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.

Intense visible lines from Be-like oxygen impurity are routinely observed in the Aditya tokamak. The spatial profile of brightness of a Be-like oxygen spectral line (2p3p 3D3–2p3d 3F4) at 650.024 nm is used to investigate oxygen impurity transport in typical discharges of the Aditya tokamak. A 1.0 m multi-track spectrometer (Czerny–Turner) capable of simultaneous measurements from eight lines of sight is used to obtain the radial profile of brightness of O4+ spectral emission. The emissivity profile of O4+ spectral emission is obtained from the spatial profile of brightness using an Abel-like matrix inversion. The oxygen transport coefficients are determined by reproducing the experimentally measured emissivity profiles of O4+, using a one-dimensional empirical impurity transport code, STRAHL. Much higher values of the diffusion coefficient compared with the neo-classical values are observed in both the high magnetic field edge region and the low magnetic field edge region of typical Aditya ohmic plasmas, which seems to be due to fluctuation-induced transport. The diffusion coefficient at the limiter radius in the low-field (outboard) region is typically ∼ twice as high as that at the limiter radius in the high-field (inboard) region.

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 Z eff of the plasma certainly indicate the improved plasma parameters in the Aditya tokamak after lithium wall conditioning.

Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z eff . Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values.

A tomographic reconstruction code had been reported by us for inferring the poloidal emissivity of tokamak plasma from tangentially acquired images. Here we present modifications to the code that account for any diffuse reflections from the surfaces of walls enclosing the plasma. It is generally recognized that such reconstruction codes are highly susceptible to noise in the data. In this work we have analysed the sensitivity to noise for varying degrees of over-determinism in the set of equations; over-determinism is defined as the ratio of the number of detector signals available to the grid resolution of reconstruction. A tractable scheme for dividing the poloidal cross section into a finite number of unknown sub-tori and voids, while still keeping the over-determinism high, is incorporated. Finally it is shown that noise level >20% can be handled with over-determinism achievable from present day detector array/cameras. The singular value decomposition of the matrix, as used here, can be expected to converge even if any ill-conditioned matrix is encountered due to computational round-off errors in the estimation of chord lengths through sub-tori and voids.