Lower hybrid (LH) power at 3.7 GHz was injected into the ADITYA-U tokamak marking the first implementation of the LH current drive on this device using a newly installed passive-active multi-junction launcher. Initial experiments achieved 80–100 kW of LH power coupling, leading to the formation of a high-energy tail in the plasma's electron energy distribution that shows a clear signature of suprathermal electrons driven by the LH waves. Bremsstrahlung emission is identified as a primary diagnostic tool for measuring these non-thermal electrons. Integrated ray tracing and Fokker–Planck simulations using the GENRAY and CQL3D has been applied to interpret the LH current drive experiments on ADITYA-U for the first time. The parametric variation of density, temperature, effective ion charge, and the coefficient of radial diffusion has been performed numerically to study the effect on the energy spectrum of bremsstrahlung emission due to the LH-driven suprathermal or fast electrons. These simulations incorporate experimentally measured parameters to model the LH wave propagation (in GENRAY) and wave–particle interactions (in CQL3D). The Fokker–Planck solver quantifies the momentum transfer from LH waves to electrons, resolving the equilibrium electron distribution function shaped by quasi-linear diffusion and collisional effects. A synthetic diagnostic module calculates the resulting bremsstrahlung emission and the pulse-height hard x-ray (HXR) spectrum as measured by a single-channel detection system. A detailed comparison between the simulated HXR spectrum and measurements from a cadmium telluride detector, shows reasonable agreement. These results provide a validated framework for interpreting the LH-driven fast electron dynamics in ADITYA-U and support future experimental and modeling studies.

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.

Microwave interferometer is one of the indispensable diagnostics tools for measuring electron density in a Tokamak. The interferometer system requires a stable microwave frequency generator to probe the plasma. Short term, random fluctuations in the phase of the microwave source like Voltage Controlled Oscillator (VCO) can significantly affect the frequency stability of the interferometer. These phase noise can degrade the performance of the interferometer system and cause errors in electron density measurements. Operating the VCO in a phase locked loop (PLL) configuration can significantly improve the frequency stability of the interferometer which can reduce the measurement errors of plasma density. The implementation of PLL configuration in interferometer generates a stable intermediate frequency ($I F$), to ensure that the changes in the $I F$ frequency during the plasma discharges are only due to the plasma density variations.

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

A first-of-its-kind, inductively driven micro-particle (Pellet) accelerator and injector have been developed and operated successfully in ADITYA-U circular plasma operations, which may ably address the critical need for a suitable disruption control mechanism in ITER and future tokamak. The device combines the principles of electromagnetic induction, pulse power technology, impact, and fracture dynamics. It is designed to operate in a variety of environments, including atmospheric pressure and ultra-high vacuum. It can also accommodate a wide range of pellet quantities, sizes, and materials and can adjust the pellets' velocities over a coarse and fine range. The device has a modular design such that the maximum velocity can be increased by increasing the number of modules. A cluster of lithium titanate/carbonate (Li2TiO3/Li2CO3) impurity particles with variable particle sizes, weighing similar to 50-200 mg are injected with velocities of the order of similar to 200 m s-1 during the current plateau in ADITYA-U tokamak. This leads to a complete collapse of the plasma current within similar to 5-6 ms of triggering the injector. The current quench time is dependent on the amount of impurity injected as well as the compound, with Li2TiO3 injection causing a faster current quench than Li2CO3 injection, as more power is radiated in the case of Li2TiO3. The increase in radiation due to the macro-particle injection starts in the plasma core, while the soft x-ray emission indicates that the entire plasma core collapses at once.

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.

This paper describes the experimental analysis and preliminary investigation of the predictability of pitch angle scattering(PAS) events through the electron cyclotron emission(ECE)radiometer signals at the ADITYA-Upgrade(ADITYA-U) tokamak. For low-density discharges at ADITYA-U, a sudden abnormal rise is observed in the ECE signature while other plasma parameters are unchanged. Investigations are done to understand this abrupt rise that is expected to occur due to PAS. The rise time is as fast as 100 μs with a single step and/or multiple step rise in ECE radiometer measurements. This event is known to limit the on-axis energy of runaway electrons. Being a repetitive event, the conditions of its repetitive occurrence can be investigated, thereby exploring the possibility of it being triggered and surveyed as an alternate runaway electron mitigation plan. Functional parameterization of such events with other discharge parameters is obtained and the possibility to trigger these events is discussed.PREDICT code is used to investigate the possible interpretations for the PAS occurrence through modeling and supporting the ECE observations. The trigger values so obtained experimentally are set as input criteria for PAS occurrence. Preliminary modeling investigations provide reliable consistency with the findings.

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.

In the ADITYA tokamak, the plasma discharge is initiated through filament pre-ionization assisted breakdown, using the conventional inductively driven electric field. Following the breakdown of the neutral gas, the discharge is sustained by a successful burnthrough phase. The nature of the breakdown and burnthrough phase is studied by varying the parameters influencing it, such as the toroidal electric field, operating pressure, and lithium wall conditioning. The plasma initiation failures in the breakdown and burnthrough phase are identified, and optimized conditions of operating parameters are derived. The value of the Lloyd parameter Eϕ×BT/Bz in the ADITYA for plasma breakdown is found to be in the range of 1200–2100 V/m, which is consistent with other conventional tokamaks. The applied Ohmic input power must overcome the power losses due to fuel ionization and fuel and impurity radiation to achieve the complete burnthrough. The power requirement for the burnthrough phase is obtained experimentally and compared with the estimated values. The required Ohmic input power is found to be ∼60 kW for the successful burnthrough. Furthermore, it has been observed that the vessel wall coating with lithium reduces the impurities influx in the burnthrough phase and, thus, reduces the Ohmic input power consumption.

We present experimental results of the trace argon impurity puffing in the ohmic plasmas of Aditya-U tokamak performed to study the argon transport behaviour. Argon line emissions in visible and Vacuum Ultra Violet (VUV) spectral ranges arising from the plasma edge and core respectively are measured simultaneously. During the experiments, space resolved brightness profile of Ar1+ line emissions at 472.69 nm (3p44s 2P3/2–3p44p 2D3/2), 473.59 nm (3p44s 4P5/2–3p44p 4P3/2), 476.49 nm (3p44s 2P1/2–3p44p 2P3/2), 480.60 nm (3p44s 4P5/2–3p44p 4P5/2) are recorded using a high resolution visible spectrometer. Also, a VUV spectrometer has been used to simultaneously observe Ar13+ line emission at 18.79 nm (2s22p 2P3/2–2s2p2 2P3/2) and Ar14+ line emission at 22.11 nm (2s2 1S0–2s2p 1P1). The diffusivity and convective velocity of Ar are obtained by comparing the measured radial emissivity profile of Ar1+ emission and the line intensity ratio of Ar13+ and Ar14+ ions, with those simulated using the impurity transport code, STRAHL. Argon diffusivities ~ 12 m2/s and ~ 0.3 m2/s have been observed in the edge (ρ > 0.85) and core region of the Aditya-U, respectively. The diffusivity values both in the edge and core region are found to be higher than the neo-classical values suggesting that the argon impurity transport is mainly anomalous in the Aditya-U tokamak. Also, an inward pinch of ~ 10 m/s mainly driven by Ware pinch is required to match the measured and simulated data. The measured peaked profile of Ar density suggests impurity accumulation in these discharges.

To reduce the risk of severe damage to the vessel and inner peripherals of any tokamak and its safe operation, a robust technique for the mitigation of runaway electrons (REs) is required. The REs in ADITYA tokamak are effectively mitigated by an application of local vertical magnetic field (LVF) perturbation. The LVF perturbation is applied using a pair of electromagnetic coils placed at the top and bottom of the ADITYA vacuum vessel in a Helmholtz configuration at one toroidal location. Powered by a capacitor bank power supply, these coils can produce a localized vertical magnetic field at the plasma center in the range of ∼150 G–260 G for a variable duration of 5–20 ms. The LVF pulse is first applied at the breakdown/current-ramp phase, where the REs are generated in the discharges initiated by the conventional ohmic breakdown in ADITYA. With the application of LVF pulse the REs are significantly reduced as indicated by the reduction in the REs generated hard x-ray flux. It has been observed that to extract the REs efficiently, an LVF pulse of magnitude at least ∼1% of the toroidal magnetic field with a minimum duration of ∼5 ms should be applied. The LVF perturbation is applied at different times into the discharge, i.e. during the breakdown/current ramp-up phase and current flat-top phase. The REs are significantly reduced in all the phases and improved discharge consistency. The LVF acts as an error field and a short-pulse of the LVF influences the REs more in comparison to the thermal electrons due to the faster velocities of the REs.

The adverse effect associated with runaway electrons (RE) requires the temporal monitoring of the Hard Xray (HX) spectrum produced by RE. This enables us to know the photon flux corresponding to a particular energy of HX in temporal space. A Lanthanum Bromide (LaBr3)-based HX spectrometer system (80 keV similar to 5 MeV) is routinely operated on the ADITYA-U tokamak for monitoring the temporal evolution of the HX spectrum. The temporal evolutions of the HX energy having maximum count and the average RE temperature (RE average energy) have been analyzed for the plasmas injected with neon (Ne) impurity. It has been found that peak energy and average runaway energy reduces significantly after the Ne gas puff and this reduction happens when the electron density rises after the Ne gas puff. The RE temperature values were similar to 620 KeV and 230 KeV before and after the Ne injection, respectively. The spectral shape, in both counts and energy, shrunk drastically, suggesting the reduction of the HX emission after the Ne gas puffing.

In this paper, we are report design, development, and characterization of indigenously developed 100 GHz Heterodyne interferometer system which provides real time plasma density measurement. The developed system consists of two free running Gunn oscillators which have been used as radio frequency source (RF) and local oscillator (LO) and produce stabilised intermediate frequency (IF) of 850 MHz using biased mixers for the down conversion at low power of LO.The IF signal is then coupled to an IQ demodulator for the generation of IQ signals. IQ for in phase and quadrature phase are known as IQ. Using highly stabilised signal conditioning electronics and DAQ system, the IQ signal is then processed by the Arctan function in MATLAB and LABVIEW for the phase measurements. Also these signals are coupled to FPGA based system to generate real-time signal. The developed system has the sensitivity or the noise floor - 69 dBm and dynamic range - 65dBm to -10dBm. The IQ phase imbalance is +/-15 degree(0.26 radian)

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.

The Ponto-Caspian (Black and Caspian seas) brackish-water fauna represents a special case of the endemic diversification in world's ancient lakes; it also involves a hotspot of continental diversity in the predominantly marine mysid crustaceans. We explored the origins and history of the mysid diversification in a phylogenetic analysis of some 20 endemic Ponto-Caspian species mainly of the genus Paramysis and their marine congeners, using sequences of two nuclear protein-coding genes, two nuclear rRNA genes, the mitochondrial COI gene and morphological data. A nearly completely resolved phylogeny was recovered, with no indication of rapid diversification bursts. Deep divergences were found among the main endemic clades, attesting to a long independent faunal history in the continental Paratethys waters. The current marine Paramysis species make a monophyletic cluster secondarily derived from the continental Paratethyan (Ponto-Caspian) Paramysis ancestors. The good phylogenetic resolution was mainly due to the two nuclear protein-coding genes, opsin and EPRS, here for the first time applied to peracarid systematics. In contrast, 'conventional' mtDNA and nuclear rRNA genes provided poor topological resolution and weak congruence of divergence rates. The two nuclear protein-coding genes had more congruent rates of evolution, and were about 10–15 times slower than the mitochondrial COI gene.

Nascent observations about the influence of kinetic instabilities on electron cyclotron emission (ECE) from runaway electrons (REs) have been detected and explored at the Aditya-Upgrade (Aditya-U) tokamak. The developed broadband ECE radiometer system offers wideband measurements by integrating several radio frequency units with a fixed intermediate frequency receiver with multiple channels, which is a novel approach to meet the needs of the localized measurements at various toroidal fields and extend the system dynamic range. The low density (ne ≤ 1 × 10-19 m-3) plasma discharges at Aditya-U are consistently accompanied by a 20%-40% increase in the ECE radiometer signal amplitude within 100 µs and sporadic step-like modulations. The Pitch Angle Scattering (PAS) of REs induced by kinetic instabilities is a potential candidate for their occurrence. This steep jump in the radiometer signals was detected due to its high temporal resolution of 10 µs. A "PREDICT" code that employs the relativistic test particle model validates these experimental findings of the radiometer diagnostic for the first time for Aditya-U tokamak. Preliminary observations of the ECE radiometer signals also show that additional gas puffs can be used to vary the trigger timings of such PAS events or even lead to their complete avoidance.

Heterodyne interferometry is an active diagnostic techniques for electron density measurement. Measured line integrated electron density is related directly to phase difference between transmitted and received microwave sig-nals. Phase variation measurement requires different phase detection techniques. Analog and digital phase measurement techniques are available for real time density estimation during plasma discharge. However, the analog circuit of the phase measurement is prone to noise handling and phase jump. This can be overcome by digital phase measurement techniques. Digital phase measurement having controller and high speed FPGA have become very popular for electron density measurement. FPGA based digital phase measurement techniques (zero cross, Fast Fourier Transform, CORDIC, ArcTAN and cross -correlation) are available for real time density measurement in tokamak. The zero cross and CORDIC plus zero cross algorithm have been developed on FPGA for phase estimation between two electromagnetic waves. The FPGA algorithm has been simulated with Vivado-15 and the density compared with density measured by 140 GHz heterodyne interferometer for functional validation.

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.