A compact Spherical Tokamak(ST) is commissioned at Institute for Plasma Research (IPR) to explore low aspect ratio tokamak physics and technologies that complement to the existing high aspect ratio tokamaks namely ADITYA-U and SST-1 by enabling studies on non-inductive startup, current drive in over dense plasmas, and shaped plasma physics on a low cost platform. The device, India's first spherical tokamak has completed major mechanical, magnetic, and electrical integration, and the coil system has been successfully tested with series of integrated commissioning. First plasma experiments have been carried out with a modest Ohmic system assisted by a 2.45GHz microwave system, supported by a centralized control and data acquisition system. An initial diagnostic set comprising visible imaging, spectroscopy, magnetics, and radiation monitors required for machine operation has been installed. This paper presents the integrated commissioning experiences and first plasma experiments of the newly installed machine.

Experiments on ADITYA-U tokamak show a marked enhancement in the sawtooth period by application of short gas puffs of fuel that cause a modification of the radial density profile. A consequent suppression of the trapped electron modes then leads to an increase in the core electron temperature. This slows down the heat propagation following a sawtooth crash causing a delay in achieving the critical temperature gradient inside the q=1 surface required for the next sawtooth crash to happen. The overall scenario has strong similarities with the behavior of sawtooth under electron cyclotron resonance heating (ECRH). Our findings suggest an alternate technique for sawtooth control that may be usefully employed in small- and/or medium-sized tokamaks that do not have an ECRH or any other auxiliary heating facility.

Enhancement of the scrape-off layer (SOL) heat flux width has been observed in the ADITYA-U Tokamak following the injection of short fuel gas pulses. A notable reduction in parallel heat flux near the last closed flux surface (LCFS) is observed after each pulse. Comparative analysis indicates that pulsed fuelling is more effective in mitigating heat flux with improved core confinement than continuous gas feeding via real-time density control. Analytical and simulation works are also carried out for validation of experimental results. The analytical model shows that SOL width modification cannot be attributed solely to the decrease of temperature due to gas pulse injection; cross-field plasma diffusion also needs to increase. Simulations with the UEDGE code suggest that an increase in both the cross-field diffusion coefficient and inward pinch velocity is necessary to replicate the experimentally observed broadening of the heat flux SOL width. These findings provide insights into efficient SOL heat flux control strategies for future fusion devices.

Application of periodic short bursts of gas leads to periodic separation of the plasma column edge from the limiter in the Ohmically heated discharges of ADITYA-U tokamak. After every gas-puff, injecting around 1017-1018 m-3 molecules of fuel gas in the edge of the plasma column, a significant reduction in plasma density, temperature, H alpha emission intensity and an increase in the H beta/H alpha ratio is observed in the vicinity of the limiter radius for a certain time duration before these parameters regain their pre-gas puff values. A decline in the ion-flux to the limiter is also observed simultaneously. The global confinement of plasma also increases after each gas-puff. These observations, which repeat after each gas-injection, imply that the plasma-neutral interaction occurs further away from the limiters after the gas-injection and indicate a periodic plasma separation from the limiter surface. Further investigation revealed that the gas-injection modifies the radial and toroidal electric field, which may be playing a role in the separation of the plasma column from the limiter surface. It was further demonstrated by an electrode bias experiment in the absence of gas-injection that an application of an external radial electric field can result in obtaining a separated plasma state. (C) 2025 Author(s). All article content, except where otherwise noted, is licensed under a Creative Commons Attribution-NonCommercial 4.0International (CC BY-NC) license (https://creativecommons.org/licenses/by-nc/4.0/).https://doi.org/10.1063/5.0286193

This article reports the development of a versatile high-speed reciprocating drive system (HRDS) with interchangeable probe heads to characterize the edge plasma region of the ADITYA-U tokamak. The system is driven by a servo motor, which is controlled by a PXI-commanded controller. It can be operated remotely with precise control over the speed, acceleration, and distance traveled by the probe head on a shot-to-shot basis. HRDS-driven linear array of Langmuir probe measurements of floating potential, electron density, and temperature revealed that the gas puff impacts these parameters up to ∼3 to 4 cm inside the last closed flux surface (LCFS). In addition, by driving an array of magnetic probes, the system facilitated measurements of magnetic fluctuations inside the LCFS, demonstrating its versatility in supporting different diagnostic probes.

Disruptions continue to pose a significant challenge to the stable operation and future design of tokamak reactors. A comprehensive statistical investigation carried out on the ADITYA-U tokamak has led to the observation and characterization of a novel disruption regime. In contrast to the conventional Locked Mode Disruption (LMD), the newly identified disruption exhibits a distinctive two-phase evolution: an initial phase characterized by a steady rise in mode frequency with a nonlinearly saturated amplitude, followed by a sudden frequency collapse accompanied by a pronounced increase in amplitude. This behaviour signifies the onset of the precursor phase on a significantly shorter timescale. Clear empirical thresholds have been identified to distinguish this disruption type from conventional LMD events, including edge safety factor, current decay coefficient, current quench (CQ) time, and CQ rate. The newly identified disruption regime is predominantly governed by the (m/n = 2/1) drift-tearing mode (DTM), which, in contrast to typical disruptions in the ADITYA-U tokamak that involve both m/n = 2/1 and 3/1 modes, consistently manifests as the sole dominant instability. Initiated by core temperature hollowing, the growth of this mode is significantly enhanced by a synergistic interplay between a strongly localized pressure gradient and the pronounced steepening of the current density profile in the vicinity of the mode rational surface.

The realisation of tokamak and equivalent devices which operate on the principle of magnetic confinement is dependent on the successful design, development and subsequent reliable operation of the magnet coils. There is a high reliance of physics experiments on the magnetic field parameters which finally depends upon the magnet cost. Almost every time, main focus is laid on characterizing the magnet's performance and insulation properties and erosion rates of the magnets are rarely measured. The main objective of this work is to identify existing problems in coil insulations under various situations and to propose strategies to condition them. This has resulted in a better understanding of operating conditions for coil insulation in ADITYA-U tokamak. The work is based on dealing with the nature of underlying physical assumptions and the sufficiency of experimental observations. The detailed analysis of scenarios provides insight into the dominant humidity responsible for a drop in insulation resistance in ADITYA-U tokamak. A detailed understanding of the varying humidity conditions imposed on magnet insulators and their response has been presented in great detail.

In this paper, we report the excitation of coherent density and potential fluctuations induced by magnetohydrodynamic (MHD) activity in the edge plasma region of ADITYA-U tokamak. When the amplitude of the MHD mode, mainly the m/n = 2/1, increases beyond a threshold value, |B-theta|/B-theta similar to 0.3%-0.4%, coherent oscillations in the density and potential fluctuations are observed having the same frequency as that of the MHD mode. The mode numbers of these MHD induced density and potential fluctuations are obtained by Langmuir probes placed at different radial, poloidal, and toroidal locations in the edge plasma region. Detailed analyses of these Langmuir probe measurements reveal that the coherent mode in edge potential fluctuation has a mode structure of m/n = 2/1, whereas the edge density fluctuation has an m/n = 1/1 structure. It is further observed that beyond the threshold, the coupled power fraction scales almost linearly with the magnitude of B-theta/B-theta fluctuations. Furthermore, the rise rates of the coupled power fraction for coherent modes in density and potential fluctuations are also found to be dependent on the growth rate of magnetic fluctuations. The disparate mode structures of the excited modes in density and plasma potential fluctuations suggest that the underlying mechanism for their existence is the coupling of even harmonics of potential to the odd harmonics of pressure due to 1/R dependence of the toroidal magnetic field. (c) 2024 Author(s). All article content, except where otherwise noted, is licensed under a Creative Commons Attribution (CC BY) license

Effective control of impurities and precise regulation of the fueling gas are supreme prerequisites for optimal operation in any fusion device. Conventional wall-conditioning methods fall short of achieving optimal wall conditioning. Conventional wall-conditioning methods, such as vessel baking and H2/(D2)-fueled glow discharge cleaning (GDC), are generally required to remove wall-absorbed impurities in bulk after vessel venting. The excess amount of hydrogen, injected during H2 GDC, can be reduced by helium (He)-fueled GDC. However, He removal from the vessel is more challenging due to its low molecular mass, very low condensation temperature, and inert characteristics. In ADITYA-U, optimal wall conditioning cannot be achieved using H2 followed by He-fueled GDC when applied for extended periods spanning hours or days. A GDC with a mixture of argon and hydrogen (Ar-H2) is introduced in the ADITYA-U tokamak to obtain better wall conditioning than H2 followed by He GDC. In Ar-H2 GDC, long-lived ArH+ ions are formed in sufficient numbers and accelerated toward the vessel wall with high momentum. This results in the breaking of high energy bonds of impurities with the wall/plasma facing components, which is not possible by H+, H2+, H3+ ions in H2 GDC due to their lower momentum. An optimal blend ratio of Ar to H2 is established at 15%-20% for the mixture. This composition ensures that the introduction of high-Z Ar does not adversely affect tokamak plasma operations. The C- and O-containing impurities are reduced beyond the limit of the prolonged operation of H2 GDC. Relative low pressures of dominant impurities such as CO, CH4, and H2O are obtained due to the Ar-H2 GDC compared to routinely operated H2 GDC. A comparison study of H2 GDC and the developed Ar-H2 GDC is performed in terms of wall conditioning and tokamak plasma operation. The encouraging results of the Ar-H2 GDC are obtained in both wall cleaning and tokamak operation scenarios in the midsize tokamak ADITYA-U. This development and application of Ar-H2 GDC are beneficial for large-sized fusion devices, leading to improved impurity reduction, reduced operational fuel consumption (H2/D2/He), and enhanced control over fuel recycling/extraction.

Intrinsic toroidal rotation velocity ( V φ ) has been measured from the Doppler shift of C 5+ carbon spectral lines (at 529.05 nm) in the edge region of the ADITYA-U tokamak without any auxiliary torque input in an ohmically heated pure hydrogen (H 2 ) plasma as well as in H 2 plasmas seeded with medium-Z (neon and argon) impurities . The toroidal rotation in the edge region is observed to reverse its direction from the counter-current to the co-current direction with an increase in plasma current beyond I p ∼ 145–150 kA. Furthermore, a systematic decrease in the co-current V φ has been observed with the edge density, which tends to decrease to almost zero velocity with an increase in the edge density. The injection of medium- Z (neon and argon) impurities is observed to influence the edge toroidal rotation significantly. In low I p discharges, argon injection leads to a reversal of edge intrinsic rotation from the counter-current to the co-current direction. In high I p discharges, both neon and argon seeding enhance the co-current rotation by about ∼5–10 km s −1 , at a constant I p compared to pure H 2 discharges. Simultaneous measurements of the edge radial electric field, E r , shows that the E r × B θ flow seems to be driving the edge toroidal rotation in ADITYA-U. With impurity injection, the E r also gets modified, leading to an observed increase in the edge toroidal rotation.

The ADITYA upgrade (ADITYA-U), a medium-sized (R0=75 cm,a=25 cm) conventional tokamak facility in India, has been consistently producing experiments findings by using circular and shaped-plasmas. Recognizing the plasma parameters aligning closely with the design parameters of circular limited plasmas, ADITYA-U shifted its focus toward exploring the operational regime for experimentation on saw-tooth and MHD phenomena. Moreover, ADITYA-U has made consistent advancements toward conducting preliminary plasma shaping experiments through the activation of top and bottom divertor coils utilizing hydrogen as well as deuterium fuels. Confinement is improved by a factor of similar to 1.5 in D2 plasmas when compared to H 2 plasmas of ADITYA-U. Further, ADITYA-U operations emphasize preventing disruptions and runaway electrons (REs) to ensure safe operations for future fusion devices. Significant suppression of REs has been achieved in ADITYA-U with the application of pulsed localized vertical magnetic field (LVF) perturbation, thereby establishing the technique's independence from the tokamak device. The successful RE mitigation requires a critical threshold of LVF pulse magnitude, which is approximately 1% of the toroidal magnetic field, and a minimum duration of similar to 5 ms. Apart from this, several novel findings have been achieved in the ADITYA-U experiments, including the modification of sawtooth duration through gas-puff, the emergence of MHD-induced geodesic acoustic mode-like oscillations, the propagation of fast heat pulses induced by MHD activity, the control of RE dynamics through Gas-puffs, the propagation of pinch-driven cold-pulses, the transport and core accumulations of argon impurities, the mass dependency of plasma toroidal rotation and the detection of 'RICE' scaling, as well as the characterization of edge plasma using wall conditioning methods, such as glow discharge cleaning using a combination of Ar-H 2 mixture, localized wall cleaning by electron cyclotron resonant plasma, and the development of machine learning-based disruption predictions, will be discussed in this paper.

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.

Plasma equilibrium in ADITYA/ADITYA-U is provided by two pairs of vertical field coils (BV1 & BV2) placed outside the vessel. A peak loop voltage of similar to 20 V is required for successful breakdown and start-up in ADITYA, which leads to a higher I-p ramp-rate similar to 6 - 8 MA/Sec during the first similar to 7 ms of discharge. To hold the plasma column in equilibrium, the vertical field should also be ramped-up at the same rate. Series connections of vertical field (BV) coils do not provide the required ramp rate due to the high L/R time-constant of the coils and 12 pulse converter firing. Therefore, additional arrangements are made to achieve it. The addition of a precharged capacitor of 500 mu F/3 kV with VF converter based power supply allows successful start-up but causes concern about a slight dip that is observed in the plasma current. To obtain proper stabilization, two techniques are used. One is the paralleling of BV coils, and second is using the combination of another capacitor bank of 19.5 mF/1.2 kV and IGBT based power supply have improved the plasma performance and raised the I-p similar to 150 kA with dI(p)/d(t) similar to 3.0 - 3.5 MA/s in ADITYA. In this paper, the effect of the equilibrium field in accordance with plasma performance is discussed in detail. (C) 2022 The Japan Society of Plasma Science and Nuclear Fusion Research

The influence of background plasma poloidal rotation on the rotation frequency of the m/n = 2/1 drift tearing mode (DTM) has been studied in ADITYA-U tokamak. The poloidal rotation velocity of the background plasma in the ion diamagnetic direction is increased or decreased by inducing an outward or inward radial electric field, respectively, through a biased-electrode placed in the edge region of the plasma. The rotation frequency of the preexisting drift tearing mode, rotating in the electron diamagnetic direction, concomitantly decreased or increased with the application of bias depending on its polarity. The positive-bias increases the background plasma rotation in the ion-diamagnetic direction from its pre-bias value, hence decreasing the DTM rotation frequency, whereas the negative bias reduces the plasma rotation velocity in the ion-diamagnetic direction, hence increasing the mode rotation. In addition to that, a short gas puff introduced during the positive and negative bias pulse further reduces the mode frequency, however, with different amplitudes in different bias-polarities. These observations suggest that the background plasma rotation contributes significantly toward the rotation of DTMs, and the rotation frequency of the magnetohydrodynamic modes can be modified by varying the poloidal rotation of the background plasma and/or the diamagnetic drift frequency.

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.