High-temperature technical superconductors are potential candidates for compact and high-field tokamak magnets. The demand for higher fusion power can be met with an on-axis high magnetic field due to toroidal magnets. An R D activity has been initiated at the Institute for Plasma Research, India, to develop a compact D-shaped superconducting magnet utilizing REBCO high-temperature superconducting tapes. Under this initiative, a toroidal configuration with a major radius of 0.42 m, consisting of eight D-shaped, four poloidal field, and a central solenoid high-temperature superconducting magnets producing an on-axis toroidal magnetic field of 0.23 T has been conceptualized. The fabrication feasibility of a D-shaped coil for this toroidal configuration also envisaged using stacked high-temperature superconducting cable. In this paper, we report the design of a compact D-shaped coil, the fabrication of a long length HTS cable, a winding pack, and its integration with a cryogenic casing and vacuum enclosure. The winding pack terminations, joints, its interfacing with the power supply, and performance testing are also reported in this paper.

The electrical joints are very important parts for any kind of superconducting magnet like solenoid magnet, hybrid magnet and superconducting bus-bar etc. Joints preparation is very crucial for superconducting magnets because of the local heating from resistive losses can drive the superconductor into unstable condition of the magnet and as a result the superconducting magnet get premature quench. In this paper a joint has been developed in shake-hand type. The joint has been made by using the two different type of cable in conduit conductor (CICC) i.e. Nb3Sn and NbTi CICC. The 15×15 mm2 Nb3Sn and 14.8×14.8 mm2 NbTi CICCs of length ~ 200 mm have been used for preparing superconducting joint. The joint length is approximately 110 mm and enclosed with copper jacket. Lead-Tin (Pb 60% & Tin 40%) solder material having melting point 188º C, is used for the preparation of the joint. The four probe method was used for the joint resistance measurement. The measured joint resistance is ~2 n?. The main aim of this type of joint fabrication is the testing of Nb3Sn based CICC solenoid magnets. In this paper we have discussed the joint fabrication procedure, & its electrical characterization with different wattage of heating pulse near liquid helium (LHe) temperature.

A case for compact gross electricity producing pilot plant is presented. The feasibility of such a plant with a moderate fusion power that is capable of delivering gross electricity to the grid is investigated. The physics and engineering considerations of such power plants are elucidated. We show that for a fusion power of about 300 MW with fusion gain of 5, a moderate plasma beta with improved confinement regime is required to prevent excessive transport power loss. The sensitivity analysis indicates a wide enough parameter range where, the fusion power and fusion gain can meet their target values. The constraints arising from the shielding, magnets and maintenance are discussed. The feasibility of steady-state gross electricity production of 160 MW is discussed using a helium-cooled solid breeder blanket with an intermediate energy storage system. It is argued that such a plant has all key technical elements of DEMO, albeit at a smaller scale, thereby providing strong technical basis for DEMO.

Gas helium cooled HTS/MgB 2 based hybrid superconducting current feeders system have potential benefits in terms of better temperature margin and cryo stability for superconducting tokamaks and accelerators. As a proof of principle, a hybrid current feeders system comprising of a 3.3 kA rated prototype HTS current leads (CL) pair equipped with a meter long in-house developed MgB 2 shunt is recently tested at the Institute for Plasma Research (IPR). The prototype HTS CL and its bottom MgB 2 shunt is tested up to the maximum rated current of 3.3 kA with a ramp rate up to 300 A/s. The heat exchanger (HX) part is cooled using liquid nitrogen (LN 2 ). The HTS module utilizes BSCCO-2223 tapes. The bottom MgB 2 shunt section is cooled near 5 K–10 K using cold helium gas from a liquid helium Dewar. Prior to helium cooling, HX is pre-cooled to 80 K by utilizing LN 2 that allows HTS module and bottom MgB 2 shunt section conduction cooing thus optimizing and saving helium consumption. The scarcity of liquid helium in the market makes HTS/MgB 2 based superconducting current feeders a prominent candidate for large scale SC magnets application. The integrated test set-up is equipped with necessary sensors and diagnostics for detailed characterization. The temperature distribution at various sections of CL are reported in detail under cold and charging conditions. Here, we report the detailed experimental characterization results for HTS CL equipped with MgB 2 shunt under cold conditions. The main outcomes from the experiment are presented and discussed in detail.

Electron cyclotron resonance heating (ECRH) two-pulse experiments are carried out on the tokamaks SST-1 and Aditya-U using single 42-GHz gyrotron. Initially, the system was used to carry out either breakdown or heating. A new anode modulator power supply with fast rise time and fall time (1 ms) has been integrated with 42-GHz gyrotron system, which facilitate to switch the gyrotron for more than one pulse within plasma shot. The first pulse is used for the plasma breakdown at low-loop voltage and second pulse is used for plasma heating. The power in the first pulse is maintained low (less than 150 kW) for the breakdown at fundamental harmonic, while power in the second pulse is more than 200 kW for plasma heating. In both the tokamaks, SST-1 and Aditya-U, two pulse experiments have been carried out and heating effect is observed clearly in Aditya-U tokamak during second ECRH pulse. This article discusses about the two ECRH pulse experiments on both the tokamaks SST-1 and Aditya-U.

The Liquid Lead lithium Magneto Hydro Dynamics (LLMHD) experimental facility has been constructed at Institute for Plasma Research (IPR), Gujarat, India to perform various R & D MHD experiments associated with the flow of electrically conducting liquid metal under strong transverse magnetic field. The electromagnet having C-shaped soft iron core has been designed and developed, to provide a uniform magnetic field of up to 1.4T within its polar volume 1000 mm (H) x400 mm (W) x370 mm (L). The magnetic field lines are aligned along the length (L). A relatively large polar volume inside the electromagnet to place the test mock up for MHD experiments is its particularity. It enables the study of MHD flows with complex flow geometries and having longer flow length perpendicular to the magnetic field. We have started running the LLMHD loop and the first MHD experiments with Pb-Li have been performed so far at 320 degrees C in a test mock-up of a basic circular flow geometry having two 90 degrees bends. So far till now, the isothermal MHD experiments have been conducted in the presence of a uniform transverse magnetic field of 0.62T (Ha 322) and 1.06T (Ha 551) for the ranges of Reynolds number 20,000-50,000. During the MHD experiments, flow rates, temperature, pressure, and induced wall electric potential have been recorded. The MHD effects on the pressure drop and flow rate has been noticed. The 3D MHD numerical simulation has also been performed, using add on MHD module of ANSYS FLUENT. Both simulation and experimental results of the induced wall electric potential have been compared.

Conventional vapor cooled current leads (CL) based on low-temperature superconductor are the major consumer of liquid helium in large-scale superconducting magnets-based nuclear fusion devices. Use of conduction-cooled, high-temperature superconductors (HTS) CL enable operation up to 80 K and saves significant cold capacity of cryo plant. Recently at Institute for Plasma Research, India, a 3.3 kA rated prototype HTS CL pair is developed with the help of Indian industries. It exploits various superconducting materials operating from 5 to 80 K. HTS module is based on commercially available rare earth cuprate oxide superconductor based BSCCO-2223 tapes. For testing purpose, we have developed bottom lap joints using composite magnesium diboride (MgB 2 ) wires as an intermediate between HTS module and NbTi: Cu cable that act as a shunt between CL pair. Such a hybrid concept is adopted as the next step to developing MgB 2 superconducting current feeder which could be operated near 20 K and save cryogenic cost in future fusion machines. Here, we report the salient features, basic details of HTS CL, superconducting joints fabrication, its test set-up, and experimental results. The inference from test results is then reported and discussed in detail.

Plasma technologies for metallurgical applications are increasingly being adopted as they have the advantages of the unique properties of plasma. Amongst the major development priorities in many metallurgical operations and the development of new materials, plasma promises improved process control, direct utilization, improved environmental compliance and increased efficiency of energy. As a result, plasma is used for a wide spectrum of applications in materials processing like waste destruction, plasma spraying, plasma cutting, plasma welding, plasma synthesis of nanopowders, iron and steel making and extractive metallurgy for recovery of precious materials. Plasma has also recently been used as a source for the future fusion reactor, where it is confined by using the magnetic field inside the doughnut-shaped vacuum vessel. The fusion reactor which is being developed is seen as a promising, clean source of energy to solve the world's energy problem in the future as it does not use radioactive materials or pollute the air by carbon emissions. There are many challenging issues related to material constraints and plasma wall interactions which need attention as the operating environment of a fusion reactor imposes radiation damage effects. Detailed metallurgical investigation studies to select the most appropriate material are being done by many researchers. Moreover, the extremely harsh loading conditions in the fusion reactor can only be met with very diligent component design and careful selection of the best-suited material; fabricating and manufacturing techniques and efforts on developing new techniques are being carried out globally by many researchers. In this contributory article, a few examples of plasma-based technologies developed at Institute for Plasma Research (IPR) for industrial applications are described in brief. IPR has also developed, materials, technologies for blanket and fusion reactors, high temperature superconductors (HTS) for fusion magnets and they are elucidated in the below sections.

Magnet System Division of IPR is working on the development of high-temperature superconducting (HTS) magnets, test facilities, and associated technologies for plasma experiments. Under this initiative, the development of a cryostat with a room-temperature bore of ~50 mm for an HTS solenoid magnet capable of producing >0.2 T at 64 K has been carried out. The integrated system consists of an outer vacuum chamber, inner cryogenic chamber, HTS solenoid magnet, and top flanges with cryogenics, electrical, and instrumentation service ports, required for its operation has been validated up to 55 K by sub-cooling of LN2. Integrated RTBC setup is demonstrated for the steady state operation of the HTS magnet up to ~ 3 Hrs. The design, development, manufacturing, assembly, integration, and validation tests of this integrated system of RTBC will be discussed in this paper.

High Temperature Superconductor (HTS) magnets are potential candidates for the future fusion reactors and electrical industries due to its compact size and operational economy. A stainless steel (SS) laminated Bi-2223 HTS solenoid coil was fabricated and tested in liquid nitrogen. The coil is a stacked pair of double-pancakes with a soldered bridge joint. This coil was charged up to 2.1 kA with maximum current ramp rate per turn of 8.5 MA/s and generated axial magnetic field of 1.1 T at 77 K, self-field. The estimated axial magnetic field ramp rate of this coil is greater than 4 kT/s. This coil has inter-turn Kapton insulation. In order to study the effect of inter-turn electrical insulation on the current ramp rate, double pancake solenoid coils with and without inter-turn Kapton insulation of similar dimensions have also been fabricated and tested up to 10 K with a cryocooler. The differences in the current and voltage profiles during current ramp up and ramp down were observed for coils with and without electrical insulation. At an operating current of 440 A, and temperature of 10 K, the coils produced a 0.2 Tesla magnetic field. The inter-double pancake joint resistance of one of the coils with overlap length of 100 mm is measured around 58 nΩ at 10 K, self-field. The first coil was operated in pulsed mode for one millisecond without any thermal damage at 77 K. I-V characteristics for all three coils, joint resistance, and axial magnetic field measurement results and analysis will be reported in this paper.

In the present work, we have irradiated the DI-BSCCO superconducting tapes with the 100 keV deuterium ions to investigate the effect of ion irradiation on their critical current (Ic). The damage simulations are carried out using the binary collision approximation method to get the spatial distribution and depth profile of the damage events in the high temperature superconducting (HTS) tape. The point defects are formed near the surface of the HTS tape. These point defects change the vortex profile in the superconducting tape. Due to the long-range interaction of vortices with each other, the Ic of the tape degrades at the 77 K and self magnetic field. The radiation dose of 2.90 MGy degrades the 44% critical current of the tape. The results of the displacement per atom (dpa) and dose deposited by the deuterium ions are used to fit an empirical relation for predicting the degradation of the Ic of the tape. We include the dpa, dose and columnar defect terms produced by the incident particles in the empirical relation. The fitted empirical relation predicts that light ion irradiation degrades the Ic in the DI-BSCCO tape at the self field. This empirical relation can also be used in neutron irradiation to predict the lifetime of the DI-BSCCO tape. The change in the Ic of the DI-BSCCO tape due to deuterium irradiation is compared with the other second-generation HTS tape irradiated with energetic radiation.

Low resistance, helium leak tight, cryo-stable superconducting (SC) joints with sufficient strength to absorb thermo-mechanical stress are commonly used in large sized magnets, their feeders and current leads (CL). Among high temperature superconductor (HTS), cuprate oxide based tapes are the popular options for CL applications. Recently composite magnesium diboride (MgB2) wires with superconducting transition temperature of 39 K are getting wide attention for low field applications. Use of Monel as a sheath material in MgBv wires poses a major challenge while fabricating SC joints due to its low wettability and flowability using PbSn alloy based solder material. To tackle this issue, we exploited Cu electroplating technique to ease fabrication of MgB2 - NbTi: Cu as well as HTS: Cu -MgB2 joints. The details of in-house electroplating technique developed to fabricate bottom joints of BSCCO-2223 based HTS CL using MgB2 wires are reported. To study the morphology of such Cu coated MgB2 wires microstructural investigation is performed using Scanning Electron Microscope with energy dispersive X-ray spectroscopy (SEM/EDX) analysis. Similar Cu plated MgB2 wires are used to fabricate bottom joints of a prototype HTS CL. The performance of joints is preliminary attempted in the HTS CL test experiment up to 1.5 kA at 20 K.

The steadystate superconducting tokamak (SST1) is aimed to demonstrate long pulse plasma discharges employing non-inductive current drive by means of lower hybrid current drive (LHCD) system. The major and minor radius of the machine is 1.1 m and 0.2 m, respectively. The LHCD system for SST1 comprises of klystrons, each rated for 0.5 MW-CW rf power at a frequency of 3.7 GHz. The grill antenna comprises of two rows, each row accommodating 32 waveguide elements. Electron cyclotron resonance breakdown assisted Ohmic plasma is formed in SST1 to overcome the issues associated with low loop voltage start-ups. With recent modifications in the poloidal coils configuration, even with narrow EC pulse (∼50 ms), good repeatable and consistent Ohmic plasmas could be produced which helped in carrying out LHCD current drive experiments on SST1. These experiments demonstrated both fully as well as partially driven non-inductive plasma current in SST1 tokamak. Discharges with zero loop voltages were obtained. The interaction of lower hybrid waves with plasma and generation of suprathermal electrons could be established using energy spectra measured by CdTe detectors. Various other signatures like drop in loop voltages, negative loop voltages, spikes in hard x-rays and increase in second harmonic ECE signal, further confirmed the current drive by LHW’s. The beneficial effect of LHW’s in suppressing hard x-rays was also demonstrated in these experiments. The longest discharge of ∼650 ms could be obtained in SST1 with the help of LHW’s. In this paper, the experimental results obtained with LHCD experiments on SST1 is reported and discussed in more details.

Conventional copper current leads (CLs) for superconducting magnets of SST-1 are fabricated to operate in dc condition for Toroidal field coils and fast current ramp up to 10 kA for Poloidal field (PF) coils at low temperature under vacuum. Two pairs of CLs are required to be installed with PF-3 coil terminals in current feeder system chamber (CFSC) for shaped plasma operation in SST-1.The PF coils of SST-1 are subjected to the induced voltages of around 1 kV due to Ohmic coil discharge. In past few campaigns, high induced voltages on PF coils due to Ohmic coil discharge have caused arcing inside the CFSC between high potential PF CLs and grounded thermal shield. This arcing resulted in severe damage to the PF current leads and helium hydraulic lines. We report the development and implementation of Paschen leak tight electrical insulation for cylindrical CLs with variations in diameter and geometry along its length. The developmental procedures include the sample preparations and optimization of insulation processes. The scales down version of CLs samples were prepared after the investigation of curing kinetics of insulation resin at 140 °C and 110 °C. The high voltage testing of these insulation samples was carried out up to 28 kV at room temperature (RT) before and after multiple thermal shocks. These insulation samples were also tested up to 5 kV dc and 80 K temperature in Paschen test setup. The insulation resistance and breakdown voltage of these samples found to be more than 100 GΩ and 28 kV, respectively at RT.

-This paper develops a light emitting diode (LED) driver circuit based on buck-boost converter including parasitic resistance. The transfer function analysis of buck boost converter and whole system has been obtained by considering parasitic resistance of the component because the resistance of an LED varies with temperature, making the circuit unstable. The forward bias of the LED used was 3.0V~4.5V, and the forward current was 0.6A. Ten 3W whitelight LEDs were driven in series in the proposed circuit. In an LED only 15% to 25% of electrical energy is converted into light; the rest is converted into heat, which increases its temperature. In some typical application like designing of lighting system for underground coalmines this change in resistance can cause hazardous effect. Hence, LED must be operated at constant current for constant illumination for at least 10-12 hours without failure. Therefore, a closed-loop control system has been designed in this paper to increase output voltage stability. The driver circuit is operated in continuous buck-boost mode and the results are then simulated using PSIM software. The transfer function analysis has been done by using averaging method. For, stability analysis MATLAB and PSIM software has been used and it has been observed that phase margin is 93.8.o

Steady State Tokamak-2 (SST-2) will be an intermediate fusion machine before Indian DEMOnstration power reactor (DEMO) development to realise the reactor technologies. It is designed for fusion gain \(Q=5\) and fusion power in the range of 100–300 MW. Nuclear design analyses of SST-2 machine have been carried out to support the conceptual design work. Analyses have been carried out for two breeding blanket concepts: Indian lead–lithium ceramic breeder (LLCB) and helium-cooled ceramic breeder (HCCB). The analyses assess the tritium production and radiation shielding capability of the machine referring to the engineering design parameters. In this study, one-dimensional radiation transport calculations have been performed to assess the SST-2 nuclear responses for 1 full power year (FPY) operation. Nuclear responses such as tritium breeding ratio (TBR), various radiation loads to toroidal field (TF) coil have been calculated to obtain the radial build-up of SST-2 capable of breeding tritium and satisfying the shielding requirements. The assessment has been made using the ANISEN code and FENDL 2.1 cross-section library. It is observed that the TBRs with LLCB and HCCB blankets are 0.85 and 0.94, respectively. Shielding calculations confirm that the radial build is sufficient to protect the superconducting TF coils for 1 FPY.

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.

Recent experiments of SST-1 have shown that cryogenic heat loads are more than installed cold capacity. Due to this fact, the system cool down goes into the status-quo in temperatures at ~12K and further cool down is not possible. This issue can be resolved in three folds i.e. by grouping and distribution of the PF coils, optimization of the cryogenic plant process and heat loads reduction. First, we replaced common PF coils distribution to three groups having equal path lengths. Secondly, providing the best possible pressure heads to each PF groups during cool down and using turbine-C to get cold capacity with active cooling of all paths. Third is the heat loads reduction at some parts of SST-1. While adopting the same, it has been shown that the simultaneous cool down of the TF and PF coils are possible while achieving superconducting transition in all the coils except PF-5(lower).