[en] The Steady-State Superconducting Tokamak (SST-1) is a medium-size tokamak with super-conducting magnetic field coils. Plasma facing components (PFC) of the SST-1, consisting of divertors, passive stabilisers, baffles, and poloidal limiters, are designed to be compatible for steady-state operation. Except for the poloidal limiters, all other PFC are structurally continuous in the toroidal direction. As SST-1 is designed to run double-null divertor plasmas, these components also have up-down symmetry. A closed divertor configuration is chosen to produce high recycling and high pumping speed in the divertor region. The passive stabilisers are located close to the plasma to provide stability against the vertical instability of the elongated plasma. The main consideration in the design of the PFC is the steady-state heat removal of up to 1 MW/m². In addition to removing high heat fluxes, the PFC are also designed to be compatible for baking at 350 deg. C. Different flow parameters and various tube layouts have been examined to select the optimum thermal-hydraulic parameters and tube layout for different PFC of SST-1. Thermal response of the PFC during baking has been performed analytically (using a Fortran code) and two-dimensional finite element analysis using ANSYS. The detailed thermal hydraulics and thermal responses of PFC baking is presented in this paper

[en] Highlights: • Non-Destructive Evaluation method based on ultrasonic technique has been developed for Inconel 625 weld joints in ELM control coils. • Probe selection, beam parameters and procedure has been simulated using CIVA software and validated by experimental trials. • Inspection procedure and parameters has been optimized and demonstrated on actual 1:1 demo ELM coil. - Abstract: Institute for Plasma Research (IPR) has been developing technologies appropriate towards realizing fusion relevant Edge Localized Mode (ELM) magnets in Tokamaks such as Joint European Torus (JET) and Steady State Superconducting Tokamak (SST-1). The winding pack of ELM coils have been encased inside the Inconel 625 casing using customized double groove and single bevel weld joints. It is critical to ensure soundness of these weld joints. With this motivation, Non-Destructive Evaluation (NDE) method using ultrasonic technique has been developed and established for inspection of weld joints in ELM control coils. The development of ultrasonic test procedure, test parameters, test validation and optimization using simulation as well as experimental approach has been discussed in this paper. The selection and locations probes along with beams parameters for weld joints have been optimized with commercially available software CIVA- versatile simulation tool developed by CEA (French Atomic Energy Commission), France. Further, the developed ultrasonic test method has been established and validated on mockup box sample of casing for ELM coils. It has been again successfully demonstrated for its performance on 1:1 demo coil equivalent to small ELM coil.

[en] Full text: Tungsten divertor target technology development is in progress at IPR for water-cooled divertors of ITER-like tokamak. Test mock-ups are fabricated using tungsten materials in macro-brush as well as mono-block fashion. Vacuum brazing technique is used for macro-brush fabrication whereas high pressure high temperature diffusion bonding technique is used for mono-block fabrication. Experimental facilities are also being set-up at IPR for Non-destructive testing and high heat flux testing of divertor targets. Present paper describes recent results on high heat flux testing of the test mock-ups and briefly mention about some of the experimental test facilities being set-up at IPR. (author)

[en] Steady State Superconducting Tokamak (SST-1) is a medium sized Tokamak. SST-1 Vacuum Vessel (VV) is one of the sub-systems of SST-1 Tokamak. The Vacuum Vessel provides an Ultra High Vacuum (UHV) environment for in-vessel components and plasma production. SST-1 vacuum vessel is a continuous torus structure fabricated using non-magnetic SS 304L material. For easy fabrication and assembly point of view, SST-1 vacuum vessel is divided into sixteen parts, out of which eight of them are vessel sectors (VS) while the other eight are vessel modules (VM). Vessel Sector (VS) is comprised of one number of Radial Port (RP), two numbers of vertical ports (top and bottom each) and one number of vessel sector ring. Vessel Module (VM) is made up of one number of Vessel Sector and two numbers of Inter Connecting Rings (ICR) on both sides. This paper will describe about the temperature measurements and calorimetric calculations of SST-1 Vacuum Vessel and its results

[en] SST-1 is a medium size tokamak with super conducting magnetic field coils. All the subsystems of SST-1 are designed for quasi steady state (∼1000 s) operation. Plasma Facing Components (PFCs) of SST-1 consisting of divertors, passive stabilizers, baffles and poloidal limiters are also designed to be compatible for steady state operation. As SST-1 is designed to run double null divertor plasmas, these components also have up-down symmetry. A closed divertor configuration is chosen to produce high recycling and high pumping speed in the divertor region. All the PFC are made of copper alloys (CuCrZr and CuZr) on which graphite tiles are mechanically attached. These copper alloy back plates are actively cooled with water flowing in the channels grooved on them with the main consideration in the design of PFCs as the steady state heat removal of about 1.0 MW/m². In addition to be able to remove high heat fluxes, the PFCs are also designed to be compatible for baking at 350 degree sign C. Extensive studies, involving different flow parameters and various cooling layouts, have been done to select the final cooling parameters and layout. Thermal response of the PFCs and vacuum vessel during baking, has been calculated using a FORTRAN code and a 2-D finite element analysis. The PFCs and their supports are also designed to withstand large electro-magnetic forces. Finite element analysis using ANSYS software package is used in this and other PFCs design. The engineering design including thermal hydraulics for cooling and baking of all the PFCs is completed. Poloidal limiters are being fabricated. The remaining PFCs, viz. divertors, stabilizers and baffles are likely to go for fabrication in the next few months. The detailed engineering design, the finite element calculations in the structural and thermal designs are presented in this paper

[en] Steady-state superconducting tokamak (SST-1) is a medium size tokamak with superconducting magnetic field coils. Plasma facing components (PFC) of SST-1 are placed inside the vacuum vessel (VV) of the tokamak and are designed to be compatible for steady-state operation. The main consideration in the design of the PFC is the steady-state heat removal of up to 1 MW/m². In addition to remove high heat fluxes, the PFC are also designed to be compatible for baking at high temperature. Since it is difficult to calculate the radiation heat loads between PFC and VV in a 3-D irregular geometry, a simplified model of concentric cylinders has been chosen for the purpose of estimation of the power requirements and the thermal responses of PFC and VV during their bakeout phases. Thermal responses of the PFC and VV have been analysed and the analytical results have been compared with 2-D finite element analysis using ANSYS. The radiation losses between PFC and VV also have been evaluated on the actual model containing all PFC inside the VV

[en] Steady-state Superconducting Tokamak (SST-1) is being designed for 1000 seconds of continuous operation with total input power up to 1.0 MW. The tokamak will have a D-shaped plasma with double null divertor. Each toroidal module of inboard as well as outboard divertor will have graphite tiles mechanically mounted on a metallic back-plate made of high strength copper alloy. The poloidal cross-section of divertor plates has been designed such that the average heat flux on graphite tiles is less than 0.6 MW/m² which is nearly the limit on incident heat flux for mechanically attached graphite tiles for plasma facing components in tokamaks. As a first step towards development of plasma facing components, preliminary experiments are conducted with a test mockup to develop the mechanical mounting scheme for attachment of graphite tiles on actively cooled copper back-plate. This paper describes the results of these experiments. (author)

[en] A conceptual design of vacuum chamber is proposed to study the thermal response of high heat flux components under energy depositions of the magnitude and durations expected in plasma fusion devices. It is equipped with high power electron beam with maximum beam power of 200 KW mounted in a stationary horizontal position from back side of the chamber. The electron beam is used as a heat source to evaluate the heat removal capacity, material performance under thermal loads and stresses, thermal fatigue etc on actively cooled mock – ups which are mounted on a flange system which is the front side door of the chamber. The tests mock – ups are connected to a high pressure high temperature water circulation system (HPHT-WCS) operated over a wide range of conditions. The vacuum chamber consists of different ports at different angles to view the mock -up surface available for mock -up diagnostics. The vacuum chamber is pumped with different pumps mounted on side ports of the chamber. The chamber is shielded from X – rays which are generated inside the chamber when high-energy electrons are incident on the mock-up. The design includes development of a conceptual design with theoretical calculations and CAD modelling of the system using CATIA V5. These CAD models give an outline on the complete geometry of HHF test chamber, fabrication challenges and safety issues. FEA analysis of the system has been performed to check the structural integrity when the system is subjected to structural and thermal loads.

[en] Tungsten is one of the most promising candidate materials for an application as plasma facing material (PFM) in future thermonuclear fusion devices. In the divertor region PFM will be exposed to very severe heat load conditions. Steady state heat loads will be around 5 MW/m"2 - 10 MW/m"2 and in some parts of the divertor even 20 MW/m"2 for short periods. These steady state heat loads are accompanied by transient events such as edge localised modes (ELMs), vertical displacement events (VDEs) and plasma disruptions which deposit very high heat loads of up to several MJ/m"2 on the PFMs. Important properties which enable tungsten to withstand such environmental conditions are its high melting point, high thermal conductivity and low erosion rate. In present study, we have fabricated Pure Tungsten (W) and Tungsten +1wt. % La_2O_3 (WL) pallets of size 50 mm (dia) and 9 mm (height) through Powder Metallurgical process using graphite mould and sintered in Direct Sintering Press (DSP) at 2390 °C with 40 MPa pressing force. The crack formations and surface modification behaviours under transient high heat load condition were investigated. Detailed characterization of the exposed sample for its various properties evaluation will be discussed and presented in the paper

[en] Steady state Superconducting Tokamak, SST-1, is a medium sized tokamak with major and minor radii of 1.10 m and 0.20 m respectively. Elongated plasma operation with double null poloidal divertor is planned with a maximum input power of 1 MW. The Plasma Facing Components (PFC) like Divertors and Baffles, Poloidal limiters and Passive stabilizers form the first material boundary around the plasma and hence receive high heat and particle fluxes. The PFC design should ensure efficient heat and particle removal during steady state tokamak operation. A closed divertor geometry is adopted to ensure high neutral pressure in the divertor region (and hence high recycling) and less impurity influx into the core plasma. A set of poloidal limiters are provided to assist break down, current ramp-up and current ramp down phases and for the protection of the in-vessel components. Two pairs of Passive stabilizers, one on the inboard and the other on the outboard side of the plasma, are provided to slow down the vertical instability growth rates of the shaped plasma column. All PFCs are actively cooled to keep the plasma facing surface temperature within the design limits. The PFCs have been shaped/profiled so that maximum steady state heat flux on the surface is less than 1 MW/m². (author)