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The present study details the generalization of a stability condition for the semi-implicit formulation of the one-dimensional impurity transport equation for tokamak plasmas in terms of the magnetic flux surface coordinate system (ρ). The radial impurity transport equation for tokamak plasmas is a set of non-linear, parabolic, partial differential equations, solving which generates the radial distributions of all impurity charge states (Z) within the plasma. The present study illustrates the application of a semi-implicit method over the ρ-based impurity transport equation, generated by applying a transformation of the coordinate for the poloidal cross-section of the torus-shaped plasma confinement system, from its geometric radius (r) to the magnetic flux surface coordinate system (ρ). The study further discusses the von Neumann stability analysis of the numerical scheme applied to this transformed (ρ-based) impurity transport equation. The von Neumann stability analysis of the semi-implicit formulation of the radial impurity transport equation has been reported earlier. The stability condition derived in this study is, therefore, a generalization to the earlier reported stability condition now applicable to all ρ(r) including the specific case ρ = r considered in the earlier study. The effects of the impurity transport coefficient (D and v) profiles and the plasma and impurity parameter profiles on the derived ρ-based stability condition are analysed in this study. The impurity element considered is oxygen (1 ≤ Z ≤ 8) and the geometry and plasma parameters of the ADITYA tokamak are applied to the cases studied for consistency.

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 spatial profile of H&alpha, 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 H&alpha, spectral emissions. Through the modeling of the spectral line profile of H&alpha, it is found that the neutral hydrogen, which is produced from molecular hydrogen and molecular hydrogen ion dissociation processes contributes 56% to the total H&alpha, 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.

The impurity ion poloidal rotation and ion temperature from the Aditya-U tokamak plasma have been measured using a high-resolution spectroscopic diagnostic. It comprises of a high resolution, 1 m, f/8.7, Czerny-Turner configuration spectrometer along with charge coupled device (CCD) detector. The system monitors the spectral line emission of C2+ impurity ions at 464.74 nm from the top port of the Aditya-U vacuum vessel with the lines of sight covering the plasma minor radius from r = 11.55 cm to 21.55 cm. The impurity ion poloidal rotation velocity and temperature have been estimated using the Doppler shift and Doppler broadening of the spectral lines respectively. The maximum poloidal rotation at a radial location of 21.55 cm in the edge of the plasma during the plasma current flat top was observed to be ~4 km/s for the analyzed discharges and the ion temperatures measured in the edge were in the range of 32&ndash, 40 eV.

The spatial profile of neutral hydrogen temperatures in Aditya-U tokamak plasma has been estimated from the spatial profile of the H&alpha, spectral emissions measured using a high-resolution multi-track spectrometer, having a spectral resolution of 0.023 nm at a 50 &mu, m entrance slit width. The neutral temperature estimation from the Doppler broadened spectral line was carried out after considering the Zeeman effect due to the magnetic field present in the tokamak. To accurately obtain the temperature of the neutral hydrogen, two temperature components (warm and hot) were required to be considered. A code was developed to obtain the neutral temperature and is used to analyze two typical plasma discharges. The temperature of warm components varies between 3 and 5 eV, while hot atoms have temperatures in the range of 15&ndash, 30 eV. It was observed that the chord-integrated neutral temperature increases slightly towards the plasma core region compared to the plasma edge of Aditya-U tokamak.

A high-resolution spectroscopic diagnostic for the measurement of spatial profiles of impurity ion toroidal rotation velocities on the ADITYA-U tokamak has been upgraded to cover the complete plasma minor radius. Earlier, the coverage of diagnostics toward the plasma edge was restricted due to the placement of collection optics on the tangential port outside the vacuum vessel. The coverage of the full plasma minor radius, from 0 to 24 cm, has been achieved using the newly designed and developed collection optics that have seven lines of sight to view the tokamak plasma mounted inside a customized re-entrant view port which is installed in the shadow of the limiter inside the vacuum vessel. The upgraded diagnostic also includes a faster charged coupled device detector with a smaller pixel size for the detection of a small wavelength shift. The complete spatial profile has been measured using the Doppler shifted passive change exchange spectral line at 529.0 nm from the C5+ ion. In this article, we present the collection optics’ design, installation, calibration, and results obtained using the upgraded diagnostic.

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 ADITYA-U tokamak (R0 = 0.75 m, a = 0.25 m) is designed to shape plasma column in both single and double null diverter configurations. It is quite well known that sustaining a shaped plasma in tokamak requires very good plasma column position control, both horizontal and vertical. An FPGA-based proportional–integral–derivative (PID) control system has been designed and operated to achieve horizontal plasma position control in ADITYA-U tokamak. The complete system has been rigorously tested with sample signals before implementing to the ADITYA-U plasma discharges. The control system is integrated and time-synchronized with the plasma discharge operation of ADITYA –U. Furthermore, the system has been trained to take appropriate actions during the disruption or plasma failure in the tokamak operation. Detailed experimental results have been obtained by the operation of the digital PID controller. The complete design, installation, operation, tuning of the system along with all the relevant testing and operating experience of the digital PID controller for real-time horizontal plasma position control is presented in the paper.

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.

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

A high resolution spectroscopic diagnostic for the measurement of the spatial profile of toroidal rotation velocity (vt) and temperature (Ti) of carbon ions has been developed and implemented on the Aditya-U tokamak. The diagnostic consists of a high resolution 1 m (f/8.7) spectrometer having an 1800 grooves/mm grating coupled with a charged couple device. The setup allows measurements from 5 toroidal lines-of-sights passing through different minor radii in the horizontal mid-plane of the Aditya-U tokamak. The carbon line emissions at 529 nm in visible and 229.6 and 227.09 nm in the ultra violet spectral range are recorded using the setup. Initial results show that typical Aditya-U plasmas have a maximum carbon ion temperature (Ti) of ∼120 eV and a rotation velocity (vt) of ∼-14 km/s in the counter plasma-current direction.

A tangential soft x-ray crystal spectrometer has been designed to measure the x-ray spectrum of He-like argon for the Aditya-U tokamak plasma. The system enables to measure electron temperature using the intensity ratio of the resonance line to the satellite line. For this purpose, an x-ray spectral line at 3.9494 A from He-like argon, Ar16+, is considered. The spectrometer consists of a cylindrically bent silicon (111) crystal and a CCD detector to measure the resonance spectral line and its satellite lines in the wavelength region of 3.94-4.0 A, viewing the plasma tangentially at an angle of 26° with respect to the toroidal direction in the magnetic axis. Considering Aditya-U tokamak plasma parameters and its geometrical constraints, plasma to crystal and crystal to detector distances have been kept at 1.47 m and 0.5 m, respectively, to detect a sufficient signal. The engineering design has been optimized after adequately addressing the issues related to port geometry and machine accessibility. Details on the design of the crystal spectrometer are presented in this paper.

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

The neutral particle temperature in the edge region of the Aditya-U tokamak has been measured by recording the hydrogen Balmer alpha emission spectra at 656.28 nm from different lines of sight in both the high and low field sides. The spatial profile of the H α emission has been recorded using a 1 m multi-track spectrometer. The neutral temperature is estimated from the Doppler broadening of the measured H α spectrum by appropriately removing the contribution from the Zeeman splitting of the spectral lines. A computer simulation code was developed to estimate the neutral temperature by including the broadening mechanisms such as Doppler broadening and the Zeeman effect to simulate the H α emission spectrum along with the proper convolution of the instrumental width of the diagnostic system. It has been observed that the high field side neutral temperature (~4–6 eV) during plasma flat top is almost twice that of the low field side (~2–3 eV) suggesting poloidal asymmetry in the neutral temperature.

Emissive probes are considered as one of the best diagnostic tools for the direct measurement of plasma potential in low temperature plasma systems. Laser-heated emissive probes (LHEPs), due to their higher lifetime, are becoming more popular than their current-heated conventional counterparts. Graphite due to its higher absorption coefficient at CO2 laser wavelength and LaB6 due to its lower work function are commonly used as the probe materials for laser heated probes. Experiments with a probe-tip made of a graphite material have revealed that the emission processes other than thermionic emission also occur on heating the material with the laser, which can affect the plasma potential measurements and change the I-V characteristics of the LHEP. This additional emission due to the laser irradiation on the graphite-tip explains the measurements of lower values of plasma potentials and also describes the observation of an increase in electron saturation current with the increase in thermionic emission current due to heating. The possible mechanisms of this negatively charged-emission other than the thermionic emission from the graphite-tip have been thoroughly characterized in the absence of plasma. This extra emission is not observed in the probe-tip made up of LaB6, and a good comparison of I-V characteristics could be established with a conventional filament emissive probe in the case of LaB6.

Lithiumization of the vacuum vessel wall of the Aditya tokamak using a lithium rod exposed to glow discharge cleaning plasma has been done to understand its effect on plasma performance. After the Li-coating, an increment of ~100 eV in plasma electron temperature has been observed in most of the discharges compared to discharges without Li coating, and the shot reproducibility is considerably improved. Detailed studies of impurity behaviour and hydrogen recycling are made in the Li coated discharges by observing spectral lines of hydrogen, carbon, and oxygen in the visible region using optical fiber, an interference filter, and PMT based systems. A large reduction in O I signal (up to ~40% to 50%) and a 20% to 30% decrease of Hα signal indicate significant reduction of wall recycling. Furthermore, VUV emissions from O V and Fe XV monitored by a grazing incidence monochromator also show the reduction. Lower Fe XV emission indicates the declined impurity penetration to the core plasma in the Li coated discharges. Significant increase of the particle and energy confinement times and the reduction of Zeff of the plasma certainly indicate the improved plasma parameters in the Aditya tokamak after lithium wall conditioning.

In the present work, polyester film and fabric were exposed to dielectric barrier discharge in air at atmospheric pressure to increase wettability and dye uptake properties. The surface of dielectric barrier discharge treated samples was analyzed using contact angle goniometer, scanning electron microscope, and Fourier transform infrared spectrometer. The plasma was characterized electrically and optically to understand plasma chemistry. Significant improvement in wettability of polyester substrate and substantial decrease in water contact angle at surface of polyester film after plasma treatment were attributed to the formation of polar functional groups and is due to morphological changes. The aging effect was also studied on plasma treated polyester surface. Polyester fabric was dyed with natural dispersive dye eco-alizarin. K/S value has increased significantly from 5.29 for untreated to 9.25 for air plasma treated polyester fabric, which clearly indicate an increase in color intensity on plasma expos...

Transport characteristics of the stochastic magnetic boundary of the Large Helical Device (LHD) are investigated, based on three-dimensional Monte-Carlo Braginskii-type fluid model code, EMC3, coup...

Dust events are recorded in the SST-1 tokamak with fast visible imaging and infrared imaging systems. The events are analyzed from the image sequences with an in-house developed dust tracking routine in MATLAB called FINDUST. Specific operating scenarios for dust observation and their deleterious effects on discharge evolution are identified. A clear correlation of the dust events frequency and strength with the successful start-up and discharge evolution is observed over eight consecutive experimental campaigns in SST-1. It has been seen that the electron cyclotron resonance heating (ECRH) generated pre-ionization slab annular-plasma around the fundamental layer is most likely to generate dust, while the strongest effect of the ECRH fundamental layer on the plasma facing components is seen when the confined plasma column disrupts by some means while the ECRH pulse is still ON.

Superdense plasma with a highly peaked electron density profile was obtained in the reduced recycling discharge in the Large Helical Device (LHD). During the discharge, a core region with a density as high as ∼5 × 1020 m−3 and a temperature of ∼0.8 keV is maintained by an Internal Diffusion Barrier (IDB) with a steep density gradient. In spite of such a high density at the core region, no serious impurity accumulation has been observed. According to the numerical calculation with the EMC3–EIRENE code, downstream parallel flow in the ergodic region plays an important role to screen the influx of impurities with its strong friction force. The divertor flux distribution and global recycling properties do not change so much as against the change in the dense core plasma.

The experimental work reported here is a validation of Langmuir probe measurements against emission spectroscopic and discharge resistivity measurements of atmospheric glow discharge plasma parameters. The probe measurements of plasma density are close to that obtained by spectroscopy and discharge resistivity measurement. However, it is found that the electron temperature measurement by the probe is higher than that obtained by the spectroscopic technique. Variation of plasma parameters with operating voltage and frequency, which are estimated by the three techniques, shows a similar trend.

In H2 discharges of large helical device (LHD), H-mode-like transition has been obtained with edge localized mode (ELM)-like bursts in an outwardly shifted configuration of R ax = 4.00 m with an extremely thick ergodic layer, where the ι/2π = 1 position is located in the middle of the ergodic layer. This H-mode-like transition can be triggered by changing P NBI (13 cm−3. The spontaneous H-mode-like transition is also observed during a stable discharge with constant heating power and density, where a steep edge pressure gradient is formed at the ι/2π = 1 position. The ELM-like bursts occurred in the ergodic layer accompanied with density bursts. It is found that the density bursts triggered at the ι/2π = 1 position in the ergodic layer instantaneously propagate outward beyond the ergodic layer with positive density pulse, and a few ms later also begin to propagate inward from the ergodic layer with negative density pulse to recover the previous edge density profile. In the case of He discharges no H-mode-like transition and ELM-like bursts are observed. The frequency of the ELM-like bursts increases with heating power, e.g. 50 Hz (3 MW)–100 Hz (5 MW) at n e = 4 × 1013 cm−3. The external resonant magnetic field to generate a static m/n = 1/1 island inside the ergodic layer is supplied to the ELM-like bursts. It is also found that the externally supplied magnetic field makes a clear response even in the ergodic layer. As a result, the frequency of the ELM-like bursts decreases, e.g. from 60 to 45 Hz. The present study strongly indicates that those phenomena observed in LHD are quite different from the ones which have been observed in tokamaks until now.

Major disruptions in Aditya tokamak are initiated by the growth and subsequent locking of m/n = 2/1 and 3/1 tearing modes, which leads to the thermal quench of the plasma. Thick filaments are seen to evolve at the low field side (LFS) of the plasma column following the thermal quench, and during the current quench. The number of filaments and inter filament spacing are observed to be related with the plasma stored energy just prior to the disruption. Rapid enhancement of the outward particle flux is seen during the thermal quench phase and the plasma conductivity reduces considerably. Interchange modes, with low poloidal wavenumber, are inferred to grow due to the reduced plasma conductivity and enhanced effective diffusivity. This may be a plausible explanation for the visualization of the thick filaments at the LFS.

A photo multiplier tube (PMT) array based spectroscopic diagnostic with fast time response of 10 mu s and spatial resolution similar to 3 cm has been developed and installed on Aditya tokamak to study the spatial and temporal behavior of H-alpha emissions from typical discharges. Collimated light has been collected from the plasma along 16 lines of sight passing through entire plasma poloidal cross section of Aditya and detected by two 8 channels PMT arrays after selecting H-alpha emission using interference filter. The studies are carried out during plasma formation phase of Aditya by changing vertical field and its delay with respect to loop voltage. It is observed that plasma initiated in the high field side in typical discharges of Aditya. The plasma formation position is matched with null field location estimated through simulation. (C) 2014 AIP Publishing LLC.

In this paper, the development of large volume double ring Penning plasma discharge source for efficient light emissions is reported. The developed Penning discharge source consists of two cylindrical end cathodes of stainless steel having radius 6 cm and a gap 5.5 cm between them, which are fitted in the top and bottom flanges of the vacuum chamber. Two stainless steel anode rings with thickness 0.4 cm and inner diameters 6.45 cm having separation 2 cm are kept at the discharge centre. Neodymium (Nd(2)Fe(14)B) permanent magnets are physically inserted behind the cathodes for producing nearly uniform magnetic field of ~0.1 T at the center. Experiments and simulations have been performed for single and double anode ring configurations using helium gas discharge, which infer that double ring configuration gives better light emissions in the large volume Penning plasma discharge arrangement. The optical emission spectroscopy measurements are used to complement the observations. The spectral line-ratio technique is utilized to determine the electron plasma density. The estimated electron plasma density in double ring plasma configuration is ~2 × 10(11) cm(-3), which is around one order of magnitude larger than that of single ring arrangement.

Four resonance transitions of CIII (977.03 A:2s2 S10-2s2p P11), CIV (1550 A:2s S2-2p P2), CV (40.27 A:1s2 S10-1s2p P11), and CVI (33.73 A:1s S2-2p P2) have been observed in vacuum ultraviolet and extreme ultraviolet regions to study the edge carbon impurity transport in the Large Helical Device ergodic layer. Here, CIII and CIV indicate the carbon influx at the outside boundary of the ergodic layer and CV and CVI indicate the ions in higher ionization stages, which have already experienced the transport in the ergodic layer. The intensity ratio of CV+CVI to CIII+CIV, therefore, represents the degree of impurity screening, which has been analyzed with different edge plasma parameters and ergodic magnetic field structures. The ratio decreases by two orders of magnitude with an increase in electron density ne in the range of 1–8×1019 m−3. The CV and CVI emissions tend to decrease with ne, whereas the CIII and CIV emissions monotonically increase with ne. The result suggests an enhancement of the impurity scr...

Z(eff) measurement system using a visible spectrometer has been newly designed and constructed instead of an old interference filter system to eliminate line emissions from the signal and to measure the Z(eff) value in low-density plasmas. The system consists of the Czerny-Turner-type spectrometer with a charge-coupled device camera and 44 optical fibers vertical array. The spectrometer is equipped with an additional toroidal mirror for further reduction in the astigmatism in addition to a flat and two spherical mirrors and three gratings (110, 120, and 1200 grooves/mm) with 30 cm focal length. The images from 44 optical fibers can be detected without astigmatism in a wavelength range of 200-900 nm. Combination of the optical fiber (core diameter: 100 microm) with the lens (focal length: 30 mm) provides spatial resolution of 30 mm at the plasma center. Results clearly indicate a very good focus image of the fiber and suggest the absence of the cross-talk between adjacent fiber images. Absolute intensity calibration has been done using a standard tungsten lamp to analyze the Z(eff) value. The bremsstrahlung profile and resultant Z(eff) profile have been obtained after Abel inversion of the signals observed in large helical device plasmas with elliptical poloidal cross section.

A flat-field extreme ultraviolet (EUV) spectrometer with a nominal 2400 grooves/mm aberration-corrected ruled grating has been developed to analyze the emission spectrum in the wavelength range of 10 to 130 A from large helical device (LHD) plasmas. Spectral properties such as resolution, sensitivity, contribution of higher-order light, and background stray light have been studied using emission spectra mainly from intrinsic impurities, e.g., C and Fe. It is found that the spectrometer well resolves closely existing spectral lines of highly ionized medium- and high-Z impurities even in a very short wavelength range such as 10 to 20 A. As a result, it allows one to study the charge state distribution of elements in high-temperature fusion plasma. The ruled grating was then replaced by a laminar type holographic grating for the comparative study. The spectral resolution for the ruled grating (Dlambda approximately 0.08 A at 33.73 A) is clearly better than the holographic grating (Dlambda approximately 0.13 A at 33.73 A). Both gratings well suppress the higher-order light, e.g., the second-order light is only less than 11% of the first-order light for C vi(33.73 A). Relative sensitivity calibration with the wavelength has been done using bremsstrahlung continuum from the LHD plasmas. Absolute intensity calibration has been done by comparing the spectral intensities directly with the absolutely calibrated 1200 grooves/mm EUV spectrometer in the overlapping range of 90-120 A due to the absence of a good branching pair in 10-130 A. As a typical result on the present spectrometer well-resolved n=2-3 full transition arrays from Ne- to Li-like ions are measured for Fe and Ti and wavelengths of the spectral array are tabulated for each charge state. Spectroscopic comparison is also made between the 1200 grooves/mm and 2400 grooves/mm gratings in a range of 50-130 A.

A flat-field extreme ultraviolet (EUV) spectrometer with a varied line spacing groove grating (1200 grooves/mm at grating center) has been developed to study the emission spectra from highly ionized medium Z impurities in large helical device (LHD). It covers a wavelength range of 50-500 A using a mechanically ruled grating, which was later replaced by a newly developed laminar-type holographic grating for comparative studies. Differences in spectral resolution, intensities of higher order spectra, and sensitivities of the spectrometer were studied between the two gratings by observing the emission spectra of LHD plasmas. Although the achieved resolution was alike between them, i.e., deltalambda approximately 0.24 A at 200 A, the holographic grating was much superior in suppressing the higher order light than the ruled grating. The relative sensitivity between the two gratings was evaluated using continuum radiation from LHD plasmas. As a result, it was found that the holographic grating has a flat response in the full wavelength range, but the sensitivity of the ruled grating drops sharply below 200 A. A new technique for the absolute calibration of the EUV holographic grating spectrometer was tried by combining the continuum radiation with a branching ratio of C IV lines (3p-3s: 5800 A/3p-2s: 312 A), and an accurate absolute sensitivity has been successfully obtained.

A steady state superconducting tokamak (SST-1) has been commissioned after the successful experimental and engineering validations of its critical sub-systems. During the ?engineering validation phase? of SST-1; the cryostat was demonstrated to be leak-tight in all operational scenarios, 80?K thermal shields were demonstrated to be uniformly cooled without regions of ?thermal runaway and hot spots?, the superconducting toroidal field magnets were demonstrated to be cooled to their nominal operational conditions and charged up to 1.5?T of the field at the major radius. The engineering validations further demonstrated the assembled SST-1 machine shell to be a graded, stress?strain optimized and distributed thermo-mechanical device, apart from the integrated vacuum vessel being validated to be UHV compatible etc. Subsequently, ?field error components? in SST-1 were measured to be acceptable towards plasma discharges. A successful breakdown in SST-1 was obtained in SST-1 in June 2013 assisted with electron cyclotron pre-ionization in the second harmonic mode, thus marking the ?first plasma? in SST-1 and the arrival of SST-1 into the league of contemporary steady state devices.Subsequent to the first plasma, successful repeatable plasma start-ups with E???0.4?V?m?1, and plasma current in excess of 70?kA for 400?ms assisted with electron cyclotron heating pre-ionization at a field of 1.5?T have so far been achieved in SST-1. Lengthening the plasma pulse duration with lower hybrid current drive, confinement and transport in SST-1 plasmas and magnetohydrodynamic activities typical to large aspect ratio SST-1 discharges are presently being investigated in SST-1. In parallel, SST-1 has uniquely demonstrated reliable cryo-stable high field operation of superconducting TF magnets in the two-phase cooling mode, operation of vapour-cooled current leads with cold gas instead of liquid helium and an order less dc joint resistance in superconducting magnet winding packs with high transport currents. In parallel, SST-1 is also continually getting up-graded with first wall integration, superconducting central solenoid installation and over-loaded MgB2-brass based current leads etc. Phase-1 of SST-1 up-gradation is scheduled by the first half of 2015, after which long pulse plasma experiments in both circular and elongated configurations have been planned in SST-1.

This paper summarizes the results of recent dedicated experiments on disruption control and runaway mitigation carried out in ADITYA, which are of the utmost importance for the successful operation of large size tokamaks, such as ITER. It is quite a well-known fact that disruptions in tokamaks must be avoided. Disruptions, induced by hydrogen gas puffing, are successfully avoided by two innovative techniques in ADITYA using a bias electrode placed inside the last closed flux surface and applying an ion cyclotron resonance pulse with a power of ∼50 to 70kW. These experiments led to better understanding of the disruption avoidance mechanisms and also can be thought of as one of the options for disruption avoidance in ITER. In both cases, the physical mechanism seems to be the control of magnetohydrodynamic modes due to increased poloidal rotation of edge plasma generated by induced radial electric fields. Real time avoidance of disruption with identifying proper precursors in both the mechanisms is successfully attempted. Further, analysing thoroughly the huge database of different types of spontaneous and deliberately-triggered disruptions from ADITYA, a significant contribution has been made to the international disruption database(ITPA).Furthermore, themitigationoftherunawayelectrongeneratedmainlyduringdisruptionsremainsachallenging topic in present tokamak research as these high-energy electrons can cause severe damage to in-vessel components and the vacuum vessel. A simple technique has been implemented in ADITYA to mitigate the runaway electrons before they can gain high energies using a localized vertical magnetic field perturbation applied at one toroidal location to extract runaway electrons.

Radial profiles of impurity spectral line emissions in extreme ultraviolet (EUV) range (60-400 A) are surveyed for impurity seeded discharges in Large Helical Device (LHD). The line emissions from light impurities such as helium, carbon and neon are located in the edge ergodic layer (1.0 ≤ ρ ≤ 1.2) except for CV and CVI, which are located in region of 0.8 ≤ ρ ≤ 1. In contract, the line emissions from heavy impurities such as iron are widely located over the whole plasma region as a function of ionization energy. On the other hand, the radial profile of bremsstrahlung continuum radiation is also clearly recorded in the EUV spectrum observed from high-density discharges with hydrogen pellet injection. Wavelength intervals available for the absolute sensitivity calibration of the present space-resolved EUV spectrometer system are listed with dominant line emissions which are useful as a wavelength marker.

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

Remarkable progress in the physical parameters of net-current free plasmas has been made in the Large Helical Device (LHD) since the last Fusion Energy Conference in Chengdu, 2006 (Motojima et al 2007 Nucl. Fusion 47 S668). The beta value reached 5% and a high-beta state beyond 4.5% from the diamagnetic measurement has been maintained for longer than 100 times the energy confinement time. The density and temperature regimes have also been extended. The central density has exceeded 1 × 1021 m−3 due to the formation of an internal diffusion barrier. The ion temperature has reached 5.2 keV at the density of 1.6 × 1019 m−3, which is associated with the suppression of ion heat conduction loss. Although these parameters have been obtained in separated discharges, each fusion-reactor relevant parameter has elucidated the potential of net-current free heliotron plasmas. Diversified studies in recent LHD experiments are reviewed in this paper.

Spectral characteristics of varied line spacing 2400 grooves/mm gratings used for a flat-field extreme ultraviolet spectrometer have been evaluated to monitor Delta n=1 (n=2-3) transitions from medium-Z impurities in 10-30 A, where many spectral lines from metallic impurities closely exist. For this purpose emission spectra of the Delta n=1 transitions from partially L-shell-ionized Ti and Fe have been measured in large helical device. At first, a ruled and a holographic grating were utilized for the comparative study. Both gratings suppressed higher order light effectively but spectral sensitivity was much higher for the holographic grating. The ruled grating, however, attained better spectral resolution (0.083 A at 18.97 A) in the 10-100 A range compared to that (0.130 A) of the holographic grating. After that a recently developed new holographic grating with different laminar shapes and line spacings has attempted to improve the spectral resolution. As a result, it has considerably improved the spectral resolution in the 10-30 A range, i.e., approximately 0.090 A. However, it is found that the experimentally obtained spectral resolution is quit poor compared to the computationally simulated one, suggesting a technical difficulty in manufacturing exactly the varied line spacing of the holographic grating at such a short wavelength range.

A diagnostic system based on a multi-fiber input high resolution spectrograph has been set up on the Aditya tokamak (Bhatt et al 1989 Ind. J. Pure Appl. Phys. 27 710) for utilizing the passive light emission to measure different kinds of plasma flow and to identify the location of emissions of hydrogen and impurities along with their temperatures. Eight simultaneous vertically collimated lines-of-sight from a top port view a poloidal cross-section of the plasma. This arrangement simplifies the analysis of spectra in terms of making the Zeeman splitting easier to account for, since each chord passes through a region of nearly constant toroidal magnetic field (BT). This paper describes the complete set-up, the wavelength and intensity calibrations performed and the initial results including the impurity emissivity profiles and simultaneous flow measurements in the inboard and outboard regions of the Aditya tokamak.