In contrast to the earlier experiments conducted in other machines, here, in SST-1 the error field measurement experiment is performed with a filled gas pressure similar to 8 x 10(-4) mbar which helped to create a luminescent toroidal beam of electron path originated due to impact excitation and guided by the toroidal magnetic field. Beam path deviations are observed and recorded from radial and top ports using visible range cameras. Such creation and detection of the electron beam path differs from the earlier works where the gun emitted electron beam deviation in ultra-high vacuum was detected on a collector-grid/fluorescent screen. In the present experiment, large beam deviations were observed. Later investigation of the experimental set-up reveals existence of a possible source of radial electric field in between the source and the vacuum vessel which are separately grounded. Thus, to understand the observed phenomena, experiments are numerically modeled with deviated TF coil set, PF coil set and the electron source location. A particle tracing code is used to follow the electron path in the magnetic field generated by the coil set of interest. Simulation results suggest that the large deviation corresponds to the E x B drifts and not due to the large field errors. Toroidally averaged field errors of the SST-1 TF coils at toroidal field of B-0 = 15 kG are negligibly small similar to B-0 x 10(-6) or less, which should not adversely affect the plasma performance.

ITER Vacuum Vessel is a torus-shaped, double wall structure. The space between the double walls of the VV is filled with In-Wall Shielding Blocks (IWS) and Water. The main purpose of IWS is to provide neutron shielding during ITER plasma operation and to reduce ripple of Toroidal Magnetic Field (TF). Although In-Wall Shield Blocks (IWS) will be submerged in water in between the walls of the ITER Vacuum Vessel (VV), Outgassing Rate (OGR) of IWS materials plays a significant role in leak detection of Vacuum Vessel of ITER. Thermal Outgassing Rate of a material critically depends on the Surface Roughness of material. During leak detection process using RGA equipped Leak detector and tracer gas Helium, there will be a spill over of mass 3 and mass 2 to mass 4 which creates a background reading. Helium background will have contribution of Hydrogen too. So it is necessary to ensure the low OGR of Hydrogen. To achieve an effective leak test it is required to obtain a background below 1 × 10-8 mbar 1 s-1 and hence the maximum Outgassing rate of IWS Materials should comply with the maximum Outgassing rate required for hydrogen i.e. 1 x 10-10 mbar 1 s-1 cm-2 at room temperature. As IWS Materials are special materials developed for ITER project, it is necessary to ensure the compliance of Outgassing rate with the requirement. There is a possibility of diffusing the gasses in material at the time of production. So, to validate the production process of materials as well as manufacturing of final product from this material, three coupons of each IWS material have been manufactured with the same technique which is being used in manufacturing of IWS blocks. Manufacturing records of these coupons have been approved by ITER-IO (International Organization). Outgassing rates of these coupons have been measured at room temperature and found in acceptable limit to obtain the required Helium Background. On the basis of these measurements, test reports have been generated and got approved by IO. This paper will describe the preparation, characteristics and cleaning procedure of samples, description of the system, Outgassing rate Measurement of these samples to ensure the accurate leak detection.

Supersonic molecular beam injection (SMBI) is one of the most effective fuelling methods for injecting neutral particles at very high velocity into the plasma core. Due to higher speed and lower divergence, the beam penetrates several centimetres into the plasma and hence increases the fuelling efficiency. For SST-1 Tokamak, two types of SMBI systems are proposed. One will be installed in the low field side (LFS) while other will be integrated in the high field side (HFS). This paper describes the design, fabrication and implementation of SMBI system in SST-1 Tokamak.

Graphite plasma facing components (PFCs) were installed inside the SST-1 vacuum vessel. Prior to installation, all the graphite tiles were baked at 1000 degrees C in a vacuum furnace operated below 1.0 x 10(-5) mbar. However due to the porous structure of graphite, they absorb a significant amount of water vapour from air during the installation process. Rapid desorption of this water vapour requires high temperature bake-out of the PFCs at > 250 C. In SST-1 the PFCs were baked at 250 C using hot nitrogen gas facility to remove the absorbed water vapour. Also device with large graphite surface area has the disadvantage that a large quantity of hydrogen gets trapped inside it during plasma discharges which makes density control difficult. Helium glow discharge cleaning (He-GDC) effectively removes this stored hydrogen as well as other impurities like oxygen and hydrocarbon within few nano-meters from the surface by particle induced desorption. Before plasma operation in SST-1 tokamak, both baking of PFCs and He-GDC were carried out so that these impurities were removed effectively. The mean desorption yield of hydrogen was found to be 0.24. In this paper the results of baking and He-GDC experiments of SST-1 will be presented in detail.

SST-1 Tokamak has recently completed the 1st phase of up-gradation with successful installation and integration of all its First Wall components. The First Wall of SST-1 comprises of similar to 3800 high heat flux compatible graphite tiles being assembled and installed on 132 CuCrZr heat sink back plates engraved with similar to 4 km of leak tight baking and cooling channels in five major sub groups equipped with similar to 400 sensors and weighing similar to 6000 kg in total in thirteen isolated galvanic and six isolated hydraulic circuits. The phase-1 up-gradation spectrum also includes addition of Supersonic Molecular Beam Injection (SMBI) both on the in-board and out-board side, installation of fast reciprocating probes, adding some edge plasma probe diagnostics in the SOL region, installation and integration of segmented and up-down symmetric radial coils aiding/ controlling plasma rotations, introduction of plasma position feedback and density controls etc. Post phase-I up-gradation spanning from Nov 2014 till June 2016, initial plasma experiments in up-graded SST-1 have begun since Aug 2016 after a brief engineering validation period in SST-1. The first experiments in SST-1 have revealed interesting aspects on the 'eddy currents in the First Wall support structures' influencing the 'magnetic Null evolution dynamics' and the subsequent plasma start-up characteristics after the ECH pre-ionization, the influence of the first walls on the 'field errors' and the resulting locked modes observed, the magnetic index influencing the evolution of the equilibrium of the plasma column, low density supra-thermal electron induced discharges and normal ohmic discharges etc. Presently; repeatable ohmic discharges regimes in SST-1 having plasma currents in excess of 65 KA (q(a) similar to 3.8, B-T = 1.5 T) with a current ramp rates similar to 1.2 MA/s over a duration of similar to 300 ms with line averaged densities similar to 0.8 x 10(19) and temperatures similar to 200 eV with copious MHD signatures have been experimentally established. Further elongation of the plasma duration up to one second or more with position and density feedback as well as coupling of Lower Hybrid waves are currently being persuaded in SST-1 apart from increasing the core plasma parameters with further optimizations and with wall conditioning.

Steady State Superconducting Tokamak (SST-1) is a `working‟ experimental superconducting device since late 2013. Referring to the last IAEA-FEC; SST-1 has been upgraded with Plasma Facing Components (PFC) with an objective towards long pulse operations in both circular and elongated configurations. The PFC integration has been completed in August 2015 and initial experiments have begun in SST-1 with circular plasma configurations. SST-1 offers a unique possibility of investigating long pulse discharges with large aspect ratio (> 5.5) compared to contemporary devices. Presently, SST-1 standard Ohmic discharges are in excess of 100 kA with typical core density ~ 1 × 10 19 m -3 and core electron temperatures ~ 200-300 eV having duration in excess of 300 ms. A 42 GHz ECR pre-ionization source at ~ 150 kW in 1.5 T central toroidal field breaks down the hydrogen gas, the current starts up at ~ 1.3 MA/s in 60-80 ms in an induced electric field of ~ 0.3 V/m. These standard discharges demonstrate copious saw teething and MHD activities as the pulse progresses including NTM, mode locking and MHD characteristics. Studies on eddy currents influencing the NULL dynamics, field errors, equilibrium index evolutions, wall influencing plasma characteristics, plasma positions, plasma rotational andTearing Mode characteristics including the island width and growth etc. have been carried out. Presently, SST-1 is attempting at multi-second long high aspect ratio plasma discharges by coupling the Lower Hybrid with the Ohmic plasma as well as with robust real time position and density controls. SST-1 device has also been upgraded with a pair of internal coil aimed at effective fast plasma control and a pair of segmented coil aimed at controlling some of the rotational aspects of plasma including the RMPs and ELMs. Supersonic Molecular Beam Injection (SMBI) from both high field and low field sides and Pellets Injection Systems have also been added with several edge plasma diagnostics aimed at both density control and edge plasma turbulence studies. This paper elaborates some of these up-gradation and results from the initial experiments.

SST-1 Tokamak was successfully commissioned in 2012 and the first plasma was achieved in June 2013 with poloidal limiters having SS 304L as vessel wall material. Due to plasma wall interactions, high -Z impurities released from the vessel wall which in turn cools the plasma by radiation loss. In order to reduce this effect, in 2 nd phase of SST-1 refurbishment PFC components were installed in the system. PFCs were integrated inside SST-1 vacuum vessel which is designed to withstand an input heat load of 1.0 MW/m 2 . Graphite was chosen as Plasma facing material considering its good thermal properties, low atomic mass. Cu -Zr& Cu-Cr-Zr alloys plates embedded with SS 304L piping were used as back plate materials for proper heat conduction. Each and every component was tested at their functional conditions to verify its functionality and to ensure operation conformity. Approximately 3800 tiles were mounted on 132 numbers of copper alloys back-plates. The total surface area of the installed PFCs exposed to plasma is about 40 m 2 which is nearly 50% of the total surface area of stainless steel vacuum chamber (~75 m 2 ). The volume of the vessel with the PFCs is ~ 16 m 3 . Gas-to-gas heat exchange method was adapted to heat nitrogen gas which is pressurized using dedicated gas blower system to bake the PFC components. All PFC components pas sed through temperature of 250 C for 8 hours flat top and working pressure of 4 bar(a) under UHV conditions in validation testes. Strict metrology and QA/QC plans were structured and executed to integrate the PFC components inside the vacuum vessel. During pump down of SST -1 main vacuum vessel, PFCs were baked at 250 °C for nearly 10 days to remove the absorbed water vapours. At this condition, this main vacuum vessel was maintained at 150 °C. In addition, initially hydrogen discharge cleaning was carried out followed with subsequent helium discharge cleaning to removeother surface impurities. With all PFCs and diagnostic integrated to the s ystem, a base pressure of 4.5 10 –8 mbar was achieved. This paper represents SST-1 post PFC Plasma-scenario, PFC requirement inciting factors, PFC architecture and lay-out details, PFC components experimental validations, metrology plan with QA/QC and fina l installation of PFC with the vacuum vessel.

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 similar to 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.