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APFA 2015
10th Asia Plasma & Fusion Association Conference
Monday, 14 December 2015 – Friday, 18 December 2015
Hosted by
Institute for Plasma Research
Gandhinagar, India
Book of Abstracts
10th Asia Plasma & Fusion Association Conference
Codes for respective disciplines
CODE Title
0 Magnetic and Inertial Confinement
Fusion Engineering and Technology including Reactor 1 Design and Materials
2 Basic Plasma Science
3 Plasma Theory, Modelling and Numerical Simulation
4 ITER related activities
5 Industrial Plasma Applications
PBN: Poster Board Number
X_XXX: X-Code, XXX-Abstract ID
10th Asia Plasma & Fusion Association Conference
Table of Contents
Progress of ITER and the Way Forward ...... 11 Japanese Strategy of Fusion Roadmap ...... 13 Fusion Roadmap in Korea ...... 14 ITER Implementation and Fusion Energy Research in China ...... 16 Fusion Research in India ...... 17 Status of ITER Project Activities in KO-DA ...... 18 ITER India R & D and ITER Package Progress ...... 19 Data Handling System for SST-1 ...... 21 The Role of Equilibrium Flows in Temperature-Gradient-Driven Modes in Hot Tokamaks ...... 22 3D Character of Plasma Transport in the Aditya Limiter Scrape-off Layer ...... 23 Fast Visible Imaging and Study of Edge Turbulence in the Aditya Tokamak ...... 24 An Overview of Experimental ICRF Research on NSTX-U ...... 25 Status of A3 Foresight Collaboration among China, Japan and Korea on Critical Physics Issues Specific to Steady State Sustainment of High-Performance Plasmas ...... 26 Comparison of Different Atomic Databases used for Evaluating Transport Coefficients in Aditya Tokamak ...... 27 Neutral Particle Profiles during ICRH Experiments in Aditya Tokamak ...... 28 Understanding of Impurity Behavior in SST-1 Plasmas Using Visible Spectroscopy ...... 29 Observation of Plasma Shift in SST-1 using Optical Imaging Diagnostics ...... 30 Estimation of Spectrally Resolved Total Radiation Power loss in Aditya Tokamak and its Comparison with Experimental Measurements ...... 31 Ponderomotive Density Modulation in Two Ion Tokamak Plasma ...... 32 Study of Neutral Particle Transport in Aditya Tokamak Plasma using DEGAS2 Code...... 33 Modeling of Eddy Current distribution and Equilibrium Reconstruction in the SST-1 Tokamak...... 34 Equilibrium Reconstruction of Plasma Discharges in the Aditya Tokamak ...... 35 Ohmic Discharges with Improved Confinement in Tokamak Aditya...... 36 Investigation of Aditya Tokamak Plasmas with Lithiumized Wall ...... 37 A Study of Anomalous Transportation of Sawtooth Generated Runaway Electrons Observed in ADITYA Tokamak ...... 38 Geodesic Acoustic Modes with Poloidal Mode Coupling ad Infinitum ...... 39
Mean EB Shear Effect on Geodesic Acoustic Modes in Tokamaks ...... 40 Estimation of Vacuum Magnetic Fields due to Ohmic Coils in Aditya Upgrade tokamak ...... 41 Divertor Coil Power Supply in Aditya Tokamak for improved Plasma Operation ...... 42
The First Results of Te Measurement with of Soft X-Ray Diagnostics in SST-1 Tokamak ...... 43 An Overview of SST-1 Diagnostics and Results from Recent Campaigns ...... 44
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Design and Development of AXUV-based Soft X-Ray Diagnostic Camera for ADITYA Tokamak ...... 45 Conceptual Design of Diagnostics for the HL-2M Tokamak ...... 46 Observation of MHD Phenomenon for SST-1 Superconducting Tokamak ...... 47
The Determination of Plasma Radial Shafranov Shift (R) and Vertical Shift (Z) Experimentally using Magnetic Probe and Flux Loop Method for SST-1 Tokamak ...... 48 Development of New Diagnostics for WEST ...... 49 Observation on Runaway Discharges in SST-1 Experiments ...... 50 Hard X-ray Diagnostic for SST-1 ...... 51 Study of MHD Activities in the Plasma of SST-1 ...... 52 A Fixed Frequency Reflectometer to Measure Density Fluctuations at Aditya Tokamak...... 53 Helium Beam Diagnostics for the Estimation Electron Temperature and Density in SST-1 ...... 54 Operation of ADITYA Thomson Scattering System: Measurement of Temperature and Density...... 55 Installation and Commissioning of SST-1 Thomson scattering system ...... 56 Limiter and Divertor Systems – Conceptual and Mechanical Design for Aditya Tokamak Upgrade ...... 57 Development of Gas Puffing System for LHCD Experiment in Aditya Tokamak ...... 58 Structural Analysis of New Vacuum Vessel for Aditya Tokamak Upgrade ...... 59 IGBT Based Active Clamping Protection Scheme for SST-1 PF Coils ...... 60 Thermal Imaging of SST-1 Limiters ...... 61 The Upgradation of Aditya Tokamak ...... 62 Development of Non-circular Metal Seal for Aditya Tokamak Upgrade Vacuum Vessel ...... 64 Study of the plasma SOL with fast reciprocating probe diagnostics on the SST-1 tokamak ...... 65 Conceptual design of Plasma position control of SST-1 Tokamak using vertical field coil ...... 66 Implementation of SST-1 plasma position control using vertical field ...... 67 Preparation of W/CuCrZr Monoblock Test Mock-up using Vacuum Brazing Technique ...... 68 Design and Performance of Vacuum System for High Heat Flux Test Facility ...... 69 Thermal Shock Behavior of Tungsten & Tungsten Alloy Materials under Transient High Heat Load Conditions .... 70 Characterization of a Segmented Plasma Torch Assisted High Heat Flux (HHF) System for Performance Evaluation of Plasma Facing Components in Fusion Devices ...... 71 Performance of Impedance Transformer for High Power ICRF Heating in LHD ...... 72 Progress of JT-60SA Construction and R&D of its Heating Systems ...... 74 Optimization, Commissioning and Operation of EAST Tungsten Divertor ...... 75 Status of the WEST Project ...... 76 Recent Advancement in Research and Planning toward High Beta Steady State Operation in KSTAR...... 77 Progress of Experiment on HL-2A ...... 79 Recent Progress and Present Status of LHD towards Deuterium Experiment ...... 80 Initial Results in SST-1 After Up-gradation ...... 81 WEST Physics Basis ...... 82 Indigenously Developed Large Pumping Speed Cryoadsorption Cryopump ...... 84
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Indian Single Pellet injection System for Plasma Fuelling Studies ...... 85 Development of Heat Sink Concept for Near-term Fusion Power Plant Divertor ...... 86 Characterization of Discharge Plasma in Cylindrical IECF Device ...... 88 Serial Interface through Stream Protocol on EPICS Platform for Distributed Control and Monitoring ...... 89 Development of Data Acquisition Set-up for Steady-state Experiments ...... 90 Prototyping of Radial Plates for Fusion Relevant Superconducting Magnets ...... 91 Application of Articulated Absolute Co-ordinate Measuring Machine for Quality Control in Manufacturing of ELM Control Coil ...... 92 Indigenously Developed Bending Strain Setup for I-V Characterization of Superconducting Tapes and Wires ...... 93 RF Assisted Glow Discharge Condition Experiment in SST-1 Tokamak ...... 94 Commissioning and Experimental Validation of SST-1 Plasma Facing Components ...... 95 Baking and Helium Glow Discharge Cleaning of SST-1 Tokamak with Graphite Plasma Facing Components ...... 96 Design and Integration of SMBI System for SST 1 ...... 97 Neutron Measurements from Beam-Target Interactions with Deuterium Ion Beam ...... 98 Electron Beam Welding: Study of Process Capabilities and Limitations towards Development of Nuclear Components ...... 99 Thermal Response of Actively Cooled Tungsten Monoblock during Inhomogeneous Surface Heat Loads ...... 100 Consistency Checks in Beam Emission Modeling for Neutral Beam Injectors ...... 102 Computational Fluid Dynamics Analysis of Heat Transfer Elements for SST-1 Neutral Beam Line ...... 103
Er2O3 Coating Development and Improvisation by Metal Oxide Decomposition Method ...... 104 Design of CPLD-DAC Based Probe Bias Generator and Current Measurement Electronics ...... 105 Nanoscale Coatings of Tungsten by Radio Frequency Plasma Assisted Chemical Vapor Deposition on Graphite .. 106 Multi-scale Modeling of Neutron Induced Radiation Damage in Tungsten ...... 107 Role of ECRH in SST-1 Tokamak Plasma ...... 109 Design of 1 MHz Solid State High Frequency Power Supply ...... 110 Neutron Induced Reaction for Long-lived Isotopes Produced in Fusion Materials ...... 111 Development of a Neutronics Facility using RFQ Accelerator as the Basic Tool ...... 112 Design of a Prototype Positive Ion Source with Slit Aperture Type Extraction System ...... 113 Optimization of Geometrical Parameters for High Heat Flux Components (Vapotrons) ...... 114 Design and Development of CRIO Based Data Acquisition and Control System for High Voltage Bushing Experiment ...... 115 Rotor-dynamic Design Aspects for a Variable Frequency Drive Based High Speed Cryogenic Centrifugal Pump in Fusion Devices ...... 116 Quench Detection, Protection and Simulation Studies on SST-1 Magnets ...... 117 Gas Fueling System for SST-1 ...... 118 Development of Electromagnetic Welding Facility of Flat Plates for Nuclear Industry ...... 119 Engineering Design & Integration of Radial Control Coil in Vacuum Vessel of SST-1 ...... 120 Engineering Design & Integration of In-vessel Single Turn Segmental Coil in Vacuum Vessel of SST-1 ...... 121
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Quality Control of FWC during Assembly/Commissioning on SST-1 ...... 122 Laser Shock Peening of Stainless Steel Surfaces: ns vis-ã-vis ps Laser Pulses ...... 123 Assembly & Metrology of First Wall Components of SST-1 ...... 124 Trap Site Formation and their Distribution Studies in Porous Lithium Titanate ...... 125 Design of a High Power Water Load for LHCD System of SST-1 Tokamak ...... 126 Design of Multiple Ferrite Tile Phase Shifters for Applications in High CW Power Differential Phase Shift Circulators ...... 127 Conceptual Design of PAM Antenna for Aditya-U Tokamak ...... 128 Assessment of Delta Ferrite in Multipass TIG Welds of 40 mm Thick SS 316L Plates: A Comparative Study of Ferrite Number (FN) Prediction and Experimental Measurements ...... 129 Study of Transients in Liquid Helium Flow during Cool Down of Cryopanel ...... 130 A Simple In-vessel/FW Component Viewing System for SST-1 ...... 131 Overall Behaviour of PFC Integrated SST-1 Vacuum System ...... 132 Assembly of Neutral Beam Injector with SST-1 ...... 133 Experience of 12 kA / 16 V SMPS during the HTS Current Leads Test ...... 134 Calibration of Low Temperature Measurement System for the Superconducting Magnet System for the SST-1 .... 135 Electronics for Coupled High Voltage Measurement on PF Magnets of SST-1...... 136 Electronics and Instrumentation for the SST-1 Superconducting Magnet System ...... 137 ITER and its Diagnostics- the Way Ahead ...... 139 Status of the Realization of the Neutral Beam Test Facility ...... 140 R & D of Tritium Technology for Fusion in CAEP: Progress and Prospect ...... 141 Precision Electronics and Measurement Techniques for the Superconducting Joint Resistance ...... 143 Preliminary Results from Electron Cyclotron Measurements at SST-1 ...... 144 PLATo (Power Load Analysis Tool) – A Module of WEST Wall Monitoring System ...... 145 Fabrication of Vacuum Vessel with Detachable Top Lid Configuration for Indian Test Facility (INTF) ...... 146 Measurement and Sweep-biasing Circuit for Langmuir Probe Diagnostic in SYMPLE ...... 147 Density Measurement Systems at SST Tokamak ...... 148 Software Upgradation of PXI Based Data Acquisition for Aditya Experiments ...... 149 Development, Integration and Testing of Automated Triggering Circuit for Hybrid DC Circuit Breaker ...... 150 Metrology Measurements for Aditya Tokamak Upgradation ...... 151 Study of Transport and Micro-structural Properties of Magnesium Di-Boride Strand under React and Bend Mode and Bend and React Mode ...... 152 Michelson Interferometer Diagnostics for Broadband ECE Measurement ...... 153 Assembly of Aditya Upgrade Tokamak ...... 154 The Refurbishment of Damaged Toroidal Magnetic Field coils for Aditya Upgrade ...... 155 Conceptual Design of Dump Resistor for Superconducting CS of SST-1 ...... 156 Safety and Environment Aspects of Tokamak-type Fusion Power Reactor - An Overview ...... 157 Fusion Blanket Materials Development and Recent R&D Activities ...... 158
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Electrical Properties of Nano Li2TiO3 for Fusion Reactors ...... 159 Design of New Superconducting Central Solenoid of SST-1 Tokamak ...... 160 Design of High Resolution Spectroscopic Diagnostics for SST-1 and ADITYA-U Tokamak ...... 161 Conceptual & Engineering Design of Plug-in Cryostat Cylinder for Superconducting Central Solenoid of SST-1 . 162 Investigation of Homoclinic Bifurcation of Plasma Fireball in a Double Plasma Device ...... 163 Determination of the Plasma Composition using Blended Stark-Broadened Emission Lines in a Self-Magnetic Pinch Diode ...... 164 Magnetic Probe Diagnostic Tool to Understand the Dynamics in a Non-transferred dc Plasma Torch ...... 165 Localized solutions in Laser Plasma Coupled System with Periodic Time Dependence ...... 166 Coupling of Drift Wave with Dust Acoustic Wave ...... 167 Resolving Issues Associated with Langmuir Probe Measurements in High Pressure Complex (Dusty) Plasmas ..... 168 On the Spatial Behavior of Background Plasma in Different Background Pressure in CPS Device ...... 169 Effect of Catalyst for the Decomposition of VOCs in a NTP Reactor ...... 171 Relativistic Cylindrical and Spherical Plasma Waves ...... 172 Observation of Early and Strong Relativistic Self-Focusing of cosh-Gaussian Laser Beam in Cold Quantum Plasma ...... 173 Electric Field Assisted Sintering (EFAST): Plasma? ...... 174 Dispersion of Linearly Polarized Electromagnetic Wave in Magnetized Quantum Plasma ...... 175 Breaking of Relativistic Electron Beam Driven Wake Waves in a Cold Plasma ...... 176 2D Turbulence Structure Observed by a Fast Framing Camera System in Linear Magnetized Device PANTA ...... 177 Production of Quiescent Collisionless Plasma over a Wide Operating Range ...... 179 Effect of Fast Drifting Electrons on Electron Temperature Measurement with a Triple Langmuir Probe ...... 180 Ponderomotive Force and Backward Raman Scattering in Dense Quantum Plasmas ...... 181 Anode Glow and Double Layer in DC Magnetron Anode Plasma ...... 182 Effect of Trapped Particle Nonlinearity in IAW Solitary Wave ...... 183 Installation of a 100 kJ Pulsed Power System to Drive Pulsed Plasma Devices ...... 184 Characterization of the Permanent Magnet Based Hydrogen Helicon Plasma Source for Ion Source Application .. 185 Investigation in Presence of External Forcing and Magnetic Field in a DC Glow Discharge Plasma and Evidence of Nonlinearity ...... 186 Radio Frequency Emissions from Plasmas due to Laser Induced Breakdown of Materials ...... 187 Effect of Transverse Magnetic Field on the Steady State Solutions of a Bursian Diode ...... 188 Wave-breaking Amplitudes of Relativistically Strong Electrostatic Waves in Cold Electron-Positron-Ion Plasmas ...... 189 Nonlinear Coherent Structures of Alfven Wave in a Collisional Plasma ...... 190 Parallel Connection Length and Flow-fluctuation Cycle in Simple Toroidal Device ...... 191 Controllable Location of Polarization Reversal in Nonuniform Helicon Plasma ...... 192 Hot Tungsten Plate Based Ionizer for Cesium Plasma in a Multi-Cusp Field Experiment ...... 193 Development of Three Dimensional Magnetic Field Probe with Signal Conditioning Electronics ...... 194
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State of Art Data Acquisition System for Large Volume Plasma Device ...... 195 Exploration of the Solar System & Beyond: The Indian Scene ...... 197 Experimental Study of Plasma Current Ramp-up by the Lower Hybrid Wave in the TST-2 Spherical Tokamak .... 199 ELM Control using Low-n RMPs in KSTAR and its Perspective to Beyond-ITER ...... 200 Development of Long Pulse Radiofrequency Heating and Current Drive Systems and Scenarios for WEST ...... 201 Behaviors of Impurity in ITER and DEMOs using BALDUR Integrated Predictive Modeling Code ...... 203 Rapid Purification of Hydrogen Isotope Gas by Palladium Alloy Membrane Separator ...... 204 Measurements and Controls Implementation for the WEST Project ...... 205 Super Rogue Wave in Plasma ...... 207 Experiment on Dust Acoustic Solitons in Strongly Coupled Dusty Plasma ...... 208 Controllable Transition from Positive Space Charge to Negative Space Charge in an Inverted Cylindrical Magnetron ...... 209 Measurement of Electron Energy Probability Function in Weakly Magnetized Plasma ...... 210 Characteristics of Dust – Density Waves in the Presence of a Floating Cylindrical Object in the DC Discharge Plasma ...... 211 Investigation of Magnetic Drift on Transport of Plasma across Magnetic Field ...... 212 High Intensity High Contrast Femtosecond Laser Absorption in Solid ...... 213 Lithium Vapor Density Diagnostics for the PWFA Plasma Source at IPR ...... 214 Turbulent, Megagauss Magnetic Fields in Intense, Ultrashort Laser Pulse Interaction with Solids ...... 215 Design and Characterization of Cesium Oven for a Multi-cusp Plasma Device ...... 216 Korteweg-de Vries-Burger (KdVB) Equation in a Five Component Cometary Plasma with Kappa Described Electrons and Ions ...... 217 Two Dimensional Imaging of Laser Produced Plasma in Magnetic field ...... 218 The Effect of Addition of Lighter Ions in a Five Component Multi-Ion Plasma ...... 219 Effect of Ablation Geometry on the Formation of Stagnation Layer in Laterally Colliding Plasmas ...... 220 Enhanced Confinement by Controlling Instability in Toroidal Electron Plasma of SMARTEX-C ...... 221 Study of Phase Space Structures in Driven 1D Vlasov Poisson Model ...... 222 Synchronization dynamics and Arnold tongues for two coupled glow discharge plasma sources ...... 223 Optical Kerr Gated Time Resolved Cherenkov Emission Produced during Ultra Intense Laser Solid Interaction ... 224 Imaging of Terahertz Emission from Intense High-Contrast Ultrashort-Pulse Laser-Solid Interaction ...... 225 Pulsed Plasma for the Study of Coherent Structure in the Electron Magnetohydrodyanamic Regime ...... 226 Chaos to Order Transitions in Chaotic Magnetic Fields ...... 227 Study of Defects in Externally Driven Dust Density Waves in Cogenerated Dusty Plasma using Time Resolved Hilbert-Huang Transform ...... 229 Efficient Hard X-ray Generation in an Interaction of Intense, Ultrashort Laser with Metal Nano-coated Dielectric Target ...... 230 Laser Heated Emissive Probe for Plasma Potential Measurement in Fusion Plasmas ...... 231 Study of Fluctuation Induced Particle Flux in the Background of ETG plasma in LVPD ...... 232
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Exhibiting Electrons in Nanoplasmas: An Estimate ...... 233 High Energy Neutral Atoms from High Intensity Laser Plasma Interaction ...... 234 Role of Magnetic Cusp for Multiple Axial Potential Structures (MAPS) Formation ...... 235 Enhanced Proton Acceleration by Ultrashort Laser Pulse Interaction with Nanostructured Thin Films ...... 236 DAQ System for Low Density Plasma Parameters Measurement ...... 237 Numerical Study of Instabilities in Magnetized Inhomogeneous Plasmas ...... 238 Modeling of Electromagnetic Fields during Plasma Startup in SST-1 Tokamak ...... 239 Oscillating Two-stream Instability of a Plasma Wave in Ion-Motion Regime ...... 240 Development of a 3D-3V PIC code to study PSI processes in Tokamak Divertor Region ...... 241 Betatron Radiation from Laser Wakefield Acceleration in a Plasma Channel ...... 242 Particle in Cell Simulations of Beam Plasma System...... 243 PIC Modeling of Negative Ion Extraction from a Dust-Seeded Plasma ...... 244 Dynamics of dusty fluid in a streaming sheared plasma ...... 245 Numerical Analysis on Bandwidth and Growth Rate of Plasma-Filled Gyrotron Devices ...... 246 Gyro-TWT in a Vane-Loaded Waveguide with Inner Dielectric ...... 247 Effect of Plasma Column on the Radial Profile of Electric Field of Gyrotron Devices ...... 248 Current Gradient Modes of Two Dimensional Electron Magnetohydrodynamics (EMHD) ...... 249 A Poynting like Theorem for Generalized Hydrodynamic Equations ...... 250 Identification of Nonlinear Resonance Absorption in a Laser Driven Deuterium Cluster using Molecular Dynamics Simulation...... 251 1D PIC simulation of relativistic Buneman instability ...... 252 Molecular Dynamics Simulation of Dust Particle Levitation in the Presence of Sheath ...... 253 Conceptual Study of High-Field LHCD in KSTAR ...... 254 Integrated Core-SOL Simulations of L-Mode Plasma in ITER and Indian DEMO ...... 255 Potential around a dust grain in collisional plasma ...... 256 Numerical simulation of a novel non-transferred arc plasma torch operating with nitrogen ...... 257 Nonlinear MHD modeling in LHD plasmas with peaked pressure profiles ...... 259 Sensitivity analysis of upstream plasma condition for SST-1 X-divertor configuration with SOLPS ...... 260 Radiation Effects on the Laser Ablative Shockwaves from Aluminum under Atmospheric Conditions ...... 261 Angular Momentum Transfer of Laguerre - Gaussian Laser Pulses and Quasi-static Magnetic Field Generation in Plasma Channels ...... 262 Real-time Horizontal Position Control for Aditya-Upgrade Tokamak ...... 263 Prediction of Temperature and Stress Distributions in Substrate and Coating during Plasma Spraying ...... 264 Design & Development of Amplitude and Phase Measurement of RF Parameter with Digital I-Q De-Modulator (DIQDM) Technique using PXI System ...... 265 Effect of Geometrical Imperfection on Buckling Failure of ITER VVPSS Tank...... 266 Nuclear Analyses of Indian LLCB Test Blanket System in ITER ...... 267
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Preferential Binding of Self-interstitial Atoms over Vacancies to Grain Boundaries of Tungsten: A Lattice Statics Study ...... 268 Alternate Design of ITER Cryostat Skirt Support System ...... 270 Neutronics Analysis, Shielding Optimization and Radiation Waste Analysis for X-Ray Crystal Spectrometer of ITER ...... 271 Preliminary Optical Design of Polarization Splitter Box for ITER ECE Diagnostic System ...... 272 Development of High Voltage and High Current Test Bed for Transmission Line Components...... 273 Development of Control System for Multi-converter High Voltage Power Supply using Programmable SoC ...... 274 Development and Validation of I-Activation Analysis Code ...... 275 Indigenous Manufacturing Realization of Twin Source and its Auxiliary System ...... 276 Wilkinson Type Lumped Element Combiner-Splitter for Indigenous Amplifier Development ...... 277 Preliminary Design Development of ITER X-ray Survey Spectrometer ...... 278 Integration & Validation of LCU with Different Sub-systems for Diacrode Based Amplifier ...... 279 Comparative Analysis on Flexibility Requirements of Typical Cryogenic Transfer Lines ...... 280 Dynamics of Cold Helium Flow inside a Cryoline used for Large Cryogenic Distribution System ...... 281 Final Configuration with Assembly Assessment of the 100kV High Voltage Bushing for the Indian Test Facility . 282 Preliminary Design of O-mode Radiometer for ITER ECE Diagnostic ...... 283 System Upgradation for Surface Mode Negative Ion Beam Extraction Experiments in ROBIN ...... 284 Thermo-mechanical Design Methodology for ITER Cryo-distribution Cold Boxes ...... 286 Preliminary Design of Bellows for the DNB Beam Source by EJMA & FE Linear Analysis ...... 287 Evolving the Inspection Techniques for determination of Volumetric Dimensions of Ground Pore in Heat Transfer Elements ...... 288 Significance of ITER IWS Material Selection and Qualification ...... 289 ITER ECE Diagnostic: Design Progress of IN-DA and its Role for Physics Study ...... 290 Manufacturing Experience of an ‘Angled’ Accelerator Grid for DNB Beam Source ...... 291 Preparation and Analysis of Helium Purge Gas Mixture to be used in Tritium Extraction System of LLCB TBM . 292 Seismic Design of ITER Component Cooling Water System-1 Piping ...... 293 Role of Outgassing of ITER Vacuum Vessel In-Wall Shielding Materials in Leak Detection of ITER Vacuum Vessel ...... 294 Manufacturing and Assembly of IWS Support Rib and Lower Bracket for ITER Vacuum Vessel ...... 295 Finite Element Analysis for ITER Ferromagnetic In-wall Shielding Block ...... 296 Development of XM-19 Fasteners for the IWS Blocks Assemblies ...... 297 Present design status of Erosion and Tritium Monitor diagnostics for ITER ...... 298 Study of Structures and Stability in Nitrogen Plasma Jet ...... 299 Pesticides Removal from Cabbage using Array of Atmospheric Pressure Plasma Jet ...... 300 Comparison of Gas and Plasma Carburizing of AISI 8620 Low Carbon Steel ...... 301 Experimental Study to Improve Anti-felting Characteristics of Merino Wool Fiber by Atmosphere Pressure Air Plasma ...... 302
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Surface Chemistry and Wettability Study of Air Plasma Treated Polyethylene by Atmospheric Pressure Dielectric Barrier Discharge ...... 303 Electrical Characteristics of a DC Non-transferrerd Arc Plasma Torch Using Theory of Dynamic Similarity ...... 304 Design and Development of 20 kW, 45 kV, 30 kHz Power Supply for Study of Pulsed Dielectric Barrier Discharges ...... 305 Plasma Sterilization for Bio-decontamination ...... 306 Superficial Layer MHD Effect and Full-cover Free Surface Flow Characterization ...... 308 Fast Wave Scrape-off Layer Losses of Tokamak Plasmas in Minority, Mid/High Harmonic, and Helicon Heating Regimes ...... 309 Manufacturing and process research of the WEST ICRH antenna ...... 310 Recent Progress of the ECRH System on HL-2A ...... 311 The 3.7GHz LHCD System on HL-2A ...... 312 Observation of Up-Down Asymmetry in Impurity Line Emissions from the Ergodic Layer of Large Helical Device ...... 314 Current Status of Safety design and Safety Analysis for China ITER Helium Coolant Ceramic Breeder Test Blanket System ...... 315 Destructive Analysis on the ITER FW Small Scale Mock-ups ...... 316 EAST Articulated Maintenance Arm for EAST and WEST ...... 317 Improvements in a Tracer-Encapsulated Solid Pellet and Its Injector for More Advanced Plasma Diagnostics ...... 318 Simulation and Modeling of Magnetic Field Dynamics in Laser Plasma Interaction ...... 320 Electrical Transverse Transport in Lorentz Plasma with Strong Magnetic Field and Collision Effect ...... 321 Spectroscopy of Laterally Colliding Plasma Plumes in Laser-blow-off of Thin Film: Role of Energetic Neutrals in Formation of Interaction Zone ...... 322 Thermionic Divertors for Tokamaks ...... 323 Modeling of ITER Disruption scenarios using TSC ...... 324 Technical Developments and Present Status of the ITER Cryolines and Cryo-distribution Systems ...... 326 Cryogenic Technology of the New Millennium – Competence of DH Industries ...... 327 Upgradation Plans of SST-1 Cryogenics System at IPR ...... 328
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Plenary Talk
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Abstract ID: 4_302
Progress of ITER and the Way Forward
Bernard Bigot1, Jean Jacquinot1
1ITER Organisation, France Email: Bernard.bigot@iter.org
The ITER project has recently benefitted from bold organizational changes which will be first discussed during this presentation. The changes focus on strengthening a common project spirit across the entire organization. This is essential for completing successfully such a challenging undertaking as ITER. All key actors, whether they are in the central team, in the DAs or in other related organizations, should feel committed to the success of the entire Project sharing a common vision and working practices. To that effect, a new organization has been put in place: - Two project teams (Buildings and Vacuum Vessel, other teams are being considered) now gather under a single leadership all components of a major critical deliverable. – An Executive Project Board (EPB), meeting twice a month, now convenes DA executives under the chairmanship of the director general (DG) who is empowered, after consultation of the EPB, to take all important technical decisions - The DG can use a recently created reserve fund for financing items or changes in the configuration which were not foreseen in procurement agreements – Finally the organigram of the central team has been simplified emphasizing project-oriented integration, building coherent technical departments and preparing for the assembly phase which will be a major undertaking.
The transitional period necessary to put in place these changes did not prevent the Project from accelerating the pace of the construction. The presentation will give examples of major achievements obtained recently. Among these an outstanding result obtained collectively is the progress in the manufacture of the superconducting cables: 90% of the needed cable-in-conduit has been produced fully complying with the technical requirements. This required a major industrial development in several Parties. India has made remarkable progress in its assigned procurements: in particular, the bottom parts of the cryostat are expected to reach the ITER site at about the time of this conference and assembly in the purpose-built hall will then start.
A major task, also undertaken by joint efforts of the central team and all the DAs, was devoted to establishing a realistic resource loaded schedule as a possible new baseline for the project. This schedule is deemed to be technically feasible assuming that the required resources will be provided. It is optimized for the earliest possible achievement of first plasma whilst preventing delays on the D/T phase by doing parallel assemblies as far as possible. The presentation will also address the physics issues which will be most important during the commissioning and the various operation phases of ITER. Many of these issues can be included in the experimental programs of existing (or about to start) experiments in Asia. As already identified in ITPA workshops, the ITER organization stresses the importance of collecting data in such subjects as disruption and ELM mitigation, H-mode threshold during the non-active phases, optimization of confinement with a tungsten divertor in a radiating mode with or without nitrogen injection, energetic particles (diagnosis and confinement) and steady state scenarios in actively cooled conditions.
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Invited Talk (Session-1)
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Abstract ID: 1_285
Japanese Strategy of Fusion Roadmap
Akio Komori1, Hiroshi Yamada1, Satoru Sakakibara1
1National Institute for Fusion Science, Japan Email: [email protected]
Technology bases required for realization of a fusion demonstration reactor (DEMO) have been discussed in the Japanese fusion community. The Join-Core Team, which was launched by ministerial council, has considered the following issues to develop strategy for the establishment of technology bases for a DEMO: (1) Concept of DEMO premised for investigation, (2) Activities requiring commitment and their goals, and (3) Scientific and technological review works for the above mentioned activities. The team summarized the issues as Joint-Core Team Report (Basic Concept of DEMO and Structure of Technological Issues) [1]. This report describes the basic concept of DEMO premised for investigation and the structure of technological issues to ensure the feasibility of the DEMO concept. Also the team clarified tasks regarding the development of the design of DEMO, and the research and development programs to resolve the issues and to provide the required evidence to support the design into the consistent timeline, which is summarized in the second Joint-Core Team Report (Chart of Establishment of Technology Bases for DEMO) [2]. The reports show that DEMO is steady power generator with several hundred thousand kilowatts being able to extend to commercialization.Also function of tritium bleeding to fulfil self-sufficiency of fuels should be equipped. Required technological activities (superconducting coils, blanket, divertor, heating and current drive systems etc.) are arranged in the chart.JAEA has established the Joint Special Design Team for Fusion DEMO in cooperation with NIFS, industry,and universities and reinforces design activity of atokamak DEMO.
Two important tasks still remain to define the roadmap of development of DEMO in future, that is, socio-economic examination of fusion energy and review of alternative approaches of helical magnetic fusion system and laser fusion system. In particular, solution of the latter task should be found under the deep discussion with Japanese fusion community.
References:
[1] H. Yamada et al., NIFS-MEMO-71 (National Institute for Fusion Science,February, 2015)
[2] H. Yamada et al., NIFS-MEMO-73 (National Institute for Fusion Science, March,2015)
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Abstract ID: 0_288
Fusion Roadmap in Korea
Keeman Kim1, Gyung-Su Lee1, Kihak Im1, Hyoung Chan Kim1, Yong-Seok Hwang1
1National Fusion Research Institute, Korea Email: [email protected]
The KSTAR (Korea Superconducting Tokamak Advanced Research) project started in 1995 as a first major step of “National Fusion Energy Development Plan” and, as a following step, Korea joined the ITER program. Korean Fusion Energy Development Promotion Law (FEDPL) was enacted in 2007 to promote a long-term cooperative fusion research and development among participating industries, universities and research institutes. And a conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) has been initiated in 2012 and “the Report on K-DEMO R&D Plan” was submitted to the Government of Korea in 2013.
One special concept of K-DEMO is a two-staged development plan. At first, K-DEMO is designed to demonstrate a net electricity generation (Qeng > 1) and a self-sustained tritium cycle (Tritium breeding ratio, TBR > 1.05), and it is also designed to be used as a component test facility. Then, at its second stage, a major upgrade is carried out by replacing in-vessel components and the net electric generation shall be on the order of 500 MWe. After a thorough 0-D system analysis, the major radius and minor radius are chosen to be 6.8 m and 2.1 m, respectively, considering practical engineering feasibilities. In order to minimize the deflection of wave and maximize the efficiency, a top launch high frequency (> 200 GHz) electron cyclotron current drive (ECCD) system is considered and, for matching the high frequency ECCD, a high magnetic field is required and the peak magnetic field can approach to 16 T with the magnetic field at the plasma center above 7 T. K-DEMO incorporates a vertical maintenance design. Pressurized water is the most prominent choice for the main coolant of K-DEMO when considering balance of plant development details. Considering the plasma performance and the peak heat flux in the divertor system, a conventional W-type double-null divertor system becomes the reference choice of K-DEMO.
The current status on the KSTAR program, ITER program and the conceptual design study of K- DEMO and the implementation plan for core technology R&D based on a gap study are presented including the Korean Fusion Energy Roadmap
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10th Asia Plasma & Fusion Association Conference
Invited Talk (Session-2)
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 4_286
ITER Implementation and Fusion Energy Research in China
Jing Zhao1, Zhaoliang Feng1, Changchun Yang1
1China International Nuclear Fusin Energy Program Execution Center, China Email: [email protected]
ITER Project is jointly implemented by China, EU, India, Japan, Korea, Russian Federation and USA, under the coordination of Center Team of ITER International Fusion Energy Organization (IO-CT). Chinese fusion research related institutes and industrial enterprises are fully involved in the implementation of China contribution to the project under the leadership of ITER China Domestic Agency (CN-DA), together with IO-CT. The progresses of Procurement Packages (PA) allocated to China and the technical issues, especially on key technology development and schedule, QA/QC issues, are highlighted in this report. The specific enterprises carrying out different PAs are identified in order to make the increasing international manufactures and producers to ITER PAs know each other well for the successful implementation of ITER project. The participation of China to the management of IO-CT is also included, mainly from the governmental aspect and staff recruited from China. On the other hand, the domestic fusion researches，including upgrade of EAST, HL-2A Tokamaks in China, TBM program, the next step design activities for fusion energy power plant, namely, CFETR and training in this area, are also introduced for global cooperation for international fusion community. Keywords: ITER, Implementation, Domestic fusion researches
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_291 Fusion Research in India
Dhiraj Bora1
1Institute for Plasma Research, India Email: [email protected]
The economic growth of our country demands a rapid increase in the energy output. Fusion is one such alternate clean source of energy to contribute in the energy mix towards the second half of the century, with a virtually inexhaustible fuel supply. The environmental impact of fusion would be acceptable and relatively safe. These advantages have driven the world fusion research programme since its inception. Till a pure fusion energy source is available, it is worthwhile to develop it for the benefit of conventional fission fuel preparation and other various usages.
Indian National Fusion Programme was initiated by indigenously developing the first Indian Tokamak, ADITYA, successfully commissioned in 1989 and has been generating interesting scientific results on various topics. The next major program at Institute for Plasma Research (IPR) has been to construct a Steady State Superconducting Tokamak (SST-1) by mix of import and indigenous development. After successful engineering validation of the subsystems in integrated operations, successful machine operation has been continued. Since then, the machine has been upgraded with a graphite first wall.
As a strategy towards leapfrogging to save time, IPR and Department of Atomic Energy (DAE) decided on India’s participation in the International Thermonuclear Experimental Reactor (ITER) as a full partner, unique features of which will be its ability to operate for long durations and at power levels ~500 MW sufficient to demonstrate the physics of burning plasma in a power plant like environment. It will also serve as a test-bed for additional fusion power plant technologies.
To accelerate the domestic fusion research programme with integration of knowledge gained from ITER, we would embark upon design of a smaller fusion machine which will use already available technologies to produce controlled fusion reactions and use it as an energetic neutron source for test of materials developed for future fusion reactors. Such a machine can also be used to accelerate utilization of Thorium in Phase-III of our Nuclear Energy Program.
Indian progress in Fusion science and technology, participation in ITER, already initiated study in gap areas along with future activities will be discussed during the talk.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 4_301
Status of ITER Project Activities in KO-DA
Kijung J Jung1
1National Fusion Research Institute, Korea Email: [email protected]
KO-DA is responsible for 11 procurement packages for ITER Project. And all of the activities for all the procurement packages have been simultaneously launched from 2007 just after procurement sharing had been agreed between Members. The first Procurement Arrangement, TF Conductors, was signed in May 2008. And KO-DA has now signed 9 PAs among 11 Packages; 93.9% of the Procurement Arrangements in kIUA value. The first delivery of the KO- DA Package, TF Conductors which is one of the most important and largest packages for KO- DA’s, has been successfully accomplished at the end of 2014. The KO-DA together with Korean industries is actually doing its best efforts to meet the delivery schedules agreed by the ITER Organization and DAs. This paper presents the current status of each procurement-packages’ activities.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 4_296
ITER India R & D and ITER Package Progress
Shishir Deshpande1
1Institute of Plasma Research, India Email: [email protected]
ITER-India is a special project of the Institute for Plasma Research (nodal agency for ITER collaboration). The mandate for ITER-India is to deliver India’s ‘in-kind’ commitments to ITER, which are defined by the nine packages: (1) Cryostat, (2) In-wall Shielding, (3) Cryodistribution & Cryolines, (4) Ion Cyclotron Heating RF-power sources (35-65 MHz) for coupling 20 MW, (5) Electron Cyclotron Heating sources for 2 MW at 170 GHz, (6) Diagnostic Neutral Beam, (7) Power Supply Systems for IC, EC and DNB, (8) Component Cooling, Chilled Water and Heat Rejection System, and (9) Diagnostics (with X-Ray, visible and microwave region) with Port Plug.
ITER has many systems, which are being built (on that scale/capacity) for the first time. A number of R&D activities are therefore needed to make sure that the systems work as expected. The talk will cover these apart from the organization of the project. Salient features of the packages and the design/manufacturing progress will also be described.
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10th Asia Plasma & Fusion Association Conference
Poster Session-1
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_3
Data Handling System for SST-1
Harish Masand1, Manisha Bhandarkar1, Aveg kumar1, Hitesh Kumar Gulati1, Kiritkumar B Patel1, Kirti Mahajan1, Jasraj Dhongde1, Hiteshkumar Chudasama1, Subrata Pradhan1
1Institute for Plasma Research, India Email: [email protected]
For carrying out experiments on Steady State Superconducting Tokamak-1 (SST-1) in the Institute for Plasma Research(IPR), Gandhinagar, a system for plant & experimental data handling and access is developed and has been used in the Institute since the experiments has began. The SAN based central storage system maintains the whole cycle of experimental data handling: from storing configuration data of plants and experiments systems to the acquisition of raw data from the fusion device (SST-1), to the presentation of processed data and support for the experiment & plant data archive. The storage system facilities the researchers to access the data both locally from within the experiment network and as well as remotely from various sites of the IPR campus network. The system developed is based on modern principle of SAN architecture that allows to produce and handle larger amounts of experimental data without single point of failure, thus providing the opportunities to intensify and extend the fusion researches. The features of the system along with the design principles are reviewed in this paper.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_7
The Role of Equilibrium Flows in Temperature-Gradient-Driven Modes in Hot Tokamaks
Deepak Verma1, Aditya K Swamy1, Rajaraman Ganesh1, Stephan Brunner2, Laurent Villard2
1Institute for Plasma Research, India 2Ecole Polytechnique Federale de Lausanne, France Email: [email protected]
In many major Tokamaks around the world, low frequency micro-instabilities and the ensuing transport are routinely suppressed by a poloidal flow. This poloidal flow could be induced from “outside/externally” or could be self-consistently generated by the plasma processes themselves, for example, shearing fields such as zonal flows [1]. The shear of the poloidal flow thus generated and produces a decorrelation of the mode structure, thus leading to stabilization. They are also believed to play role in L-H transition, which is a phase transition like phenomena from low (L) confinement mode to high (H) confinement mode [1] [2].
In the first part of the work, we present linear global gyro-kinetic formulation to incorporate equilibrium flows in the poloidal and toroidal direction which includes key physics elements such as Landau damping, passing and trapped particle physics, radial and poloidal coupling due to magnetic drifts, FLR effects to all orders and is fully electromagnetic in nature [4] [5]. In the second part we, study the effect of the equilibrium flow on the stability in hot Tokamaks and present some preliminary results.
References:
[1] Zonal flows in plasma – a review, P. H. Diamond et.al. Plasma phy. Control Fusion, 47 (2005).
[2] The JET Team, Nuc. Fusion 32, 187 (1991)
[3] K. H. Burrell, Plasma Phys. Controlled Fusion 36, A291 (1994)
[4] Matteo Maccio Thesis No. [2219] EPFL, Lusanne (2000)
[5] Paolo Angelino Thesis No. [3559] EPFL, Lusanne (2006)
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_8
3D Character of Plasma Transport in the Aditya Limiter Scrape-off Layer
Bibhu Prasad Sahoo1, Devendra Sharma1, Ratneshwar Jha1, Yuhe Feng2
1Institute for Plasma Research, India 2Max-Planck-instite fur Plasmaphysik, Germany Email: [email protected]
Strong 3-dimensional character of plasma transport was identified in the SOL plasma of tokamak Aditya in 3D edge plasma transport simulation using the code EMC3-EIRENE [1, 2, 3]. Quasi- periodic flow structure and associated density modulations are observed in the poloidal direction resulting in complex sheared flow from locations far upstream to the limiter. The modulations in plasma parameters are estimated to generate secondary drifts contributing to the SOL flows routinely measured in the device. The divergence free diamagnetic drift resulting from the quasi- periodic modulations is estimated in the SOL region and shown to result in a flow vorticity along the parallel direction, contributing to the generation of large scale structures that determine an effective cross field diffusivity in the SOL region [4].
References:
[1] Y. Feng, F. Sardei, J. Kisslinger, J. Nucl. Mater 266–269 (1999) 812
[2] Devendra Sharma, Ratneshwar Jha, Yuhe Feng and Francesco Sardei, J. Nucl. Mater. 438 (2013) S554-S558
[3] D. Sangwan, R. Jha, J. Brotankova and M.V. Gopalakrishna, Phys. Plasmas 19, 092507 (2012)
[4] B. P. Sahoo, D. Sharma, R. Jha, Y. Feng , Nucl. Fusion 55, 063042 (2015)
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_10
Fast Visible Imaging and Study of Edge Turbulence in the Aditya Tokamak
Santanu Banerjee1, Ranjana Manchanda1, Malay Bikas Chowdhuri1, Nilam Ramaiya1, Navin Parmar1, Joydeep Ghosh1, Rakesh L Tanna1, Braj Kishore Shukla1, Pramod K Sharma1, Aditya team1
1Institute for Plasma Research, India Email: [email protected]
Fast visible imaging is used on toroidal magnetic confinement devices for a wide variety of purposes. This includes monitoring of the plasma evolution, transient effects in the plasma and the study of edge turbulence. Two fast visible imaging systems are installed on the Aditya tokamak. One of the system is for imaging the plasma evolution with a wide angle lens covering a major portion of the vacuum vessel. The imaging fiber bundle along with the objective lens is installed inside a radial re-entrant viewport, specially designed for the purpose. Another system is intended for tangential imaging of the plasma column.
During the termination phase of Aditya plasma, large filaments are seen predominantly across all types of discharges. These filaments are apparent just after the strong interaction of the plasma column with the high field side limiter surface almost at the end of the discharge. Statistical features of these filaments [1, 2] and their role during the termination of plasma is studied. Further, there are many interesting visual impacts of either the experiments carried out or several inherent phenomena in Aditya like the ECRH and LHCD operations, dynamics of the runaway dominated discharges and plasma equilibrium at various discharge scenarios. Such observations and the gained physical insights will be reported.
References:
[1] S. Banerjee et al., “Statistical features of coherent structures at increasing magnetic field pitch investigated using fast imaging in QUEST,” Nucl. Fusion, 52, 123016 (2012).
[2] S. Banerjee et al., “Role of stochasticity in turbulence and convective intermittent transport at the scrape off layer of Ohmic plasma in QUEST,” Phys. Plasmas, 21, 072311 (2014).
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_26
An Overview of Experimental ICRF Research on NSTX-U
Rory Perkins1, Joel Hosea1, Nicola Bertelli1, John Caughman2, Cornwall Lau2, Cynthia Phillips1, Gary Taylor1, James Wilson1
1Princeton Plasma Physics Laboratory, Princeton NJ, USA 2Oak Ridge National Laboratory, USA Email: [email protected]
The National Spherical Torus eXperiment Upgrade (NSTX-U) will begin operation in 2015. The same twelve-strap 30 MHz fast-wave antenna from NSTX will be available on NSTX-U with up to 6 MW of source power; however the higher magnetic-field strength of NSTX-U will put the waves at the fourth to sixth harmonic of the ion cyclotron frequency. This regime, the “high harmonic fast wave” (HHFW) regime, is intermediate to conventional ICRF heating (minority heating, second-harmonic heating) and fast waves in the lower hybrid frequency range (helicon current drive), which makes for good comparison of fast-wave physics across a broad range of machines [1]. Additional grounding points have been added to the HHFW antenna to improve the high-voltage standoff. Diagnostic upgrades include an infrared camera to monitor the heat flux to the antenna, high-frequency Langmuir divertor probes with electronics suitable for detection of 30 MHz waves, a wide-angle IR camera for edge loss studies, and a reflectometer suited for SOL density profile measurements.
Planned ICRF experiments will first focus on characterization of SOL losses of HHFW power and later to study interactions between beam ions and fast waves and well as solenoid-free start- up. Significant losses of HHFW power were sometimes observed along SOL field lines in NSTX, leading to bright and hot spirals on both upper and lower divertors [2]. The diagnostic upgrades described above will allow for a quantitative characterization of these losses and help determine optimum conditions for coupling HHFW power into H-mode plasmas. Later experiments will focus on the interaction of beam ions with fast waves, both as a power-loss mechanism and as a potential tool to influence fast-ion modes, as well as solenoid-free start-up. Other more technical issues to be addressed on NSTX-U are the compatibility of the HHFW antenna with the new neutral beam (and higher level of NBI power), high-voltage standoff of the antenna, and performance of the double-feed antenna in boronized and lithiated conditions. Additionally, work on a test stand will elucidate the role of induced currents in the antenna sidewalls on the launched wave spectrum and on outgassing [3].
References:
[1] N. Bertelli et al., work presented at this conference, APFA (2015)
[2] J. Hosea et al., “High harmonic fast wave heating efficiency enhancement and current drive at longer wavelength on the National Spherical Torus Experimen,” Phys. Plasmas, 15, 056104 (2008).
[3] R. J. Perkins et al., “High Voltage Test-Stand Research Done on ICRF Antenna Elements of the High-Harmonic Fast-Wave System of NSTX,” to appear in proceedings from 21st Topical Conference on Radiofrequency Power in Plasmas Lake Arrowhead, US (2015).
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_49
Status of A3 Foresight Collaboration among China, Japan and Korea on Critical Physics Issues Specific to Steady State Sustainment of High- Performance Plasmas
Shigeru Morita1, Liqun Hu2, Yeong-kook Oh3
1National Institute for Fusion Science, Japan 2Institute of Plasma Physics Chinese Academy of Sciences, China 3National Fusion Research Institute, Korea Email: [email protected]
The collaboration among China, Japan and Korea based on the A3 foresight program on plasma physics has newly started from August 2012 under the auspice of The National Natural Science Foundation of China (NSFC, China), The Japan Society for the Promotion of Science (JSPS, Japan), National Research Foundation of Korea (NRF, Korea). The period of cooperation is set as five years. The A3 Foresight collaboration on critical physics for the steady state operation of high-performance plasmas is mainly made by three superconducting devices of EAST, KSTAR and LHD, while small devices also contribute to this collaboration program. The A3 collaboration activity is categorized by the following four issues;
(I) Steady state sustainment of magnetic configuration - Current drive and profile control (II) Edge and divertor plasma control - (IIa) Transport of edge and divertor plasmas - (IIb) Stability of edge plasma (III) Confinement of alpha particles - Interaction of energetic particle and bulk plasma (IV) Theory and simulation for (I) - (III)
During past three years several productive results have been obtained with fruitful discussions through personal exchange among three countries of China, Japan and Korea. The purpose of A3 Foresight program is also focused on education of young scientists including graduate students. For the purpose A3 Foresight seminar has been held twice in a year in which the status of on-going collaborations in each category is also presented with discussions on coming collaboration (1st: 22 Aug. 2012, Korea, 2nd: 22-25 Jan. 2013, Japan, 3rd: 20-23 May 2013, China, 4th: 3-4 Nov. 2013, Korea, 5th: 23-26 Jun. 2014, Japan, 6th: 6-9 Jan. 2015, China, 7th: 18-23 May 2015, Korea and 8th: 1-4 Dec. 2015, Japan).
In the conference the status and recent progress in A3 Foresight program are presented with productive results from the collaboration.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_73
Comparison of Different Atomic Databases used for Evaluating Transport Coefficients in Aditya Tokamak
Malay Bikas Chowdhuri1, Joydeep Ghosh1, Santanu Banerjee1, Ranjana Manchanda1, Nilam Ramaiya1, Parveen Kumar Atrey1, Y Shankara Joisa1, Rakesh L Tanna1, Prabal K Chattopadhyay1, Chet Narayan Gupta1, Shailesh B Bhatt1, Motoshi Goto2, Izumi Murakami2
1Institute for Plasma Research, India 2National Institute for Fusion Science, Japan Email: [email protected]
Oxygen impurity transport in typical discharges of Aditya tokamak has been estimated using 4+ 3 3 spatial profile of brightness of Be-like oxygen (O ) spectral line (2p3p D3-2p3d F4) at 650.024 nm. This O4+ spectrum is recorded using a 1.0 m multi-track spectrometer (Czerny-Turner) capable of simultaneous measurements from eight lines of sights. 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 then determined by reproducing the experimentally measured emissivity profiles of O4+, using a one-dimensional empirical impurity transport code, STRAHL. To calculate the emissivity, photon emissivity coefficient (PEC) is required along with electron and O4+ density, which is the output of STRAHL. The PEC values depend on both electron density and temperature and are obtained from ADAS and NIFS atomic databases. Using both the databases, much higher values of diffusion coefficient compared to the neo-classical values are observed in the high and low magnetic field edge regions of typical Aditya Ohmic plasmas. The obtained values of diffusion coefficients using PEC values from both the databases are compared with the diffusion coefficients calculated from the fluctuation induced transport in both the inboard and outboard edge regions. Although similar profiles for diffusion coefficients are obtained using PEC values from both databases, the magnitude differs considerably.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_89
Neutral Particle Profiles during ICRH Experiments in Aditya Tokamak
Nilam Ramaiya1, Ritu Dey1, Ranjana Manchanda1, Malay Bikas Chowdhuri1, Santanu Banerjee1, Niral Virani1, Rakesh L Tanna1, Jayesh V Raval1, Y Shankara Joisa1, Parveen Kumar Atrey1, Shailesh B Bhatt1, Chet Narayan Gupta1, Sanjay V Kulkarni1, Prabal K Chattopadhyay1, Joydeep Ghosh1
1Institute for Plasma Research, India Email: [email protected]
In magnetically confined fusion plasmas, the transition from Low confinement mode (L-mode) to High confinement mode (H-mode) is characterized by improved energy and particle confinement times. One of the characteristic and directly observable signatures of H-mode is sudden decrease in Hα radiation. To explore the effect of ICRH on confinement [1] of the Aditya plasma, radial profiles of Hα with high temporal and radial resolution have been measured using Photomultiplier tube (PMT) array based spectroscopic system [2]. The PMT array module incorporates 8 PMTs, which provides high gain, high sensitivity, wide dynamic range, fast time response & high S/N ratio. Light collected from 8 different vertical chords spanned over the poloidal cross-section in the low-field side edge region of the plasma is transferred to the PMT array through an interference filter having center wavelength at 656.3 nm with 1 nm bandwidth. The chord integrated data is inverted using Abel-like matrix inversion technique [3] to obtain the radial profiles of Hα profile. In this paper, the modification of neutral particle profiles, which suggests the change in the penetration of neutral particle, will be discussed during the ICRH experiments. Comparison with radial profiles of Hα emissivity obtained using DEGAS2 code has been attempted for proper understanding of the role of neutral particle penetration on plasma confinement.
References:
[1] K. Steinmetz et al, “Observation of High-Confinement Regime in a Tokamak Plasma with Ion Cyclotron-Resonance Heating”, Physical Review Letters, Vol. 58, No.2 (1987).
[2] M. B. Chowdhuri et al, “Measurement of spatial and temporal behavior of H emission from Aditya tokamak using a diagnostics based on a photomultiplier tube array”, Review of Scientific Instruments 85, 11E411 (2014).
[3] J. Ghosh et al. “Radially resolved measurements of plasma rotation and flow-velocity shear in the Maryland Centrifugal Experiment”, Physics of Plasmas 13, 022503 (2006).
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_96
Understanding of Impurity Behavior in SST-1 Plasmas Using Visible Spectroscopy
Ranjana Manchanda1, Nilam Ramaiya1, Malay Bikas Chowdhuri1, Santanu Banerjee1, Joydeep Ghosh1, SST-1 Team1
1Institute for Plasma Research, India Email: [email protected]
Studies of impurity behavior in SST-1 plasma have been carried out using visible spectroscopic systems installed on the tokomak. This has been carried out using a low resolution and broadband survey spectrometer covering a 350-900 nm wavelength range, 0.5 m visible spectrometer having 600 and 1200 grooves/mm grating coupled with CCD camera and interference filter and photomultiplier (PMT) tube based systems. Temporal evolution of the hydrogen (Hα, Hβ ) and impurities emissions like, C II, C III, O I, O II, O III , O V and a visible Continuum at 536.0 nm have been monitored using the PMT based system to understand impurity charge state evolution during plasma discharges. All systems are absolutely calibrated for impurity influx and plasma parameter estimations.
Observed spectral lines in the visible range have been identified to recognize the presence of various impurities in the SST-1 plasmas. Comparison of impurities emission has been made for different plasma currents and toroidal magnetic fields. An analysis has been carried out to understand the impurities activities in plasmas of SST-1 tokomak in presence and absence of installed plasma facing components (PFC). Significantly higher carbon emissions have been observed indicating higher carbon content in the plasma with graphite PFCs installed.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_110
Observation of Plasma Shift in SST-1 using Optical Imaging Diagnostics
Manoj Kumar Gupta1, Chesta Parmar1, Vishnu K Chaudhari1, Ajai Kumar1, SST-1 Team1
1Institute for Plasma Research, India Email: [email protected]
A tangential viewing optical imaging system at SST-1 is used to observe the plasma shift both vertical and horizontal during experimental campaigns. The images from the plasma are transferred through optical imaging fibre and coupled to a CCD camera which operates at 31 frames/sec. The data from the CCD camera is transferred through gigabit Ethernet cable to acquisition PC placed in diagnostics lab. The whole system is fully automated for operation and data acquisition of the imaging data. The complete imaging system will be explained in this presentation. With this optical imaging system, the shift in plasma position both in vertical and horizontal direction is observed. The plasma shape and diameter can also be estimated with this system. The estimated diameter during some of the plasma shots is ~50 cm and shape is circular. The data from this diagnostics is very useful from the operation point of view of the machine.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_115
Estimation of Spectrally Resolved Total Radiation Power loss in Aditya Tokamak and its Comparison with Experimental Measurements
Kumuduni Tahiliani1, Malay Bikas Chowdhuri1, Ratneshwar Jha1, Parveen Kumar Atrey1, Y Shankar Joisa1, Joydeep Ghosh1, Rakesh L Tanna1, Aditya Team1
1Institute for Plasma Research, India Email: [email protected]
The radiation power loss in Aditya tokamak is routinely measured using AXUV diodes [1]. Both single channel and arrays of AXUV diode are used for the measurement. In addition, filtered channels are used for the measurement of spectrally resolved radiation loss in the VUV region and to estimate the effective responsivity in the operation regimes where there is a significant contribution of lower energy radiation to the total power loss [2].
In the present work, the steady state radiation power loss in Aditya tokamak is modeled using one dimensional impurity transport code, STRAHL under the assumption of toroidal and poloidal symmetries of the plasma. For this purpose, photon emissivity coefficients from ADAS database of the main impurities, such as carbon and oxygen, have been used to estimate the spectrally resolved radiation power loss. The simulated radiation power loss is compared to the experimentally measured radiation power loss from a typical Aditya plasma discharge and the similarities and discrepancies are discussed.
References:
[1] Kumudni Tahiliani, et al., “Radiation power measurement on the Aditya tokamak, Plasma Physics and Controlled Fusion, 51, 085004 (2009).
[2] Kumudni Tahiliani, et al., “Estimation of effective responsivity of AXUV bolometer in Aditya tokamak by spectrally resolved radiation power measurement”, Plasma Fusion Research, 8, 2402124 (2013).
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_129
Ponderomotive Density Modulation in Two Ion Tokamak Plasma
J K Atul1, S K Singh1, S Sarkar2, O V Kravchenko3
1Magadh University, India 2FCIPT-Institute for Plasma Research, India 3Department of higher mathematics, BMSTU, Moscow, Russia Email: [email protected]
Many efficient heating methods have been proposed to heat Tokamak plasma. It includes various techniques such as compressional heating [1] (magnetic field, electric field, shock wave, beam pressure), wave heating [2] (radio waves, microwaves, laser beams), particle beam injection (electron beams, ion beams, neutral beams) as well as alpha particle heating [3]. Particularly for wave heating processes, it seems that the transport of RF energy into the core regions is one of the major problem in the auxiliary heating of plasma to the thermonuclear temperatures. Nonlinear effects such as pump self-induced filamentation and parametric decays further complicate the heating process [4]. In context with it, an exact nonlinear solution of the Two ion hybrid mode is estimated under the influence of adiabatic perturbations in a Two ion species magnetized plasma. The dominant nonlinearity arises through the ion ponderomotive force term thereby modulating the plasma density profile. The nonlinear equation which has KorteVeg De Vries [KdV] soliton as its solution, represents the nonlinear stage of a purely growing mode. It turns out that these solitons exists only if the wave frequency is lower than the Buschbaum frequency [5] and if the concentration of the lighter ions is less than the heavier one. The application of the theory is discussed in terms of Proton and Tritium minority species in a Deuterium plasma.
References:
[1] Avinash, K., and P. K. Kaw. "Plasma Heating by Electric Field Compression."Physical review letters,112, 185002,( 2014).
[2] Cairns R. A., Radiofrequency heating of plasmas. Institute of Physics Publishing, 1991.
[3] T. J. Dolan (ed.), Magnetic Fusion Technology, Lecture Notes in Energy 19, Springer-Verlag London, (2013).
[4] Tagare, S. G., and P. Rolland. "The nonlinear filamentation of lower‐hybrid waves by ion‐ion hybrid perturbations." Physics of Fluids (1958-1988) 25.11. 2012-2018. (1982).
[5] Buchsbaum, S. J. "Ion resonance in a multicomponent plasma." Physical Review Letters 5.11, 495. (1960).
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_133
Study of Neutral Particle Transport in Aditya Tokamak Plasma using DEGAS2 Code
Ritu Dey1, Joydeep Ghosh1, Malay Bikas Chowdhuri1, Ranjana Manchanda1, Santanu Banerjee1, Nilam Ramaiya1, Aditya Team1
1Institute for Plasma Research, India Email: [email protected]
Aditya tokamak is a medium sized air-core tokamak having a limiter configuration. The circular poloidal ring limiter is placed at one particular toroidal location. The spatial profile of neutral particles are experimentally observed in this tokamak [1] and the observation suggests important roles of charge exchange processes into the penetration of neutral particle in plasma core. Therefore, to understand the neutral dynamics in Aditya tokamak, the neutral particle transport studies have been carried out using the DEGAS2 code [2]. This code is based on Monte Carlo algorithms and extensively used for investigating the dynamics of neutrals in various tokamaks having divertors as the plasma facing component. The required modification has been carried out in the machine geometries and plasma parameter files through the user developed programs for ADITYA tokamak plasma. Modifications are successfully implemented in this code and the radial profile of Hemissivityhas been obtained. The simulated results are then compared with the experimental observations. In this paper, details on the implementation of the code on Aditya tokamak plasmas are presented and the simulation results are compared with the experiments to understand the neutral particle behaviour in Aditya tokamak plasma.
References:
[1] S. Banerjee, J. Ghosh, R. Manchanda, et al., “Observations of Hα emission profiles in Aditya tokamak”, J. Plasma Fusion Res. Series 9, 29 (2010).
[2] D. P. Stotler, C. F. F. Karney, “Neutral gas transport modeling with Degas2”, Contrib. Plasma Phys. 34, 392 (1994).
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_142
Modeling of Eddy Current distribution and Equilibrium Reconstruction in the SST-1 Tokamak
Santanu Banerjee1, Amit Kumar Singh2, Deepti Sharma1, Srinivasan Radhakrishnana1, Raju Daniel1, Y Shankara Joisa1, Parveen Kumar Atrey1, Surya Kumar Pathak1, SST-1 Team1
1Institute for Plasma Research, India 2ITER-India, Institute for Plasma Research, India Email: [email protected]
Toroidal continuity of the vacuum vessel and the cryostat leads to the generation of large eddy currents in these passive structures during the Ohmic phase of the steady state superconducting tokamak SST-1. This reduces the magnitude of the loop voltage seen by the plasma as also delays its buildup. During the ramping down of the Ohmic transformer current (OT), the resultant eddy currents flowing in the passive conductors play a crucial role in governing the plasma equilibrium. Amount of this eddy current and its distribution has to be accurately determined such that this can be fed to the equilibrium reconstruction code as an input. For the accurate inclusion of the effect of eddy currents in the reconstruction, the toroidally continuous conducting structures like the vacuum vessel and the cryostat with large poloidal cross-section and any other poloidal field (PF) coil sitting idle on the machine are broken up into a large number of co-axial toroidal current carrying filaments. The inductance matrix for this large set of toroidal current carrying conductors is calculated using the standard Green’s function and the induced currents are evaluated for the OT waveform of each plasma discharge. Consistency of this filament model is cross-checked with the 11 in-vessel and 12 out-vessel toroidal flux loop signals in SST-1. Resistances of the filaments are adjusted to reproduce the experimental measurements of these flux loops in pure OT shots and shots with OT and vertical field (BV). Such shots are taken routinely in SST-1 without the fill gas to cross-check the consistency of the filament model.
A Grad-Shafranov (GS) equation solver, named as IPREQ [1], has been developed in IPR to reconstruct the plasma equilibrium through searching for the best-fit current density profile. Ohmic transformer current (OT), vertical field coil current (BV), currents in the passive filaments along with the plasma pressure (p) and current (Ip) profiles are used as inputs to the IPREQ code to reconstruct the equilibrium consistently with the flux loop measurements and the poloidal flux, plasma shape, βp and the safety factor (q) are inferred.
References:
[1] Tokamak equilibrium code-IPREQ, R. Srinivasan and S. P. Deshpande IPR/RR-393/2007 (August, 2007)
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_144
Equilibrium Reconstruction of Plasma Discharges in the Aditya Tokamak
Deepti Sharma1, Santanu Banerjee1, Amit Kumar Singh1, Srinivasan Radhakrishnana1, Raju Daniel1, Rakesh L Tanna1, Joydeep Ghosh1, Y Shankara Joisa1, Parveen Kumar Atrey1, Surya Kumar Pathak1, Aditya Team1 1Institute for Plasma Research, India Email:[email protected]
External magnetic measurements with flux loops and magnetic pick-up coils in tokamaks have provided vital information on the shape of the plasma column and also global current profile parameters, such as the sum of the poloidal beta (βp) and the internal inductance (ℓi) [1]. Such a reconstruction needs to be fast and sufficiently accurate such that it can be used routinely as a complementary input with other experimentally measured parameters for any sort of physics analysis of the plasma discharges.
Here we present a method which can be used to proficiently reconstruct the current profile parameters, the plasma shapes, and a current density profile satisfying the MHD equilibrium constraint, reasonably conserving the external magnetic measurements. A Grad-Shafranov (GS) equation solver, named as IPREQ [2], has been developed in IPR to search for the best-fit current density profile. GS equation is a nonlinear elliptical differential equation describing axisymmetric toroidal equilibria. Ohmic transformer current (OT), vertical field coil current (BV) along with the plasma pressure (p) and current (Ip) profiles are used as inputs to the IPREQ code to reconstruct the equilibrium and the poloidal flux, plasma shape, βp and the safety factor (q) are inferred.
At the four corners of the square cross-section vacuum vessel of Aditya, there are four magnetic pick-up coils aligned to measure the poloidal magnetic field (B) during a plasma discharge. Further, there are two toroidal flux loops at the shadow of the limiter on the high field side to measure the loop voltage inside the vacuum vessel. Vacuum shots with OT and BV and no fill gas are used to calibrate these coils and loops. Measurement from these coils and flux loops are used to reconstruct the equilibrium consistently with the peak density and temperature measurements. Finally, the reconstructed equilibria are validated against the visible images from the fast visible imaging diagnostic on Aditya.
References:
[1] L. L. Lao et al., “Reconstruction of current profile parameters and plasma shapes in tokamaks,” Nucl. Fusion, 25, 1611 (1985).
[2] Tokamak equilibrium code-IPREQ, R. Srinivasan and S. P. Deshpande IPR/RR-393/2007 (August, 2007)
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_149
Ohmic Discharges with Improved Confinement in Tokamak Aditya
Rakesh L Tanna1, Harshita Raj1, Joydeep Ghosh1, Prabal K Chattopadhyay1, Sharvil Patel2, Kumarpalsinh A Jadeja1, Kaushal M Patel1, Shailesh B Bhatt1, Chet Narayan Gupta1, Kunal Shah1, Motibhai Makwana1, Narendra Patel1, Vipul K Panchal1, Chhaya Chavda1, Pramod Sharma1, Malay Bikas Chowdhuri1, Santanu Banerjee1, Nilam Ramaiya1, Ranjana Manchanda1, Raju Daniel1, Sameer Kumar Jha1, Kumuduni Tahiliani1, Praveenlal Edappala1, Shishir Purohit1, Y Shankar Joisa1, Jayesh V Raval1, C V S Rao1, Parveen Kumar Atrey1, Surya Kumar Pathak1, Ratneshwar Jha1, Amita Das1, Dhiraj Bora1
1Institute for Plasma Research, India 2Gujarat University, India Email:[email protected]
ADITYA (R0 = 75 cm, a = 25 cm), an ohmically heated circular limiter tokamak is regularly being operated to carry out several experiments related to controlled thermonuclear fusion research. In recent experimental schedule, special efforts are made to enhance the plasma parameters to achieve Ohmic discharges with improved confinement. Repeatable plasma discharges of maximum plasma current of ~ 160 kA and discharge duration beyond ~ 250 ms with plasma current flattop duration of ~ 140 ms has been obtained for the first time in the first Indian tokamak ADITYA. The discharge reproducibility has been improved with Lithium wall conditioning and much-improved plasma discharges are obtained by precisely controlling the plasma position. Improved discharges are attempted over a wider parameter range to carry out various confinement scaling experiments. In these discharges, chord-averaged electron density ~ 1.0 – 4.0 1019 m–3 using multiple hydrogen gas puffs, plasma temperature of the order of ~ 400 – 700 eV has been achieved. The measured confinement time matches quite well with ALCATOR scaling for most of the discharges. It is also observed that in new discharges, the confinement time crosses the L-mode scaling. Detailed analysis of these discharges along with the possible reasons for obtaining higher confinement times will be addressed in this paper.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_168
Investigation of Aditya Tokamak Plasmas with Lithiumized Wall
Niral Virani1, Malay Bikas Chowdhuri1, Kumarpalsinh A Jadeja2, Joydeep Ghosh1, Ranjana Manchanda1, Nilam Ramaiya1, Santanu Banerjee1, Jayesh V Raval1, Y Shankara Joisa1, Umeshkumar C Nagora1, Parveen Kumar Atrey, Rakesh L Tanna, Prabal K Chattopadhyay1, Chet Narayan Gupta1, Shailesh B Bhatt1, Aditya team1
1Institute for Plasma Research, India Email: [email protected]
The Lithium coating on plasma facing components of tokamak leads to better plasma properties through the reduction in impurities and controlling the hydrogen recycling. In Aditya tokamak, lithiumization of vacuum vessel wall is regularly carried out prior to its daily operation using lithium rod exposed to overnight glow discharge-cleaning plasma. Spectroscopic studies of Aditya tokamak plasmas shows the reduction of hydrogen (H at 656.3 nm) and oxygen (O II at 441.6 nm) as compared to discharges without the lithium coated walls. This clearly indicates reduction of recycling and impurity influxes from the wall, respectively. After Li coating, plasma stored energy increases significantly and plasmas with higher electron densities are obtained. Estimation of energy confinement time shows that it increases after lithimization and becomes comparable to the values predicated by Neo-Alcator scaling for ohmically heated tokamak plasma. Further analysis indicates that recycling must be low to achieve better plasma confinement.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_173
A Study of Anomalous Transportation of Sawtooth Generated Runaway Electrons Observed in ADITYA Tokamak
Harshita Raj1, Joydeep Ghosh1, Rakesh L Tanna1, Prabal K Chattopadhyay1, Raju Daniel1, Sameer Kumar Jha1, Jayesh V Raval1, Y Shankara Joisa1, Shishir Purohit1, C V S Rao1, Umeshkumar C Nagora1, Parveen Kumar Atrey1, Malay Bikas Chowdhuri1, Ranjana Manchanda1, Yogesh C Saxena1, Rabindranath Pal2 and Aditya Team1
1Institute for Plasma Research, India 2Saha Institute of Nuclear Physics, 1ndia Email: [email protected]
In Aditya tokamak, Hard X-Ray spikes coinciding with the sawtooth crashes have been observed at the plateau phase of plasma current in many discharges. Owing to the fact that the runaway electrons generate the hard X-ray spikes while hitting the limiter, generation of runaway electrons during the sawtooth crash and their radial propagation to the limiter has been investigated. The electric field generated during sawtooth crash is estimated and found to be more than critical electric filed required for the electrons to run away. The energy gained by Runaway electrons due to this Electric field matches quite well with energy of Hard X-Ray spikes observed. Further investigation reveals that in later part of the same discharge no HXR bursts were observed in spite of presence of similar sawtooth oscillation. To understand this anomaly in HXR burst pattern, radial transport of runaway electrons is thoroughly examined. The hard X-ray bursts are only observed in presence of high Mirnov activity whereas during low Mirnov activity the hard X-ray bursts are absent. Estimations of magnetic island width reveals that overlapping of the magnetic islands m = 2 and m = 3 takes place when hard X-ray bursts are observed, which may be causing the faster transportation of runaway electrons. The detail exploration of the anomaly observed will be presented in this paper.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_177
Geodesic Acoustic Modes with Poloidal Mode Coupling ad Infinitum
Rameswar Singh1, Ozgur D Gurcan1
1Laboratoire de Physique des Plasmas, Ecole Polytechnique, France Email:[email protected]
Geodesic acoustic mode (GAM) is studied including all poloidal mode (m) couplings using drift reduced fluid equations. The nearest neighbor coupling pattern, due to geodesic curvature, leads to a semi-infinite chain model of the GAM with the mode-mode coupling matrix elements proportional to the radial wave number k_r. The infinite chain can be reduced to a renormalized bi-nodal chain with a matrix continued fraction. Convergence study of linear GAM dispersion with respect to k_r and the m -spectra confirms that high m couplings become increasingly important with k_r. Theoretical predictions were compared against the experimental observations on GAM frequency profiles in Tore Supra shots showing that the radially sorted theoretical roots down shift to overlap with experimental GAM frequency profile for low collisionality shot in the radial wave number range 0.1< kr <0.15. This is proposed as a possible resolution of the GAM frequency reduction in Tore Supra compared with the previous theories.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_178
Mean EB Shear Effect on Geodesic Acoustic Modes in Tokamaks
Rameswar Singh1, Ozgur D Gurcan1
1Laboratoire de Physique des Plasmas, Ecole Polytechnique, France Email: [email protected]
E × B shearing effect on geodesic acoustic mode (GAM) is investigated for the first time both as an initial value problem in the shearing frame and as an eigenvalue value problem in the lab frame. The nontrivial effects are that E × B shearing couples the standard GAM perturbations to their complimentary poloidal parities. The resulting GAM acquires an effective inertia increasing in time leading to GAM damping. Eigenmode analysis shows that GAMs are radially localized by E × B shearing with the mode width being inversely proportional and radial wave number directly proportional to the shearing rate for weak shear.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_187
Estimation of Vacuum Magnetic Fields due to Ohmic Coils in Aditya Upgrade tokamak
Krishana Kumari K1, Rohit Kumar1, Rakesh L Tanna1, Joydeep Ghosh1, Prabal K Chattopadhyay1, Srinivasan Radhakrishnana1, Sharvil Patel2, Raju Daniel1, Someswar Dutta1, Dhiraj Bora1, Yogesh C Saxena1, Aditya Team1
1Institute for Plasma Research, India 2Gujarat University, India Email: [email protected]
The magnetic null is of utmost importance in plasma formation in any tokomak. In Ideal case, the radial (Br) and vertical (Bz) component of magnetic field produced by the ohmic transformer coil should be approximately zero at some specific location inside the vacuum vessel. Non-zero Br & Bz within the plasma region acts as error field and causes difficulties in gas breakdown. Auxiliary transformer coils TR2, TR3, TR4, in series with central solenoid TR1 are used for error field compensation in Aditya tokamak within the plasma volume. The main sources of error field are imperfection in coil positions, geometry of the coils, small variation in the coil fabrication or misalignment of the large coil systems and stray fields. Furthermore, the error fields remain present when the compensation provided by auxiliary coils is not sufficient. Therefore, vacuum magnetic fields due to ohmic transformer coils need to be estimated for precisely placing the magnetic null location inside the vacuum vessel for better plasma current inception. The magnetic field and its components generated due to ohmic transformer coils has been estimated using different computer codes such as ANSYS, EFFI etc. The codes are first validated using analytical calculations. In this paper, the optimization of the coil positions in order to obtain null position at the desired location inside the vacuum vessel has been presented.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_190
Divertor Coil Power Supply in Aditya Tokamak for improved Plasma Operation
Vaibhav Ranjan1, Kunal Shah1, Motibhai N Makawana1, Chet Narayan Gupta1, A Varadharajulu1, Joydeep Ghosh1, Rakesh L Tanna1, Prabal K Chattopadhyay1, Raju Daniel, Srinivasan Radhakrishnana1, Yogesh C Saxena1
1Institute for Plasma Research, India Email:[email protected]
The existing Aditya tokamak, a medium sized tokamak with limiter configuration is being upgraded to a tokamak with divertor configuration. This moderate field tokamak is capable of producing 250 kA of plasma current with 300 ms duration. Two new sets of diverter coils will be added to the system with an objective of producing double null plasmas in Aditya Upgrade tokamak. Diverter coils, made up of continuously transposed conductor, are low voltage high current carrying poloidal field coils. One set of inner divertor coil has radius of 460 mm containing 6 turns and the other set of 1075 mm radius coil with 1 turn makes the outer divertor coils. The simulated plasma double null equilibrium demands 150 kAT of NI for the inner divertor coils and 10 – 20 kAT of NI for outer divertor coils. To energize the divertor coils with required power, a pulsed DC power supply of 3 MW (100V, 30 kA) has been designed. The designed pulsed DC power supply will be a 3-phase, 12-pulse rectifier based convertor power supply having a duty cycle of 300 ms on-time and 15 minutes off-time. The current rise time in the divertor coils will be ~ 0.6 MA/sec. Detailed design of the divertor power supply with active controls for real time control of the plasma shape will be discussed in this paper.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_203
The First Results of Te Measurement with of Soft X-Ray Diagnostics in SST-1 Tokamak
Jayesh V Raval1, Shishir Purohit1, Y Shankara Joisa1, Ajai Kumar1
1Institute for Plasma Research, India Email: [email protected]
Soft X-Ray (SXR) is one of the important diagnostics for high temperature tokamak plasma. It can be used to measure the relative intensity of the emission in Soft X-Ray region (100eV to 20keV) of the spectrum. Radiated Soft X-Ray flux mainly depends on basic characteristics of plasma density, electron temperature and impurity. Soft X-Ray diagnostics has been designed on SST-1 tokamak to measure chord average electron temperature based on absorption foil thickness principle [1] with certain conditions. In the early experiment phases of SST-1 tokamak, plasma density and electron temperature were very low to measure by conventional measurement techniques. SXR diagnostics is modified in way that estimation of temperature with low plasma density is possible at the cost of spatial resolution and accuracy. Present system consists of pair of two silicon surface barrier detector (SBD) with different foil thickness viewing plasma through 8mm pin-hole camera, covered with Be-filter of 10µm thickness. In this COMMUNICATION, modified diagnostics system and first results of recent experimental campaign are discussed. Using Soft X-Ray intensity from the two detectors with different foil thickness, chord averaged electron temperature 80-100eV is estimated.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_204
An Overview of SST-1 Diagnostics and Results from Recent Campaigns
Ajai Kumar1, Asha N Adhiya1, Hemchandra C Joshi1, Janmejay U Buch1, Jayesh V Raval1, Jinto Thomas1, Joydeep Ghosh1, Kiran Patel1, Kumar Ajay1, Kumudni Tahiliani1, M V Gopalakrishna1, Malay Bikas Chowdhuri1, Manoj Kumar1, Neha Singh1, Nilam Ramaiya1, Parveen Kumar Atray1, Pabitra K Mishra1, Ratneshwar Jha1, Raju Daniel1, Rajwinder Kaur1, Ranjana Manchanda1, Sameer Kumar Jha1, Santanu Banerjee1, Santosh P Pandya1, Shishir Purohit1, Shwetang N Pandya1, Snehlata Gupta1, Surya Kumar Pathak1, Umeshkumar C Nagora1, Varsha Siju1, Vishnu K Chaudhari1, Y Shankar Joisa1
1Institute for Plasma Research, India Email:[email protected]
SST-1 is a large aspect ratio tokomak with superconducting magnets designed to operate in steady-state mode for around 1000 seconds. All essential diagnostics for the machine operation and advance diagnostics are commissioned in SST-1 during the different phases of its operation. This report describes the various diagnostics in SST-1 and the results of recent SST-1 campaign with Plasma Facing components. The chord averaged electron density of SST-1 plasma is recorded in the range of 2-5 1012 /cc and the electron temperature is estimated around 100 eV. Various spectral line emissions from plasma and temporal evolutions of some of them have been recorded by spectroscopy diagnostics to understand the impurity behaviour in the SST-1 plasma. The radiation power loss and the power deposited on limiter has been estimated using bolometry and IR thermography respectively. Plasma evolution recorded using visible imaging diagnostics. The energy distribution of non-thermal electron has been characterised using LaBr spectrometer and NaI detector. This article will also be discussing about the possible additions and modification planned for the near future.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_205
Design and Development of AXUV-based Soft X-Ray Diagnostic Camera for ADITYA Tokamak
Jayesh V Raval1, Shishir Purohit1, Y Shankara Joisa1, Joydeep Ghosh1, Rakesh L Tanna1, Kumarpalsinh A Jadeja1, Ajai Kumar1, Shailesh B Bhatt1, Praveena Kumari1, Vismaysinh Raulji1, Minsha Shah1, Rachana Rajpal1
1Institute for Plasma Research, India Email: [email protected]
The hot tokamak plasma emits Soft X-rays (SXR) in accordance with the temperature and density which are important to be studied. A silicon photo diode array (AXUV16ELG, Opto- diode, USA) based prototype SXR diagnostics is designed and developed for ADITYA tokamak for the study of SXR radial intensity profile, internal disruption (Saw-tooth crash), MHD instabilities. The diagnostic is having an array of 16 detector of millimeter dimension in a linear configuration. Absolute Extreme Ultra Violate (AXUV) detector offers compact size, improved time response with considerably good quantum efficiency in the soft X-ray range(200 eV to10 keV). The diagnostic is designed in competence with the ADITYA tokamak protocol. The diagnostic design geometry allows detector view the plasma through a slot hole (0.5 cm 0.05 cm), 10 µm Beryllium foil filter window, cutting off energies below 750 eV .The diagnostic was installed on Aditya vacuum vessel at radial port no 7 enabling the diagnostics to view the core plasma. The spatial resolution designed for diagnostic configuration is 1.3 cm at plasma centre. The signal generated from SXR detector is acquired with a dedicated single board computer based data acquisition system at 50 kHz. The diagnostic took observation for the ohmically heated plasma. The data was then processed to construct spatial and temporal profile of SXR intensity for Aditya plasma. This information was complimentary to the Silicon surface barrier detector (SBD) based array for the same plasma discharge. The cross calibration between the two was considerably satisfactory under the assumptions considered.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_210
Conceptual Design of Diagnostics for the HL-2M Tokamak
Qingwei Yang1, Yi Liu1, Z B Shi1, HL-2A Diagnostic Group1
1Southwestern Institute of Physics, China Email: [email protected]
The HL-2M (intention) tokamak is a new experimental device, which is on constructing at SWIP (Southwestern Institute of Physics) in Chengdu, China. The dimensions are the major radial of R = 1.78 m, minor radial of a = 0.64 m, the maximum elongation of κ = 2.0 and the triangularity of δ = 0.4 ~ 0.8. The main plasma parameters are the plasma current of IP = 2.5 MA, toroidal 20 -3 magnetic field of BT = 2.2 T, line average electron density of ne = 1.6 10 m ，the electron temperature Te and ion temperature Ti expected of 6.0 keV and 12.0 keV during ECRH and NBI respectively. The conceptual design of the diagnostic systems is described in this paper.
To meet the need of experimental studies, about 50 kinds of diagnostics will be utilized on HL- 2M for the parameter measurements and physics analysis, which include:
1. Magnetic coils: to measure the plasma current, plasma position, plasma energy, halo current and MHD instabilities. 2. Laser aided diagnostics: Thomson scattering (YAG) is used to measure Te(r) (at plasma core, edge and divertor). Dispersion interferometer (CO2) and Polarimeter (HCOOH) are used to measure ne(r) and q(r). 3. Beam aided diagnostics: CXRS system for Ti(r), MSE system for q(r) and FIDA for fast ion measurements are arranged. 4. Passive spectrum: The visible spectrometer is designed for low-Z (Carbon and Oxygen) impurities and VUV/EUV spectrometer is used for higher-Z (Si, Al, Ti, Ni, W, etc.) material monitors and further their transport studies. The Zeff(r) will employ the continuum spectrum detection. 5. Microwave systems: ECE radiometer for Te(r), reflectometer for ne(r), interferometer for ne (at divertor) measurements are proposed. 6. Ion and neutral particle measurement: NPA (neutral particle analyzer) for Ti(r) and fast ion energy-spectra profile measurements, and fast lost ion probe (FLIP) for lost ion detection are designed. 7. X-ray and neutron detection: hard X ray emission detection for runaway electron monitor and super-thermal electron detection are planed. The fission chamber, 3He detector and liquid scintillator (in the neutron camera) are utilized to monitor the neutron flux, spectrum and profile respectively. 8. Operational diagnostics: the bolometer arrays, soft X ray camera, divertor Langmuir probe arrays, Hα arrays, fast neutral gas pressure gauge, visible and infrared cameras, etc. are employed to detect HL-2M plasma during discharge. 9. Besides, to meet the need of physics studies, the ECE imaging, reflectometer imaging, CO2 laser scattering, fast movable Langmuir probe, BES (beam emission spectrum), Doppler reflectometer etc. also be programmed.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_211
Observation of MHD Phenomenon for SST-1 Superconducting Tokamak
Manisha Bhandarkar1, Jasraj Dhongde1, Subrata Pradhan1
1Institute for Plasma Research, India Email: [email protected]
Steady State Superconducting Tokamak (SST-1) is a medium size Tokamak (major radius=1.1m, minor radius=0.2m) 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 ~500ms 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, X-Ray etc.) and heating system (ECRH) for the recent campaigns of SST-1.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_212
The Determination of Plasma Radial Shafranov Shift (R) and Vertical Shift (Z) Experimentally using Magnetic Probe and Flux Loop Method for SST-1 Tokamak
Subrata Jana1, Jasraj Dhongde1, Harish Masand1, Subrata Pradhan1
1Institute for Plasma Research, India Email: [email protected] @ipr.res.in
The Radial Shafranov shift (ΔR) and Vertical Shift (ΔZ) has been calculated for steady state Superconducting tokamak (SST-1) [1] experimentally using magnetic probes [2] and Flux loops [3]. The SST-1 plasma at the present phases of operations is circular in shape and leans against the limiters. The Radial and vertical shift formulated from Shafranov equation have been used for computation.Flux loops and magnetic probes are used according to machine geometries for ΔR and ΔZ measurements. The results obtained from these two methods for numerous numbers of shots for SST-1 campaigns are found to be in good agreement, repeatable and reliable. Since the control of plasma position plays an important role in plasma confinement and optimized tokamak operations, this mentioned methodology and results (ΔR, ΔZ as control parameter) could later be used as a plasma position feedback control in long duration SST-1 plasma experiments.
References:
[1] S Pradhan et al, “The First Experiment in SST-1”, IOPScience, Nuclear Fusion 55(2015) 104009 (10pp).
[2] A Salar Elahi, M Ghoranneviss, “Measurement of the plasma boundary Shift and approximation of the Magnetic Surfaces on the IR-T1 tokamak”, Brazilian Journal of Physics, Vol.40 no.3 (2010).
[3] A Salar Elahi,M Ghoranneviss, “Estimation of plasma horizontal displacement using flux loop and Comparison with analytical solution in IR-T1 Tokamak”, Journal of Nuclear and Particle Physics, 2(6),142-146 (2012).
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_213
Development of New Diagnostics for WEST
P Lotte1, P Moreau1, C Gil1, J Bucalossi1, M H Aumeunier1, J M Bernard1, C Bottereau1, C Bourdelle1, Y Camenen2, M Chernyshova7, F Clairet1, T Czarski7, M Choi6, G Colledani1, Y Corre1, X Courtois1, R Daniel3, D Davis3, P Devynck1, D Douai1, D Elbeze1, A Escarguel2, C Fenzi1, W. Figacz7, J C Giacalone1, R Guirlet1, J Gunn1, S Hacquin1, X Hao4, J Harris9, G T Hoang1, F Imbeaux1, S Jablonski7, A Jardin1, H C Joshi3, G Kasprowicz8, C Klepper9, E Kowalska-Strzeciwilk7, M Kubkowska7, A Kumar3 , V Kumar3, W Lee5, B Lyu4, P Malard1, L Manenc1, Y Marandet2, D Mazon1, O Meyer1, M Missirlian1, D Molina1, G Moureau1, Y Nam6, E Nardon1, T Nicolas1, R Nouailletas1, H Park5, J Y Pascal1, K Pozniak8, N Ravenel1, R Sabot1, F Samaille1, J Shen4, J M Travere1, E Tsitrone1, S Varshney3, S Vartanian1, D Volpe1, F D Wang4, G Yun6, W Zabolotny8 and WEST team1
1IRFM, CEA, F-13108 Saint Paul lez Durance, France 2CNRS, Aix-Marseille Université, PIIM UMR 7345, Marseille, France 3IPR, Near Indira Bridge, Bhat, Gandhinagar- 382 428, Gujarat, India 4ASIPP, Hefei Institutes of Physical Science, Hefei 230031, Anhui, P. R. China 5Ulsan National Institute of Science and Technology, Ulsan, Korea 6Pohang University of Science and Technology, Pohang, Korea 7IPPLM, Hery 23, 01-497 Warsaw, Poland 8Warsaw University of Technology, Nowowiejska 15/19, 00-665 Warsaw, Poland 9Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37830-6288, USA Email: [email protected]
WEST, the upgraded superconducting tokamak Tore Supra, will be an international experimental platform aimed to support ITER Physics program. The main objective of WEST is to provide relevant plasma conditions for validating plasma facing component (PFC) technology, in particular the actively cooled Tungsten divertor monoblocks, and also assessing high heat flux and high fluence plasma wall interactions with Tungsten in order to prepare ITER divertor operation. In parallel, WEST will also open new experimental opportunities for developing integrated H mode operation and exploring steady state scenarios in a metallic environment.
In order to fulfil the Scientific Program of WEST, new diagnostics have been developed in addition to the already existing diagnostics of Tore Supra, modified and improved during the shutdown. For the PFC technology validation program, new tools have been implemented, like a full infrared survey of the PFC, a new calorimetry system, local temperature measurements (thermocouple and Bragg grating optical fiber), and several sets of Langmuir probes. For the analysis of long pulse H mode operation, new plasma diagnostics will be implemented, among which the Visible Spectroscopy diagnostic for W sources and transport studies, the Soft-Xray diagnostic based on gas electron multiplier detectors for transport and MHD studies, the X-ray imaging crystal spectroscopy diagnostic with advanced solid state detector properties for ion temperature, ion density and plasma rotation velocity measurements, and the ECE Imaging diagnostic for MHD and turbulence studies. Most of these new diagnostics are developed with the participation of French Universities or through international collaborations. This paper focuses on the description of these four plasma diagnostics.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_220
Observation on Runaway Discharges in SST-1 Experiments
Kiritkumar B Patel1, Subrata Pradhan1
1Institute for Plasma Research, India Email: [email protected]
During ECRH assisted Plasma discharge experiments in Superconducting Steady State Tokamak-1 [SST-1], spike in loop voltage signals are observed. These spikes are modeled using the one-dimensional diffusion equation proposed by I El Chamaa Neto et al. [1] for SST-1 parameters and compared with the experimental data to explain the relaxation instability of Plasma in the runaway dominated discharges. This best fitting of experimental data with the modeled data helps in concluding on plasma conductivity and runway parameters. The best fit gives g << 1. It is observed that these spikes are correlated in time with Plasma current, H-α, ECE, OV and Hard X-ray lines.
References:
[1] I El Chamaa Neto, Yu K Kuznetsov, I C Nascimento, R M O Galvao and V S Tsypin, Physics Plasmas 7, 2894 (2000).
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_221
Hard X-ray Diagnostic for SST-1
Shishir Purohit1, Jayesh V Raval1, Y Shankara Joisa1, Ajai Kumar1, SST-1 Team1
1Institute for Plasma Research, India Email: [email protected]
An experimental study of runaway electrons for the SST-1 tokamak has been performed by the investigation of Hard X-ray measurements by dedicated HXR diagnostics which explores the time evolution and energy distribution of the HXR (energy > 150 keV) from the SST-1 plasma. Both of the objectives have been accomplished by two different detector configurations, stationed on the SST-1 platform, viewing to the SST plasma radially. The Hard X-ray time evolution is addressed by the NaI scintillator detector based diagnostic system. The NaI crystal is 3”x 3” in dimension supported with compatible electronic to have observation with 1 MHz frequency. The system is calibrated, with Cs137 and works within the energy range of 200 keV to 10 MeV depending on the gain settings. The Energy distribution of the Hard X-rays is performed by a Hard X-ray spectrometer which is LaBr (Ce) based. The crystal employed is 1.5” 1.5” with sophisticated electronic capable enough t o handle 250k counts/sec. The diagnostic is fast enough to handle the heavy Hard X-ray flux from the crystal side too. The detector is calibrated in energy space and shows a fairly good energy resolution, 3% @ 662 keV. The operational energy range of this detector system is 150 keV -5 MeV. The range is variable in to the higher energy side by introduction of gain settings. The lower detection limit is restricted by the crystal properties. The Hard X-rays from the SST-1 plasma was observed to be less in energy. The centroid of the population curve resides well below the 200 keV mark. Occasionally high energy spikes are registered which is possible associated with the minor disruptions or with the application of auxiliary power systems.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_223
Study of MHD Activities in the Plasma of SST-1
Jasraj Dhongde1, Manisha Bhandarkar1, Subrata Pradhan1, Sameer Kumar Jha1, SST-1 Team1
1Institute for Plasma Research, India Email: [email protected]
Steady State Superconducting Tokamak (SST-1) [1] is a medium size Tokamak in operation at the Institute for Plasma Research, India. SST-1 has been consistently producing plasma currents and plasma durations in excess of 60kA, 400ms respectively at a central field of 1.5T over last few experiment campaigns of 2014. Investigation of these experimental data of Mirnov coils [4, 5] suggests the presence of MHD activity in the SST-1 plasma. Further analysis clearly explains the behavior of MHD instabilities observed [2, 3], modes present (i.e. m=2, n=1), estimates the characteristic growth time, growth rate for an island and island width etc in the SST-1 Plasma. MHD activity i.e. Poloidal magnetic field and Toroidal magnetic field fluctuations in SST-1 are observed using Mirnov coils. Onsets of disruptions in connection with MHD activities have been correlated with other diagnostics such as ECE, Density, and Hα etc. The observations have been cross compared with the theoretical calculations and are found to be in good agreement.
References:
[1] S. Pradhan, Z. Khan, V.L. Tanna et al, “ The first experiments in SST-1”, Nuclear Fusion,55, (2015).
[2] M. Asif, X. Gao et al, “Study of MHD activity in the HT-7 superconducting tokamak”, Physics Letters, A 342, (2005).
[3] Pravesh Dhyani, J. Ghosh et al, “A novel approach for mitigating disruptions using biased electrode in Aditya tokamak”, Nuclear Fusion, 54, (2014).
[4] C Nordone, “Multichannel fluctuation data analysis by the Singular Value Decomposition method. Application to MHD modes in JET”, Plasma Physics and Controlled Fusion, .34, (1992).
[5] D. Raju, R. Jha et al., “Mirnov coil data analysis for tokamak Aditya”, Pramana Journal of Physics, 55, (2000)
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_230
A Fixed Frequency Reflectometer to Measure Density Fluctuations at Aditya Tokamak
Parveen Kumar Atrey1,2, Dhaval Pujara2, Subroto Mukherjee1
1Institute for Plasma Research, India 2Nirma University, India Email: [email protected]
Amongst modern diagnostics of fusion plasmas, microwave methods, both passive and active, play an important role. Microwave Reflectometer is used to measure the plasma density and its fluctuations in fusion research device like tokamak. A fixed frequency (O – mode) microwave 12 -3 reflectometer at 22 GHz (cut – off density nc = 6 10 cm ) has been designed, developed and used to measure the critical density layer and its fluctuations in Aditya. It can measure the plasma density fluctuations from r = 11 to 22 cm for central electron density 7.5 1012 cm-3 and more, respectively.
The output signal of reflectometer is analyzed and compared with the density measurement from the microwave interferometer. When the density measured by interferometer is constant, then the fluctuations of local density are seen from the reflectometer signal. Analysis of initial results show that density fluctuation at r = 21 cm in the main plasma has correlation time of about 8 sec and frequency spectrum is broad. Use of 22 GHz incident wave allows the observation of density fluctuation with wave number in the range of 0 – 9.2 cm-1 from the reflecting region at the receiving horn. Radial variation of the fluctuation level is observed from 5% to 22% for minor radius 11 to 22 cm, respectively.
References:
[1] TFR Group, “Local Density Fluctuations Measurements by Microwave Reflectometry on TFR,” Plasma Physics and Cont. Fusion, 27(11), 1299 (1985).
[2] W. A. Peebles, et. al., “Fluctuation Measurements in the DIII-D and TEXT tokamaks via Collective Scattering and Reflectometry”, Rev. Sci. Instrum., 61(11), 3509 (1990).
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_231
Helium Beam Diagnostics for the Estimation Electron Temperature and Density in SST-1
Vishal Pillai1, Neha Singh1, Jinto Thomas1, Rajesh Kumar Singh1, Hem Chandra Joshi1, Ajai Kumar1
1Institute for Plasma Research, India Email: [email protected]
Supersonic helium beam Diagnostics is used to estimate edge electron density and temperature in tokamaks [1] Ratio of line emission intensities from neutral helium is used to estimate electron temperature and density. Temperature is estimated from the ratio of intensities (728.1nm /706.3 nm) whereas density is estimated from ratio (668.1nm/728.1nm). We have designed and tested a supersonic helium beam injector for edge plasma temperature and density for SST-1 tokamak. The system consists of a supersonic injector and an imaging system. The emission is collected by the imaging system and optical fibers and an EMMCD coupled spectrograph is used to record the spectra from various spatial locations. The spatial resolution is around 5 mm.
In a recent campaign in SST-1, we tried to estimate these parameters using the residual helium after the helium GDC. The spectrometer and detection system was calibrated and signal was optimized. The spectra were good enough to use these helium lines to estimate electron temperature and density with an integration time of 10 ms. The observed line ratios are compared with the line ratios obtained from CR model/ Atomic Data and Analysis Structure (ADAS) to get an estimate of electron temperature and density. The estimated electron density is in the range of 51011 - 21012 cm-3 and electron temperature 30-55 eV. The obtained parameters provide reasonable estimates when compared with other diagnostics considering the diffusion and ionization of neutral helium inside the tokamak.
References:
[1] U. Kruezi, H. Stoschus, B. Schweer, G. Sergienko and U. Samm, “Supersonic helium Beam diagnostics for fluctuation measurements of electron temperature and density at the Tokamak Textor” Rev. Sci. Instrum., 83, 065107, 2012
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_239
Operation of ADITYA Thomson Scattering System: Measurement of Temperature and Density
Jinto Thomas1, Vishal Pillai1, Neha Singh1, Kiran Patel1, Lingeshwari G1, Zalak Hingrajiya1, Ajai Kumar1
1Institute for Plasma Research, India Email: [email protected]
ADITYA Thomson scattering (TS) system is a single point measurement system operated using a 10 J ruby laser and a 1 meter grating spectrometer. Multi-slit optical fibers are arranged at the image plane of the spectrometer so that each fiber slit collects 2 nm band of scattered spectrum. Each slit of the fiber bundle is coupled to high gain Photomultiplier tubes (PMT). Standard white light source is used to calibrate the optical fiber transmission and the laser light itself is used to calibrate the relative gain of the PMT. Rayleigh scattering has been performed for the absolute calibration of the TS system. The temperature of ADITYA plasma has been calculated using the conventional method of estimation (calculated using the slope of logarithmic intensity vs the square of delta lambda). It has been observed that the core temperature of ADITYA Tokamak plasma is in the range of 300 to 600 eV for different plasma shots and the density 2-3 1013/cc. The time evolution of the plasma discharge has been studied by firing the laser at different times of the discharge assuming the shots are identical. In some of the discharges, the velocity distribution appears to be non Maxwellian.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_240
Installation and Commissioning of SST-1 Thomson scattering system
Jinto Thomas1, Vishal Pillai1, Neha Singh1, Kiran Patel1, Vishnu K Chaudhari1, Ajai Kumar1
1Institute for Plasma Research, India Email: [email protected]
SST-1 Thomson scattering (TS) system is designed with 6 Nd:YAG lasers of 1.6 J energy each at the fundamental wavelength of 1064 nm. The 90 degree scattered photons are imaged to an array of optical fibers which transfer the photons for spectral dispersion and detection to a five channel interference filter polychromator. Avalanche photodiodes with nearly 3 mm active area are being used for the signal detection along with appropriate signal conditioning electronics and data acquisition. The commissioning of 6 laser system is in progress.
At present, a single laser based TS system with 1.6 J energy at 30 Hz has been commissioned in SST-1 with 5 spatial points having spatial resolution around 10 mm. Inter channel calibration of filter polychromator and absolute calibration of Thomson scattering system has been performed. We discuss the data from different calibrations performed for commissioning of a single laser based Thomson scattering system on SST-1.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_249
Limiter and Divertor Systems – Conceptual and Mechanical Design for Aditya Tokamak Upgrade
Kaushal Patel1, Kulav Rathod1, Kumarpalsinh A Jadeja1, Shailesh B Bhatt1, Deepti Sharma1, Srinivasan Radhakrishnana1, Raju Daniel1, Rakesh L Tanna1, Joydeep Ghosh1, Prabal K Chattopadhyay1, Yogesh C Saxena1, Aditya Team1
1Institute for Plasma Research, India Email:[email protected]
Existing Aditya tokamak with limiter configuration is being upgraded into a machine to have both the limiter and divertor configurations. Necessary modifications have been carried out to accommodate divertor coils by replacing the old vacuum vessel with a new circular section vacuum vessel [1]. The upgraded Aditya tokamak will have different set of limiters and divertors, such as (1) Safety limiter, (2) Toroidal Inner limiter, (3) outer limiter of smaller toroidal extent, (4) Upper and lower divertor plates. The limiter and divertor locations inside the Aditya tokamak upgrade are decided based on the numerical simulation of the plasma equilibrium profiles. Initially graphite will be used as plasma facing material (PFM) in all the limiter and divertor plates. The dimensions of the limiter and divertor tiles are decided based on their installation inside the vacuum vessel as well as on the total plasma heat loads (~ 1 MW) falling on them. Depending upon the heat loads; the thickness of graphite tiles for limiter and divertor plates is estimated. Shaped graphite tiles will be fixed on specially designed support structures made out of SS-304L inside the torus shaped vacuum vessel. In this paper mechanical structural design of limiter & divertor of Aditya Upgrade Tokamak is presented.
References:
[1] "Design of Vacuum Vessel for Aditya Upgrade Tokamak", S. B. Bhatt, et al. XXVI Int. Symp. on Discharge and Electrical Insulation in Vacuum. Mumbai, India-2014, Conference proceeding Page 681-684.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_250
Development of Gas Puffing System for LHCD Experiment in Aditya Tokamak
Kumarpalsinh A Jadeja1, Kaushik S Acharya1, Kaushal M Patel1, Nilesh D Patel1, Kalpesh M Chaudhary1, Shailesh B Bhatt1, Pramod K Sharma1, Kirankumar K Ambulkar1, Pramod R Parmar1, Chetan G Virani1, Saifali Dalakoti1, Arvindkumar L Thakur1, Rakesh L Tanna1, Santanu Banerjee1, Joydeep Ghosh1
1Institute for Plasma Research, India Email: kumarpal@ ipr.res.in
Lower hybrid (LH) wave coupling experiments have been successfully carried out in Aditya tokamak using 120 kW, pulsed LHCD system based at 3.7 GHz [1]. To enhance the coupling of LH waves in the edge plasma region, an especially designed gas puffing system is developed to inject Hydrogen gas from the electron side of the grill antenna. The developed new gas puffing system consists of a multi-hole gas injection manifold with precisely fabricated holes. The dimensions of the manifold are determined so as to spread the gas uniformly in front of antenna. We achieved precise control of neutral gas injection near the antenna by this new gas puffing system of LHCD as observed by the images taken by fast camera. The gas puff using the manifold near the LH antenna led to considerable reduction in the reflection co-efficient of LH power indicating enhance absorption in plasma. The number of particles injection through gas puffing system has been estimated to figure out the optimum LH power coupling in Aditya tokamak. This paper presents detail of the developed gas puffing system for LHCD experiments and its implication on LHCD experiments.
References:
[1] Sharma, P. K., S. L. Rao, D. Bora, R. G. Trivedi, et. al., Fusion Engg. & Design, 82, 41, 2007.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_251
Structural Analysis of New Vacuum Vessel for Aditya Tokamak Upgrade
Kulav Rathod1, Joydeep Ghosh1, Shailesh B Bhatt1, Rakesh L Tanna1, Kumarpalsinh A Jadeja1, Kaushal M Patel1
1Institute for Plasma Research, India Email: [email protected]
The new toroidal-shaped vacuum vessel for Aditya Tokamak Upgrade is fabricated by joining two semi tori of circular cross section, equipped with as many as 115 ports of different sizes and shapes for pumping and diagnostics. Both semi tori are identical and are made up of stainless steel 304L. The major radius of toroidal chamber is 750 mm and minor radius is 305 mm. The vacuum vessel is subjected to different loads such as vacuum load and electromagnetic loads. As the vacuum level required inside the vessel is ~ 1 x 10-9 mbar, the vessel wall should sustain compressive forces due to atmospheric pressure from outside and should not deform. Hence, the wall thickness of the vessel wall has been chosen after carrying out the detailed stress analysis in ANSYS workbench. Meshing has been carried out using the method of Tetrahedron in the workbench. The maximum stress on vessel due to pressure difference comes out to be ~ 70 MPa. The maximum deformation for a wall thickness of 10 mm is ~ 0.45 mm. The vacuum vessel is also planned to be baked up to 150oC, and the maximum stress on vessel due to combined thermal load and vacuum load (10-9 mbar) becomes ~ 80 MPa and maximum deformation is 2.95 mm for 10 mm thick walls. As the yield strength of stainless steel 304L is 170 MPa, the stress generated by various load acting on vacuum vessel is under safety limit. Detailed design consideration thoroughly substantiated by ANSYS analysis for the new vacuum vessel of Aditya Tokamak Upgrade will be presented in this paper.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_267
IGBT Based Active Clamping Protection Scheme for SST-1 PF Coils
Azad Makwana1, Deven Kanabar1, Chiragkumar Dodiya 1, Kalpesh Doshi1, Yohan Khristi1, Subrata Pradhan1
1Institute for Plasma Research, India Email: [email protected]
Steady State Superconducting Tokomak (SST-1) [1] is a medium size Tokamak in operation at the Institute for Plasma Research, India. In SST-1, during the central solenoid discharge, voltage is induced in the PF coils due to magnetic coupling with central solenoid. Induced voltage in PF coils has to be within safe limit of coil insulation capability. To restrict the induced voltage, the central solenoid current profile needs to be regulated during the plasma experiments. A novel concept of active IGBT clamping is introduced to limit induced voltage in PF coils. In this scheme, whenever induced voltage of PF coils crosses the predefined level, resistor is inserted dynamically in parallel resistive network using the fast IGBT switches. It will reduce loop resistance and clamp the high voltage spike across coils. Initially scheme was tested and validated in lab-scale prototype consisting of capacitor charging circuit and transformer. Subsequently, concept of active IGBT clamping has been successfully implemented for PF3 coils during the SST-1 plasma shots.
References:
[1] S. Pradhan, Z. Khan, V.L.Tanna et al, “The first experiment in SST-1”, Nuclear Fusion, 55, (2015).
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_268
Thermal Imaging of SST-1 Limiters
Shwetang N Pandya1, Santosh P Pandya1, Kumar Ajay1
1Institute for Plasma Research, India Email: [email protected]
The main power loss channels from plasma are impurity radiation, charge exchange neutrals and transport losses. The radiation and charge-exchange losses are estimated by the bolometric and Neutral Particle Analyzer measurements respectively. The measurement of power losses through convection and conduction is carried out by thermal imaging of the Plasma Facing Components (PFCs) which are heated due to the plasma surface interaction. SST-1 is a medium sized tokamak with graphite PFCs. Two sets of poloidal limiters, separated toroidally by 180°, are used during the limiter phase of the machine operation. The plasma-surface interaction is monitored and studied using thermal imaging cameras. The spatio-temporal temperature evolution monitored by this thermal imager is processed to estimate the heat loads on the SST-1 limiters. Thermographic observations carried out during recent SST-1 experimental campaigns are reported here and power drawn by the limiters is estimated. This estimate will serve as a valuable input while budgeting the global power balance.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_271
The Upgradation of Aditya Tokamak
Shailesh B Bhatt1, Joydeep Ghosh1, Rakesh L Tanna1, Chhaya Chavda1, Chet Narayan Gupta1, Prabal K Chattopadhyay1, Raju Daniel1, Srinivasan Radhakrishnana1, Kaushik S Acharya1, Kalpesh M Chaudhary1, Someswar Dutta1, Kumarpalsinh A Jadeja1, Madan B Kalal1, Sanjay V Kulkarni1, Kumari (K) Krishna1, Moti Makwana1, Rohitkumar Panchal1, Vipul K Panchal1, Kaushal M Patel1, Narendra Patel1, Nilesh Patel1, Sharvil Patel1, Vijay Patel1, Harshita Raj1, Ramasubramanian Narayanan1, Vaibhav Ranjan1, Kulav Rathod1, Devraj H Sadharkiya1, Kunal Shah1, Krishnamachari Sathyanarayana1, Deepti Sharma1, Pramod K Sharma1, Braj Kishore Shukla1, A Varadharajulu1, Dinesh S Varia1, Ajai Kumar1, Ratneshwar Jha1, Amita Das1, Abhijeet Sen1, Yogesh C Saxena1, Predhiman Krishan Kaw1, Dhiraj Bora1
1Institute for Plasma Research, India Email: [email protected]
Aditya Tokamak is the first Indian tokamak, indigenously built and commissioned at the Institute for Plasma Research, Gandhinagar, Gujarat, India, in September, 1989. Aditya Tokamak has been in operation since more than 25 years. More than 30,000 discharges are taken and a large number of experiments are carried out, with plasma current ranging from 50 KA to 150 KA, lasting for 100 to 250 milliseconds. Various types of wall conditioning techniques and different hot plasma diagnostics are tested and operated on Aditya Tokamak. The experiments for turbulent particle transport and turbulence in the edge plasma, gas puffing, lithium coating, mitigation, plasma disruption, limiter and electron biasing, runaway discharges etc. led to many interesting results contributing immensely to the world of thermonuclear fusion. Experiments on Pre-ionization and Plasma heating by ICRH and ECRH are also worked out.
The worldwide effort on magnetic fusion is devoted to the present generations of large tokamaks like DIII-D, TCV, EAST, SST-1 etc., which are operational emphasizing on divertor and tungsten wall ITER-like operation. There are very few small / medium-sized tokamaks operational around the world with divertor facility and technical capabilities to provide able support for operation and trouble shooting of these big tokamaks. The high-risk experiments, such as studying disruptions and runaways can be carried out in these small machines as they are not desirable in bigger machines and manpower to run and operate bigger tokamaks can be easily trained in the smaller machines. Hence, converting and upgrading the existing small / medium-sized tokamaks with limiters into more state of art facilities like divertor operation, tungsten first wall, good plasma control, are the need of hour to provide able support for the existing big tokamaks and for the future tokamaks. Therefore, after a long successful operation of Aditya Tokamak, it has been planned to upgrade the existing Aditya tokamak into a state of art machine with divertor operation and very good plasma control to support the future Indian Fusion program in a big way.
The scientific objectives of Aditya tokamak Upgrade include Low loop voltage plasma start-up with strong pre-ionization having a good plasma control system. The upgrade is designed keeping in mind the experiments, disruption mitigation studies relevant to future fusion devices, runway mitigation studies, demonstration of Radio-frequency heating and current drive etc. This
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10th Asia Plasma & Fusion Association Conference upgraded Aditya tokamak will be used for basic studies on plasma confinement and scaling to larger devices, development and testing of new diagnostics etc. This machine will be easily accessible compared to SST-I and will be very useful for generation of technical and scientific expertise for future fusion devices. In this paper, especial features of the upgrade including various aspects of designing of new components for Aditya Upgrade tokamak is presented.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_277
Development of Non-circular Metal Seal for Aditya Tokamak Upgrade Vacuum Vessel
Kaushik S Acharya1, Kaushal M Patel1, Kumarpalsinh A Jadeja1, Kulav Rathod1, Nilesh D Patel1, Kalpesh M Chaudhary1, Shailesh B Bhatt1, Aditya Team1
1Institute for Plasma Research, India Email: [email protected]
Existing Aditya Tokamak is being upgraded into a machine with divertor operation. To accommodate divertor magnet coils, existing vacuum vessel will be replaced with new circular section vacuum vessel having volume of ~1.5 m3 [1]. This vacuum vessel is fabricated by SS 304L and can be baked upto 150oC. The ultimate vacuum envisaged in the vessel is ~10-9 torr. The vacuum vessel has 112 ports opening of various sizes and shapes, viz. circular, rectangular and triangular types. The circular ports are vacuum sealed using CF metal seal, while the non- circular ports are sealed using metal wire-seals. Customized shaped aluminium wire seals are designed and fabricated for new vacuum vessel. The designed and fabricated aluminium wire seals are tested on local set up in laboratory to confirm its validation as appropriate metal seal for new vacuum vessel for Aditya Tokamak Upgrade. The challenging task of achieving a leak rate less than ~10-9 torr-l/s with baking upto 150oC is successfully accomplished on the test bench. The same wire-seals are then successfully used in Aditya Upgrade vessel achieving a base vacuum ~ 10-9torr. The flanges with wire seals are required to be tightened specific torque range (25 – 35 N-m) to obtain optimum symmetrical sealing. The wire seals are fabricated in-house using butt welding machine and the stiffness of joints are checked using radiography. This paper presents design, fabrication technique and test results of the wire-seals successfully used in ultra- high vacuum vessel of Aditya Upgrade.
References:
[1] "Design of Vacuum Vessel for Aditya Upgrade Tokamak", S. B. Bhatt, et al. XXVI Int. Symp. on Discharge and Electrical Insulation in Vacuum. Mumbai, India-2014, Conference proceeding Page 681-684.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_298
Study of the plasma SOL with fast reciprocating probe diagnostics on the SST-1 tokamak
M V Gopalakrishna1, , Sameer Kumar Jha1, , Santanu Banerjee1, Manoj Kumar Gupta1, Pramila Gautam1, Dilip Raval1, Snehal Jaiswal1, Pradeep Chauhan1, Subrata Pradhan1, SST-1 Team1
1Institute for Plasma Research, India Email:[email protected]
A reciprocating probe drive system has been designed, fabricated and successfully installed at the bottom port of Steady State Superconducting Tokamak (SST-1). The probe system has been designed to measure the local plasma parameters such as temperature (1 eV to 50 eV range), density (up to ~ 1018 m-3) and floating potential (~100V) near the lower X-point of the plasma column at the plasma current flat top. The probe head can move a total distance of 390 mm from its reference position during plasma shot with a combination of two pneumatic cylinders (slow and fast) and edge welded bellows. Slow movement is achieved from rest position to reference position (200mm) in 2sec. From the reference position, the fast movement over 190 mm of length is made in 300 ms. A programmable logic controlling (PLC) system records the number of scan and delay with reference to loop voltage. Timing between the scans is synchronized with the of SST-1 control system sequence. The density at 370 mm below the mid plane is measured to be 0.3-1 1011 cm-3 at a bias voltage of – 70 V. Interaction of the plasma with the probe tip and the probe movement during a plasma shot can be traced with the fast visible imaging in SST- 1. The measured density and probe-plasma interaction will be correlated with the radiated power measured using bolometer diagnostics. Density fluctuations and radial electric field at the scrape off layer (SOL) and their implications on plasma performance will be reported. Further, these signals will also serve as an input for power balance studies.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_303
Conceptual design of Plasma position control of SST-1 Tokamak using vertical field coil
Hitesh Kumar Gulati1, Kiritkumar B Patel, Jasraj Dhongde1, Kirti Mahajan1, Aveg Kumar1, Harish Masand1, Manisha Bhandarkar1, Hiteshkumar Chudasama1, Subrat Jana, Chet Narayan Gupta, Subrata Pradhan1
1Institute for Plasma Research, India [email protected]
SST-1 (steady state superconducting Tokamak) is a plasma confinement device in Institute for Plasma Research (IPR) India. SST-1 has been commissioned successfully and has been carrying out plasma experiments since the beginning of 2014 achieved a maximum plasma current of 75 kA at a central field of 1.5 T and the plasma duration ~ 500 ms. SST-1 looks forward to carrying out elongated plasma experiments and stretching plasma pulses beyond 1s.
Based on the solution of Grad–Shafranov equation the shift of plasma column center from geometrical centre of vacuum chamber is measured using various magnetic probes and flux loops installed in the machine. The closed feedback loop uses plasma current (Ip), Delta R as feedback signal and manipulate the vertical field current (Ivf). The discharge starts with feed forward loop using initially provided reference then the active feedback starts after discharge by few msec once plasma column is completely formed. The feedback loop time is of the order of 10 msec.
The primary objective is to acquire plasma position control related signals, compute plasma position and generate position correction signal for VF coil power supply, communicate correction to VF coil power supply and modify VF power supply output in a deterministic time span.
In this we present the methodology used for plasma horizontal displacement control using vertical field and discuss the preliminary results.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_304
Implementation of SST-1 plasma position control using vertical field
Kirti Mahajan1, Jasraj Dhongde1, Kiritkumar B Patel1, Hitesh Kumar Gulati1, Aveg Kumar1, Harish Masand1, Manisha Bhandarkar1, Hiteshkumar Chudasama1, Subrat Jana, Chet Narayan Gupta, Subrata Pradhan1
1Institute for Plasma Research, India [email protected]
SST-1 (steady state superconducting Tokamak) is a plasma confinement device in Institute for Plasma Research (IPR) India. SST-1 has been commissioned successfully and has been carrying out plasma experiments since the beginning of 2014 and achieved a maximum plasma current of 75 kA at a central field of 1.5 T and the plasma duration ~ 500 ms.After commissioning of first wall components SST-1 looks forward to carrying out elongated plasma experiments and stretching plasma pulses beyond 1sec.
The plasma is very unstable in nature. In order to confine it for a longer duration various controls need to be act upon simultaneously. The most important control is the position control. One of the means to control plasma position is by adjusting Vertical Field (VF) which forces plasma to remain in centre of the Tokamak.
The SST-1 Plasma control system is a distributed real-time system based on VME architecture. The plasma position and current are computed in real time on Digital Signal Processor (DSP) module and then transmitted to VF power supply controller over Reflective Memory (RFM) based data network where VF coil current is modified based on plasma position drift from the centre. The SST-1 recent campaign shows that real time control of VF enhanced the plasma duration by the order of few msec.
This paper focuses on the architecture and implementation aspects of the plasma position feedback control system and presents the initial results observed in recent SST-1 experiment campaign.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_6
Preparation of W/CuCrZr Monoblock Test Mock-up using Vacuum Brazing Technique
Kongkham Premjit Singh1, Samir S Khirwadkar1, Kedar Bhope1, Nikunj Patel1, Prakash K Mokaria1, Mayur Mehta1
1Institute for Plasma Research, India Email: [email protected]
Development of the joining for W/CuCrZr monoblock PFC test mock-up is an interest area in Fusion R&D [1-5]. W/Cu bimetallic material has prepared using OFHC copper casting approach on the radial surface of W monoblock tile surface. The W/Cu bimetallic material has been joined with CuCrZr tube (heat sink) material with the vacuum brazing route. Vacuum brazing of W/Cu- CuCrZr has been performed @ 970 C for 10 mins using NiCuMn-37 filler material under deep vacuum environment (10–6 mbar). Graphite fixtures were used for OFHC copper casting and vacuum brazing experiments [1]. The joint integrity of W/Cu-CuCrZr monoblock mock-up on W/Cu and Cu-CuCrZr has been checked using ultrasonic immersion technique. Micro-structural examination and Spot-wise elemental analysis have been carried out using HR-SEM and EDAX. The results of the experimental work will be discussed in the paper.
References:
[1] S.S. Khirwadkar et.al, Fusion Eng. Des., 86 (2011), pp. 1736-1740
[2] Pietro Appendino et.al, J. Nucl. Mater., 329–333 (2004), pp. 1563-1566
[3] V. Casalegno et.al, J. Nucl. Mater., 393 (2009), pp. 300-305
[4] M. Singh et.al, Mater. Sci. Eng. A, 412 (2005), pp. 123-128
[5] K. P. Singh et.al,Fusion Science and Technology, 65 (2014), pp.235-240
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Abstract ID: 1_12
Design and Performance of Vacuum System for High Heat Flux Test Facility
Rajamannar Swamy Kidambi1, Prakash K Mokaria1, Samir S Khirwadkar1, Sunil Belsare1, Mohammed Shoaib Khan1, Tushar Patel1, Deepu S Krishnan1
1Institute for Plasma Research, India Email: [email protected]
High heat flux test facility (HHFTF) at IPR is used for testing thermal performance of plasma facing material or components [1]. It consists of various subsystems like vacuum system, high power electron beam system, diagnostic and calibration system, data acquisition and control system and high pressure high temperature water circulation system. Vacuum system consists of large D-shaped chamber, target handling system, pumping systems and support structure. The net volume of vacuum chamber is 5m3 was maintained at the base pressure of the order of 10-6 mbar for operation of electron gun with minimum beam diameter [2]. Inorder to achieve the ultimate vacuum, turbo-molecular pump (TMP) and cryo pump are installed. Each TMP and cryo-pump unit has an electro-pneumatic gate valve of respective size to isolate the pump in the case of either vacuum break in the D-shaped chamber or in case of the pump failure to protect each in either condition. A variable conductance gate valve is used for maintaining required vacuum in the chamber. Initial pumping of the chamber was carried out by using suitable rotary and root pumps. PXI and PLC based faster real time data acquisition and control system is implemented for performing the various operations like remote operation, online vacuum data measurements, display and status indication of all vacuum equipments. This paper describes in detail the design and implementation of various vacuum subsystems including relevant experimental details.
References:
[1] Yashashri Patil, S. S. Khirwadkar, S. M. Belsare, Rajamannar Swamy, M. S. Khan, S.Tripathi, K. Bhope, D. Krishnan, P. Mokaria, N. Patel, I. Antwala, K. Galodiya, M. Mehta, T. Patel, “Performance of straight tungsten mono-block test mock-ups using new high heat flux test facility at IPR”, Fusion Engineering and Design, 84-90, 95 (2015)
[2] M. S. Khan, Rajamannar Swamy, S. Khirwadkar, “Conceptual design of vacuum chamber for testing of high heat flux components using electron beam as heat source,” Journal of Physics conference series, 012060, 390 (2012).
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_13
Thermal Shock Behavior of Tungsten & Tungsten Alloy Materials under Transient High Heat Load Conditions
Shailesh Kanpara1, Samir S Khirwadkar1, Sunil Belsare1, Kedar Bhope1, Rajamannar Swamy Kidambi1, Prakash K Mokaria1, Nikunj Patel1, Tushar Patel1, Narendra Chauhan1, Nirav Jamnapara1
1Institute for Plasma Research, India Email: [email protected]
Present paper is concerned with investigation of damaged studies of Plasma Facing Materials, different grades of Tungsten material under transient heat load conditions relevant to the ITER- like Divertor. Pure tungsten-reference material (Hot rolled), Pure Tungsten and Tungsten lanthanum (Direct Sintering Processed) have been tested under transient heat loads expected in ITER. These experiments were carried out using newly established High Heat Flux Test Facility (HHFTF) at the Institute for Plasma Research (IPR)-India using electron beam as a heat source. The targets were exposed by series repeated pulsed surface heat loads for 500 cycles in energy density range of 1.0–3.14 MJ/m2 and a pulse duration of 20 ms with 1 second off time. The crack formations and surface modification behaviors under transient heat load were investigated. Microstructural characterization clearly shows the large network of macro and micro crack in Tungsten-lanthanum with crack width of 15 micron (µm), and other two grades of Tungsten remain un-damaged. FEM simulation was carried out for thermo-mechanical stresses developed in exposed tungsten material during transient heat load events. Detailed characterization of the exposed sample for its various properties i.e., structural, microstructural and mechanical properties will be presented in the paper.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_15
Characterization of a Segmented Plasma Torch Assisted High Heat Flux (HHF) System for Performance Evaluation of Plasma Facing Components in Fusion Devices
Aomoa Ngangom1, Trinayan Sarmah1, Puspa Sah1, Joydeep Ghosh2 and Mayur Kakati1
1Centre of Plasma Physics - Institute for Plasma Research, India 2Institute for Plasma Research, India Email: [email protected]
A wide variety of high heat and particle flux test facilities are being used by the fusion community to evaluate the thermal performance of plasma facing materials/components, which includes electron beam, ion beam, neutral beam and thermal plasma assisted sources. In addition to simulate heat loads, plasma sources have the additional advantage of reproducing exact fusion plasma like conditions, in terms of plasma density, temperature and particle flux.
At CPP-IPR, Assam, we have developed a high heat and particle flux facility using a DC, non- transferred, segmented thermal plasma torch system, which can produce a constricted, stabilized plasma jet with high ion density. In this system, the plasma torch exhausts into a low pressure chamber containing the materials to be irradiated, which produces an expanded plasma jet with more uniform profiles, compared to plasma torches operated at atmospheric pressure.
The heat flux of the plasma beam was studied by using circular calorimeters of different diameters (2 and 3 cm) for different input power (5-55 kW). The effect of the change in gas (argon) flow rate and mixing of gases (argon + hydrogen) was also studied. The heat profile of the plasma beam was also studied by using a pipe calorimeter. From this, the radial heat flux was calculated by using Abel inversion. It is seen that the required heat flux of 10 MW/m2 is achievable in our system for pure argon plasma as well as for plasma with gas mixtures.
The plasma parameters like the temperature, density and the beam velocity were studied by using optical emission spectroscopy. For this, a McPherson made 1.33 meter focal length spectrometer; model number 209, was used. A plane grating with 1800 g/mm was used which gave a spectral resolution of 0.007 nm. A detailed characterization with respect to these plasma parameters for different gas (argon) flow rate and mixing of gases (argon+hydrogen) for different input power will be presented in this paper. The plasma temperature was measured to be around 0.2 – 2.5 eV and plasma density of about 1020/m3. The plasma jet velocity was measured to be around 1 to 1.5 km/sec.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_18
Performance of Impedance Transformer for High Power ICRF Heating in LHD
Kenji Saito1, Tetsuo Seki1, Hiroshi Kasahara1, Ryohsuke Seki1, Shuji Kamio1, Goro Nomura1, Takashi Mutoh1
1National Institute for Fusion Science, Japan Email: [email protected]
There are two types of ion cyclotron range of frequencies (ICRF) antennas in the Large Helical Device (LHD). These are referred to as Field-Aligned-Impedance-Transforming (FAIT) antennas and handshake form (HAS) antennas. The HAS antenna has high performance in the heating efficiency in minority ion heating at the 0- current phase. However, the loading resistance Rp 2 defined by Rp=2P(zc/Vmax) was small and the maximum injection power was limited by the voltage on the transmission line, where P is the injected power from antenna and Vmax is the maximum voltage on the transmission line with the characteristic impedance of zc. In LHD zc is 50 , and the interlock level of Vmax was set to 35 kV. The typical loading resistance of HAS antenna was only 2 . The maximum injection power calculated with the loading resistance and the interlock level is only 490 kW. FAIT antenna has a smaller antenna head than HAS antenna, however, it has higher loading resistance of typically 5 due to the optimized in-vessel impedance transformer between the antenna head and the feed-through. Voltage on the coaxial line limits the power to 1.2 MW.
In order to increase the loading resistance and decrease the maximum voltage on the transmission line, pre-matching is necessary. Pre-stub tuner is one of the candidates, but space is limited around the antenna port. In-vessel impedance transformer for FAIT antenna worked well. Therefore, we designed ex-vessel impedance transformer for HAS and FAIT antennas. They are designed to be inserted in the transmission line outside of the vacuum vessel close to ceramic feed-throughs. The diameter of the outer conductor is 241.2 mm, which is the same size as that of the transmission line, and the diameter of the inner conductor is 185.6 mm. This means that the characteristic impedance is 15.7 . The flange to flange length is 628 mm, and it is not enough for the perfect matching for the frequency of 38.5 MHz but it is effective for increasing the loading resistance.
Electromagnetic simulation was performed with HFSS in order to estimate the increment of loading resistance and the electric field which cause the breakdown. The estimated enhancement factors of loading resistances are 2.5 and 1.65 for FAIT and HAS antennas, respectively. The ex- vessel impedance transformers were attached to HAS antennas in 2014. The loading resistance was compared without and with the ex-vessel impedance transformer for the lower HAS antenna changing the distance between the antenna and the last closed flux surface. The upper antenna was turned off in order to avoid mutual coupling effect. The loading resistance was increased from 1.5 to 2 times with the ex-vessel impedance transformer, which agreed with the simulation. We also installed the ex-vessel impedance transformers for FAIT antennas in 2015. High power injection is expected with FAIT antennas owing to the increase of the loading resistance.
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Invited Talk (Session-3)
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_151
Progress of JT-60SA Construction and R&D of its Heating Systems
Yoshitaka Ikeda1
1Japan Atomic Energy Agency, Japan Email: [email protected]
The JT-60SA (JT-60 Super Advanced) project is a combined project of JAEA’s program for national use and JA-EU Satellite Tokamak Program collaborating with Japan and EU fusion community. The main objectives of the JT-60SA are to demonstrate steady-state high-beta plasma, and to support ITER through the optimization of ITER operation scenario. To attain these objectives, the JT-60SA is designed to be the superconducting tokamak with a wide range of diverted plasma configurations at the maximum plasma current of 5.5 MA. Powerful heating systems of total power of 41 MW (negative-ion NBI: 10MW at 500 keV, positive-ion NBI: 24 MW at 85 keV, ECRF: 7 MW at 110 GHz) for 100s is required to allow the JT-60SA to be operated in break-even equivalent conditions for a long pulse duration. The NBI and ECRF systems of JT-60 are being upgraded to increase the power and pulse duration up to 100 s.
Design and fabrication of JT-60SA components, shared by the EU and Japan, started in 2007. Assembly in the torus hall started in 2013, and welding work of the vacuum vessel sectors is currently ongoing on the cryostat base. Other components such as TF coils, PF coils, power supplies, cryogenic system, cryostat vessel, thermal shields and so on were or are being delivered to the Naka site for installation, assembly and commissioning towards the first plasma in 2019. In parallel with this construction activity, developments of the heating systems have been remarkably progressed. To realize the 500keV negative-ion NBI on JT-60SA, long pulse negative ion production and high voltage holding capability have been independently developed. The long pulse production of negative ion beams has achieved 100 s at the beam current of 15 A by modifying the JT-60 negative ion source. This beam current is 68% of the target of JT-60SA (22 A). The reliable voltage holding on the accelerator was achieved up to 500 kV by adjusting the acceleration gap, where the effects of the surface area and the number of multi-apertures on the large accelerator (diameter of ~2 m) are taken into account. In ECRF, oscillations at 1 MW for 100 s as the development target of the JT-60SA gyrotron were achieved at both 110 GHz and 138 GHz in 2014. In addition, 82 GHz oscillation was achieved at 1.0 MW for 1 sec on this gyrotron in 2015. This additional frequency would be applicable to plasma start-up assistance and wall conditioning at the fundamental EC resonance in JT-60SA.
In the conference, these technical progress on construction of JT-60SA jointly by European and Japanese fusion communities, as well as progress of the development of its heating systems, will be presented.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_248
Optimization, Commissioning and Operation of EAST Tungsten Divertor
Damao M Yao1, G N Luo1, L Cao1, Z B Zhou1, Q Li1, W J Wang1, L Li1, P F Zi1
1Institute of Plasma Physics Chinese Academy of Sciences, China Email: [email protected]
The EAST tungsten divertor was designed and manufactured in 2012-2014 shut down. First commissioning is during EAST 2014 plasma summer operation due May to July. Some weak points exposed and brought damages on some divertor modules. Reasons were analized and optimization was made. Around half year spent for analysis, divertor modules structure optimized manufacturing and recostruction.
The optimized divertor operated during 2015 summer plasma operation and demonstrate optimizations are efficiency. There is no issue for tungsten divertor during operation. EAST plasma heating power increased step by step and will up to 20MW in 2015 winter campaign plasma operation and will validate tungsten divertor heat exausting capability
References:
[1] D M. Yao, G. N. Luo, L, Cao et al., SOFE 2013 San Francisco US
[2] D.M. Yao, L. Cao, Z.B. Zhou et al., SOFT 2014 Sebastian Spain
[3] D.M. Yao, G.N. Luo, L. Cao, et al, PFMC 2015 Aix-en-Provance France
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_206 Status of the WEST Project
Jérôme Bucalossi1, WEST Team1
1CEA, DSM, IRFM, France Email: [email protected]
Power exhaust has been identified as a challenge for ITER and a potential showstopper on the roadmap towards fusion energy. Reliable power exhaust requires a thorough integration of physics and technology. The WEST project main objective is the minimization of risks for ITER divertor plasma facing components (PFC) construction and operation. The WEST project also fills the gap on long pulse tokamak operation in the European fusion program. It offers a readily available integrated tokamak environment for ITER but also at a later stage for DEMO divertor testing.
The assessment of ITER PFC performance and lifetime as well as innovative PFC under relevant power fluxes and particle fluence is the central thrust of the WEST program. Other issues including operation at high radiated fraction in compact divertor geometry, demonstration of detachment control over long pulse, exhaust physics at large aspect ratio and operation in double null are key topics which will be also tackled in the perspective of the fusion reactor.
The WEST project consists in the transformation of the French Tore Supra facility into a diverted tokamak with ITER-like divertor PFC. The limited Tore Supra circular cross section plasma is turned into D-shape diverted plasma by the addition of two poloidal field coils inside the upper and lower region of the vacuum vessel. The carbon environment is changed into a tungsten environment with a set of new actively cooled plasma facing components. 9 MW of ICRH and 7 MW of LHCD will provide relevant heat and particle load conditions on the divertor PFC over duration up to 1000s. A new infrared thermography system, together with embedded temperature sensors and calorimetry of the cooling circuits will ensure PFC protection and accurate power balance. A new visible spectroscopy system will monitor all potential tungsten sources.
The WEST assembly phase has started in October 2014. The manufacturing of the divertor coils casing, conductors and supporting structures is now completed and the assembly of the divertor coils inside the vacuum vessel is underway. First plasmas are expected in 2016. The WEST platform will be run as a user facility, open to the EUROfusion Consortium and all ITER partners. Major fusion partners have already demonstrated their interest for WEST and participate to the construction.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_237
Recent Advancement in Research and Planning toward High Beta Steady State Operation in KSTAR
Hyeon Keo Park1,2, S Hong1, D Humphreys3, Y K In1, Y M Jeon1, J G Kwak1, J M Kwon1, Y U Nam1, Y K Oh1, J K Park4, S Sabbagh5, S J Wang1, S W Yoon1, G S Yun6, KSTAR Team
1National Fusion Research Institute, Daejeon, Korea 2Ulsan National Institute of Science and Technology, Ulsan, Korea 3General Atomics, San Diego, USA 4 Princeton Plasma Physics Laboratory, Princeton, USA 5Columbia University, New York, USA 6Pohang University of Science and Technology, Pohang, Korea Email:[email protected]
The goal of Korean Superconducting Tokamak Advanced Research (KSTAR) research is to explore stable improved confinement regimes and technical challenge for superconducting tokamak operation and thus, to establish the basis for predictable high beta steady state tokamak plasma operation. To fulfil the goal, the current KSTAR research program is composed of three elements: 1) Exploration of anticipated engineering and technology for a stable long pulse operation of high beta plasmas including Edge Localized Mode (ELM) control with the low n (=1, 2) Resonant Magnetic Perturbation (RMP) using in-vessel control coils and innovative non- inductive current drives. The achieved long pulse operation up to ~50s and fully non-inductive current drive will be combined in the future. Study of efficient heat exhaust will be combined with an innovative divertor design/operation. 2) Exploration of the operation boundary through establishment of true stability limits of the harmful MagnetoHydroDynamic (MHD) instabilities and confinement of the tokamak plasmas in KSTAR, making use of the lowest error field and magnetic ripple simultaneously achieved among all tokamaks ever built. The intrinsic machine error field has a long history of research as the source of MHD instabilities and magnetic ripple is known to be a cause of energy loss in the plasma. The achieved high beta discharges at N ~4 and stable discharges at q95 (~2) will be further improved. 3) Validation of theoretical modeling of MHD instabilities and turbulence toward predictive capability of stable high beta plasmas. In support of these research goals, the state of the art diagnostic systems, such as Electron Cyclotron Emission Imaging (ECEI) system in addition to accurate profile diagnostics, are deployed not only to provide precise 2D/3D information of the MHD instabilities and turbulence but also to challenge unresolved physics problems such as the nature of ELMs, ELM-crash dynamics and the role of the core current density in the sawtooth crash. Work is supported by NRF of Korea (grant No. NRF- 2014 M1A 7A1A03029865)
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10th Asia Plasma & Fusion Association Conference
Invited Talk (Session-4)
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_287
Progress of Experiment on HL-2A
Xuru Duan1, HL-2A Team1
1Southwestern Institute of Physics, China Email: [email protected]
In recent years, HL-2A tokamak has experienced a series of upgrades on heating and diagnostic systems. It is now equipped with 5MW ECRH, 3MW NBI, and 2MW LHCD; New diagnostics such as motional stark effect (MSE), far infrared laser interferometer, charge exchange recombination spectroscopy (CXRS), electron cyclotron emission imaging (ECEI), microwave imaging reflectometer (MIR), scintillator-based lost fast-ion probe (SLIP), etc., have also been installed.
Physics experiment on the HL-2A has also progressed substantially. Two types of limit-cycle- oscillations (LCOs) were observed in the intermediate phase (I-phase), which indicated a second type of predator-prey process between turbulence and pressure gradient in addition to the conventional predator–prey involving zonal flows and turbulence. Kink-type MHD mode crash was found to play a crucial role in triggering H-mode through the increase of the edge pressure gradient and E × B flow shear. Series of I-H-I transitions were also found to be caused by impurity concentration in the pedestal region. Besides, an electromagnetic oscillation with the frequency of 50–100 kHz was found to be associated with pedestal density gradient saturation, and help to realize the ELM-free H-mode. For the first time, two types of magnetic fluctuations with n = 0 were identified to be generated through the nonlinear coupling between Alfvén eigenmodes (AEs) and low-frequency MHD modes. Up- and down-sweeping reverse shear Alfvén eigenmodes (RSAEs) were observed experimentally. It was found that fishbone could transit from/to LLM and even trigger tearing modes (TMs). For non-local transport, key characteristics of enhanced avalanches in the theory of self-organized criticality (SOC) were identified, including high Hurst exponents and large-scale radial propagation. Experimental results also revealed a close correlation between NTMs and non-local transport. For the first time, LHCD passive active module (PAM) antenna was tested and studied quantitatively in H-mode plasmas. Important progress has also been made in controlling neoclassical tearing modes (NTM) by using ECRH.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_293
Recent Progress and Present Status of LHD towards Deuterium Experiment
Tomohiro Morisaki1, M Osakabe1, M Y okoyama1, R Sakamoto1, K Y Watanabe1, K Nagasaki2, M Sakamoto3, S Inagaki4, S Hamaguchi1, M Isobe1
1National Institute for Fusion Science, Japan 2Institute of Advanced Energy, Kyoto University, Japan 3Plasma Research Center, University of Tsukuba, Japan 4Research Institute for Applied Mechanics, Kyushu University, Japan Email: [email protected]
Finalization of the hydrogen experiments towards the deuterium experiment are going on in LHD, together with the preparation of the hardware. In order to see the effect of deuterium, it is crucially important to know the ability of the hydrogen plasma before the deuterium experiment starts. Some trials to extend the parameter regime have been performed in the last experimental campaign.
As for the plasma heating, the mega-watt-class gyrotron for the electron cyclotron heating (ECH) has been developed in the collaboration program with University of Tsukuba. Two 154 GHz tubes have recently been installed in LHD, which has increased the total power of ECH up to 5.4 MW. In the experiment, fine tuning of antennas to inject microwaves was also performed to optimize the power deposition on the resonant surface, using a newly developed ray-tracing code. The upgrade of the ECH scheme resulted in the achievement of the central electron temperature of 10 keV with the averaged electron density of 2 1019 m-3. Simultaneous achievement of high ion temperature with high electron temperature was also achieved. Superimposing the ECH on NB heated plasma, central ion and electron temperatures, Ti and Te, of 6.0 keV and 7.6 keV were obtained, respectively. Especially, electron temperature increased in the core region, forming the internal transport barrier. On the other hand, for ions, improvement in Ti gradient can only be observed in the edge region. To explain the experimental result, transport analyses are being performed. Effect of ion species on high-Ti discharges were also investigated, changing the helium ion concentration in the hydrogen plasma. It was qualitatively observed that Ti tends to increase with the increase in the helium concentration. On the other hand, little dependence of Te on helium concentration is seen. However we should be quite careful when we discuss the experimental result. Detailed analyses are necessary to know the reason for the Ti increase in the helium-mixed plasma.
In the conference, present status of the preparation for the deuterium experiment will be presented, focusing on the diagnostics. The experimental plan and issues to be explored will also be discussed.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 0_300
Initial Results in SST-1 After Up-gradation
Subrata Pradhan1, Ziauddin Khan1, Vipul L Tanna1, Dilip Raval1, Upendra Prasad1, Harish Masand1, Aveg Kumar1, Kiritkumar B Patel1, Manisha Bhandarkar1, Jasraj Dhongde1, Braj Kishore Shukla1, Imran Mansuri1, Yohan Khristi1, Yuvakiran Paravastu1, Chet Narayan Gupta1, Dinesh Sharma1, Kalpeshkumar R Dhanani1, Pratibha Semwal1, Siju George1, Subrata Jana1, Pradip Panchal1, Rohitkumar Panchal1, Rakeshkumar Patel1, Hitesh Kumar Gulati1, Kirti Mahajan1, Mohammad Shoaib Khan1, Prashant Thankey1, Azadsinh Makwana1, Gaurang Mehsuriya1, Pradeep Chauhan1, Arun Parkash A1, Murtuza Vora1, Akhilesh Singh1, Dashrath Sonara1, Pankaj Varmora1, G Srikanth1, Dikens Christian1, Atul Garg1, Arun Panchal1, Nitin Bairagi1, Manika Sharma1, Gattu Ramesh Babu1, Prosenjit Santra1, Tejas Parekh1, Hiteshkumar Patel1, Prabal Biswas1, Snehal Jayswal1, Tusharkumar Raval1, Hiteshkumar Chudasama1, Atish Sharma1, Amit Ojha1, Bhadresh R Praghi1, Moni Banaudha1, Ketan Patel1, Hiren Nimavat1, Pankil Shah1, Jayant C Patel1, Rajiv Sharma1, A Varadharajulu1, Ranjana Manchanda1, Parveen Kumar Atrey1, Surya Kanth Pathak1, Y Sankar Joisa1, Kumudni Tahiliani1, Manoj Kumar1, Santanu Banerjee1, Debashis Gosh1, Bhoomi Chaudhary1, Amita Das1, Dhiraj Bora1
1Institute for Plasma Research, India Email: [email protected]
Steady State 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 ~ 4500 high heat flux compatible graphite tiles being assembled and installed on 136 Cu-alloy heat sink back plates engraved with ~ 4 km of leak tight baking and cooling channels in five major sub groups equipped with ~ 400 sensors and weighing ~ 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 2015, initial plasma experiments in up-graded SST-1 have begun since Aug 2015 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 (qa~3.8, BT=1.5 T) with a current ramp rates ~ 1.2 MA/s over a duration of ~ 300 ms with line averaged densities ~ 0.8 1019 per cc and temperatures ~ 400 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. This paper will elaborate the salient features of the SST-1 up-gradation spectrum, the subsequent engineering validations and important aspects of the first results in up- graded SST-1 apart from the immediate future experimental plans in SST-1.
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Abstract ID: 0_198
WEST Physics Basis
C Bourdelle1, J F Artaud1, Vbulandi Basiuk1, M Bécoulet1, S Brémond1, J Bucalossi1, H Bufferand1, G Ciraolo1, L Colas1, Y Corre1, X Courtois1, J Decker2, L Delpech1, P Devynck1, G Dif-Pradalier1, R P Doerner3, D Douai1, R Dumont1, A Ekedahl1, N Fedorczak1, C Fenzi1, M Firdaouss1, J Garcia1, P Ghendrih1, C Gil1, G Giruzzi1, M Goniche1, C Grisolia1, A Grosman1, D Guilhem1, R Guirlet1, J Gunn1, P Hennequin4, J Hillairet1, G T Hoang1, F Imbeaux1, I Ivanova- Stanik5, E Joffrin1, A Kallenbach6, J Linke7, T Loarer1, P Lotte1, P Maget1, Y Marandet8, M L Mayoral9,10, O Meyer1, M Missirlian1, P Mollard1, P Monier-Garbet1, P Moreau1, E Nardon1, B Pégourié1, Y Peysson1, R Sabot1, F Saint-Laurent1, M. Schneider1, J M Travère1, E Tsitrone1, S Vartanian1, L Vermare4, M Yoshida11, R Zagorski5, JET contributors*
1 CEA, IRFM, F-13108 Saint-Paul-lez-Durance, France. 2 Centre de Recherches en Physique des Plasmas, Ecole Polytechnique Fédérale de Lausanne, Switzerland 3 Center for Energy Research, University of California in San Diego, USA 4 Ecole Polytechnique, LPP, CNRS UMR 7648,91128 Palaiseau, France 5 Institute of Plasma Physics and Laser Microfusion, Warsaw, Poland 6 Max Planck Institute for Plasma Physics, Boltzmannstr 2, D-85748 Garching, Germany. 7 Forschungszentrum Jülich, D-52425 Jülich, Germany. 8 Aix-Marseille Université, CNRS, PIIM, UMR 7345, 13013 Marseille, France 9 CCFE, Culham Science Centre, Abingdon, Oxon, OX14 3DB, UK 10 EUROfusion Programme Management Unit, D-85748 Garching, Germany 11 Japan Atom Energy Agency, Naka, Ibaraki, Japan. *UK See the Appendix of F. Romanelli et al., Proceedings of the 25th IAEA Fusion Energy Conference 2014, Saint Petersburg, Russia. EUROfusion Consortium, JET, Culham Science Centre, Abingdon, OX14 3DB Email: [email protected]
With WEST (Tungsten (W) Environment in Steady State Tokamak) [1], the Tore Supra facility and team expertise [2] is used to pave the way towards ITER divertor procurement and operation. It consists in implementing a divertor configuration and installing ITER-like actively cooled tungsten monoblocks in the Tore Supra tokamak, taking full benefit of its unique long pulse capability. WEST is a user facility platform, open to all ITER partners.
This paper describes the physics basis of WEST: the estimated heat flux on the divertor target, the planned heating schemes, the expected behaviour of the L-H threshold and of the pedestal the potential W sources. A series of operating scenarios has been modelled, showing that ITER relevant heat fluxes on the divertor can be achieved in WEST long pulse H mode plasmas [3].
References:
[1] J. Bucalossi et al., Fusion Engineering and Design 89 (2014) 907–912 [2] R.J. Dumont et al., Plasma Phys. Control. Fusion 56 (2014) 075020 [3] C. Bourdelle et al, Nuclear Fusion volume 55 (2015) 063017
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Poster Session-2
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Abstract ID: 1_35
Indigenously Developed Large Pumping Speed Cryoadsorption Cryopump
Ranjana Gangradey1, Samiran Shanti Mukherjee1, Jyoti Agarwal1, Manoahstephen Manuelraj1, Paresh Panchal1, Pratik Kumar Nayak1, Jyoti Shankar Mishra1, Vrushabh Lambade1, Pawan Bairagi1, Vijay Kumar1, Reena Sayani1, Srinivasan Kasthurirengan1, Swarup Udgata2, Vijay Shankar Tripathi2
1Institute for Plasma Research, India 2I-Design Engineering Solutions Ltd., India Email: [email protected]
Indigenous cryoadsorption cryopump with large pumping speeds for fusion reactor application has been developed at the Institute for Plasma Research (IPR). Towards its successful realization, technological bottlenecks were identified, studied and resolved. Hydroformed cryopanels were developed from concept leading to the design and product realization with successful technology transfer to the industry. This has led to the expertise for developing hydroformed panels for any desired shape, geometry and welding pattern. Activated sorbents were developed, characterized using an experimental set up which measures adsorption isotherms down to 4K for hydrogen and helium. Special techniques were evolved for coating sorbents on hydroformed cryopanels with suitable cryo-adhesives. Various arrangements of cryopanels at 4 K surrounded by 80 K shields and baffles (which are also hydroformed) were studied and optimized by transmission probability analysis using Monte Carlo techniques. CFD analysis was used to study the temperature distribution and flow analysis during the cryogen flow through the panels. Integration of the developed technologies to arrive at the final product was a challenging task and this was meticulously planned and executed. This resulted in a cryoadsorption cryopump offering pumping speeds as high as 50,000 to 70,000 l/s for helium and 1,50,000 l/s for hydrogen with a 3.2 m2 of sorbent panel area.
The first laboratory scale pump integrating the developed technologies was a Small Scale CryoPump (SSCP-01) with a pumping speed of 2,000 l/s for helium. Subsequently, Single Panel CryoPump (SPCP-01) with pumping speed 10,000 l/s for helium and a Multiple Panel CryoPump (MPCP-08) with a pumping speed of 70,000 l/s for helium and 1,50,000 l/s for hydrogen respectively were developed. This paper describes the efforts in realizing these products from laboratory to Industrial scales.
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Abstract ID: 1_36
Indian Single Pellet injection System for Plasma Fuelling Studies
Ranjana Gangradey1, Jyotishankar Mishra1, Samiran Shanti Mukherjee1, Paresh Panchal1, Pratik Kumar Nayak1, Hardik Sharma1, Haresh Patel1, Pramit Dutta1, Naveen Rastogi1, Jyoti Agarwal1
1Institute for Plasma Research, India Email: [email protected]
A single barrel hydrogen pellet injection system is developed at Institute for plasma research (IPR), India. The injector is able to produce 1.6 mm length × 1.8 mm diameter pellets. The achieved velocity of pellet is in the range of 700 to 900 m/s and is controlled by regulating the propellant pressure. The size and speed of pellet are decided by considering the neutral gas shielding model (NGS) based calculations.
The injector is an in-situ pipe gun type injector, in which, a solid hydrogen pellet is formed at the freezing zone maintained at a temperature < 10 K and is accelerated to high speed using high pressure propellant gas. A GM cycle based cryocooler is used to maintain temperature at freezing zone. Proper care has been taken to minimize heat load on freezing zon using MLI. Pellet formed at the freezing zone is dislodged and accelerated to higher speed by using high pressure helium propellant gas through a fast opening valve of (opening duration < 2 millisecond). A three-stage differential pumping system is employed to remove propellant gas from injection line. Appropriate diagnostics is used to measure pellet parameters. Speed of pellet is measured by time of flight measurement using light gate diagnostic system. Pellet quality and its size, and also speed are measured using fast camera based imaging system. A Labview based GUI is used to communicate between control system and Pellet injector. The reliability of pellet formation and injection in present experimental system is greater than 95 %.
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Abstract ID: 1_37
Development of Heat Sink Concept for Near-term Fusion Power Plant Divertor
Sandeep Rimza Sandy1, Samir S Khirwadkar1, Karupanna Velusamy 2
1Institute for Plasma Research, India 2Indira Gandhi Centre for Atomic Research (IGCAR), India E-mail: [email protected]
The development of the efficient divertor concept is an important task to meet in the scenario of the future fusion power plant (DEMO). The divertor has to discharge the considerable fraction ∼15% of the total fusion thermal power incident on the divertor, therefore it has to survive very high thermal loads (~10 MW/m2) [1-3]. In the present study, a new high efficient divertor heat sink (HEDHS) concept is proposed for the future post ITER tokamak called as ‘DEMO’. The first wall of the diverter made-up of several modules to overcome the stresses caused by high heat flux, in the present design. Thermal hydraulic performance of one such HEDHS module is numerically investigated using the Fluent software. The effects of critical thermal hydraulic and geometric parameters on the heat transfer characteristics of HEDHS are presented with the Reynolds number (Re) range of 1.2×104 - 3.0×104. The stresses induced in the HEDHS by the thermal and pressure loads are an important factor that limits the performance and life of the divertor. Therefore, heat transfer coefficient received from the computational fluid dynamics (CFD) analysis is used to perform the thermo-mechanical analysis through finite element based approach. The result revealed that, the proposed design is capable to accommodate the design loads at the acceptable pumping power ratio, and stresses are well within the allowable limits. In addition, detailed of fluid flow and heat transfer mechanism associated with geometric variation have also been studied for the HEDHS to enhance the thermal performance.
Fig. 1. Schematic diagram of high efficient divertor heat sink concept.
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Temperature [k]
Max. temp. = 2093 k
Max. temp. = 1843 k
Max. thimble temp. = 1683 k Max. thimble temp. = 1493 k
Dthimble = 25mm Dthimble = 16mm
Fig. 2. Comparisons of temperature distribution at various thimble diameter at same Reynolds number.
References:
[1] D. Maisonnier, I. Cook, S. Pierre, B. Lorenzo et al., DEMO and fusion power plant conceptual studies in Europe, Fusion Engineering and Design 81 (2006) 1123–1130.
[2] S. Rimza, K. Satpathy, S. Khirwadkar, K. Velusamy, Numerical studies on helium cooled divertor finger mock up with sectorial extended surfaces, Fusion Engineering and Design 89(2014) 2647-2658.
[3] S. Rimza, S. Khirwadkar, K. Velusamy, An experimental and numerical study of flow and heat transfer in helium cooled divertor finger mock-up with sectorial extended surfaces, Applied Thermal Engineering 82 (2015) 390- 402.
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Abstract ID: 1_40
Characterization of Discharge Plasma in Cylindrical IECF Device
Neelanjan Buzarbaruah1, Nilam Jyoti Dutta1, Davashree Borgohain1, Smruti Ranjan Mohanty1
1Centre of Plasma Physics-Institute for Plasma Research,India Email: [email protected]
Inertial Electrostatic Confinement Fusion (IECF) device [1] draws a considerable attention, during last decade, because of its application in neutron activation analysis, land mind detection, plasma space propulsion etc. Its simple construction and ability to provide high fusion rate in small volume prompt the researchers to use this device as a portable neutron source. This source mainly comprises of a concentric coaxial cylindrical grid assembly housed inside a cylindrical vacuum chamber, a gas injection system, a high voltage feedthrough and a high voltage negative polarity power supply. On application of high negative potential of few tens of kV to the inner grid of the device, the ions would overcome the coulomb barrier force and fuse together to produce neutrons of the order 108 n/s.
A compact cylindrical IECF device is currently under development at Centre of Plasma Physics- Institute for Plasma Research. The installation of the cylindrical IECF chamber of diameter 50cm and height 30cm has been completed. The chamber is integrated with all necessary components namely the Turbo Molecular Pump (TMP), gate valve, pressure gauges and high voltage DC feedthrough (150 kV) [2]. Presently, we are producing the filamentary glow discharge plasma using deuterium gas inside the chamber. The plasma is characterized using electrostatic probes. Plasma parameters such as the electron temperature (Te), plasma potential (Vp) and plasma 15 -3 density (ni) are evaluated [3]. Plasma density of the order 10 m is achieved and this would enable us to generate neutrons in the above mentioned range. The details on the experimental studies will be presented in the paper.
References:
[1] S. K. Murali, G. A. Emmert, J. F. Santarius, G. L. Kulcinski, “Effects of chamber pressure variation on the grid temperature in an inertial electrostatic confinement device”, Physics of Plasmas. 17 102701 (2010).
[2] N. Buzarbaruah, N. J. Dutta, J. K. Bhardwaz and S. R. Mohanty, “Design of a linear neutron source,” Fusion Engineering and Design, 90, 97-104 (2015).
[3] R. L. Merlino, “Understanding Langmuir probe current-voltage characteristics”, American Journal of Physics 75, 1078 (2007)
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Abstract ID: 1_57
Serial Interface through Stream Protocol on EPICS Platform for Distributed Control and Monitoring
Arnab Das Gupta1, Amit Srivastava1, Sunil Susmithan1, Ziauddin Khan1
1Institute for Plasma Research, India Email: [email protected]
Remote operation of any equipment or device is implemented in distributed systems in order to control and proper monitoring of process values. For such remote operations, Experimental Physics and Industrial Control System [1] (EPICS) is used as one of the important software tool for control and monitoring of a wide range of scientific parameters. A hardware interface is developed for implementation of EPICS software so that different equipment such as data converters, power supplies, pump controllers etc. could be remotely operated through stream protocol. EPICS base was setup on windows as well as Linux operating system for control and monitoring while EPICS modules such as asyn and stream device were used to interface the equipment with standard RS-232/RS-485 protocol. Stream Device protocol communicates with the serial line with an interface to asyn drivers. Graphical user interface and alarm handling were implemented with MEDM (Motif Editor and Display Manager) and ALH (Alarm Handler) command line channel access utility tools. This paper will describe the developed application which was tested with different equipment and devices serially interfaced to the PCs on a distributed network.
References:
[1] http://www.aps.anl.gov/epics/
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Abstract ID: 1_58
Development of Data Acquisition Set-up for Steady-state Experiments
Amit Srivastava1, Arnab Das Gupta1, Sunil Susmithan1, Ziauddin Khan1
1Institute for Plasma Research, India Email: [email protected]
For short duration experiments, generally digitized data is transferred for processing and storage after the experiment whereas in case of steady-state experiment the data is acquired, processed, displayed and stored continuously in pipelined manner. This requires acquiring data through special techniques for storage and on the go viewing data to display the current data trends for various physical parameters. A small data acquisition set-up is developed for continuously acquiring signals from various physical parameters at different sampling rate for long duration experiment. This includes the hardware set-up for signal digitization, FPGA based timing system for clock synchronization and event/trigger distribution, time slicing of data streams for storage of data chunks to enable viewing of data during acquisition and channel profile display through down sampling etc. To store a long data stream of indefinite/long time duration, the data stream is divided into data slices/chunks of user defined time duration. Data chunks avoid the problem of non-access of data until the channel data file is closed at the end of the long duration experiment. A graphical user interface has been developed in LabVIEW application development environment for configuring the data acquisition hardware and storing data chunks on local machine as well as at remote data server for further data access. The data plotting and analysis utilities have been developed with Python software, which provides tools for further data processing. This paper describes the detailed development and implementation of data acquisition for steady-state experiment.
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Abstract ID: 1_59
Prototyping of Radial Plates for Fusion Relevant Superconducting Magnets
Mahesh Ghate1, Dhaval Bhavasar1, Arun Panchal 1, Swaroop Udgata2, Subrata Pradhan1
1Institute for Plasma Research, India 2I-DESIGN Engineering Solutions Ltd., Waghoili, Pune Email: [email protected]
“Magnet Technology Development Division” is engaged in focused research and development of indigenous fusion relevant superconducting magnet along with its associated technologies at Institute for Plasma Research in association with various R&D organizations. Under this initiative, prototyping trials for radial plate to validate its conceptual design and feasibility for manufacturing have been discussed in this paper. The simulation approach with CAD to formulate machining sequences for prototyping of radial plates has been presented. The extensive trials had been done on SS316LN plates to estimate and establish machining sequences, machine parameters, machining tools to achieve required stringent tolerances. The critical machining operation and parameters for prototype radial plates has been discussed in this paper. The inspection methodology with Articulated Arm Coordinate Measuring Machine (AACMM) for prototype radial plate has been established and verified.
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Abstract ID: 1_60
Application of Articulated Absolute Co-ordinate Measuring Machine for Quality Control in Manufacturing of ELM Control Coil
Dhaval Bhavsar1, Mahesh Ghate1, Subrata Pradhan1
1Institute for Plasma Research, India Email: [email protected]
Under India-EU collaboration, Institute For Plasma Research had undertaken an engineering feasibility initiative aimed at developing a 1:1 prototype Edge Localized Modes control coils (ELM CC) for Joint European Torus (JET). The ELM coils comprised of winding pack made of CuCrZr conductor encased in Inconel 625 casing. The ELM control coils are designed in saddle coil configuration having toroidal and poloidal curves similar to that of JET vacuum vessel. ELM coil are in-vessels coils forming the primary boundary with torus vacuum which demands stringent requirement for its quality aspects. The dimensional accuracies of winding pack and casing are critical for its encasing and remote assembly inside vacuum vessel. The articulated arm co-ordinate measuring machine (AACMM) has been extensively used for dimensional metrology of ELM CC from winding to its encasing. The inspection methodology and procedures using noncontact technique for ELM CC with AACMM has been developed and established with extensive trials. The winding pack, their formers and final ELM control coils has been systematically investigated for their dimensional accuracies with AACMM. The effectiveness of AACMM based evaluation for quality control in fabrication of 1:1 prototype of ELM CC has been presented in this paper.
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Abstract ID: 1_61
Indigenously Developed Bending Strain Setup for I-V Characterization of Superconducting Tapes and Wires
Arun Panchal1, Anees Bano1, Mahesh Ghate1, Piyush Raj1, Subrata Pradhan1
1Institute for Plasma Research, India Email: [email protected]
The indigenously developed bending strain setup to examine the effect of pure bending on critical current of superconducting tapes and strands has been presented in this paper. This set up is capable of applying various bending radius at cryogenic temperature with rack and pinion gear mechanism. The strain applied on samples can be controlled externally by torsional input which transferred in the form of bending radius during experiments. The working principle, design and optimization of this set up have been discussed. The performance and validation of set up has been done on various HTS tapes and copper strands at 77 K in actual experimental facility. The effect of bending radius (15.5 mm- 48 mm) i.e strain and ramp rate (2 amp/s – 8 amp/s) is observed on current capability of various HTS Tapes. The critical current capability of BSCCO tape without any strain is 133 A which is reduced to 93 A with 0.83% strain at ramp rate of 8 amp/s. The critical current capability of DI-BSCCO tape without any strain is 139.6 A which reduced to 97 A with 1% strain at ramp rate of 8 amp/s. The critical current capability of YBCO tape without any strain is 80 A which is reduced to 77.8 A with 1% strain at ramp rate of 8 amp/s. In mentioned conditions, it is observed that in uniform bending condition, the degradation in current carrying capacity BSCOO and DIBSCCO (~30%) is more as compare to YBCO (~2.75%) at 77K. The effect of pure mechanical strain has been experimentally observed and presented in this paper.
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Abstract ID: 1_64
RF Assisted Glow Discharge Condition Experiment in SST-1 Tokamak
Dilip Raval1, Ziauddin Khan1, Siju George1, Kalpeshkumar R Dhanani1, Yuvakiran Paravastu1, Pratibha Semwal1, Prashant Thankey1, Mohammad Shoaib Khan1, Subrata Pradhan1
1Institute for Plasma Research, India Email: [email protected]
Impurity control reduces the radiation loss from plasma and hence enhances the plasma operation. Oxygen and water vapors are the most common impurities in tokamak devices. Water vapour can be reduced with extensive baking while in order to have a significant reduction in oxygen it is necessary to use glow discharge condition (GDC). RF assisted glow discharge cleaning system was implemented to remove low z impurities. Discharge cleaning with both pure helium and with 20% hydrogen was used. It was observed that the ultimate impurity level was reduced significantly below to the accepted level for plasma operation. In this paper, the detailed design aspect and the implementation of RF assisted Glow discharge conditioning on PFC installed SST-1 vacuum vessel is discussed.
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Abstract ID: 1_67
Commissioning and Experimental Validation of SST-1 Plasma Facing Components
Yuvakiran Paravastu1, Dilip Raval1, Ziauddin Khan1, Hiteshkumar Patel1, Prabal Biswas1, Tejas Parekh1, Siju George1, Prosenjit Santra1, Gattu Ramesh Babu1, Prashant Thankey1, Pratibha Semwal1, Arun Prakash A1, Kalpeshkumar R Dhanani1, Snehal Jaiswal1, Pradeep Chauhan1, Subrata Pradhan1
1Institute for Plasma Research, India Email: [email protected]
Plasma facing components of SST-1 are designed to withstand an input heat load of 1.0 MW/m2. They also protect vacuum vessel, auxiliary heating source i.e. RF antennas, NBI and other diagnostic in-vessel components from the plasma particles and high radiative heat loads. PFC’s are positioned symmetric to mid-plane to accommodate with circular, single and double null configuration. Graphite is used as plasma facing material which is fixed on copper alloy (CuCrZr and CuZr) back plate with mechanical attachment followed with graphoil in between. SS cooling/baking tubes are brazed on copper alloy back plates for efficient heat removal of incident heat flux. Benchmarking of PFC assembly was first carried out in prototype vacuum vessel of SST-1 to develop understanding and methodology of co-ordinate measurements. Based on such hands-on-experience, the final assembly of PFC’s in actually vacuum vessel of SST-1 was carried out. Initially, PFC’s are to be baked at 250 C for wall conditioning followed with cooling for heat removal of incident heat flux during long pulse plasma operation. For this purpose, the supply and return headers are designed and installed inside the vacuum vessel in such a way that it will cater water as well as hot nitrogen gas depending up on the cycle. This paper will discuss the successful installation of PFC’s and its operation respecting all design criteria for plasma operation.
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Abstract ID: 1_68
Baking and Helium Glow Discharge Cleaning of SST-1 Tokamak with Graphite Plasma Facing Components
Pratibha Semwal1, Ziauddin Khan1, Dilip Raval1, Kalpeshkumar R Dhanani1, Siju George1, Yuvakiran Paravastu1, Arun Prakash A1, Prashant Thankey1, Gattu Ramesh Babu1, Mohammad Shoaib Khan1, Partha Saikia1, Subrata Pradhan1
1Institute for Plasma Research, India Email: [email protected]
Graphite plasma facing components (PFCs) were installed inside SST-1 vacuum vessel. Prior to installation, all the graphite tiles were baked at 1000 C in a vacuum furnace operated below 1.0 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 water vapour requires high temperature bake-out of the PFCs at 250 C. In SST-1 the PFCs were baked at 250C 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 (He) glow discharge cleaning (GDC) effectively removes this stored hydrogen as well as other impurities like oxygen and hydrocarbon within few nanometers 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 0.48. In this paper, the results of effect of baking and He-GDC experiments of SST-1 will be presented in detail.
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Abstract ID: 1_79
Design and Integration of SMBI System for SST 1
Siju George1, Yuvakiran Paravastu1, Mohammad Shoaib Khan1, Kalpeshkumar R Dhanani1, Dilip Raval1, Ziauddin Khan1, Santanu Banerjee1, Subrata Pradhan1
1Institute for Plasma Research, India Email: [email protected]
Supersonic molecular beam injection (SMBI) is one of the most effective fuelling methods for injecting neutral particle at very high velocity into the plasma core. Due to higher speed and lower divergence, the beam penetrates several centimeters into the plasma and hence increases the fuelling efficiency.
In SST-1, two types of SMBI systems are proposed. One will be installed in the low field side (LFS) while two are integrated in the high field side (HFS). This paper will describe the design, fabrication and implementation of SMBI system in SST-1 Tokamak.
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Abstract ID: 1_90
Neutron Measurements from Beam-Target Interactions with Deuterium Ion Beam
Sudhirsinh Vala1, A T T Mostako1, Mitul Abhangi1, C V S Rao1, Rajnikant Makwana2, T K Basu1
1Institute for Plasma Research, India 2M. S. University, India Email: [email protected]
Neutron measurements can be used as an important diagnostic tool for studying beam homogeneity in Neutral Beam Injection (NBI) facility. Neutrons are produced due to fusion reaction between beam deuterons and deuterons implanted in the beam-dump. The penetration and saturation concentration of deuterium ions implanted into copper beam dump is studied using TRIM-Monte Carlo simulation code.
In the present study deuterium ions are extracted from the SILHI ECR Ion source facility at Fusion Neutronics Laboratory. 2.5 MeV neutron emission from beam-target DD reaction is measured using NE-213 liquid scintillation detector. Neutron transport effects in the beam dump is investigated using MCNP code.
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Abstract ID: 1_93
Electron Beam Welding: Study of Process Capabilities and Limitations towards Development of Nuclear Components
Gautam Vadolia1, Kongkham Premjit Singh1
1Institute for Plasma Research, India Email: [email protected]
Electron beam (EB) welding technology is an established and widely adopted technique in nuclear research and development area. Electron Beam welding is thought of as a candidate process for ITER Vacuum Vessel Fabrication. Dhruva Reactor @ BARC, Mumbai and Niobium Superconducting accelerator Cavitity @ BARC has adopted the EB welding technique as a fabrication route. The highly concentrated energy input of the electron beam has added the advantages over the conventional welding as being less HAZ and provided smooth & clean surface. EB Welding has also been used for the joining of various reactive and refractory materials. EB system as heat source has also been used for vacuum brazing application.
The Welding Institute (TWI) has demonstrated that EBW is potentially suitable to produce high integrity joints in 50 mm pure copper. TWI has also examined 150 kV Reduced Pressure Electron Beam (RPEB) gun in welding 140 mm and 147 mm thickness Nuclear Reactor Pressure Vessel Steel (SA 508 grade). EBW in 10 mm thick SS316 plates were studied at IPR and results were encouraging.
In this paper, the pros and cons and role of electron beam process will be studied to analyze the importance of electron beam welding in nuclear components fabrication. Importance of establishing the high precision Wire Electro Discharge Machining (WEDM) facility will also be discussed.
References:
[1] T K Saha and A K Ray, Vacuum – The Ideal Environment for Welding of Reactive Materials, J. Phys.: Conf. Ser. 114 012047 (Issue 1 2008)
[2] R. Lindau et.al, Mechanical and microstructural characterization of electron beam welded reduced activation oxide dispersion strengthened – Eurofer steel, J. Nucl. Mater., 416, 1-2 (2011), pp.22-29
[3] K. P. Singh et.al, Curved small tungsten macrobrush test mock-up fabrication using vacuum brazing for divertor Target elements, Fusion Science and Technology, 65 (2014), pp.235-240
[4] Online reference (www.msm.cam.ac.uk/phase-trans/2014/Duffy.pdf) accessed on 26 August 2015 and online reference (www.ndt.net/article/nde-india2011/pdf/2-24B-2.pdf) accessed on 26 August 2015.
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Abstract ID: 1_99
Thermal Response of Actively Cooled Tungsten Monoblock during Inhomogeneous Surface Heat Loads
Yashashri Patil1, Samir S Khirwadkar1, Deepu S Krishnan1
1Institute for Plasma Research, India Email: [email protected]
Vertical targets of the ITER Divertor consist of continuous actively cooled plasma facing units (PFU). Tungsten (W) monoblocks joined with Copper Chromium Zirconium (CuCrZr) alloy tube are the basic building block of PFU. Tungsten monoblocks are exposed to the non-homogeneous heat loads arises due to exponentially decaying power flux along Scrap of Layer (SOL) away from the separatrix as well as fabrication tolerances and misalignment of divertor targets in poloidal and toroidal directions. Thermal and structural studies of tungsten monoblock under such non-homogeneous heat loads are needed.
This Paper presents the non-homogeneous high heat flux studies carried out on the ITER Tungsten monoblock. Surface temperature and thermal stresses observed on tungsten monoblock were calculated using Finite Element Analysis (FEA). FEA model of a tungsten monoblock exposed to the homogeneous heat flux of 10 MW/m2 & 20 MW/m2 was developed using Comsol Multiphysics software 5.1. Further FEA were carried out for two ITER non- homogeneous heat flux scenarios. Tungsten monoblock exposed to different heat flux values along toroidal direction with fraction varies 20-95 % by steps of 15 % with incident flux 4 MW/m2on shadow and 10 MW/m2 rest of the tungsten monoblock. Other case heat flux on shadow is 0 MW/m2 and rest part of tungsten monoblock is 20 MW/m2. Heat transfer coefficient of 46,000 W/m2 K applied on inner surface of heat sink tube. Surface temperature was calculated by FEA on the tungsten monoblock during Non-homogeneous studies are as shown in figure 1.
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Figure1. Surface temperature observed on tungsten monoblock by FEA for non-homogeneous heat flux with fraction varies 20-95 % by steps of 15 % with incident flux 4 MW/m2on shadow and 10 MW/m2 rest of the tungsten monoblock.
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Abstract ID: 1_101
Consistency Checks in Beam Emission Modeling for Neutral Beam Injectors
Bharathi Punyapu1, Prahlad Vattipalle1, Sanjeev Kumar Sharma1, Ujjwal Kumar Baruah1, Brendan Crowley2
1Institute for Plasma Research, India 2DIII-D, General Atomics, USA Email: [email protected]
In positive neutral beam systems, the beam parameters such as ion species fractions, power fractions and beam divergence are routinely measured using Doppler shifted beam emission spectrum. The accuracy with which these parameters are estimated depend on the accuracy of the atomic modeling involved in these estimations. In this work, an effective procedure to check the consistency of the beam emission modeling in neutral beam injectors is proposed. As a first consistency check, at a constant beam voltage and current, the intensity of the beam emission spectrum is measured by varying the pressure in the neutralizer. Then, the scaling of measured intensity of un-shifted (target) and Doppler shifted intensities (projectile) of the beam emission spectrum at these pressure values are studied. If the un-shifted component scales with pressure, then the intensity of this component will be used as a second consistency check on the beam emission modeling. As a further check, the modeled beam fractions and emission cross sections of projectile and target are used to predict the intensity of the un-shifted component and then compared with the value of measured target intensity. An agreement between the predicted and measured target intensities provide the degree of discrepancy in the beam emission modeling.
In order to test this methodology, a systematic analysis of Doppler shift spectroscopy data obtained on the JET neutral beam test stand data was carried out. The analysis showed that the intensity ratios of full, half and one third components did not scale with different pressure values, indicating the significance of the role played by the beam–beam interactions, electron impact excitations and non-radiative decay due to collisional quenching. These processes are usually neglected in the beam emission modelling. The intensity vs pressure scaling showed a discrepancy of ~57% for full energy component and ~60% for fractional energy components respectively. Whereas the scaling of un- shifted component showed a discrepancy of up to 5%. A further consistency check was then done by modeling the intensity of the un-shifted component. The measured and modeled un-shifted intensities agreed within ~25% indicating the discrepancies in the modeling of beam fractions and the cross section data base used in these calculations. This procedure can be established as a tool to identify the role of the mentioned process in the modeling for improving the accuracy of the measurements of beam parameters. The complete procedure is discussed and the detailed formulation to model the target intensity is presented in this paper.
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Abstract ID: 1_105
Computational Fluid Dynamics Analysis of Heat Transfer Elements for SST-1 Neutral Beam Line
Ravi Patel1, Mahesh Ghate2, Bharathi Punyapu1, Rajesh Patel2, Prahlad Vattipalle1
1Pandit Deendayal Petroleum University, India 2Institute for Plasma Research, India Email: [email protected]
A 5 MW Neutral Beam Injector (NBI) is designed and commissioned to deliver a heating power of 1.7 MW to the SST-1 tokomak. To sustain the high heat flux in these injection experiments, heat transfer elements (IPR-HTE) were successfully developed and fabricated. These HTEs are actively cooled elements which rely on internal fins and boiling heat transfer to maximise the heat transfer capability. In this work the performance of HTE is analysed using analytical models and a commercially available Computational Fluid Dynamics (CFD) software. Validation of these CFD models are accomplished by comparing these with the available experimental results obtained on similar neutral beam systems.
For an initial assessment on performance of HTE, a 2-D thermal analysis using transient thermal module of ANSYS software was performed in which the heat transfer coefficient (h) was calculated for the single phase flow for establishing the procedure and preliminary study. For improving the accuracy in these results, a 3-D single phase flow CFD analysis using CFX module of ANSYS software was carried out for detailed study flow characteristics. These results were then compared with the published experimental results of hypervapotron of JET neutral beams which has similar geometry of IPR-HTE. The computational results were found to be in good agreement with the experimental result for heat flux values up to 5 MW/m2 beyond which they deviated from experimental results (32% of deviation) indicating the onset of two phase flow. Hence, a two phase flow analysis was further attempted with Eulerian approach and RPI boiling model in CFX module of ANSYS. With the inclusion of the two phase models and user defined functions, the results agreed well with the experimental results (<15 % deviation). This analysis significantly improved the understanding of the flow characteristics such as velocity streamlines, eddies formulation, temperature distribution and their effect on performance of IPR- HTE at different heat flux regimes.
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Abstract ID: 1_106
Er2O3 Coating Development and Improvisation by Metal Oxide Decomposition Method
Pratipalsinh A Rayjada1, Amit Sircar1, Prakash M Raole1, Lalit M Manocha2, Raseel Rahman1
1Institute for Plasma Research, India 2DMSRD establishment, Kanpur, India Email: [email protected]
Compact, highly resistive and chemically as well as physically stable ceramic coatings are going to play vital role in successful and safe exploitation of tritium breeding and recovery system in the future fusion reactors. Due to its stability and high resistivity, Er2O3 was initially studied for resistive coating application to mitigate Magneto Hydro Dynamic (MHD) forces in liquid Li cooled blanket concept [1]. Subsequently, its excellence as tritium permeation barrier (TPB) was also revealed [2]. Ever since, there is a continual thrust on studying its relevant properties and application methods among the fusion technology and materials community. Metal Oxide Decomposition is a chemical method of coating development. One of the major advantages of this process over most of the others is its simplicity and ability to coat complex structures swiftly. The component is dipped into a liquid solution of the Er2O3 and subsequently withdrawn at an optimized constant speed, so as to leave a uniform wet layer on the surface. This can be repeated multiple times after drying the surface to obtain the required thickness. Subsequently, the component is heat treated to obtain crystalline uniform Er2O3 coating over it. However, the porosity of the coatings and substrate oxidation are the challenges for in MOD method [3].
We successfully develop Er2O3 coating in cubic crystalline phase on P91 steel and fused silica substrates using 3 wt% erbium carboxylic acid solution in a solvent containing 50.5 wt% turpentine, 25.5 wt% n-butyl acetate, 8.4 wt% ethyl acetate, a stabilizer, and a viscosity adjustor. A dip coating system equipped with 800 C quartz tube furnace was used to prepare these coatings. The withdrawal speed was chosen as 72 mm/min from the literature survey. The crystallization and microstructure are studied as functions of heat treatment temperature in the range of 500-700 C. We also try to improvise the uniform coverage and porosity of the coating by altering the multiple dipping cycle so that to provide heat treatment after every sub-layer formation. We would report significant improvement in the porosity reduction and completeness of the surface coverage as viewed from systematic microscopic studies in combination with X- ray Diffraction for crystallization.
References:
[1] T. Muroga, “Ceramic Coatings as Electrical Insulators in Fusion Blankets,” Comprehensive Nuclear Materials, 4, 691 (2012).
[2] D. Levchuk, S. Levchuk, H. Maier, H. Bolt, A. Suzuki, J. Nucl. Mater., 367–370, 1033 (2007).
[3] Z. Yao, A. Suzuki, D. Levchuk, T. Chikada, T. Tanaka, T. Muroga, et al., J. Nucl. Mater. 386–388, 700, (2009).
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Abstract ID: 1_108
Design of CPLD-DAC Based Probe Bias Generator and Current Measurement Electronics
Minsha Shah1, Rachna Rajpal1, Amitkumar D Patel1, Meenakshee Sharma1, Narayanan Ramasubramanian1
1Institute for Plasma Research, India Email: [email protected]
In Institute for Plasma Research Multi-Cusp Plasma experiment is being pursued. In Multi-Cusp Plasma experiment contact-ionized cesium ions will be confined by a multi-cusp magnetic field configuration. The cesium ions will be produced by impinging collimated neutral Cesium atoms on a hot tungsten plate. The temperature of the tungsten plate will also be made high enough such that it will contribute electrons also to charge neutralize the plasma. Since this plasma will not emitting any visible or UV radiations, electric probes are the only diagnostics planned for the time being.
The probe needs to be biased with ramp or triangular signal waveforms of fixed amplitude of +/- 15 V & frequency of 50 Hz for different plasma experiments. A combination of Complex programmable Logic Device (CPLD) and Digital to Analog Convertor (DAC) based waveform generator is conceptually designed. The programmable devices play a very important role where flexibility of programming the amplitude as well as the frequency of the bias waveform is required. The accuracy of the bias waveform can be set to as low as in the micro volt ranges. At present a 16 bit DAC will be interfaced with the CPLD. The requirement also involves development of signal conditioning electronics for the probe current measurement. Plasma current measurement is very tricky as it needs to extract the low amplitude AC signal from DC bias voltage. Accurate difference amplifier is designed to achieve maximum AC CMRR to extract the real signal. The signal conditioning electronics is a combination of I-V convertor, a precision difference amplifier, optical isolator and driver. SMD components are selected to make the circuit compact and rugged. This paper describes the hardware and software design aspects of electronics.
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Abstract ID: 1_111
Nanoscale Coatings of Tungsten by Radio Frequency Plasma Assisted Chemical Vapor Deposition on Graphite
Uttam Sharma1, Sachin Singh Chouhan1, Amulya Kumar Sanyasi2, Kumarpalsinh A Jadeja2, Joydeep Ghosh2, J. Sharma3
1Shri Vaishnav Institute of Technology and Science, India 2Institute for Plasma Research, India 3 M. B. Khalsa College, India Email: [email protected]
Future thermonuclear fusion reactors including ITER are heading towards full scale operations with tungsten being the material for the divertor, limiter and probably the first wall too. Tungsten has several superior properties over its low Z competitors in terms of higher melting point, lower sputtering yield, low fuel retention (D - T) etc. So far, fusion experimentalists have gained enough experience and have rich databases with carbon as its first wall as well as target materials in tokamaks. However, database for tungsten line radiation in variety of plasmas i.e. basic laboratory scale to high density and high temperature plasmas is rare and this requires immediate attention to construct a database with experimental evidences. Such studies are not limited to only large scale fusion reactors but small and medium scale toroidally confined devices can be suitably utilized. Present day tokamaks are now switching to plasma facing components made up of tungsten. As the complete replacement of the wall and target materials from carbon to tungsten in existing tokamaks is challenging and time consuming exercise, tungsten coatings on selected target materials remains a very feasible option for the purpose.
This paper will present the development of indigenous tungsten coating reactor which has successfully produced tungsten coated graphite tiles of sample dimensions. The tungsten coated graphite tiles are produced by RF plasma assisted chemical vapor deposition of tungsten on graphite substrates. The RF plasma is produced with 60 – 100 W power and tungsten nano ions are produced by dissociating the precursor gas tungsten hexa-fluoride (WF6) in sufficient hydrogen background. Further, challenges in handling WF6 plasma at high pressures and in-situ spectroscopy results during the coating process will be presented.
References:
[1] Deposition and qualification of tungsten coatings produced by plasma deposition in WF6 precursor gas. Phys. Scr. T145, 014030 (2011).
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Abstract ID: 1_113
Multi-scale Modeling of Neutron Induced Radiation Damage in Tungsten
Maya P N1, Shishir P Deshpande1, Manoj Warrier2, Prithwish Nandi1, Prakash M Raole3, Samir S Khirwadkar3
1ITER-India, Institute for Plasma Research, India 2Bhabha Atomic Research Centre-Visakhapatnam, India 3Institute for Plasma Research, India Email: [email protected]
Tungsten will be used in ITER divertor which is also one of the candidate materials for future fusion reactors such as DEMO [1]. When exposed to 14.1 MeV fusion neutrons and 3.5 MeV alpha particles, tungsten is going to accumulate radiation damage. Since the fusion relevant conditions cannot be realized except in the reactor itself, there is an urgent need to approach this problem from carefully validated multi-scale models. Surrogate ion-irradiation experiments can be used to validate the multi-scale models of radiation damage in tungsten.
In this work, we discuss a consistent multi-scale scheme of neutron damage in tungsten starting from neutronics calculations [2]. We specially emphasize the radiation damage studies using Molecular dynamics simulations. The MD simulations are performed using parallel molecular dynamics codes ParCas [3] and LAMMPS [4]. We show the vacancy and interstitial clustering in single crystal tungsten due to irradiation of energetic W primary knock-on atoms (PKA) for a range of energy starting from 500 eV to 20 keV using different interatomic potentials [5, 6, 7]. The PKA are initialized along 100 random directions within the sample bulk. During the collision cascade a part of the energy is assumed to be transferred to electronic system via inelastic collisions. These electronic losses are taken into account by applying constant frictional losses with a cut-off [3, 8].
The collision cascade generates stable vacancy-interstitial pairs (Frenkel pairs) with in MD simulation time scales. At higher irradiation energies we have observed interstitial clustering. The vacancies are immobile in the simulation time scales. In comparison to the standard binary collision models, the observed number of stable Frenkel pairs in MD simulations is much smaller which is attributed to the recombination of the displaced atoms during the thermal spike. The difference in the thermal spike relaxation with and without electronic losses will be shown [8]. The difference between interacting and non-interacting cascades in defect formation will also be discussed where the latter can contribute to the observed micro-structure in ion-irradiation experiments. The vacancy clusters observed in experiments due to surrogate ion-irradiation will also be briefly discussed in the presentation, where we discuss the challenges in interpreting the experimental data of radiation damage.
References:
[1] H. Bolt, V. Barabash, G. Federici et.al., J. Nucl. Mater., 307–311:43–52 (2002)
[2] S.P. Deshpande, P.N. Maya, P. V. Subhash et.al., 15th PFMC conference,18-22 May (2015), Aix- en-Provence, France
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[3] K. Nordlund and J. Keinionen. Phys. Rev. Lett., 77:699, (1996)
[4] S. Plimpton, J Comp Phys, 117, 1-19 (1995)
[5] N. Juslin, P. Erhart, et.al. J. Appl. Phys., 98:123520,(2005)
[6] Xiao-Chun Li, Xiaolin Shu et.al., J. Nucl. Mater. 408, 12-17 (2011)
[7] C. Björkas, K. Nordlund et.al., Nucl. Instr. Methods.Phys.B, 267, 3204-3208 (2009)
[8] P.N. Maya and S.P. Deshpande, Swift Heavy ions in Materials Engineering & Characterization (SHIMEC 2014) October 14-17 New Delhi
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Abstract ID: 1_117
Role of ECRH in SST-1 Tokamak Plasma
Braj Kishore Shukla1, Dhiraj Bora1, Ratneshwar Jha1, Subrata Pradhan1
1Institute for Plasma Research, India Email: [email protected]
In SST-1, the Electron Cyclotron resonance Heating (ECRH) system has been used extensively to carry out various experiments related to fundamental and second harmonic ECRH assisted breakdown and start-up of tokamak. The ECRH further contributes to plasma current during long pulse operation. The 42 GHz ECRH system delivers 500 kW power for 500 ms duration and corrugated waveguide ( 63.5mm) based transmission line at normal atmospheric pressure is used to launch power in HE11 mode. The mirror based launcher is used to launch focused beam in SST-1 plasma. Since the loop voltage of SST-1 is low (~3.0 V), the ECRH assisted start-up is mandatory for reliable plasma discharges. In the beginning of each plasma campaign, it is observed that impurity dominates results in small discharges. In such cases ECRH is used at higher power for long duration to overcome the impurity burn-through and get the good plasma discharges. In SST-1, the ECRH is also used to drive some current to support plasma current. The ECRH power from 150 kW to 350 kW has been launched in fundamental O-mode and second harmonic X-mode. The ECRH is used for short pulse ~80 to 120 ms (for breakdown and start-up) and long pulse duration up to 430ms (for start-up as well as support plasma current with electron cyclotron current drive ECCD). As the first pass absorption of ECRH is not good in breakdown phase (at low density and temperature), the ECRH power transmits up to inboard side wall of tokamak. At the inboard side, a profiled reflector is installed at an angle to launch focused beam from high field side with an angle to toroidal magnetic field (BT). This is similar to co-injection launch of ECRH power in X-mode from high field side to support plasma current with ECCD. The experiments show that the plasma current profile is different in two cases (ECRH short pulse and long pulse). In the long pulse ECRH, the plasma current profile is smooth with some increase (~ 5 to 10%) in plasma current, which confirms the role of ECRH on plasma current. The paper discusses about the role of ECRH and explains the various experiments related to ECRH assisted breakdown and ECCD carried out on tokamak SST-1 at fundamental and second harmonic.
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Abstract ID: 1_119
Design of 1 MHz Solid State High Frequency Power Supply
Darshan Kumar Parmar1, N P Singh, Sandip Gajjar, Aruna Thakar, Amit Patel, Bhavin Raval, Hitesh Dhola, Rasesh Dave, Dishang Upadhay, Vikrant Gupta, Niranjan Goswami, Kush Mehta, Ujjwal Kumar Baruah
1ITER-India, Institute for Plasma Research, India Email: [email protected]
A High Voltage High Frequency (HVHF) Power supply is used for various applications, like AM Transmitters [1], metallurgical applications [2], Wireless Power Transfer [3], RF Ion Sources [4], etc. The Ion Source for a Neutral beam Injector at ITER-India uses inductively coupled power source at High Frequency (~1MHz). Switching converter based topology used to generate 1MHz sinusoidal output is expected to have advantages on efficiency and reliability as compared to traditional RF Tetrode tubes based oscillators.
In terms of Power Electronics, thermal and power coupling issues are major challenges at such a high frequency. A conceptual design for a 200kW, 1MHz power supply and a prototype design for a 600W source been done. The prototype design is attempted with Class-E amplifier [5] topology where a MOSFET is switched resonantly. The prototype uses two low power modules and a ferrite combiner to add the voltage and power at the output. Subsequently solution with class-D H-Bridge [6] configuration have been evaluated through simulation [7] where module design is stable as switching device do not participate in resonance, further switching device voltage rating is substantially reduced. The rating of the modules is essentially driven by the maximum power handling capacity of the MOSFETs and ferrites in the combiner circuit. The output passive network including resonance tuned network and impedance matching network caters for soft switching and matches the load impedance to 50ohm respectively. This paper describes the conceptual design of a 200kW power supply and experimental results of the prototype 600W, 1MHz source.
References:
[1] H. Swanson, “Digital AM transmitters”, Broadcasting, IEEE Transactions on, Volume:35, (2002) [2] J. Tsujino, Recent developments of ultrasonic welding, Ultrasonics Symposium, Proceedings, Volume:2 (1995) [3] S. Bani, “A Wireless Power Transfer system optimized for high efficiency and high power applications”, Power Electronics Conference (IPEC-Hiroshima 2014 - ECCE-ASIA), (2014) [4] M. J. Singh, “RF‐Plasma Source Commissioning in Indian Negative Ion Facility”, AIP Conf. Proc. 1390, 604 (2011) [5] Nathan Sokal , "Class-E RF power amplifiers", QEX Jan/Feb 2001, WA1HQC [6] Herbert L. Krauss, “Solid State Radio Engineering”, 0471514101, John Wiley & Sons Incorporated, 1992 [7] PSIM user manual, version 8.0 (by Powersim Inc)
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Abstract ID: 1_120
Neutron Induced Reaction for Long-lived Isotopes Produced in Fusion Materials
Bhawna Pandey1, C V S Rao1, Jyoti Pandey2, Mayank Rajput1, G Vaitheeswaran1, T K Basu1, H M Agrawal2
1Institute for Plasma Research, India 2G. B. Pant University of Agriculture & Technology, India Email: [email protected]
Neutron cross-section data are required to predict the extent of activation, nuclear heating, radioactive waste generation, radiation damage and radiation dose-rate on all critical components of the fusion reactor. A large numbers of long-lived radio nuclides in the mass region ~50-60 are 53 55 produced inside the fusion reactor (D-T fuel cycle), such as Mn (T½=3.74e+6 year), Fe (T½= 60 60 59 63 2.73 year), Fe (T½=1.5e+6 year), Co (T½ = 5.27 year), Ni (T½ =7.6e+4 year), Ni (T½ = 100.1 year) originating from neutron induced transmutation reactions with the elements in the pristine Stainless Steel (SS) structural material. This may lead to significant long term waste disposal and radiation damage issues [1-2]. Fusion neutronics studies have been performed so far considering only the stable isotopes of Cr, Fe, Ni. But in D-T fusion reactor, large amounts of radio-nuclides are produced during reactor operation as well as after shut down, which affects the neutronic response of the reactor. There is an urgent need to study the neutron induced cross- section on the long-lived radio-isotopes produced in fusion materials [3].
In present work the focus is to study the interaction of neutrons with long-lived isotopes (A=50- 60) using nuclear reaction modular codes [4]. There is no experimental data of neutron induced reactions for such radionuclides, because of the non-availability of these materials in nature. In this case the only way to generate such data is by the use of theoretical model codes. Recent advancement in this field with new codes such as TALYS & EMPIRE [4] it is imperative that if optimized model parameters (along with proper validation of the code) are used then sufficiently accurate data could be obtained for many of the reactions. The neutron induced reactions cross-sections have been calculated for the long-lived radioisotopes and compared with the available discrepant values in the data libraries. Recently the surrogate technique has been used to measure the 55Fe(n,p) reaction in the neutron energy range of 8-20 MeV [5-6].
References:
[1] H. Iida et al., “Nuclear Analysis Report (NAR) ITER”, G73 DDD2 W0.2, July (2004). [2] A. Wallner et al., “Production of Long-lived Radionuclides 10Be,12C,53Mn,55Fe,59Ni and 202gPb in a Fusion Environment” Journal of the Korean Physical Society, 59, 1378 (2011). [3] R.A. Forrest, “Data requirements for neutron activation Part I: Cross-sections”, Fusion Engineering and Design, 81, 2143 (2006). [4] http://www.nndc.bnl.gov/nndcscr/model-codes/modlibs/ [5] Bhawna Pandey et al. “Estimation of (n, p) Cross-section for Radio-Nuclide 55Fe using EMPIRE and TALYS”, Nuclear Science and Engineering, 179, 313 (2015). [6] Bhawna Pandey et al., “Neutron Induced Proton Emission Cross-section by Surrogate Reaction Method for Fusion Technology Applications”, Physical Review Letter (submitted).
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Abstract ID: 1_124
Development of a Neutronics Facility using RFQ Accelerator as the Basic Tool
Renu Bahl1, Biswanath Sarkar2, Anurag Shyam1, Rajesh Kumar1, Mridula Mittal1, Sumit Kumar1
1Institute for Plasma Research, India 2ITER-India, Institute for Plasma Research Email: [email protected]
One among the many challenges in a fusion reactor is the qualification of materials for the in service conditions, such as, the compatibility with high energy thermal neutrons. Therefore, it is prudent to envision that functional and structural materials to be used in construction of fusion reactor would need qualification to the extent possible.
Along with the global efforts, the Indian domestic fusion program also initiated a project on “Development of a RFQ for accelerators” at Institute for Plasma Research, Gandhinagar, as a first step to create a neutronic facility for material qualification in a large scale.
The facility at IPR will consist of an Electron Cyclotron Resonance (ECR) high intensity ion (H+/D+) source coupled to (copper) vane type Radio Frequency Quadrupole (RFQ) Accelerator through a LEBT to produce 5MeV, 40 mA deuterium ions ultimately. The radio frequency quadrupole (RFQ) is a linear accelerator and is very efficient at low velocities. Its inherent property of bunching the beam adiabatically and carrying out the task of focussing and accelerating the beam simultaneously, has made it a preferred choice as front end injectors of high current linacs. The accelerated ion beam produced by RFQ and the subsequent reaction of the beam with a target (possibly ‘Be’) will produce a spectrum of neutrons. These spectrum of neutrons will then be used to interpret the effect of intense neutron fluxes on materials to be used. The facility will also support the qualification of electronics and instrumentation to be used in neutron environment in the fusion reactor facility as required. A copper four vane type RFQ [1] @ 352 MHz frequency has been designed to accelerate deutrons upto 1 MeV energy. The physical design of a RFQ includes two main aspects: a) the beam dynamics design to generate the vane tip modulation and b) the electromagnetic design of the resonator cavity [2]. The physics design has been completed, where the basic design has been able to separate out the dipole and quadrupole modes distinctly. The end losses have been taken care with proper state of the art end-cut design. The harmonization of the vane design, vane tips and realistic possible manufacturing has been also studied to have the most realistic design. The basic dimensions have been worked out along with the primary integration and assembly plan. Both the aspects of the RFQ accelerator design will be discussed in detail in this paper along with brief introduction of the facility as a whole. References:
[1] Thomas P. Wangler, “RF Linear Accelerator”, WILEY-VCH Verlag GmbH & Co. kGaA, Michigan State University USA, 2008
[2] Thomas P. Wrangler, “Lumped Circuit Model of Four-Vane RFQ Resonator”, Proceedings of the 1984 Linear Accelerator Conference, Seeheim, Germany, pg. no. 332-334.
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Abstract ID: 1_127
Design of a Prototype Positive Ion Source with Slit Aperture Type Extraction System
Sanjeev Sharma1, Prahlad Vattipalle1, Bhargav Choksi1, Bharathi Punyapu1, Rambabu Sidibomma1, Sridhar B1, Ujjwal Kumar Baruah1
1Institute for Plasma Research, India Email: [email protected]
The neutral beam injector group at IPR is developing a positive ion source capable of delivering H+ ion beam having energy of 30 – 40 KeV and carrying an ion beam current of 5 – 10 A for constructing a diagnostic neutral beam for SST-1. The slit aperture based extraction system is chosen for extracting and accelerating the ions so as to achieve low divergence of the ion beam (< 0.5°). For producing ions a magnetic multi-pole bucket type plasma chamber is selected. A design study is carried out to optimize the magnetic configuration and the ion extraction- acceleration system.
The magnetic multi-pole bucket type plasma chamber is one of the most prominent sources for application in diagnostics neutral beam systems. The spatial uniformity of the source plasma depends on the spatial distribution of the magnetic field near the extraction plane. The basic characteristics of the ion source such as uniformity of magnetic field and distribution of primary electrons are examined by analyzing the magnetic field and trajectories of primary electron. A computer program is used to calculate the magnetic field and trajectories (orbits) of the primary electrons to investigate the role of the two magnetic configurations i.e. line cusp and checker board.
Numerical simulation is carried out by using OPERA-3D to study the characteristic performance of the slit aperture type extraction-acceleration system. Beam divergence, perveance and emittance were estimated for the slit apertures having widths of 4mm and 8 mm. The distribution of the slit apertures on the extraction plane of the ion source was fixed by optimizing power density profile and focal length of the extracted ion beam.
We report here the results of the studies carried out on the aspects of design of slit aperture type positive ion source.
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Abstract ID: 1_132
Optimization of Geometrical Parameters for High Heat Flux Components (Vapotrons)
Sajal Thomas1, Shrishail B Padasalagi1
1ITER-India, Institute for Plasma Research, India Email: [email protected]
One of the major requirements of Fusion Reactors is to handle the high heat flux in the Tokomak and its auxiliary systems. The expected flux density is in the range 5 MW/m2 to 20 MW/m2. These power densities within fusion device necessitate the need of High Heat Flux Components; one of the candidates for such heat flux requirements is Hypervapotrons.
Hypervapotrons are water cooled devices with internal fins or cavities oriented perpendicular to the flow of water.
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Abstract ID: 1_139
Design and Development of CRIO Based Data Acquisition and Control System for High Voltage Bushing Experiment
Himanshu Tyagi1, Sejal Shah1, Jignesh Soni2, Ratnakar Kumar Yadav1, Kartik J Patel2, Hiren Mistri2, Deepak Parmar1, Jignesh Bhagora1, Dheeraj Kumar Sharma2, Mainak Bandyopadhyay1, Arun Kumar Chakraborty1
1ITER-India, Institute for Plasma Research, India 2Institute for Plasma Research, India Email: [email protected]
In Diagnostic Neutral Beam (DNB) [1], High Voltage (HV) bushing is an interface between the HV transmission line and Beam source. For validating the design of HVB [2] of DNB, a scaled down configuration of the Bushing is fabricated, referred to as PHVB and assembled. This PHVB is to be subjected to long duration HV tests up to 60 kV under vacuum conditions for verifying the voltage holding capacity of the bushing.
For automating the entire experimental process of HVB experiment and acquiring important experimental data up to 3600 sec for post analysis, a dedicated Data Acquisition and Control System (DACS) is required. This will help in understanding the behavior of the PHVB in High voltage and vacuum environment. Also it is required that high speed breakdown events are monitored. CRIO (compact reconfigurable input output) is a rugged, small sized hardware platform which combines the power of real time processor and FPGA bus. It is becoming an upcoming standard for medium sized experiment. For ensuring smooth control operation of the experiment; NI CRIO was selected as the controller for DACS.
In this experiment the CRIO provides a user interface for setting of important control set points of power supply and vacuum system. Also it provides seamless control and acquisition for pulse durations of up to 3600 sec. The voltage signal is generated as a ramp signal with upper voltage set point. The ramp rate applied is 1.5kV/min and is user configurable.
The testing of PHVB has been successfully completed using the developed DACS. In this paper the technical details of DACS design, implementation and test results shall be discussed.
References:
[1] Diagnostic Neutral Beam for ITER—Concept to Engineering, A. Chakraborty et. al, at IEEE Transactions on Plasma Science, Vol. 38, no. 3, March 2010
[2] Design optimization of the 100 kV HV bushing for ITER-DNB, S Shah et al, at Symposium of Fusion Technology (SOFT), 2009
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Abstract ID: 1_145
Rotor-dynamic Design Aspects for a Variable Frequency Drive Based High Speed Cryogenic Centrifugal Pump in Fusion Devices
Jotirmoy Das1, Hitensinh Vaghela1, Ritendra Bhattacharya1, Pratik Patel1, Vinit Shukla1, Nitin Shah1, Biswanath Sarkar1
1ITER-India, Institute for Plasma Research, India Email: [email protected]
Superconducting magnets of large size are inevitable for fusion devices due to high magnetic field requirements. Forced flow cooling of the superconducting magnets with high mass flowrate of the order ~3 kg/s is required to keep superconducting magnets within its safe operational boundaries during various plasma scenarios. This important requirement can be efficiently fulfilled by employing high capacity and high efficiency cryogenic centrifugal pumps. The efficiency > 70% will ensure overall lower heat load to the cryoplant. Thermo-hydraulic design of cryogenic centrifugal pump revealed that to achieve the operational regime with high efficiency, the speed should be ~ 10,000 revolutions per minute. In this regard, the rotor-dynamic design aspect is quite critical from the operational stability point of view. The rotor shaft design of the cryogenic pump is primarily an outcome of optimization between thermal heat-in leak at cryogenic temperature level from ambient, cryogenic fluid impedance and designed rotation speed of the impeller wheel. The paper describes the basic design related to critical speed of the rotor shaft, rotor whirl and system instability prediction to explore the ideal operational range of the pump from the system stability point of view. In the rotor-dynamic analysis, the paper also describes the Campbell plots to ensure that the pump is not disturbed by any of the critical speeds, especially while operating near the nominal and enhanced operating modes.
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Abstract ID: 1_156
Quench Detection, Protection and Simulation Studies on SST-1 Magnets
Aashoo N Sharma1, Yohan Khristi1, Subrata Pradhan1, Kalpesh Doshi1, Upendra Prasad1, Moni Banaudha1, Pankaj Varmora1, Bhadresh R Praghi1
1Institute for Plasma Research, India Email: [email protected]
Steady-state Superconducting Tokamak-1 (SST-1) is India’s first tokamak with superconducting toroidal field (TF) and Poloidal Field (PF) magnets [1]. These magnets are made with NbTi based Cable-In-Conduit-Conductors [2].
The quench characteristic of SST-1 CICC has been extensively studied both analytically and using simulation codes. Dedicated experiments like model coil test program, TF coil test program and laboratory experiments were conducted to fully characterize the performance of the CICC and the magnets made using this CICC.
Results of quench experiments performed during these tests have been used to design the SST-1 quench detection and protection system. Simulation results of TF coil quenches and slow propagation quench of TF busbars have been used to further optimize these systems during the SST-1 tokamak operation. Redundant hydraulic based quench detection is also proposed for the TF coil quench detection. This paper will give the overview of these development and simulation activities.
References:
[1] S. Pradhan et al. “Superconducting Magnets of SST-1 Tokamak”, 20th IAEA Fusion Energy Conference (2004), FT3-4Rb
[2] S. Pradhan et al. “Superconducting cable-in-conduit-conductor for SST-1 magnets”, in Proc. 2nd IAEA TCM on Steady-state Operation of fusion devices, Fukuoka, Japan, 1999, vol. II, p. 482.A.B. Author, “Title of paper,” Title of Journal, 1, 100 (2009).
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Abstract ID: 1_166
Gas Fueling System for SST-1
Kalpeshkumar R Dhanani1, Ziauddin Khan1, Dilip Raval1, Pratibha Semwal1, Siju George1, Yuvakiran Paravastu1, Prashant Thankey1, Mohammad Shoaib Khan1, Subrata Pradhan1
1Institute for Plasma Research, India Email: [email protected]
SST-1 Tokamak, the first Indian Steady-state Superconducting experimental device is at present under operation in Institute for Plasma Research. For plasma break down & initiation, the piezoelectric valve based gas feed system is implemented as primary requirement due to its precise control, easy handling, low costs for both construction and maintenance and its flexibility in working gas selection. The main functions of SST-1 gas feed system are to feed the required amount of ultrahigh purity hydrogen gas for specified period into the vessel during plasma operation and ultrahigh helium gas for glow discharge cleaning. In addition to these facilities, the gas feed system is used to feed a mixture gas of hydrogen and helium as well as other gases like nitrogen and Argon during divertor cooling etc. The piezoelectric valves used in SST-1 are remotely driven by a PXI based platform and are calibrated before the plasma operation during each SST-1 plasma operation with precise control. This paper will present the technical development and the results of gas fueling in SST-1.
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Abstract ID: 1_169
Development of Electromagnetic Welding Facility of Flat Plates for Nuclear Industry
Rajesh Kumar1, Subhanarayan Sahoo1, Biswanath Sarkar1, Anurag Shyam1
1Institute for Plasma Research, India Email: [email protected]
Electromagnetic pulse welding (EMPW) process, one of high speed welding process uses electromagnetic force from discharged current through working coil, which develops a repulsive force between the induced current flowing parallel and in opposite direction. For achieving the successful weldment using this process the design of working coil is the most important factor due to high magnetic field on surface of work piece [1].
In case of high quality flat plate welding factors such as impact velocity, angle of impact standoff distance, thickness of flyer and overlap length have to be chosen carefully. All the parameters should be optimized because above or below the optimized value, it is impossible to get high quality welding of flat components. Electromagnetic pulse welding of flat components has been studied in detail by many researches due to its advantages of increased formability and reduced spring back than other welding methods [2].The feasibility of electromagnetic welding of sheets has been established, but the effect of process parameters on the weld quality has not been justified properly.
The present study investigates the effect of parameters on welding quality of flat sheets, which has wide applications in nuclear industry, automotive industry, aerospace, electrical industries. However formability and weld ability still remain major issues. The EMPW process for flat sheets and axi-symmetric components has been studied in details by many researchers. Due to ease in controlling the magnetic field enveloped inside tubes, the EMPW has been widely used for tube welding [3]. In case of flat components control of magnetic field is difficult. Hence the application of EMPW gets restricted.
The present work attempts to make a novel contribution by investigating the effect of process parameters on welding quality. The work emphasizes the approaches and engineering calculations required to effectively use of actuator in EMPW.
References:
[1] Ji-Yeon Shim, Bong-Yong Kang, “Distribution of Electromagnetic Force of Square Working Coil for High-Speed Magnetic Pulse Welding Using FEM” Materials Sciences and Applications, 4, 856-862 (2013).
[2] S. D. Kore, J. Imbert, M. J. Worswick and Y. Zhou “Electromagnetic impact welding of Mg to Al sheets” Science and Technology of Welding and Joining,14,549-553(2009).
[3] S. D. Kore, P.P. Date, S.V. KulkarniEffect of process parameters on electromagnetic impact weldingof aluminum sheets, International Journal of Impact Engineering 34, 1327–1341(2007).
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_170
Engineering Design & Integration of Radial Control Coil in Vacuum Vessel of SST-1
Pradeep Chauhan1, Prosenjit Santra1, Snehal Jaiswal1, Prabal Biswas1, Kirit R Vasava1, Tejas Parekh1, Hiteshkumar Patel1, Subrata Pradhan1
1Institute for Plasma Research, India Email: [email protected]
Due to unsymmetrical placement of toroidal field coil inside the vacuum vessel, which generates field error and tend to push the plasma from its major radius 1100 mm to towards inboard side. Hence it was require install the Radial control coil (RCC) at a location of 1300 mm radius and elevation of 350 mm above and below the mid-plane of the toroidal field coil. The radial control coil is decided to make from multi-strand flexible super conducting cable encased inside the prefabricated SS 304 L piped casing made in four segment and in-situ welded together inside the vacuum vessel to form the shape of coil. The radial control coil is open to atmosphere and experiencing the vacuum inside the vacuum vessel. To maintain the circular shape of the copper cable inside the SS casing, very close tolerances are maintained e.g. super-conducting cable has outer diameter of 14 mm and after FRP insulation and Teflon rapping the outer diameter reaches to 16 mm while inner diameter of the pipe is 18 mm. This paper will present the design drivers, material selection, advantages and constraints of the RC coils, its conceptual and engineering design, CAD models, finite element analysis using ANSYS, its fabrication, quality assurance/control and assembly/integration aspects inside vacuum vessel of SST-1.
120 | P a g e
10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_171
Engineering Design & Integration of In-vessel Single Turn Segmental Coil in Vacuum Vessel of SST-1
Snehal Jaiswal1, Pradeep Chauhan1, Prosenjit Santra1, Kirit R Vasava1, Tejas Parekh1, Hiteshkumar Patel1, Prabal Biswas1, Subrata Pradhan1
1Institute for Plasma Research, India Email: [email protected]
SST-1 tokamak is having the error field due to unsymmetrical positioning of Toroidal field (TF) coils which displace the plasma to inboard side from its major radius of 1100 mm, hence it is required to install the In-vessel Coil (PF6) at a radial location of 1350 mm and elevation of 350 mm above and below the mid-plane of the toroidal field coils for proper plasma positioning. The In-Vessel coil was decided to make in eight segments for futuristic use, to control the individual localized error field correction by supplying the different current. A single turn, eight segments, copper conductor with 18 mm diameter with GFRP insulation and housed in SS304 L casing to carry 8000 Ampere current for 10 sec duration was designed, fabricated and installed in vacuum vessel of SST-1. This paper will present the design drivers, material selection, advantages and constraints of the in-vessel coils, its conceptual and engineering design, CAD models, finite element analysis using ANSYS, its fabrication, quality assurance/control and assembly/integration aspects inside vacuum vessel of SST-1.
121 | P a g e
10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_174
Quality Control of FWC during Assembly/Commissioning on SST-1
Hiteshkumar Patel1, Prosenjit Santra1, Snehal Jaiswal1, Pradeep Chauhan1, Yuvakiran Paravastu1, Siju George1, Gattu Ramesh Babu1, Arun Prakash A1, Pratibha Semwal1, Prasant Thankey1, Kalpeshkumar R Dhanani1, Dilip Raval1, Ziauddin Khan1, Subrata Pradhan1, Tejas Parekh1, Prabal Biswas1
1Institute for Plasma Research, India Email: [email protected]
First Wall components (FWC) of SST-1 tokamak, which are in the immediate vicinity of plasma comprises of limiters, divertors, baffles, passive stabilizers are designed to operate long duration (1000 s) discharges of elongated plasma. All FWC consists of a copper alloy heat sink modules with SS cooling tubes brazed onto it, graphite tiles acting as armour material facing the plasma, and are mounted to the vacuum vessels with suitable Inconel support structures at ring & port locations. The FWC are very recently assembled and commissioned successfully inside the vacuum vessel of SST-1 under going a rigorous quality control and checks at every stage of the assembly process. This paper will present the quality control and checks of FWC from commencement of assembly procedure, namely material test reports, leak testing of high temperature baked components, assembled dimensional tolerances, leak testing of all welded joints, graphite tile tightening torques, electrical continuity of passive stabilizers, and electrical isolation of passive stabilizers from vacuum vessel, baking and cooling hydraulic connections inside vacuum vessel.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_175
Laser Shock Peening of Stainless Steel Surfaces: ns vis-ã-vis ps Laser Pulses
Prem Kiran P1, Pardhu Yella1, Koteswararao V Rajulapati1, Venkateshwarlu Pinnoju1, Ramesh Kumar Buddu2, Bhanu Sankara Rao Kota3
1University of Hyderabad, India 2Institute for Plasma Research, India 3Mahatma Gandhi Institute of Technology, India Email: [email protected]
Laser Shock Peening (LSP), an advanced surface treatment technique used to improve the strength of structural materials used in aeronautical and reactor industries has become the most sought after material processing technique [1, 2]. Laser pulses of different durations are employed to generate plasma from the materials of interest in either direct ablation or confined mode. This expanding plasma launches a shockwave into the material due to momentum conservation generating compressive residual stresses which in turn enhancing the material strength.
We present the response of SS304 and SS 316L(N) specimens to LSP using nanosecond (ns) and picosecond (ps) laser pulses in direct and confined ablation mode. Structural changes are studied using optical microscopy (OM), scanning electron microscopy (SEM) and X-ray diffraction (XRD). The microstructure has changed considerably in both the ablation modes. Though the direct ablation mode has shown a tensile residual stress on the surface of the sample, as expected, but up to a depth of 0.5 mm the compressive nature remained intact. In the confined ablation mode, the role of different sacrificial layers used, studied using the extracted precise lattice parameters from XRD indicated the presence of microstrain in ns- and ps-LSP. In both the ns- LSP and ps-LSP, the confined ablation mode has shown an improved material properties. The correlation between the laser energy coupled to the specimens, residual stresses induced to the specimens during ns- and ps-LSP in both the ablation modes will be discussed to bring out the advantages the technique.
References:
[1] K. Ding and L. Ye., “Laser Shock Peening Performance and process simulation”, Woodhead Publishing Limited, (2006).
[2] Z. Yongkang L. Zinjhong, L. Kaiyou, “Laser shock processing of FCC metals – Mechanical Properties and micro-structural strengthening mechanism”, Springer series (2013).
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_176
Assembly & Metrology of First Wall Components of SST-1
Tejas Parekh1, Prosenjit Santra1, Prabal Biswas1, Hiteshkumar Patel1, Yuvakiran Paravastu1, Snehal Jaiswal1, Pradeep Chauhan1, Gattu Ramesh Babu1, Arun Prakash A1, Dhaval Bhavsar1, Dilip Raval1, Ziauddin Khan1, Subrata Pradhan1
1Institute for Plasma Research, India Email: [email protected]
First Wall components (FWC) of SST-1 tokamak, which are in the immediate vicinity of plasma comprises of limiters, divertors, baffles, passive stabilizers are designed to operate long duration (1000 s) discharges of elongated plasma. All FWC consists of a copper alloy heat sink modules with SS cooling tubes brazed onto it, graphite tiles acting as armour material facing the plasma, and are mounted to the vacuum vessels with suitable Inconel support structures at ring & port locations. The FWC are very recently assembled and commissioned successfully inside the vacuum vessel of SST-1 under going a meticulous planning of assembly sequence, quality checks at every stage of the assembly process. This paper will present the metrology aspects & procedure of each FWC, both outside the vacuum vessel, and inside the vessel, assembly tolerances, tools, equipment and jig/fixtures, used at each stage of assembly, starting from location of support bases on vessel rings, fixing of copper modules on support structures, around 3800 graphite tile mounting on 136 copper modules with proper tightening torques, till final toroidal and poloidal geometry of the in-vessel components are obtained within acceptable limits, also ensuring electrical continuity of passive stabilizers to form a closed saddle loop, electrical isolation of passive stabilizers from vacuum vessel.
124 | P a g e
10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_189
Trap Site Formation and their Distribution Studies in Porous Lithium Titanate
Chandan Danani1, Manoj Warrier2, Paritosh Chaudhuri1
1Institute for Plasma Research, India 2Bhabha Atomic Research Centre-Visakhapatnam, India Email:[email protected]
Lithium based ceramics (LiSiO4, Li2TiO3) in the packed pebble form, used in solid breeder blankets, are promising candidates for future fusion reactors [1]. Neutron interactions with these ceramic breeders not only generates tritium but also causes formation of Primary Knock on Atoms (PKA) which trigger collision cascade and it leads to the creation of open bonds & vacancies. Tritium can diffuse and get re-trapped at the damage sites produced by the PKAs. As it diffuses it can also react to form molecules which have different diffusion properties as compared to its constituents. Diffusion of tritium is affected by the damage of the ceramic pebbles and quantification of damage and its distribution plays an important role in estimating the tritium extraction from the solid breeder blankets. Radiation transport calculation can provide the neutron spectrum in ceramic pebbles which can be used as a input to radiation damage tool SPECTER [2] to obtain the PKA energy spectrum. The PKA energy spectrum is then used to find the number of displacements using Monte Carlo simulations based on the binary collision approximation (BCA) [3] and compared with the NRT model [4].
References:
[1] L. Giancarli, V. Chuyanov, M. Abdou,M. Akiba, B.G. Hong, R. Lasser, C. Pan, Y. Strebkov, and the TBWG, “Breeding blanket modules testing in iter: An international program on the way to demo”, Fusion Eng. Design, 81:393–405, (2006).
[2] L. R. Greenwood and R. K. Smither, “SPECTER: Neutron Damage Calculations for Materials Irradiations”, ANL/FPP-TM-197, (1985).
[3] W. Eckstein., “Computer simulations of ion–solid interactions”, Springer series in material science 10, Springer-Verlag, (1991).
[3] M.J. Norgett, M.T. Robinson, I.M. Torrens, “A Proposed Method for Calculating Displacement Dose Rates”, Nucl. Eng. Des. 33 (1975) 50.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_191
Design of a High Power Water Load for LHCD System of SST-1 Tokamak
Harish V Dixit1, Aviraj Jadhav1, Yogesh M Jain2, Alice Cheeran1, Vikas Gupta3, Pramod K Sharma2
1Veermata Jijabai Technological Institute, India 2Institute for Plasma Research, India 3Vidyavardhini College of Engineering and Technology, India Email: [email protected]
Water loads are traditionally used in lower hybrid current drive (LHCD) system to condition klystrons at full power or used in conjunction with a circulator to protect the tube from high VSWR. Various designs of water load are available in the literature. However the availability of indigenous design of such water loads is limited. This paper presents a novel indigenous structure of S Band water load capable of absorbing 250 kW CW power at 3.7 GHz for LHCD installed on SST1 machine. Further this paper also presents generalized equations through which the design can be reproduced at different frequencies without much effort. As water is an excellent absorber at S band (loss tangent=0.0036), it is used as an absorber in this design. Conventional designs of water load usually use a quarter wave transformer to match the impedance of water to the air filled waveguide. However this matching along with the issue of bonding the ceramic to the waveguide is often critical and if not done properly can lead to arcing and breakdown. In this design, water flowing in Teflon/polypropylene/glass channels arranged in a tapered configuration is used to absorb the incident microwave power. The tapering provides impedance matching. The flow is regulated and maintained at a rate so as to avoid localized boiling of water. The water load is designed and tested in COMSOL and ANSYS and exhibits a VSWR of < 1.1 and a bandwidth of 10%.
126 | P a g e
10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_192
Design of Multiple Ferrite Tile Phase Shifters for Applications in High CW Power Differential Phase Shift Circulators
Harish V Dixit1, Aviraj Jadhav1, Yogesh M Jain2, Alice Cheeran1, Vikas Gupta3, Pramod K Sharma2
1Veermata Jijabai Technological Institute, India 2Institute for Plasma Research, India 3Vidyavardhini College of Engineering and Technology, India Email: [email protected]
The LHCD System of SST-1 Tokamak at IPR, Gandhinagar uses four Klystrons (TH 2103D) each supplying 500 kW power to drive the plasma current non inductively. However these tubes are often susceptible to damage due to a high VSWR. As such, circulators are employed to route the reflections to a dummy load thereby protecting the klystron from damage.
Differential Phase Shift Circulators (DPSC) which are usually composed of magic tee, ferrite phase shifters and couplers are often preferred at a high CW power level over conventional junction circulators due to their higher power handling capacity. The power handling capacity of circulators is often limited by the power handling capacity of the phase shifter section. At a high CW power level, the cooling of the ferrite is of prime importance. As such it is required to have a larger contact area of the ferrite material with the waveguide so that better cooling arrangements are provided.
It is thus advantageous to use multiple ferrite tiles in the phase shifter to maximize the power handling capacity of the phase shifter. This paper presents the low power prototype design of a ferrite phase shifter at 3.7 GHz to be used in the circulator of LHCD System of SST-1 tokamak. The paper also analyses the phase characteristics and the thermal power dissipation of the phase shifter with single, two and four tile ferrites along with stacked waveguides multiple ferrite tiles configuration.
127 | P a g e
10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_193
Conceptual Design of PAM Antenna for Aditya-U Tokamak
Yogesh M Jain1, Pramod K Sharma1, Jagabandhu Kumar1, Harish V Dixit2, Kirankumar K Ambulkar1, Pramod R Parmar1, Chetan G Virani1
1Institute for Plasma Research, India 2Veermata Jijabai Technological Institute, India Email: [email protected]
ADITYA Tokamak is being upgraded (ADITYA-U) to operate the machine at enhanced plasma parameter. This also provides an opportunity to upgrade lower hybrid current drive (LHCD) system to drive plasma current non-inductively and enhance the coupling of RF power to the plasma. It is proposed to replace existing grill antenna [1] by a new type of antenna which is often referred as passive active multijunction (PAM) antenna [2]. The PAM antenna has an advantage of providing efficient RF coupling to the plasma, even at edge densities close to cut- off. Further it provides a lower reflection from the plasma as compared to the conventional grill antenna.
ADITYA-U would operate at toroidal magnetic field of 1.5T and may have line average density lying in the range of [0.8 – 3.0] 1019 m-3. For LHW’s to access to the plasma center, the waves would be launched having parallel refractive index (N||) which is well above the critical accessible condition given by Stix [3]. Thus the PAM antenna is designed to launch N|| of 2.25 0.28. Our analysis reveals that periodicity for the PAM antenna would be 27mm to launch the design value of N|| with three passive and three active waveguide in a single PAM module having phase shift of 270o between adjacent active waveguides. The size of the radial port (490 mm x 190 mm) of ADITYA-U tokamak determines the number of PAM modules which may be accommodated in the new scheme. It turns out that two modules of PAM antenna may be accommodated in the said radial port. Mode convertors (TE10 to TE30 mode) would be employed for dividing the RF power in three poloidal locations. A thermal and electro-mechanical analysis is also discussed in this paper.
References:
[1] P. K. Sharma, S. L. Rao, D. Bora, R. G. Trivedi, et. al., "Commissioning of 3.7GHz LHCD system on ADITYA tokamak and some initial results", Fusion Engg. & Design, 82, 41 (2007).
[2] P. Bibet, X. Litaudon, D. Moreau, "Conceptual Study of a Reflector Waveguide Array for Launching Lower Hybrid Waves In Reactor Grade Plasmas", Nuclear Fusion, Vol. 35, No. 10 (1995).
[3] T. H. Stix “Waves in Plasmas”, Springer Science & Business Media (1992).
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_195
Assessment of Delta Ferrite in Multipass TIG Welds of 40 mm Thick SS 316L Plates: A Comparative Study of Ferrite Number (FN) Prediction and Experimental Measurements
Ramesh Kumar Buddu1, Shamsuddin Shaikh1, Prakash M Raole1, Biswanath Sarkar2
1Institute for Plasma Research, India 2ITER-India, Institute for Plasma Research, India Email: [email protected]
Austenitic stainless steels are widely used in the fabrication of fusion reactor major systems like vacuum vessel, divertor, cryostat and other major structural components development. AISI SS316L materials of different thicknesses are utilized due to the superior mechanical properties, corrosion resistance, fatigue and stability at high temperature operation. The components are developed by using welding techniques like TIG welding with suitable filler material. Like in case of vacuum vessel, the multipass welding is unavoidable due to the use of high thickness plates (like in case of ITER and DEMO reactors). In general austenitic welds contains fraction of delta ferrite phase in multipass welds. The quantification depends on the weld thermal cycles like heat input and cooling rates associated with process conditions and chemical composition of the welds. Due to the repeated weld thermal passes, the microstructure adversely alters due to the presence of complex phases like austenite, ferrite and delta ferrite and subsequently influence the mechanical properties like tensile and impact toughness of joints. Control of the delta ferrite is necessary to hold the compatible final properties of the joints and hence its evaluation vital before the fabrication process. The present paper reports the detail analysis of delta ferrite phase in welded region and heat affected zones of 40 mm thick SS316L plates welded by special design multipass narrow groove TIG welding process under three different heat input conditions (1.67 kJ/mm, 1.78 kJ/mm, 1.87 kJ/mm). The correlation of delta ferrite microstructure with optical microscope and high resolution SEM has been carried out and different type of acicular and vermicular delta ferrite structures is observed. This is further correlated with the non destructive magnetic measurement using Ferrite scope. The measured ferrite number (FN) is correlated with the formed delta ferrite phase. The chemical composition of weld samples is used to predict the FN with Schaeffler’s, DeLong and WRC-1992 diagram by calculating the Creq and Nieq ratios and compared with experimental data of FN from Feritescope measurements. The low heat input conditions (1.67 kJ /mm) have produced higher FN (7.28) , medium heat input (1.72 kJ/mm) shown FN (7.04) where as high heat input (1.87 kJ/mm) conditions has shown FN (6.68).
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_197
Study of Transients in Liquid Helium Flow during Cool Down of Cryopanel
Reena Sayani1, Samiran Shanti Mukherjee1, Ranjana Gangradey1
1Institute for Plasma Research, India Email: [email protected]
In cryo-sorption Cryopump hydroformed cryopanels are cooled down below temperature 5 K to adsorb hydrogen and helium gases. The panels are coated with activated carbon as sorbent. Sorbent with micro-pores adsorbs gases and the pores get saturated after certain duration of pumping operation. On regenerating by increasing the panel temperature adsorbed gases get removed. A cycle of operation is thus followed comprising, cool down from 300 K to ~ 5 K and warm up from ~ 5 K to 300 K. The work presented in this paper describes cool down process of an indigenously developed sorbent based cryopump by flowing compressed helium through its cryopanel. The pump was tested for its pumping speed at small scale cryopump facility (SSCF). SSCF hydraulic network is described with hydro formed panel of size 500 mm (l) 100 mm (w) with a sheet thickness of 1.5mm connected by inlet and outlet tubes. Numerical analysis of cool down process of SSCF is done by solving equations of mass, momentum and energy conversation. Cool down time required to reach steady state flow conditions is approximated. Also transient parameters of helium are estimated during cool down of SSCF.
References:
[1] Basic and applications of Cryopump, C Day, ITP, Forschungzentram, Karlsruhe, Germany, 2011.
[2] Cryogenic subsystem to provide for nominal operation and fast regeneration of the ITER primary cryo-sorption vacuum pump, V. Kalinine, R. Haange, N. Shatil, F. Millet, I. Guiliment, M. Wykes, C. Day, A. Mack AIP proceeding, 2008. Alaska.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_202
A Simple In-vessel/FW Component Viewing System for SST-1
Prosenjit Santra1, Prabal Biswas1, Kirit R Vasava1, Snehal Jaiswal1, Tejas Parekh1, Pradeep Chauhan1, Hiteshkumar Patel1, Subrata Pradhan1
1Institute for Plasma Research, India Email: [email protected]
A simple compact system is being proposed for in-situ visual inspection of around 3800 First Wall (FW) graphite (armour) tiles in the vacuum vessel of SST-1 tokamak. The 2 DOF, manual driven system (permanently stationed inside vacuum vessel behind outer passive stabilizer) at top and bottom mid-plane locations consist of a rack and pinion mechanism operating a arm with a CCD camera/LED mounted on it, moving over a cam profile to cover approximately 1/8th of the toroidal span of the vacuum vessel both at interior top/bottom locations with in the FW modules. The camera and LED light should withstand the ultrahigh vacuum conditions, prolonged baking temperatures of around 200o C along with high electromagnetic forces inside the vessel. This system can be operated remotely in-between shots from outside the VV through a linear motion feed through providing linear moment to a rack & pinion mechanism connected to the arm.
This mechanism provides a better viewing of the inside FW components and vessel wall surface of tokamak with simple engineering & operational effort. Any information can be acquired from system regarding damages to FWC due to interaction with plasma as well as damage of other support structures inside VV.
In comparison to more complicated and complex inspection system used in other tokamaks, this mechanism can be used for frequent in vessel visual inspection, which limits the system to be small, simple, occupying less space and custom made. This system is cheap with a minimum time for realization of the concept.
The paper will present the conceptual and engineering design aspect of the in-viewing system, CAD images, its advantages and limitations, camera & LED details, data acquisition and the present status of realization of the project.
131 | P a g e
10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_208
Overall Behaviour of PFC Integrated SST-1 Vacuum System
Ziauddin Khan1, Dilip Raval1, Yuvakiran Paravastu1, Pratibha Semwal1, Kalpeshkumar R Dhanani1, Siju George1, Mohammad Shoaib Khan1, Arun Prakash A1, Gattu Ramesh1, Prashant Thankey1, Firozkhan Pathan1, Subrata Pradhan1
1Institute for Plasma Research, India Email: [email protected]
As a part of phase-I up-gradation of Steady-state Superconducting Tokamak (SST-1), Graphite Plasma Facing Components (PFCs) have been integrated inside SST-1 vacuum vessel as a first wall (FW) during Nov 14 and May 2015. The SST-1 FW has a total surface area of the installed PFCs exposed to plasma is ~ 40 m2 which is nearly 50% of the total surface area of stainless steel vacuum chamber (~75 m2). The volume of the vessel with the PFCs is ~ 16 m3. After the integration of PFCs, the entire vessel as well as the PFC cooling/baking circuits has been qualified with an integrated leak tightness of < 1.0 10–8 mbar l/s. The pumping system of the SST-1 vacuum vessel comprises of one number of Roots’ pump, four numbers of turbomoleculars and a cryopump. After the initial pump down, the PFCs were baked at 250 °C for nearly 200 hours employing hot nitrogen gas to remove the absorbed water vapours. Thereafter, Helium discharges cleaning were carried out towards removal of surface impurities. The pump down characteristics of SST-1 vacuum chamber and the changes in the residual gaseous impurities after the installation of the PFCs will be discussed in this paper.
132 | P a g e
10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_209
Assembly of Neutral Beam Injector with SST-1
Rambabu Sidibomma1, Prahlad Vattipalle1, Sanjeev Kumar Sharma1, Sridhar B1, Laxmi Narayan Gupta1, Ujjwal Kumar Baruah1
1Institute for Plasma Research, India Email: [email protected]
Neutral Beam Injector (NBI) is capable of delivering a hydrogen beam of power 1.7 MW to the SST-1 tokomak for the purpose of heating its plasma. The Steady State Superconducting Tokamak (SST-1) is the core project aimed at producing high temperature plasma. The Neutral Beam Injector (NBI) is a system meant for heating the SST-1 plasma.
NBI system is used for generating a beam of energetic hydrogen particles and then launches them into the SST-1. The NBI system is currently being operated for production of such a beam on a designated test stand in the NBI hall. As a next step, it is now required to transfer the entire NBI system from the test stand (in NBI hall) to the NBI-SST-1 area and then integrate with the SST-1 Tokamak.
The NBI system comprises of a huge vacuum vessel with an ion source and gate valve mounted on it. The vacuum vessel contains the following major sub-systems such as neutralizer, electromagnet (magnet), magnet liner, calorimeter, Ion dump, Beam Transmission Duct, Shine- Through and cryo-condensation pumps (cryopumps). It also contains headers and distribution systems for liquid nitrogen, liquid helium and cooling water, external vacuum system, external cryogenic distribution, external cooling water distribution and snubber deck.
NBI integration with SST-1 involves assembly sequence of activities, Heat Transfer Elements welding with neutraliser, ion dump, magnet liner and calorimeter, dis-mantling of existing cooling water lines, dis-assembly of snubber deck, shifting of Vacuum Vessel (VV), lifting of VV and placing VV on the Support Structure, and alignment of VV with SST-1 at pre-defined position. In this paper, we present the planning, sequence of assembly activities, VV lifting methodology.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_216
Experience of 12 kA / 16 V SMPS during the HTS Current Leads Test
Pradip Panchal1, Dikens Christian1, Rohitkumar Panchal1, Dashrath Sonara1, Gaurav Purwar1, Atul Garg1, Hiren Nimavat1, Gaurav Kumar Singh1, Jal Patel1, Vipul L Tanna1, Subrata Pradhan1
1Institute for Plasma Research, India Email: [email protected]
As a part of Upgradation plans in SST-1 Tokamak Machine, a (+/-) pair of 3.3 kA rated prototype hybrid current leads were developed using Di-BiSCCO as High Temperature Superconductors (HTS) and the Copper Heat exchanger (77 K – 300 K). In order to validate the manufacturing procedure prior to go for series production of such current leads, it was recommended to test these current leads using dedicated and very reliable switch mode DC power supply. A dedicated 12 kA, 16 VDC high current, low voltage programmable switch mode power supply (SMPS) is already installed and successfully commissioned and tested as part of test facility. This power supply has special features such as modularity (8 modules), N+1 redundancy, very low ripple voltage (< 8 mVrms), precise current measurements with Direct Current - Current Transformer, CC/CV modes with auto-crossover and auto-sequence programming. As a part of acceptance of this converter, A 5.8 mΩ water-cooled low resistive dummy load and PLC based SCADA system is designed, developed for commissioning of power supply. The same power supply was used for the testing of the prototype HTS current leads connected via 1 mΩ feeder cooled using liquid nitrogen at 77 K. The paper describes the salient features and state-of-art of power supply. Experience and results obtained from this converter during the HTS current leads test especially for lower current operation are discussed.
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Abstract ID: 1_217
Calibration of Low Temperature Measurement System for the Superconducting Magnet System for the SST-1
Bhadresh R Praghi1, Yohan Khristi1, Subrata Pradhan1, Pankaj Varmora1, Upendra Prasad1
1Institute for Plasma Research, India Email: [email protected]
SST- 1 Magnet Division, IPR is involved in the operations and maintenance activity of sensor signal conditioning of ~150 temperature measurement channels used for low temperature measurement of CICC made super conducting magnets. Measurement of cryogenic temperature of different active and passive locations in the Steady State superconducting Tokamak (SST-1) machine is needed to be accurate and reliable. For reliable and safe operation of the magnet system, it is necessary to measure the temperature information with 0.1 K accuracy on 4 to 5 K operation temperature of the magnets [1].
The signal conditioning, excitation current sources and VMEbus chassis Hardware with associated analog acquisition cards add offset in order of mV to final acquired voltage and hence in converted temperature. The non-linear negative temperature coefficient (NTC) temperature sensor has lower sensitivity at room temperature, therefore gives the 3K to 5K errors at the room temperature measurement. An error in the measurement makes difficult to establish relation between cryogenic condition and temperature of particular portion of the machine. Therefore a combine calibration of the temperature system with maximum error of 0.5 to 1 K between actual and measured temperature with DAQ [2] is needed to be carried out. Prior to plasma campaigns is required to minimize the error due to thermal drift in offset voltage (mV/K) and permanent DC shift (order of 1 to 3 mV)in signal conditioning electronics, error in excitation current sources (order of 10 nA) and offset (order of 1 to 3 mV ) in data acquisition analog input modules [2]. Systematically offset is compensated at each stage by proper calibration techniques to obtain minimum required accuracy in measured temperature. Conclusively we were able to minimize error in temperature measurement up to 0.5 to 1 K between actual and measured temperature through DAQ [2].
This poster describes the temperature measurement system, results and its calibration procedure of the SST-1 superconductor magnet system.
References:
[1] Kalpesh Doshi, “Title of Precision Signal Conditioning Electronics for Cryogenic Temperature and Magnetic Field Measurements in SST-1paper,” IEEE Transactions on Plasma science,” Vol 40, No.3 (2012).
[2] Kalpesh Doshi, “Design and implementation of data acquisition system for magnets of SST-1”, Fusion Engineering and Design 89 (2014) 679–683
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Abstract ID: 1_219
Electronics for Coupled High Voltage Measurement on PF Magnets of SST-1
Moni Banaudha1, Yohan Khristi1, Subrata Pradhan1, Azadsinh Makwana1, Upendra Prasad1, Devenkuram H Kanabar1
1Institute for Plasma Research, India Email: [email protected]
Steady State Superconducting Tokamak-1 (SST-1) machine is in operation phase and the different plasma campaigns are carried out for the plasma study. The Toroidal Field (TF) magnet system of SST-1 is operated up to 5 kA of nominal current at the helium cooled condition of 4.5 K. During Plasma discharge, PF magnets are coupled high voltage (> 1 kV) due to the operation of Ohmic Transformer (OT) and Vertical Field Coils (VF) as a phenomenon of transformer action,may damage the PF magnet system if high voltage cross the insulation level of 1 kV. 40 Channels of highly reliable and accurate signal conditioning electronic system developed with differential signal input, isolated output and improved signal to noise ratio for high noise spicks coupled from the noisy tokanak environments. The electronics system is working fine and reliable throughout the all plasma campain for the each and individual 9 PF-Magnets as well as the different layers of the PF Magnet Systems. The information about induced high voltage on PF magnets by the electronics which helps to carry on the experiment in Safe margin and decision making for experimental parameters.
This paper describes the reliable electronics measurement system, a precaution taken on measurement and results of the electronics system in all plasma campaigns.
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Abstract ID: 1_224
Electronics and Instrumentation for the SST-1 Superconducting Magnet System
Yohan Khristi1, Subrata Pradhan1, Pankaj Varmora1, Moni Banaudha1, Bhadresh R Praghi1, Upendra Prasad1
1Institute for Plasma Research, India Email: [email protected]
Steady State Superconducting Tokamak-1 (SST-1) at Institute for Plasma Research (IPR), India is now in operation phase. The SST-1 magnet system consists of sixteen superconducting (SC), D-shaped Toroidal Field (TF) coils and nine superconducting Poloidal Field (PF) coils together with a pair of resistive PF coils, inside the vacuum vessel of SST-1. The magnets were cooled down to 4.5 K using either supercritical or two-phase helium, after which they were charged up to 10 kA of transport current. Precise quench detection system, cryogenic temperature, magnetic field, strain, displacement, flow and pressure measurements in the Superconducting (SC) magnet were mandatory.
The Quench detection electronics required to protect the SC magnets from the magnet Quench therefore system must be reliable and prompt to detect the quench from the harsh tokamak environment and high magnetic field interference. A ~200 channels of the quench detection system for the TF magnet are working satisfactorily with its design criteria. Over ~150 channels Temperature measurement system was implemented for the several locations in the magnet and hydraulic circuits with required accuracy of 0.1K at bellow 30K cryogenic temperature. Whereas the field, strain and displacement measurements were carried out at few predefined locations on the magnet. More than 55 channels of Flow and pressure measurements are carried out to know the cooling condition and the mass flow of the liquid helium (LHe) coolant for the SC Magnet system.
This report identifies the different in-house modular signal conditioning electronics and instrumentation systems, calibration at different levels and the outcomes for the SST-1 TF magnet system.
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Invited Talk (Session-5)
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Abstract ID: 4_200
ITER and its Diagnostics- the Way Ahead
Michael Walsh1, ITER Team1
1ITER Organization, France Email: [email protected]
ITER will be a large, technically advanced fusion device that can produce 500MW of power. It is currently under construction in France. It is building on the success of many smaller devices that have already been built and tested around the world. Designed to carry 15MA and with a major radius of 6 m and plasma diameter of approximately 4m, it will push many boundaries both technical and logistical. It will also be the first nuclear installation based on magnetic fusion. Managing the performance of this device will bring many new challenges with a particular need to have precise monitoring and control.
The monitoring will be performed by a suite of diagnostics or measurement systems; there are close to 50 of these systems in total. These diagnostics are crucial for successful operation and will have to handle both routine and advanced operation and also physics exploitation. These diagnostics must perform reliably and robustly in a wide range of operating scenarios. The requirements for the diagnostics have been developed and the flow-down of these requirements dictate the exact systems that are needed. These include systems that work in the fields of magnetics, neutrons, bolometer, optical, microwave and operational systems. The latter including pressure gauges, infrared systems and a range of observation systems for tritium and dust.
All the diagnostics have to handle electromagnetic, seismic and many other loads. The large electromagnetic loads come from disruptions to the plasma. The system has also to maintain its integrity in the event of an earthquake. The harsh environment for these systems also includes neutrons, activation and ultra-high vacuum. More specifically, most of the optical diagnostics also have to handle the issue of first mirror contamination.
Incorporation of all these systems to ITER is a challenging task from both a technical and integration perspective. For example, managing the interfaces is a specifically complex task. It involves interacting with many teams; those working on diagnostics and those working on the rest of the device. Many teams are working remotely in the partner countries around the world. To ensure as smooth as possible integration of the designs, there is need for strong coordination between the different teams.
To date, a significant amount of work has been performed to support the design and development of these systems. Now, the designs are progressing. To date, more than 95% of the systems are in the intermediate and final design stage. Some systems are already past the final design and are in manufacturing. The systems that are most advanced in the design are those that are needed for early installation or are design drivers for many other components. For example, the former includes a plasma current measuring diagnostic that is integrated to the Toroidal field coils while the latter includes items such as the port plug structures. This talk will focus on the approaches and the challenges of the development and implementation of all the systems.
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Abstract ID: 4_292
Status of the Realization of the Neutral Beam Test Facility
Vanni Toigo1
1Consorzio RFX, Italy Email: [email protected]
The ITER Neutral Beam Injectors (NBI) are required to deliver 16.5 MW of additional heating power to the plasma, accelerating negative ions up to -1 MV with a beam current of 40A lasting up to 1 hour. Since these outstanding requirements were never achieved all together so far, the realization of a Neutral Beam Test Facility (NBTF), called PRIMA, currently under construction in Padova, was launched with the aim to test the operation of the NB injector and to study the relevant physical and technological issues, in advance to the implementation in ITER. Two projects are under development: MITICA and SPIDER.
MITICA is a full scale prototype of the ITER NB injector; the design is based on a similar scheme and layout, with the same power supply system and also the control and protection systems are being designed according to the ITER rules and constraints. The HV components are procured by JADA; the low voltage ones and the injector are procured by F4E.
SPIDER project is an ion source with the same characteristics of the ITER one, specifically addressed to study the issues related to the RF operation; for this reason, the beam energy is limited to 100keV. It can generate both Hydrogen and Deuterium Ions; the design includes provisions to filter electrons and also to allow the use of cesium to attain the high values of current density required. SPIDER is procured by F4E and INDA. The construction of PRIMA buildings and auxiliaries, started in autumn 2008, was completed in summer 2015.
SPIDER plant systems procurement is well advanced and some systems are under installation or site acceptance tests. In 2016 integrated commissioning and power supply integrated tests will be performed followed by the beginning of the first experimental phase.
MITICA design was completed; many procurement contracts have been signed or will be launched in the next months. Installation activity will start in December 2015 with the installation of the first HV power supply components provided by JADA.
The paper, after an overview of the main characteristics of SPIDER and MITICA experiments, will present the status of the realization of the NB Test Facility including plant systems and experimental components.
The work leading to this publication has been funded partially by Fusion for Energy under the Contract F4E-RFX-PMS_A-WP-2015. This publication reflects the views only of the author, and Fusion for Energy cannot be held responsible for any use which may be made of the information contained therein. The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.
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Abstract ID: 4_283
R & D of Tritium Technology for Fusion in CAEP: Progress and Prospect
Song Jiangfeng 1, Meng Daqiao1, Li Rong1, Huang Zhiyong1, Huang Guoqiang1, Chen Chang- an1, Deng Xiaojun1, Qin Cheng1, Qian Xiaojing1, Zhang Guikai1
1Institute of Materials, China Academy of Engineering Physics, China Email: [email protected]
China has decided to develop its own fusion engineering test reactor and has also joined ITER. Tritium plant is one of the key systems of fusion system. Programs supposed by China ministry of Science and technology named “Conceptual design and key technologies research on TBM tritium system” and “Conceptual design and key technologies research on tritium plant for fusion reactor” were finished in 2013 and 2014. After several years of research, we have finished the design of TBM tritium system, TEP, SDS, WDS, ISS and tritium safety system. The key technologies such as TES, CPS, hydrogen storage materials for SDS, catalysts for WDS, palladium alloy membranes for TEP are under research. In this paper, the progress and prospect of tritium technology for R&D of fusion is introduced.
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Poster Session-3
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Abstract ID: 1_225
Precision Electronics and Measurement Techniques for the Superconducting Joint Resistance
Yohan Khristi1, Subrata Pradhan1, Kalpesh Doshi1
1Institute for Plasma Research, India Email: [email protected]
The Superconducting Tokamak has different superconducting magnet systems, such as The Toroidal Field (TF), Poloidal field (PF) and Center solenoid (CS) magnet system. Each magnet coil has ~ 1-2 nΩ low DC resistance joints as per the construction criteria and mechanical constraints. The measurement of such a low resistances is critical at the operating condition of 5 K helium temperature and 10 kA DC transport current. The development of electronics and instrumentation are challenging due to the measured signal intensity, large temperature gradient, large DC as well as time varing magnetic field ~ 3 T and tokamak harsh noisy environment. A signal-conditioning electronics with large signal gain of 125 × 103 was developed for the low-DC superconducting joint resistance measurements. The measurement techniques followed to carry out these measurements by taking into account the thermo-electric potentials, lead resistance, non ohmic contacts, device heating etc.
This paper presents the electronics design, measurement precision and different measurement techniques to measure the low-DC superconducting joint resistance.The paper also identifies the role of standard instruments and results of supercounducting joints resistance measurement in the laboratory scale experiment.
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Abstract ID: 1_226
Preliminary Results from Electron Cyclotron Measurements at SST-1
Varsha Siju1, Praveena Kumari1, Surya Kumar Pathak1
1Institute for Plasma Research, India Email: [email protected]
An 8-channel heterodyne radiometer system is developed and installed for the measurements of second harmonic electron cyclotron emission (ECE) at magnetic field of 1.5T at SST-1. This system covers a spectral range of 75.4 to 84.5 GHz at a spatial resolution of less than 1 cm, sensitivity of 9.51 106 V/W. The calculated noise temperature of the system is 1.66eV. The system is calibrated using Hot/cold technique, wherein, a silicon carbide based source at 600 °C acts as the hot body and the room temperature (RTP) acts as the cold body. This paper presents the preliminary observations of the heterodyne radiometer system at SST-1. The measured radiation temperature is around 100eV.
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Abstract ID: 1_232
PLATo (Power Load Analysis Tool) – A Module of WEST Wall Monitoring System
Sutapa Ranjan1, Jean-marcel Travere2, P Moreau2, C Balorin2, J Bucalossi2, V Chaudhari1, Y Corre2, M Firdaouss2, M Jouve2, E Nardon2, R Nouailletas2, N Ravenel2, B Santraine2
1Institute for Plasma Research, India 2CEA-IRFM, France Email: [email protected]
The mandate of the WEST (W Environment for Steady-state Tokamak) [1] project, is to upgrade the medium- sized superconducting Tokamak, Tore Supra in a major scale. One of it's objectives, is to also act as a test- bed for ITER divertor components, to be procured and used in ITER.
WEST would be installing actively cooled Tungsten divertor elements, like the ones to be used in ITER. These components would be tested under two experimental scenarios: high power (Ip=0.8MA, lasting 30s with 15MW injected power) and high fluence (Ip=0.6 MA, lasting 1000s with 12 MW injected power). Heat load on the divertor target will range from a few MW/m2 up to 20 MW/m2 depending on the X point location and the heat flux decay length.
The tungsten Plasma Facing Components (PFCs) are less tolerant to overheating than their Carbon counterparts and prevention of their burnout is a major concern. It is in this context that the Wall Monitoring System (WMS) – a software framework aimed at monitoring the health of the Wall components, was conceived.
WMS has been divided into three parts: a) a pre-discharge power load analysis tool to check compatibility between plasma scenario and PFC's operational limits in terms of heat flux b) a real-time system during discharge, to take into account all necessary measurements involved in the PFCs protection c) a set of analysis tools that would be used post-discharge, that would access WEST database and compare predicted and experimental results.
This paper presents an overview of PLATo – the pre-pulse module of WMS that has been recently developed under IPR–IRFM research collaboration. PLAto has two major components – one that produces heat flux information of the PFCS and the other that produces energy graphs depending on shot profile defined by time variant magnetic equilibrium and injected power profiles. Preliminary results will be presented based on foreseen WEST plasma reference scenarios.
References:
[1] J. Bucalossi et al., Fusion Engineering and Design 89 (2014) 907–912
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Abstract ID: 1_235
Fabrication of Vacuum Vessel with Detachable Top Lid Configuration for Indian Test Facility (INTF)
Jaydeep Joshi1, Ashish Yadav1, Dhananjay Kumar Singh1, Hiteshkumar K Patel1, S Ulahannan2, A Vinaykumar2, M Girish3, M Khan3, Mahohar3, Chandramouli Rotti1, Mainak Bandyopadhyay1, Arun Kumar Chakraborty1
1 ITER-India, Institute for Plasma Research, India 2Airframe Aerodesigns Pvt. Ltd., India 3Vacuum Techniques Pvt. Ltd, India Email: [email protected]
Indian Test Facility Vacuum Vessel (INTF Vessel) with customized configuration has been designed and manufactured as per ASME Sec VIII Div. 1 to house and provide an ultra-high vacuum environment for Diagnostic Neutral Beam (DNB) components for the qualification of beam parameters. DNB is expected to deliver 18–20A hydrogen neutral beam in ITER plasma to measure the helium ash density, produced by the D-T reaction through Charge Exchange Recombination Spectroscopy in ITER.
As per design and operational requirements, INTF vessel is fabricated from AISI 304L materials, in cylindrical shape with the diameter of 4.5m and length of 9m. The unique attribute of this vacuum vessel is, it has a detachable top lid to allow access for internal components during installation and maintenance. Considering the fact that it is the biggest vacuum vessel with this kind of configuration realized ever, as per the best of authors’ knowledge, there were many areas of manufacturing where prior experience is not available. For present manufacturing, top lid is cut from the shell itself which is critical in terms of controlling the deflection which may arise because of relaxation of internal stresses caused by welding and shell rolling. This has been realized by defining the proper cutting sequencing, controlling heat input by very slow cutting, designing additional temporary stiffeners. Further, a systematic approach was essential towards the welding of large flanges and their machining to achieve the flatness in the range of 1.2mm over the area of 9m x 5m for achieving vacuum integrity of the level of 10-9 mbar.l/sec. Though, the welding of flange to collar and flange to top lid has been carried out in very controlled manner, it was felt mandatory to adopt the methodology of stage machining for top flange assembly to nullify the distortion caused by large amount of welding.
This paper presents some of the unique experiences, methodologies and learning gathered from manufacturing of large vacuum vessel with detachable top lid configuration.
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Abstract ID: 1_236
Measurement and Sweep-biasing Circuit for Langmuir Probe Diagnostic in SYMPLE
Pramila Gautam1, Jignesh Patel1, Rachana Rajpal1, Chandresh Hanasalia1, Anitha V P1, Krishnamachari Sathyanarayana1, Ratneshwar Jha1
1Institute for Plasma Research, India Email: [email protected]
A device named SYMPLE is being developed at IPR to study high power microwave - plasma interaction physics. The plasma that enables the proposed investigation needs to satisfy certain criteria in terms of its density ((1-10) 1018/m3), uniform axial (~1 m) and radial (~ 10 cm) extends and a sharp gradient, with scale length of the order of the wavelength of the microwave, in the microwave-plasma interaction regime. In order to identify the right parametric regime where the plasma meets with the required pre-requisites conditions, Langmuir Probe based measurements need to be routinely carried out to measure various plasma parameters such as the electron density (ne), the electron temperature (Te), the floating potential(Vf), and the plasma potential (Vp).
The Langmuir Probe diagnostics electronics along with biasing power supplies is installed in standard industrial racks with an isolation of 2KV provided by the isolation transformer .The electronics system is populated inside the standard 4U- chassis based system. The front end electronics is designed using high common mode differential amplifiers which can measure small differential signal in presence of high common mode dc- bias or ac ramp voltage, which is given to the probes. The front end is populated by programmable gain instrumentation amplifier and programmable filter modules. There is a provision to take optically isolated output signal which can be acquired by data acquisition system. The electronics system is designed for both dc bias (around -70V) and sweep bias mode(10µs rise time in 10KHz ramp) operation. The paper will describe the detailed design of the system with experimental results.
References:
[1] “High-speed dual Langmuir probe”, Review of scientific instruments 81, 073503 _2010.
[2] “A comparison of emissive probe techniques for electric potential measurements in a complex plasma”, Physics of plasmas 18, 073501 (2011).
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Abstract ID: 1_242
Density Measurement Systems at SST Tokamak
Umeshkumar C Nagora1, Surya Kumar Pathak1, Parveen Kumar Atrey1
1Institute for Plasma Research, India Email: [email protected]
Electromagnetic wave experiences a phase difference while passing through the plasma with respect to the reference arm. This phase information gives line averaged electron plasma density. At SST-1 Tokamak, two microwave interferometer systems - (1) 100 GHz homodyne system and (2) 140 GHz phase locked heterodyne system, have been designed, developed and installed. In this paper developed systems performances as well as measurement descriptions are explained. A comparative study has been done to understand the measurement capabilities of the two independent systems and a good agreement is obtained. The measured density of the recent plasma discharges after first wall installation is in the range of 2 - 5 1012/ cm3.
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Abstract ID: 1_245
Software Upgradation of PXI Based Data Acquisition for Aditya Experiments
Vipul K Panchal1, Chhaya Chavda1, Vijay Patel1, Narendra Patel1, Joydeep Ghosh1
1Institute for Plasma Research, India Email: [email protected]
Aditya Data Acquisition and Control System is designed to acquire data from diagnostics like Loop Voltage, Rogowski, Magnetic probes, X-Rays etc and for control of gas feed, gate valve control, trigger pulse generation etc. CAMAC based data acquisition system was updated with PXI based Multifunction modules. The System is interfaced using optical connectivity with PC using PCI based controller module. Data is acquired using LabVIEW graphical user interface (GUI) and stored in server. The present GUI based application doses not have features like module parameters configuration, analysis, webcasting etc. So a new application software using LabVIEW is being developed with features for individual module support considering programmable channel configuration – sampling rate, number of pre & post trigger samples, number of active channel selection etc. It would also have facility of using multi-functionality of timer & counter. The software would be scalable considering more modules, channels and crates along with security of different access level of user privileges.
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Abstract ID: 1_253
Development, Integration and Testing of Automated Triggering Circuit for Hybrid DC Circuit Breaker
Deven Kanabar1, Swati Roy1, Chiragkumar Dodiya1, Subrata Pradhan1
1Institute for Plasma Research, India Email: [email protected]
A novel concept of Hybrid DC circuit breaker having combination of mechanical switch and static switch provides arc-less current commutation into the dump resistor during quench in superconducting magnet operation. The triggering of mechanical and static switches in Hybrid DC breaker can be automatized which can effectively reduce the overall current commutation time of hybrid DC circuit breaker and make the operation independent of opening time of mechanical switch. With this view, a dedicated control circuit (auto-triggering circuit) has been developed which can decide the timing and pulse duration for mechanical switch as well as static switch from the operating parameters. This circuit has been tested with dummy parameters and thereafter integrated with the actual test set up of hybrid DC circuit breaker. This paper deals with the conceptual design of the auto-triggering circuit, its control logic and operation. The test results of Hybrid DC circuit breaker using this circuit have also been discussed.
References:
[1] Swati Roy, Deven Kanabar, Chiragkumar Dodiya, and Subrata Pradhan, “Development of a Prototype Hybrid DC Circuit Breaker for Superconducting Magnets Quench Protection”, IEEE Trans. on Applied Superconductivity, 24, 6 (2014).
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Abstract ID: 1_255
Metrology Measurements for Aditya Tokamak Upgradation
Sharvil Patel1, Kulav Rathod2, Snehal Jayaswal2, Pradeep Chauhan2, Joydeep Ghosh2, Rakesh L Tanna2, Prabal K Chattopadhyay2, Mohan Parmar3, Jinto Thomas2, Madan B Kalal2, Krishnamachari Sathyanarayana2, Mohsin Malek3, Pratik Patel3, Ramkrushna Panchal2, Nilesh Patel2
1Gujarat University, India 2Institute for Plasma Research, India 3Shell-N-Tube Pvt. Ltd., India Email: [email protected]
After 25 years of Aditya tokamak (midsized, air-core, R0= 75 cm, a = 25 cm) operationachieving high temperature circular plasmas in limiter configuration, upgrading it to Aditya-U tokamak with divertor configuration has been planned and the upgradation is under progress. The upgradation process include dismantling of the existing Aditya tokamak to its base level and re- erect it by placing new subsystems like new vacuum vessel of circular cross-section, new buckling cylinder etc. Apposite alignment of subsystems, mainly all the magnetic coil systems in all grades and scales of tokamak is very crucial and essential, as misaligned magnetic coil system scan generate error magnetic fields, which can significantly impact the plasma formation and sustainment in a tokamak.
With this motivation, position and alignment measurement of the existing magnetic coils and structural components of ADITYA tokamak is carried out for the very first time with the optical metrology instrument. Prior to carrying out measurement exercise, machine datum has been transferred to the reference on the wall of tokamak hall using five-point laser and the machine center has been transformed to the four wall of tokamak hall. All position measurements are done with respect to machine major axis in cylindrical geometry. Measurement includes existing radial (R) and elevation (Z) positions of all magnetic coils and various structural components within the accuracy of ± 1 mm. More than 5000 data points are recorded using optical metrology instrument. Again the elevation references are transferred to the primary network established and the angular references are transformed on the floor of the tokamak hall. These results will serve as ready reference for reassembly and alignment of Aditya – Upgrade tokamak. In this paper detailed position measurements of different subsystems of old Aditya tokamak and the relocation of them along with new ones using the optical metrology instruments will be presented.
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Abstract ID: 1_258
Study of Transport and Micro-structural Properties of Magnesium Di- Boride Strand under React and Bend Mode and Bend and React Mode
Ananya Kundu1, Subrat Kumar Das1, Anees Bano1, Subrata Pradhan1
1Institute for Plasma Research, India Email:[email protected]
I-V characterization of commercial multi-filamentary Magnesium Di-Boride (MgB2) wire of diameter 0.83 mm were studied in cryocooler based self –field characterization system under both react and bent mode and bent and react mode for a range of temperature 6 K-25 K. This study is of practical technical relevance where the heat treatment of the superconducting wire makes the sample less flexible for winding in magnet and in other applications. There are limited reported data [1],[2] available on degradation of MgB2 wire with bending induced strain in react and wind and wind and react method. In the present work the bending diameter were varied from 80 mm to 20 mm in the interval of 10 mm change of bending diameter and for each case critical current (Ic) of the strand is measured for the above range of temperature. An ETP copper made customized sample holder for mounting the MgB2 strand was fabricated and is thermally anchored to the cooling stage of the cryocooler. It is seen from the experimental data that in react and bent mode the critical current degrades from 105 A to 87 A corresponding to bending diameter of 80 mm and 20 mm respectively. The corresponding bending strain was analytically estimated and compared with the simulation result. It is also observed that in react and bent mode, the degradation of the transport property of the strand is less as compared to react and bent mode. For bent and react mode in the same sample, the critical current (Ic) was measured to be ~145 A at 15 K for bending diameter of 20 mm. Apart from studying the bending induced strain on MgB2 strand, the tensile test of the strand at RT was carried out. The electrical characterizations of the samples were accompanied by the microstructure analyses of the bent strand to examine the bending induced degradation in the grain structure of the strand. All these experimental findings are expected to be used as input to fabricate prototype MgB2 based magnet.
References:
[1] Q.Wang et al.,“Influence of bending strain on mono- and multi-filamentary MgB2/Nb/Cu wires and tapes,” Physica C, 484, 163 (2013).
[2] K.Katagiri et al., “Stress–strain effects on powder-in-tube MgB2 tapes and wires,” Superconducting Science and technology, 18, 12 (2005).
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Abstract ID: 1_259
Michelson Interferometer Diagnostics for Broadband ECE Measurement
Abhishek Sinha1, Surya Kumar Pathak1
1Institute for Plasma Research, India Email: [email protected]
A Michelson interferometer (MI) diagnostic is capable of measuring broadband intensity spectra in the microwave and near infrared spectral range. The Michelson interferometer diagnostics is dedicated to probe the full electron cyclotron emission (ECE) spectrum emitted by plasmas in tokamak experiments with magnetic confinement. At the SST-1 Tokamak at IPR, Michelson interferometer will be used to measure the spectrum of the electron cyclotron emission in the spectral range 70–500 GHz. The interferometer is absolutely calibrated using the hot/cold technique and, in consequence, the spatial profile of the plasma electron temperature is determined from the measurements. The Michelson interferometer has spectral resolution of 3.66 GHz and temporal resolution of about 16.67 ms. Installation of the Michelson interferometer diagnostics is in process at IPR.
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Abstract ID: 1_265
Assembly of Aditya Upgrade Tokamak
Madan B Kalal1, Rakesh L Tanna1, Joydeep Ghosh1, Shailesh B Bhatt1, Dinesh S Varia1, Sharvil Patel1, Vaibha Ranjan1, Devraj H Sadharkiya1, Ramkrushna Panchal1, Rohit Kumar1, Harshita Raj1, Krishnamachari Sathyanarayana1, Kulav Rathod1, Kumarpalsinh A Jadeja1, Kaushal M Patel1, Kaushik S Acharya1, Prabal K Chattopadhyay1, Ashok V Apte2, Yogesh C Saxena1, Dhiraj Bora1, Shell-N-Tube Team3
1Institute for Plasma Research, India 2Space Application Center, India 3Shell-N-Tube Pvt. Ltd., India Email: [email protected]
The existing Aditya tokamak, a medium sized tokamak with limiter configuration is being upgraded to a tokamak with divertor configuration. At present the existing ADITYA tokamak has been dismantled up to bottom plinth on which the whole assembly of toroidal field (TF) coils and vacuum vessel rested. The major components of ADITYA machine includes 20 TF coils and its structural components, 9 Ohmic coils and its clamps, 4 BV coils and its clamps as well as their busbar connections, vacuum vessel and its supports and buckling cylinder, which are all being dismantled.
The re-assembly of the ADITYA Upgrade tokamak started with installation and positioning of new buckling cylinder and central solenoid (TR1) coil. After that the inner sections of TF coils are placed following which in-situ winding, installation, positioning and support mounting of two pairs of new inner divertor coils have been carried out. After securing the TF coils with top I-beams the new torus shaped vacuum vessel with circular cross-section in 2 halves have been installed. The assembly of TF structural components such as top and bottom guiding wedges, driving wedges, top and bottom compression ring, inner and outer fish plates and top inverted triangle has been carried out in an appropriate sequence. The assembly of outer sections of TF coils along with the proper placements of top auxiliary TR and vertical field coils with proper alignment and positioning with the optical metrology instrumentmainly completes the reassembly. Detailed re-assembly steps and challenges faced during re-assembly will be discussed in this paper.
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Abstract ID: 1_266
The Refurbishment of Damaged Toroidal Magnetic Field coils for Aditya Upgrade
Devraj H Sadharakiya1, Rakesh L Tanna1, Joydeep Ghosh1, Prabal K Chattopadhyay1, Sharvil Patel1, Vaibhav Ranjan1, Rohit Kumar1, Harshita Raj1, Krishnamachari Sathayanarayana1, Madan B Kalal1, Dinesh S Varia1, Ramkrushna Panchal1, Kulav Rathod1, Shailesh B Bhatt1, A Vardharajulu1, Yogesh C Saxena1, Dhiraj Bora1, Shell-N-Tube Team2
1Institute for Plasma Research, India 2Shell-N-Tube Pvt. Ltd., India Email:[email protected]
Aditya tokamak (R0 = 75 cm, a = 25 cm), is a machine in which high temperature plasma is produced and contained using magnetic fields. After 25 years of Aditya tokamak operation, Upgradation of the Aditya tokamak with limiter configuration to Aditya-U tokamak with divertor configuration is under progress. There are 20 numbers of toroidal magnetic field coils which produces 1.5 Tesla of magnetic field at plasma centre, when 50 kA of current is passed through them. Each of the TF coils is a picture frame type coil, with a bore of 0.78 m 0.9 m and outer dimension of 1.03 m 1.26 m, having weight of 500 kg each. A single TF coil is made up of two C’s (Big–C and Small–C) joined together using 16 bolts. Each C is made up of six C shaped copper plates pressed together with pre-impregnated epoxy glass insulation between the copper plates. For joining the two C’s to make one picture frame type TF coil, at the open ends of C, all six plates are tapered to accommodate the respective open ends of small C which has similar tapering. The tapered portion of each C is named as Fingers. Each Finger in both the C’s is silver plated having dimension of 160 mm x 160 mm x 6.5 mm thickness and 6.5 mm apart from each other.
During the dis-assembly of Toroidal magnetic field (TF) coils, it was realised that 6 numbers of TF coils (Coil No. 2, 3, 8, 9, 17 and 20) are damaged at the fingers joints of two C’s sections constituting a TF coil. The copper material have been melted and eroded mainly at the edges of fingers joints of small–C and big–C of TF coils especially in the middle fingers. Large depositions of carbon have been found near melted copper. The Aditya Upgrade team has found in house technique of refurbishing these TF coils. After repairing the damaged TF coils, they are assembled one by one on Test Stand by joining both C’s sections and the electrical parameter testing (Resistance and Inductance) of these coils have been carried out. The resistance and inductance measurements of each damaged coil after repairing showed that electrical parameters are within satisfactory limits and are in good condition to be reused again. This remarkable task has saved lot of cost and time for Aditya Upgrade re-assembly. The details of damaged TF coils refurbishment and its electrical parameters measurements will be discussed in this paper.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_270
Conceptual Design of Dump Resistor for Superconducting CS of SST-1
Swati Roy1, Subrata Pradhan1, Arun Panchal1
1Institute for Plasma Research, India Email: [email protected]
During the upgradation of SST-1, the resistive central solenoid (CS) coil has been planned to be replaced with Nb3Sn based superconducting coil. The superconducting CS will store upto 3.5MJ of magnetic energy per operation cycle with operating current upto 14kA. In case of coil quench, the energy stored in the coils is to be extracted rapidly with a time constant of 1.5s. This will be achieved by inserting a 20mOhm dump resistor in series with the superconducting CS which is normally shorted by circuit breakers. As a vital part of the superconducting CS quench protection system, a conceptual design of the 20mOhm dump resistor has been proposed. In this paper, the required design aspects and a dimensional layout of the dump resistor for the new superconducting CS has been presented. Natural air circulation is proposed as cooling method for this dump resistor. The basic structure of the proposed dump resistor comprises of stainless steel grids connected in series in the shape of meander to minimize the stray inductance and increase the surface area for cooling. The entire dump resistor will be an array of such grids connected in series and parallel to meet electrical as well as thermal parameters. The maximum temperature of the proposed dump resistor is upto 350 degC during dump 3.5MJ energy. The proposed design permits indigenous fabrication of the dump resistor using commercially available welding techniques.
References:
[1] S. Kedia, S. Roy, S. Pradhan, “Finite-Element Analysis of Dump Resistor for Prototype Superconducting Magnet Carrying 3.60 MA-t”, IEEE Transactions on Applied Superconductivity, 20, 6, (2010).
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Abstract ID: 1_272
Safety and Environment Aspects of Tokamak-type Fusion Power Reactor - An Overview
Bharatkumar Doshi1, D Chenna Reddy1
1Institute for Plasma Research, India Email: [email protected]
Naturally occurring thermonuclear fusion reaction (of light atoms to form a heavier nucleus) in the sun and every star in the universe, releases incredible amounts of energy. Demonstrating the controlled and sustained reaction of deuterium-tritium plasma should enable the development of fusion as an energy source here on Earth. The promising fusion power reactors could be operated on the deuterium-tritium fuel cycle with fuel self-sufficiency. The potential impact of fusion power on the environment and the possible risks associated with operating large-scale fusion power plants is being studied by different countries. The results show that fusion can be a very safe and sustainable energy source. A fusion power plant possesses not only intrinsic advantages with respect to safety compared to other sources of energy, but also a negligible long term impact on the environment provided certain precautions are taken in its design. One of the important considerations is in the selection of low activation structural materials for reactor vessel. Selection of the materials for first wall and breeding blanket components is also important from safety issues. It is possible to fully benefit from the advantages of fusion energy if safety and environmental concerns are taken into account when considering the conceptual studies of a reactor design. The significant safety hazards are due to the tritium inventory and energetic neutron fluence induced activity in the reactor vessel, first wall components, blanket system etc. The potential of release of radioactivity under operational and accident conditions needs attention while designing the fusion reactor. This paper describes an overview of safety and environmental merits of fusion power reactor, issues and design considerations and need for R&D on safety and environmental aspects of Tokamak type fusion reactor.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_274
Fusion Blanket Materials Development and Recent R&D Activities
Chandra Sekhar Sasmal1, Shiju Sam1, Atikkumar N Mistry1, Atul Prajapati1, Hardikkumar M Tailor1, Jignesh P Chauhan1, Kinkar Laha2, Arun Kumar Bhaduri2, Rajendra Kumar E1
1Institute for Plasma Research, India 2Indira Gandhi Center for Atomic Research (IGCAR), India Email: [email protected]
Development of structural materials plays an important role in the feasibility of fusion power plant. The candidate structural materials for future fusion reactors are Reduced Activation Ferritic Martensitic (RAFM) steel, nano structured ODS Steel, vanadium alloys and SiC/SiCf composite etc. RAFM steel is presently considered as the structural material for Lead Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) because of its high void swelling resistance and improved thermal properties compared to austenitic steel. Development of RAFM steel in India is being carried out in full swing in collaboration with various research laboratories and steel industries.
India’s participation in the ITER TBM testing program necessitates the development of India Specific RAFM (IN-RAFM) steel for LLCB TBM. Comprehensive research studies are being carried out for the development of IN-RAFM steel with optimized W and Ta content. Based on extensive testing and evaluations on various composition of RAFM steel having tungsten content in the range 1-2 wt.% and tantalum content in the range 0.06 – 0.014%, it was found that 9Cr-1.4W-0.06Ta RAFM steel possesses better combination of strength and toughness and hence was the chosen composition for IN-RAFM steel. Commercial heats of IN-RAFM steel have been produced and characterization of these heats is under process. Various mechanical properties like tensile strength and impact energy of this commercial grade heat are found similar to the laboratory scale of IN-RAFM steel. Heat treatment has been carried out at 770 C for 2 hours followed by furnace cooling to co-relate the effect of tempering temperature (close to PWHT temperature) on the mechanical properties of base metal.
This paper presents an overview of the Indian activities on fusion blanket materials and describes in brief the efforts made to develop IN-RAFM steel as structural material for the LLCB TBM.
In Future, due to enhanced properties of vanadium base alloy and nano structured materials like ODS RAFMS, RAFM steel may be replaced by these materials for its application in DEMO relevant fusion reactor. Future R&D activities will be specifically towards the development of these structural materials for fusion reactor.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_278
Electrical Properties of Nano Li2TiO3 for Fusion Reactors
S K S Parashar1, Kajal Parashar1, Paritosh Chaudhuri2
1KIIT University, Bhubaneswar, India 2Institute for Plasma Research, India Email: [email protected]
In the development of tritium breeding blankets for fusion reactor, lithium based ceramic such as lithium orthosilicate (Li4SiO4), lithium titanate (Li2TiO3), lithium zirconate (Li2ZrO3), and lithium oxide (Li2O) for breeding blankets. Among them Lithium titanate (LT) is one of the most promising tritium breeding materials due to their reasonable lithium atom density, low activation, good compatibility with structural materials, excellent tritium release performance and chemical stability. Electrical properties may reflect some characteristic features, hence analysis of electrical charge transport in small grained Li2TiO3 ceramics, as envisaged for tritium breeding, may contribute to gain information of certain high energy ball-milling process. The main attribute of current study analyzes the electrical conductivity behavior of Li2TiO3 ceramics. 0 The 10h milled 3 powder of Li2TiO3 by High energy ball milling (HEBM) calcined at 700 C for 2h. The calcinied powder was pressed uniaxially with 3wt. % PVA (polyvinyl alcohol) solution added as binder. The rectangular disk samples of diameter 12.7mm and thickness 12mm was made by by hydraulic press with 400Mpa pressure. The samples were sintered at 700 0C, 800 0C, 900 0C and 1000 0C 2h in conventional sintering. Silver contacts were made on the opposite disc faces and heated at 700 0C for 15 minutes with a heating 50C per minute for electrical measurement. It was found that microwave sintered samples shows higher thermal conductivity then conventional sintered one.
The Ea value decreases with increase in frequency, due to the increase in ionic conductivity. The ionic conductivity is a combination of both macroscopic and microscopic conduction, which is indirectly depend on the bulk Rb and grain boundary Rgb resistance. At high temperature only single semicircle could be found, using high frequency data, indicate dominant behavior of grain. The value of activation energy (0.238eV) and conductivity range (10-3 to 10-4 S/cm) says that material is a semiconductor.
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Abstract ID: 1_279
Design of New Superconducting Central Solenoid of SST-1 Tokamak
Upendra Prasad1, Subrata Pradhan1, Mahesh Ghate1, Piyush Raj1, Vipul L Tanna1, Ziauddin Khan1, Swati Roy1, Prosenjit Santra1, Prabal Biswas1, Aashoo N Sharma1, Yohan Khirsti1, Pankaj Varmora1
1Institute for Plasma Research, India Email:[email protected]
The key role of the central solenoid (CS) magnet of a Tokamak is for gas breakdown, ramp up and maintaining of plasma current for longer duration. The magnetic flux change in CS along with other PF coils generates magnetic null and induces electric field in toroidal direction. The induced toroidal electric field accelerates the residual electrons which collide with the neutrals and an avalanche takes place which led to the net plasma in the vacuum vessel of a Tokamak. In order to maximize the CS volt-sec capability, the higher magnetic field with a greater magnetic flux linkage is necessary. In order to facilitate all these requirements of SST-1 a new superconducting CS has been designed for SST-1. The design of new central solenoid has two bases; first one is physics and second is smart engineering in limited bore diameter of ~655 mm. The physics basis of the design includes volt-sec storage capacity of ~0.8 volt-sec, magnetic field null around 0.2 m over major radius of 1.1 m and toroidal electric field of ~0.3 volt/m.The engineering design of new CS consists of Nb3Sn cable in conduit conductor (CICC) of operating current of 14 kA @ 4.5 K at 6 T, consolidated winding pack, smart quench detection system, protection system, housing cryostat and conductor terminations and joint design. The winding pack consists of 576 numbers of turns distributed in four layers with 0.75 mm FRP tape soaked with cyanide Easter based epoxy resin turn insulation and 3 mm of ground insulation. The inter- layer low resistance (~1 nΩ) at 14 kA @ 4.5 K terminal praying hand joints has been designed for making winding pack continuous. The total height of winding pack is 2500 mm. The stored energy of this winding pack is ~3 MJ at 14 kA of operating current. The expected heat load at cryogenic temperature is ~10 W per layer, which requires helium mass flow rate of 1.4 g/ s at 1.4 bars @ 4.5 K. The typical diameter and height of housing cryostat are 650 mm and 2563 mm with 80 K shield respectively. The protection system consists of SS310 made array of dump resistor of 20 mΩ. The detail physics and engineering design of new superconducting CS of SST-1 will be discussed in this presentation.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_281
Design of High Resolution Spectroscopic Diagnostics for SST-1 and ADITYA-U Tokamak
Gaurav Shukla1, Kajal Shah1, Malay Bikas Chowdhuri1, Ranjana Manchanda1, Santanu Banerjee1, Nilam Ramaiya1, Joydeep Ghosh1
1Institute for Plasma Research, India Email: [email protected]
High Resolution spectroscopic diagnostics are proposed for SST-1 and ADITYA-U Tokamak for the measurement of plasma rotation and ion temperature using line radiations emitted by impurity ions. A high resolution Charge eXchange Recombination Spectroscopy (CXRS) using line emission from C VI (n=87) at 529 nm is proposed for SST-1 Tokamak. SST-1 Tokamak is equipped with a heating neutral beam of 40keV energy with a beam power of 1.2MW for the measurement of impurity rotation and temperature [1-4]. The CXRS diagnostic for SST-1 will have a high spatial resolution of ~ 1cm and a high time resolution of ~5ms.
Imaging X-ray crystal spectroscopy diagnostic (XCS) [5-6] is proposed for ADITYA-U Tokamak [7] to provide spatially and temporally resolved measurement of plasma rotation and impurity ion behavior. The spectrometer will consist of a spherically bent crystal and CCD detector to measure Ne IX line emission at 13.4474 Å (w) in the toroidal plane of the vacuum vessel with spatial resolution of ~ 2.8 cm. The diagnostic will provide multiple line of sight measurement to estimate toroidal rotation velocity profile and understand impurity transport for ADITYA-U plasma.
Feasibility study for the design of the CXRS diagnostic including a detailed calculation of the photon budget and Etendue budget is presented in this article. Moreover, details of the XCS diagnostic design and system integration with ADITYA-U tokamak are also presented.
References:
[1] R.C. Isler, Plasma Phys. Control. Fusion 36 (1994) 171.
[2] K.H. Burrell, P. Gohil, R. Groebner, D. Kaplan, J. Robinson, W. Solomon, et al.,Rev. Sci. Instrum. 75 (2004) 3455.
[3] R.P.Seraydarian, K. Burrell, N. Brooks, R. Groebner, C. Kahn, Rev. Sci. Instrum. 57 (1986) 155.
[4] R.J. Fonck, D. Darrow, K. Jaehnig, Phys. Rev. A 29 (6) (1984) 3288.
[5] M. Bitter, K. W. Hill, B. Stratton, et al., Rev. Sci. Instrum. 75, 3660 (2004).
[6] A. Ince-Cushman, J. E. Rice, M. Bitter, et al., Rev. Sci. Instrum. 79, 10E302 (2008).
[7] Bhatt S.B. et al 1989 Indian J. Pure Appl. Phys. 27 710.
161 | P a g e
10th Asia Plasma & Fusion Association Conference
Abstract ID: 1_295
Conceptual & Engineering Design of Plug-in Cryostat Cylinder for Superconducting Central Solenoid of SST-1
Prabal Biswas1, Prosenjit Santra1, Kirit R Vasava1, Snehal Jayswal1, Tejas Parekh1, Pradeep Chauhan1, Hiteshkumar Patel1, Subrata Pradhan1
1Institute for Plasma Research, India Email: [email protected]
SST-1, country’s first indigenously built steady state super-conducting tokamak is planned to be equipped with a Nb3Sn based super-conducting 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 SS304L 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.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 2_4
Investigation of Homoclinic Bifurcation of Plasma Fireball in a Double Plasma Device
Arun Sarma1, Vramori Mitrra1, Bornali Sarma1
1VIT University Chennai Campus, India Email: [email protected]
Plasma fire balls are generated due to localized discharge and it is a sharp boundary of the glow region, which suggests a localized electric field such as an electrical sheath or double layer structure. In this paper, homoclinic bifurcation phenomena in the plasma fireball dynamics which is produced in the target chamber of double plasma device have been explored. Homoclinic bifurcation is noticed in the plasma fireball as the system evolving from large time period oscillation to small time period oscillation. The control parameters of this observations are density ratio of target to source chamber (nT/nS), applied electrode voltage to produce fireball, grid bias voltage etc. The dynamical transition of plasma fire balls have been investigated by recurrence quantification analysis (RQA) and by different statistical measures. The gradual increment of kurtosis and decrement of skewness with the change of nT has been observed which are strongly indicative of homoclinic bifurcation in the system. The visual changes of recurrence plot and the gradual changes in recurrence quantifiers reflect the bifurcation with the variation in the control parameter of the double plasma device. The combination of RQA and statistical measures like 1/f power spectrum, clearly conjectured the homoclinic bifurcation due to plasma fire ball in the experimental conditions.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 2_14
Determination of the Plasma Composition using Blended Stark- Broadened Emission Lines in a Self-Magnetic Pinch Diode
Subir Biswas1, R Doron1, V Bernshtam1, Y Maron1, M D Johnston2, M L Kiefer2, M E Cuneo2
1Weizmann Institute of Science, Israel 2Sandia National Laboratories, USA Email: [email protected]
We analyzed visible spectra obtained in self-focusing, relativistic-electron diode experiments performed on the RITS-6 [1] accelerator facility at Sandia National Laboratories (SNL). An electron beam emitted from the cathode strikes a planar anode surface with high current densities (~1 MA/cm2), forming a plasma in the anode-cathode (A-K) gap. Radiation emitted from the plasma is imaged onto a spectrometer input slit via an optical fiber bundle. The spectrometer output is coupled to a gated, intensified charge-coupled device (ICCD) camera, yielding spatially resolved (2mm) spectra. The spectra from the high-density plasma region mainly exhibit emission that appears to be from a continuum source. However, the radiation intensity distribution cannot be explained by free-free or free-bound emissions. Rather, we suggest that the spectrum originates from the blending of many Stark-dominated spectral lines. Accordingly, the spectral intensity distribution provides information on the plasma composition and thermodynamic parameters.
References:
[1] K. D. Hahn, N. Bruner, M. D. Johnston et. al., “Overview of Self-Magnetically Pinched-Diode Investigations on RITS-6,” IEEE Trans. Plasma Sci., 38, 2652 (2010).
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 2_16
Magnetic Probe Diagnostic Tool to Understand the Dynamics in a Non- transferred dc Plasma Torch
Vidhi Goyal1, Ravi Ganesh1
1Institute for Plasma Research, India Email: [email protected]
In the present work, magnetic field driven dynamics and influence of J × B forces on the plasma column inside a dc non-transferred plasma torch are experimentally investigated using magnetic probe as a diagnostic tool. The magnetic diagnostic is a powerful tool which has been used in a number of, plasma experiments including those on waves and tokamaks. However, use of magnetic diagnostic in plasma torches is unheard of, except in one earlier work [1]. In the present work, arrays of magnetic probes are incorporated inside the plasma torch cooling channel and elsewhere; experiments are carried out for a wide range of gas flow rates (20 to 100 lpm) in the presence of external magnetic field (100 to 500 G) on a non-transferred dc plasma torch at atmospheric pressure with nitrogen as working gas. The diagnostic is used primarily to estimate the arc root rotational velocity [2]. Results can also be interpreted to figure out whether the arc root is constricted, has multiple attachments or diffused attachment [3]. More experiments are underway and it is speculated that that this diagnostic can also be used to reveal more useful information [4] on the nature of the arc root attachment and of the entire plasma channel.
References:
[1] Magno Pinto Collares, Characteristics of DC plasma torches and the use of magnetic probes for diagnostics , Ph.D. thesis ,university of Minnesota , (1996)
[2] R N Szente, R J Munz and M G Drouet,Arc velocity and cathode erosion rate in a magnetically driven arc burning in nitrogen, J.phys.D:Appl.Phys.21, 909-913 (1988)
[3] S Ghorui, S N Sahasrabudhe and A K Das, Current transfer in dc non-transferred arc plasma torches, J.phys.D:Appl.Phys.43, 245201 (2010)
[4] Boulos, M. I., Fauchais, P., and Pfender E., Thermal Plasmas: Fundamentals and Applications,Plenum Press, New York, (1994)
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Abstract ID: 2_22
Localized solutions in Laser Plasma Coupled System with Periodic Time Dependence
Deepa Verma1, Amita Das1, Bhavesh Patel1, Predhiman Krishan Kaw1
1Institute for Plasma Research, India Email: [email protected]
There are well known varieties of exact nonlinear localized solutions for the laser plasma system [1] which have been studied extensively. In these solutions the ponderomotive pressure of light wave expels and evacuates the electrons from the center creating a cavity of electron density. The electrons are pulled up by the electrostatic force of the ions which are left behind at in the central region. The balance of ponderomotive and the electrostatic forces leads to a configuration wherein the electrons are piled up at the edge region of the solutions. The higher electron density at the edge in turn confines the radiation and prevents its leaking out. Both stationary as well as moving structures with constant group velocities have been obtained and studied in detail in some of our previous work [2]. Here we report a new variety of solutions showing periodic time dependence. These solutions have been shown to exist in both fluid and Particle – in – Cell simulations. A physical understanding of such solutions will also be provided.
References:
[1] Vikrant Saxena, Amita Das, Sudip Sengupta, Predhiman Kaw, and Abhijeet Sen, “Stability of 1D Laser Pulse Solitons in a Plasma,” Physics of Plasmas, 14, 072307(2007 ).
[2] Sita Sundar, Amita Das, Vikrant Saxena, Predhiman Kaw, and Abhijeet Sen, “Relativistic electromagnetic flat top solitons and their stability,” Physics of Plasmas, 18, 112112(2011).
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Abstract ID: 2_23
Coupling of Drift Wave with Dust Acoustic Wave
Atul Kumar1, Amita Das1, Predhiman Krishan Kaw1
1Institute for Plasma Research, India Email: [email protected]
Drift wave occurs universally in magnetized plasmas producing the dominant mechanism for the transport of particles, energy and momentum across the magnetic field lines. It is a local wave, which propagates in the direction of diamagnetic drift velocity [1] in an inhomogeneous region of plasma. In a magnetized plasma, there can be many collective modes but the lowest frequency modes i.e. << ci (strong magnetic field approximation) dominate the transport. The coupling of these low frequency modes with the dust acoustic waves have been studied for both weakly and strongly coupled [2] dusty plasma. We find a typical acoustic wave for large k in the perpendicular direction in weakly coupled regime. We also observe the coupling of shear wave with drift wave in strongly coupled regime. Instabilities have also been observed in the strongly coupled regime, which depends on the density gradient scale length, viscoelastic effects, depletion of electrons from the plasma etc.
References:
[1] Stationary spectrum of strong turbulence in magnetized non-uniform plasma, Akira Hasegawa and Kunioki Mima, Phys. Rev. Lett. 39, 205, 1977.
[2] Low frequency modes in strongly coupled dusty plasma, Kaw, P. K. and Sen, A., Physics of Plasmas (1994-present), 5, 3552-3559 (1998).
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 2_30
Resolving Issues Associated with Langmuir Probe Measurements in High Pressure Complex (Dusty) Plasmas
Manjit Kaur1, Sayak Bose1, Prabal K Chattopadhyay1, Joydeep Ghosh1, Yogesh C Saxena1
1Institute for Plasma Research, India Email: [email protected]
The Langmuir probe measurements in high pressure dusty plasmas are not straightforward. There exist two major issues which needs attention during Langmuir probe measurements in high pressure dusty plasmas. First is the deposition of dust particles on the probe head. Being negatively charged the dust particles get attracted towards it when the probe bias rises above floating potential. The dust deposition alters the probe I–V characteristics significantly leading to gross errors in estimating the plasma parameters. Secondly, when used in high pressure ( ) plasmas, the elastic scattering of ions due to their collisions with neutrals reduces the ion collection current and substantially decreases the signal to noise ratio. The effect of collision on ion current has to be taken into account to interpret the probe data correctly. The details will be discussed in the presentation. After taking into consideration the above mentioned complications, a specially designed Langmuir probe system is described that is immune to dust contamination and is capable of working in high pressure plasmas giving correct estimates of plasma parameters. The biasing circuit of the probe has been suitably designed to minimize the effects of capacitive current and noise on the probe characteristics using tri-axial cable having a driven shield. Experimental results and proper analysis of Langmuir probe measurements from Complex Plasma Experimental Device (CPED) with levitated dust particles has been presented.
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10th Asia Plasma & Fusion Association Conference
Abstract ID: 2_32
On the Spatial Behavior of Background Plasma in Different Background Pressure in CPS Device
Subrata Samantaray1, Rita Paikaray2, Gourishankar Sahoo2, Parthasarathi Das2, Joydeep Ghosh3, Amulya Kumar Sanyasi3
1Christ College, India 2Ravenshaw University, India 3Institute for Plasma Research, India Email: [email protected]
Blob formation and transport is a major concern for investigators as it greatly reduces the efficiency of the devices [1-4]. Initial results from CPS device [5, 6] confirm the role of fast neutrals [7, 8] inside the bulk plasma in the process of blob formation and transport. 2-D simulation of curvature and velocity shear instability in plasma structures suggest that in the presence of background plasma, secondary instability do not grow non-linearly to a high level and stabilizes the flow [9]. Adiabaticity effect also creates a radial barrier for interchange modes [10]. In the absence of background plasma the blob fragments even at the modest level of viscosity [9]. The fast neutrals outside bulk plasma supposed to stabilize the system. The background plasma set up is aimed at creating fast neutrals outside main plasma column, hence; the background plasma set up is done in CPS device. The spatial behavior of plasma column in between electrodes is different for different base pressure in CPS device. The spatial variation of electron temperature of plasma column between electrodes is presented in this communication. Electron temperature is measured from emission spectroscopy data. The maximum electron temperature (line averaged) is ~ 1.5 eV.
References:
[1] M. Endler, Turbulent SOL transport in stellarators and tokamaks, J. Nucl. Mater 266-269, 84 (1999)
[2] V. Naulin, Turbulent transport and the plasma edge, J. Nucl. Mater, 363–365, 24 (2007)
[3] O. E. Garcia, Blob transport in plasma edge: a review, Plasma and Fusion Research: Review Articles, 4(019), 1-7 (2009) [4] D. A. D’Ippolito, J. R. Myra. and S. J. Zweben, Convective transport by intermittent blob- filaments: comparison of theory and experiment, Phys. Plasmas, 18, 060501: 1 (2011)
[5] G. Sahoo, R. Paikaray, S. Samantaray, P. Das, J. Ghosh, A. Sanyasi, M. B. Chowdhuri., Base pressure plays an important role for production of plasma blob in argon plasma., Journal of Physical Science and Application, 4 (6), 348 (2014)
[6] G. Sahoo, R. Paikaray, S. Samantaray, P. Das, J. Ghosh, A. Sanyasi, On the role of fast neutrals in the process of blob formation in low temperature plasmas, Kathmadu University Journal of Science, Engineering and Technology 10 (II), 50 (2014) [7] A. Niemczewski, I. H. Hutchinson, B. LaBombard, B. Lipschultz, G. M. McCracken, Neutral particle dynamics in the Alcator C-Mod tokamak, Nucl. Fusion 37(2), 21997: 151 (1997)
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[8] S. I. Krasheninnikov, A. I. Smolyakov, On neutral wind and blob motion in linear devices Phys. Plasmas 10(7), 3020 (2003)
[9] D. A. D’Ippolito, J. R. Myra, S I Krasheninnikov, G. Q. Yu, A. Yu. Pigarov, Blob transport in the TOKAMAK Scrape-off-layer, Contrib. Plasma Phys, 44(1-3), 205 (2014)
[10] D. A. D’Ippolito, J. R. Myra, D. A. Russell, Turbulent transport regimes and scrape-off- layer heat flux width LRC-15-160, Submitted to Phys Plasma in march 2015.
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Abstract ID: 2_33
Effect of Catalyst for the Decomposition of VOCs in a NTP Reactor
Suchitra Mohanty1, Smrutiprava Das1, Rita Paikaray1, Gourishankar Sahoo1, Subrata Samantaray2
1Ravenshaw University, India 2Christ College, India Email: [email protected]
Air pollution has become a major cause of human distress both directly and indirectly [1]. VOCs are becoming the major air pollutants. So the decomposition of VOCs is present need of our society. Non-thermal plasma reactor (NTP) is proven to be effective for low concentration VOCs decomposition. For safe and effective application of DBD, optimization of treatment process requires different plasma parameter characterization. So electron temperature and electron density parameters of VOCs show the decomposition path ways. In this piece of work by taking the emission spectra and comparing the line intensity ratios, the electron temperature and density were determined. Also the decomposition rate in terms of the deposited products on the dielectric surface was studied. Decomposition rate increases in presence of catalyst as compared to the pure compound in presence of a carrier gas. Decomposition process was studied by UV-VIS, FTIR, OES Spectroscopic methods & by GCMS [2-5]. Deposited products are analyzed by UV- VIS and FTIR spectroscopy. Plasma parameters like electron temperature, density are studied with OES. And gaseous products are studied by GCMS showing the peaks for the byproducts.
References:
[1] G.Xiao, W. Xu, R. Wu, M.Ni, C.Du, X. Gao, Z. Luo, K. Cen, “Non-Thermal Plasmas for VOCs Abatement”, Plasma Chem Plasma Process, 34, 1033-1065(2014).
[2] S. Mohanty, S. P. Das, “Analysis of Deposited Byproducts of Volatile Organic Compounds (VOCs) Like Toluene, Xylene Subjected to Di-Electric Barrier Discharge (DBD)” International Journal of Science & Research, 3, 1360 (2014).
[3] S. Mohanty, S. P. Das, G. Sahoo, R. Paikaray, P. S. Das, S. Samantaray, D. S. Patil, “effect on plasma parameters in a dielectric barrier discharge reactor with volatile organic compounds”, KUSET, , 10, 24 (2014).
[4] S. Mohanty, S.P. Das, R. Paikray, A.K Patnaik, “Plasma Assisted Destruction of Volatile Pollutants using Dielectric Barrier Discharge”, IJACSA, 1, 1(2013).
[5] T. N. Das, G. R. Dey, “Methane from benzene in argon dielectric barrier discharge”, J Haz. Mat., 248-249, 469 (2013).
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Abstract ID: 2_38
Relativistic Cylindrical and Spherical Plasma Waves
Arghya Mukherjee1, Sudip Sengupta1
1Institute for Plasma Research, India Email: [email protected]
Breaking of relativistically intense nonlinear space charge oscillations is studied analytically and numerically using Sheet Model proposed by Dawson in cylindrical and spherical geometries [1- 3]. It is found that fundamental modes that exist in cylindrical and spherical symmetric system break via the process of phase mixing due to additional anharmonicity induced by geometrical and relativistic effects. A general expression of phase mixing time is given and it is shown that for all cases under consideration phase mixing time scales as the inverse of the cube of the amplitude of applied perturbation. Finally this analytical dependence is also verified by numerical simulations based on Dawson Sheet Model [4].
References:
[1] J. M. Dawson, Phys. Rev. Lett, 62, 383, (1959)
[2] L. M. Gorbunov, A. A. Frolov, E. V. Chizkonov and N. E. Andreev, Plasma Phys. Rep, 36, 345, (2010)
[3] S. V. Bulanov, A. Maksimchuk, C. B. Schroeder, A. G. Zhidkov, E. Esarey, Phys. Plasmas., 19, 020702 (2012)
[4] Arghya Mukherjee and Sudip Sengupta, Phys. Plasmas., 21, 112104, (2014)
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Abstract ID: 2_41
Observation of Early and Strong Relativistic Self-Focusing of cosh- Gaussian Laser Beam in Cold Quantum Plasma
Vikas Nanda1, Niti Kant1
1Lovely Professional University, Punjab, India Email: [email protected]
Relativistic self-focusing of cosh-Gaussian laser beam in the cold quantum plasma has been investigated theoretically using Wentzel-Kramers-Brillouin (WKB) and paraxial ray approximation. The comparative study between self-focusing of cosh-Gaussian laser beam in cold quantum case and classical relativistic case has been made for decentered parameter b 0.9 and it is observed that as the beam propagates deeper inside the cold quantum plasma, the self- focusing ability of the laser beam enhances and shifted towards lower value of normalized propagation distance due to quantum contribution. The variation of beam width parameter with normalized propagation distance for various values of relative density parameter of the medium and intensity parameter has also been studied. It is observed that with the increase in the value of relative density parameter, self-focusing of laser beam becomes stronger. Observation of early and strong self-focusing for higher values of relative density parameter and intensity parameter are reported. Also, with the increase in the value of the relative density parameter of the medium and intensity parameter, self-focusing ability shifted towards lower value of normalized propagation distance due to relativistic effect. The present study might be very useful in the applications like the generation of inertial fusion energy driven by lasers, laser driven accelerators, scribing type of applications in electronics etc.
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Abstract ID: 2_44
Electric Field Assisted Sintering (EFAST): Plasma?
Prasad Mattipally1, Devarasetty Suresh Babu1
1Osmania University, India Email: [email protected]
Electric Field Assisted Sintering (EFAST) discovered in third decade of nineteenth century. This field assisted sintering (FAST) plays a key role in compacting Nano-composites. This sintering technique is also named as Plasma Assisted Sintering (PAS). Commercially this technique is named as Spark Plasma Sintering (SPS). Plasma discharge plays a major role in consolidation of composites. The electrical discharge between powder particles results in confined to a small area and short-lived heating of the particles surfaces up to more than a few thousand degrees Celsius. Since the micro-plasma discharges form uniformly all through the sample volume, the generated heat is also consistently distributed. The particles surfaces are purified and activated due to the elevated temperature causing vaporization of the impurities strenuous on the particle surface. The purified surface layers of the particles melt and mingle to each other forming “necks” between the particles.
Several modeling and experimental studies carried out by researchers, but nobody confirms absolutely the existence of plasma in this technique till today. Present revise focusing on survival of PLASMA in Plasma Assisted Sintering (PAS) and its necessity.
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Abstract ID: 2_45
Dispersion of Linearly Polarized Electromagnetic Wave in Magnetized Quantum Plasma
Abhisek Kumar Singh1, Punit kumar1
1University of Lucknow, India Email: [email protected]
The generation of harmonic radiation is significant in terms of laser-plasma interaction and has brought interesting notice due to the diversity of its applications. The odd harmonics of laser frequency are generated in the majority of laser interactions with homogenous plasma [1, 2]. It has been remarked that second harmonic generation takes place in the presence of density gradient [3, 4] which gives rise to perturbation in the electron density at the laser frequency. The density perturbation coupled with the quiver motion of the electrons produces a source current at the second harmonic frequency. Second harmonic generation has also been related with filamentation [5, 6]. In the present paper, A study of second harmonic generation by propagation of a linearly polarized electromagnetic wave through homogeneous high density quantum plasma in the presence of transverse magnetic field. The nonlinear current density and dispersion relations for the fundamental and second harmonic frequencies have been obtained using the recently developed quantum hydrodynamic (QHD) model. The effect of quantum Bohm potential, Fermi pressure and the electron spin have been taken into account. The second harmonic is found to be less dispersed than the first.
References:
[1] W. B. Mori, C. D. Decker, and W. P. Leemans, “Relativistic harmonic content of nonlinear electromagnetic waves in underdense plasmas,” IEEE Trans. Plasma Sci., 21, 110 (1993).
[2] G. Zeng, B. Shen, W. Yu, and Z. Xu, “Relativistic harmonic generation excited in the ultrashort laser pulse regime,” Phys. Plasmas, 3, 4220 (1996).
[3] E. Esarey, A. Ting, P. Sprangle, D. Umstadter, and X. Liu, “Nonlinear analysis of relativistic harmonic generation by intense lasers in plasmas”IEEE TransPlasma Sci. 21, 95(1993).
[4] V. Malka, J. Modena, Z. Nazmudin, A. E. Danger, C. E. Clayton, K. AMarsh, C. Joshi, C. Danson, D. Neely, and F. N. Walsh, “Second harmonic generation and its interaction with relativistic plasma waves driven by forward Raman instability in underdense plasmas,” Phys. Plasmas 4, 1127 (1997).
[6] J. A. Stamper, R. H. Lehmberg, A. Schmitt, M. J. Herbst, F. C. Young, J.H. Gardener, and S. P. Obenschain, Phys. Fluids 28, 2563 (1985).
[6] J. Meyer and Y. Zhu, Phys. Fluids 30, 890 (1987).
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Abstract ID: 2_46
Breaking of Relativistic Electron Beam Driven Wake Waves in a Cold Plasma
Ratan Kumar Bera1, Arghya Mukherjee1, Sudip Sengupta1, Amita Das1
1Institute for Plasma Research, India Email: [email protected]
Excitation of relativistic electron beam driven wakefield in a cold, over-dense plasma, is studied using 1D-numerical fluid simulation techniques. For the beam density less or equal to the half of the plasma density, simulation results are found to be in good agreement with the analytical work by Rosenzweig et al. [1] for several plasma periods. For the beam density larger than the half of the plasma density, analytical calculations are presented and compared with simulation results here. At later times, the wakefield profile shows an irregular behavior and finally breaks via the gradual process of phase mixing. The excited wakefield profile follows longitudinal Akhiezer- Polovin (AP) mode [2] exactly. The breaking of wake wave is understood in terms of AP wave breaking phenomena and results are compared with the existing theoretical calculations.
References:
[1] J. B Rosenzweig, “Nonlinear Plasma Dynamics in the Plasma Wake-Field Accelerator,” Physical Review Letters, 58, 555 (1987).
[2] A. I. Akhiezer and R. V. Polovin, “Theory of Wave Motion of an Electron Plasma,” Sov. Phys. JETP, 3, 5 (1956).
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Abstract ID: 2_47
2D Turbulence Structure Observed by a Fast Framing Camera System in Linear Magnetized Device PANTA
Satoshi Ohdachi1, S Inagaki1, T Kobayashi2, 3, M Goto2, 3
1Kyushu University, Japan 2National Institute for Fusion Science, Japan 3SOKENDAI (The Graduate University for Advanced Studies), Japan Email: [email protected]
Mesoscale structure, such as the zonal flow and the streamer plays important role in the drift- wave turbulence. The interaction of the mesoscale structure and the turbulence is not only interesting phenomena but also a key to understand the turbulence driven transport in the magnetically confined plasmas. In the cylindrical magnetized device, PANTA, the interaction of the streamer and the drift wave has been found by the bi-spectrum analysis of the turbulence [1]. In order to study the mesoscale physics directly, the 2D turbulence is studied by a fast-framing visible camera system view from a window located at the end plate of the device. The parameters of the plasma is the following; Te~3eV, n ~ 1x1019 m-3, Ti~0.3eV, B=900G, Neutral pressure Pn=0.8 mTorr, a~ 6cm, L=4m, Helicon source (7MHz, 3kW). Fluctuating component of the visible image is decomposed by the Fourier-Bessel expansion method. Several rotating mode is observed simultaneously. From the images, m = 1 (f~0.7 kHz) and m = 2, 3 (f~-3.4 kHz) components which rotate in the opposite direction can be easily distinguished. Though the modes rotate constantly in most time, there appear periods where the radially complicated node structure is formed (for example, m=3 component, t = 142.5~6 in the figure) and coherent mode structures are disturbed. Then, a new rotating period is started again with different phase of the initial rotation until the next event happens. The typical time interval of the event is 0.5 to 1.0 times of the one rotation of the slow m = 1 mode. The wave-wave interaction might be interrupted occasionally. Detailed analysis of the turbulence using imaging technique will be discussed.
References:
[1] T. Yamada, et. al., “Observation of Quasi-Two-Dimensional Nonlinear Interactions in a Drift- Wave Streamer”, Phys. Rev. Lett. 105, 225002 (2010).
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Fig. Time evolution of the rotation phase of the dominant component of the fluctuations and fluctuating component of the images (image of the fluctuations with the frequence of 2 178 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_48 Production of Quiescent Collisionless Plasma over a Wide Operating Range Sayak Bose1, Manjit Kaur1, Prabal K Chattopadhyay1, Joydeep Ghosh1, Yogesh C Saxena1 1Institute for Plasma Research, India Email: [email protected] Experimental study of waves and instabilities and their contribution to transport of particles and energy in a magnetized plasma device has been an integral part of the research dedicated towards the achievement of controlled nuclear fusion. However, it is quite difficult to study them comprehensively in complex systems like Tokamak, because of the noisy environment and small time and space scales involved. To study delicate effects like universal instabilities and its effect in anomalous diffusion, a number of linear devices with low noise plasma source have been designed in the past with varying degrees of success, however with several limitations. We report the production of quiescent magnetized plasma column over a wide operating range using multifilamentary source with low filament spacing in cusp geometry along with a flexible transition magnetic field region between the plasma source chamber and the main chamber [1, 2]. The new device has a much wider operating range and much greater flexibility than other existing quiescent plasma sources like Q machines etc. Quiescent magnetized plasma (n n 1%) is produced over a wide operating range by operating the system in low mirror ratio Rm Bmain Bsource configuration. Demonstrating the effectiveness of this method magnetized argon plasma with low density fluctuation have been produced in the pressure range ~ 5105 to 103 mbar, 109 to 1090 G magnetic field achieving a density of ~ 1010 to 1012 cm-3 and temperature of ~ 2 to 5 eV. The cause for the reduction in density fluctuation at lower mirror ratio is discussed. References: [1] Bose et al., “Inverse mirror plasma experimental device – A new magnetized linear plasma device with a wide operating range”, Rev. Sci. Instrum. 86, 063501 (2015) [2] Bose et al. “Inverse mirror plasma experimental device (IMPED) – a magnetized linear plasma device for wave studies”, J. Plasma Phys. 81, 345810203 (2015) 179 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_50 Effect of Fast Drifting Electrons on Electron Temperature Measurement with a Triple Langmuir Probe Satyajit Chowdhury1, Subir Biswas2, Rabindranath Pal1 1Saha Institute of Nuclear Physics, India 2Weizmann Institute of Science, Israel Email: [email protected] Triple Langmuir Probe (TLP) [1] is a widely used diagnostics for instantaneous measurement of electron temperature and density in low temperature laboratory plasmas as well as in edge region of fusion plasma devices. Presence of a moderately energetic flowing electron component, constituting only a small fraction of the bulk electrons, is also a generally observed scenario in plasma devices where plasmas are produced by electron impact ionization of neutrals. A theoretical analysis [2] of its effect on interpretation of the TLP data for bulk electron temperature measurement is to be presented assuming electron velocity distribution not deviating substantially from a Maxwellian. The study predicts conventional expression from standard TLP theory to give overestimated value of bulk electron temperature. Correction factor is significant and largely depends on population density, temperature and energy of the fast component. Experimental verification of theoretical results is obtained in the Magnetized Plasma Linear Experimental (MaPLE) [3] device of Saha Institute of Nuclear Physics where plasma is produced by ECR method and known to have a fast flowing electron component. References: [1] S. Chen and T. Sekiguchi, “Instantaneous direct-display system of plasma parameters by means of triple probe,” J. Appl. Phys. 36, 2363 (1965). [2] Subir Biswas, Satyajit Chowdhury, Yaswanth Palivela and Rabindranath Pal, “Effect of fast drifting electrons on electron temperature measurement with a triple Langmuir probe” J. Appl. Phys. 118, 063302 (2015). [3] Rabindranath Pal, et al. "The MaPLE device of Saha Institute of Nuclear Physics: Construction and its plasma aspects." Review of Scientific Instruments 81.7 (2010): 073507. 180 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_51 Ponderomotive Force and Backward Raman Scattering in Dense Quantum Plasmas Punit kumar1, Nisha Singh Rathore1 1University of Lucknow, India Email: [email protected] Over the last decade the field of quantum plasma has attracted attention of physicists due to its wide range of applications in modern technology [1]. Quantum plasma where the density is quite high and the de–Broglie thermal wavelength associated with the charged particle approaches the electron Fermi wavelength and exceeds the electron Debye radius is significantly different from the low-density, high-temperature ‘classical plasma’ obeying Maxwell-Boltzmann distribution [2,3]. The present paper is devoted to the study of a laser pulse propagating through high density quantum plasma. The plasma is embedded in a transverse magnetic field. The ponderomotive force imparts a longitudinal velocity to electrons [4-7]. The second harmonic plasma wave undergoes Raman scattering resulting in the excitation of an upper hybrid Langmuir wave and a backscattered second harmonic electromagnetic wave. The interaction dynamics has been built-up using the recently developed quantum hydrodynamic (QHD) model. References: [1] A. P. Mishra, “Dust ion-acoustic shocks in quantum dusty pair-ion plasmas”, Phys. Plasmas 16, 033702 (2009) and references cited therein. [2] Q. Haque, S. Mahmood and A. Mushtaq, “Nonlinear electrostatic drift waves in dense electron- positron-ion plasmas “ Phys. Plasmas 15, 082315(2008). [3] P. K. Shukla and B. Eliasson, “Nonlinear aspects of quantum plasma physics”, Physics-Uspekhi 53 (1), 51(2010) and references cited therein. [4] P. Mora and R. Pellat, “Ponderomotive effects in a magnetized plasma”, Phys. Fluids 22, 2408 (1979) [5] M. K. Srivastava, S. V. Lawande, M. Khan, C. Das, and B. Chakraborty, “Axial magnetic field generation by ponderomotive force in a laser‐produced plasma”, Phys.Fluids B 4, 4086 (1992). [6] H-B. Cai, W. Yu, S-P. Zhu, and C. Zhou, “Generation of strong quasistatic magnetic fields in interactions of ultraintense and short laser pulses with overdense plasma targets”, Phys. Rev. E 76, 036403 (2007). [7] T. Lehner, “Intense self-generated magnetic field in the interaction of a femtosecond laser pulse with an underdense plasma”, Europhys. Lett. 50, 480 (2000). 181 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_52 Anode Glow and Double Layer in DC Magnetron Anode Plasma Samir Chauhan1, Mukesh Ranjan1, Subroto Mukherjee1 1Institute for Plasma Research, India Email: [email protected] Sputtering magnetron is widely used device in research and industry alike. DC planar magnetron employs series of magnets to create magnetic field above the electrode surface which traps electrons in closed drift. Similar device used in reversed polarity power was reported for use in various applications [1, 2]. In contrast to its normal counterpart there is no closed drift effect in there. This device has very limited understanding. We here investigate this device for its discharge properties. Our device is dominated by anode glow. The anode glow is expected to have the electron sheath which provides energy to electron to excite the neutrals. Where as many experimental studies have been reported for anode glow and anode double layer, many of them uses auxiliary anode in the discharge [3, 4]. Most of the cases anode double layer (fire ball/ fire rod) is small structures very near to anode surface which in itself is required to be small. The DC planar magnetron biased in reverse polarity have glow only near anode. Measurements confirm it as anode glow and the presence of electrons sheath is proven. The double layer structure was observed and measured in two mutually perpendicular directions. The double layer shows sub MHz oscillation that is typical of the unstable anode double layer [4, 5]. The dimension of anode glow is relatively large and is primarily in magnetic field free region making it easy to probe. The potential structure still shows large cathode fall but surprisingly visible cathode glow is not present. The device operates very stable for pressure bellow 0.01 mbar. But it shows instabilities such as unstable anode double layer above said pressure. References: [1] Zhao, J. G., and H. Yasuda. "Cathodic plasma polymerization and treatment by anode magnetron torch" Journal of Vacuum Science & Technology A 17.6, 3157 (1999) [2] Ranjan, Mukesh, et al. "Characterization of the Plasma Properties of a Reverse Polarity Planar Magnetron Operated as an Ion Source." Plasma Processes and Polymers 4.S1, S1030 (2007) [3] Baalrud, S. D., B. Longmier, and N. Hershkowitz. "Equilibrium states of anodic double layers." Plasma Sources Science and Technology 18.3, 035002 (2009) [4] Stenzel, R. L., C. Ionita, and R. Schrittwieser. "Dynamics of fireballs." Plasma Sources Science and Technology 17.3, 035006 (2008) [5] Mujawar, M. A., S. K. Karkari, and M. M. Turner. "Properties of a differentially pumped constricted hollow anode plasma source." Plasma Sources Science and Technology 20.1 015024 (2011) 182 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_53 Effect of Trapped Particle Nonlinearity in IAW Solitary Wave Debraj Mandal1, Devendra Sharma1 1Institute for Plasma Research, India Email: [email protected] Plasma support a great variety of coherent nonlinear structures. These include shocks, double layers, solitary wave, vortex etc. In the formation of this coherent structures in collision-less or collisional plasma, involves dispersion and nonlinearities together. Fluid and kinetic models are frequently used to investigate the formation and evolution of this structure. In addition with the macroscopic fluid nonlinearity there is also microscopic trapped particle nonlinearity (TN) which are responsible for the formation of the coherent structures. Including this trapped particle effect the various phenomena in plasma can be explained where the plasma follows the Nonlinear Dispersion Relation [NDR][1]. Solitary electron hole (SEH), Solitary potential dip (SPD), cnoidal electron hole wavelet (CEHWL) are the examples of three special type of trapped particle structures, which are found in both laboratory and space plasma, following NDR. In our present Vlasov simulation, in the regime of small amplitude limit, SEHs are also found to grow on IAW branch of plasma wave, in presence of electron current. In this small amplitude limit the trapped particle nonlinearity is shown to dominate over predominate over the hydrodynamic nonlinearity in the formation of the solitary wave. References: [1] H. Schamel, Phys. of Plasmas, 7, 4831 (2000). [2] D. Mandal, D. Sharma, Phys. Plasmas 21, 102107 (2014) 183 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_54 Installation of a 100 kJ Pulsed Power System to Drive Pulsed Plasma Devices Suramoni Borthakur1, Nayan Talukdar1, Nirod Neog1, Tridip Borthakur1, Rajesh Kumar2, Rishi Verma3, Anurag Shyam Shyam3 1Centre of Plasma Physics-Institute for Plasma Research, India 2Institute for Plasma Research, India 3Bhabha Atomic Research Centre-Visakhapatnam, India Email: [email protected] A pulsed-plasma accelerator is being developed at CPP-IPR, Assam. The accelerator consists of a co-axial electrode assembly housed inside an evacuated chamber that can produce high speed plasma stream of density approximately equal to 1022 m-3. For driving this plasma accelerator, a Pulsed Power System (PPS) of energy nearly 200kJ will be coupled to the electrode assembly. The voltage appearing across the electrode assembly will breakdown the gas present in the inter- electrode gap and create high density plasma. In this paper, the installation of a 100kJ PPS will be discussed, which is one module of the 200 kJ PPS of the plasma accelerator. In general, the conventional high voltage PPS is basically for producing fast output pulses (time periods of few microseconds) according to their uses. In contrast to that, the newly installed pulsed power system at CPP-IPR will produce relatively longer pulse of time period around 1.0ms. This PPS consists of 5 capacitors of rating 180µF, 15 kV each, connected in parallel by using two parallel plates of SS. The newly installed 100kJ bank has been tested and the detailed report of installation and testing will be presented. 184 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_63 Characterization of the Permanent Magnet Based Hydrogen Helicon Plasma Source for Ion Source Application Arun Pandey1, Dass Sudhir Kumar2, Arun Kumar Chakraborty2 1Institute for Plasma Research, India 2ITER-India, Institute for Plasma Research Email: [email protected] The helicon wave plasma (HWP) sources have been found to produce higher density plasmas compared to standard capacitively coupled plasma (CCP) or inductively coupled plasma (ICP) and can be of great importance for ion source development. Due to highly efficient nature of helicon plasma sources, they are also being used in the fields of plasma processing and space exploration. A permanent ring magnet based Helicon plasma source using hydrogen gas has been developed on the basis of the optimized design [1]. The uniqueness of the design is having minimum auxiliary interfaces like cooling system and electrical power system, which are normally required for electromagnet based HWP. In the present configuration, the permanent magnet, instead of electromagnet provides the necessary axial magnetic field [2]. The plasma is generated with the help of a single loop, m = 0 antenna [3, 4] using a 13.56 MHz, 1.2kW source. To characterize the HWP few diagnostic systems are incorporated and used in the experiment which includes a double Langmuir probe for the density measurements and a B-dot probe [5] for identifying the helicon mode by measuring the helicon wave magnetic field. The paper will describe the experimental system and report the experimental characterization data. References: [1] A. Pandey et al., “Conceptual Design of a Permanent Ring Magnet based Helicon Plasma Source module intended to be used in large size fusion grade ion source,” Fusion Engineering and Design (under review). [2] Chen, Francis F, “Physics of helicon discharges,” Physics of Plasmas, 3, 1783-1793 (1996). [3] F. Chen, “Plasma Ionization by Helicon Waves,” Plasma Phys. Controlled Fusion 33, 339 (1991). [4] D. Arnush and Chen, Francis F., “Generalized theory of helicon waves. II. Excitation and absorption,” Physics of Plasmas, 5, 1239-1254 (1998). [5] R. Piejak et al., “Magnetic field distribution measurements in a low-pressure inductive discharge,” Journal of Applied Physics 78, 5296 (1995). 185 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_65 Investigation in Presence of External Forcing and Magnetic Field in a DC Glow Discharge Plasma and Evidence of Nonlinearity Debajyoti Saha1, Pankaj Kumar Shaw1, Sabuj Ghosh1, M S Janaki1, A N S Iyengar1 1Saha Institute of Nuclear Physics, India Email: [email protected] Detection of nonlinearity has been carried out in periodic and aperiodic floating potential fluctuations of DC glow discharge plasma (GDP) in presence of forcing and magnetic field respectively by generating surrogate data using iterative amplitude adjusted Fourier transform (IAAFT) method. We introduce ‘Delay vector variance’ analysis (DVV) for the first time which allows reliable detection of nonlinearity and provides some easy to interpret diagram conveying information about the nature of the experimental floating potential fluctuations (FPF). We have paced the system with a periodic forcing (1, 1.5 KHz) below the dominant frequency keeping the plasma in a periodic regime. An informal test for the bicoherency has been applied to detect the interaction amongst the dominant coherent structures obtained by performing Emperical mode decomposition to strengthen our nonlinearity analysis. An attempt to model the experimental observations by a second order nonlinear ordinary differential equation derived from the fluid equations of plasma has revealed convincing results. 186 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_66 Radio Frequency Emissions from Plasmas due to Laser Induced Breakdown of Materials P Prem Kiran1, L Vinoth Kumar1, Elle Manikanta1 1University of Hyderabad, India Email: [email protected] Laser pulses of short duration, with sufficient intensity, can breakdown target materials into plasma state. This plasma while cooling down, based on the conditions, emits energy across the entire electromagnetic spectrum. These radiations are successfully employed in different fields. However, the low frequency emissions (radio frequency (RF) and microwaves) from these plasmas need more understanding to be utilized to develop into a potential standoff sample identification technique. Hence, there is a need to understand the role of the plasma constituents in the phenomena of low frequency emissions from laser induced breakdown (LIB). The results on the RF emissions, scanned over broad spectral range (30MHz–1 GHz), from single shot nanosecond (7 ns) and picosecond (30 ps) LIB of different target materials are presented. The laser-matter interaction that leads to different plasma current density (Jp) values for different materials, determines the plasma frequency (ωp) which in turn determines the frequencies to be emitted. Hence, the dominant emissions from the LIB of the target materials (conductors, insulators, dielectrics and organic molecules) fall in different specific spectral bands. Thus, with a particular laser and target material, the emissions were observed to be spectral selective [1]. The higher strength of RF emissions from ns-LIB than that with ps-LIB of materials reveals the role of interaction of charged particles with atomic and molecular clusters (in the plasma) in the emission of radiation. The increase in RF emissions from LIB, upto certain input laser energy, shows the importance of the seed electrons in the plasma buildup and the associated RF emissions. At higher input laser energies, the emissions were observed to reduce owing to the increase in the plasma frequency coming closer the laser frequency thus reducing input laser- plasma interaction. The role of laser produced plasma parameters and the interaction of plasma constituents in RF emissions were further confirmed by the studies on RF emissions from ns and ps LIB of atmospheric air under different focusing conditions [2]. Besides, the role of target surface in plasma formation and the resulting emissions during laser interactions were studied using the LIB of compacts of copper micro powders of different particles sizes. To summarize, RF emissions from LIB of different materials can be tailored, for various applications, by tuning the laser and target parameters. In addition, they also give an insight into the target material generating the plasma, due to their spectral selective nature which scales with Iλ2 that has potential applications in the detection of hazardous materials. References: [1] L. Vinoth Kumar, E. Manikanta, C. Leela, P. Prem Kiran, “Spectral selective radio frequency emissions from laser induced breakdown of target materials,” Appl. Phys. Lett. 105, 064102 (2014). [2] L. Vinoth Kumar, E. Manikanta, C. Leela, P. Prem Kiran, “Characteristics of radio frequency emissions from laser induced breakdown of atmospheric air,” (under review). 187 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_69 Effect of Transverse Magnetic Field on the Steady State Solutions of a Bursian Diode Sourav Pramanik1, Nikhil Chakrabarti1, Victor Kuznetsov2 1Saha Institute of Nuclear Physics, Kolkata 2Ioffe Institute, Russia Email: [email protected] The effect of external transverse magnetic field on a steady-state planar vacuum diode [1] driven by a cold electron beam is presented. Three distinct situations are studied and they are as follows: (a) when no electrons are reflected back by the magnetic field [2], (b) when electrons are reflected partially and (c) totally. The emitter electric field is evaluated as a characteristics function for the existence of solutions depending on diode length, applied voltage and magnetic field strength. All steady state solutions corresponding to the cases stated above, are visualized through the "emitter electric field strength vs diode gap" parametric plot. It is shown that, due to the inclusion of magnetic field a new region of non-unique solutions appear. An external magnetic field seems to have profound effect in controlling fast electronic switches based on Bursian diode. 188 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_70 Wave-breaking Amplitudes of Relativistically Strong Electrostatic Waves in Cold Electron-Positron-Ion Plasmas Mithun Karmakar1, Chandan Maity2, Nikhil Chakrabarti1 1Saha Institute of Nuclear Physics, India 2Government General Degree College Singur, Hooghly, India Email: [email protected] A one-dimensional nonlinear propagation of relativistically strong electrostatic waves in cold electron-positron-ion (EPI) plasmas has been analyzed in pseudo potential approach. The motion of all the three species, namely, electron, positron, and ion has been treated to be relativistic. The wave breaking electric field amplitude of such an electrostatic wave has been derived, showing its dependence on the relativistic Lorentz factor associated with the phase velocity of the plasma wave, on the electron/positron to ion mass ratio, and on the ratio of equilibrium ion density to equilibrium electron/positron density. 189 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_71 Nonlinear Coherent Structures of Alfven Wave in a Collisional Plasma Sayanee Jana1, Nikhil Chakrabarti1, Samiran Ghosh2 1Saha Institute of Nuclear Physics, Kolkata, India 2University of Calcutta, India Email: [email protected] Low-frequency Magneto Hydrodynamic waves [1] in general and Alfv´en wave, in particular, occurs in various physical problems starting from laboratory to space plasma [2]. These low frequency disturbances make the magnetic fluctuations large enough so that nonlinear coupling becomes finite [3]. Among these low-frequency waves, nonlinear Alfv´en wave has become a topic of intense research due to its applications in various physical processes, related to particle energization in magnetized plasma, self-modulation in strongly magnetized plasma, tokamak plasma heating, interplanetary shocks, turbulence etc. In the present work, we have investigated weakly nonlinear Alfv´en wave dynamics in the framework of Lagrangian two-fluid theory in a compressible cold magnetized plasma in presence of finite electron inertia effect. The electron-ion collision induced dissipation effect is also taken into account. In the finite amplitude limit, we have shown that the collisionless Alfv´en wave is governed by a modified Korteweg-de Vries (mKdV) equation. In presence of collision it becomes a modified Korteweg-de Vries -Burgers (mKdVB) equation, where the electron inertia is found to act as a dispersive effect and the electron-ion collision serves as a dissipation which is responsible for the Burgers term. In the long wavelength limit, we have also investigated another important physical phenomenon, known as the wave modulation instability [4]. The dynamics of this modulated wave is shown to be governed by a nonlinear Schrödinger equation (NLSE) [5] with a linear damping term arising due to electron-ion collision. These two nonlinear equations are analyzed by means of analytical and numerical simulation to elucidate the various aspects of the phase-space dynamics of the nonlinear wave. Both the results reveal that nonlinear Alfven wave exhibits shock, dissipative envelope and breather like structures. Numerical simulation also predicts the formation of Alv´enic rogue wave and giant breathers. References: [1] H. Alfven, “Existence of Electro-Magnetic Hydrodynamic Waves,” Nature (London), 150, 405 (1942). [2] N. F. Cramer, “The Physics of Alfven Waves,” WILEY-VCH, Germany, 2001. [3] S. Spangler, “Nonlinear Waves and Chaos in Space Plasmas,” Terra Scientific Publishing Company, TERRA-PUB, Tokyo, 1997. [4] P. M. Bellan, “Fundamental of Plasma physics,” The Cambridge Univ. Press, Cambridge, 2006. [5] M. Saito, S. Watanabe, and H. Tanaka, “Modulational Instability of Ion Wave in Plasma with Negative Ion,” J. Phys. Soc. Japan, 53, 2304 (1984). 190 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_72 Parallel Connection Length and Flow-fluctuation Cycle in Simple Toroidal Device Umesh Kumar1, Shekar Goud Thatipamula1, Rajaraman Ganesh1, Yogesh C Saxena1, Raju Daniel1 1Institute for Plasma Research, India Email: [email protected] In a recent series of experiments [1], it was demonstrated that fluctuations drive flow in simple current less toroidal device BETA. In particular, the effect of magnetic field strength has been experimentally studied in great detail [2]. It has been found that with increasing toroidal field strength, these systems undergo a coherent to turbulent transition [2]. It was further shown that the transition was reinforced by flow generated due to fluctuations. In this work, external vertical magnetic field has been applied to experimentally vary the parallel connection length Lc, which in turn controls the nature of the fluctuations by varying k|| [3]. Extensive experimental results on flow-fluctuation dynamics and possible explanations will be presented. References: [1] T. S. Goud thesis, Institute for Plasma research, Gandhinagar, 2012. [2] T. S. Goud et al Phys. Plasmas 19, 032307, 2012. [3] Muller et al in Phys. Rev. Lett. 93, 16 (2004). 191 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_80 Controllable Location of Polarization Reversal in Nonuniform Helicon Plasma Sonu Yadav1, Prabal K Chattopadhyay1, Joydeep Ghosh1, Soumen Ghosh1 1Institute for Plasma Research, India Email: [email protected] The experiments have been performed using m = +1 half wavelength helical antenna using rf power at frequency 13.56MHz. The right hand circularly polarization (RHCP) m = +1 mode propagates in the direction of applied magnetic field, whereas the left hand circularly polarization (LHCP) m = -1 mode propagates in the opposite direction. The radial wave field component and phase measurement shows the polarization of wave gets reversed i.e. RHCP become LHCP or vice versa [1, 2]. A theoretical model has been presented for radially nonuniform cylindrical plasma. The radial profiles of helical wave magnetic field component Br, Bθ and Bz, and phase profile of same are computed for different radial density profiles. Observation shows that polarization of wave gets reversed at certain radial location. The polarity reversal (or zero crossing) of amplitude of azimuthal component (Bθ) is related to wave polarization reversal. It is shown that location of polarization reversal can be controlled by the radial wavelength and nonuniform density profile. References: [1] Barada et al., “Experimental observation of left polarized wave absorption near electron cyclotron resonance frequency in helicon antenna produced plasma” Phys. Plasma 20, 012123 (2013). [2] J. P. Klozenberg, et al., “The dispersion and attenuation of helicon waves in a uniform cylindrical plasma” Journal of Fluid Mechanics, vol. 21, pp. 545-563, (1965). 192 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_82 Hot Tungsten Plate Based Ionizer for Cesium Plasma in a Multi-Cusp Field Experiment Amitkumar D Patel1, Meenakshee Sharma1, Narayanan Ramasubramanian1, Prabal K Chattopadhyay1 1Institute for Plasma Research, India Email: [email protected] In a newly proposed basic experiment, contact-ionized cesium ions will be confined by a multi cups magnetic field configuration. The cesium ion will be produced by impinging collimated neutral atoms on an ionizer consisting of the hot tungsten plate. The temperature of the tungsten plate will also be made high enough (~2700 K) such that it will contribute electrons also to the plasma. It is expected that at this configuration the cesium plasma would be really quiescent and would be free from even the normal drift waves observed in the classical Q-machines. For the ionizer a design based on F. F. Chen’s design [1] was made. This ionizer is very fine machining and exotic material like Tungsten plate, Molybdenum screws, rings, and Boron Nitride ceramics etc. The fine and careful machining of these materials was very hard. In this paper, the experience about to join the tungsten wire to molybdenum plate and alloy of tantalum and molybdenum ring is described. In addition experimental investigations have been made to measure 2D temperature distribution profile of the Tungsten hot plate using infrared camera and the uniformity of temperature distribution over the hot plate surface is discussed. References: [1] F. F. Chen, “Coaxial cathode design for Plasma source” Rev. Sci. Instrum., 40, 8 (1969). 193 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_107 Development of Three Dimensional Magnetic Field Probe with Signal Conditioning Electronics Kiran Patel1, Narayan Behera1, Rajesh Kumar Singh1, Ajai Kumar1 1Institute for Plasma Research, India Email: [email protected] Three dimensional magnetic field probes have been constructed and calibrated to measure self- generated magnetic field in laser produced plasma. The magnetic probe was made on the 3.2 mm Teflon cube where twisted copper wire of Gauge 40 wounded on it. Each axis having two loops with 5 turns which are connected in opposite direction to reduce the stray noise. Coil area, number of turns, self-inductance and shielding are carefully optimized to achieve the accurate measurement of magnetic field with reduced noise level. A separate differential amplifier with variable gain is designed and developed for the amplification of the each axis signal. The calibration of the probe is carried out with the known field of Helmholtz coil. Details of technical aspect, optimization, and performance tests of the developed probe are briefly described. References: [1] Eveson ET et al., “Design, construction and calibration of a three axis, high frequency magnetic probe (B-dot probe) as a diagnostic for exploding plasmas,” Rev. Sci. Instruments, 80(11), 113505 (2009). 194 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_121 State of Art Data Acquisition System for Large Volume Plasma Device Ritesh Sugandhi1, Pankaj Srivastava1, Amulya Kumar Sanyasi1, Prabhakar Srivastav1, Lalit Mohan Awasthi1, Shiban Krishna Mattoo1, Vijay Parmar2, Keyur Makadia2, Ishan Patel2, Sandeep Shah2 1Institute for Plasma Research, India 2Optimized Solutions Private Limited, India Email: [email protected] The Large volume plasma device (LVPD) is a cylindrical device (=2m, L= 3m) dedicated for carrying out investigations on plasma physics problems ranging from excitation of whistler structures to plasma turbulence especially, exploring the linear and nonlinear aspects of electron temperature gradient(ETG) driven turbulence, plasma transport over the entire cross section of LVPD. The machine operates in a pulsed mode with repetition cycle of 1 Hz and acquisition pulse length of duration of 15ms, presently, LVPD has VXI data acquisition system [1] but this is now in phasing out mode because of non-functioning of its various amplifier stages, expandability and unavailability of service support. The VXI system has limited capabilities to meet new experimental requirements in terms of numbers of channel (16), bit resolutions (8 bit), record length (30K points) and calibration support. Recently, integration of new acquisition system for simultaneous sampling of 40 channels of data, collected over multiple time scales with high speed is successfully demonstrated, by configuring latest available hardware and in- house developed software solutions. The operational feasibility provided by LabVIEW platform is not only for operating DAQ system but also for providing controls to various subsystems associated with the device. The new system is based on PXI express instrumentation bus [2] and supersedes the existing VXI based data acquisition system in terms of instrumentation capabilities. This system has capability to measure 32 signals at 60MHz sampling frequency and 8 signals with 1.25 GHz with 10 bit and 12 bit resolution capability for amplitude measurements. The PXI based system successfully addresses and demonstrate the issues concerning high channel count, high speed data streaming and multiple I/O modules synchronization. The system consists of chassis (NI 1085), 4 high sampling digitizers (NI 5105), 2 very high sampling digitizers (NI 5162), data streaming RAID drive (NI-8266) and timing and synchronization module (NI-6674T). The system is developed on LabVIEW 2014 using object oriented design patterns. The software provides the configuration and handling horizontal, vertical and trigger parameters for I/O modules and archives raw data into binary and configuration data in XML format. The paper will highlight the requirements, rationales for hardware and software selection, design architecture, development, integration and test results. References: [1] G. B. Patel et.al., “Data acquisition system for large volume plasma device, Rev. Sci. Instruments ,73,1779(2002) [2] PXI bus: http://www.ni.com/pxi 195 | P a g e 10th Asia Plasma & Fusion Association Conference Popular Talk 196 | P a g e 10th Asia Plasma & Fusion Association Conference Exploration of the Solar System & Beyond: The Indian Scene Jitendra Nath Goswami1 1Physical Research Laboratory, Ahmedabad Email: [email protected] The success of India’s first planetary mission, Chandrayaan-1, ushered a new era in the Indian Space Program. The discovery of water on moon and other novel results provided impetus for further exploration of the solar system and beyond. Successful placement of a spacecraft around Mars in the very first attempt is another landmark achievement. Preparation for Chandrayaan-2, with Orbiter-Lander-Rover combination, is currently in progress. A dedicated astronomy satellite, Astrosat, was launched recently. Future plans include a mission for solar observation and proposal for a mission to Venus. 197 | P a g e 10th Asia Plasma & Fusion Association Conference Oral Session-1 198 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 0_27 Experimental Study of Plasma Current Ramp-up by the Lower Hybrid Wave in the TST-2 Spherical Tokamak Yuichi Takase1, Akira Ejiri1, Tsujii Naoto1, Takahiro Shinya1, Hirokazu Furui1, Hiroto Homma1, Kenta Nakamura1, Masateru Sonehara1, Wataru Takahashi1, Toshihiro Takeuchi1, Hiro Togashi1, Kazuya Toida1, Satoru Yajima1, Hibiki Yamazaki1, Yusuke Yoshida1 1The University of Tokyo, Japan Email:[email protected] The development of an effective plasma current ramp-up method is an important issue for future applications of the spherical tokamak as a fusion neutron source, a demonstration reactor, or a commercial reactor, for which the elimination of the central solenoid is considered to be a necessity. Plasma initiation and plasma current ramp-up have been studied on the TST-2 spherical tokamak at the University of Tokyo (R0 = 0.38 m, a = 0.25 m, B0 = 0.3 T, Ip = 0.1 MA) [1] using waves in various frequency ranges from the ion cyclotron range to the electron cyclotron range. Presently, the most effective wave is believed to be the lower hybrid wave. It is critically important to keep the plasma density low enough during the plasma current ramp-up phase for effective ramp-up [2]. Up to now, plasma current ramp-up to nearly 20 kA has been achieved. In plasmas with such low currents, the confinement of energetic electrons, which carry most of the plasma current, is expected to be poor because of the large deviations of their orbits from the flux surface. RF power modulation experiments indicate that a substantial fraction of energetic electrons are lost promptly. In addition, it is suspected that a significant fraction of the wave energy is lost in the peripheral region of the plasma. Numerical calculations using wave codes (ray-tracing or full- wave) and a Fokker-Planck code have been performed in order to identify possible ways to improve the efficiency of plasma current ramp-up, Operation at higher magnetic fields is favorable for improving the accessibility of the lower hybrid wave to the plasma core. Wave power losses in the edge plasma could be reduced by improving the single-pass absorption. This can be accomplished by launching the lower hybrid wave from the inboard top region instead of the outboard midplane. The parallel wavenumber of the lower hybrid wave launched from the top antenna upshifts quickly, and is absorbed efficienctly by electrons. Unlike the wave launched from the outboard midplane which travels through the edge plasma after the first pass through the plasma, the wave launched from the top is absorbed during its first pass through the plasma, resulting in the added benefit of preventing wave losses in the edge plasma. A top- launch antenna has been developed and will be installed during 2015. References: [1] Y. Takase, et al., “Initial results from the TST-2 spherical tokamak,” Nucl. Fusion 41, 1543 (2001). [2] T. Shinya, et al., “Non-inductive plasma start-up experiments on the TST-2 spherical tokamak using waves in the lower-hybrid frequency range,” Nucl. Fusion 55, 073003 (2015). 199 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 0_152 ELM Control using Low-n RMPs in KSTAR and its Perspective to Beyond- ITER Youngmu M Jeon1, Jong-kyu Park2, Yongkyoon In1, Jayhyun Kim1, Siwoo W Yoon1, Gunyoung Y Park1, Yeong-kook Oh1, Hyeon Park1,3, KSTAR Team1 1National Fusion Research Institute, Japan 2Princeton Plasma Physics Laboratory, USA 3Ulsan National Institute of Science and Technology, Korea Email: [email protected] In this talk, we present the recent experimental progresses on ELM control using low-n magnetic perturbations in KSTAR, where the non-axisymmetric perturbation fields are provided by using three rows of toroidally segmented coil system similarly with those in ITER. First of all, we have successfully demonstrated that type-I ELMs can be completely suppressed using low-n (n=1 [1] or n=2 [2]) magnetic perturbations in a wide range of q95 (3.5~7.5). Particularly for n=1 cases, each narrow q95 windows within the range show a clear correlation with the edge rational surfaces such as q95=5.0 (~5/1), 6.0 (~6/1), and 7.5 (~7/1). Therefore, it suggests that the physics mechanism of ELM-suppression under the edge magnetic perturbations is strongly associated with a resonant plasma response and thus it is important for ELM-suppression to control a specific rational surface into a proper edge region. Furthermore, the ELM-suppression has been achieved by using a single row of coil (mid-plane coil alone), two off-mid plane coils, or all three rows of coils. It shows the flexibility of magnetic perturbations, the redundancy for coil failure (the mid-plane coil alone corresponds to eight coils failure among 12 coils), and the possibility of a simpler coil design for ITER and beyond. On the other hand, the influence of magnetic perturbations on global confinement and transport is also addressed with importance. A certain amount of reduction of global confinements is a well-known phenomenon associated with magnetic perturbations, such as a strong density pump- out and reductions of plasma stored energy and beta. In addition, we have found a variety of effects depending on the perturbed field configurations, such as a strong rotation damping with the ‘mid-plane alone’ configuration and a distinctive confinement improvement with the ‘all three coils’ configuration accompanying the ELM-suppression. Overall experimental observations described above show a practical possibility and a potential of using low-n magnetic perturbations on ITER ELM control. However simultaneously it reveals out several critical physics issues in application to ITER, which will be discussed further for the application to beyond-ITER. References: [1] Y. M. Jeon, et al., “Suppression of Edge Localized Modes in High-Confinement KSTAR Plasmas by Nonaxisymmetric Magnetic Perturbations”, Phys. Rev. Lett., 109, 035004 (2012). [2] Y. M. Jeon, et al., “Successful ELM Suppressions in a Wide Range of q95 Using Low n RMPs in KSTAR and its Understanding as a Secondary Effect of RMP”, IAEA Fusion Energy Conference, St. Petersburg, Russia (2014) 200 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 0_179 Development of Long Pulse Radiofrequency Heating and Current Drive Systems and Scenarios for WEST Annika Ekedahl1, Clarisse Bourdelle1, Jean-Francois Artaud1, Jean-Michel Bernard1, Laurent Colas1, Joan Decker1, Léna Delpech1, Rémi Dumont1, Marc Goniche1, Walid Helou1, Julien Hillairet1, Gilles Lombard1, Roland Magne1, Patrick Mollard1, Eric Nardon1, Yves Peysson1, Emmanuelle Tsitrone1, Zhaoxi Chen2, Bojiang Ding2, Xianzu Gong2, Miaohui Li2, Yuntao Song2, Yongsheng Wang2, Qingxi Yang2, Yanping Zhao2 and Tore Supra / WEST Team 1CEA, IRFM, France 2Institute of Plasma Physics, Chinese Academy of Sciences,China Email: [email protected] The longstanding expertise of the Tore Supra Team in long pulse radiofrequency (RF) heating and current drive systems will now be exploited in WEST (tungsten-W Environment in Steady- state Tokamak) [1]. WEST will allow an integrated long pulse tokamak programme for testing W-divertor components at ITER-relevant heat flux (10-20MW/m2), while treating crucial aspects for ITER-operation, such as avoidance of W-accumulation in long discharges, monitoring and control of heat fluxes on the metallic plasma facing components (PFCs) and coupling of RF waves in H-mode plasmas. Scenario modelling using the METIS-code shows that ITER-relevant heat fluxes are compatible with the sustainment of long pulse H-mode discharges, at high power (up to 15MW/30s at IP=0.8MA) or high fluence (up to 10MW, up to 1000s at IP=0.6MA) [2], all based on RF heating and current drive using Ion Cyclotron Resonance Heating (ICRH) and Lower Hybrid Current Drive (LHCD). To allow coupling to H-mode plasmas, three ELM-resilient ICRH antennas have been designed for WEST. They will be fabricated and provided as in-kind contribution by ASIPP (Hefei), within the framework of the Associated Laboratory IRFM-ASIPP. Furthermore, the ICRH generator has been upgraded to allow high power operation (9MW/30s) at high reflected power (VSWR=2). The WEST ICRH system is thus the first ever ICRH system combining continuous wave (CW) operation at high power and load tolerance capability for coupling on H-modes. The nominal operating frequencies are 53±2MHz and 55.5± 2MHz, in order to allow flexibility in the location of the resonance layer around the magnetic axis. The LHCD system, with capability to inject 7MW/1000s, is an indispensable tool for long pulse scenarios. The LH power deposition and current profiles have been modelled with the recent “Tail LH” model in C3PO/LUKE, which has proven to reproduce well the experimental LHCD results on Tore Supra, as well as on EAST [3]. The simulations show that the LH wave 19 -3 absorption (n//0=2.0) takes place in the region r/a=0.3-0.6 at pedestal a density of nped=3×10 m . The WEST device with its relevant diagnostics will allow bringing new insight into the LHCD physics and will allow validating the Passive-Active-Multijunction as a viable LHCD launcher concept for ITER and long pulse tokamaks. 201 | P a g e 10th Asia Plasma & Fusion Association Conference References: [1] J. Bucalossi et al., Fusion Eng. Des. 89, 907 (2014). [2] C. Bourdelle et al., Nucl. Fusion 55, 063017 (2015). [3] Y. Peysson et al., submitted to Plasma Phys. Control. Fusion (2015). 202 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 0_184 Behaviors of Impurity in ITER and DEMOs using BALDUR Integrated Predictive Modeling Code Thawatchai Onjun1, Wannapa Buangam1, Apiwat Wisitsorasak2 1Sirindhorn International Institute of Technology, Thailand 2King Mongkut’s University of Technology Thonburi, Thailand Email: [email protected] The behaviors of impurity are investigated using self-consistent modeling of 1.5D BALDUR integrated predictive modeling code, in which theory-based models are used for both core and edge region. In these simulations, a combination of NCLASS neoclassical transport and Multi- mode (MMM95) anomalous transport model is used to compute a core transport. The boundary is taken to be at the top of the pedestal, where the pedestal values are described using a theory- based pedestal model. This pedestal temperature model is based on a combination of magnetic and flow shear stabilization pedestal width scaling and an infinite-n ballooning pressure gradient model. The time evolution of plasma current, temperature and density profiles is carried out for ITER and DEMOs plasmas. As a result, the impurity behaviors such as impurity accumulation and impurity transport can be investigated. 203 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_284 Rapid Purification of Hydrogen Isotope Gas by Palladium Alloy Membrane Separator Xiong Yifu1, Song Jiangfeng1，Jing Wenyong1，He Mingmin1，Ba Jingwen1, Shiyan1 1Institute of Materials, China Academy of Engineering Physics, China Email: [email protected] Efficient and rapid purification of hydrogen isotopes is one of the core technologies of deuterium-tritium fuel cycle in fusion reactor. Applying this technology during operation, not only can a large amount of unreacted (also called unburned) deuterium / tritium gas be cyclic utilized, but the environmental release amount of tritium can also be controlled efficiently. In this paper, a fast purification of hydrogen isotope gas was carried out via a device employing spiral palladium-yttrium alloy tube as its core component. The result indicated that under different temperatures and pressures, the overall leakage rate was down to less than 1.510 – 9 Pa.m3.s-1, the recovery rate for hydrogen isotopes of low content was up to more than 99%, and the daily processing capacity had approximately a tenfold increase to 20ml comparing with the conventional straight palladium alloy tube. The fundamental solution was achieved on the rapid removal of tiny amount of 3He gas in a large batch of hydrogen isotope gas thus the significant increase was also acquired on the purification capacity for hydrogen isotopes. 204 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 0_214 Measurements and Controls Implementation for the WEST Project Raju Daniel1, P Moreau2, Manisha Bhandarkar1, S Brémond2, J Bucalossi2, Vishnu K Chaudhari1, X Courtois2, Jasraj Dhongde1, C Gil2, Aveg Kumar1, Praveena Kumari1, M Lewerentz4, P Lotte2, Imran Mansuri1, Harish Masand1, O Meyer2, M Missirlian2, E Nardon2, R Nouailletas2, Kiritkumar B Patel1, Sutapa Ranjan1, C Rapson3, G Raupp3, N Ravenel2, F Samaille2, Manika Sharma1, J Signoret2, A Spring4, J M Travere2, W Treuterrer3, A Werner4, WEST team2 1Institute for Plasma Research (IPR), Near Indira Bridge, Bhat, Gandhinagar- 382 428, Gujarat, 2IRFM, CEA, F-13108 Saint Paul lez Durance, France 3Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstraße 2, 85748 Garching, Germany 4Max-Planck-Institut für Plasmaphysik, EURATOM Association, Teilinstitut Greifswald, Wendelsteinstraße 1, D-17491 Greifswald, Germany Email: [email protected] The WEST (W Environment for Steady-state Tokamak) [1] project consists in a major upgrade of the superconducting medium size tokamak Tore Supra to minimize risks for the ITER divertor procurement in terms of cost, delays and performance. This modification consists in changing the present circular magnetic configuration to a divertor configuration and implementing an ITER like actively cooled Tungsten divertor. Tests of the divertor will be performed according to 2 main scenarios: high power (Ip=0.8MA lasting 30s with 15MW injected power) and high fluence (Ip=0.6MA lasting 1000s with 12MW injected power). Heat load on divertor target will range from a few MW/m2 up to 20MW/m2 depending on the X point location and the heat flux decay length. To reach these goals while ensuring the protection of the machine, major changes and significant developments are on-going on the measurement systems (diagnostics); the control, data access and communication (CODAC); the plasma control system (PCS), the monitoring and protection of the first wall and modelling to prepare the restart of the plasma. This paper provides an overview of the diagnostics implemented on WEST and gives more details on the infra-red system which is one of the main systems used to analyze the heat loads and ensure the machine protection. The modification of the CODAC and communications networks is also discussed. The new functionalities and architecture of the WEST PCS are detailed; especially it ensures the orchestration of many subsystems such as diagnostics, actuators and allows handling asynchronous off-normal events during the plasma discharge. In correlation the plasma discharge is now seen as a set of elementary pieces (called segments) joints together. Development of new plasma controllers will be addressed. An overview of the first wall monitoring activity and development is provided. Finally preparing the plasma restart requires control oriented modelling and simulations devoted to the control of the plasma shape will be presented. References: [1] J. Bucalossi et al., Fusion Engineering and Design 89 (2014) 907–912 205 | P a g e 10th Asia Plasma & Fusion Association Conference Poster Session-4 206 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_122 Super Rogue Wave in Plasma Pallabi Pathak1, Sumita Kumari Sharma1, Heremba Bailung1 1Institute of Advanced Study in Science and Technology, India Email:[email protected] The evolution of super rogue wave having amplitude ~5 times the background wave has been observed in multicomponent plasma with critical concentration of negative ions in a double plasma device. In normal electron-ion plasma the ion acoustic solitons are described by the Korteweg-de Vries (KdV) equation [1, 2]. At a critical concentration of negative ions, the ion acoustic modified KdV solitons are found to propagate [3]. Multicomponent plasma also supports the propagation of a special kind of soliton namely ‘Peregrine soliton’ at critical concentration of negative ions. Peregrine soliton is a doubly localized solution of the nonlinear Schrodinger equation (NLSE) having amplitude 3 times the background carrier wave [4, 5]. In a double plasma device, ion-acoustic Peregrine soliton is excited by applying slowly varying amplitude modulated continuous sinusoidal signal to the source anode and described by the rational solution of NLSE. The ion acoustic wave is modulationally unstable in multicomponent plasma with critical concentration of negative ions and an initial modulated wave perturbation is found to undergo self-modulation to form localized structures by balancing the nonlinearity with the dispersion. In presence of higher order nonlinearity, propagation of a high amplitude (~5 times of background carrier wave) ion acoustic Peregrine soliton has been observed experimentally. The existence of such types of higher order wave has been reported in other dispersive media [6, 7]. These are considered to be the prototype of super rogue wave in deep water. In this work, experimental results on the evolution of super rogue wave in a double plasma device are presented and compared with the numerical solution of NLSE. References: [1] H. Washimi and T. Taniuti, “Propagation of Ion-Acoustic Solitary Waves of Small Amplitude”, Phys. Rev. Lett. 17, 996 (1966). [2] H. Ikezi, R. J. Taylor and D.R. Baker, “Formation and Interaction of Ion-Acoustic Solitons” Phys. Rev. Lett. 25, 11 (1970). [3] Y. Nakamura and I. Tsukabayashi, “Observation of Modified Korteweg-de Vries Solitons in a Multicomponent Plasma with Negative Ions”, Phys. Rev. Lett. 52, 2356 (1984). [4] D. H. Peregrine, “Water Waves, Nonlinear Schrödinger Equations and their Solutions”, J. Austral. Math. Soc. Series B, Appl. Math 25, 16 (1983). [5] H. Bailung, S. K. Sharma and Y. Nakamura, “Observation of Peregrine Solitons in a multicomponent Plasma with Negative Ions” Phys. Rev. Lett. 107, 255005 (2011). [6] A.Chabchoub, N. Hoffmann, M. Onorato and N. Akhmediev, “Super Rogue Waves: Observation of Higher Order Breather in Water Waves”, Phys. Rev. X 2, 011015 (2012). [7] M. Akbari-Moghanjoughi, “Electrostatic Rogue-Waves in Relativistically Degenerate Plasmas”, Phys. Plasmas, 21, 102111 (2014). 207 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_123 Experiment on Dust Acoustic Solitons in Strongly Coupled Dusty Plasma Abhijit Boruah1, Sumita Kumari Sharma1, Heremba Bailung1 1Institute of Advanced Study in Science and Technology, India Email: [email protected] Dusty plasma, which contains nanometer to micrometer sized dust particles along with electrons and ions, supports a low frequency wave called Dust Acoustic wave, analogous to ion acoustic wave in normal plasma [1, 2]. Due to high charge and low temperature of the dust particles, dusty plasma can easily transform into a strongly coupled state when the Coulomb interaction potential energy exceeds the dust kinetic energy [3]. Dust acoustic perturbations are excited in such strongly coupled dusty plasma by applying a short negative pulse (100 ms) of amplitude 5 – 20 V to an exciter [4]. The perturbation steepens due to nonlinear effect and forms a solitary structure by balancing dispersion present in the medium. For specific discharge conditions, excitation amplitude above a critical value, the perturbation is found to evolve into a number of solitons. The experimental results on the excitation of multiple dust acoustic solitons in the strongly coupled regime are presented in this work. The experiment is carried out in radio frequency discharged plasma produced in a glass chamber at a pressure 0.01 – 0.1 mbar. Few layers of dust particles (~ 5 m in diameter) are levitated above a grounded electrode inside the chamber. Wave evolution is observed with the help of green laser sheet and recorded in a high resolution camera at high frame rate. The high amplitude soliton propagates ahead followed by smaller amplitude solitons with lower velocity. The separation between the solitons increases as time passes by. The characteristics of the observed dust acoustic solitons such as amplitude- velocity and amplitude- Mach number relationship are compared with the solutions of Korteweg- de Vries (KdV) equation. References: [1] N. N. Rao, P. K. Shukla and M. Y. Yu, “Dust-acoustic waves in dusty plasma”, Planetary and Space Science, 38, 543 (1990). [2] P. Bandyopadhyay, G. Prasad, A. Sen and P. K. Kaw, “Experimental study of nonlinear dust acoustic solitary waves in a dusty plasma”, Physical Review Letters, 101, 065006 (2008). [3] H. Ikezi, “Coulomb solid of small particles in plasma”, Physics of Fluids, 29, 1764 (1986). [4] S. K. Sharma, A. Boruah, and H. Bailung, “Head-on collision of dust-acoustic solitons in a strongly coupled dusty plasma” Physical Review E, 89, 013110 (2014). 208 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_125 Controllable Transition from Positive Space Charge to Negative Space Charge in an Inverted Cylindrical Magnetron Ramkrishna Rane1, Mainak Bandyopadhyay2, Mukesh Ranjan1, Subroto Mukherjee1 1FCIPT-Institute for Plasma Research, India 2ITER-India, Institute for Plasma Research,India Email: [email protected] The combined effect of magnetic field (B), gas pressure (P) and the corresponding discharge voltage on the discharge properties of argon in inverted cylindrical magnetron has been investigated. In the experiment, anode is biased with continuous 10 ms sinusoidal half wave. It is observed that at a comparatively high magnetic field (i.e. greater than 200 gauss) and low operating pressure (i.e. less than 1x10-3 mbar) the discharge extinguishes and demands a high voltage to reignite the discharge. Discharge current increases with increase in magnetic field and start reducing at sufficiently high magnetic field for a particular discharge voltage due to restricted electron diffusion towards anode. It is observed that B/P ratio plays an important role in sustaining the discharge and is constant for a discharge voltage. The B/P ratio varies linearly with discharge voltage. The discharge is transformed to negative space charge regime from positive space charge regime at that constant B/P ratio. Radial profile of the floating potential in between the two electrodes has been measured for different magnetic fields for inverted configuration. At a particular higher magnetic field (beyond 100 gauss), the floating potential increases gradually with the radial distance from cathode whereas it remains almost constant at lower magnetic field. 209 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_126 Measurement of Electron Energy Probability Function in Weakly Magnetized Plasma Deiji Kalita1, Bharat Kakati2, Bipul Kumar Saikia1, Mainak Bandyopadhyay3, Siddhartha Sankar Kausik1 1Centre of Plasma Physics-Institute for Plasma Research, India 2Institute for Plasma Research, India 3ITER-India, Institute for Plasma Research, India Email: [email protected] The electron energy probability function (EEPF) is one of the key factors for the evaluation of the plasma parameters by the Langmuir probe (LP) theories. It is known that the presence of magnetic field can influence the anisotropy of the electron energy probability function (EEPF) [1-3]. Knowledge of the real EEDF is of great importance in understanding the underlying physics of processes occurring at the magnetized plasma, such as the formation of transport barriers, cross-field diffusion coefficients and plasma–substrate interactions. Although the electric probe method is one of the oldest methods in plasma physics itself, it is yet not fully understood in presence of magnetic field [4]. In the present experiment, the application of LPs to evaluate EEPF in presence of magnetic fields within the range (594 –32) G is investigated. The data recorded for EEPFs in magnetic fields and in dust is acquired using current–voltage characteristics measured in low pressure hydrogen plasma. The values of plasma density, electron temperature and EEPF are evaluated with a single cylindrical Langmuir probe at different axial positions (1cm to 6 cm) from the magnet. From the recent EEPF observations in presence of magnetic field, it shows a bi-Maxwellian EEPF structure at different magnetic fields. But at different magnetic field, it is observed that the low energy electron population changes whereas the high-energy electron population remains almost constant. EEPF measurement shows almost identical behaviour with the unmagnetized plasma when the larmour radius of electron is greater than or equal to 10 times of the probe radius. References: [1] M. Tich, P. Kudrna, J.F. Behnke, C. Csambal and S. Klagge, “Langmuir Probe Diagnostics for Medium Pressure and Magnetised Low-Temperature Plasma”, J. Phys IV France 7 (1 997) [2] A. Aanesland, J. Bredin, P. Chabert, and V. Godyak , “Electron energy distribution function and plasma parameters across magnetic filters”, Applied Physics Letters 100, 044102 (2012) [3] V. A. Godyak, R. B. Piejak and B. M. Alexandrovich, “Electron energy distribution function measurements and plasma parameters in inductively coupled argon plasma” Plasma Sources Sci. Technol.11 525–543(2002) [4] Tsv K. Popov, P. Ivanov, M. Dimitrova ,J. Kovacic, T. Gyergyek and M. Cercek, “Langmuir probe measurements of the electron energy distribution function in magnetized gas discharge plasmas”, Plasma Sources Sci. Technol. 21 025004 (2012) 210 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_136 Characteristics of Dust – Density Waves in the Presence of a Floating Cylindrical Object in the DC Discharge Plasma Mangilal Choudhary1, Subroto Mukherjee1, Rajaraman Ganesh1, Abhijeet Sen1 1Institute for Plasma Research, India Email: [email protected] Dusty plasma provides a unique opportunity to study a variety of collective modes. One such collective mode is dust – acoustic wave (DAW) [1-3].In experiments, this low frequency mode is self or externally excited below a critical gas pressure. In most D. C. discharges, dust particles are trapped in the anodic plasma where self – excited waves are a result of ion – streaming instability [4]. Our experimental studies are on the self – excited non – linear dust density waves in a dust cloud trapped in an elliptical potential well of the cathode sheath (D.C. Discharge initiated dusty plasma). Since, ions are streaming towards the cathode they pass through the dust cloud and exert a drag force on the dust particles as long as velocity ui uthi where ui , uthi are ions streaming velocity and ion thermal speeds respectively. This ion – dust streaming instability is the main free energy source for non – linear dust – density waves. The wave propagates in the direction of ion flow and gravity. In our experimental studies, we have observed different characteristics of the dust medium when it is perturbed by a floating cylindrical object with r >> λd where r, λd are radius of cylinder and Debye length respectively. For lower discharge voltages, a void is formed around a vertically oriented floating cylindrical object. For higher discharge voltages, the propagation characteristics of dust-density waves get significantly changed when a floating cylindrical object is placed near the upper side of dust cloud. Nonlinear dust- density wave propagation give rise to interesting and novel wave pattern which are explained on the basis of the modified equipotential surfaces created by the cylindrical floating object. References: [1] N. Rao, P. K. Shukla, and M. Y. Yu. Planet Space Sci. 38, 543 (1990) [2] A. Barkan, R. L. Merlino, and N. D’Angelo, Phys. Plasmas. 2, 3563 (1995) [3] C. Thompson, A. Barkan, N. D’Angelo, and R. L. Merlino Phys. Plasmas, 4 (7), 2331 (1997 [4] E. Thomas, Jr. and R. L. Merlino, IEEE Trans. Plasma Sci. 29, 152 (2001). 211 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_140 Investigation of Magnetic Drift on Transport of Plasma across Magnetic Field Parismita Hazarika1, Bidyut Das2, Monojit Chakraborty1, Mainak Bandyopadhyay2 1Centre of Plasma Physics-Institute for Plasma Research, India 2 ITER-India, Institute for Plasma Research, India Email: [email protected] When a metallic body is inserted inside plasma chamber it is always associated with sheath which depends on plasma and wall condition. The effect of sheath formed in the magnetic drift and magnetic field direction on cross field plasma transport has been investigated in a double Plasma device (DPD) .The drifts exist inside the chamber in the transverse magnetic field (TMF) region in a direction perpendicular to both magnetic field direction and axis of the DPD chamber. The sheath are formed in the magnetic drift direction in the experimental chamber is due to the insertion of two metallic plates in these directions and in the magnetic field direction sheath is formed at the surface of the TMF channels. These metallic plates are inserted in order to obstruct the magnetic drift so that we can minimised the loss of plasma along drift direction and density in the target region is expected to increased due to the obstruction. It ultimately improves the negative ion formation parameters. The formation of sheath in the transverse magnetic field region is studied by applying electric field both parallel and antiparallel to drift direction. Data are acquired by Langmuir probe in source and target region of our chamber. 212 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_141 High Intensity High Contrast Femtosecond Laser Absorption in Solid Kamalesh Jana1, Amitava Adak1, Moniruzzaman Shaikh1, Deep Sarkar1, Indranuj Dey1, Amit D Lad1, G Ravindra Kumar1 1Tata Institute of Fundamental Research, India Email: kamales. jana @ tif.res.in Several mechanisms of high intensity short pulsed laser absorption by solids have been explored with numerical simulations, analytical works and experimental studies. We investigate high contrast femtosecond laser absorption in a polished fused silica target at near relativistic laser intensities. Absorption measurements are performed for p- and s-polarized laser light and as a function of the incident laser energy and the angle of incidence. Results show absorptivity for p- polarized laser increases with angle of incidence up to ~ 65˚ and beyond this angle it starts decreasing. But for s-polarized laser absorptivity decreases with angle of incidence up to ~ 55˚ and beyond that it remains almost constant. Also it is observed that, separation between absorption curves (for p and s-polarized laser) increases with angle of incidence for all incident laser energies. Vacuum heating process is found to be dominant at large oblique incidence. 213 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_153 Lithium Vapor Density Diagnostics for the PWFA Plasma Source at IPR Mohandas Kizhupadathu Krishnan1, Sivakumaran Valluvadasan1, Sneha Singh1, Ravi A V Kumar1 1Institute for Plasma Research, India Email: [email protected] A photo-ionized Lithium vapor plasma source for Plasma Wakefield Acceleration (PWFA) experiment at Institute for Plasma Research (IPR), Gujarat has been developed as part of the ongoing Accelerator Programme. The plasma source is a 40 cm long Li vapor based heat pipe oven photo-ionized by a UV laser (193 nm) to produce a uniform column of Li plasma. Li vapor in the oven is produced by heating solid Li in helium buffer gas. In PWFA experiment, an accurate measurement of Li vapor density is important as it has got a direct consequence on the electron density of the plasma formed by single photon ionization. Three different optical diagnostics (White light absorption, UV absorption and Hooks method) have been employed in the present experiment to measure the Li neutral column density in the plasma source. The characterization and optimization studies of the Li vapor column formed in the oven have been carried out using these different optical diagnostics as a function of external oven temperature and the He buffer gas pressure. Here, we present the comparative study of the three different measurements carried out for the estimation of the line integrated Lithium vapor density. 214 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_157 Turbulent, Megagauss Magnetic Fields in Intense, Ultrashort Laser Pulse Interaction with Solids Amit D Lad1, Gourab Chatterjee1, Kevin Schoffler2, Prashant Singh1, Sudip Sengupta3, Predhiman Krishan Kaw3, Luis Silva2, Amita Das3, G Ravindra Kumar1 1Tata Institute of Fundamental Research, India 2Instituto Superior Tecnico, Universidade De Lisboa 3Institute for Plasma Research, India Email: [email protected] Intense laser-plasma interactions provide a novel and fascinating platform to simulate astrophysical scenarios [1]. Giant magnetic fields (102 – 103 megagauss) are created when a relativistic intensity >1018 W/cm^2, ultrashort laser pulse interacts with plasma created on a solid. Here we present snapshots of these megagauss magnetic fields, capturing their picosecond- scale evolution with micron-precision. The plasma created by an 800 nm laser is probed at density of ~1022 electrons/cc at 266 nm. This density is so far the highest at which plasma probing has been performed. The Fourier spectrum of these megagauss magnetic fields shows a power-law behaviour for the magnetic energy, which is provides the signature of magnetic turbulence [2]. Detailed particle-in-cell simulations have shown that the relativistic hot electron transport in a hot dense laser-generated plasma suffers from several instabilities including the Weibel instability [3], which leads to the spatial separation of forward and return currents and eventually lead to the filamentary structure. The currents subsequently get Weibel-separated, followed by the tearing and coalescence instabilities, which produce current channels and thereby filamentary magnetic field structures. These results are fundamentally interesting in the context of fast ignition of laser fusion [1], laser-based acceleration of protons, ions and neutral particles [4], the feasibility of experimentally verifying such instability mechanisms in astrophysical magnetic fields [1], mimic observations of kinetic Alfven wave turbulence in the earth’s magneto-sheath, solar flares and solar wind and simulating intra-planetary matter existing at ultrahigh pressures. References: [1] R. P. Drake, High-energy-density Physics-fundamentals, inertial fusion and experimental astrophysics (Springer, Berlin, Heidelberg) (2006). [2] S. Mondal et al., “Direct observation of turbulent magnetic fields in hot, dense laser produced plasmas”, Proc. Natl. Acad. Sci. USA 109, 8011 (2012). [3] E. S. Weibel, “Spontaneously growing transverse waves in a plasma due to an anisotropic velocity distribution”, Phys. Rev. Lett. 2, 83 (1959). [4] R. Rajeev et al., “A compact laser-driven plasma accelerator for megaelectronvolt-energy neutral atoms”, Nature Phys. 9, 185 (2013). 215 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_158 Design and Characterization of Cesium Oven for a Multi-cusp Plasma Device Meenakshee Sharma1, Amitkumar D Patel1, Narayan Ramasubramanian1 1Institute for Plasma Research, India Email: [email protected] In the Multi-cusp Plasma Device, contact ionized Cesium plasma will be confined by a multi- cusp magnetic field. Since magnetic field at the axis of the device is near to zero, the drift waves based fluctuations (as present in classical Q-machine [1]) are expected to be not present. This may help to record the real thermodynamic fluctuations. A hot oven feeding well collimated Cesium beam maintaining an uniform vapor pressure into a high vacuum (10-7 mbar) region is designed for this device. In this paper a brief description of the design of the Cesium oven including critical components and its characterization will be discussed. References: [1] Q-machines, Robert W. Motely, Academic Press, 1975. 216 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_160 Korteweg-de Vries-Burger (KdVB) Equation in a Five Component Cometary Plasma with Kappa Described Electrons and Ions Manesh Michael1, Sreekala G1, Sijo Sebastian1, Neethu Jayakumar1, Anu Varghese1, Neethu Theresa Willington2, Venugopal Chandu1 1Mahatma Gandhi University, India 2C. M. S. College Kottayam, India Email: [email protected] We investigate the existence of Ion-Acoustic solitary waves in a five component cometary plasma consisting of positively and negatively charged oxygen ions, kappa described hydrogen ions, hot solar electrons and slightly colder cometary electrons. The KdVB equation is derived for the system and its solution plotted for different kappa values, oxygen ion densities, as well as for the temperature ratios of ions. It is found that the amplitude of solitary wave decreases with increasing kappa values. While it increases with increasing temperature of positively charged oxygen ions and density of negatively charged oxygen ions. 217 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_165 Two Dimensional Imaging of Laser Produced Plasma in Magnetic field Narayan Behera1, Rajesh Kumar Singh1, Ajai Kumar1 1Institute for Plasma Research, India Email: [email protected] A new experimental set which consist of pulse magnetic field system has been developed for two dimensional imaging of laser produced plasma across the transverse magnetic field. A pair of coils coupled with capacitor bank system is used to generate uniform magnetic field varying from 0-0.8 T magnetic. The coils, target and ablation geometry are set in such a way that it facilitate the plume imaging in both across and along the magnetic field lines. Internally synchronized two ICCD cameras, mounted in orthogonal direction have been used to capture the temporal evolution of expending plasma plume. The design, optimization and performance of the above system will discuss in detail. Apart from the technical aspect of the experimental setup, test results related to effect of magnetic field on the geometrical aspect of the expanding plasma across as well as along the magnetic field will discuss briefly. 218 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_172 The Effect of Addition of Lighter Ions in a Five Component Multi-Ion Plasma Sijo Sebastian1, Manesh Michael1, Sreekala G1, Neethu Theresa Willington2, Anu Varghese1, Chandu Venugopal1 1Mahatma Gandhi University, India 2C. M. S. College Kottayam, India Email: [email protected] We investigate the effect of adding another light ion component on solitary waves in a five component plasma consisting of pair ions, electrons of solar and cometary origin and hydrogen ions. Both the electron components are modeled by kappa distribution function. The Zakharov- Kuznetzov (ZK) equation is derived and solutions plotted for different physical variables relevant to comet Halley. From the plots, it is seen that the addition of another lighter ion component has a significant effect on both the width and polarity of the solitary waves. 219 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_182 Effect of Ablation Geometry on the Formation of Stagnation Layer in Laterally Colliding Plasmas Alamgir Mondal1, Rajesh K Singh1, Ajai Kumar1 1Institute for Plasma Research, India Email: [email protected] Interaction between two parallel propagating plasma plumes have been investigated in two different ablation schemes e.g. laser-blow-off (LBO) of thin film and conventional laser ablation (LPP). Fast imagine technique is used to study the dynamical and geometrical aspect of seed plasmas and induced stagnation layer in between the two expanding seed plasmas. Interaction between the energetic particles, coming from the seed plasmas are responsible for formation of stagnation layer. It has been found that geometrical shape, size, kinetic energy and divergence of plasma plumes are highly dependent on the ablation geometry. These variations in seed plasmas initiate the significant differences in the stagnation layer formed by LBO and LPP geometry. In this presentation, characteristic feature of stagnation layer which includes density, initiation time, emissive life time and geometry in both LBO and LPP geometry are briefly discussed. A comparative study of present results suggests that the plume composition and directionality of seed plasma play crucial role in mechanistic aspect of stagnation layer. 220 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_183 Enhanced Confinement by Controlling Instability in Toroidal Electron Plasma of SMARTEX-C Lavkesh Lachhvani1, Sambaran Pahari2, Manu Bajpai1, Prabal K Chattopadhyay1, Yogesh Yeole1 1Institute for Plasma Research, India 2Bhabha Atomic Research Centre-Visakhapatnam, India Email: [email protected] Experiments have been carried out on a toroidal non-neutral plasma in a tight aspect ratio partial torus SMARTEX-C[1]. Different triggering mechanisms for diocotron instability are identified i.e. presence of ions[2], neutrals and finite wall impedance[3]. Respective scaling of growth rates and their effect on confinement has been investigated. Controlling transport triggered by the instability results in enhanced confinement of electron plasma for ~ few 100 ms. While confinement may presently appear to be challenged by the magnetic pumping transport[4], experiments in SMARTEX-C, reported in this paper, explores the possibility of overcoming this theoretical limit. References: [1] S. Pahari, H. S. Ramachandran, and P. I. John, Phys. Plasmas, vol. 13, no. 9, p. 092111, 2006. [2] R. H. Levy, J. D. Daugherty, and O. Buneman, Physics of Fluids, vol. 12, no. 12, pp. 2616–2629, Dec. 1969. [3] W. D. White, J. H. Malmberg, and C. F. Driscoll, Phys. Rev. Lett., vol. 49, no. 25, pp. 1822– 1826, Dec. 1982. [4] S. M. Crooks and T. M. O’Neil, Physics of Plasmas, vol. 3, no. 7, pp. 2533–2537, Jul. 1996. 221 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_185 Study of Phase Space Structures in Driven 1D Vlasov Poisson Model Pallavi Trivedi1, Rajaraman Ganesh1 1Institute for Plasma Research, India Email: [email protected] Electrostatic waves in a collisionless, unmagnetized plasma are known to interact with particles that stream with velocities close to the wave phase speed to produce damping effects, particle trapping and interesting nonlinear coherent structures [1, 2]. For example, it is well known that if the initial amplitude of the wave is large enough, the damping effects can be overcome to form BGK structures. In the present work, we consider a 1D driven Vlasov-Poisson plasma model. It is demonstrated that by a careful choice of drive phase and for drive amplitudes smaller than or comparable to the linear limit, it is possible to generate surprisingly large amplitude coherent structures in phase space [3]. This and other details will be presented. References: [1] G. Manfredi, Physical Review Letters 79, 2815 (1997) [2] M. Raghunathan and R. Ganesh, Physics of Plasmas 20, 032106 (2013) [3] Pallavi Trivedi and Rajaraman Ganesh (Manuscript under preparation 2015) 222 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_186 Synchronization dynamics and Arnold tongues for two coupled glow discharge plasma sources Neeraj Chaubey1, Subroto Mukherjee1, A N Sekar Iyengar2, Abhijeet Sen1 1Institute for Plasma Research, India 2Saha Institute of Nuclear Physics, India Email: [email protected] Two DC glow discharge plasma sources, whose cathode and anode diameters were 70 mm and 2 mm respectively, and operating at a neutral pressure of 0.1 mbar, have been deployed for synchronization experiments. In each of the chambers, a Langmuir probe was placed for measuring floating potential fluctuations. Plasma was produced in both the chambers and floating potential frequencies were monitored. It has been observed that floating potential fluctuation frequencies in both the chambers were increasing linearly with increase in the discharge voltage. For the synchronization experiment we have kept fixed the discharge voltage of one of the systems to an oscillation frequency (f) and coupled the oscillation of frequencies of the other system with an increase in the discharge voltage. Nonlinear phenomenon like frequency entrained states were observed when oscillation frequencies of two coupled systems were close to the harmonic frequencies like f/2, 2f, 3f and 4f and far from this region frequency pulling and chaos were observed. Experimental results of frequency entrainment, frequency pulling and chaos have produced a very nice picture of the Arnold tongue between the two coupled glow discharge plasma systems. Some region of the experimental results was modeled by numerical simulation of two coupled asymmetric Van der Pol type equations and these results were found to be in good agreement. References: [1] Neeraj Chaubey et. al. Phys. Plasmas 22, 022312 (2015) [2] T. Fukuyama, Y. Watanabe, K. Taniguchi, PRE 74, 016401 [3] T. Fukuyama et al., PRL 96, 024101 (2006) [4] Catalin M. Ticos, PRL, Volume 85, Number 14, 2929 (2000) [5] S. Boccaletti et al. / Physics Reports 366 (2002 [6] Sync: How Order Emerges From Chaos In the Universe by S. Strogatz [7]B. E. Keen and W. H. W. Fletcher, PRL,Vol.-23, Number-14(1968) [8]T. Gyergyek et.al. Contrib. Plasma Phys. 37 (1997) 5, 399-416 [9]T. Klinger, et al. Physical Review E Volume 52, Number 4 (1995) 223 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_194 Optical Kerr Gated Time Resolved Cherenkov Emission Produced during Ultra Intense Laser Solid Interaction Moniruzzaman Shaikh1, Amit Lad1, Deep Sarkar1, Sheroy Tata, Indranuj Dey, Gourab Chatterjee1, Rajeev P P2 , G Ravindra Kumar1 1Tata Institute of Fundamental Research, India 2Rutherford Appleton Laboratory, United Kingdom Email: [email protected] We present the optical Kerr gated time resolved Cherenkov emission emitted by fast electrons produced in intense femtosecond laser-solid interaction. The fast electrons are produced in thick (3.0-10.0 mm) BK-7 glass targets irradiated by infrared laser pulse of intensity ≈ 5 × 1E19 W cm-2. The optically polished targets are coated with 100 nm aluminium on the irradiated side. It is a big challenge to measure the time dynamics of a physical phenomenon which is lasting only for a few picoseconds. Even the best electronic time window is hundreds of picosecond in width. To measure Cherenkov emission we have replaced intensified CCD electronic gate of hundreds of picoseconds in width with a gate created by optical Kerr effect of a width of two picoseconds. We have observed for the first time that the Cherenkov emission is lasting for more than tens of picoseconds inside the thick target. This finding offers crucial information the fast electron transport dynamics. 224 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_199 Imaging of Terahertz Emission from Intense High-Contrast Ultrashort- Pulse Laser-Solid Interaction Indranuj Dey1, Deep Sarkar1, Moniruzzaman Shaikh1, Sheroy Tata1, Amit D Lad1, G Ravindra Kumar1 1Tata Institute of Fundamental Research, India Email: [email protected] 18 19 2 -9 When an intense (Io ~ 10 - 10 W/cm , ~ 800 nm), high-contrast (pedestal/peak ~ 10 ), ultrashort pulse (td ~ 30 fs, rep. rate 10 Hz) is obliquely focused on to a solid surface, a very 22 -3 high density plasma (ne ~ 10 m ) is generated within a very small volume (diameter ~ 10 - 15 m). The high density plasma furnishes hot electrons and energetic ions, accompanied by various radiation processes like bremsstrahlung, second-harmonic generation, two-plasmon decay, Cherenkov, and plasma emission [1–3]. The energetic charged particle dynamics and the various emissions have been quite well studied over the years [3]. However, the emission of low frequency radiations from such laser-solid interactions, especially in the Terahertz (THz) regime has only received attention over the last few years [4, 5]. The intense radiation from such interaction can be used to study the hot electron dynamics resulting from the laser-plasma interaction, and also as a probe for the emerging field of non-linear spectroscopy. In this work, we intend to image THz emission from the interaction region, to study the evolution of the non-linear current induced on the surface of the solid, which leads to the generation of THz. Since the repetition rate of the laser is low, the standard techniques of time domain spectroscopy and imaging used in low power kHz repetition rate lasers would not be effective here [6], due to large shot to shot fluctuations and limited target area. In contrast, the THz radiation would be directly imaged by a THz camera. A technique similar to the Kerr gating would be employed to study the time evolution of the THz from the interaction region. It is expected that the THz emission would furnish information regarding the electron dynamics on the surface of the solid target. References: [1] L. Gizzi, et al. Simultaneous measurements of hard x-rays and second harmonic emission in fs laser-target interactions. Phys. Rev. Lett. 76, 2278 (1996). [2] L. Gremillet, et al. Time-Resolved Observation of Ultrahigh Intensity Laser-Produced Electron Jets Propagating through Transparent Solid Targets. Phys. Rev. Lett. 83, 5015 (1999). [3] D. Umstadter, Review of physics and applications of relativistic plasmas driven by ultra-intense lasers. Phys. Plasmas 8, 1774 (2001). [4] Y. T. Li, et al. Strong terahertz radiation from relativistic laser interaction with solid density plasmas. Appl. Phys. Lett. 100, 1 (2012). [5] W. J. Ding, et al. High-field half-cycle terahertz radiation from relativistic laser interaction with thin solid targets. Appl. Phys. Lett. 103, 2011 (2013). [6] Q. Wu, et al. Two-dimensional electro-optic imaging of THz beams. Appl. Phys. Lett. 69, 1026 (1996). 225 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_218 Pulsed Plasma for the Study of Coherent Structure in the Electron Magnetohydrodyanamic Regime Garima Joshi1,2, Ravi Ganesh1, Subroto Mukherjee1 1FCIPT-Institute for Plasma Research, India 2Nirma University, India Email: [email protected] In any study involving cold plasma waves in the laboratory, the general requirement is of a uniform, quiescent and collision-less plasma devoid of any instabilities. The conditions of plasma extent and uniformity impose additional constraints when a study of wave phenomena in the context of electron magneto hydrodynamics (EMHD) is involved [1]. In order to investigate 11 12 -3 phenomena in this regime, the plasma conditions required are approximately ne ~10 – 10 cm , Te ~ 1 – 2 eV & extent of uniformity is orders of several wavelengths. Many experimental devices have been developed in the past by researchers [2-3] for the study of EMHD phenomena; however the focus has been mainly on whistler waves and accompanying physics. In our laboratory, we have developed a new device for the exploration of nonlinear EMHD vortices, coherent structures which have been predicted and studied extensively in theoretical works [4] but unexplored in experiments. The most important element of the device is the plasma source that we have built in-house. It is a multi-filamentary type plasma cathode source [5] coupled with a broken line cusp arrangement that prevents the loss of primary electrons and helps enhance the plasma density. Our source is capable of producing a uniform, quiescent (δn/n ≈1%), low temperature (~1 - 2 eV) pulsed plasma of moderately high density (1018 m-3) in the main glow and (1017 m-3) in the afterglow regime. The plasma is pulsed so that streaming primary electrons from the filament are cut off & bulk electron temperature is reduced to (~1 - 2 eV). Plasma decays slowly due to the presence of external magnetic field and forms a region of uniform density ~1017 m-3 in afterglow over a large area 2.5 m2 plasma & therefore well suitable to study wave-plasma phenomena. References: [1] A. S. Kingsep, K. V. Chukbar, and V. V. Yankov, “Electron Magnetohydrodynamics”. Reviews of Plasma Physics _ Consultants Bureau, New York, Vol. 16 (1990) [2] W. Gekelman, H. Pfister, Z. Lucky, J. Bamber, D. Leneman, and J. Maggs, “Design, Construction, and properties of the large plasma research device−The LAPD at UCLA”, Rev. Sci. Instrum. 62, 2875 (1991). [3] S. K. Mattoo, V. P. Anitha, L. M. Awasthi, G. Ravi, and LVPD Team, “A large volume plasma device”, Rev. Sci. Instrum., 72, 3864 (2001) [4] M. B. Isichenko and A. M. Marnachev, “Nonlinear wave solution of electron MHD in uniform plasma”, Sov. Phys. JETP, 66, 702 (1987). [5] L M Awasthi, G Ravi, V P Anitha, P K Srivastava and S K Mattoo, “A large area Multifilamentary plasma source”, Plasma Sources Science and Technology, 12, 2, p158 (2003) 226 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_228 Chaos to Order Transitions in Chaotic Magnetic Fields A N Sekar Iyengar1, M S Janaki1, Pankaj Kumar Shaw1, Subha Samanta1 1Saha Institute of Nuclear Physics, India Email: [email protected] The study of the structure of magnetic fields including formation of magnetic surfaces as well as field line chaos is of much help in understanding the problems of plasma confinement and instabilities in the context of fusion devices. Chaotic magnetic fields in astrophysical environments and fusion physics have been directly and indirectly postulated by many earlier workers. Parker [1] was the first to point out the effect of irregular and chaotic magnetic field on charged particle motion in cosmic plasma. These works were subsequently elaborated in a series of papers by Jokipii [2] to study the cosmic ray propagation in a random magnetic field. Lee and Parks [3] studied the evolution of nonlinear magnetic field in MHD plasmas by casting these equations in the form of a forced Duffing’s equation which showed chaotic behavior. In fusion physics, existence of chaotic magnetic fields has been conjectured by several authors due to its relevance to enhanced heat transport. Recently, the ubiquity of the chaotic magnetic fields in an asymmetric combination of current carrying wire loop system has been demonstrated [4]. Following kinetic treatment[5], equilibrium magnetic fields in current carrying systems have been recently shown to be governed by a Yang-Mills-Higgs type equation giving rise to a coupling of x and y components of the magnetic fields with the variation in the z-direction. Interestingly, it was shown in that paper that for a given value of the coupling parameter, above certain energy the system is always chaotic. Recently we carried out a numerical solution of the same equation and some very interesting results were obtained as a function of both the initial conditions and coupling parameter. It was observed that the system is periodic for values of the coupling parameter given by 2, 4, 8 and... 2n. Together with the initial condition have to be equal, i.e x(0)=y(0), and xdot(0)=ydot(0). If the above condition is not satisfied the solution only yields chaotic solutions. The Fourier spectral analysis shows that as the mutual coupling parameter is increased the peak frequency shifts towards higher frequencies, but for the chaotic data it exhibits a broad spectrum which is confined to a certain band. This study may lead to an understanding of the phenomena responsible for particle acceleration in space plasmas, fusion plasmas etc. and also in other areas of physics. References: [1] E. N. Parker, J. Geophys. Res. 69, 1755, (1964). [2] J. R. Jokipii, Astrophys. J. 146, 480 (1966). [3] N. C. Lee and G. K. Parks, Theory, Geophys. Res. Lett. 19, 637 (1992). [4] A. K. Ram and B. Dasgupta, Phys. Plasmas 17, 122104 (2010). [5] Abhijit Ghosh, M.S. Janaki, B. Dasgupta and A. Banerjee, Chaos, 24, 013117 (2014). 227 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_229 Investigation of Coherent Modes in the Chaotic Fluctuation in the SINP Tokamak Pankaj Kumar Shaw1, Debajyoti Saha1, Sabuj Ghosh1, Subha Samanta1, M S Janaki1, A N Sekar Iyengar1 1Saha Institute of Nuclear Physics, India Email: [email protected] The study of coherent structures in plasma turbulence has been of great interest in view of their importance in the transport of momentum and energy [1]. In this paper, we have used empirical mode decomposition method, which resolves the signal into modes of various time scales called intrinsic mode functions, for the identification of coherent structures in a chaotic time series [2]. The estimation of the log-variance and correlation coefficients of the intrinsic mode functions helps in identifying the coherent modes in the chaotic time series. By this technique, coherent modes were detected in the chaotic floating potential fluctuations, obtained from a glow discharge plasma device, as test cases. Then we applied this to our chaotic fluctuations obtained from the low qa discharges of the SINP tokamak [3]. We also carried out a bicoherency analysis on the coherent modes extracted using empirical mode decomposition to detect the interactions amongst them. References: [1] Marie Farge, Kai Schneider and Pascal Devynck, “Extraction of coherent bursts from turbulent edge plasma in magnetic fusion devices using orthogonal wavelets,” Physics of Plasmas, 13, 042304 (2006). [2] Pankaj Kumar Shaw, D. Saha, S. Ghosh, M.S.Janaki, A.N.SekarIyengar, “Investigation of coherent modes in the chaotic time series using empirical mode decomposition and discrete wavelet transform analysis,” Chaos, Solitons and Fractals, 78, 285 (2015). [3] S. Lahiri, A. N. S. Iyengar, S. Mukhopadhyay, R. Pal,"Investigation of low qa discharges in the SINP tokamak," Pramana, 58, 79 (2002). 228 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_233 Study of Defects in Externally Driven Dust Density Waves in Cogenerated Dusty Plasma using Time Resolved Hilbert-Huang Transform Sanjib Sarkar1, Mridul Bose2, Subroto Mukherjee1 1FCIPT-Institute for Plasma Research, India 2Department of Physics, Jadavpur University, Kolkata, India Email: [email protected] Spatiotemporal study of defects in positively biased electrode induced dust density wave (DDW) in cogenerated dusty plasma is reported. DDW is excited for threshold positive bias through another electrode which is placed in between two main discharge electrodes. Spatiotemporal evolution of DDW reveals wave defect and non-propagating wave mode in the DDW field. Space-time plot and time resolved Hilbert-Huang transform (HHT) was employed to analyze the spatiotemporal wave data. 229 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_238 Efficient Hard X-ray Generation in an Interaction of Intense, Ultrashort Laser with Metal Nano-coated Dielectric Target Deep Sarkar1, Amitava Adak1, Moniruzzaman Shaikh1, Indranuj Dey1, Amit D Lad1, G Ravindra Kumar1 1Tata Institute of Fundamental Research, India Email: [email protected] Hard x-ray emission in intense laser-matter interaction studies is a topic of great interest due in significant part to its various applications [1].We measure the hard x-ray yield from Ag nano- coated thick BK-7 glass target interacting with an intense femtosecond laser and compare the results with those from an uncoated BK-7 target. The enhancement in integrated hard x-ray yield is measured as a function of thickness of Ag nano-coating which was varied from tens of nanometer to hundreds of nanometer. The effect of laser polarization on hard x-ray yield is studied. Maximum enhancement (20x) is observed for a coating thickness of 35 nm for a p- polarized pump laser of relativistic intensity (~1019 W/cm2). For the coating thicknesses of more than 100 nm, the x-ray enhancement factor is found to be flat. The x-ray yield from uncoated BK-7 target is found to be the same for the two polarizations of the pump laser. Additionally, it is observed that the X-ray enhancement for coating thickness of 42 nm is greater for the p- polarized pump laser as compared to that for s-polarized pump laser. We compare our results with those from earlier studies [2, 3] and discuss the implications. Fig.1. X-ray enhancement factor as a function of thickness of silver coating. Blue line corresponds to enhancement factor of unity. References: [1] T. Pfeifer et.al, Rep. Prog. Phys. 69 443-505 (2006). [2] P.P. Rajeev et al., Phys. Rev. Lett. 90, 115002 (2003). [3] S. Mondal et al., Phys. Rev. B 83, 035408 (2011). 230 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_244 Laser Heated Emissive Probe for Plasma Potential Measurement in Fusion Plasmas Vara Prasad Kella1, Payal Mehta2, Joydeep Ghosh1, Arun Sarma2 1Institute for Plasma Research, India 2Venus International College of Technology, India Email: [email protected] Emissive probes are the successful tools in measuring plasma potentials in various conditions up to much better accuracy. Plasma potential structures study in tokomak edge region has much importance, which gives valuable information on the ion flow rates and energy loss to the walls. Conventional Emissive probes (Filament) has some limitations in this scenario, owing to their short life time, effect of filament heating current in external magnetic fields. Laser heated emissive probe (LHEP) address solution to these problems and produce better results [1, 2]. In this present work, we demonstrated LHEP with measuring sheath potential profile in low temperature filament discharge, which gives better agreement with the theoretical potential profile and sheath thickness estimation. Low work function LaB6 material has been used as probe material for this study and CW CO2 LASER of wavelength 10.6 μm and maximum power 55 watt used for heating LHEP [3]. References: [1] Roman Schrittwieser et al, “Laser-heated emissive plasma probe”, Review of scientific instruments, 79, 083508 (2008). [2] Roman Schrittwieser et al, “A Radially Movable Laser-Heated Emissive Probe”, J. Plasma Fusion Res. SERIES, Vol. 8 (2009). [3] Payal Mehta et al, “Measurement of emission current and temperature profile of emissive probe materials using CO2 LASER”, Current Applied Physics, 11,1215-1221,(2011). 231 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_246 Study of Fluctuation Induced Particle Flux in the Background of ETG plasma in LVPD Prabahkar Srivastav1, Lalit Awasthi1, Amulya Kumar Sanyasi1, Pankaj Srivastava1, Ratneshwar Jha1, Raghvendra Singh2, Predhiman Krishan Kaw1 1Institute for Plasma Research, India 2WCI Center for Fusion Theory, National Fusion Research Institute, Korea Email: [email protected] Theoretical and numerical investigations on plasma transport in fusion devices suggest that turbulent electron thermal transport due to ETG turbulence is probably a major source for plasma loss in fusion devices. Direct identification of this instability in fusion devices is a very difficult task because of its extremely small-scale length but indirect inferences support the theoretical work [1, 2]. Unambiguous excitation of Electron Temperature Gradient (ETG) turbulence is demonstrated in the steady state, collision less Argon plasma of Large Volume Plasma Device (LVPD) [3] because of measurable length-scale and time-scale. In this paper, we report the fluctuation-induced particle transport in ETG unstable regime of the LVPD. In LVPD, investigations on fluctuation induced plasma transport is carried out for two different plasmas namely, 1) when ETG turbulence is present i.e., = Ln/LT > 2/3, where Ln and LT are the gradient scale lengths of plasma density and electron temperature and 2) when ETG is absent. We have measured particle flux by measuring the fluctuations of plasma density and plasma potential, making use of a 6- pin probe assembly. Poloidally separated pair of emissive probes is used to measure the potential fluctuations as finite temperature fluctuations (Te/Te ~ 13 %) are present in the background plasma. The time-averaged flux is calculated in both scenarios and initial results do indicate that the direction of flux is radially inward in an ETG dominated regime and the ETG absent regime has radially outward particle flux. The calculated Probability Distribution Function (PDF) for the particle flux, density and potential fluctuations is found to be non- Gaussian. It is leptokurtic with its peak at the centre and fatter wings [4]. The inward particle flux in the background of ETG is an interesting observation and detailed experimental and theoretical discussion on it will be presented at the conference. References: [1] Y. C. Lee, J. Q. Dong and P. N. Guzdar, Phys. Plasmas 30, 1331(1987). [2] P. N. Guzdar, C. S. Liu, J. Q. Dong et al., Phys. Rev. Lett. 57, 2818 (1986). [3] S. K. Mattoo, S. K. Singh, L. M. Awasthi et al., Phys. Rev. Lett. 108, 255007 (2012). [4] R. Jha, P. K. Kaw, S. K. Mattoo et al., Phys. Rev. Lett. 69, 1375 (1992). 232 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_252 Exhibiting Electrons in Nanoplasmas: An Estimate Jagannath Jha1 1Tata Institute of Fundamental Research, India Email: [email protected] Using the method of ion kinetic energy spectrometry, we measure the degree of outer ionization in cluster nanoplasma. We show that the degree of outer ionisation in Ar7000 clusters nearly doubles if the intensity is increased from 1014 W/cm2 to 1015 W/cm2. Molecular dynamics simulation is used to infer the degree of outer ionization and is found to compare well with the experimental measurements. 233 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_261 High Energy Neutral Atoms from High Intensity Laser Plasma Interaction Sheroy Tata1, Malay Dalui1, T Madhu Trivikram1, Jagannath Jha1, M Krishnamurthy1 1Tata Institute of Fundamental Research, India Email: [email protected] Interaction of a high intensity laser with solid targets leads to acceleration of ions from the surface of the target [1, 2]. Ion acceleration is governed by electron dynamics at the target vacuum interface setting up a charge separation. This electron cloud near the target interface can also provide a neutralizing background for ions that have been accelerated. The accelerated ions are thus detected as a high energy neutral atom on a detector. Further, due to the inherent contrast profile of high intensity lasers a pre-plasma is almost always formed and neutral atoms can be detected. The ion and neutral atom energies are measured by a Thomson parabola spectrometer coupled with a ‘time of flight’ measurement. The neutral atom energies are obtained from the time of flight. The TIFR 20TW laser with an intensity contrast 10-5 was used to carry out the experiment. Defocusing the target led to a 2 fold increase in the neutral atom yield suggesting the role of the pre-plasma. Using a high contrast laser we attempt to tune the recombination dynamics for efficient neutralization of ions by using a controlled pre-plasma. References: [1] Wilks, S. et al., “Energetic proton generation in ultra-intense laser-solid interactions”, Phys. Plasmas 8, 542 (2001). [2] M. Hegelich et. al., “MeV Ion Jets from Short-Pulse-Laser Interaction with Thin Foils”, Phys. Rev. Lett. 89, 085002 (2002). 234 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_262 Role of Magnetic Cusp for Multiple Axial Potential Structures (MAPS) Formation Soumen Ghosh1, Pabal K Chattopadhyay1, Joydeep Ghosh1, Dhiraj Bora1 1Institute for Plasma Research, India Email: [email protected] Cusp like magnetic field profile in expanding helicon experimental system is studied for the formation of multiple axial potential structures (MAPS). Double layer like this potential structures formation in this kind of expanding helicon system produces thrusts along the axial direction. Observation of multiple ion beams is an indirect evidence for the formation of multiple double layers like potential structures. However, there is no such direct evidence available to identify the strength and location for the formation of these structures, in magnetic and geometric expanding helicon plasma systems. Transition from single to multiple axial potential structures is observed by varying the magnetic field topology from diverging to cusp. A localized threshold density is required to maintain the steady state potential structure inside the bulk plasma. Cusp like magnetic field profile inside the expansion controls this downstream density rise and beyond the threshold limit of this density rise, the second potential structure is formed. In this presentation, quantitative discussion will be presented to understand the root causes to maintain the critical density for the formation of MAPS and the mechanism responsible for maintaining this density inhomogeneity [1] in these expanding plasmas. References: [1] Soumen Ghosh et. al., “Localized electron heating and density peaking in downstream helicon plasma,” Plasma Sources Sci. Technol. 24, 034011 (2015). 235 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_263 Enhanced Proton Acceleration by Ultrashort Laser Pulse Interaction with Nanostructured Thin Films Angana Mondal1, Malay Dalui1, Sheroy Tata1, Subhrangshu Sarkar1, Jagannath Jha1, Amit Lad1, W -m Wang2, Z m Sheng3, M Krishnamurthy1, P Ayyub1 1Tata Institute of Fundamental Research, India 2Chinese Academy of Sciences, China 3Shanghai Jiao Tong University, China Email: [email protected] Enhancement of local electromagnetic field in nanostructured targets as opposed to plain polished targets has been experimentally observed and studied [1]. This increase in field strength leads to enhanced hot electron generation, which gives rise to highly energetic ions through Target Normal Sheath Acceleration [2]. As the laser energy coupled to the electrons increases, the sheath magnitude is expected to increase, leading to an enhancement in ion acceleration [3]. We investigate energy enhancements in ions generated as a result of intense femtosecond laser interaction with nanostructured thin film targets, comprising 2 µm Ta foil coated with 100-200 nm diameter Ta clusters. The optimum nanoparticle size of 100 nm corresponding to maximum laser energy absorption has been predetermined through PIC simulations. The accelerated ions have been studied using Thompson parabola spectrometer at a laser intensity of 15x10^19 W/cm^2 at the TIFR high contrast 100 TW Ti:Sapphire laser facility. The proton cut-off energy is observed to increase rapidly with increasing cluster density till a saturation is reached. The enhancement in the proton cut-off energy is observed to be three-fold as compared to the proton cut-off energy for unstructured foils. 236 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_276 DAQ System for Low Density Plasma Parameters Measurement Rashmi S Joshi1, Suryakant B Gupta1 1FCIPT-Institute for Plasma Research, India Email: [email protected] In various cases where low density plasmas (number density ranges from 1E4 to 1E6 cm-3) exist for example, basic plasma studies or LEO space environment measurement of plasma parameters becomes very critical. Conventional tip (cylindrical) Langmuir probes often result into unstable measurements in such lower density plasma. Due to larger surface area, a spherical Langmuir probe is used to measure such lower plasma densities. Applying a sweep voltage signal to the probe and measuring current values corresponding to these voltages gives V-I characteristics of plasma which can be plotted on a digital storage oscilloscope. This plot is analyzed for calculating various plasma parameters. The aim of this paper is to measure plasma parameters using a spherical Langmuir probe and indigenously developed DAQ system. DAQ system consists of Keithley source-meter and a host system connected by a GPIB interface. An online plasma parameter diagnostic system is developed for measuring plasma properties for non- thermal plasma in vacuum. An algorithm is developed using LabVIEW platform. V-I characteristics of plasma are plotted with respect to different filament current values and different locations of Langmuir probe with reference to plasma source. V-I characteristics is also plotted for forward and reverse voltage sweep generated programmatically from the source meter. 237 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_0 Numerical Study of Instabilities in Magnetized Inhomogeneous Plasmas Jyoti Chaudhary1 1Manipal University Jaipur, Rajasthan, India Email: [email protected] The continuity and the momentum equation which take into account the ionization constant are formulated for ions and the electrons including the effect of finite temperature of ions along with the ionization effect. Using normal mode analysis along with linear approximation, potential is found from Poisson’s equation neglecting higher order perturbed terms. From Potential equation, dispersion relation is generated which is solved numerically for obtaining the value of ɷ using typical laboratory as well as space plasma parameters. The behavior of growth rate with magnetic field and the propagation angle along with ionization constant has been studied with different plasma oscillation wavelength to Debye length ratio. We observe two types of instability in both the cases. In case of laboratory plasma one of the instability is growing at larger plasma oscillation wavelength and another one at lower wavelength while in the case of space plasma both the instabilities grow only at smaller plasma oscillation wavelength but with different growth rates. All the instabilities has higher growth rate at smaller wave length of oscillations. Effect of finite ion temperature is studied with respect to different electron temperature both in the laboratory as well as in space plasma. 238 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_11 Modeling of Electromagnetic Fields during Plasma Startup in SST-1 Tokamak Amit Kumar Singh1, Indranil Bandyopadhyay2, Srinivasan Radhakrishnana2, SST-1 Team1 1ITER-India, Institute for Plasma Research, India 2Institute for Plasma Research, India Email: [email protected] The time varying currents in the Ohmic transformer in SST-1 tokamak induce large eddy currents in the passive structures like the vacuum vessel and cryostat. Especially since the vacuum vessel and the cryostat are toroidally continuous without breaks in SST-1, this leads to a shielding effect on the flux penetrating the vacuum vessel. This reduces the magnitude of the loop voltage seen by the plasma as also delays its buildup. Also the induced currents alter the null location of magnetic field. Studying the effective loop voltage and magnetic null location during the plasma breakdown and startup is important, as corrective measures may be required in case of an insufficient loop voltage or an improper null. The dynamics of the evolution of the loop voltage and the magnetic null due to the toroidal eddy currents in SST-1 passive structure has been studied in the breakdown phase of SST-1. At the time of the plasma initiation, the Ohmic transformer current is discharged by short-circuiting the central solenoid (CS) coil through a resistance. The flux stored in the CS coil is linked to the plasma region, as also the conductors surrounding the plasma region. The resulting eddy currents flowing in the passive conductors lead to Joule heating losses of the stored flux in the CS coil. The amount of this eddy current and the associated flux loss has to be accurately determined in order to estimate the external loop voltage seen by the plasma required for plasma breakdown and current ramp-up. We have studied the effect of the induced currents on the loop voltage and the magnetic null using a toroidal-filament model. As the vessel and cryostat are conductors with large poloidal cross-section, for the approximation to be valid and results to be accurate, they are broken up into a large number of co-axial toroidal current carrying filaments. The inductance matrix for this large set of toroidal current carrying conductors is calculated using the standard Green functions and the induced currents evaluated by solving a set first order ODEs for the circuit equations. Of course the induced flux of the Ohmic transformer will also generate local non-toroidal eddies around, for example, around the port structures; however they are expected to provide space localized higher order correction to the field due to toroidal components of the induced current and are neglected in this work. The loop voltages calculated on flux loop locations in SST-1 from these circuit simulations match very well with the experimentally measured loop voltage signals, which prove that this simple model indeed works very well. We also investigate the magnetic null evolution, which indicate a gradual inward shift of the null location during the plasma breakdown. 239 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_19 Oscillating Two-stream Instability of a Plasma Wave in Ion-Motion Regime Pinki Yadav1, Devki Nandan Gupta1, Avinash Khare1 1University of Delhi, India Email: [email protected] It is known that the laser interactions with high-density plasmas can excite a large amplitude plasma wave near critical layer [1, 2]. This large amplitude plasma wave may be susceptible for oscillating two-stream instability by exciting a pair of two electrostatic sidebands and a purely growing low-frequency mode. We propose to revisit this study in the time scale of the order of ion-plasma period by incorporating the ion motion in estimation of the growth rate of the instability. The growth of plasma wave strongly modifies due the ion motion and thus the growth rate of the instability is modified in a specific parameter region. The present study shows that there is a narrow parameter space where the oscillating two-stream instability exists in this regime. References: [1] Y. C. Lee and P. K. Kaw, “Temporal electrostatic instabilities in inhomogeneous plasmas,” Phys. Rev. Lett., 32, 4 (1974). [2] D. N. Gupta, Pinki Yadav, D. G. Jang, M. S. Hur, H. Suk, and K. Avinash, “Onset of stimulated Raman scattering of a laser in a plasma in the presence of hot drifting electrons,” Phys. Plasmas, 22, 052101 (2015). 240 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_20 Development of a 3D-3V PIC code to study PSI processes in Tokamak Divertor Region Sayan Adhikari1, Kalyan Sindhu Goswami1 1Centre of Plasma Physics, Institute for Plasma Research, India Email: [email protected] A limited overview of the theoretical understanding as well as PIC simulation of edge plasmas in fusion devices is given. The effect of grazing angle on solid surface (divertor) erosion due to ion sputtering in magnetic fusion devices is studied by a 3D-3V PIC-MCC code. For an oblique magnetic field, there exists a different kind of region in front of the solid surface named as Chodura sheath (CS) [1]. Important factors like ion energy and impact angle for physical sputtering are highlighted. Because of the presence of the surface itself, the ion distribution in front of the wall is generally not Maxwellian [2]. In spite of this even for an unmagnetized case, presence of sheath can modify the ion distribution, which has been found in different numerical simulation and laboratory experiments [3-4]. For magnetized plasmas, the distribution can have several peaks at different energies [5], which brings further complexity in erosion calculation. The dependence of these two parameters on grazing angle is investigated in detail. The code has been written in java and the plots has been generated in VTK based software Paraview developed by Los Alamos National Laboratory. References: [1] S Devaux and G Manfredi, “Magnetized plasma–wall transition—consequences for wall sputtering and erosion,” Plasma Phys. Control. Fusion, 50, (2008) 025009. [2] R. Chodura, “Plasma–wall transition in an oblique magnetic field,” Phys. Fluids 25, 1628 (1982). [3] Chung K-S and Hutchinson I H, “Kinetic theory of ion collection by probing objects in flowing strongly magnetized plasmas,” Phys. Rev. A, 38, (1988) 4721 [4] Tskhakaya D, Eliasson B, Shukla P K and Kuhn S, “On the theory of plasma-wall transition layers,” Phys. Plasmas, 11, (2004) 3945 [5] Devaux S and Manfredi G, “Vlasov simulations of plasma-wall interactions in a magnetized and weakly collisional plasma”, Phys. Plasmas, 13, (2006) 083504 241 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_21 Betatron Radiation from Laser Wakefield Acceleration in a Plasma Channel Devki Nandan Gupta1, Inhyuk Nam2, Hyyong Suk2 1University of Delhi, India 2Gwangju Institute of Science and Technology, South Korea Email: [email protected] Laser wakefield acceleration by a high-power laser pulse and a plasma has attracted lots of attention in recent years as it can generate quasi-monoenergetic high-energy electron beams and may be used for a compact x-ray source on a table-top scale [1-3]. In the laser wakefield acceleration, plasma electrons can be self-injected into the acceleration phase of the wake wave and they are accelerated with an extremely high gradient in the longitudinal direction. In addition to the longitudinal acceleration, the wake wave also gives an ultra-strong focusing force in the transverse direction. As a result, the accelerated electrons execute the betatron oscillations which can produce the betatron radiation. We propose a method to increase the betatron oscillation amplitude by off-axis injection of a laser pulse into a capillary plasma waveguide. The capillary plasma waveguide has been used only for optical guiding and electron acceleration, where the transverse plasma density profile is nearly parabolic. In our work, we found that the betatron oscillation amplitude can be significantly increased by off-axis injection of the laser pulse into the capillary plasma waveguide, which can be utilized for generation of shorter wavelength x-ray radiation. In order to demonstrate the proposed idea for increasing the betatron oscillation amplitude, we performed two-dimensional (2D) particle in-cell (PIC) simulations in addition to analytical studies [4]. References: [1] S. P. D. Mangles, C. D. Murphy, Z. Najmudin, A. G. R. Thomas, J. L. Collier, et. al., “Monoenergetic beams of relativistic electrons from intense laser–plasma interactions,” Nature (London), 431, 538 (2004). [2] A. Rousse, K. Phuoc, R. Shah, A. Pukhov, E. Lefebvre, V. Malka, S. Kiselev, F. Burgy, J. P. Rousseau, D. Umstadter, and D. Hulin, “Production of a keV x-ray beam from synchrotron radiation in relativistic laser-plasma interaction,” Phys. Rev. Lett., 93, 135005 (2004). [3] V. B. Pathak, J. L. Martins, J. Vieira, R. A. Fonseca and L. O. Silva, “Laser wakefield acceleration in corrugated plasma channel,” Proc. 41st EPS Conference on Plasma Physics, Berlin, Germany, p2.110 (2014). [4] S Lee, T H Lee, D N Gupta, H S Uhm and H Suk, “Enhanced betatron oscillations in laser wakefield acceleration by off-axis laser alignment to a capillary plasma waveguide,” Plasma Phys. Control. Fusion 57, 075002 (2015). 242 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_39 Particle in Cell Simulations of Beam Plasma System Chandrasekhar Shukla1, Atul Kumar1, Bhavesh Patel1, Amita Das1, Kartik Patel2 1Institute for Plasma Research, India 2Bhabha Atomic Research Centre, India Email: [email protected] The propagation of relativistic electron beam in dense plasma is studied with the help of Particle in Cell simulations for both 2D and 3D configurations. The background plasma system provides for the return currents balancing the beam current. These two current systems are unstable to Weibel destabilization as a result of which the forward and return currents separate spatially [1]. This leads to the generation of magnetic fields. The present paper focuses on the study of the spatial and temporal profiles of the generated magnetic fields. In the normal case of infinite and/or periodic simulation box with homogeneous plasma density the observed magnetic field dominates at the scale length of skin depth. The role of plasma density inhomogeneity and the finite transverse width of the beam electrons are investigated in the work. It is shown that when the plasma density inhomogeneity with scales sharper than the skin depth is chosen, the magnetic field structures with similar short scales form [2]. It is also observed that when the beam width is finite magnetic fields with structures at the scale length of beam width form. References: [1] Erich S. Weibel. Phys. Rev. Lett. 2, 83 (1959) [2] Chandrasekhar Shukla, Amita Das and Kartik Patel, “1D3V PIC simulation of propagation of relativistic electron beam in an inhomogeneous plasma” Phys. Scr. 90, 085605 (2015) 243 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_43 PIC Modeling of Negative Ion Extraction from a Dust-Seeded Plasma Ananya Phukan1, Pranjal Bhuyan1 1Centre of Plasma Physics-Institute for Plasma Research, India Email: ananya.phukan26 @gmail.com Plasma behavior near the plasma grid of Negative Ion (NI) sources [1] is studied by a 3D – electrostatic Particle-In-Cell (PIC) code. The computational domain is assumed to be a cuboid volume around a single hole of the plasma grid. The plasma is assumed to be seeded with Cs coated dusts [2] that provides additional surfaces for NI production throughout the volume of the source. The dusts are not explicitly modeled; rather, constant charges are assumed to remain distributed randomly throughout the volume of the plasma mimicking the dust particles. The effect of dust on NI extraction is studied by considering its effect on controlling parameters like meniscus formation for different combination of the system variables. Figure 1: The figure shows scatter plot of ion and electron distribution near a conical hole in the plasma grid (along a plane at the middle of x-axis), showing meniscus formation by Ions (left), and also electron beam coming out of the hole (right) due to the absence of electron magnetizing fields. The plot is overlaid with contour plot of extraction potential penetrating into the plasma core through the hole. The contours range between 3000V (rightmost curve) and 10V (leftmost curve). References: [1] S Mochalskyy, D Wunderlich, B Ruf, U Fantz, P Franzen and T Minea, “On the meniscus formation and the negative hydrogen ion extraction from ITER neutral beam injection relevant ion source,” Plasma Phys. Control. Fusion 56, 105001 (2014). [2] A Phukan, K. S. Goswami, and P. J. Bhuyan, “Potential formation in a collisionless plasma produced in an open magnetic field in presence of volume negative ion source” Phys of Plasmas 21, 084504 (2014) 244 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_62 Dynamics of dusty fluid in a streaming sheared plasma Modhuchandra Laishram1, Devendra Sharma1, Predhiman Krishan Kaw1 1Institute for Plasma Research, India Email: [email protected] Experimental observations of toroidal dust vortex flow dynamics in an azimuthally symmetric cylindrical plasma setup [1] are analyzed using 2-dimensional fluid model for the electrically levitated and confined dust in a dynamical flow equilibrium. Driven by an unconfined sheared flow of a streaming plasma the single and multiple poloidal dust vortex structures are recovered [2]. Analytic structure of the dust vortex flow shows departure from correlations with the driving ion flow field even at the low dust Reynolds numbers as a result of finer scales introduced by the boundaries. Characterization of boundary layer width and effective Reynolds number with respect to kinematic viscosity reveal existence of a definite exponent with respect to the viscosity over a substantially large range of Reynolds number [3]. These orderings are observed to be modified by increasing degree of plasma flow turbulence indicating a correlation between dust dynamics and properties of plasma flow and transport. References: [1] Manjit Kaur et al. Phys. of Plasmas., 22, 033703 (2015). [2] M. Laishram, D. Sharma, and P. K. Kaw, Phys. of Plasmas., 21, 073703 (2014). [3] M. Laishram, D. Sharma, and P. K. Kaw, Phys. Rev. E., 91, 063110 (2015). 245 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_75 Numerical Analysis on Bandwidth and Growth Rate of Plasma-Filled Gyrotron Devices Elaheh Allahyari1, Shahrooz Saviz1 1Science and Research Branch of Islamic Azad University, Iran Email: [email protected] The linear theory of a plasma–loaded gyrotron amplifier is studied in the fast and mixed wave modes. The analysis is done for an infinitely hollow thin electron beam, as the electrons have the same energy and angular momentum. The plasma is assumed to be cold. In the numerical analysis, the plasma has electrons and ions, with dielectric coefficient . The system configuration is consist of the cylindrical plasma column loaded inside the electron beam and is placed parallel to the axis of conductive cylinder. There is a strong magnetic field, Be0 ˆz along the axis of the cylinder. The dispersion relation is derived with the Vlasov-Maxwell’s equations. The effects of beam location, plasma column radius, electron beam parameters and azimuthal harmonic number on the growth rate for fast and mixed wave modes are investigated. Results show that the growth rate and bandwidth of the mixed wave mode is larger than the fast wave mode. It is shown that the bandwidth of this structure is largest for small value of the axial momentum spread. 246 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_76 Gyro-TWT in a Vane-Loaded Waveguide with Inner Dielectric Reyhaneh S. Hashemi1, Elaheh Allahyari1, Shahrooz Saviz1 1Science and Research Branch of Islamic Azad University, Iran Email: [email protected] In recent years, there have been numerous theoretical investigations of the gyrotron amplifier in a dielectric loaded waveguide, motivated by properties of the gyrotron amplifier, including influence of inner dielectric material on stability behavior. In this study, a Gyro-TWT in a vane- loaded waveguide with inner dielectric is investigated. The hollow electron beam propagates between the dielectric rod and the waveguide wall, so it interacts with electromagnetic wave. The waveguide is a vane-loaded one in which we analyze the parameters including the number of vanes, the angle between the vanes, beam location, dielectric radius and electron beam parameters. Effect of the vane-loaded waveguide on the growth rate and bandwidth of two different modes, a fast wave mode and a mixed wave mode, are discussed. The plot of growth rate versus frequency is illustrated for the fast wave mode. The results show that the existence of metal vanes would decrease the growth rate. The presence of dielectric material on the stability behavior of the fast wave mode does not have any influence. That is to say, the stability properties are almost independent of the dielectric constant. The growth rate plot of mixed wave mode for various values of axial momentum spread is illustrated. The results show that for a small axial momentum spread (Δ less than 0.005), the growth rate and bandwidth would increase. 247 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_77 Effect of Plasma Column on the Radial Profile of Electric Field of Gyrotron Devices Elaheh Allahyari1, Shahrooz Saviz1 1Science and Research Branch of Islamic Azad University, Iran Email: [email protected] In the present work the radial behavior of the electric field is investigated. In this analysis we consider the system in the absence of the electron beam in the fast wave mode. The system configuration is consist of the cylindrical plasma column loaded inside the cylindrical waveguide. The external magnetic field, Be0 ˆz , exists along the axis of the waveguide. By using Maxwell’s equations the differential equation for the axial component of the electric field is evaluated. The solution for the electric field considering the boundary conditions in each region of this configuration is determined. As the plots shown the electric field at the plasma edge is greater than at the plasma column center. It is clear that when the distance between the plasma column and the cylinder wall decreases, the electric field oscillates less. It is also shown that the ratio of electric field in cylinder radius to electric field in plasma column radius, outside the plasma becomes small, and the mode becomes similar to the transverse electromagnetic wave that propagates on a coaxial line. 248 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_83 Current Gradient Modes of Two Dimensional Electron Magnetohydrodynamics (EMHD) Gurudatt Gaur1, Predhiman Krishan Kaw1 1Institute for Plasma Research, India Email: [email protected] A two dimensional electron magnetohydrodynamic (EMHD) model [1] has been evoked to study the current gradient driven modes of a one-dimensional equilibrium sheared electron current configuration. No variations along the equilibrium current are chosen which excludes the Kelvin- Helmholtz modes [2] in the system. A linear analysis shows that the perturbations parallel to equilibrium magnetic field B0, driven by the current-gradients, lead to two different modes. The first mode is the tearing mode [3] having a non-local behavior which requires the null-line in the magnetic field profile. Whereas, the second mode is a non-tearing local mode [4-5] which does not require the null-line in the ′′ magnetic field. No unstable mode exists when the quantity B0 −B0 does not change the sign. We also have carried out the numerical simulations to understand the nonlinear regime in the presence of one or both the modes. References: [1] A. V. Gordeev, A. V. Gordeev, A. S. Kingsep, and L. I. Rudakov, “Electron magnetohydrodynamics,” Physics Reports, 243, 215 (1994). [2] A. Das and P. Kaw, “Nonlocal sausage-like instability of current channels in electron magnetohydrodynamics,” Phys. Plasmas, 8, 4518 (2001). [3] S. V. Bulanov, F. Pegoraro, and A.S. Sakharov, “Magnetic reconnection in electron magnetohydrodynamics,” Phys. Fluids B, 4, 2499 (1992). [4] N. Jain, A. Das and P. Kaw, “Kink instability in electron magnetohydrodynamics,” Phys. Plasmas, 11, 4390 (2004). [5] V. S. Lukin, “Stationary nontearing inertial scale electron magnetohydrodynamic instability,” Phys. Plasmas, 16, 122105 (2009). 249 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_86 A Poynting like Theorem for Generalized Hydrodynamic Equations Vikram Singh Dharodi1, Bhavesh Patel1, Amita Das1, Predhiman Krishan Kaw1 1Institute for Plasma Research, India Email: [email protected] The generalized hydrodynamic (GHD) model depicts the behaviour of a visco – elastic fluid. The model has often been invoked for the understanding of the behaviour of strongly coupled dusty plasma medium below its crystallization limit. The model supports both compressive acoustic and tranverse shear modes. Restricting to the incompressible limit, we obtain a Poynting like conservation equation for the system. Simulation studies have also been performed which confirm the validity of the theorem and help identify the contribution for the loss of conserved quantity from that which is lost through convective dissipation. 250 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_100 Identification of Nonlinear Resonance Absorption in a Laser Driven Deuterium Cluster using Molecular Dynamics Simulation Sagar Sekhar Mahalik1, Mrityunjay Kundu1 1Institute for Plasma Research, India Email: [email protected] Collisionless laser energy absorption in nanometer size atomic clusters may occur through linear and nonlinear resonance (NLR). During the linear resonance which typically requires long laser pulses > 100 fs, Mie-plasma frequency of the expanding cluster becomes equal to the laser frequency and electrons leave the cluster by absorbing good amount of laser energy [1]. However, for very short infrared (800 nm wavelength) laser pulses of duration < 30 fs linear resonance processes do not contribute and laser energy absorption by cluster electrons mainly happen by NLR which occurs in the anharmonic potential of the spherical cluster when a driven electron's frequency meets the laser frequency. Earlier NLR absorption mechanism was studied by particle-in-cell (PIC) simulations and simple analytical model [2]. But it is not rigorously verified so far by first principle methods e.g. molecular dynamics (MD) simulation. In this work, we identify NLR mechanism in a laser driven deuterium cluster by a newly developed three dimensional MD simulation code with soft-core Coulomb interaction among the charge particles. By following the trajectory of each individual electron and identifying its time-dependent frequency in the self-consistent anharmonic potential it is found that electron leaves the potential only when NLR condition is met. Thus we bridge the gap between PIC simulations, analytical model and first principle MD calculations and prove that NLR processes are a universal dominant mechanism of absorption in the short pulse regime or early time of longer pulses. References: [1] T. Ditmire et al., “Interaction of intense laser pulses with atomic clusters”, Phys. Rev. A, 53, 5 (1996). [2] M. Kundu and D. Bauer, “Nonlinear Resonance Absorption in the Laser-Cluster Interaction”, Phys. Rev. Lett., 96, 123401 (2006). 251 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_128 1D PIC simulation of relativistic Buneman instability Roopendra Singh Rajawat1, Sudip Sengupta1 1Institute for Plasma Research, India Email: [email protected] Buneman instability in the relativistic regime has been studied using a 1D electrostatic particle- in-cell code. In the non-relativistic case, Hirose et al. (Plasma Phys. 20, 481(1978)) has shown 2 that breakdown of linear growth (saturation) occurs when |E| /16πW0 ~ max, where W0 is the initial beam kinetic energy density and max is maximum growth rate of the instability. In the weakly relativistic case, it has been confirmed using PIC simulation that scaling of saturation of Buneman instability follows a similar behavior as the non-relativistic case, whereas in the strongly relativistic case our simulation results show significant deviation from Hirose's results. In the strongly relativistic case, growth rate reduces due to relativistic corrections; so saturation occurs at a lower value compared to the non-relativistic/weakly relativistic case. 252 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_130 Molecular Dynamics Simulation of Dust Particle Levitation in the Presence of Sheath Sandeep Kumar1, Amita Das1, Sanat Kumar Tiwari1, Predhiman Krishan Kaw1 1Institute for Plasma Research, India Email: [email protected] Two dimensional Molecular Dynamics simulation has been carried out to study the phenomena of dust levitation in the presence of the sheath electric fields. The dust particles are typically made to levitate in experiments by applying an electric field to the electrode, which balances the gravitational pull felt by the heavy dust particles. In the simulations the potential due to the applied external electric field has been chosen to be exponentially decaying from the wall, taking account of the shielding by the lighter plasma species. The inter dust potential is represented by a Yukawa potential again considering the effect of shielding from the lighter electrons and ion species in the plasma. The simulation considers both single and two kinds of dust species (at present only their respective masses are chosen to be different). The equilibrated density profile as a function of the height is plotted. It is observed that the heavier dust species accumulates at the bottom and on top of it the lighter species settles. The peak of the density occurs at a location where there is a balance between the sheath and the gravitational forces. The sheath width is dependent on the total number of particles and the temperature of the system. By reversing the direction of the gravitational force a configuration where the heavier particle reside on top of lighter one results, and the evolution shows appearance of the Rayleigh - Taylor instability. The simulations show the reduction of the growth rate of instability in the presence of strong coupling in conformity with the predictions of the Generalized Hydrodynamics (GHD) fluid model for visco - elastic systems. References: [1] G. Foroutan and A. Akhoundi, Numerical study of an electrostatic plasma sheath containing two species of charged dust particles, Journal of Applied Physics, 112, 073301, (2012). [2] Jin-yuan Liu and J. X. Ma, Effects of various forces on the distribution of particles at the boundary of a dusty plasma, Physics of Plasmas, 4, 2798, (1997). 253 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_135 Conceptual Study of High-Field LHCD in KSTAR Young-soon Bae1, S Shiraiwa2, P Bonoli2, S J Wang1, G Wallace2, J C Wright2, R Parker2, Won Namkung2, M H Cho3, H. Park1,4 1National Fusion Research Institute, Korea 2Plasma Science Fusion Center, MIT,USA 3Pohang University of Science and Technology, Korea 4Ulsan National Institute of Science and Technology, Korea Email: [email protected] An innovative lower-hybrid (LH) current drive scheme in KSTAR tokamak is being studied in order to achieve high performance advanced tokamak operation. Taking advantage of less plasma wall interaction and good LH wave accessibility at the high toroidal magnetic field, the inside LH wave launch would provide good opportunity to study reactor-relevant operation scenario using LHCD. We investigated the LH wave launch parameters and plasma operation conditions to provide efficient current drive by inside high-field LH wave launch using the ray tracing code (GENRAY) and Fokker-Planck code (CQL3D). The conventional launcher structure is very unlikely to be used due to the limited space in the inboard side and complicated path of the waveguide through divertor section. The new concept of LH launcher structure is therefore suggested and discussed in this paper. 254 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_138 Integrated Core-SOL Simulations of L-Mode Plasma in ITER and Indian DEMO Apiwat Wisitsorasak1, Thawatchai Onjun2, Wittawat Kanjanaput2 1King Mongkut's University of Technology Thonburi, Thailand 2Sirindhorn International Institute of Technology, Thailand Email: [email protected] Core-SOL simulations are carried out using 1.5D BALDUR integrated predictive modeling code to investigate tokamak plasma in ITER and Indian DEMO reactors operating in low confinement mode (L-Mode). In each simulation, the plasma current, temperature, and density profiles in both core and SOL region are evolved self-consistency. The SOL is simulated by integrating the fluid equations, including sources, along the field lines. The solutions in SOL subsequently provide as the boundary conditions of core plasma region on low-confinement mode. The core plasma transport model is described using a combination of anomalous transport by Multi-Mode-Model version 2001 (MMM2001) and neoclassical transport calculated by NCLASS module together with the toroidal velocity based on the torque due to Neoclassical Toroidal Viscosity (NTV). In addition, a sensitivity analysis is explored by varying plasma parameters, such as plasma density and auxiliary heating power. Furthermore, the ignition tests are conducted to observed plasma response in each design after shutting down an auxiliary heating. 255 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_150 Potential around a dust grain in collisional plasma Rakesh Moulick1, Kalyan Sindhu Goswami1 1Centre of Plasma Physics-Institute for Plasma Research, India Email: [email protected] The ion neutral collision can lead to interesting phenomena in dust charging, totally different from the expectations based on the traditional OML theory. The potential around a dust grain is investigated for the collisional plasma considering the presence of ion neutral collisions. Fluid equations are solved for the one dimensional radial coordinate. It is observed that with the gradual increase of ion neutral collision, the potential structure around the dust grain changes its shape and is different from the usual Debye- Hückel potential. The shift however, starts from a certain value of ion neutral collision and the electron-ion density varies accordingly. The potential variation is interesting and reconfirms the fact that there exists a region of attraction for negative charges. The collision modeling is done for the full range of plasma i.e. considering the bulk and sheath jointly. The potential variation with collision is also shown explicitly and the variation is found to cope up with the earlier observations. 256 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_167 Numerical simulation of a novel non-transferred arc plasma torch operating with nitrogen Gavisiddayya Hiremath1, Ramachandran Kandasamy2, Ravi Ganesh3 1Karunya University, India 2UGC-Associate Professor, India 3FCIPT-Institute for Plasma Research, India Email: [email protected] High power plasma torches with higher electro-thermal efficiency are required for industrial applications. To increase the plasma power and electrothermal efficiency, conventional torches are being modified to operate with molecular gases such as air and nitrogen. Since increasing arc current enhances the heat loss to the anode, torches are being developed to operate under high voltage and low current. The plasma flow dynamics and electromagnetic coupling with plasma flow inside the torch etc. are highly complex and knowledge on the same is required to develop high torches with higher efficiency. Unfortunately detailed experimentation on the same is very difficult. Numerical modeling and simulation is one of the best tools to understand the physics involved in such complex processes. A 2D numerical model is developed to simulate the characteristics of the plasma inside the torch. Though plasma is not in local thermodynamic equilibrium (LTE) close to the electrodes, LTE is assumed everywhere in the plasma to avoid complex and time consuming calculations. Other valid assumptions used in the model are plasma flow is optically thin, laminar and incompressible. Flow, energy and electromagnetic equations are solved with appropriate boundary conditions and volume sources using SIMPLE algorithm with finite volume method. Temperature dependent thermophysical properties of nitrogen are used for the simulations. Simulations are carried out for different experimental conditions. The effects of arc current, gas flow rate of plasma generating gas and sheath gas injected above the bottom anode on the arc voltage, electrothermal efficiency of the torch, plasma temperature and plasma velocity are simulated. Predicted results are compared with experimental results. 257 | P a g e 10th Asia Plasma & Fusion Association Conference Poster Session-5 258 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_180 Nonlinear MHD modeling in LHD plasmas with peaked pressure profiles Yasuhiro Suzuki1,2 1National Institute for Fusion Science, Japan 2SOKENDAI, The Graduate University for Advanced Studies, Japan Email: [email protected] Nonlinear dynamics in Heliotron plasmas using a 3D nonlinear MHD simulation code in heliotron plasmas is studied. In the Large Helical Device (LHD) experiment, many MHD instabilities are observed. Especially, if the peaked pressure profile was sustained by the pellet injection, a collapse event, so-called the core density collapse (CDC), was happen. In nonlinear MHD simulations, it is expected the CDC is driven by the resistive ballooning mode [1]. Recently, a new imaging diagnostics of the two-dimensional soft-X ray arrays is installed in the LHD. Using the new diagnostics, perturbations localized at the outward of the torus. That is a characteristic of the ballooning mode. So, it seems the ballooning mode is observed in the LHD experiments. However, to interpret the experimental observation, we need to know what kind mode patterns should be observed. In this study, we study 3D MHD equilibria with reconstructed pressure profile using a 3D MHD equilibrium code, which does not assume nested flux surfaces [2]. And then, we will study nonlinear MHD simulations based on the 3D MHD equilibrium with the magnetic island [3]. In this study, we note nonlinear saturation to compare with the experimental observation. References: [1] N. Mizuguchi, et al., Nucl. Fusion 49 (2009) 095023 [2] Y. Suzuki, et al., Nucl. Fusion. 46 (2006) L19 [3] Y. Todo, et al., Plasma Fusion Res. 5 (2010) S2062 259 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_188 Sensitivity analysis of upstream plasma condition for SST-1 X-divertor configuration with SOLPS Himabindu Manthena1, Anil K Tyagi1, Deepti Sharma1, Devendra Sharma1, Srinivasan Radhakrishnana1 1Institute for Plasma Research, India Email: [email protected] The extensive power exhausts and target heat loads are anticipated in reactor grade fusion devices. Prototyping of an X-Divertor based power exhaust scheme is being attempted by means of simulations of Scrape-off Layer plasma transport in the diverted plasma equilibria of SST-1 tokamak using SOLPS5.1.Evaluation of the relative advantages of an X-Divertor configuration involves simulating the SST-1 standard divertor scheme plasma transport for the reference and then achieving equivalent upstream plasma conditions in the X-divertor equilibrium to ensure an equivalent core plasma in both the cases. The first optimization is to be achieved by simulating effects of an external gas puff in the SOL region for controlling separatrix density in the X- divertor configuration with visible modifications in the downstream plasma conditions. The present work analyzes sensitivity of the upstream SOL plasma conditions to the gas puff intensity and its effect on the plasma neutral transport in the divertor region. 260 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_247 Radiation Effects on the Laser Ablative Shockwaves from Aluminum under Atmospheric Conditions Sai Shiva S1, Leela C H1, Sijoy C D2, Shashank Chaturvedi2, Prem Kiran P1 1University of Hyderabad, India 2Bhabha Atomic Research Centre-Visakhapatnam, India Email: [email protected] The evolution of laser ablative shockwaves (LASW) from Aluminum under atmospheric pressures is numerically modeled using a one-dimensional, three-temperature (electron, ion and thermal radiation temperatures), non-equilibrium, radiation hydrodynamic (RHD) model. The governing RHD equations in Lagrangian form are solved by using an implicit scheme. Similarly, the energy relaxation between the electrons and ions and the electrons and thermal radiation are determined implicitly. Apart from these, the energy equation takes into account the flux-limited electron thermal heat flux. The RHD equations are closed by using a two temperature QEOS model for the Al [1]. The MULTI-fs code is modified to incorporate the nanosecond laser absorption model via the photoionization (PI) and the inverse bremsstrahlung (IB) processes. The spatio-temporal evolution of the laser ablative shockwaves generated by focusing a second harmonic (532 nm, 7ns) of Nd:YAG laser on to Aluminum target under atmospheric pressures in air is captured using a shadowgraphy technique. These measurements are made from 200 ns to 10 s after the laser pulse with a temporal resolution of 1.5 ns [2]. We report the details of the RHD model and compare the simulated and experimental results for input laser energies in the range of 25 – 175 mJ per pulse. The evolution of the plasma parameters like electron density, charge states and the shockwaves launched into the ambient atmosphere due to expanding plasma plume are compared. The role of thermal radiation on the evolution of LASW from Al is discussed. References: [1] R. Ramis, K. Eidmann, J. Meyer-ter-Vehn, and S. Huller, “MULTI-fs - A computer code for laser-plasma interaction in the femtosecond regime,” Comp. Phys. Comm., 183, 637 (2012). [2] Ch. Leela, P. Venkateshwarlu, R.V. Singh, P. Verma, and P.P. Kiran, “Spatio-temporal dynamics behind the shock front from compacted nanopowders,” Optics Express, 22, A268 (2014). 261 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_269 Angular Momentum Transfer of Laguerre - Gaussian Laser Pulses and Quasi-static Magnetic Field Generation in Plasma Channels Bhawani Shankar Sharma1, Ramesh Chand Dhabhai2, Naveen Kumar Jaiman3 1RR Autonomous Gov College Alwar, India 2Govt. P.G.College Kota, India 3University of Kota, India Email: [email protected] To generate a strong axial and azimuthal quasi-static magnetic field, we propose to study the interaction of Laguerre-Gaussian laser beams in a parabolic plasma channel. Our study shows that the higher-order modes with orbital angular momentum generate a stronger magnetic field in comparison to the lower-order modes of the laser beam. The contribution of the effective mass of photon on the orbital angular momentum and the polarization state of the beam are analyzed analytically and with 2D Particle in Cell (PIC) simulation. These effects have been put forwarded in analyzing the magnetic field generation. 262 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_275 Real-time Horizontal Position Control for Aditya-Upgrade Tokamak Rohit Kumar1, Joydeep Ghosh1, Rakesh L Tanna1, Praveen Lal1, Prabal K Chattopadhyay1, Chhaya Chavda1, Vipul K Panchal1, Vijay Patel1, Chet Narayan Gupta1, Raju Daniel1, Aditya Team1 1Institute for Plasma Research, India Email: [email protected] Position of plasma column is required to be controlled in real time for improved operation of any tokamak [1]. A PID based system for real-time horizontal plasma position control has been designed for Aditya Upgrade tokamak. Modelling of transfer functions of actuators, plasma and diagnostic system [2] are carried out for ADITYA-U tokamak. The PID controller is optimized using MATLAB-SIMULINK for horizontal position control. Further feed-forward loop is implemented where disturbance due to density variation is suppressed [3], which results in improved performance as compared to conventional PID operation. In this paper the detailed design of the whole system for real time control of plasma horizontal position in Aditya Upgrade tokamak is presented. References: [1] V. Mukhovatov et al., “Plasma equilibrium in a tokamak’’, Nucl. Fusion, 11, 1971. [2] W.Z Yu et al., “Robust control design for the plasma horizontal position control on J-TEXT tokamak”, Fusion Engineering and Design, 88, 2013. [3] W.Z Yu et al., “Plasma horizontal position control for the J-TEXT tokamak based on feed- forward density compensation”, Plasma Phys. Control. Fusion, 56, 2014. 263 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_294 Prediction of Temperature and Stress Distributions in Substrate and Coating during Plasma Spraying Raja Mohan1, H Gavisiddayya1, K Ramachandran2, P V A Padmanabhan3, T K Thiyagarajan3 1Karunya University, Coimbatore, Tamil Nadu, India 2Bharathiar University, Coimbatore, Tamil Nadu, India 3Bhabha Atomic Research Centre, Mumbai, India Email: [email protected] A numerical model is developed to predict the temperature distribution in the coating and substrate during plasma spraying. A transient heat conduction equation is solved using finite volume method for coating (Al2O3), substrate (Cu) and substrate holder (SS) regions with appropriate boundary conditions. The heat flux received by the substrate/ previously deposited layer from the plasma and particles is calculated from the previous measurements. The variation of the coating and/or substrate temperatures with spraying time is shown for different deposition rates, % of porosities, bond coat thicknesses and spraying distances. Further an analytical model is developed to predict quenching, cooling and residual stresses in the coating-substrate system during spraying using predicted temperature distribution. Effects of micro-cracks in the coating on the residual stress in the coating are discussed. 264 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_5 Design & Development of Amplitude and Phase Measurement of RF Parameter with Digital I-Q De-Modulator (DIQDM) Technique using PXI System Dipal Soni1, Rajnish Kumar1, Sriprakash Verma1, Hriday Patel1, Rajesh Trivedi1, Aparajita Mukherjee1 1ITER-India, Institute for Plasma Research Email: [email protected] ITER-India, the Indian domestic agency for ITER project, is responsible to deliver one of the packages, called ICH&CD Radio Frequency Power Sources (RFPS). Total 20 MW of RF power is required for ITER plasma from RFPS system using 8 nos. of identical sources. Each power source is capable to deliver 2.5 MW @ 35 to 65 MHz frequency range with a load condition up to VSWR 2:1 & any reflection coefficient of phase angle [1]. Each source should be operated independently as well as in slave mode with synchronization of central plant control system of ITER. To fulfill the desired specifications of constant power and fixed relative phase, the real time control loop is required. The real time control loops would be used for maintaining the Amplitude and Phase as requested from central plant control system. Since, there are methods available for the measurement of amplitude and phase but the accuracy and linearity of the measurement is one of the important parameters, thus after survey and analysis ITER-India has chosen a digital I-Q demodulator based technique for amplitude and phase detection. RFPS is having two cascaded amplifier chains (10kW, 130kW & 1.5MW) combined to get 2.5 MW RF power output. Directional couplers are inserted at the output of each stage to extract forward power and reflected power as samples for measurement of relative amplitude and phase. Using passive mixer, forward power and reflected power are down converted to 1MHz Intermediate frequency (IF). This IF signal is used as an input to the DIQDM. Digital I-Q demodulator consists of National Instruments make PXI hardware, like PXI-8108 RT controller, PXI-7841R FPGA and PXI-6133. To realize the application of measurements, LabVIEW software tool is used. To generate 0, π/2, π, 3π/2 and 2π positions, 1 MHz IF sampled with 4 times higher sampling frequency. There are samples like, 0 as I, π/2 as Q, π as –I and 3π/2 as –Q. These samples are used for the algorithm used in DIQDM technique. In this paper, Amplitude and Phase measurement of RF signal with DIQDM technique using PXI system is described in detail, with various test results with dummy signals and low power RF systems. References: [1] A. Mukherjee et. al., “Ion Cyclotron Power Source System for ITER,” Fusion Science & Technology, Vol 65, (2013). 265 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_9 Effect of Geometrical Imperfection on Buckling Failure of ITER VVPSS Tank Saroj Kumar Jha1, Girish Kumar Gupta2, Manish Kumar Pandey1, Avik Bhattacharya1, Gaurav Jogi1, Anil Kumar Bhardwaj1 1ITER-India, Institute for Plasma Research 2Institute for Plasma Research, India Email: [email protected] The ‘Vacuum Vessel Pressure Suppression System’ (VVPSS) is Part of ITER machine, which is designed to protect the ITER Vacuum Vessel and its connected systems, from an over-pressure situation. It is comprised of a partially evacuated tank of stainless steel approximately 46 meters long and 6 meters in diameter and thickness 30mm. It is to hold approximately 675 tonnes of water at room temperature to condense the steam resulting from the adverse water leakage into the Vacuum Vessel chamber. For any vacuum vessel, geometrical imperfection has significant effect on buckling failure and structural integrity. Major geometrical imperfection in VVPSS tank depends on form tolerances. To study the effect of geometrical imperfection on buckling failure of VVPSS tank, finite element analysis (FEA) has been performed in line with ASME section VIII division 2 part 5 [1], ‘design by analysis method’. Linear buckling analysis has been performed to get the buckled shape and displacement. Geometrical imperfection due to form tolerance is incorporated in FEA model of VVPSS tank by scaling the resulted buckled shape by a factor ‘60’. This buckled shape model is used as input geometry for plastic collapse and buckling failure assessment. Plastic collapse and buckling failure of VVPSS tank has been assessed by using the elastic–plastic analysis method. This analysis has been performed for different values of form tolerance. The results of analysis show that displacement and load proportionality factor (LPF) vary inversely with form tolerance. For higher values of form tolerance LPF reduces significantly with high values of displacement. References: [1] ASME, “Rules for the Construction of Pressure Vessels,” ASME Boiler and Pressure Vessel Code, Section VIII, Division 2, Alternative Rules, (2010). 266 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_17 Nuclear Analyses of Indian LLCB Test Blanket System in ITER H L Swami1, Akshaya Kumar Shaw1, Chandan Danani1, Paritosh Chaudhuri1 1Institute for Plasma Research, India Email: [email protected] Heading towards the Nuclear Fusion Reactor Program, India is developing Lead Lithium Ceramic Breeder (LLCB) tritium breeding blanket for its future fusion Reactor [1, 2]. A mock-up of the LLCB blanket is proposed to be tested in ITER equatorial port no.2 [3], to ensure the overall performance of blanket in reactor relevant nuclear fusion environment. Nuclear analyses play an important role in LLCB Test Blanket System development. It is required for tritium breeding estimation, thermal-hydraulic design, coolants process design, radio-active waste management, equipments maintenance & replacement strategies and nuclear safety. To predict the nuclear behaviour of LLCB test blanket module in ITER, nuclear responses like tritium production, nuclear heating, neutron fluxes and radiation damages are estimated. As a part of ITER machine, LLCB TBS has to follow certain nuclear shielding requirements i.e. shutdown dose rates should not exceed the defined limits in ITER premises (inside bio-shield ~100 Sv/hr after 12 days cooling & outside bio-shield ~10 Sv/hr after 1 day cooling). Hence nuclear analyses are performed to assess & optimize the shielding capability of LLCB TBS inside and outside bio-shield. To state the radio-activity level of LLCB TBS components which support the rad-waste and safety assessment, nuclear activation analyses are executed. Nuclear analyses of LLCB TBS are performed using ITER recommended nuclear analyses codes (i.e. MCNP, EASY), nuclear cross section data libraries (i.e. FENDL 2.1, EAF) and neutronic model (ITER C-lite v.1). The paper describes comprehensive nuclear performance of LLCB TBS in ITER. References: [1] R. Srinivasan, S P Deshpande, et al., “Strategy for the Indian DEMO design”, Fusion Engineering and Design, 83 (2008) 889–892. [2] Paritosh Chaudhuri, Chandan Danani, et al., “Thermal–hydraulic and thermo-structural analysis of first wall for Indian DEMO blanket module”, Fusion Engineering and Design, Volume 84, Issues 2–6, June 2009, Pages 573–577 [3] L.M. Giancarli, M. Abdou, et al., “Overview of the ITER TBM Program”, Fusion Engineering and Design, Volume 87, Issues 5–6, August 2012, Pages 395–402 267 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_28 Preferential Binding of Self-interstitial Atoms over Vacancies to Grain Boundaries of Tungsten: A Lattice Statics Study Prithwish K Nandi1 1ITER-India, Institute for Plasma Research, India Email: [email protected] Understanding materials properties under prolonged radiation exposure in a nuclear reactor is important. Especially, the future design of a fusion reactor will produce a much larger amount of self-interstitial atoms (SIA), vacancies and transmutation impurities in plasma facing component materials like tungsten (W). Therefore, the structural stability of such components over longer period very much depends on how W can accumulate such defects efficiently to decelerate structural degradation. Both experimental and computational efforts have been put by materials engineers to understand the spatial and temporal evolution of defect microstructure in W. Though ion irradiation is a widely used technique to quickly simulate neutron damage in materials, it has its own disadvantage – it creates mostly superficial damage to a material and therefore, the conclusive scenario of in-depth microstructure is often debatable. Moreover, it is quite challenging to understand the formation, diffusion and segregation of such defects in the atomistic level from even the state-of-the art experimental setups. Hence the atomistic modeling comes into picture which can enlighten such gray areas of the damage process with a certain level of confidence that owes to the supremacy of the interatomic potential model, the delicacy of the computational techniques and the complexity of the representative physical model to mimic the real system. Almost all the materials are polycrystalline. Therefore, to understand the radiation induced segregation of solute and impurities, it is crucial to introduce grain boundaries (GB) in materials models while simulating defect microstructure and its subsequent dynamical evolution that starts with the transfer of kinetic energy to a single primary knock-on atom. Employing grain boundary engineering (GBE) method, Wanatabe first demonstrated that it is possible to tailor the intergranular fracture mechanism by elevating the fraction of the low order coincident site lattice (CSL) boundaries in the materials [1]. CSL boundaries are thus very special in GBE though they occupy only a small portion of the five-dimensional geometric GB phase space. In the present study, we have explored the equilibrium structures of more than 25 symmetric tilt CSL boundaries with <100> tilt axis and their effects on both vacancies and SIA bindings in W. A substantially large number of initial configurations for each CSL type are sampled to identify the equilibrium stable and metastable GB structures at 0 K using lattice statics in conjunction with embedded atom method potentials. Accessibility of equilibrium structure for each CSL type is also computed. Formation energies for both the vacancies and SIA are calculated when placed in a distance of ±15 Å from the GB plane. Our data shows that the formation energies vary significantly in and around the GB plane (within ±5 Å) as compared to its bulk value. Comparing the binding energies of both vacancies and SIA for each site, we conclude that interstitials are more prone to bind to grain boundary sites over vacancies – this is an important observation to understand segregation processes in polycrystalline W. 268 | P a g e 10th Asia Plasma & Fusion Association Conference References: [1] T. Wanatabe,“The impact of grain boundary character distribution on fracture in polycrystals”, Mater. Sci. Eng. A, 176, 39 (1994). 269 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_29 Alternate Design of ITER Cryostat Skirt Support System Manish Kumar Pandey1, Girish Kumar Gupta1, Anil Kumar Bhardwaj, Saroj Kumar Jha1 1ITER-India, Institute for Plasma Research, India Email: [email protected] The skirt support of ITER cryostat is a support system which takes all the load of cryostat cylinder and dome during normal and operational condition [1]. The present design of skirt support has full penetration weld joints at the bottom (shell to horizontal plate joint). To fulfill the requirements of tolerances and control the welding distortions, we have proposed to change the full penetration weld into fillet weld. A detail calculation is done to check the feasibility and structural impact due to proposed design. The calculations provide the size requirements of fillet weld. To verify the structural integrity during most severe load case, finite element analysis (FEA) has been done in line with ASME section VIII division 2 [2]. By FEA ‘Plastic Collapse’ and ‘Local Failure’ modes has been assessed. 5° sector of skirt clamp has been modeled in CATIA V5 R21 and used in FEA. Fillet weld at shell to horizontal plate joint has been modeled and symmetry boundary condition at ± 2.5°applied. ‘Elastic Plastic Analysis’ has been performed for the most severe loading case i.e. Category IV loading [3]. The alternate design of Cryostat Skirt support system has been found safe by analysis against Plastic collapse and Local Failure Modes with load proportionality factor 2.3. Alternate design of Cryostat skirt support system has been done and validated by FEA. As per alternate design, the proposal of fillet weld has been implemented in manufacturing. References: [1] Backhouse A., “Cryostat Skirt Support Analysis Report,” ITER Organization, (2011). [2] ASME, “Rules for the Construction of Pressure Vessels,” ASME Boiler and Pressure Vessel Code, Section VIII, Division 2, Alternative Rules, (2010). [3] Sannazzaro G., “Load Specification,” ITER Organization, (2012). 270 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_31 Neutronics Analysis, Shielding Optimization and Radiation Waste Analysis for X-Ray Crystal Spectrometer of ITER P V Subhash1, Gunjan Indauliya2, Sai Chaitanya Tadepalli1, Shrichand Jakhar3, Sanjeev Kumar Varshney1, Siddharth Kumar1, K Raja Krishna4, Nirav Bhaliya2, Robin Barnsley3, Bernascolle Phillipe3, Shrishail B Padasalagi1, Sapna Mishra1 and Vinay Kumar1 1ITER-India, Institute for Plasma Research 2Bhakti Consultants, India 3ITER-Organization, France 4UPES Dehradun, India Email: [email protected] Neutronics and activation analysis have been carried out for the X-ray Crystal Spectrometer (XRCS) system, which will be installed in equatorial port no. 11 assigned for the ITER diagnostics. ITER diagnostic port plugs are subject to severe nuclear environment that presents a critical design challenge. A neutron shield has been designed for the aforesaid XRCS sight tube which is a torus vacuum extension used for X-rays transmission, placed in the interspace. The neutron transport calculations are performed using Monte-Carlo N-Particle code (MCNP). The transport results are used for the design and optimization of a proper radiation shield for the sight tube. The base C-lite neutronics ITER model is grossly modified to include all required details of the port plug, diagnostic apertures and the diagnostic system. A local modelling approach has been used and cross talks from adjacent upper and lower ports are not considered in this analysis. The shield designed is effectively found to reduce the neutron flux to an acceptable level. ITER regulations demands the shutdown dose rate (SDDR) should be below a specified limit at a distance of 1 meter from port closure plate. While designing the radiation shield this factor plays an important role. A complete radioactive inventory calculation along with contact doses and nuclear activity levels are obtained for two different materials of sight tube. FISPACT 2007 used for this purpose. The analysis for this particular sight tube can be used to obtain a material preference based on radiation point of view. Various other nuclear responses like nuclear heating, DPA calculations and clearance index are also presented. This analysis specifically addresses the impact of material type, the sight tube made up of, on the SDDR. 271 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_78 Preliminary Optical Design of Polarization Splitter Box for ITER ECE Diagnostic System Ravinder Kumar1, Suman Danani1, Hitesh Kumar Pandya1, Vinay Kumar1 1ITER-India, Institute for Plasma Research Email: [email protected] In tokamak, electron cyclotron emission (ECE) leaves the magnetically confined plasma with two polarizing modes, one with electric field parallel to magnetic field known as ordinary mode or O-Mode polarization, and other with the electric field perpendicular to magnetic field, extraordinary Mode or X-Mode. These radiation modes will be collected simultaneously in the ITER ECE measurement line. Therefore, it is necessary to split the radiation into O and X-mode polarizations before transmission otherwise there might be polarization mixing during transmission of the ECE radiation from tokamak to the measurement instruments. The overall ITER ECE Diagnostic system is described in reference [1]. The collected radiation in O and X-mode polarization is coupled to the transmission line via polarization splitter unit. The polarizing modes will be simultaneously shared with the ITER ECE instruments, which are located in diagnostic room, consists of two Michelson interferometers that can simultaneously measure ordinary and extraordinary mode from 70 to 1000 GHz, and two heterodyne radiometer systems, one is covering 122-230 GHz (O-Mode radiometer) and other 244-355 GHz (X-Mode radiometer) frequency band. The X-mode radiometer is being developed by US ITER team. Proposed design of the polarization splitter box consists of two Gaussian beam telescopes built from three ellipsoidal mirrors [2] and one flat mirror. A wire grid beam splitter separates the O and X-Mode polarization emission. The box is covered with microwave absorber to minimize scattering of the radiation. The design is being optimized by simulation using the Gaussian beam Mode software [3] to achieve the desired performance, details will be discussed. References: [1] G. Taylor, et al., Status of the design of the ITER ECE Diagnostic, EPJ Web of Conferences, 87, 03002, 2015 [2] J.A. Murphy, Distortion of a simple Gaussian beam on reflection from off axis ellipsoidal mirrors, Int. J. Infrared and millimeter waves, 8(9), pp 1165-1168, 1987 [3] Gaussian Beam Mode 3D Engine (running within PTC's Creo Parametric 2.0) Thomas Keating Ltd. 272 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_87 Development of High Voltage and High Current Test Bed for Transmission Line Components Akhil Jha1, Manojkumar Patel1, J V S Harikrishna1, Ajesh P1, Rohit Anand1, Rajesh Trivedi1, Aparajita Mukherjee1 1ITER-India, Institute for Plasma Research Email: [email protected] India is responsible for delivery of 8+1(prototype) RF sources to ITER project. Each RF source will provide 2.5 MW of RF power at VSWR 2:1 in the frequency range of 35 to 65 MHz. Eight such RF sources will generate total 20 MW of RF power. A large number of high power transmission line components are required for connecting various stages of RF source. To test these passive transmission line components at high voltage and current level, similar to the level expected during operation, a test facility is required. A test bed based on the concept of standing wave resonator is being developed at ITER-India RFPS lab, which can be configured and operated for various lengths of the resonator for optimum requirement, for example, it may be quarter wave (/4), half wave (/2) and three quarter wave (3/4). RF power is fed to the resonator through a 12inch coaxial Tee. Input impedance of the resonator is matched with external RF source (50 ohm) using a tunable matching capacitor, which provides impedance matching for different operating conditions at resonance frequency. Peak voltage and current level of ~ 32 kV and ~ 900 A can be achieved inside the resonator during operation with an estimated input power of ~ 20 kW. The Device Under Test (i.e. transmission line components for testing) needs to be connected in-line during operation. In this paper, detailed design and simulation results are presented for the test bed. A brief description of future development and test plan for the test bed is described. 273 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_94 Development of Control System for Multi-converter High Voltage Power Supply using Programmable SoC Rasesh Dave1, Jagruti Dharangutti2, N P Singh1, Aruna Thakar1, Hitesh Dhola1, Sandip Gajjar1, Darshan Kumar Parmar1, Tanish Zaveri2, Ujjwal Kumar Baruah1 1ITER-India, Institute for Plasma Research 2Nirma University, India Email: [email protected] Multi-converter based High Voltage Power Supplies (HVPSs) find application in multi- megawatt accelerators, RF systems. Control system for HVPS must be a combination of superior parallel processing, real time performance, fast computation and versatile connectivity. The hardware platform is expected to be robust, easily scalable for future developments without any cost overhead [1]. Typical HVPS control mechanism involves communication, generation of precise control signals/pulses for few hundred Nos of chopper and closed loop control in microsecond range for regulated output [2]. Such kind of requirements can be met with Zynq All Programmable SoC, which is a combination of Dual core ARM Cortex A-9 Processing System (PS) and Xilinx 7 series FPGA based Programmable Logic (PL) [3]. Deterministic functions of power supply control system such as generation of control signals with precise inter-channel delay of nanosecond range and communication with individual chopper at 100kbps can be implemented on PL. PS should implement corrective tasks based on field feedback received from individual chopper, user interface and OS management that allows to take full advantage of system capabilities. PS and PL are connected with on-chip AXI-4 interface with low latency and higher bandwidth through 9 AXI ports. Typically PS boots first, this ensures secure booting and prevents external environment from tampering PL. This paper describes development of control system on Zynq All Programmable SoC for HVPS. References: [1] Altera’s White Paper, “Architecture Matters: Choosing the Right SoC FPGA for Your Application”, November 2013. [2] Young-Min Parky, Han-Seong Ryu, Hyun-Won Lee, Myung-Gil Jung and Se-Hyun Lee, “Design of a cascaded H-bridge multilevel inverter based on power electronics building blocks and control for high performance”, Journal of Power Electronics, vol 10, no 3, May 2010. [3] Louise H. Crockett, Ross A. Elliot, Martin A. Enderwitz, Robert W. Stewart, “The Zynq Book Embedded Processing with the ARM® Cortex®-A9 on the Xilinx® Zynq®-7000 All Programmable SoC,” 1st edition. 274 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_95 Development and Validation of I-Activation Analysis Code Sai Chaitanya Tadepalli1, P V Subhash1, Gunjan Indauliya2 1ITER-India, Institute for Plasma Research 2Bhakti Consultants, India Email: [email protected] I-Activation Analysis Code (IAAC) is a nuclear depletion code which solves coupled Bateman equations for radioactive-transmutation and growth-decay system for large numbers of isotopes to get time evolution of decay products and nuclear activity. It is currently being developed primarily for neutron activation and radiation waste analysis, as a part of the code development activities. The code functions by separating long and short-lived isotopes and then uses the well-known matrix exponential method to quickly solve a large system of coupled, linear, first-order ordinary differential equations with constant coefficients for long-lived isotopes. This method allows a faster treatment of complex decay and transmutation schemes. The short-lived isotopes are solved using approximated decay-chain method. FENDL 3.0 neutron activation files are used for data library. Separate set of code modules are designed to read, decode, convert and condense the continuous-energy ACE formatted data into 175 VITAMIN-J energy groups. The new compiled library that includes half-lives and neutron absorption cross sections is then used as input source for nuclear data. The code is readily suitable for calculations pertaining to nuclear transmutation, activation and decay studies in mainly fusion applications and activation analyses. Details of the code and its primary validation performed for various test cases and material compositions, largely related to current ITER project specific neutronic and radiation analyses will be presented. The nuclear activity calculations are validated against FISPACT, available under EASY code system. 275 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_97 Indigenous Manufacturing Realization of Twin Source and its Auxiliary System Ravi Pandey1, Mainak Bandyopadhyay2, Deepak Parmar2, Ratnakar Kumar Yadav2, Himanshu Tyagi2, Jignesh Soni1, Hardik Shishangiya2, Dass Sudhir Kumar2, Sejal Shah2, Gourab Bansal1, Kaushal Pandya1, Kanubhai Parmar1, Mahesh Vuppugalla1, Agrajit Gahlaut1, Arun Kumar Chakraborty1 1Institute for Plasma Research, India 2ITER-India, Institute for Plasma Research Email: [email protected] Indian negative ion source development program has gained momentum with planned integration of Indian Test Facility (INTF) for ITER – Diagnostic Neutral Beam (DNB) characterization at Institute for Plasma Research (IPR). Eight RF drivers based negative ion source, being developed for DNB will be tested and operated in INTF. The TWIN (Two driver based indigenously built Negative ion source) source provides a bridge between the operational single driver based negative ion source test facility, ROBIN in IPR and an ITER-type multi driver based ion source. The source is designed to be operated in CW mode with 180kW, 1MHz, 5s ON/600s OFF duty cycle and also in 5Hz modulation mode with 3s ON/20s OFF duty cycle for 3 such cycle. The complete design of TWIN source and its test facility, from conceptual to detailed engineering, has been carried out in IPR. The manufacturing design has been optimized to match the capability of Indian manufacturers, without compromising on the specifications. Some examples of optimization are i) an improvised design of the Faraday shields where electro-deposition has been replaced by vacuum brazing, ii) a simplified design of the side walls of the plasma source, where jointing process is simplified, without the application of Electron Beam Welding (EBW), iii) introduction of an FRP based integrated electrical and vacuum isolation scheme that replaces the application of a large ceramic. Finite element analysis (FEA) based on heat load and structural load calculation ensure the functionality and structural integrity of each components of the source. Due to non-nuclear environment in TWIN source experimental area, vacuum brazing is an acceptable manufacturing process. The contract for manufacturing of the ion source has been awarded to an Indian manufacturing company for the first indigenous production of a large size fusion grade ion source. The uniqueness of the TWIN source design is that, it can be operated both in Air mode (ROBIN type operation) as well as Vacuum mode (DNB type vacuum immersed operation). The Twin Source shall be manufactured as per ASME guidelines for pressure vessel. Experiments on the Twin Source are foreseen in the near future, as all the auxiliary systems like 180 kW, RF generator system, vacuum vessel with Pumping station, Cooling water system, Data acquisition and control system (DACS) and other power supply systems are already installed in the lab premises. The paper discusses the FEA based engineering design, simplified manufacturing design, manufacturing experience with highlighting quality control and the system integration activities undertaken for the TWIN source test facility. 276 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_98 Wilkinson Type Lumped Element Combiner-Splitter for Indigenous Amplifier Development Manojkumar Patel1, Akhil Jha1, J V S Harikrishna1, Rajesh Trivedi1, Aparajita Mukherjee1 1ITER-India, Institute for Plasma Research Email: [email protected] India is developing ITER like Ion Cyclotron Heating & Current Drive (ICH&CD) RF source in the frequency of 35 to 65 MHz. Three cascaded amplifiers will be used. Tube based driver (~150 kW) and final (1.7 MW) stage amplifier are driven by a solid state power amplifier (~ 10 kW). Development of wideband solid state power amplifier in above frequency range is ongoing. The goal is to achieve power level of ~ 12 kW/CW. 16 pallet amplifier modules, each of ~ 1kW, will be combined using 16x1 wideband combiners. 16 RF signals, with equal phase, will be required to drive each pallet module. 1x16 wideband splitter will be used at input side. Study has been carried out on two options mainly coaxial type & lumped element based Wilkinson splitter/combiner. Tentative power level of both input N-Type ports of combiner is ~ 1kW. Design and simulation for coaxial type Wilkinson combiner is done. Quarter wave length for center frequency is ~ 1500 mm. To reduce mechanical dimension of combiner, PTFE dielectric is used with complicated arrangement. Coaxial combiner required unique fabrication process. Alternate option is proposed as a lumped element based Wilkinson combiner with reduced size, cost & development time. Design and simulation was carried out. Required PCB design & fabrication was done accordingly. Same design will be implemented for splitter as well. Design scheme for the splitter/combiner will be finalized depending on the achieved performance of both the designs. In this paper, detailed design, simulation and test results are presented for both types of combiners. A detailed comparison of combiners is provided. 277 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_102 Preliminary Design Development of ITER X-ray Survey Spectrometer Sanjeev Kumar Varshney1, Siddharth Kumar1, Sapna Mishra1, Subhash Puthenveetil1, Kuashal Joshi1, Shivakant Jha1, Vinay Kumar1, Robin Barnsley2, Philippe Bernascolle2, Gunter Bertschinger2, Stefan Simrock2, Jean-Marc Drevon2, and Michael Walsh2 1ITER-India, Institute for Plasma Research, India 2ITER-Organization, France Email: [email protected] In order to reliably operate the ITER machine and make physics measurements, a large number of plasma and first wall diagnostics have been envisaged [1]. Similar to JET X-ray crystal spectrometers (XRCS) [2], which operated successfully in D-T phase, advanced design of crystal spectrometers are in development to work in harsh radiation environment and to satisfy the ITER measurement requirements. ITER-India, the domestic agency of ITER in India, is developing X- ray Crystal Spectrometers for ITER. These are based on X-ray spectroscopy of Hydrogen or Helium like ions of low to high Z impurities in the plasmas. The XRCS-Survey, a broad-band X- ray spectrometer, is one of the first plasma diagnostics to help the start-up of the plasma operations. The primary function is to measure plasma impurities due to various in-vessel components exposed to the plasma or from plasma dopants. The performance of the optical setup has been simulated and results have shown that the specified ITER measurement requirements are mostly realizable [3]. The preliminary design of XRCS Survey has been developed addressing many challenges such as, (1) designing a 7.5 meter long, vacuum extending sight-tube that interfaces spectrometer, placed in the port-cell, with equatorial port-plug (EPP-11) while allowing ~50 mm machine movements, (2) optimizing neutron shield design so that systems can fit into the available space and still the shutdown dose rates (SDDR) remains within the safe limits (3) designing tightly bent crystals (radius curvature ~ 250 mm) and estimating the modifications to the image properties etc. To meet these requirements, design detailing has been done for the sight-tube layout and its components. Engineering and neutronic analyses are completed for estimating the thermal displacement, stresses in the front-end components, neutron flux on the sight-tube components, SDDRs in the interspace region etc. Pressure profile inside vacuum chamber has been simulated. Effects of tight bending on the crystal are assessed using ANSYS. Shadow-XOP ray-tracing simulations are performed to simulate optical performance for a group of crystals and crystal bending effects. The spectrometer performance using Si-cuts for the high energy channels has also been analyzed. Furthermore, much progress has been made in the design of the plant I&C in terms of requirements, operating procedures, functional analysis and variable definition, hardware architecture, signal list and automation. This paper will focus on the design developments made to the ITER X-ray Survey spectrometer, and will discuss some of the key results of analysis towards the preliminary design. 278 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_112 Integration & Validation of LCU with Different Sub-systems for Diacrode Based Amplifier Rajnish Kumar1, Sriprakash Verma1, Dipal Soni1, Hriday Patel1, Gajendra Suthar1, Hrushikesh Dalicha1, Hitesh Dhola1, Amit Patel1, Dishang Upadhyay1, Akhil Jha1, Manojkumar Patel1, Rajesh Trivedi1, Raghuraj Singh1, Harsha Machchhar1, Aparajita Mukherjee1 1ITER-India, Institute for Plasma Research, India Email: [email protected] ITER-India as a domestic agency for the ITER project, is responsible to deliver one of the packages called ICH&CD Radio Frequency Power Sources (RFPS) to ITER system. Each power source is capable to deliver 2.5 MW at 35 to 65 MHz frequency range with a load condition up to VSWR 2:1 & any reflection coefficient of phase angle. Each source should be operated independently as well as in synchronization with central plant control system of ITER. For remote operation of different subsystems, like auxiliary power supply, high voltage power supply, low power RF system, Solid state power amplifier, Mismatched transmission line and 3MW-RF dummy load, Local Control Unit (LCU) is developed. LCU is developed using PXI hardware and Schneider PLC with LabVIEW-RT developmental environment. All the protection function of the amplifier is running on PXI 7841R module that ensures hard wired protection logic. There are three level of protection function- first power supply itself detects overcurrent/overvoltage and trips itself and generate trip signal for taking further action by protection function. There are some direct hardwired signal interfaces between power supplies (Anode trip to Screen Grid-off) to protect the amplifier. Second level of protection is generated through Command Control Embedded (CCE) against arc and Anode di/dt. Third level of Protection is through LCU where different fault signals are received and based on fault, off command of different sub-systems is generated within 1μs. Before connecting different subsystem with High power RF amplifiers (Driver & Final stage), each subsystem is individually tested through LCU. All protection functions are tested before hooking up the subsystems with main amplifier and initiating RF testing. The entire testing procedures and validation result, that was carried out by amplifier manufacturer along with ITER-India team during Site Acceptance Test of R&D amplifier will be discussed in this paper. 279 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_114 Comparative Analysis on Flexibility Requirements of Typical Cryogenic Transfer Lines Mohit Jadon1, Uday Kumar1, Ketan Choukekar1, Nitin Shah1, Biswanath Sarkar1 1ITER-India, Institute for Plasma Research, India Email: [email protected] The cryogenic systems and their applications, primarily in large Fusion devices, utilize multiple cryogen transfer lines of various sizes and complexities in terms of layout to transfer cryogenic fluids from plant to the various user/ applications. These transfer lines are composed of various critical sections like tee sections, elbows, flexible components etc. The mechanical sustainability (under failure circumstances) of these transfer lines are primary requirement for safe operation of the system and applications. The transfer lines need to be designed for multiple design constraint conditions like line layout, support locations and space restrictions. The transfer lines are subjected to single load and multiple load combinations, such as operational loads, seismic loads, leak in insulation vacuum etc. [1]. The analytical calculations and flexibility analysis using CAESAR II software are performed for the typical transfer lines without any flexible component, the results were analysed for functional and mechanical load conditions. The failure modes were identified along the critical sections. The same transfer line was then refurbished with the flexible components and analysed for failure modes. Inclusion of these components provides additional flexibility to the transfer line system and makes it safe. The optimization was performed by selection of the appropriate flexible components to meet the design requirements as per ASME B31.3/ EN 13480 codes. This paper describes the results obtained from the analytical calculations, which are compared and validated with those obtained from the flexibility analysis software calculations. References: [1] S Badgujar, L Benkheira, M Chalifour, A Forgeas, N Shah, H Vaghela, and B Sarkar, “Loads specification and embedded plate definition for the ITER cryoline system”, Cryogenic Engineering Conference and International Cryogenic Materials Conference, Tucson, Arizona; (2015) 280 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_116 Dynamics of Cold Helium Flow inside a Cryoline used for Large Cryogenic Distribution System Uday Kumar1, Mohit Jadon1, Ketan Choukekar1, Vinit Shukla1, Pratik Patel1, Himanshu Kapoor1, Nitin Shah1, Srinivasa Muralidhara1, Biswanath Sarkar1 1ITER-India, Institute for Plasma Research, India Email: [email protected] The Cryolines, which by definition transfers cryogens from the source, normally a cryogenic plant, to several systems requiring cooling at cryogenic temperature to the level of 4 K and 80 K. The operations of cryolines are normally assumed to be steady state following a cool down from room temperature to the cryogenic temperature. It is to be noted that in a distributed cryogenic system, especially in a nuclear facility such as ITER having confinement definition due to the regulatory requirements, do also attract the attention in the system design that the release from safety valves cannot be allowed inside a building. Therefore, all safety valves need to be discharged inside a confined space, which is a specific space requiring fulfillment of definition for a cryogenic line. The specificity in such cases is that such cryogenic lines will realize dynamic conditions for each release of safety valves or a combination of safety valves in terms of pressure, temperature and flow, leading to unexpected failures. Such operating scenarios also lead to serious impact on fatigue with a question mark on the reliability. Therefore, one can define such cryolines as Relief Collection Header (RCH) which collects discharged helium and transport it to the appropriate place as defined in the system design. The discharges of cold helium from safety relief discharge ports of equipment can result into significantly unsteady and compressible flow in RCH [1]. The proper design of the RCH has to be supported by detailed dynamic of expected flow phenomena for specific cases. The paper presents the dynamics of cold helium flow inside the typical RCH that has been performed to investigate the variation in flow parameters (pressure, temperature, velocity and density) along the axis of RCH and predictions on its reliability. References: [1] R. Andersson, “Numerical simulation of cold helium safety discharges into a long relief line” ScienceDirect Physics Procedia (2014). 281 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_137 Final Configuration with Assembly Assessment of the 100kV High Voltage Bushing for the Indian Test Facility Dheeraj Kumar Sharma1, Sejal Shah2, M Venkata Nagaraju2, Mainak Bandyopadhyay2, Chandramouli Rotti2, Arun Kumar Chakraborty2 1Institute for Plasma Research, India 2ITER-India, Institute for Plasma Research, India Email: [email protected] The Indian Test Facility (INTF) of Neutral Beam (NB) system is an Indian voluntary effort for the full characterization of the diagnostic neutral beam which is the part of ITER's neutral beam system. The design activities of INTF NB system are completed. The INTF High Voltage Bushing (HVB), which is one of the component of NB system, is designed [1] to connect all the required feedlines, e.g. electrical busbars, RF co-axial lines, diagnostic lines and hydraulic & gas feed lines, carried by the transmission line from the HV deck to the Beam Source of NB system. It forms the primary vacuum boundary and provides 100 kV isolation for INTF beam operation. The entire feedlines pass through a metallic plate of HVB called Dished Head (DH) where all the feedlines converge. The overall diameter of DH is 847 mm which is governed by the diameter of the Porcelain insulator which is meant for 100 kV isolation. The effective diameter where all the feedlines converge at the dished head is ~60 0mm which is quite a challenge to accommodate 26 feedlines each of average diameter 60 mm. Electrical feedlines require Vacuum-Electrical feedthroughs for voltage isolation whereas water and gas lines are considered to be directly welded with the DH except one water line which requires 12 kV voltage isolation with respect to DH. For RF lines, different scheme is considered which includes separate Electrical Feedthrough and Vacuum Barrier. To provide connection to electrical cables of heaters and thermocouples, 4 numbers of multipin vacuum compatible electrical feedthroughs are provided which can accommodate ~250 cables. Due to space constraints, Vacuum-Electrical Feedthroughs are considered to be welded with the DH and therefore they shall be of metal-ceramic-metal configuration to allow welding. To avoid undue loading on the ceramic part, the feedlines are supported additionally at DH using vacuum compatible and electrically insulating material. One more important aspect of the INTF HVB is addressed which is related to the assembly of the INTF HVB with INTF Vessel. During Assembly, INTF HVB will be rotated from vertical to horizontal orientation (as per port orientation on INTF vessel) which requires support to all the feedlines to avoid deflection, in the long unsupported span of the feedlines, due to gravity effect. This paper describes the final configuration with assembly assessment of INTF HVB. References: [1] 100-kV Feedthrough for the Indian Test Facility (INTF) - Design and Analysis, S Shah et al, at Asian Plasma and Fusion Association (APFA), 2013 282 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_143 Preliminary Design of O-mode Radiometer for ITER ECE Diagnostic Suman Danani1, Hitesh Pandya1, Ravinder Kumar1, Max E Austin2, Victor S Udintsev3, Vinay Kumar1 1ITER-India, Institute for Plasma Research, India 2The University of Texas at Austin, USA 3ITER Organization, France Email: [email protected] The Electron Cyclotron Emission (ECE) diagnostic system in ITER provides essential information for plasma control and for evaluating the plasma performance. It measures the electron temperature profile (edge/core), electron temperature fluctuations and radiated power in electron cyclotron frequency range from the plasma. These measurements yield information about important plasma parameters [1] such as NTM, TAE, δT/Te, βp, ELM associated temperature perturbations and runaway electrons and are vital for understanding the evolution of plasma, thereby contributing significantly to transport studies and plasma confinement. The spatially resolved temperature measurement in the first harmonic ECE frequency range from 122-230 GHz (for BT = 5.3 T) is obtained by using an O-mode heterodyne radiometer. From the ITER measurement requirements, the electron temperature profile needs to be measured with a spatial resolution of ~6.7 cm, temporal resolution of 10 ms and accuracy of 10% in the plasma core, for the temperature range 0.5 – 40 keV. The principal limitations of the system are restricted radial region of observation due to harmonic overlap and degraded spatial resolution due to the relativistic broadening. The ECE frequency range 122-230 GHz is very wide and it is difficult to cover this wide frequency band by one radiometer, due to technological challenges in achieving wide bandwidth for the mixers. So, the present radiometer design has been optimized by considering four receivers, each of bandwidth ~ 30 GHz which can provide reliable temperature measurements. The splitting of frequency band into four receiver bands is efficiently achieved by considering a combination of quasi-optical and waveguide diplexers, optimizing power loss and cross-talk between the channels. The target spatial resolution is achieved by choosing Radiometer IF filter bandwidth of 1-2 GHz. Further, the radiometer is designed to achieve noise temperature < 10 eV. In this paper, the present design and performance of O-mode Radiometer will be discussed. The simulated ECE Radiation temperature profile using the ECESIM code [2], the effects limiting the ITER ECE measurement [3] and the power loss due to ECE will also be presented. References: [1] V. S. Udintsev et al., Proc. FEC (San Diego) IAEA-CN-197/ITR/P5-41 (2012) [2] M. E. Austin and H.K.B. Pandya, Report FRC-534, University of Texas Fusion Research Center. [3] S. Danani, Hitesh Kumar B. Pandya, P. Vasu, M. E. Austin, Fusion Science & Technology, Vol. 59, 4, May 2011, 651-656. 283 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_146 System Upgradation for Surface Mode Negative Ion Beam Extraction Experiments in ROBIN Kaushal Pandya1, Gourab Bansal1, Jignesh Soni1, Agrajit Gahlaut1, Ratnakar Kumar Yadav2, Mahesh Vuppugalla1, Himanshu Tyagi2, Kanubhai Parmar1, Hiren Mistri2, Jignesh Bhagora1, Bhavesh K Prajapati1, Kartik J Patel1, Manar Bhuyan2, Mainak Bandyopadhyay1, Arun Kumar Chakraborty2, 1Institute for Plasma Research, India 2ITER-India, Institute for Plasma Research, India Email: [email protected] ROBIN (Replica Of BATMAN source in India) is a replica of BATMAN source of IPP, Garching [1], [2], [3]. Plasma production (inductively coupled, RF produced plasma), plasma diagnostic (langmuir probe, optical emission spectroscopy), negative ion beam extraction in volume mode with reduced extraction area of 2 cm2 (4 apertures) using small bench top type power supply (10kV, 400mA), with increase extraction area of 73 cm2 (146 apertures) and using actual power supplies (Extraction Power Supply System, EPSS (11kV, 35A), and Accelerator Power Supply System, APSS (35kV, 15A)) and beam diagnostic etc have been performed successfully in ROBIN. Now, the negative ion source, ROBIN, has been prepared for surface mode experiments with cesium. In surface mode, the metallic cesium is injected into source which helps in enhancing the negative ion production by surface process. For the same, a cesium oven has been designed, fabricated, tested and calibrated prior to installation in ROBIN. Cesium oven has been installed in ROBIN with all necessary equipment and instrumentation. For optimum performance of the source, the cesium feeding at a typical flow rate of ~10 mg/hr is required. In order to avoid any cold regions and proper recirculation of the cesium in the source and uniform deposition on the plasma grid (plasma facing grid) source components temperature are kept around 50-60°C. To achieve this, a heat transfer unit has been integrated with ROBIN which supplies the warm water to the source components. Cesium being reactive makes cesium compounds easily and gets contaminated. To avoid cesium contamination, source is vented using argon gas and the filling pressure is controlled by a pressure switch. A Doppler shift spectroscopy, beam dump for beam current measurements and visible cameras for beam viewing have been installed for the beam diagnostic. A spectroscopic diagnostic of cesium line emission has been implemented for the measurement of the cesium inventory in the source. 284 | P a g e 10th Asia Plasma & Fusion Association Conference This paper will describe the details of the system upgradation for surface mode negative ion experiments and its performance in ROBIN. References: [1] E. Speth. et al, “Overview of the RF source development programme at IPP Garching”, Nucl. Fusion 46 (2006) S220–S238 [2] M.J. Singh et al, “RF - Plasma Source Commissioning in Indian Negative Ion Facility”, AIP Conference Proceedings, Volume 1390, pp. 604-613 (2011) [3] G. Bansal, et al, “Negative ion beam extraction in ROBIN”, Fusion Eng. Des., Volume 88, Issues 6–8, October 2013, Pages 778–782 285 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_147 Thermo-mechanical Design Methodology for ITER Cryo-distribution Cold Boxes Vinit Shukla1, Pratik Patel1, Hiten Vaghela1, Jotirmoy Das1, Nitin Shah1, Ritendra Bhattacharya1, Hyun-sik Chang2, Biswanath Sarkar1 1ITER-India, Institute for Plasma Research, India 2ITER-Organisation, France Email: [email protected] The ITER cryo-distribution system is in charge of the proper distribution of the cryogen at required mass flow rate, pressure and temperature level to the users namely; the superconducting magnets and cryopumps. The cryo-distribution also acts as a thermal buffer in order to run the cryo-plant as much as possible at a steady state condition. A typical cryo-distribution cold box is equipped with mainly liquid helium bath with heat exchangers, cryogenic valves, cold circulating pump and cold compressor. During the intended operation life of ITER, several loads on the cryo-distribution system are envisaged, these are, gravity/assembly loads, nominal pressure/temperature, test pressure/temperature, purge pressure, thermo-mechanical loads due to break of insulation vacuum, transport acceleration and seismic loads. Single loads or combinations of them can act on the cryo-distribution system and its components; therefore, it is very important to analyze the behavior of the system and components under the influence of these loads or combinations. Possible load combinations for the cryo-distribution system will be analyzed and will lead to the basis of the design. This paper will focus on the understanding of the nature of the loads and their combinations for the ITER cryo-distribution as well as their impacts on the design. A representative model of a cold box is considered on which the load combinations have been applied in order to understand their impacts on the design of the cryo-distribution. Also the worst-impact loads or their combination which drive the design of cryo-distribution cold boxes will be derived. 286 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_148 Preliminary Design of Bellows for the DNB Beam Source by EJMA & FE Linear Analysis Shobhit Trapasiya1, Venkata Nagaraju Muvvala2, Rambilas P2, Dheeraj Kumar Sharma3, Roopesh Gangadharan2, Chandramouli Rotti2, Arun Kumar Chakraborty2 1Pandit Deendayal Petroleum University, India 2ITER-India, Institute for Plasma Research, India 3Institute for Plasma Research, India Email: [email protected] In piping system, U-shaped Bellows are widely used among flexible elements. In general, bellows are typically design for Fatigue behavior according to the EJMA standard based on empirically generated fatigue curves. The present work proposes a methodology in the design of bellows by design by analyses and validates its design by EJMA standard. A linear FE approach is chosen to in line with the EJMA standard. The proposed methodology is benchmarked with the available literatures. The same practice is implemented in the preliminary design of a U- shaped bellows in the water line circuits of DNB beam source. DNB Beam Source is a negative ion source-based neutral beam generator for ITER operates at 100KV. The beam divergence (intrinsic) and magnetic fields from ITER torus causes deflection of beams. This calls for beam optic alignment, which are assured by BS Movement mechanism system. To accomplish the above movement requirements, bellows, which is a stringent of its kind (± 22 mm axial, ± 45 mm lateral within 400mm available space with single ply), is designed between the beam source and possible rigid interface-cooling lines coming from HVB. The paper describes right from conceptual stage to preliminary design. Optimization tools are adopted in the selecting bellow dimensions using MATLAB. At the end a coordinated approach between FE based assessment (in ANSYS) and widely applied code, EJMA is implemented for the validation of design and found FE approach is a very conservative than later in the present case. 287 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_154 Evolving the Inspection Techniques for determination of Volumetric Dimensions of Ground Pore in Heat Transfer Elements Hitesh Kumar Kantilal Patel1, Jainish Topiwala3, Kedar Bhope2, Alpesh Patel2, Chandramouli Rotti1, Arun Kumar Chakraborty1 1ITER-India, Institute for Plasma Research, India 2Institute for Plasma Research, India 3Pandit Deendayal Petroleum University, India Email: [email protected] Ground Pore is the inherent defect observed in the weld joint of Heat Transfer Element (HTE) made up of CuCrZr where the two subcomponents are joined in lap configuration. It is therefore essential to ensure that such defects would not affect the desired function of Heat Transfer Element during its operation. A study has been initiated to assess the behavior of the ground pore by simulating the heat flux & other operational parameters on the welded sample cut from HTE. Determination of the volumetric dimensions of the pores before and after application of HTE operational condition is primary essential requirement to understand the behavior of the pores. It was assessed during initial efforts that, it is almost impossible to get the volumetric dimensions of the pores with the help of conventional volumetric examination methods in partial penetration joint configuration, where the expected pore dimensions are as low as 100 microns. Therefore, advanced nondestructive examination (NDE) techniques like computer tomography (CT) having accuracy of detecting defect up to 1 micron was explored and applied for the purpose. Considering the present generation devices, thickness up to 20mm for Copper alloys can be investigated, which meets the requirement for most applications in fusion devices. Apart from this, approach of using combination of two NDE techniques like radiography & water submerged ultrasonic techniques was also explored to determine defect volume. In this paper, the approach taken to establish & validate the inspection technique to determine the volumetric dimensions of weld defect in partial penetration configuration shall be presented. 288 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_159 Significance of ITER IWS Material Selection and Qualification Bhoomi K Mehta1, Haresh A Pathak1, Gurlovleen Singh Phull1, Rahul Kumar Laad1, Abha Maheshwari1, Jigar Raval1 1ITER-India, Institute for Plasma Research, India Email: [email protected] In-Wall Shielding (IWS) is one of the important components of ITER Vacuum Vessel (VV) which fills the space between double walls of VV with cooling water. Procurement Arrangement (PA) for IWS has been signed with Indian Domestic Agency (IN DA). Procurement of IWS materials, fabrication of IWS blocks and its delivery to respective Domestic Agency (DA) or ITER Organization (IO) are the main scope of this PA. Hence, INDIA is the only country which is contributing to VV IWS among all seven ITER partners. The main functions of the IWS are to provide Neutron Shielding with blanket, VV shells and water during plasma operations and to reduce ripple of the Toroidal Magnetic Field. To meet these functional requirements IWS blocks are made up of special materials (Borated Steels SS304 B4 & SS304 B7, Ferritic Steels SS 430, Austenitic Steel SS 316 L (N)-IG, XM-19 and Inconel-625) which are qualified, reliable and traceable for the design assessment. The choice of these materials has a significant influence on performance, maintainability, licensing, detailed design parameters and waste disposal. The main reasons for the materials selected for IWS are its high mechanical strength at operating temperatures, water chemistry properties, excellent fabrication characteristics and low cost relative to other similar materials. The materials are qualified by ASTM or EN standards with additional requirements as described in RCC-MR code 2007 and ITER requirements. Agreed Notified Body (ANB) has control conformity of materials certificates with approved material specification and traceability procedure for Safety Important Component (SIC). The procurement strategy for all the IWS materials has been developed in close collaboration with IO, ANB and Industries as per Product Procurement Specification (PPS). The R&D for sample, bulk material production, testing, inspection and handling as required are carried out by IN DA and IO. At present almost all IWS materials (~2500 Tons) has been procured by IN DA with spares to manufacture ~9000 IWS blocks. This paper summarizes IWS material selection, qualification and procurement processes in detail. 289 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_196 ITER ECE Diagnostic: Design Progress of IN-DA and its Role for Physics Study Hitesh Kumar Pandya1, Ravinder Kumar Jakhmola1, Suman Danani1, Shrishail B Padasalagi1, Sajal Thomas1, Vinay Kumar1, G Taylor2, A Khodak2, W L Rowan3, S Houshmandyar3, V S Udintsev4, N Casal4, M Walsh4 1ITER-India, Institute for Plasma Research, India 2Princeton Plasma Physics Laboratory, USA, 3Institute for Fusion Studies,The University of Texas at Austin, USA, 4ITER Organization, France Email: [email protected] The ECE Diagnostic system in ITER will be used for measuring the electron temperature profile evolution, electron temperature fluctuations, the runaway electron spectrum, and the radiated power in the electron cyclotron frequency range (70-1000 GHz), These measurements will be used for advanced real time plasma control (eg. steering the electron cyclotron heating beams), and physics studies. The ITER ECE Diagnostic system has two measurement views: one radial and the other oblique. The diagnostic system consists of two sets of port plug optics, two high temperature (~ 700 oC) calibration sources, two polarization splitter units, four sets of broadband long transmission lines and ECE radiation measurement Instruments (Michelson Interferometers and heterodyne radiometers). The scope of the Indian domestic agency (IN-DA) is to design and develop the polarizer splitter units, the broadband (70 to 1000 GHz) transmission lines, a high temperature calibration source in the Diagnostics Hall, two Michelson Interferometers (70 to 1000 GHz) and an O-mode Radiometer (122-230 GHz). The remainder of the ITER ECE diagnostic system is the responsibility of the US domestic agency and the ITER Organization. The polarization splitter unit consists of a Gaussian beam telescope with wire grid polarizer selector and the transmission system that includes straight waveguide sections, miter bends, a vacuum window and some quasi-optical components. Waveguide sections are joined together to transmit the emission to the Diagnostics Hall nearly 40 meters away from the port plug optics. The required transmission loss ≤15 dB (up to 400GHz) and ≤ 22 dB (for 400 to 1000 GHz) is a significant design challenge. The high throughput Michelson interferometer with frequency resolution ≤ 3.75 GHz and scanning repetition rate ≥ 50 Hz in a low vacuum is yet another design challenge. The design also needs to conform to the ITER Organization’s strict requirements for reliability, availability, maintainability and inspectability. Progress in the design and development of various subsystems and components considering various engineering challenges and solutions will be discussed in this paper. Enhancing the understanding of plasma physics using various measurements of ECE diagnostics will also be highlighted. 290 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_234 Manufacturing Experience of an ‘Angled’ Accelerator Grid for DNB Beam Source Jaydeep Joshi1 1ITER-India, Institute for Plasma Research, India Email: [email protected] The acceleration system of Neutral Beam Source (BS) is composed of water cooled Copper Oxygen Free CuOF multi aperture grid systems which is designed for focusing of the beamlets to a point located at 20.665 m from the grounded grid. The focusing is obtained using a combination of segment bending and aperture offsets. In the vertical direction, the segments 1 and 2 are bent by 0.549° and 1.647° respectively so that the center line of each segment points to the focal point. In the horizontal direction, grid segment is to be shaped in horizontal direction (over length of ~825mm) to have angles in two stages (i.e. 0.222°, 0.665°). Manufacturing of this kind of ‘Bend Segment’ has been undertaken for the first time to the best of author’s knowledge and therefore, the need arose to establish a method to achieve these angles. Moreover, each of the apertures are to be drilled perpendicular to their own plane which calls for complex machining on angled plate and with very tight tolerances on positions (50 microns) to meet the operational needs. Further, there is a need for high degree of planarity (40 microns) and its stability with very thin material being left after milling of water channels. The case is even more stringent and demanding in case of Plasma Grid as it has scooping of material and balance thickness in some sections is as low as 1mm. To address to the above issue and assess the interdependence of manufacturing operation (i.e. milling of water cooling channel, aperture drilling, copper electro deposition, material scooping, bending of plate / machining of plate to achieve desired angle, stages of stress relieving / annealing) a full scale prototype of plasma grid has been manufactured and significant data is now available on the manufacturing tolerances and handling of angled grid. This information generated out of this experience provides a recipe for the best practices for manufacturing the accelerator for NB system for ITER and upcoming devices. The paper shall present the technical data generated out of manufacturing this full scale prototype grid, summarizing the recommendations on: optimize machining of apertures, machining surfaces at desired angle, handling during manufacturing, handling and transportation, checking of process reliability and identifying the measurement techniques. 291 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_243 Preparation and Analysis of Helium Purge Gas Mixture to be used in Tritium Extraction System of LLCB TBM V Gayathri Devi1, Deepak Yadav1, Amit Sircar1 1Institute for Plasma Research, India Email: [email protected] Hydrogen isotopes are extracted from the ceramic breeder and liquid breeder zones of Lead Lithium Ceramic Breeder (LLCB) Test Blanket Module (TBM) with Helium purge gas. 1000 ppm of hydrogen gas is mixed with the purge helium gas to facilitate improved extraction of hydrogen isotopes due to hydrogen swamping reactions. An experimental set-up is developed for making up the purge gas mixture with a composition similar to the purge gas composition provided at the inlet of the ceramic breeder zones and detritation column of LLCB TBM. This is achieved by introducing different ppm levels (500-5000 ppm) of hydrogen in helium gas by flow control mechanism. The analysis of the purge gas mixture is performed using a highly sensitive gas chromatograph system. In this work, parametric analysis is performed to optimize the process parameters viz., flow rates, temperatures etc. for achieving the required mixture of hydrogen and helium. This paper describes the detailed design of the experimental set-up along with parametric analysis results leading to optimized experimental conditions. 292 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_254 Seismic Design of ITER Component Cooling Water System-1 Piping Aditya Prakash Singh1, Mahesh Jadhav1, Lalit Kumar Sharma1, Dinesh Kumar Gupta1, Nirav Patel1, Rakesh Ranjan1, Guman Gohil1, Hirenkumar A Patel1, Jinendra Dangi1, Mohit Kumar1, A G Ajith Kumar1 1ITER-India Institute for Plasma Research, India Email: [email protected] The successful performance of ITER machine very much depends upon the effective removal of heat from the in-vessel components and other auxiliary systems during Tokamak operation. This objective will be accomplished by the design of an effective Cooling Water System (CWS). The optimized piping layout design is an important element in CWS design and is one of the major design challenges owing to the factors of large thermal expansion and seismic accelerations; considering safety, accessibility and maintainability aspects. An important sub-system of ITER CWS, Component Cooling Water System-1 (CCWS-1) has very large diameter of pipes up to DN1600 with many intersections to fulfill the process flow requirements of clients for heat removal. Pipe intersection is the weakest link in the layout due to high stress intensification factor. CCWS-1 piping up to secondary confinement isolation valves as well as in-between these isolation valves need to survive a Seismic Level-2 (SL-2) earthquake during the Tokamak operation period to ensure structural stability of the system in the Safe Shutdown Earthquake (SSE) event. This paper presents the design, qualification and optimization of layout of ITER CCWS-1 loop to withstand SSE event combined with sustained and thermal loads as per the load combinations defined by ITER and allowable limits as per ASME B31.3, This paper also highlights the Modal and Response Spectrum Analyses done to find out the natural frequency and system behavior during the seismic event. 293 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_256 Role of Outgassing of ITER Vacuum Vessel In-Wall Shielding Materials in Leak Detection of ITER Vacuum Vessel Abha Maheshwari1, Haresh A Pathak1, Bhoomi K Mehta1, Gurlovleen Singh Phull1, Rahul Laad1, Moin Shaikh1, Siju George2, Kaushal Joshi2, Ziauddin Khan2 1ITER-India, Institute for Plasma Research, India 2Institute for Plasma Research, India Email: [email protected] ITER Vacuum Vessel (VV) 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. On a leak detector there will be a spillover 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 10-9 Pa m3s-1 and hence the maximum Outgassing Rate of IWS Materials should comply with the maximum Outgassing Rate required for hydrogen i.e. 1 10-7 Pa m3s-1m-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 of samples, description of the system, Outgassing Rate Measurement of these samples to ensure the accurate leak detection. 294 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_257 Manufacturing and Assembly of IWS Support Rib and Lower Bracket for ITER Vacuum Vessel Rahul Laad1, Yatin Sarvaiya1, Haresh A Pathak1, Raval Jigar2, Chang-ho Choi2 1ITER-India, Institute for Plasma Research, India 2ITER-Organization, France Email: [email protected] ITER Vacuum Vessel (VV) is made of double walls connected by ribs structure and flexible housings, space between these walls is filled up with In Wall Shielding (IWS) blocks to (1) shield neutrons streaming out of plasma and (2) reduce toroidal magnetic field ripple. These blocks will be connected to the VV through a supporting structure of Support Rib (SR) and Lower Bracket (LB) assembly. SR and LB are two independent components manufactured from SS316L (N)-IG material using water jet cutting followed by CNC machining. Water jet cutting is used to prevent Heat Affected Zone, while CNC machining is required to meet the desired surface roughness. Total 1584 support ribs and 3168 lower bracket of different sizes and shapes will be manufactured for the IWS. Two lower brackets will be welded with one support rib to make an assembly. The welding between SR and LB is a full penetration welding by combining Tungsten Inert Gas (TIG) welding and Shielded Metal Arc Welding (SMAW). K type weld joint has been selected for assembly to minimise the welding distortion and a unique welding fixture has been designed to facilitate this weld joint. This unique fixture has an arrangement of rotation of assembly and maintaining appropriate flow of purging gas (Argon) to minimise the welding defects and distortion. Total 1584 assemblies of different sizes and shapes will be manufactured within fastened tolerance to support IWS blocks in the VV. Various mock ups have been manufactured to establish and validate the manufacturing processes, welding and inspection procedures. Process qualification documents (WPS, PQR and WPQR) have been developed. With sufficient experience gained from manufacturing and testing of mock ups, final manufacturing of IWS support rib and lower bracket has been started at the site of IWS manufacturer M/s. Avasarala Technologies Limited, India. This paper will describe, optimization of water jet cutting speed on IWS material, selection criteria for K type weld joint, unique features of fixture designed for SR and LB assembly, manufacturing of Mock ups, welding process to minimise distortion, and the status of manufacturing of SR, LB and Assembly. 295 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_260 Finite Element Analysis for ITER Ferromagnetic In-wall Shielding Block Moinuddin Shaikh1, Haresh A Pathak1, Raval Jigar2, Tailhardat Oliver3 1ITER-India Institute for Plasma Research 2ITER-Organization, France 3Assystem EOS, France Email: [email protected] The In-wall shielding (IWS) located between two shells of the vacuum vessel is part of the vacuum vessel of ITER. The function of the IWS is to provide neutron shielding and to reduce toroidal magnetic field ripple. The material of plates in IWS blocks are SS 304 B7, SS 304 B4 and SS 430. The IWS plates are fastened using M30 bolts to hold them securely and the IWS blocks are mounted to the support ribs using the brackets and M20/M24 bolts. The IWS blocks are subjected to various loads during Vacuum Vessel operation and off-normal condition. It is essential to evaluate design strength of IWS block and individual IWS components. This paper discusses about analysis carried out using ANSYS in three consecutive load steps (1) Pretension on M30 (2) Pretension on M30 and M20 and (3) Pretension on M30 and M20 along with Electromagnetic forces, dynamic forces, Seismic forces, thermal load etc. The stresses of individual IWS components are evaluated against their allowable stress limits as per ASME III Div. 1 for each load step. Stresses and displacements pattern as well peak stress values for the concerned regions are evaluated and discussed here. The results show the stresses and displacements are within allowable limits with safe margin, this confirms the design. Other IWS blocks can also be analysed with similar steps of the analysis and using their own loading conditions. 296 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_264 Development of XM-19 Fasteners for the IWS Blocks Assemblies Sunil Dani1, Gurlovleen Singh1, Haresh A Pathak1, Jigar Raval 2, Chang-ho Choi2 1ITER-India Institute for Plasma Research, India 2ITER-Organization, France Email: [email protected] Fasteners of XM-19 material were developed for the first time in INDIA for the IWS block assemblies for the ITER Vacuum Vessel. Total quantity of fasteners required for the IWS block assembly is around 97000. Fasteners are manufactured from chromium-manganese-nickel austenitic stainless steel type XM-19,UNS S20910 bars in accordance with A479/A479M-04 Standard Specification for Stainless Steel Bars and Shapes for Use in Boilers and Other Pressure Vessels (identical to SA- 479/SA-479M ASME Edition 2007). The high strength, corrosion resistance, and low magnetic permeability of this alloy allow it to be used for IWS block assembly. XM-19 possesses strength and corrosion resistance that is higher than stainless steel grades 316, 316/316L, 317, and 317/317L. M30146_LM_Bolt (Flange) with head thickness 19mm, M30145_NM_Bolt (Flange), M30139_NM_Bolt (Flange), M30 Nut, M2032 Cap Screw, M2046 Cap Screw (Flange) and M2458 SP_Bolt (Flange) has been developed and Approval for the Bulk production has been given by ITER-INDIA and ITER-IO. M30 Bolts will be manufactured as per ANSI B18.2.3.7M-1979, M30 nut will be manufactured as per ANSI B.18.2.4.1M:2002 and cap screw will be manufactured as per ANSI B18.3.1M- 1986.Development of fasteners for the first time with XM-19 material itself is associated with challenges to acquire the required mechanical properties after heat treatment. Other activities which are important for the manufacturing of fasteners are tolerance to be kept while hot forging, development of die for hot forging, shrinkage allowance, thread rolling, slotting on the threads, and selection of heat treatment method to retain the mechanical properties. The various stages of manufacturing of M30 bolts, M30 nuts and M20 cap crews from raw material to the finished product, challenges faced during manufacturing and how it were resolved will be explained in this paper. 297 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_299 Present design status of Erosion and Tritium Monitor diagnostics for ITER Govindarajan Jagannathan1, Nancy Ageorges2, David Anthoine3, Sharath Delanthabettu4, Roger Reichle1, George Vayakis1 and Michael Walsh1 1ITER Organization, France 2Kampf Telescope Optics GmbH, Munich, Germany 3Bertin Technologies, France 4Indira Gandhi Center for Atomic Research (IGCAR), India Email: [email protected] Due to plasma – wall interaction large amount of erosion, dust production and tritium retention are expected to occur in ITER. These have functional as well as safety implications during the operation of the machine. In fact, there is a safety limit of 1000 kg of dust and 1 kg of retained tritium. Hence, continuous monitoring of all these phenomena by multiple diagnostics is essential. Suits of diagnostics have been planned and are currently being developed to monitor the dust, erosion and tritium amount within ITER. This presentation will highlight the present status of the frontline diagnostics for erosion and tritium retention monitoring, which are under the process of design. Dust monitoring has already been reported elsewhere [1]. Erosion at the divertor targets will be monitored by Speckle interferometry and the thickness of deposition at the baffle region will be measured by Lock – in Thermography method. Tritium retention at the baffle region will be monitored by Laser Induced Desorption cum Residual Gas Analyser. This presentation provides an overview of the measurement requirements, the techniques chosen and the concept of these diagnostics. It concludes with the issues and challenges related to the implementation of these diagnostics and the possible solutions to address them. References: [1] http://pos.sissa.it/archive/conferences/240/026/ECPD2015_026.pdf 298 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 5_81 Study of Structures and Stability in Nitrogen Plasma Jet Nirupama Tiwari1, K C Meher1, Srikumar Ghorui1 1Bhabha Atomic Research Centre, India Email: [email protected] Stability of a dc non-transferred arc plasma jet and its internal structures are important for any application related to material processing like plasma spraying, nano synthesis etc. Plasma jet fluctuation and structure formation inside arc plasma jets occur due to reasons like arc root rotation, power supply fluctuation, air entrainment and interaction between electromagnetic and fluid dynamic body forces. Isolated temperature islands originated through such interactions affects particle trajectory, physical processes and process chemistry in a significant manner. In this paper, plasma jet images are recorded at frame rate 7000 FPS for argon and nitrogen plasma. Images are synchronized with voltage signal using camera trigger signal as a trigger to the digital storage oscilloscope. In the experiment, gas flow rate is varied from 10 lpm to 30 lpm in step of 5 lpm keeping torch power constant. All parameter of the camera (exposure time, aperture, focal length) are kept fixed throughout. It has been observed that the luminous length of the plasma jet decreases with increase in gas flow rate for nitrogen, while the reverse happens for argon. It is also observed that while the plasma jet remains fairly steady for low flow rate, variety of different interesting structures are observed inside the plasma jet at higher flow rates. The intensity variation and intensity contours inside the plasma jet are probed using image analysis software. It has been observed that these structures are relatively independent of the arc voltage but highly dependent on gas flow rate and torch power. Reasons for observed behavior are investigated. As thermal and chemical processes are highly dependent on temperature, observed isolated temperature zones inside the plasma jet are of great importance from application point of view. Plasma blob movement observed inside the jet is used for a rough estimate of the plasma jet velocity. 299 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 5_84 Pesticides Removal from Cabbage using Array of Atmospheric Pressure Plasma Jet Akshay Vaid1, Chirayu Patil1, Ramkrishna Rane1, Subroto Mukherjee1, Sudhir Nema1, Hetal Bhatt2, R V Prasad2 1Institute for Plasma Research, India 2Anand Agriculture University, India Email: [email protected] Cold plasmas found their applications in many societal based problems. One such application is the removal of pesticides from vegetables. As these days farmers put enormous amount of pesticides on the vegetables to protect them from pests. Some of these pesticides remain on the vegetables even if they are washed with water resulting in the contamination of food chain. We have developed an array of atmospheric pressure plasma jet which is useful in decreasing the conc. of pesticides on the surface without affecting the bulk properties. In this paper we will present the effect of plasma treatment on the cabbage doped with known amount concentration of pesticide. We have found that after 9 min plasma treatment, the pesticide concentration goes down to 3 times to the original value. Comparative study with different plasma forming gases (Argon, oxygen, helium) will be shown. 300 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 5_85 Comparison of Gas and Plasma Carburizing of AISI 8620 Low Carbon Steel Alphonsa Joseph1, Ghanshyam Jhala1, Akshay Vaid1, Suryakant Gupta1, Keena Kalaria1, Naresh Vaghela1, Subroto Mukherjee1 1FCIPT-Institute for Plasma Research, India Email: [email protected] Case hardening by carburizing is an old art that has recently experienced a growth sprint in new equipment designs and processes. Plasma carburizing represents a new technology and is being accepted in the heat treating industry. Plasma carburizing differs from conventional gas carburizing process as it is carried out in a vacuum chamber at sub atmospheric pressure. The atmosphere used is acetylene gas. The carbon source is ionized and accelerated to the work pieces due to an electrical potential between the work piece and the surroundings. This is manifested as a glow discharge around the work piece. The glow is very uniform, creating a very uniform carbon profile over the entire surface of the work piece. Plasma carburizing is slightly done at a higher temperature than gas carburizing process. In addition, the glow supplies carbon so effectively that the surface of the work is saturated with carbon for during the carburizing time. This combination shortens the cycle time without having a detrimental effect on the product quality. The work piece has less distortion than conventional carburizing process. Moreover, the glow formed during this process can penetrate surface irregularities much better resulting in a more uniform product. Because, plasma carburizing is not limited by the gases ability to supply carbon to surfaces, it saturates the surface with carbon very quickly. As a result, plasma carburizing process could attain the same carbon gradient. As, this process is carried out in a vacuum chamber, there is no intergranular oxidation as observed in gas carburized work pieces when it was done at higher temperatures. The present project aims to compare gas and plasma carburizing process on AISI 8620 low carbon steel. 301 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 5_92 Experimental Study to Improve Anti-felting Characteristics of Merino Wool Fiber by Atmosphere Pressure Air Plasma Nisha Chandwani1, Purvi Deva1, Vishal Jain1, Sudhir Nema1, Subroto Mukherjee2 1FCIPT-Institute for Plasma Research, India 2Institute for Plasma Research, India Email: [email protected] Felting is an inherent property of wool fibers leading to shrinkage and pilling of garments while laundering. Felting occurs mainly because of presence of outermost hydrophobic cuticle layer having sharp scales. [1] Atmospheric pressure plasma processing of wool offers an eco-friendly technique suitable for industry to impart anti-felting characteristics to wool. Dielectric Barrier Discharge (DBD) is a technique to generate non-thermal plasmas at atmospheric pressure. The present work investigates the effect of high frequency (2.5 MHz) Dielectric Barrier Discharge (DBD) air plasma on surface characteristics of Merino wool as a function of plasma exposure time (5 seconds to 15 seconds). The FE-SEM (Field Emission Scanning Electron Microscopy), EDS (Energy Dispersive X-ray spectrum) and Derivative ATR-FTIR (Attenuated Total Reflection- Fourier Transform Infrared) Spectroscopy are used to study physio-chemical changes induced by plasma. These physio-chemical properties of fibers can be co-related with the felting behavior of the wool fiber. [2] The FE-SEM analysis of wool fiber reveal that after plasma exposure the overlapping scales become smoother and nano-scale roughness is induced on wool fiber surface. This leads to reduction in directional friction of the fibers. The analysis of second order derivative of ATR- FTIR spectrum demonstrate the formation of sulphur-oxygen groups such as bunte’s salt (-S- SO3- ), cysteic acid (-SO3-), cystine monoxide(-SO-S-), cysteine di-oxide (-SO2-S-) after plasma processing. The concentration of these groups is found to increase with plasma exposure time. The EDS analysis shows reduction in sulphur concentration with increase in plasma exposure time. A combined effect of morphological and chemical changes on wool fiber surface results in minimizing the felting of the fibers. References: [1] Liu et.al “Comparative study on the felting propensity of animal fibers”, Textile research journal, Vol. 77, No. 12 (2007) [2] Masukuni Mori et.al “Relationship Between Anti-Felting Properties and Physicochemical Properties of Wool Treated with Low-Temperature Plasma” RJTA Vol. 10, No. 1(2006) 302 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 5_103 Surface Chemistry and Wettability Study of Air Plasma Treated Polyethylene by Atmospheric Pressure Dielectric Barrier Discharge Purvi Deva1, Nisha Chandwani1, Vishal Jain1, Sudhir Nema1, Subroto Mukherjee2 1FCIPT, Institute for Plasma Research, India 2Institute for Plasma Research, India Email: [email protected] Polymeric materials are playing an ample role in various industrial applications such as biomedical, automotive, food packaging etc. due to their excellent mechanical properties, easy processing and good resistance to chemicals. Synthetic polymers such as polyethylene (PE) have very low wetting properties and high chemical resistance. Treatment of such polymer surfaces by different types of plasma is often used for modification of wettability, printability, adhesion, durability, stretch resistance, hardness, permeability; wear resistance etc. [1]. In the present work, high frequency (2.5 MHz) Dielectric Barrier Discharge (DBD) air plasma is used to investigate the effect of plasma treatment time on wettability and surface chemistry of polyethylene (PE). PE surface is exposed to air plasma for different time durations from 5-30 seconds. Water contact angle reduces from 101° to 45° in this study, unlike in the case of 50 Hz AC DBD air plasma, where surface exposure time was 30 minutes to achieve water contact angle 70° as reported in our previous work [2]. Partial hydrophobic recovery is observed during extended plasma treatment time. Efforts have been made to understand the phenomena responsible for partial hydrophobic recovery during extended period of plasma treatment time. ATR-FTIR spectroscopy results confirm C-C and C-H bond dissociation followed by formation of double bonds, Low Molecular Weight Oxidized Material (LMWOM) and / or oxygen containing functional groups on PE surface. Scanning Electron Microscopy is done for observing plasma induced morphological changes on the PE surface. Present study gives fair understanding about occurrence of chemical processes on the surface during plasma exposure. Further work to understand the aging behavior of plasma activated PE surface is underway. References: [1] Alina Kaminska et. al, “The influence of side groups and polarity of polymers on the kind and effectiveness of their surface modification by air plasma action,” European Polymer Journal, 38, 1915-1919 (2002). [2] P. Kikani et. al, “comparison of low and atmospheric pressure air plasma treatment” Surface Engineering, 29, 211-221 (2013). 303 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 5_227 Electrical Characteristics of a DC Non-transferrerd Arc Plasma Torch Using Theory of Dynamic Similarity Yugesh V1, Ravi Ganesh2, K Ramachandran3 1Karunya University, India 2Institute for Plasma Research, India 3Bharathiar University, India Email: [email protected] The key component of any industrial thermal plasma system is the plasma torch. Due to its ease of operation, the dc non-transferred plasma torch is more popular than RF or microwave torches. As a result of this and unique properties of the thermal plasma produced, systems employing dc torches have found many industrial applications such as spraying, waste treatment etc. However, there are several experimental parameters such as torch geometry, power, flow, field etc. which influence the torch properties; a generic relationship cannot be constructed that relates the voltage or electro-thermal efficiency to all these parameters. Each class/configuration of plasma torch is unique and its operating regime different. In order to scale up the powers and predict the operating regime, a well-known technique is that of the theory of dynamic similarity, first invented by Russian group [1] and then by others [2, 3]. We have also constructed a functional form relating the voltage and efficiency to all controllable parameters for the torch we use in our laboratory, by using the theory of dynamic similarity. The torch configuration is unique in the sense that it employs all three, viz. gas, wall and magnetic stabilization mechanisms. First, we carried out exhaustive experiments at low powers ~ 25 kW that yielded the current-voltage (C-V) characteristics. The next step involved construction of several dimensionless numbers such as enthalpy number, Reynold’s number, electromagnetic field number etc. by using momentum & energy equation, Maxwell equations and boundary conditions in non-dimensionlized forms. Then the theory of dynamic similarity was combined with experimentally obtained data to build a unique relationship of the form Using this relation, we have been able to predict operational regimes of the same class of torch and has helped us develop similar torches capable of working at higher powers. References: [1] O. I. Yas’ko, “Correlation of the characteristics of the electric arc,” J.Phys.D:Appl.Phys. 2, 733 (733). [2] Gang Li, Wenexia Pan et al “Application of similarity theory to the characterization of a non- transferred laminar plasma jet generation.” Plasma sources science and technology, 14, (2005). [3] A. M. Paingankar et al “Prediction of electrical characteristics dc non- transferred arc torch Theory of dynamic similarity theory,” Plasma sources Sciences Technology, 8, 100-109, (199). 304 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 5_241 Design and Development of 20 kW, 45 kV, 30 kHz Power Supply for Study of Pulsed Dielectric Barrier Discharges Surender Kumar Sharma1, Anurag Shyam1,2 1Bhaba Atomic Research Center-Visakhapatnam, India 2Insttitute forPlasma Research, India Email: [email protected] Dielectric barrier discharges are frequently used for industrial [1], environmental [2] and biomedical application [3, 4] such as for UV sources, ozone production, toxic gas treatment, water treatment surface treatment and plasma medicine applications. These discharges are generated inside the insulated chamber placed between parallel plated by applying the pulsed high voltage at a frequency ranging from few 100’s Hz to 1 MHz, the high voltage pulse ionizes the gas in the chamber and produces radiations for various applications. The voltage ranges from 1 kV to 100 kV depending on the gas, dielectric material, geometry and the dimension of the discharge chamber. A high voltage power supply is designed to generate and study dielectric barrier discharges at atmospheric, higher and lower pressures. A 20 kW, 45 kV power supply with the pulse frequency ranging from 1 kHz to 30 kHz is designed. The power supply consists of dc rectifier, high frequency inverter using MOSFET switches switching up to 30 kHz, high voltage transformer and feedback control circuit. The voltage of the power supply can be adjusted from 2 kV to 45 kV. The frequency of the high voltage pulse can also be varied from 1 kHz to 30 kHz with the pulse duration of 1 µs. The rise time and fall time of the high voltage pulse is < 200 ns. The power supply is short circuit proof and can withstand variable load condition from overloads to arcs. The discharge chamber is made of evacuated quartz tube of 50 mm diameter with SS mesh electrodes on the external surface. The design details and the performance of the power will be discussed in the paper. References: [1] Falkenstein Zoran, “Application of dielectric barrier discharge”, IEEE Conf Proc. of 12th International Conference on High Power Particle Beams, BEAMS -98, Vol 1, pp 117-120 (1998) [2] Daniel S.L, “On the ionization of air for removal of noxious effluvia”, IEEE Trans. on Plasma Science, 30 (4), 1471 – 1481 (2002) [3] Weltmann K D, Von Woedtke T“Campus PlasmaMed – From basic research to clinical proof”, IEEE Trans on Plasma Science, 39 (4), 1015- 1025 (2011) [4] Kim Y et. al., “Plasma apparatus for biomedical applications”, IEEE Trans. on Plasma Science, 43 (4), 944 – 950 (2015) 305 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 5_280 Plasma Sterilization for Bio-decontamination Suryakant Gupta1, Sudhir Nema1 1Institute for Plasma Research, India Email:[email protected] Due to continuous emergence of new infectious microorganisms, particular attention should be paid to avoid iatrogenic diseases by minimizing the contamination of medical instruments with infectious microorganisms. It is well known that one of the most effective ways to prevent hospital-acquired infection is to implement a sterilization and disinfection system that includes physical and chemical inactivation methods. Conventional sterilization techniques, such as those using autoclaves, ovens and ethylene oxide (EtO) have certain drawbacks while sterilizing heat sensitive devices. EtO adsorb on the surface of the device and has many side effects when it come in contact with human organs. Plasma sterilization technique is emerging rapidly in the world for effective killing of thermally stable microorganisms such as spores, viruses and prions. Plasma of specific gas mixture is a source of Hydroxyl radicals, Hydroperoxyl radical, UV radiations and reactive oxygen species such as atomic oxygen etc. Combined effect of all these is a good recipe to kill microorganism present in the system. The basic mechanism behind this process is based upon rupture of cell membrane and breaking of DNA molecules of microorganism. This paper presents information on the current status and future perspectives of a state-of-art plasma sterilization technique, initial work carried out at FCIPT, IPR and our future plan to develop compact plasma sterilizer for the safe sterilization of medical devices. 306 | P a g e 10th Asia Plasma & Fusion Association Conference Oral Session-2 307 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 1_56 Superficial Layer MHD Effect and Full-cover Free Surface Flow Characterization Zengyu Xu1, Chuanjie Pan1, Xiujie Zhang1, Lu Peng1 1Southwestern Institute of Physics, China Email: [email protected] Up to now, no realistic liquid metal (LM) free surface flow has been successfully used in magnetic confinement fusion devices because of MHD instability and unavoidable rivulet flow of the free surface. Recently, after performing a guidable free curve-surface flow investigation theoretically and experimentally, seeking for other way to get a full-cover free surface flow is in implementing. The superficial layer MHD effect, rivulet flow enhancement effect by magnetic field and thin film flow rivulet effect are experimentally observed. Compared with the experimental results and the characteristic parameters of the free surface flow, new variables of surface cover ratio and rivulet flow index are introduced to characterize the flowing characteristic of the full-cover free surface flow under magnetic field. According to the analysis rule, there are different unique conditions to meet full-cover free surface flow for different liquid metal under a magnetic field. Meanwhile, one inherent full-cover free surface flow is addressed for alternative application to liquid metal plasma facing component system. The experiments were carried out at Liquid Metal Experimental Loop Upgrade (LMEL–U) facility in Southwestern Institute of Physics, China. The free surface flow was measured 58 mm in width and 900 mm in length. The flowing angle is 60 degree to gravity direction in order to differentiate the effect of MHD from gravity for the flow under a gradient magnetic field. The average velocity of the free surface flow is from 0.4 to 4.34 m/s. The magnetic field is from 0 to 1.851 Tesla. To seek for the best free surface flow, the thickness of free surface flow was designed from 1 mm to several millimeter. Due to a limitation by the current liquid metal fluid diagnosis technology, the free surface flow is recorded by normal and super high speed camera. 308 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_25 Fast Wave Scrape-off Layer Losses of Tokamak Plasmas in Minority, Mid/High Harmonic, and Helicon Heating Regimes Nicola Bertelli1, E F Jaeger2, J C Hosea1, C Lau3, R J Perkins1, C K Phillips1, G Taylor1 1Princeton Plasma Physics Laboratory, USA 2XCEL Engineering Inc, USA 3Oak Ridge National Laboratory, USA Email: [email protected] This paper examines fast wave propagation and power loss in the scrape-off layer (SOL) of tokamak plasmas by using the full wave code AORSA, with the edge plasma beyond the last closed flux surface (LCFS) included in the solution domain and with a collisional damping parameter used as a proxy to represent the real, and most likely nonlinear, damping processes. In [1], 2D and 3D AORSA results for the low aspect ratio National Spherical Torus eXperiment (NSTX), show a strong transition to higher SOL power losses (driven by the RF field) when the FW cut-off is removed from in front of the antenna by increasing the edge density. This result is consistent with previous NSTX observations [2] and it will be further verified in the upcoming NSTX-Upgrade (NSTX-U) experimental campaign [3]. Here, full wave simulations have been extended to “conventional” tokamaks with higher aspect ratios, such as the DIII-D, Alcator C-Mod, and EAST devices. DIII-D results show behavior similar to that found in NSTX and NSTX-U, and consistent with previous DIII-D experimental observations. In contrast, a different behavior is found for Alcator C-Mod and EAST, which unlike NSTX/NSTX-U and DIII-D that operate in the mid/high harmonic regime, operate in the minority heating regime. In the minority heating regime AORSA results indicate lower SOL power losses with increasing density in front of the antenna, in agreement with the experimental observation that increasing the density in front of the antenna leads to better antenna-plasma coupling. The effect of the pitch angle of the magnetic field and a comparison of the minority heating and mid/high harmonic heating regimes are presented. It is found that for NSTX-U scenarios the behavior of the RF field in the SOL region changes with plasma current. Finally, the impact of the SOL region on the evaluation of helicon current drive efficiency in DIII-D is presented and compared to the other heating regimes mentioned above. References: [1] N. Bertelli et al., Nucl. Fusion, 54, 083004 (2014). [2] C. K. Phillips et al., Nucl. Fusion, 49, 075015 (2009). [3] R. J. Perkins et al., work presented in this conference APFA 2015. 309 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 1_55 Manufacturing and process research of the WEST ICRH antenna Qingxi Yang1 1Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), China Email: [email protected] An important issue for the WEST (Tungsten (W) Environment in Steady-state Tokamak) project, which aims at modifying Tore Supra to an X-point divertor machine, is to provide a heat flux at 10MW/m2 during 1000 sec and 20MW/m2 during 20s. To obtain this level of flux, the operation of three Ion Cyclotron Resonant Heating (ICRH) launchers at a level of 9MW during 30s or 3MW during 1000s is necessary. The WEST ICRH system has to deal with two challenging issues that no other ICRH system before ITER has faced simultaneously so far, i.e. ELMs resilience and Continuous Wave (CW) RF operation. Three antennas have the same structure and components, with front face components (Faraday screen and straps, matching capacitors), the matching unit (capacitors, bridge and actuating system), the feeding line (impedance transformer and vacuum window), the external structure (real flange, bottom fame, auxiliary systems) and the instrumental devices (RF probes, arc detection system, reflectometer waveguides and feed through). The current WEST ICRH antenna was designed based on a former tested load-resilient “2007 prototype”, and optimized to improve the coupling performance while adding CW operation capability by introducing water cooling in the whole antenna. The present WEST ICRH antenna design aim to bridge the operational gap, and also technological gap towards the ITER ICRH antenna. Since WEST ICRH antenna has a CW operation requirements under high power, like the ITER ones, it forces to cope with high level of specifications for the manufacturing, like material choice, high precision machining process. Thus, it requires specific fixture tools and jigs designed for fabrication and assembly, optimized welding process study for reducing deformation, assembly workflow optimization. This paper is mainly focused on the manufacturing of the WEST ICRH antenna components, currently under fabrication in collaboration with CAS/ASIPP. Based on the high accuracy required for the ICRH components, machining process is introduced, followed by assembly, welding, qualification tests, 3D scanning and metrological analysis. 310 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 1_104 Recent Progress of the ECRH System on HL-2A Shaodong Song1, Mei Huang1, He Wang1, Jun Rao1, Bo Lu1, Zihua Kang1, Mingwei Wang1, Kun Feng1, Chao Wang1, Jieqiong Wang1, Jiruo Ye1, Feng Zhang1, Xiaolan Zou2, Gerardo Giruzzi2, Roland Magne2, Zhongbing Shi1, Qingwei Yang1, Weimin Xuan1, Xuru Duan1 1Southwestern Institute of Physics, China 2 CEA, IRFM, France Email: [email protected] The electron cyclotron resonant heating (ECRH) system is one of the most important heating methods for magnetic confinement fusion device. It has been widely used in plasma heating, current drive, sawteeth tailoring, NTM control and current profile shaping etc, due to its localized heating and highly controllable characteristics. The ECRH system on HL-2A tokamak has been equipped step by step since 2005, and has been upgraded to 5MW in 2012. The total power of 5MW is provided by six 0.5 MW/68 GHz/1s gyrotrons and two 1MW/140(105) GHz/2s gyrotrons. The ECRH system is composed of power source, transmission line and antenna. The gyrotrons are equipped with depressed collectors, which are produced by GYCOM Ltd. For 68GHz gyrotrons, BN barrier windows are used, and for 140(105) GHz gyrotrons, CVD diamond windows are taken due to the higher power. The output beam is horizontal linearly polarized Gaussian beam after the matching-optical-unit (MOU). The conversion efficiency from electricity to RF wave is 50% for both types of gyrotrons. Helium-free superconducted magnet is used to provide magnetic field. For the 68GHz gyrotrons, the transmission line is equipped with 80mm diameter corrugated waveguides; for the 140(105) GHz gyrotrons, the transmission line uses 63.5mm diameter evacuated corrugated waveguides. Measurement of RF power is carried out with calorimeter method, which is located at MOU and RF window in front of antenna. A variety of polarization can be achieved with polarizers equipped in the transmission lines. Two antennas are used to deliver the RF power into the plasma, each integrating four beams, capable of changing the poloidal and toroidal injection angles. The poloidal angle can be tuned in real time for the purpose of MHD control. The maximum injected power into plasma is 2.5 MW for the 3MW/68GHz system, while the 2 MW/140(105) GHz system is not yet in use which needs appropriate magnetic field. With the 3MW ECRH system, many physical experiments have been carried out. First H-mode on HL-2A is achieved in 2009 with 0.8 MW neutral beam injection (NBI) and 1.2MW EC power. Maximum electron temperature of 4.93 keV is obtained with 1.57 MW EC power. The newly developed NTM real-time control system firstly demonstrated its capacity on (2, 1) tearing modes control. The density pump-out effect in ECRH phase is quite clear during divertor discharges, which provides a possible regime to remove helium ash in future fusion reactor. The modulation frequency can be up to 500Hz, and series of modulation experiments have been done to pursue the transport issues in different areas of plasma. Evidence of transition from L- to H- mode in purely ECRH heated plasma has been observed. Intensive electron cyclotron current drive (ECCD) Experiments have been done and comparison between theoretical simulation and experimental results shows good agreement. The ECRH system has also been applied in assisted start-up, which saves the flux consumption of ohmic coils. Together with neutral beam, neoclassical tearing modes have also been observed under high beta discharges. 311 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 1_109 The 3.7GHz LHCD System on HL-2A Bo Lu1, Roland Magne2, Xingyu Bai1, Hao Zeng1, Yali Chen1, Chao Wang1, Emmanuel Bertrand2, Lena Delpech2, Annika Ekedahl2, Julien Hillairet2, Jun Liang1, Jieqiong Wang1, Zhihua Kang1, Kun Feng1 and Jun Rao1 1Southwestern Institute of Physics, China 2CEA, IRFM, France Email: [email protected] A 3.7 GHz lower hybrid current drive (LHCD) system was built on HL-2A in 2014. A Passive- Active Multijunction (PAM) concept antenna is installed and there are 16 active and 17 passive waveguides in each row. The peak parallel refractive index is 2.75 with a low theoretical Reflection Coefficient (RC). The antenna is fed by four high power pulsed klystrons TH2103A. The klystrons are protected from the reflected power by high power circulators. The RF power is transmitted to the launcher in TE10 propagation mode through the rectangular waveguide (WR284) transmission lines. The transmission line is pressurized with 2 bar of nitrogen to prevent arcing. The schematic layout of the system is shown in Fig.1. The 4 klystrons are fed by a pulse step modulation (PSM) high voltage power supply (HVPS). The fast switch-off time is less than 10 microseconds. A simple method to control the beam current of the klystron was developed. Unlike the traditional beam control method using anode modulator based on a non-linear high voltage tetrode, the anode is fed by a simple voltage divider. The beam is then controlled by the cathode voltage. Both the power supply and the beam current feedback control component are simplified. The stable working region was investigated by experimental study. The transmitters are commissioned on matched loads, the total output power reaches 2 MW when the pulse duration achieves 2s. The coupling experiment is carried out at a relative lower power level. The reflected power decreased to less than 10% by changing the parameters of the plasma and gas puffing near the front of the antenna. The loop voltage was decreased, the stored energy and the total radiated power increase significantly and supra thermal electrons were observed during lower hybrid wave injection. The maximum injected power reaches 800kW after the system commissioning. 312 | P a g e 10th Asia Plasma & Fusion Association Conference Oral Session-3 313 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 0_42 Observation of Up-Down Asymmetry in Impurity Line Emissions from the Ergodic Layer of Large Helical Device Tetsutarou Oishi1, Shigeru Morita1, Xianli Huang2, Hongming Zhang2, Motoshi Goto1 1National Institute for Fusion Science, Japan 2Graduate University for Advanced Studies, Japan Email: [email protected] Development of diagnostics for edge impurity emission profiles contributes quantitative evaluation for a total amount of impurity radiation for the discharges in which impurity ions play significant roles in the edge plasmas such as divertor detachment discharges. Therefore, we conducted a vacuum ultraviolet (VUV) spectroscopy diagnostics to measure the spatial profiles of impurity emissions released from edge plasmas in Large Helical Device (LHD). The edge plasma of the LHD is characterized by stochastic magnetic fields with three-dimensional structure intrinsically formed by helical coils called the “ergodic layer [1],” while well-defined magnetic surfaces exist inside the last closed flux surface (LCFS). Line radiations from impurity ions in the ergodic layer are significantly emitted in the VUV wavelength range because the electron temperature around the LCFS ranges from 10 to 500 eV. A space-resolved spectroscopy using a 3 m normal incidence VUV spectrometer was developed to measure the VUV emission profiles in wavelength range of 300-3200 Å from impurities in the ergodic layer [2]. The emission intensity, the ion temperature, the impurity ion flow, and their vertical profiles are derived by measuring the Doppler profile of impurity line spectra. The optical axis of the spectrometer is arranged perpendicular to the toroidal magnetic field at holizontally-elongated plasma cross section. The observation range covers the full vertical profile of the emission from plasmas. The VUV spectroscopy has revealed that intensity profiles of the impurity emission from the ergodic layer have strong up-down asymmetries, namely, the emission intensities are quite different between top and bottom plasma edges. In this paper, observations of the up-down asymmetries are summarized and its dependence on the experimental parameters, such as the electron density, position of the magnetic axis, and direction of the toroidal field, is discussed on the following impurity line emissions: (1) CII 1335.71 Å (2s-2p), CIII 977.02 Å (2s-2p), and CIV 1548.20 Å (2s-2p) from intrinsic carbon impurity ions sputtered from the carbon divertor plates, (2) NeVII 465.22 Å (2s-2p), NeVIII 770.41 Å (2s-2p), ArVII 585.75 Å (3s-3p), and ArVIII 700.24 Å (3s-3p) from Ne or Ar ions introduced in plasmas by gas puffing, and (3) WVI 639.68 Å (5d-6p) from W ions introduced in plasmas by pellet injection. This work was partially conducted under the LHD project financial support (NIFS14ULPP010), Grant-in-Aid for Young Scientists (B) 26800282, and the JSPS-NRF-NSFC A3 Foresight Program in the field of Plasma Physics (NSFC: No.11261140328, NRF: No.2012K2A2A6000443). 314 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 1_290 Current Status of Safety design and Safety Analysis for China ITER Helium Coolant Ceramic Breeder Test Blanket System Long Zhang1, Qixiang Cao1, Yanling Wang1, Yanjing Chen1, Qijie Wang1, Fengchao Zhao1, Fen Wang1, Xinghua Wu1, Xiaoyu Wang1, Kaiming Feng1 1 Email: [email protected] Helium Coolant Ceramic Breeder (HCCB) Test Blanket System (TBS) designed by China are planned to be tested in ITER to validate key technologies, including demonstration of nuclear safety, for future fusion reactor breeding blankets. Furthermore, in order to be operated in ITER, a nuclear facility (INB) recognized by French nuclear safety authority, safety design and safety analysis of the TBS are mandatory for the licensing procedures. This paper summarizes the status at current design phase with following main elements: The main radiological source terms in the system are tritium and activation products. Nuclear and tritium analysis are performed to identify their inventories and distributions in system. Multiple confinement barriers are considered to be the most essential safety feature. French regulation for pressure equipment and nuclear equipment (ESP/ESPN regulations) will be followed to ensure the system integrities. ALARA principle is kept in mind during the whole safety design phases. Protective actions including choice of advanced materials, improvement of shielding, optimization of operation and maintenance activities, usage of remote handling operations, zoning and access control have been considered. Passive safety is emphasized in the system design, only minimal active safety functions including call for fusion plasma shutdown and isolation of TBM from ex-vessel ancillary systems. High reliability and redundancies are required for components related to these functions. Several accidents have been identified and analyzed. Consider the limited inventories in the system and the intrinsic safety of fusion device, positive conclusions have been obtained. 315 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_91 Destructive Analysis on the ITER FW Small Scale Mock-ups Pinghuai Wang1, Jiming Chen1, Danhua Liu1, Fanya Jin1, Bo Yang1 1Southwestern Institute of Physics, China Email: [email protected] As one of the core components of ITER, the first wall (FW) panel of shield blanket defines a physical boundary for the plasma transients and exhausts the majority of the plasma heat flux. China will undertake 12.64% of FW manufacturing tasks, and all of them are enhanced heat flux (EHF) components which will suffer surface heat flux of 4 - 5MW/m2. The FW will be manufactured by a combination technology of explosion bonding CuCrZr alloy/316L (N) stainless steel plate and hot iso-static pressing (HIP) joining of beryllium tiles/CuCrZr alloy. The Be/Cu joint qualities is the key issue for the manufacturing of the FW panels. Several small scale mock-ups were manufactured for the qualification of the HIP technology for the FW. To avoid the brittle Be-Cu phase formed during the HIPing process, different thick Ti and pure Cu were coated on the beryllium tiles before HIPing to CuCrZr alloy. Ultrasonic testing was conducted on the mock-ups and destructive analysis was carried out on the mock-ups. For the failed ones, the results show that in the UT indication area brittle fracture occurs at the Be/Ti interface and then Ti/Cu interface in other areas. Based on these results, the manufacturing technology was improved mainly on the beryllium tiles quality, coating process and canister design. 316 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 1_131 EAST Articulated Maintenance Arm for EAST and WEST Yong Cheng1, Yuntao T Song1, Eric Villedieu2, Vincent Bruno2 1Institute of Plasma Physic, Chinese Academy of Sciences, China 2CEA, IRFM, France Email: [email protected] A project to upgrade the Articulated Inspection Arm (AIA) [1] into a fully operational robot EAMA (EAST Articulated Maintenance Arm) has been set between IRFM and ASIPP in an associated laboratory. EAMA consists of an articulated serial arm with 7 degrees of freedom (DOF) and a 3-DOF gripper. The total length is 8.867 meters. As planned, it will work in Experimental Advanced Superconductor Tokamak (EAST) and W/Tungsten Environment in Steady-state Tokamak (WEST) vacuum vessel (VV) to perform a remote inspection and maintenance tasks after plasma shutting down. The EAMA system will be demonstrated under EAST conditioning, namely ultra-high vacuum and temperature conditions [2]. The robot system has been extensively upgraded. The effort has been focused on three areas: 1) Increasing of the 3-DOF gripper, which was developed to inspect the condition of PFCs and remove the debris dropping flexibly from the first wall; 2) Two kinds of supervisor software will be used , one is built based on Actin system, a commercial off-the-shelf package, enriched with specific functionalities, the other is built based on Robot Operating System (ROS), permissive licensing and collaborative environment, excellent robot development platform. 3) Improvement of the robot algorithm system, feedback algorithm and visual servo control algorithm have been developed to increase the operational stability and robustness of the grasping task with high efficiency [3]. The aim of this paper is to detail the architecture of the EAMA system and present the obtained results of the test campaign. References: [1] Perrot Y, Cordier J J, Friconneau J P, et al. ITER articulated inspection arm (AIA): R&D progress on vacuum and temperature technology for remote handling [J]. Fusion Engineering & Design, 2005, 75:537-541. [2] Shi S S, Song Y T, Cheng Y, et al. Design and Implementation of Storage Cask System for EAST Articulated Inspection Arm (AIA) Robot [J]. Journal of Fusion Energy, 2015, 34:1-6. [3] L. L. Lin, Y. T. Song, Y. Yang, et al. Computer vision system R&D for EAST Articulated Maintenance Arm robot [J] Fusion Engineering & Design 317 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 0_88 Improvements in a Tracer-Encapsulated Solid Pellet and Its Injector for More Advanced Plasma Diagnostics Naoki Tamura1, Shigeru Sudo2, Chihiro Suzuki1, Hisamichi Funaba1, Masaru Takagi3, Nakahiro Satoh3, Hiromi Hayashi1, Hiroya Maeno1, Mitsuhiro Yokota1, Hideki Ogawa1 1National Institute for Fusion Science, Japan 2Chubu University, Japan 3Hamamatsu Photonics Inc., Japan Email: [email protected] Although we are facing the age of the International Thermonuclear Experiment Reactor (ITER), many physics issues related to the confinement of magnetically-confined toroidal plasma still remain to be clarified. For example, under some conditions, impurities inside the magnetically- confined toroidal plasma tend to accumulate into the core region of the plasma. This will cause a dilution of fusion fuel. Moreover, a radiation loss from the core plasma will be enhanced due to the impurity accumulation, and then the temperature in the core region will be decreased dramatically. Consequently, fusion plasma performance will be degraded below the acceptable level. In order to develop strategy for obviating and suppressing the impurity accumulation, it is significantly important to gain a full understanding of the impurity transport in the magnetically- confined toroidal plasma. In consideration of such a situation, we have developed a Tracer- Encapsulated Solid Pellet (TESPEL) [1, 2] for promoting a precise study of the impurity transport. To put it plainly, the TESPEL is a double-layered impurity pellet. This form enables us to produce a both poloidally and toroidally localized “tracer” impurity source in the plasma, and to specify the total amount of the tracer impurity deposited in the plasma precisely. In this contribution, we introduce new-type TESPELs [3], which are greatly improved in regard to the above-mentioned features. Owing to this improvement, we have achieved a shallower penetration of the TESPEL into the plasma with sufficient quantities of the tracer particles, which can be measured with the existing diagnostics. In addition, we also introduce a new TESPEL injector, which enables us to inject the TESPEL obliquely into the plasma. This injector can also contribute to a further shallower penetration of the TESPEL into the plasma. Moreover, we will discuss a future strategy of the TESPEL in the research of fusion plasma and plasma application. References: [1] S. Sudo, “Diagnostics of Particle Transport by Double-Layer Pellet,” J. Plasma Res, 69, 1349 (1993). [2] S. Sudo and N. Tamura, “Tracer-encapsulated solid pellet injection system,” Rev. Sci. Instrum., 83, 023503 (2012). [3] N. Tamura, “Creation of Impurity Source inside Plasmas with Various Types of Tracer- Encapsulated Solid Pellet,” Plasma Fusion Res., 10, 1402056 (2015). 318 | P a g e 10th Asia Plasma & Fusion Association Conference Oral Session-4 319 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_297 Simulation and Modeling of Magnetic Field Dynamics in Laser Plasma Interaction Amita Das1, Chandrashekhar Shukla1, Atul Kumar1, Bhavesh Patel1, Predhiman Krishan Kaw1 1Institute for Plasma Research, Inida Email: [email protected] The dynamical evolution of magnetic field plays an important role in variety of contexts ranging from astrophysical phenomena to laboratory plasmas. It is well known that when a high power laser impinges on an overdense plasma target (and/or solid which can get ionized to form a plasma) it generates energetic electrons. The current due to these energetic electrons are balanced by the return plasma current from the background electrons. It is believed that the Weibel destabilization of the two currents leads to the magnetic field generation. This has been illustrated by the Particle – in – Cell simulations (PIC) of periodic infinite plasma medium. In these studies the realistic role of finite laser spot size resulting in an electron beam of finite transverse extent were not considered. With the help of PIC simulations it has been shown by us that the development of Kelvin Helmholtz (KH) instability at the beam edge occurs much faster than the usual Wiebel destabilization process leading to magnetic field generation having a typical scale size of the transverse extent of the beam. The Weibel mediated magnetic field generation on the other hand gets generated at the short skin depth scale. This difference yields interesting differences in the magnetic field turbulent characteristics which will be discussed in detail in the talk. 320 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_74 Electrical Transverse Transport in Lorentz Plasma with Strong Magnetic Field and Collision Effect Baisong Xie1, Chong LV1, Ziliang Li 1Beijing Normal University, China Email: [email protected] In inertial confinement fusion (ICF), the spontaneous magnetic field formed from laser interacting with the pellet may reach few hundreds of Megagauss (MG) which results in the cyclotron frequency at the same order of the collision frequency [1,3]. Electrical transverse transport in this case would become very important so that we study it by the Boltzmann equation for different electron density distribution. For the Maxwell distribution, it is shown that transport coefficients decrease with the increase of (the ratio of to ), which means the electrons would be highly collimated by strong magnetic field. This is attributed to that the electron’s gyroradius is smaller than the collisional mean free paths [2, 3]. Moreover, the electrical transverse transport is also studied for quasi-monoenergy distribution with different width , which is different from the Maxwell one. It is found that the transport coefficients decrease greatly as quasi-monoenergy degree increases. In particular when approaches to zero, i.e. the Delta distribution with almost perfect monoenergy electron density, the electric conductivity doesn’t change while the thermal conductivity decreases with . On the other hand the smaller the is the less amount the transverse transport exhibits. Our study indicates that they are beneficial to limit the electric transverse transport. References: [1] E. M. Lifshitz and L. P. Pitaevskii, Physical Kinetics (World Publishing Corporation, Beijing, 1999). [2] M. A. Bake, B. S. Xie, S. Zhang and H. Y. Wang, “Energetic protons from ultraintense laser with a symmetric parabolic concave target”, Phys. Plasmas, 20, 033112 (2013). [3] H. B. Cai, S. P. Zhu and X. T. He, “Effects of the imposed magnetic field on the production and transport of electron beams”, Phys. Plasmas, 20, 072701 (2013). 321 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 2_155 Spectroscopy of Laterally Colliding Plasma Plumes in Laser-blow-off of Thin Film: Role of Energetic Neutrals in Formation of Interaction Zone Ajai Kumar1, Bhupesh Kumar2, Rajesh Kumar Singh1 1Institute for Plasma Research, India 2Weizmann Institute of Science, Israel Email:[email protected] Laser Laser-blow-off (LBO) plasma plumes formed by two spatially separated laser beams has been studied using optical emission spectroscopy and fast imaging technique. The two parallel expanding plasma plumes lead to the formation of an interaction zone in between the seed plasmas. Dynamics, geometry and optical features of both seed as well as interaction zone are investigated. Transport mechanism of seed plasma species to the interaction zone and consequently the plausible formation mechanism of interaction zone are briefly described. In contrast to conventional laser, our spectral analysis suggest that that fast neutral formed by charge exchange with fast ions play the important role in generation of interaction zone [1,2]. Dominance of neutral emission and depletion of ionic emission in the interaction zone are agreeing with the simulation of emission lines at similar plasma parameters. References: [1] Bhupesh Kumar, R.K. Singh, Sudip Sengupta, P. K. Kaw, and Ajai Kumar, Phys. of Plasma 21, 083510 (2014). [2] Bhupesh Kumar, R K Singh, Sudip Sengupta, P K Kaw and Ajai Kumar, Phys. of Plasma 22, 063505 (2015). 322 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 0_34 Thermionic Divertors for Tokamaks Avinash Khare1, Sanjay K Mishra2, Predhiman Krishan Kaw2 1University of Delhi, India 2Institute or Plasma Research, India Email: [email protected] Thermionic emission and conversion is suggested as a viable technique to cool divertor plates in a tokamak and remove the excessive heat flux from the tokamak scrape off layer (SOL) recycling nearly half of it into electrical power with good efficiency (~55% of the corresponding Carnot efficiency). The thermionic divertor, for this purpose consists of divertor plate which is directly heated by the concentrated high heat flux from the SOL to a high temperature ( 2500K ) and a collector plate maintained at lower temperature. Outer side of the divertor plate is fabricated with micro-hemispherical tips and is coated with low work function material to enhance the thermionic emission losses and the consequent cooling of the divertor plate. The electrical circuit is completed via an external load connected to divertor and collector plates, which gives electrical power output. In fusion reactor producing 3GW of fusion power, the thermionic divertor removes 600MW from SOL and recycles approximately 270MW of it into electrical power via direct conversion. 323 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 3_181 Modeling of ITER Disruption scenarios using TSC Indranil Bandyopadhyay1, Amit K Singh 1ITER-India, Institute for Plasma Research, India Email: [email protected] Plasma Disruptions are one of the major concerns in ITER as they would subject the ITER vacuum vessel and other first wall components to largest electromagnetic and thermal loads. Thus these components have to be designed and built to withstand these forces during disruptions. The input to their design comes through accurate simulations of these events to predict the possible magnitudes of halo and eddy currents flowing through the first wall components and the vacuum vessel. The Tokamak Simulation Code (TSC) [1] has been used over many years to simulate disruptions and vertical displacement events (VDEs) in many different tokamaks and has been well validated against experimental data in those machines. Even then, uncertainties remain over some of the critical parameters, which determine the peak halo currents, namely the width and temperature of the halo region and their evolution during the disruptive events. In the absence of any physics model to determine these parameters, empirical models are used to best fit experimental observations. Recently the ITER model in TSC has been fine-tuned to match the earlier predictions done for ITER using the DINA code [2]. Presently the halo current model as also the halo current diagnostics model in TSC is being refined to have better predictions for ITER. A new halo current diagnostics has recently been added to space resolve the halo current that flows along the open flux lines in the plasma halo region to the first wall and vacuum vessel as a function of distance from the separatrix location in the first wall. This can then be directly compared and validated against tile current measurements in existing tokamaks as also in ITER. Cases of slow and fast current quenches in ITER following major central disruptions, depending on post thermal quench plasma temperature are simulated using TSC. The peak halo current and their spatial resolution on the first wall/blanket modules are presented. This model will be validated against experimental data in DIII-D and CMOD, which is already underway. References: [1] S. C. Jardin, N. Pomphrey and J. Delucia, J. Comput. Phys. 66 (1986) 481 [2] S. Miyamoto et al, Nucl. Fusion 54 (2014) 083002 324 | P a g e 10th Asia Plasma & Fusion Association Conference Oral Session-4 325 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 4_118 Technical Developments and Present Status of the ITER Cryolines and Cryo-distribution Systems Biswanath Sarkar1, Nitin Shah1, Hitensinh Vaghela1, Ketan Choukekar1, Pratik Patel1, Himanshu Kapoor1, Srinivasa Muralidhara1, Jotirmoy Das1, Uday Kumar1, Anuj Garg1, Vinit Shukla1, Mohit Jadon1, Vikas Gaur1, Bikash Dash1, Shk Madeenavalli1 1ITER-India, Institute for Plasma Research, India Email:[email protected] ITER cryolines and cryo-distribution system is an important and critical link between ITER cryoplants and end users, which are mainly the superconducting magnets and cryopumps. The overall system design has considerably evolved and matured now from the baseline 2010 design. The 5 km of ITER cryolines are spread over tokamak building, plant bridge and cryoplant area of ITER site. The sizes of cryolines vary from DN100 to DN1000 with number of process pipes up to 7 in single cryoline which carries either cold helium or nitrogen at varying temperature, pressure and flow rates as per the process and functional requirements of end users. The cryolines are divided in to 2 groups, X (complex cryolines with number of process pipe per line typically more than 3) and Y (comparatively simpler cryolines with number of process pipe per line typically less than or equal to 3). Each group is further divided in to 5 lots (X1 to X5, Y1 to Y5) for ease of project execution in terms of design and production activities. ITER cryo-distribution system, which mainly manages and controls the primary cooling loops of ITER end users, consists of four number of auxiliary cold boxes (ACB) for cooling loop of superconducting magnets and structures, one ACB for cooling loop of cryopumps, one cold valves box for cooling loop of thermal shield system of tokamak and one cryoplant termination cold box (CTCB), which acts as an interconnection between cryoplants, liquid helium tank, test cryostat, 80 K plant and other cold boxes of ITER cryo-distribution system. The cold circulators inside ACBs plays an important role of maintaining the required pressure and flow rate of cold helium inside cooling loops of superconducting magnets and cryopumps. Preliminary design of lot Y1 cryolines as well as final design of lot Y2 cryolines is completed while preliminary design of lot X3 cryolines and CTCB is ongoing. The prototype cryoline (by one industry) as well as cold circulators (by two industries) have been designed, manufactured and cold tested. The paper describes the major activities, achievements and current status of the cryo-distibution and cryolines project as well as summarizes the outcome of the prototype tests of the components. References: [1] B. Sarkar et al, “Value Engineering in System of Cryoline and Cryo-distribution for ITER: In-kind Contribution from India,” Advances in Cryogenic Engineering, Volume 58, in press. 326 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 1_289 Cryogenic Technology of the New Millennium – Competence of DH Industries Ronald Den Heijer1 1DH Industries, The Netherlands Email: [email protected] Two decades of the present new millennium has experienced exciting development in the field of cryogenic application especially in the field of high temperature superconductivity for large scale application. Application of large capacity magnets in Plasma and Fusion application enhanced the importance of an efficient cryogenic support system for its right level of performance. Being strongly associated with the world of cryogenics for last six decades DH Industries BV, The Netherlandshas developed itself as an unmatched competence center through its products, knowledge and support services right at your doorstep. This presentation will share interesting details of some of the latest projects carried out in the mentioned segment along with information on new possibilities of association. 327 | P a g e 10th Asia Plasma & Fusion Association Conference Abstract ID: 1_207 Upgradation Plans of SST-1 Cryogenics System at IPR Vipul L Tanna1, SST-1 Cryo Team1, Subrata Pradhan1 1Institute for Plasma Research, India Email: [email protected] Steady State Superconducting Tokamak (SST-1) is India’s First Superconducting Tokamak and has Toroidal (TF) and Poloidal (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 facilities the cooling down of the cold mass and cyo-stable operation of SST-1TF magnets. SST-1 experimental campaigns have revealed that the existing plant is just sufficient for the heat loads acting on the plant. 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 coil. Detailed estimates have been made and it has been found that an additional ~ 850 W at 4.5 K of cryo power is required towards (a) cooling all the PF magnets (b) the cooling down and the operation of a new Nb3Sn based central solenoid of SST-1. This paper will elaborate on (a) the experimental heat loads acting on the cryo system (b) the `thermal runaway amongst the PF magnets observed in the SST-1 campaign’ (c) the robust need of a higher operation pressure up to 2.1 bar (a) (d) the need of the flow optimizations as per the hydraulic paths (e) the engineering solutions at each of these described (a)-(d) above. 328 | P a g e