The fast and accurate measurement of plasma position for the tokamak control system is determined by the application of function parametrization on the magnetic probe data. In Aditya-U tokamak, a poloidal array of 16 probes is installed around the plasma column, and the plasma position has been estimated by an algorithm based on the function parametrization technique. The validation of the method is done by placing a current loop inside the vacuum vessel at different positions. It is found that the measurement accuracy improved as the number of probes was increased from 7 to 16. Also, to enable the real-time position measurement in the future, time integration of the probe output is enabled using the integrator circuits in 7 probes, which will be extended to all the probes.

A data-driven long short-term memory (LSTM)-based deep learning model has been implemented on ADITYA-U to predict disruption events using various important diagnostic signals. The model proves effective with simulated data and has been validated on previous shots as well as unseen data using TensorFlow/keras which is redeveloped using PyTorch. PyTorch's dynamic computation enables developers to modify models on the fly, making it ideal for research and experimentation. Additionally, PyTorch offers excellent support for complex recurrent neural networks (RNNs) such as (LSTM) and gated recurrent units (GRUs), which are commonly using sequential data in time series forecasting. While TensorFlow may provide slightly better optimization for deployment. PyTorch's speed and flexibility along with its growing ecosystem (e.g., PyTorch Forecasting) often result in more efficient solutions during the model building phase, making it the preferred choice for time series forecasting tasks in research or quick prototyping. The both models are analyzed, and the results have been compared. This article describes the comparative analysis of TensorFlow and PyTorch model implementations in time series neural networks for ADITYA-U. The keras model inference is made using object code conversion runs at similar to 66 & micro;s while PyTorch model runs the same code within similar to 41 & micro;s using object code inference for each prediction steps. Additionally, this study investigates the impact of different neural network frameworks on prediction accuracy, latency requirements, and resource utilization in fusion energy applications.

A specially shielded CdTe detector based hard X-ray (HXR) monitoring system equipped with a lead collimator has been developed and installed on the Aditya-U tokamak to investigate the dynamics of fast electrons ( 20-200 keV) generated during sawtooth activity. The pre-existing HXR monitor in Aditya-U is exposed to the entire HXR bremsstrahlung emission from the plasma volume, peripheral limiters, and other structural components, which limits its ability to separately study the dynamics of lost and confined runaway electrons (REs). In contrast, the newly developed diagnostic has successfully measured the chord-averaged thin-target HXR bremsstrahlung emission encompassing the core plasma region, particularly within and around the sawtooth inversion radius. The measured HXR spectra are validated through forward modelling code that incorporates plasma parameters, confined RE characteristics, and the geometric configuration of the diagnostic system. The results confirm the capability of the developed HXR monitor to probe the fast-electron dynamics during internal plasma instabilities.

The ADITYA tokamak, a medium-sized device with a limiter configuration is upgraded to ADITYA-U tokamak with divertor configuration for shaped plasma operation. Three pairs of divertor coils and two pairs of position control coils are introduced in ADITYA-U tokamak. The strategy was focused on optimising the available space for new magnetic field coils installation without relying on the active cooling arrangements. The primary challenge was to install the coil along with the bus bar without any joints while utilizing the same conductor. The demountable TF coils in ADITYA-U tokamak simplified the feasibility for the new coils. The physical and technical boundary conditions were met using copper-based continuous transposed conductor (CTC) for the in-situ coil winding. This simple design is low cost and offers winding flexibility while ensuring accurate coil dimensions. A detailed description of the assembly procedure and solution to different engineering challenges during the fabrication of the divertor and position control coils are presented in this paper.

This work presents the design aspects of a new power supply, termed the divertor power supply (DPS), for shaped plasma experiments in the ADITYA-U tokamak. The primary objective of the shaped plasma configuration is to improve plasma performance compared to the limiter-mode operation in the ADITYA tokamak. The divertor coil current profile plays a vital role in plasma operation and in controlling vertical displacement events that may be triggered during plasma shaping in the ADITYA-U tokamak. A comprehensive consideration of system requirements, design, and testing is essential to ensure the reliable operation of the DPS. This paper presents the design of a converter-based power supply and its operational strategies in accordance with the existing plasma operation scenarios in the ADITYA-U tokamak.

The ADITYA tokamak (R₀ = 0.75 m, a = 0.25 m), originally operated with a limiter configuration, has been successfully upgraded to ADITYA-U with an open divertor configuration to enhance plasma confinement and operational flexibility. Limiter and divertor systems define the plasma boundary within the vacuum vessel, protecting in-vessel components by minimizing direct plasma-wall interactions. In both single-null and double-null divertor configurations, ADITYA-U is designed to produce circular and shaped plasmas with a triangularity (δ) of ∼0.45, elongation (κ) of ∼1.1–1.2, and plasma currents (Ip) in the range of 100–150 kA.Using the plasma equilibrium simulation code IPREQ, the optimal locations for limiter and divertor plates were determined and validated for the new vacuum vessel geometry. Plasma-facing components (PFCs) based on graphite—including toroidal, poloidal, and safety limiters, as well as divertor tiles—were installed in a staged manner to facilitate a progressive operational strategy. Limiters were installed prior to the initial operation phase to manage plasma-wall interactions during early campaigns, while divertor plates were added subsequently, following operational experience with impurity behavior during the burn-through phase.This paper details the preparatory studies, simulation-guided design process, and the in-situ installation challenges associated with retrofitting graphite limiter and divertor assemblies in the ADITYA-U tokamak. The work provides insights into the phased upgrade process of a medium-sized tokamak and highlights practical strategies for integrating advanced plasma boundary configurations in existing devices.

Robust control of plasma current ramp-up is an absolute necessity, as an efficient and uncontaminated plasma current ramp-up is essential for achieving prolonged, high-pressure tokamak plasma discharges. In conventional tokamaks with Ohmic breakdown, the plasma current ramp-up is achieved primarily with pre-fixed temporal profiles of the applied toroidal electric field and the equilibrium magnetic field (Bv). The pre-fixed temporal profiles of these fields are often insufficient to maintain a successful plasma current ramp-up, as several unquantified dynamical variables, such as the condition of the vessel wall and plasma-facing components, influence the plasma current rise. Fuel gas injection in an appropriate quantity at a suitable time during the current ramp-up is therefore used to control the plasma current rise rate, ensuring successful plasma current start-up in Aditya-U. The gas injection time and gas quantity are controlled based on real-time measurement of plasma current rise rate using a digital signal processor (DSP) controller. This special control scheme is capable of achieving the plasma current to rise nearly at the desired rate, resulting in a successful start-up and a stable plasma discharge.

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.

A Fast Visible Imaging Diagnostic (FVID) has been installed in Aditya-U tokamak to capture the visible image of poloidal cross-section of the vacuum vessel, where first interaction of pellet (micro-particle) with plasma takes place. The pellet particles are injected into the plasma using Inductively Driven Pellet Injector to investigate the assimilation process during plasma-pellet impurity interaction which induces the plasma disruptions. A complementary metal-oxide-semiconductor (CMOS) based high speed camera has been used that can capture upto150 kilo frames per second at resolution of 16x8 pixels. The system has been operated at high temporal and spatial resolution (26 kfps at 256x256 pixels) of 38 mu s/frame and 2.9 mm/pixel at tangency plane respectively. In Aditya-U tokamak, plasma experiments with inductively driven pellet injector were aimed to study the suitable disruption control mechanism for ITER like fusion machines where a projectile containing an impurity payload, weighing similar to 50-200 mg, is accelerated towards the core of the plasma with velocities of similar to 200-220 m/s using a linear coil gun accelerator, which can deposit micron-sized particles deep into the plasma core and radiate away the plasma stored energy within few milliseconds. The spatial and temporal evolution of visible radiation during plasma-pellet impurity interaction have been investigated for different payload weights of lithium-titanate (Li2TiO3) and Lithium carbonate (Li2CO3) with particle size of 50-80 micron. Dynamics of plasma column in aspect of size and its movement was studied using various image processing techniques.

This paper presents design considerations of a new power supply and a new coil system for the vertical plasma position control in ADITYA-U tokamak. The main objective is to improve overall plasma performance compared to the existing open-loop system. Feedback stabilization of the plasma vertical position in ADITYA-U tokamak will play a vital role. Vertical Displacement Events (VDE) can be triggered during shaped plasma operation and requires proper control to avoid plasma disruptions. A radial magnetic field created by exterior coils will arrest the vertical plasma position against unwanted VDEs. Therefore, procurement of a new power supply namely, active position magnet control power supply (APMC-PS) is proposed to fulfill the vertical position control requirement in the ADITYA-U tokamak. This work presents the overall requirement for an active control system to improve the vertical control capability in the mid-size machine without passive stabilisers. The proposed scheme will support a reduction in time response with relatively less apparent power for controlling the vertical plasma position.

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.

High Current Power Supplies (HCPS) fed to inductive coils are used for magnetic confinement of the hot plasma in fusion machines, such as ITER in southern France. The output current of the HCPS typically ranges from few kA up to 70kA, and employs thyristors and IGBTs as an elementary switching device. The press-pack (capsule) thyristor modules are commonly used in multi-pulse converters and has demonstrated very high reliability. 5kA/500V rectifier stack based on 3 Phase, full wave, full controlled thyristor bridge has been developed utilizing two thyristors in parallel in each arm. The DSP controller is used for implementing the control and protection functions. This paper presents the validation tests conducted on the selected high current press-pack thyristors along with test results of 5kA/500V Rectifier Stack on RL dummy load. The lessons from initial synchronous operation of Rectifier Stack (on RL load) during Aditya-U plasma pulse are also presented.

The former ADITYA, a medium-sized tokamak with a limiter configuration was upgraded to ADITYA-U tokamak with divertor configuration. Two pairs of new divertor coils, a single pair of auxiliary divertor coils and position control coils have been introduced in ADITYA-U tokamak to achieve shaped plasma operation using the existing Toroidal field, Ohmic transformer and vertical field coils. Currently, copper-based continuous transposed conductor (CTC) has been introduced for in-situ winding of the coils. Coil insulation materials are selected to withstand high current (15 kA), high voltage (5 kV) and sustain high temperature (120 degrees C) during the experiment. The primary challenge was to install the coil with the bus bar using the same conductor without any joints. The design of new divertor coils mainly includes electrical and thermal considerations within the limited space available for installation. A detailed description of the installation of coils, insulation fabrication, insulation curing process and testing of the divertor coils is presented in this paper.

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.

A free-boundary, axisymmetric magnetohydrodynamic equilibrium code, pyIPREQ, has been developed for Tokamak plasmas using the finite difference and Green's function approach. The code builds upon the foundational frameworks of the PEST and IPREQ codes, introducing several enhancements and new capabilities. Notably, pyIPREQ supports the specification of limiter boundaries and enables the computation of key physical quantities. The code has also been extended to compute equilibria constrained by a prescribed magnetic axis position, which is particularly useful when such information can be inferred from the diagnostics data. In addition, pyIPREQ includes functionality to address vertical instabilities, a requirement for accurately modeling elongated plasma configurations. Benchmarking has been carried out against published results and the original IPREQ code. Applications are demonstrated for ADITYA-U Tokamak experiments, where magnetic axis measurements are available, and predictions are also made for SST-1 and ADITYA-U Tokamaks under various operational scenarios. (c) 2025 Author(s). All article content, except where otherwise noted, is licensed under a Creative Commons Attribution (CC BY) license (https:// creativecommons.org/licenses/by/4.0/).

In this study, we investigate the generation of submillimeter radiation which lies in terahertz (THz) frequency range using two unidentical laser profiles in a hot, collisional, corrugated plasma. Utilizing a beat scheme methodology, we achieve resonance conditions that significantly enhance the THz-field amplitude. The novelty of our approach lies in leveraging the thermal velocity of electrons to mitigate collisional effects by increasing plasma nonlinearity and creating nonequilibrium state of plasma by application of external field that amplifies the THz field. Furthermore, the applied external periodic electrostatic field not only serves to enhance the THz radiation amplitude but also provides an additional tuning parameter in this scheme. Our findings reveal a 38% enhancement in efficiency when thermal velocity varying from 0% to 20% of c (speed of light). This enhancement is observed in both the beat-induced mechanisms and the coupling mechanism of individual lasers with external electrostatic field.

Pairs of external magnetic field coils placed outside the vessel maintain plasma equilibrium in tokamaks. The shape and position of the plasma column are determined and controlled by the temporal current profile in the external field coils. The temporal profile of the coil current mainly controls the horizontal movement of the plasma column, which keeps the plasma column at the vessel center. Hence, it is important to set and control the equilibrium profile according to the plasma current evolution and its profile. For this purpose, a controlled power supply is used to set the current in the coils as per the given reference profile. Currently, during plasma operation in the ADITYA-U and State Superconducting Tokamak (SST-1) tokamaks, a preprogramed profile is used to generate the coil current profile, which is a passive form of control. For modern tokamak operations, however, real-time plasma control is essential. A novel technique has been devised for controlling the equilibrium field profile in ADITYA-U and SST-1 tokamaks using the real-time plasma current. To match the output current profile with the plasma current profile, a preconditioning system called Plasma Current Conditioning (PCC) is developed. To accomplish this, a MATLAB Simulink model of a 35-year-old power supply and a mathematical model of inverse control were developed, which generates the required reference profile in real time based on the current plasma profile. Using prior experimental data, the model has been exhaustively tested and validated.

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.