experiments requires interpretative modeling of current experiments. Multiple experiments have been carried out on ADITYA-U tokamak to design optimal plasma discharge scenarios for RE avoidance and suppression. In this article, we have numerically studied two representative pairs of ADITYA-U plasma discharges. In the first pair of discharges, plasma density was increased by gas puffs during the flap-top phase which is shown to suppress RE generation. In the second pair of the representative discharges, the effect of a lower ratio of peak electric field during breakdown to the pre-fill pressure is shown on RE generation. We simulate these plasma discharges using the PREDICT code to study the dynamics of Res. The results are consistent with the experimental Hard X-Ray diagnostic observations for both the pair of representative discharges. Additionally, the lowest RE-current is obtained in the discharges with a low ratio of the peak electric field during breakdown to the pre-fill pressure and high plasma density during the flap-top phase due to gas puffs. The suppression of REs is demonstrated by showing the effect of increasing plasma density on the separatrix in momentum space between thermal electrons and REs.

Since the 2018 IAEA-FEC conference, in addition to expanding the parameter horizons of the ADITYA-U machine, emphasis has been given to dedicated experiments on inductively driven particle injection (IPI) for disruption studies, runaway electron (RE) dynamics and mitigation, plasma rotation reversal, radiative-improved modes using Ne and Ar injection, modulation of magneto–hydrodynamic modes, edge turbulence using periodic gas puffs and electrode biasing (E-B). Plasma parameters close to the design parameters of circular plasmas with H2 and D2 as fuel have been realized, and the shaped plasma operation has also been initiated. Consistent plasma discharges having I P ∼ 100–210 kA, t ∼ 300–400 ms, n e ∼ 3–6 × 1019 m−3, core T e ∼ 300–500 eV were achieved with a maximum B T of ∼1.5 T. The enhanced plasma parameters are the outcome of repeated cycles of baking (135 °C), followed by extensive wall conditioning, which includes pulsed glow discharge cleaning in H, He and Ar–H mixture, and lithiumization. A higher confinement time has been observed in D2 compared to H2 plasmas. Furthermore, shaped plasmas are attempted for the first time in ADITYA-U. A first of its kind inductively driven particle injection for disruption mitigation studies has been developed and operated. The injection of solid particles into the plasma core leads to a fast current quench. Two pulses of electron cyclotron resonance wave at 42 GHz are launched in a single discharge: one pulse is used for pre-ionization and the second for heating. In a novel approach, a positively biased electrode is used to confine REs after discharge termination. E-B is also used for controlling the rotation of drift-tearing modes by changing the plasma rotation. Cold pulse propagation and signatures of detachment are observed during the injection of short gas puffs. A correlation between the plasma toroidal rotation and the total radiated power has been observed with neon gas injection-induced improved confinement modes.

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 generation and subsequent loss of runaway electrons (REs) during the operation sequence in a tokamak is a potent threat to the plasma-facing components and the interface of actively cooled parts. Control and mitigation of REs are of prime importance to the safe operation and machine health of a fusion device. A supersonic molecular beam injection (SMBI) system has been installed in the ADITYA-U tokamak to explore the effects of the high Mach number molecular beam on the REs and ways to mitigate the REs. In the majority of discharges in which SMBI has been injected, a burst in hard x-rays has been observed accompanying the SMBI pulse, indicating significant RE loss. This is followed by a long RE-mitigated phase in the discharge. The most plausible explanation of the mitigation of REs is minor disruption caused by SMBI. This in turn triggers field line stochastization and subsequent rapid RE loss. Finally, this leads to reorganization of the flux surfaces, resulting in bigger islands with the potential of trapping any surviving RE fraction.

On ADITYA-U tokamak, a spectroscopic diagnostic has been developed to measure the radial profile of visible continuum radiation for determining the plasma effective charge, Z(eff), to study the impurity transport and MHD driven instabilities. It consists of the collimating lenses, optical fibers, a multi-channel wavelength selection system, and photo multiplier tubes. The optical system allowing continuum radiation measurements around 536 nm (the wavelength selection system) consists of set of lenses, optical fibers and an interference filter with diameter of 5 cm and bandwidth of 3 nm. The spatial profile of radiation with a spatial resolution of similar to 3 cm has been recorded from eight lines of sight viewing the plasma using an UHV compatible rectangular view port placed on the bottom port of the ADITYA-U tokamak. The centrally peaked spatial profile of visible continuum radiation has been recorded from the ADITYA-U tokamak plasmas. The chord averaged Z(eff) values estimated from the brightness measured along the central chord fall within 2.5 to 4.1 for the electron densities of 1.0 - 2.2 x 10(19) m(-3).

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.

Plasma burnthrough and current ramp-up phases in an SST-1 superconducting tokamak are simulated by the OOPS code. The main purpose of this study is to optimize the operation regime in SST-1 through the input parameter scan for OOPS for both successful and failed shots from the SST-1 database. SST-1 is now equipped with carbon plasma facing components (PFCs). Hence, PFCs are expected to offer a carbon dominated impurity environment. An electrically continuous vacuum vessel and cryostat hinder sufficient loop voltage during start-up, thereby rendering electron cyclotron resonance heating (ECRH) pre-ionization an absolute requirement. OOPS simulation is optimized for the essential plasma initiation parameters like the seed electron density ne = 3 × 1017/m3 and the initial neutral density n0 = 1–4 × 1018/m3 (corresponding to a tailored pre-fill pressure) and for an error magnetic field Berr ≤ 20 G. The simulation shows that the ECRH pre-ionization threshold power required for the successful plasma start-up under low available loop voltage conditions is ≥180 kW.

The rate of plasma current-quench during tokamak plasma disruptions determines the electromagnetic forces on the in-vessel components/vacuum vessel. Also halo currents and rapid changes of poloidal field due to the plasma vertical displacement contribute to loads on vessel and in-vessel components and hence needs to be studied thoroughly to safeguard these tokamak peripherals. The plasma current quench occurrence during the spontaneous major disruption has been investigated for a set of ADITYA tokamak disrupted discharges and average plasma current quench and instantaneous current quench rates have been estimated. The fastest area-normalized plasma current (I-P) quench time is observed to be similar to 5 ms m(-2). The estimated post disruption plasma electron temperatures (PDET) are observed to be similar to 15-35 eV and proportional to area-normalized plasma current quench time. Further analysis of several disruptive discharges of ADITYA tokamak reveal that the current quench time is inversely proportional to the pre-disruptive values of edge safety factor, q(a), and the current quench properties are strongly correlated with the prevailing pre-disruptive plasma magnetohydrodynamic (MHD) activities. For larger values of pre-disruptive q(a), the larger island widths of m= 2 and m = 3 MHD modes leads to a significant overlap of these islands. Such an overlap along with the deeper locations of the islands inside the plasma column, as compared to discharges having smaller values of pre-disruptive q(a), seems to facilitate the faster current quench.

Study of hard X-ray (HX) spectrum is an important tool to understand the runaway electrons (RE) behavior present in the tokamak plasma as these energetic electrons, especially those generated during disruption, represent a major threat to tokamak plasma devices in terms of melting and damaging the plasma-facing component (PFC) and thereby reduced the lifetime of the first wall. A LaBr3(Ce) detector based spectroscopic diagnostic has been set up on the Aditya tokamak to measure the HX spectrum. This diagnostic consists of a photomultiplier tube and multi-channel analyzer along with the detector, viewing the whole plasma tangentially along a line of sight terminating on the limiter. The spectrum has been observed in the energy range of around ∼ 75 keV to 3 MeV and is continuum in nature. The spectrum is having the peaks within ∼ 85–150 keV and the estimated RE temperatures have been found to be in the range of 100–600 keV for the analyzed discharges. This temperature decreases with the increase of the thermal electron temperature of the discharges. The ratios of the estimated confinement time of runaway electron and energy confinement time of thermal particles are mostly inversely proportional to the plasma electron density.

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 v to the diffusion coefficient D profile, but with two different Fe concentrations. The ratio v/D varies from the value of approximate to - 0.22 m(-1) at the plasma normalized radius rho = 0.2 to a maximum value of approximate to - 0.35 m(-1) at rho = 0.6. 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 spatial profile of Hα spectrum is regularly measured using a high-resolution multi-track spectrometer in ADITYA tokamak to study the neutral particle behavior. The Monte Carlo neutral particle transport code DEGAS2 is used to model the experimental Hα spectral emissions. Through the modeling of the spectral line profile of Hα, it is found that the neutral hydrogen, which is produced from molecular hydrogen and molecular hydrogen ion dissociation processes contributes 56% to the total Hα emission, and the atoms which are produced from charge-exchange process have 30% contribution. Furthermore, the experimentally measured spatial profile of chord integrated brightness was modeled for the two plasma discharges having relatively high and low density to understand the neutral particle penetration. The presence of neutrals inside the core region of the ADITYA tokamak is mainly due to the charge-exchange process. Furthermore, it is observed that neutral particle penetration is lower in higher density discharge.

A washer-gun based plasma system has been developed to enable high power microwave (HPM)-plasma interaction in a system for microwave plasma experiments. The critical pre-requisites of the plasma are density, ne ∼ (1-10) × 1017 m-3, uniformity over a radial extent ≈10 cm and axial extent ≈20-30 cm, an axial density gradient of scale-length Ln ≈ wavelength of HPM, and ambient pressure low enough to maintain electron-neutral collision frequency much less than plasma frequency. The system developed deploys a ten stage pulse forming network, discharged to the washer-gun to produce pulsed (τpulse ∼ 100 μs) discharges that get ejected into an experimental chamber. The system is capable of generating ne ∼ 1018 m-3 and Te ∼ 10 eV. Temporal and spatial regimes are identified to obtain the required extents of radial and axial ne uniformity of 10 cm and 20 cm, respectively, and a steep axial gradient Ln ≈ 10 cm. Based on the desired frequency of the interacting HPM (in the range 3-5 GHz) planned for a particular experimental campaign, the density and spatial density profiles of the plasma can be tailored. The present paper presents an account of the plasma source and characterization of the plasma.

The first Indian tokamak, ADITYA, operated for over two decades with a circular poloidal limiter, has been upgraded to a tokamak named ADITYA Upgrade (ADITYA-U) to attain shaped-plasma operations with an open divertor in single and double-null configurations. Experimental research using ADITYA-U has made significant progress since the last FEC in 2016. After installation of a plasma facing component and standard tokamak diagnostics in ADITYA-U, the Phase-I plasma operations were initiated in December 2016 with a graphite toroidal belt limiter. Ohmically heated circular plasmas supported by filament pre-ionization with plasma parameters I-p similar to 80-95 kA, duration similar to 80-180ms, with a maximum toroidal field similar to 1 T and chord averaged electron density similar to 2.5 x 10(19) m(-3), have been obtained. The runaway electron (RE) generation, transport and mitigation experiments, along with magneto hydrodynamic (MHD) activities and density enhancement with H-2 gas puffing experiments were carried out in Phase-I, which was completed in March 2017. Preparation for the Phase-II operation in ADITYA-U includes calibration of magnetic diagnostics followed by commissioning of major diagnostics and installation of a baking system. After repeated cycles of baking the vacuum vessel up to similar to 135 degrees C, the Phase-II operations resumed in 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 similar to 135 kA in Phase-II, with a maximum chord averaged electron density of similar to 4 x 10(19)m(-3). Hydrogen gas breakdown has been observed in more than 2000 discharges, including Phase-I and Phase-II operations, without a single failure. Several experiments have been carried out, including the control of REs with the fuelling of supersonic molecular beam injection as well as sonic H-2 gas puffing during current flat-top, MHD mode studies using multiple periodic gas puffs, and radiative improved modes using neon gas puffs. The experimental results from Phase-I and Phase-II operations of the ADITYA-U tokamak are discussed in this paper.

In order to understand the atomic and molecular processes involved in the emission of the hydrogen Balmer alpha (H-alpha) spectral line from tokamak plasmas, the measured radial profile of H-alpha emissivity in ADITYA tokamak discharges has been modeled using the DEGAS2 neutral particle transport code. The radial profile of emissivity has been measured using a 1.0 m multi-track spectrometer and with PMT array based space-resolved visible spectroscopic diagnostics involving interference filters. The radial profile of the neutral hydrogen density has been obtained using the DEGAS2 code by reproducing the experimentally observed H-alpha emissivity profile. It has been found that the neutral density falls by 80-200 times at the plasma center compared to its maximum value at the plasma edge near the limiter (rho = 0.92). Detailed investigation of the contributions of atomic and molecular processes involved in H-alpha emission reveals the significant presence of hydrogen molecules and molecular ions within similar to 4cm of the limiter radius. The contributions from processes involving H-2(+) ions are found to be higher than those from the molecular hydrogen dissociation with H-2(+) dissociative recombination dominating over H-2(+) dissociation at the extreme edge of the plasma.

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.

Sawteeth oscillations are commonly observed in temporal profiles of temperatures in almost all tokamaks including ADITYA tokamak. In typical discharges of ADITYA, bursts of hard x-ray (HXR) due to interaction of runaway electrons (RE) with limiter, during sawteeth crash are regularly observed in initial period of plasma current flat-top. These HXR bursts are highly correlated with the sawteeth-crash in time, which suggests that sawtooth crash generates these runaway electrons, which then yield the HXR bursts due to their interaction with the limiter. The electric field induced in the toroidal direction due to change in poloidal magnetic field during the sawtooth crash is found to be higher than the critical electric field required for runaway generation. This induced toroidal electric field during each sawtooth crash generates the REs, which then travel to the limiter to give a HXR burst. Furthermore, it is observed that in the later period of the plasma current flat-top of the discharge, no HXR bursts accompany the sawtooth crash. This indicates different transport mechanisms of runaway electrons in the initial and later parts of plasma current flat-top. Further investigation revealed that the overlapping inflated magnetic islands seem to be responsible for loss of sawtooth-crash generated REs and subsequent HXR bursts in the initial phase. Whereas no overlapping of magnetic islands and presence of good magnetic surfaces in-between the islands restricts the transportation of REs leading to absence of correlated HXR bursts in the later phase of plasma current flat-top.

Study of medium and high Z impurities in high temperature tokamak plasma have become important, considering molybdenum and tungsten are being considered as the first wall materials due to its high melting points and capabilities to handle high heat load. Their presence in plasma is mainly due to neutral charge exchange and sputtering processes involving plasma fuel dilution and overall degrade plasma properties. Thus the study of behaviour of the Iron impurity is carried out in Aditya and Aditya-U tokamak. VUV spectra from impurities is regularly monitored using an absolutely calibrated VUV survey spectrometer having operation in the spectral range of 10 180 nm, which covers the important lines of partially ionized low and mid-Z impurities and emissions from higher excited states of highly ionized low-Z impurities. Absolute intensity calibration of this system has been carried out using branching ratio technique and by simulating the VUV spectra and then comparing those with experimental counts. VUV spectral lines at 28.41 nm (3p 3s S0 3s 3p P1) from Fe, 33.54 nm (2p 3s S1/2 2p 3p P3/2) and 36.08 nm (2p 3s S1/2 2p 3p P1/2) from Fe are measured during the current flat-top phase of plasma discharge. Transport analysis of Iron has been carried out by modelling ratio of these spectral line emission using one dimensional impurity transport code STRAHL. Transport coefficients are expressed in terms of ratio of convective velocity to the diffusion coefficient and it is found that direction of convective velocity is inward with peak magnitude equal to -0.4 at normalized radius equal to 0.6. Estimated diffusion coefficient used to reproduce the observed spectral line emission ratio in the edge region is around 10 m/sec, while in core it is 0.1 m/sec, is explained using ITG driven mode and neo-classical estimates.