Steady State Superconducting Tokamak Research Articles

3D computational fluid dynamics analysis of PINI ion source back plate under high heat flux condition

Neutral Beam Injection (NBI) plays an essential role in Tokamak plasma heating and current drive. A positive ion-based NBI (PNBI) system is adopted in Steady State Superconducting Tokamak- 1 (SST-1). This PNBI system is capable of generating neutral hydrogen beam power of 1.7 MW at 55 keV. This system has a JET PINI (Positive Ion Neutral Injector) type of ion source. The Back Plate (BP) is a component of the PINI ion source. It consists of an SS 304 L magnet positioning & cover plate, and an OFE copper cooling plate. The back plate plays a vital role in removing the high heat load intercepted during beam operation. The present study describes 3D Computational Fluid Dynamics (CFD) analysis of the actual size back plate using the ANSYS version of R121. The 3D CFD analysis has been done under a 2.5 MW/m2 steady heat load incident on the OFE copper plate. Cooling water is provided in the Inlet header of the back plate with a mass flow rate of 1 kg/s at 34 °C. The analysis gives the surface temperature distribution over the OFE copper cooling plate with a maximum surface temperature of 174 °C and the average temperature is 122 °C. This result is in good agreement with High Heat Flux experimental results.

Development of fast steerable launcher for ECRH system

Electron cyclotron resonance heating (ECRH) launcher is an electro-mechanical system mainly consisting of mirrors (flat and profiled), a steering mechanism, a vacuum barrier window and a UHV (Ultra High Vacuum) compatible gate valve which focuses and steers the ECRH beam into plasma. Presently, Steady-state Superconducting Tokamak (SST)-1 ECRH system has a 42 GHz Gyrotron capable of delivering 500 kW of RF power for the maximum duration of 500 ms. A new steerable ECRH launcher has been developed which can steer the EC beam along both poloidal and toroidal directions. The span of mirror movement in poloidal and toroidal directions is ±20° and ±50°, respectively. The poloidal motion of the mirror can be done in real-time during plasma operation and the mirror can be rotated by 20° in 100 ms. Two AC servo motors independently control the mirror movements through an assembly of motion feedthroughs, spur gears and mechanical linkages. The accuracy, repeatability and speed of both the motions have been measured under UHV in a specially designed vacuum chamber.

Design and development of a liquid nitrogen cooled test cryopump for application in Steady-state Superconducting Tokamak-1

Cryocoolers and liquid helium cooled sorption cryopumps are widely used for the generation of high and ultra-high vacuum in large size experimental systems. Application areas of the sorption cryopumps using only liquid nitrogen is not explored much. Aiming to that, liquid nitrogen cooled test cryopump is designed, developed and fabricated to confirm the application feasibility on Steady-state Superconducting Tokamak-1 (SST-1) for pumping nitrogen and water vapor. After rough vacuum, base pressure ≤2.0E-7 mbar is achieved in the pumping dome using the cryopump. In addition, the pumping performance tests of the cryopump were carried out as per the AVS recommended method. The system offers pumping speed of ∼3616 l/s for nitrogen gas and 15928 l/s (N2 equivalent) for water vapor using liquid nitrogen only and the temperature on the panels were 80–81 K. Adsorption capacity for nitrogen gas was tested up to 6123 mbar-l and maximum gas load applied was 8.0E-2 mbar-l/s. In this paper, the pump design concept, analysis, pumping speed simulations, selected sorbent application and experimental investigations are discussed systematically.

Development of a novel spiral antenna system for low loop voltage current start-up at the Steady State Superconducting Tokamak (SST-1)

In general, superconducting tokamaks require low loop voltage current start-up for the safety purpose of their poloidal field coils. The loop voltage inside the vacuum vessel of Steady State Superconducting Tokamak (SST-1) is low in nature since its central solenoid is located outside the cryostat. The low loop voltage current start-up of the SST-1 is routinely performed by using the electron cyclotron resonance (ECR) method at the toroidal magnetic field = 1.5 T (first harmonic) and 0.75 T (second harmonic). Recently, an alternative RF-based plasma current start-up system has been planned for operating the machine, especially for a higher toroidal magnetic field regime . The system was already developed based on an antenna system, made of a series of combinations of two flat spiral antennas, to assist plasma current start-up at a lower inductive electric field. It has already been tested and installed in the SST-1 chamber. The system testing was performed without a background magnetic field within the frequency regime of 35 MHz–60 MHz at present. The test results show that it can produce an electron density of measured by the Langmuir probe at the expense of 500 W RF power. The spectroscopy results indicate its capability of producing plasma density greater than and an electron temperature of –6 eV. In addition, the results also show the presence of a turbulent electric field of the order of 106 V m−1 at the antenna center and a finite anomalous temperature of neutral particles. Calculations show that the obtained density is sufficient for SST-1 low loop voltage plasma breakdown. The antenna system is also capable of producing plasma at higher frequencies. This article will discuss the development of the prototype and the installed antenna system along with their test results in detail.

Computational studies on fast wave current drive in high beta SST-1 and SST-2 plasmas

The non-inductive current drive in low toroidal beta (β) steady-state superconducting tokamak-1 (SST-1) is obtained using slow waves or Lower Hybrid waves (LHW). If β of the plasma is increased either by increasing the plasma parameters (density/temperature) or by decreasing the toroidal magnetic field, the non-inductive current drive may be supported by Fast Waves (FWs), as wave accessibility is limited for slow waves at these high β plasmas. Non-inductive current drive using FWs is explored for these high β plasmas to identify the optimized parameters (operating frequency and parallel refractive index (n||)) of the fast wave. The studies indicate that FWs may be launched with an operating frequency of 0.5 GHz RF source and n|| around 4 in SST-1 plasma. The GENRAY code is employed to carry out detailed studies on Fast Wave Current Drive (FWCD) analysis using ray tracing techniques. Computational studies on FWCD are carried out where central electron density (ne0) and central electron temperature (Te0) of 6×1019m−3 and 3 keV respectively is assumed. The FWCD displays central depositions of RF power and is attributed to fast radial propagation compared to the parallel propagation along the toroidal magnetic field, resulting in ray transits with large number of reflections. The off-axis current drive in SST-1 plasma requires reduced radial penetration of the wave and demands unrealistic plasma parameter in terms of β. However, for SST-2 plasmas, having enhanced plasma parameters (βN∼1.8), it is found that if FWs are launched with n||∼3 and f∼1.5 GHz, the radial penetration of the wave is impeded and single pass absorption in toroidal direction is observed, thus providing conducive condition for off axis current drive. It indicates that FWs would drive plasma current of 11 kA/MW at ρ∼0.7 for ne0∼2.2×1020m−3 and Te0∼7.7 keV values.

Thermo-structural analysis of SST-1 cryopump

A liquid nitrogen-based cryopump for Steady State Superconducting Tokamak (SST-1) is designed for pumping of water vapour from vacuum vessel during baking and under heavy gas load conditions. During baking, the load of the water vapour molecules increases while the resultant speed of available Turbo-Molecular Pumps is less for water vapour which increases the pressure of the vacuum vessel. Cryopump with the same opening area as of TMPs has more pumping speed for water vapour molecules. Hence, the said cryopump will pump water vapour molecules with higher pumping speed. In principle, water vapour molecules get condensed at a surface temperature of <120 K which corresponds to the saturation pressure of <10−10 mbar. The pump is based on the cryosorption principle where nitrogen gas and water vapour get adsorbed on to the cryosufaces at a temperature of <80 K. The amount of gas pumped and accumulated on cryosurfaces depends on many factors such as the temperature of a gas, microscopic surface roughness of cryosurface, its temperature and thermal conductivity, etc. [1]. Various components of this cryopump are thermal shield, cryopanels coated with suitable sorbent, baffles and vacuum chamber. For reliable and long duration operation of this pump, detailed thermal analysis is done to study the temperature distribution and thermal stresses generated for the efficient operation of the pump. Thermo-structural analysis of the developed pump is carried out for failure protection. The design by analysis method is followed to realize the concept of cryopump for SST-1 and the same is described in this paper.

Design and Analysis of liquid nitrogen cooled sorption cryopump for SST-1 Tokamak

Abstract SST-1 is a Steady State Superconducting Tokamak for a proposed plasma discharge of long duration. This plasma device has two vacuum vessels, the cryostat and the vacuum vessel (VV). Cryostat houses the superconducting magnet systems (TF and PF coils), LN2 cooled thermal shields and hydraulics for these circuits and the plasma will be confined inside the VV [ 1 ]. To have plasma discharges, ultra-high vacuum (UHV) is required to be maintained in VV [ 2 ]. To attain a vacuum level of ∼1 × 10-8 mbar in VV, the plasma-facing components (PFCs) are baked to a temperature of 250 °C and the vacuum vessel is baked to a temperature of 150 °C to remove impurities from the bulk of materials which are adsorbed/absorbed. In baking, the gas load of mainly water vapour and hydrogen gas is increased. The pumping system for vacuum vessel consists of six sets of turbo-molecular pumps (each set consist of 3 TMP) and one cryopump which is sufficient to achieve a vacuum of

Steady-State Operation of High CW Power Circulator: Challenges and Solutions through Simulation and Experiments

The lower hybrid current drive (LHCD) system of the steady-state superconducting tokamak-1 (SST-1) has been augmented with two more klystrons, each rated for 500-kW continuous wave (CW) power at 3.7 GHz to increase its power rating to 2-MW CW. To protect these klystrons at rated power, the existing 250-kW CW circulators, connected to each of the klystrons, are required to be upgraded to 500-kW CW rating. Simulation studies are carried out using COMSOL Multiphysics to ascertain the key parameters which limit its performance to the desired value. In this article, we present the simulation and experimental studies of the high power circulator, highlighting major challenges and its solutions. The effect of ferrite temperature on the RF performance of the circulator is analyzed with the help of multiphysics simulations. Accordingly, the configuration of the circulator is modified. A series of low power tests are carried out to characterize the circulator performance. High power tests, up to a few seconds, have been successfully carried out to ensure that no breakdown is observed at rated power. The results from the simulation studies and experiments have enabled us to conclude that the circulator can handle a forward power of 400-kW CW under matched load conditions.

Generic power supply feedback controller for control of plasma parameters in SST-1

Steady State Superconducting Tokamak (SST-1) uses Radial Control Coil (RCC) for radial control of the plasma in initial experiments. RCC is a resistive single turn coil. There are two radial coils placed symmetrically up and down. These coils are connected either in saddle or parallel configurations. The requirement of rate of rise of current in RCC is 1 MA/s with peak current as ±10 kA. The power supply consists of IGBT based 72 H bridge inverter modules connected in parallel, L-C filter, 12 pulse SCR rectifier circuit, water cooling system and transformer. A versa module euro card (VME) bus based Generic Controller (GC) has been developed to control current in RCC in feedback loop. GC controls gate pulse firing of 72 IGBT based H bridge inverter module using hysteresis band control. GC provides 2.5 kHz to 10 kHz switching with fiber optic gate pulse interface to RCC power supply. This paper describes existing controller along with associated issues, development of VME bus based GC, its hardware and software with flow chart, RTOS VxWorks programming, integration of GC with H bridge inverter, MATLAB simulation, testing results of random and step perturbation with power supply operations.

Development, optimization and validation of ultrasonic testing for NDE of ELM coils

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.

Studies on radial and poloidal particle transport at the edge of SST-1 tokamak

The radial and poloidal particle fluxes occurring at the plasma edge are essential towards understanding the plasma confinement in the tokamak device. In tokamaks, the edge transport barriers play a critical role in the transitions from low confinement (L-mode) to high confinement (H-mode). Recently, the edge plasma profiles have been studied for steady-state superconducting tokamak-1 (SST-1) with the help of an array of Langmuir probes. The floating potential and ion saturation current fluctuations have been measured at different radial and poloidal distance at the edge of SST-1. Increases in magnetic fluctuations associated with enhanced magneto-hydrodynamic (MHD) activities have been found to increase the radial particle flux drastically. It indicates that the MHD activity leads to anomalous particle transport during the tokamak discharge at SST-1. It is found that the average poloidal velocity decreases from ∼4 km/s to ∼3.7 km/s whereas the average radial velocity decreases from ∼7.2 km/s to 4.7 km/s. Further, during MHD activity, it is found that as the magnetic island grows, the radial electric field changes from negative to positive. It is observed that the turbulent particle flux at SST-1 tokamak is predominantly positive and bursty. The local flux probability distribution function shows a clear non-Gaussian character, and it is skewed negatively.

Upgradation plans of SST-1 Cryogenics system at IPR

Steady State Superconducting Tokamak (SST-1) is India’s First Superconducting Tokamak and has Toroidal Field (TF) and Poloidal Field (PF) superconducting coils along with the cold mass support structure weighing about 38 ton of cold mass. A 1.3 kW Helium refrigeration and liquefaction (HRL) at 4.5 K along with its distribution network facilitates the cooling down of the cold mass and cryo-stable operation of SST-1TF magnets. SST-1 experimental campaigns have revealed that the existing HRL is just sufficient for the heat loads acting on it. Further, the SST-1 PF magnets require a higher pressure head and mass flow rate than the nominal values on account of the longer paths of some of the PF magnets. In order to make SST-1 being fully superconducting device, we are introducing superconducting central solenoid (CS). Detailed estimates have been made and it has been found that an additional ∼ 900 W at 4.5 K of cryo power is required towards (i) cooling all the PF magnets (ii) the cooling down and the operation of a new Nb3Sn based CS of SST-1. This paper will elaborate on (i) the experimental heat loads acting on the cryo system (ii) the ‘thermal runaway amongst the PF magnets observed in the SST-1 campaign’ (iii) the robust need of a higher operation pressure up to 2.1 bar (a) (iv) the need of the flow optimizations as per the hydraulic paths (v) the engineering solutions at each of these described (i)-(v) above.

Observation of MHD phenomenon for SST-1 superconducting tokamak

Steady State Superconducting Tokamak (SST-1) is a medium size Tokamak (major radius = 1.1 m, minor radius = 0.2 m) and is operational at the Institute for Plasma Research (IPR), India. In the last few experimental campaigns SST-1 has successfully achieved plasma current in order of 60-70kA and plasma duration in excess of ∼ 500 ms at a central magnetic field of 1.5T. An attempt has made to study the behavior of the magneto-hydrodynamic (MHD) activity during different phases of plasma pulse which leads to major/minor disruptions, its present modes (poloidal/toroidal mode number i.e. m = 2, n = 1) impact on plasma confinement and signature of lock mode and its frequency in the SST-1 plasma using experimental data from Mirnov signals. Observed MHD phenomenon has also been correlated with other diagnostics (i.e. ECE, Density, Soft X-Ray etc.) and heating system (ECRH) for the recent campaigns of SST-1.

Conceptual & Engineering Design of Plug-in Cryostat Cylinder for Super-Conducting Central Solenoid of SST-1

SST-1, country’s first indigenously built steady state super-conducting tokamak is planned to be equipped with an Nb3Sn based superconducting central solenoid, which will replace the existing copper conductor TR1 coil for the purpose of Ohmic breakdown. This central solenoid (CS) of four layers with each layer having 144 turns with an OD of 573 mm, ID of 423 mm length of 2483 mm will be housed inside a high vacuum, CRYO compatible plug-in cryostat thin shell having formed from SS 304L plate duly rolled and welded to form cylinder along with necessary accessories like LN2 bubble panel, current lead chamber, coil and cylinder support structure etc. This paper will present the design drivers, material selection, advantages and constraints of the plug-in cryostat concept, sub-systems of plug-in cryostat, its conceptual and engineering design, CAD models, finite element analysis using ANSYS, safety issues and diagnostics, on-going works about fabrication, quality assurance/control and assembly/integration aspects with in the existing SST-1 machine bore.

MHD mode evolutions prior to minor and major disruptions in SST-1 plasma

Steady State Superconducting Tokamak (SST-1) is a medium size Tokamak (R0=1.1m, a=0.2m, BT=1.5T, Ip∼110kA) in operation at the Institute for Plasma Research, India. SST-1 uniquely experiments large aspect ratio (∼5.5) plasma in different operation regimes. In these experiments, repeatable characteristic MHD phenomena have been consistently observed. As the large aspect ratio plasma pulse progresses, these MHD phenomena display minor-major disruptions ably indicated in Mirnov oscillations, Mirnov oscillations with saw teeth and locked modes etc. Even though somewhat similar observations have been found in some other machines, these observations are found for the first time in large aspect ratio plasma of SST-1. This paper elaborates the magnetic field perturbations and mode evolutions due to MHD activities from Mirnov coils (poloidal and toroidal), Soft X-ray diagnostics, ECE diagnostics etc. This work further, for the first time reports quantitatively different regimes of MHD phenomena observed in SST-1 plasma, their details of mode evolutions characteristics as well as the subsequently observed minor, major disruptions supported with the physical explanations. This study will help developing disruption mitigation and avoidance scenarios for having better confinement plasma experiments.

Electrical testing and performance evaluation of 1:1 prototype JET ELM control coils

Magnet Technology Development Division at Institute for Plasma research is engaged in extensive R & D for appropriate technologies towards manufacturing of Edge Localized Mode (ELM) magnets for Large Tokamak such as Joint European Torus (JET) as well as for Steady State Superconducting Tokamak (SST-1). Under this project, manufacturing of 1:1 prototype of Large and Small JET ELM control coils (CC) is completed incorporating indigenously developed manufacturing and insulation technologies. Performance evaluation of both the types of coil has been completed for its current carrying capability and insulation resistance as required by various operational scenarios of JET. Experimental setups, test procedures and measurements for electrical characterization of both type of ELM control coil has been discussed in this paper.

Vessel eddy current characteristics in SST-1 tokamak

Eddy current distribution in the vacuum vessel of the Steady state superconducting (SST-1) tokamak has been determined from the experimental data obtained using an array of internal voltage loops (flux loop) installed inside the vacuum vessel. A simple circuit model has been employed. The model takes into account the geometric and constructional features of SST-1 vacuum vessel. SST-1 vacuum vessel is a modified ‘D’ shaped vessel having major axis of 1.285m and minor axis of 0.81m and has been manufactured from non-magnetic stainless steel. The Plasma facing components installed inside the vacuum vessel are graphite blocks mounted on Copper Chromium Zirconium (CuCrZr) heat sink plates on inconel supports. During discharge of the central solenoid, eddy currents get generated in the vacuum vessel and passive supports on it. These eddy currents influence the early magnetic NULL dynamics and plasma break-down and start-up characteristics. The computed results obtained from the model have been benchmarked against experimental data obtained in large number of SST-1 plasma shots. The results are in good agreement. Once bench marked, the calculated eddy current based on flux loop signal and circuit equation model has been extended to the reconstruction of the overall B- field contours of SST-1 tokamak in the vessel region. A comparison of the field lines with and without the plasma column in identical conditions of the central solenoid and equilibrium field profiles has also been done with an aim to quantify the diagnostics responses in vacuum shots.

Overview of data acquisition and central control system of steady state superconducting Tokamak (SST-1)

Steady State Superconducting Tokamak (SST-1) has been commissioned successfully and has been carrying out limiter assisted ohmic plasma experiments since the beginning of 2014 achieving a maximum plasma current of 75kA at a central field of 1.5T and the plasma duration ∼500ms. In near future, SST-1 looks forward to carrying out elongated plasma experiments and stretching plasma pulses beyond 1s. The data acquisition and central control system (CCS) for SST-1 are distributed, modular, hierarchical and scalable in nature The CCS has been indigenously designed, developed, implemented, tested and validated for the operation of SST-1. The CCS has been built using well proven technologies like Redhat Linux, vxWorks RTOS for deterministic control, FPGA based hardware implementation, Ethernet, fiber optics backbone for network, DSP for real-time computation & Reflective memory for high-speed data transfer etc.CCS in SST-1 controls & monitors various heterogeneous SST-1 subsystems dispersed in the same campus. The CCS consists of machine control system, basic plasma control system, GPS time synchronization system, storage area network (SAN) for centralize data storage, SST-1 networking system, real-time networks, SST-1 control room infrastructure and many other supportive systems. Machine Control System (MCS) is a multithreaded event driven system running on Linux based servers, where each thread of the software communicates to a unique subsystem for monitoring and control from SST-1 central control room through network programming. The CCS hardware and software architecture is capable to fulfill the present operation and control requirement of SST-1. The CCS is successfully validated in several operation campaigns of SST-1 since year 2013.The lossless PXI based data acquisition of SST-1 is capable of acquiring around 130 channels of the ranging from 10KHz to 1MHz sampling frequency and is capable to acquire the volume of 500Mbytes of data in each shot. The indigenously developed Matlab based software utility is being used to analyze the data. The complete system is also validated in several operation campaigns of SST-1 since year 2013.This paper will provide the overview of all the above mentioned subsystems of SST-1 DAQ and SST-1 CCS focusing on the design, architecture, performance, lesson learned and future upgrade plans.

Design of mass flow rate measurement system for SST-1 superconducting magnet system

Superconducting Magnet System (SCMS) of Steady State Superconducting Tokamak – 1 (SST-1) is forced-flow cooled by a closed cycle 1.3kW (at 4.5K) class Helium Refrigerator cum Liquefier (HRL) system. An accurate measurement of helium mass flow rate in different coils is required to ensure the uniform cooling of the cold mass in the entire range of operating temperature (300K to 4.5K) and pressure (0.9–0.4MPa). To meet this requirement, indigenously designed and fabricated venturi meters are installed on 27 different coils of SST-1 SCMS. A VME based Data Acquisition System (DAS) has been developed and used to acquire the flow measurement data from different flowmeters. The details of the design of venturi meter, its different measurement and signal conditioning components, the data acquisition system and the mass flow rate calculation method are described in this paper. The mass flow rate measurement data from cryogenic acceptance and SST-1 magnet commissioning experiments are also presented and discussed in this paper.

Upgradation in SCADA and PLC of existing LN2 control system for SST-1

Helium Refrigerator/Liquefier system of Steady State Superconducting Tokamak (SST-1) incorporates Liquid Nitrogen (LN2) pre-cooling system. LN2 is used for 80K thermal shields of SST-1, current feeder system and integrated flow distribution and control system. The LN2 management system is distributed system and requires automatic control. Initially LN2 control system had Citect based Supervisory Control and Data Acquisition (SCADA) and Koyo make Programmable Logic Controller (PLC). With the passage of time and due to unavailability of their hardware, it is being obsoleted. So, the requirements of new PLC and SCADA systems have been envisaged to make uninterruptable operation of SST-1 cryogenic system. Therefore, Wonderware SCADA and Schneider Electric make PLC is programmed to replace Citect SCADA and Koyo PLC. New control features have been added in upgraded control system for better management of LN2 system. This upgradation of SCADA and PLC is completed, tested successfully and in operation. Operational performance highlights of the new upgraded system are presented in this paper.