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10th Asia Plasma & Fusion Association Conference
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
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Codes for respective disciplines
CODE Title
0 Magnetic and Inertial Confinement
1 Fusion Engineering and Technology including Reactor 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
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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 Jacquinot
1
1ITER Organisation, France
Email: [email protected]
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 Yamada
1, Satoru Sakakibara
1
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 Lee
1, Kihak Im
1, Hyoung Chan Kim
1, Yong-Seok Hwang
1
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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Invited Talk (Session-2)
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Abstract ID: 4_286
ITER Implementation and Fusion Energy Research in China
Jing Zhao1, Zhaoliang Feng
1, Changchun Yang
1
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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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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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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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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Poster Session-1
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Abstract ID: 0_3
Data Handling System for SST-1
Harish Masand1, Manisha Bhandarkar
1, Aveg kumar
1, Hitesh Kumar Gulati
1, Kiritkumar B Patel
1,
Kirti Mahajan1, Jasraj Dhongde
1, Hiteshkumar Chudasama
1, Subrata Pradhan
1
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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Abstract ID: 0_7
The Role of Equilibrium Flows in Temperature-Gradient-Driven Modes in Hot Tokamaks
Deepak Verma1, Aditya K Swamy
1, Rajaraman Ganesh
1, Stephan Brunner
2, Laurent Villard
2
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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Abstract ID: 0_8
3D Character of Plasma Transport in the Aditya Limiter Scrape-off Layer
Bibhu Prasad Sahoo1, Devendra Sharma
1, Ratneshwar Jha
1, Yuhe Feng
2
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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Abstract ID: 0_10
Fast Visible Imaging and Study of Edge Turbulence in the Aditya Tokamak
Santanu Banerjee1, Ranjana Manchanda
1, Malay Bikas Chowdhuri
1, Nilam Ramaiya
1, Navin
Parmar1, Joydeep Ghosh
1, Rakesh L Tanna
1, Braj Kishore Shukla
1, Pramod K Sharma
1, 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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Abstract ID: 0_26
An Overview of Experimental ICRF Research on NSTX-U
Rory Perkins1, Joel Hosea
1, Nicola Bertelli
1, John Caughman
2, Cornwall Lau
2, Cynthia Phillips
1,
Gary Taylor1, James Wilson
1
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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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 Hu
2, Yeong-kook Oh
3
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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Abstract ID: 0_73
Comparison of Different Atomic Databases used for Evaluating Transport Coefficients in Aditya Tokamak
Malay Bikas Chowdhuri1, Joydeep Ghosh
1, Santanu Banerjee
1, Ranjana Manchanda
1, Nilam
Ramaiya1, Parveen Kumar Atrey
1, Y Shankara Joisa
1, Rakesh L Tanna
1, Prabal K
Chattopadhyay1, Chet Narayan Gupta
1, Shailesh B Bhatt
1, Motoshi Goto
2, Izumi Murakami
2
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
spatial profile of brightness of Be-like oxygen (O4+
) spectral line (2p3p 3D3-2p3d
3F4) 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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Abstract ID: 0_89
Neutral Particle Profiles during ICRH Experiments in Aditya Tokamak
Nilam Ramaiya1, Ritu Dey
1, Ranjana Manchanda
1, Malay Bikas Chowdhuri
1, Santanu Banerjee
1,
Niral Virani1, Rakesh L Tanna
1, Jayesh V Raval
1, Y Shankara Joisa
1, Parveen Kumar Atrey
1,
Shailesh B Bhatt1, Chet Narayan Gupta
1, Sanjay V Kulkarni
1, Prabal K Chattopadhyay
1, 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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Abstract ID: 0_96
Understanding of Impurity Behavior in SST-1 Plasmas Using Visible Spectroscopy
Ranjana Manchanda1, Nilam Ramaiya
1, Malay Bikas Chowdhuri
1, Santanu Banerjee
1, Joydeep
Ghosh1, SST-1 Team
1
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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Abstract ID: 0_110
Observation of Plasma Shift in SST-1 using Optical Imaging Diagnostics
Manoj Kumar Gupta1, Chesta Parmar
1, Vishnu K Chaudhari
1, Ajai Kumar
1, SST-1 Team
1
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
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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 Chowdhuri
1, Ratneshwar Jha
1, Parveen Kumar Atrey
1, Y
Shankar Joisa1, Joydeep Ghosh
1, Rakesh L Tanna
1, Aditya Team
1
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
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Abstract ID: 0_129
Ponderomotive Density Modulation in Two Ion Tokamak Plasma
J K Atul1, S K Singh
1, S Sarkar
2, O V Kravchenko
3
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
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Abstract ID: 0_133
Study of Neutral Particle Transport in Aditya Tokamak Plasma using DEGAS2 Code
Ritu Dey1, Joydeep Ghosh
1, Malay Bikas Chowdhuri
1, Ranjana Manchanda
1, Santanu Banerjee
1,
Nilam Ramaiya1, Aditya Team
1
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
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Abstract ID: 0_142
Modeling of Eddy Current distribution and Equilibrium Reconstruction in the SST-1 Tokamak
Santanu Banerjee1, Amit Kumar Singh
2, Deepti Sharma
1, Srinivasan Radhakrishnana
1, Raju
Daniel1, Y Shankara Joisa
1, Parveen Kumar Atrey
1, Surya Kumar Pathak
1, SST-1 Team
1
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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Abstract ID: 0_144
Equilibrium Reconstruction of Plasma Discharges in the Aditya Tokamak
Deepti Sharma1, Santanu Banerjee
1, Amit Kumar Singh
1, Srinivasan Radhakrishnana
1, Raju
Daniel1, Rakesh L Tanna
1, Joydeep Ghosh
1, Y Shankara Joisa
1, Parveen Kumar Atrey
1, Surya
Kumar Pathak1, Aditya Team
1
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
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Abstract ID: 0_149
Ohmic Discharges with Improved Confinement in Tokamak Aditya
Rakesh L Tanna1, Harshita Raj
1, Joydeep Ghosh
1, Prabal K Chattopadhyay
1, Sharvil Patel
2,
Kumarpalsinh A Jadeja1, Kaushal M Patel
1, Shailesh B Bhatt
1, Chet Narayan Gupta
1, Kunal
Shah1, Motibhai Makwana
1, Narendra Patel
1, Vipul K Panchal
1, Chhaya Chavda
1, Pramod
Sharma1, Malay Bikas Chowdhuri
1, Santanu Banerjee
1, Nilam Ramaiya
1, Ranjana Manchanda
1,
Raju Daniel1, Sameer Kumar Jha
1, Kumuduni Tahiliani
1, Praveenlal Edappala
1, Shishir Purohit
1,
Y Shankar Joisa1, Jayesh V Raval
1, C V S Rao
1, Parveen Kumar Atrey
1, Surya Kumar Pathak
1,
Ratneshwar Jha1, Amita Das
1, Dhiraj Bora
1
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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Abstract ID: 0_168
Investigation of Aditya Tokamak Plasmas with Lithiumized Wall
Niral Virani1, Malay Bikas Chowdhuri
1, Kumarpalsinh A Jadeja
2, Joydeep Ghosh
1, Ranjana
Manchanda1, Nilam Ramaiya
1, Santanu Banerjee
1, Jayesh V Raval
1, Y Shankara Joisa
1,
Umeshkumar C Nagora1, Parveen Kumar Atrey, Rakesh L Tanna, Prabal K Chattopadhyay
1,
Chet Narayan Gupta1, Shailesh B Bhatt
1, Aditya team
1
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
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Abstract ID: 0_173
A Study of Anomalous Transportation of Sawtooth Generated Runaway Electrons Observed in ADITYA Tokamak
Harshita Raj1, Joydeep Ghosh
1, Rakesh L Tanna
1, Prabal K Chattopadhyay
1, Raju Daniel
1,
Sameer Kumar Jha1, Jayesh V Raval
1, Y Shankara Joisa
1, Shishir Purohit
1, C V S Rao
1,
Umeshkumar C Nagora1, Parveen Kumar Atrey
1, Malay Bikas Chowdhuri
1, Ranjana
Manchanda1, Yogesh C Saxena
1, Rabindranath Pal
2 and Aditya Team
1
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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Abstract ID: 0_177
Geodesic Acoustic Modes with Poloidal Mode Coupling ad Infinitum
Rameswar Singh1, Ozgur D Gurcan
1
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
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Abstract ID: 0_178
Mean EB Shear Effect on Geodesic Acoustic Modes in Tokamaks
Rameswar Singh1, Ozgur D Gurcan
1
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.
Page 44
10th Asia Plasma & Fusion Association Conference
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Abstract ID: 0_187
Estimation of Vacuum Magnetic Fields due to Ohmic Coils in Aditya Upgrade tokamak
Krishana Kumari K1, Rohit Kumar
1, Rakesh L Tanna
1, Joydeep Ghosh
1, Prabal K
Chattopadhyay1, Srinivasan Radhakrishnana
1, Sharvil Patel
2, Raju Daniel
1, Someswar Dutta
1,
Dhiraj Bora1, Yogesh C Saxena
1, Aditya Team
1
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.
Page 45
10th Asia Plasma & Fusion Association Conference
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Abstract ID: 0_190
Divertor Coil Power Supply in Aditya Tokamak for improved Plasma Operation
Vaibhav Ranjan1, Kunal Shah
1, Motibhai N Makawana
1, Chet Narayan Gupta
1, A
Varadharajulu1, Joydeep Ghosh
1, Rakesh L Tanna
1, Prabal K Chattopadhyay
1, Raju Daniel,
Srinivasan Radhakrishnana1, Yogesh C Saxena
1
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.
Page 46
10th Asia Plasma & Fusion Association Conference
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Abstract ID: 0_203
The First Results of Te Measurement with of Soft X-Ray Diagnostics in SST-1 Tokamak
Jayesh V Raval1, Shishir Purohit
1, Y Shankara Joisa
1, Ajai Kumar
1
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
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Abstract ID: 0_204
An Overview of SST-1 Diagnostics and Results from Recent Campaigns
Ajai Kumar1, Asha N Adhiya
1, Hemchandra C Joshi
1, Janmejay U Buch
1, Jayesh V Raval
1, Jinto
Thomas1, Joydeep Ghosh
1, Kiran Patel
1, Kumar Ajay
1, Kumudni Tahiliani
1, M V Gopalakrishna
1,
Malay Bikas Chowdhuri1, Manoj Kumar
1, Neha Singh
1, Nilam Ramaiya
1, Parveen Kumar Atray
1,
Pabitra K Mishra1, Ratneshwar Jha
1, Raju Daniel
1, Rajwinder Kaur
1, Ranjana Manchanda
1,
Sameer Kumar Jha1, Santanu Banerjee
1, Santosh P Pandya
1, Shishir Purohit
1, Shwetang N
Pandya1, Snehlata Gupta
1, Surya Kumar Pathak
1, Umeshkumar C Nagora
1, Varsha Siju
1, Vishnu
K Chaudhari1, Y Shankar Joisa
1
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
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Abstract ID: 0_205
Design and Development of AXUV-based Soft X-Ray Diagnostic Camera for ADITYA Tokamak
Jayesh V Raval1, Shishir Purohit
1, Y Shankara Joisa
1, Joydeep Ghosh
1, Rakesh L Tanna
1,
Kumarpalsinh A Jadeja1, Ajai Kumar
1, Shailesh B Bhatt
1, Praveena Kumari
1, Vismaysinh Raulji
1,
Minsha Shah1, Rachana Rajpal
1
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.
Page 49
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Abstract ID: 0_210
Conceptual Design of Diagnostics for the HL-2M Tokamak
Qingwei Yang1, Yi Liu
1, Z B Shi
1, HL-2A Diagnostic Group
1
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
magnetic field of BT = 2.2 T, line average electron density of ne = 1.6 1020
m-3,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.
Page 50
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Abstract ID: 0_211
Observation of MHD Phenomenon for SST-1 Superconducting Tokamak
Manisha Bhandarkar1, Jasraj Dhongde
1, Subrata Pradhan
1
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.
Page 51
10th Asia Plasma & Fusion Association Conference
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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 Dhongde
1, Harish Masand
1, Subrata Pradhan
1
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).
Page 52
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Abstract ID: 0_213
Development of New Diagnostics for WEST
P Lotte1, P Moreau
1, C Gil
1, J Bucalossi
1, M H Aumeunier
1, J M Bernard
1, C Bottereau
1, C
Bourdelle1, Y Camenen
2, M Chernyshova
7, F Clairet
1, T Czarski
7, M Choi
6, G Colledani
1, Y
Corre1, X Courtois
1, R Daniel
3, D Davis
3, P Devynck
1, D Douai
1, D Elbeze
1, A Escarguel
2, C
Fenzi1, W. Figacz
7, J C Giacalone
1, R Guirlet
1, J Gunn
1, S Hacquin
1, X Hao
4, J Harris
9, G T
Hoang1, F Imbeaux
1, S Jablonski
7, A Jardin
1, H C Joshi
3, G Kasprowicz
8, C Klepper
9, E
Kowalska-Strzeciwilk7, M Kubkowska
7, A Kumar
3 , V Kumar
3, W Lee
5, B Lyu
4, P Malard
1, L
Manenc1, Y Marandet
2, D Mazon
1, O Meyer
1, M Missirlian
1, D Molina
1, G Moureau
1, Y Nam
6,
E Nardon1, T Nicolas
1, R Nouailletas
1, H Park
5, J Y Pascal
1, K Pozniak
8, N Ravenel
1, R Sabot
1,
F Samaille1, J Shen
4, J M Travere
1, E Tsitrone
1, S Varshney
3, S Vartanian
1, D Volpe
1, F D
Wang4, G Yun
6, W Zabolotny
8 and WEST team
1
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.
Page 53
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Abstract ID: 0_220
Observation on Runaway Discharges in SST-1 Experiments
Kiritkumar B Patel1, Subrata Pradhan
1
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).
Page 54
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Abstract ID: 0_221
Hard X-ray Diagnostic for SST-1
Shishir Purohit1, Jayesh V Raval
1, Y Shankara Joisa
1, Ajai Kumar
1, SST-1 Team
1
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.
Page 55
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Abstract ID: 0_223
Study of MHD Activities in the Plasma of SST-1
Jasraj Dhongde1, Manisha Bhandarkar
1, Subrata Pradhan
1, Sameer Kumar Jha
1, SST-1 Team
1
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)
Page 56
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Abstract ID: 0_230
A Fixed Frequency Reflectometer to Measure Density Fluctuations at Aditya Tokamak
Parveen Kumar Atrey1,2
, Dhaval Pujara2, Subroto Mukherjee
1
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
reflectometer at 22 GHz (cut – off density nc = 6 1012
cm-3
) 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).
Page 57
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Abstract ID: 0_231
Helium Beam Diagnostics for the Estimation Electron Temperature and Density in SST-1
Vishal Pillai1, Neha Singh
1, Jinto Thomas
1, Rajesh Kumar Singh
1, Hem Chandra Joshi
1, 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
Page 58
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Abstract ID: 0_239
Operation of ADITYA Thomson Scattering System: Measurement of Temperature and Density
Jinto Thomas1, Vishal Pillai
1, Neha Singh
1, Kiran Patel
1, Lingeshwari G
1, Zalak Hingrajiya
1,
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.
Page 59
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Abstract ID: 0_240
Installation and Commissioning of SST-1 Thomson scattering system
Jinto Thomas1, Vishal Pillai
1, Neha Singh
1, Kiran Patel
1, Vishnu K Chaudhari
1, Ajai Kumar
1
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.
Page 60
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Abstract ID: 0_249
Limiter and Divertor Systems – Conceptual and Mechanical Design for Aditya Tokamak Upgrade
Kaushal Patel1, Kulav Rathod
1, Kumarpalsinh A Jadeja
1, Shailesh B Bhatt
1, Deepti Sharma
1,
Srinivasan Radhakrishnana1, Raju Daniel
1, Rakesh L Tanna
1, Joydeep Ghosh
1, Prabal K
Chattopadhyay1, Yogesh C Saxena
1, Aditya Team
1
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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Abstract ID: 0_250
Development of Gas Puffing System for LHCD Experiment in Aditya Tokamak
Kumarpalsinh A Jadeja1, Kaushik S Acharya
1, Kaushal M Patel
1, Nilesh D Patel
1, Kalpesh M
Chaudhary1, Shailesh B Bhatt
1, Pramod K Sharma
1, Kirankumar K Ambulkar
1, Pramod R
Parmar1, Chetan G Virani
1, Saifali Dalakoti
1, Arvindkumar L Thakur
1, Rakesh L Tanna
1,
Santanu Banerjee1, Joydeep Ghosh
1
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.
Page 62
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Abstract ID: 0_251
Structural Analysis of New Vacuum Vessel for Aditya Tokamak Upgrade
Kulav Rathod1, Joydeep Ghosh
1, Shailesh B Bhatt
1, Rakesh L Tanna
1, Kumarpalsinh A Jadeja
1,
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.
Page 63
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Abstract ID: 0_267
IGBT Based Active Clamping Protection Scheme for SST-1 PF Coils
Azad Makwana1, Deven Kanabar
1, Chiragkumar Dodiya
1, Kalpesh Doshi
1, Yohan Khristi
1,
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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Abstract ID: 0_268
Thermal Imaging of SST-1 Limiters
Shwetang N Pandya1, Santosh P Pandya
1, Kumar Ajay
1
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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Abstract ID: 0_271
The Upgradation of Aditya Tokamak
Shailesh B Bhatt1, Joydeep Ghosh
1, Rakesh L Tanna
1, Chhaya Chavda
1, Chet Narayan Gupta
1,
Prabal K Chattopadhyay1, Raju Daniel
1, Srinivasan Radhakrishnana
1, Kaushik S Acharya
1,
Kalpesh M Chaudhary1, Someswar Dutta
1, Kumarpalsinh A Jadeja
1, Madan B Kalal
1, Sanjay V
Kulkarni1, Kumari (K) Krishna
1, Moti Makwana
1, Rohitkumar Panchal
1, Vipul K Panchal
1,
Kaushal M Patel1, Narendra Patel
1, Nilesh Patel
1, Sharvil Patel
1, Vijay Patel
1, Harshita Raj
1,
Ramasubramanian Narayanan1, Vaibhav Ranjan
1, Kulav Rathod
1, Devraj H Sadharkiya
1, Kunal
Shah1, Krishnamachari Sathyanarayana
1, Deepti Sharma
1, Pramod K Sharma
1, Braj Kishore
Shukla1, A Varadharajulu
1, Dinesh S Varia
1, Ajai Kumar
1, Ratneshwar Jha
1, Amita Das
1,
Abhijeet Sen1, Yogesh C Saxena
1, Predhiman Krishan Kaw
1, Dhiraj Bora
1
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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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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Abstract ID: 0_277
Development of Non-circular Metal Seal for Aditya Tokamak Upgrade Vacuum Vessel
Kaushik S Acharya1, Kaushal M Patel
1, Kumarpalsinh A Jadeja
1, Kulav Rathod
1, Nilesh D Patel
1,
Kalpesh M Chaudhary1, Shailesh B Bhatt
1, Aditya Team
1
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 ~ 10-9
torr. 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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Abstract ID: 0_298
Study of the plasma SOL with fast reciprocating probe diagnostics on the SST-1 tokamak
M V Gopalakrishna1, , Sameer Kumar Jha
1, , Santanu
Banerjee1, Manoj Kumar Gupta
1, Pramila Gautam
1, Dilip Raval
1, Snehal Jaiswal
1, Pradeep
Chauhan1, Subrata Pradhan
1, SST-1 Team
1
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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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 Dhongde
1, Kirti Mahajan
1, Aveg Kumar
1,
Harish Masand1, Manisha Bhandarkar
1, Hiteshkumar Chudasama
1, 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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Abstract ID: 0_304
Implementation of SST-1 plasma position control using vertical field
Kirti Mahajan1, Jasraj Dhongde
1, Kiritkumar B Patel
1, Hitesh Kumar Gulati
1, Aveg Kumar
1,
Harish Masand1, Manisha Bhandarkar
1, Hiteshkumar Chudasama
1, 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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Abstract ID: 1_6
Preparation of W/CuCrZr Monoblock Test Mock-up using Vacuum Brazing Technique
Kongkham Premjit Singh1, Samir S Khirwadkar
1, Kedar Bhope
1, Nikunj Patel
1, Prakash K
Mokaria1, Mayur Mehta
1
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 Mokaria
1, Samir S Khirwadkar
1, Sunil Belsare
1,
Mohammed Shoaib Khan1, Tushar Patel
1, Deepu S Krishnan
1
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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Abstract ID: 1_13
Thermal Shock Behavior of Tungsten & Tungsten Alloy Materials under Transient High Heat Load Conditions
Shailesh Kanpara1, Samir S Khirwadkar
1, Sunil Belsare
1, Kedar Bhope
1, Rajamannar Swamy
Kidambi1, Prakash K Mokaria
1, Nikunj Patel
1, Tushar Patel
1, Narendra Chauhan
1, 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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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 Sarmah
1, Puspa Sah
1, Joydeep Ghosh
2 and Mayur Kakati
1
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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Abstract ID: 1_18
Performance of Impedance Transformer for High Power ICRF Heating in LHD
Kenji Saito1, Tetsuo Seki
1, Hiroshi Kasahara
1, Ryohsuke Seki
1, Shuji Kamio
1, Goro Nomura
1,
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
defined by Rp=2P(zc/Vmax)2 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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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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Abstract ID: 1_248
Optimization, Commissioning and Operation of EAST Tungsten Divertor
Damao M Yao1, G N Luo
1, L Cao
1, Z B Zhou
1, Q Li
1, W J Wang
1, L Li
1, P F Zi
1
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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Abstract ID: 1_206
Status of the WEST Project
Jérôme Bucalossi1, WEST Team
1
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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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 Humphreys
3, Y K In
1, Y M Jeon
1, J G Kwak
1, J M Kwon
1,
Y U Nam1, Y K Oh
1, J K Park
4, S Sabbagh
5, S J Wang
1, S W Yoon
1, G S Yun
6, 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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Invited Talk (Session-4)
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Abstract ID: 0_287
Progress of Experiment on HL-2A
Xuru Duan1, HL-2A Team
1
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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Abstract ID: 0_293
Recent Progress and Present Status of LHD towards Deuterium Experiment
Tomohiro Morisaki1, M Osakabe
1, M Y okoyama
1, R Sakamoto
1, K Y Watanabe
1, K Nagasaki
2,
M Sakamoto3, S Inagaki
4, S Hamaguchi
1, M Isobe
1
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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Abstract ID: 0_300
Initial Results in SST-1 After Up-gradation
Subrata Pradhan1, Ziauddin Khan
1, Vipul L Tanna
1, Dilip Raval
1, Upendra Prasad
1, Harish
Masand1, Aveg Kumar
1, Kiritkumar B Patel
1, Manisha Bhandarkar
1, Jasraj Dhongde
1, Braj
Kishore Shukla1, Imran Mansuri
1, Yohan Khristi
1, Yuvakiran Paravastu
1, Chet Narayan Gupta
1,
Dinesh Sharma1, Kalpeshkumar R Dhanani
1, Pratibha Semwal
1, Siju George
1, Subrata Jana
1,
Pradip Panchal1, Rohitkumar Panchal
1, Rakeshkumar Patel
1, Hitesh Kumar Gulati
1, Kirti
Mahajan1, Mohammad Shoaib Khan
1, Prashant Thankey
1, Azadsinh Makwana
1, Gaurang
Mehsuriya1, Pradeep Chauhan
1, Arun Parkash A
1, Murtuza Vora
1, Akhilesh Singh
1, Dashrath
Sonara1, Pankaj Varmora
1, G Srikanth
1, Dikens Christian
1, Atul Garg
1, Arun Panchal
1, Nitin
Bairagi1, Manika Sharma
1, Gattu Ramesh Babu
1, Prosenjit Santra
1, Tejas Parekh
1, Hiteshkumar
Patel1, Prabal Biswas
1, Snehal Jayswal
1, Tusharkumar Raval
1, Hiteshkumar Chudasama
1, Atish
Sharma1, Amit Ojha
1, Bhadresh R Praghi
1, Moni Banaudha
1, Ketan Patel
1, Hiren Nimavat
1,
Pankil Shah1, Jayant C Patel
1, Rajiv Sharma
1, A Varadharajulu
1, Ranjana Manchanda
1, Parveen
Kumar Atrey1, Surya Kanth Pathak
1, Y Sankar Joisa
1, Kumudni Tahiliani
1, Manoj Kumar
1,
Santanu Banerjee1, Debashis Gosh
1, Bhoomi Chaudhary
1, Amita Das
1, Dhiraj Bora
1
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 Artaud
1, Vbulandi Basiuk
1, M Bécoulet
1, S Brémond
1, J Bucalossi
1, H
Bufferand1, G Ciraolo
1, L Colas
1, Y Corre
1, X Courtois
1, J Decker
2, L Delpech
1, P Devynck
1, G
Dif-Pradalier1, R P Doerner
3, D Douai
1, R Dumont
1, A Ekedahl
1, N Fedorczak
1, C Fenzi
1, M
Firdaouss1, J Garcia
1, P Ghendrih
1, C Gil
1, G Giruzzi
1, M Goniche
1, C Grisolia
1, A Grosman
1, D
Guilhem1, R Guirlet
1, J Gunn
1, P Hennequin
4, J Hillairet
1, G T Hoang
1, F Imbeaux
1, I Ivanova-
Stanik5, E Joffrin
1, A Kallenbach
6, J Linke
7, T Loarer
1, P Lotte
1, P Maget
1, Y Marandet
8, M L
Mayoral9,10
, O Meyer1, M Missirlian
1, P Mollard
1, P Monier-Garbet
1, P Moreau
1, E Nardon
1, B
Pégourié1, Y Peysson
1, R Sabot
1, F Saint-Laurent
1, M. Schneider
1, J M Travère
1, E Tsitrone
1, S
Vartanian1, L Vermare
4, M Yoshida
11, R Zagorski
5, 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 Mukherjee
1, Jyoti Agarwal
1, Manoahstephen Manuelraj
1,
Paresh Panchal1, Pratik Kumar Nayak
1, Jyoti Shankar Mishra
1, Vrushabh Lambade
1, Pawan
Bairagi1, Vijay Kumar
1, Reena Sayani
1, Srinivasan Kasthurirengan
1, Swarup Udgata
2, 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 Mishra
1, Samiran Shanti Mukherjee
1, Paresh Panchal
1, Pratik
Kumar Nayak1, Hardik Sharma
1, Haresh Patel
1, Pramit Dutta
1, Naveen Rastogi
1, Jyoti Agarwal
1
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 Khirwadkar
1, 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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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.
Max. temp. = 2093 k
Max. thimble
temp. = 1683 k
Dthimble = 25mm
Temperature [k]
Dthimble = 16mm
Max. temp. = 1843 k
Max. thimble
temp. = 1493 k
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Abstract ID: 1_40
Characterization of Discharge Plasma in Cylindrical IECF Device
Neelanjan Buzarbaruah1, Nilam Jyoti Dutta
1, Davashree Borgohain
1, Smruti Ranjan Mohanty
1
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
density (ni) are evaluated [3]. Plasma density of the order 1015
m-3
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 Srivastava
1, Sunil Susmithan
1, Ziauddin Khan
1
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/
Page 93
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Abstract ID: 1_58
Development of Data Acquisition Set-up for Steady-state Experiments
Amit Srivastava1, Arnab Das Gupta
1, Sunil Susmithan
1, Ziauddin Khan
1
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.
Page 94
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Abstract ID: 1_59
Prototyping of Radial Plates for Fusion Relevant Superconducting Magnets
Mahesh Ghate1, Dhaval Bhavasar
1, Arun Panchal
1, Swaroop Udgata
2, Subrata Pradhan
1
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.
Page 95
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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 Ghate
1, Subrata Pradhan
1
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.
Page 96
10th Asia Plasma & Fusion Association Conference
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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 Bano
1, Mahesh Ghate
1, Piyush Raj
1, Subrata Pradhan
1
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.
Page 97
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Abstract ID: 1_64
RF Assisted Glow Discharge Condition Experiment in SST-1 Tokamak Dilip Raval
1, Ziauddin Khan
1, Siju George
1, Kalpeshkumar R Dhanani
1, Yuvakiran Paravastu
1,
Pratibha Semwal1, Prashant Thankey
1, Mohammad Shoaib Khan
1, Subrata Pradhan
1
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.
Page 98
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Abstract ID: 1_67
Commissioning and Experimental Validation of SST-1 Plasma Facing Components
Yuvakiran Paravastu1, Dilip Raval
1, Ziauddin Khan
1, Hiteshkumar Patel
1, Prabal Biswas
1, Tejas
Parekh1, Siju George
1, Prosenjit Santra
1, Gattu Ramesh Babu
1, Prashant Thankey
1, Pratibha
Semwal1, Arun Prakash A
1, Kalpeshkumar R Dhanani
1, Snehal Jaiswal
1, Pradeep Chauhan
1,
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/m
2.
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.
Page 99
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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 Khan
1, Dilip Raval
1, Kalpeshkumar R Dhanani
1, Siju George
1,
Yuvakiran Paravastu1, Arun Prakash A
1, Prashant Thankey
1, Gattu Ramesh Babu
1, Mohammad
Shoaib Khan1, Partha Saikia
1, Subrata Pradhan
1
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.
Page 100
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Abstract ID: 1_79
Design and Integration of SMBI System for SST 1
Siju George1, Yuvakiran Paravastu
1, Mohammad Shoaib Khan
1, Kalpeshkumar R Dhanani
1,
Dilip Raval1, Ziauddin Khan
1, Santanu Banerjee
1, Subrata Pradhan
1
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.
Page 101
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Abstract ID: 1_90
Neutron Measurements from Beam-Target Interactions with Deuterium Ion Beam
Sudhirsinh Vala1, A T T Mostako
1, Mitul Abhangi
1, C V S Rao
1, Rajnikant Makwana
2, 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.
Page 102
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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 Singh
1
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.
Page 103
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Abstract ID: 1_99
Thermal Response of Actively Cooled Tungsten Monoblock during Inhomogeneous Surface Heat Loads
Yashashri Patil1, Samir S Khirwadkar
1, Deepu S Krishnan
1
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/m
2 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/m
2 rest of the tungsten monoblock. Other case heat flux on
shadow is 0 MW/m2 and rest part of tungsten monoblock is 20 MW/m
2. 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.
Page 104
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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.
Page 105
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Abstract ID: 1_101
Consistency Checks in Beam Emission Modeling for Neutral Beam Injectors
Bharathi Punyapu1, Prahlad Vattipalle
1, Sanjeev Kumar Sharma
1, Ujjwal Kumar Baruah
1,
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.
Page 106
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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 Ghate
2, Bharathi Punyapu
1, Rajesh Patel
2, Prahlad Vattipalle
1
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 Sircar
1, Prakash M Raole
1, Lalit M Manocha
2, Raseel Rahman
1
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 Rajpal
1, Amitkumar D Patel
1, Meenakshee Sharma
1, 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.
Page 109
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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 Chouhan
1, Amulya Kumar Sanyasi
2, Kumarpalsinh A Jadeja
2,
Joydeep Ghosh2, J. Sharma
3
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).
Page 110
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Abstract ID: 1_113
Multi-scale Modeling of Neutron Induced Radiation Damage in Tungsten
Maya P N1, Shishir P Deshpande
1, Manoj Warrier
2, Prithwish Nandi
1, Prakash M Raole
3, 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 Bora
1, Ratneshwar Jha
1, Subrata Pradhan
1
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 Rao
1, Jyoti Pandey
2, Mayank Rajput
1, G Vaitheeswaran
1, T K Basu
1, 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
produced inside the fusion reactor (D-T fuel cycle), such as 53
Mn (T½=3.74e+6 year),55
Fe (T½=
2.73 year), 60
Fe (T½=1.5e+6 year), 60
Co (T½ = 5.27 year), 59
Ni (T½ =7.6e+4 year), 63
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 55
Fe(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 10
Be,12
C,53
Mn,55
Fe,59
Ni and 202g
Pb 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 55
Fe 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 Sarkar
2, Anurag Shyam
1, Rajesh Kumar
1, Mridula Mittal
1, Sumit Kumar
1
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 Vattipalle
1, Bhargav Choksi
1, Bharathi Punyapu
1, Rambabu
Sidibomma1, Sridhar B
1, Ujjwal Kumar Baruah
1
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 Padasalagi
1
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/m
2.
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 Shah
1, Jignesh Soni
2, Ratnakar Kumar Yadav
1, Kartik J Patel
2, Hiren
Mistri2, Deepak Parmar
1, Jignesh Bhagora
1, Dheeraj Kumar Sharma
2, Mainak Bandyopadhyay
1,
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 Vaghela
1, Ritendra Bhattacharya
1, Pratik Patel
1, Vinit Shukla
1, Nitin
Shah1, Biswanath Sarkar
1
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 Sharma
1, Yohan Khristi
1, Subrata Pradhan
1, Kalpesh Doshi
1, Upendra Prasad
1, Moni
Banaudha1, Pankaj Varmora
1, Bhadresh R Praghi
1
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 Khan
1, Dilip Raval
1, Pratibha Semwal
1, Siju George
1,
Yuvakiran Paravastu1, Prashant Thankey
1, Mohammad Shoaib Khan
1, Subrata Pradhan
1
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 Sahoo
1, Biswanath Sarkar
1, Anurag Shyam
1
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).
Page 123
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Abstract ID: 1_170
Engineering Design & Integration of Radial Control Coil in Vacuum Vessel of SST-1
Pradeep Chauhan1, Prosenjit Santra
1, Snehal Jaiswal
1, Prabal Biswas
1, Kirit R Vasava
1, Tejas
Parekh1, Hiteshkumar Patel
1, Subrata Pradhan
1
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.
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Abstract ID: 1_171
Engineering Design & Integration of In-vessel Single Turn Segmental Coil in Vacuum Vessel of SST-1
Snehal Jaiswal1, Pradeep Chauhan
1, Prosenjit Santra
1, Kirit R Vasava
1, Tejas Parekh
1,
Hiteshkumar Patel1, Prabal Biswas
1, Subrata Pradhan
1
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.
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Abstract ID: 1_174
Quality Control of FWC during Assembly/Commissioning on SST-1
Hiteshkumar Patel1, Prosenjit Santra
1, Snehal Jaiswal
1, Pradeep Chauhan
1, Yuvakiran Paravastu
1,
Siju George1, Gattu Ramesh Babu
1, Arun Prakash A
1, Pratibha Semwal
1, Prasant Thankey
1,
Kalpeshkumar R Dhanani1, Dilip Raval
1, Ziauddin Khan
1, Subrata Pradhan
1, Tejas Parekh
1,
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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Abstract ID: 1_175
Laser Shock Peening of Stainless Steel Surfaces: ns vis-ã-vis ps Laser Pulses
Prem Kiran P1, Pardhu Yella
1, Koteswararao V Rajulapati
1, Venkateshwarlu Pinnoju
1, Ramesh
Kumar Buddu2, Bhanu Sankara Rao Kota
3
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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Abstract ID: 1_176
Assembly & Metrology of First Wall Components of SST-1
Tejas Parekh1, Prosenjit Santra
1, Prabal Biswas
1, Hiteshkumar Patel
1, Yuvakiran Paravastu
1,
Snehal Jaiswal1, Pradeep Chauhan
1, Gattu Ramesh Babu
1, Arun Prakash A
1, Dhaval Bhavsar
1,
Dilip Raval1, Ziauddin Khan
1, Subrata Pradhan
1
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.
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Abstract ID: 1_189
Trap Site Formation and their Distribution Studies in Porous Lithium Titanate
Chandan Danani1, Manoj Warrier
2, Paritosh Chaudhuri
1
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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Abstract ID: 1_191
Design of a High Power Water Load for LHCD System of SST-1 Tokamak
Harish V Dixit1, Aviraj Jadhav
1, Yogesh M Jain
2, Alice Cheeran
1, Vikas Gupta
3, 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%.
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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 Jadhav
1, Yogesh M Jain
2, Alice Cheeran
1, Vikas Gupta
3, 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.
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Abstract ID: 1_193
Conceptual Design of PAM Antenna for Aditya-U Tokamak
Yogesh M Jain1, Pramod K Sharma
1, Jagabandhu Kumar
1, Harish V Dixit
2, Kirankumar K
Ambulkar1, Pramod R Parmar
1, Chetan G Virani
1
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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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 Shaikh
1, Prakash M Raole
1, Biswanath Sarkar
2
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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Abstract ID: 1_197
Study of Transients in Liquid Helium Flow during Cool Down of Cryopanel
Reena Sayani1, Samiran Shanti Mukherjee
1, Ranjana Gangradey
1
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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Abstract ID: 1_202
A Simple In-vessel/FW Component Viewing System for SST-1
Prosenjit Santra1, Prabal Biswas
1, Kirit R Vasava
1, Snehal Jaiswal
1, Tejas Parekh
1, Pradeep
Chauhan1, Hiteshkumar Patel
1, Subrata Pradhan
1
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.
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Abstract ID: 1_208
Overall Behaviour of PFC Integrated SST-1 Vacuum System
Ziauddin Khan1, Dilip Raval
1, Yuvakiran Paravastu
1, Pratibha Semwal
1, Kalpeshkumar R
Dhanani1, Siju George
1, Mohammad Shoaib Khan
1, Arun Prakash A
1, Gattu Ramesh
1, Prashant
Thankey1, Firozkhan Pathan
1, Subrata Pradhan
1
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 m
3. 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.
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Abstract ID: 1_209
Assembly of Neutral Beam Injector with SST-1
Rambabu Sidibomma1, Prahlad Vattipalle
1, Sanjeev Kumar Sharma
1, Sridhar B
1, Laxmi Narayan
Gupta1, Ujjwal Kumar Baruah
1
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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Abstract ID: 1_216
Experience of 12 kA / 16 V SMPS during the HTS Current Leads Test
Pradip Panchal1, Dikens Christian
1, Rohitkumar Panchal
1, Dashrath Sonara
1, Gaurav Purwar
1,
Atul Garg1, Hiren Nimavat
1, Gaurav Kumar Singh
1, Jal Patel
1, Vipul L Tanna
1, 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 Khristi
1, Subrata Pradhan
1, Pankaj Varmora
1, Upendra Prasad
1
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
Page 139
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Abstract ID: 1_219
Electronics for Coupled High Voltage Measurement on PF Magnets of SST-1
Moni Banaudha1, Yohan Khristi
1, Subrata Pradhan
1, Azadsinh Makwana
1, Upendra Prasad
1,
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.
Page 140
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Abstract ID: 1_224
Electronics and Instrumentation for the SST-1 Superconducting Magnet System
Yohan Khristi1, Subrata Pradhan
1, Pankaj Varmora
1, Moni Banaudha
1, Bhadresh R Praghi
1,
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.
Page 141
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Invited Talk (Session-5)
Page 142
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Abstract ID: 4_200
ITER and its Diagnostics- the Way Ahead
Michael Walsh1, ITER Team
1
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.
Page 143
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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.
Page 144
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Abstract ID: 4_283
R & D of Tritium Technology for Fusion in CAEP: Progress and Prospect
Song Jiangfeng 1, Meng Daqiao
1, Li Rong
1, Huang Zhiyong
1, Huang Guoqiang
1, Chen Chang-
an1, Deng Xiaojun
1, Qin Cheng
1, Qian Xiaojing
1, Zhang Guikai
1
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.
Page 145
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Poster Session-3
Page 146
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Abstract ID: 1_225
Precision Electronics and Measurement Techniques for the Superconducting Joint Resistance
Yohan Khristi1, Subrata Pradhan
1, Kalpesh Doshi
1
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.
Page 147
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Abstract ID: 1_226
Preliminary Results from Electron Cyclotron Measurements at SST-1
Varsha Siju1, Praveena Kumari
1, Surya Kumar Pathak
1
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.
Page 148
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Abstract ID: 1_232
PLATo (Power Load Analysis Tool) – A Module of WEST Wall Monitoring System
Sutapa Ranjan1, Jean-marcel Travere
2, P Moreau
2, C Balorin
2, J Bucalossi
2, V Chaudhari
1, Y
Corre2, M Firdaouss
2, M Jouve
2, E Nardon
2, R Nouailletas
2, N Ravenel
2, B Santraine
2
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
Page 149
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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 Yadav
1, Dhananjay Kumar Singh
1, Hiteshkumar K Patel
1, S Ulahannan
2,
A Vinaykumar2, M Girish
3, M Khan
3, Mahohar
3, Chandramouli Rotti
1, Mainak Bandyopadhyay
1,
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.
Page 150
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Abstract ID: 1_236
Measurement and Sweep-biasing Circuit for Langmuir Probe Diagnostic in SYMPLE
Pramila Gautam1, Jignesh Patel
1, Rachana Rajpal
1, Chandresh Hanasalia
1, Anitha V P
1,
Krishnamachari Sathyanarayana1, Ratneshwar Jha
1
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).
Page 151
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Abstract ID: 1_242
Density Measurement Systems at SST Tokamak
Umeshkumar C Nagora1, Surya Kumar Pathak
1, Parveen Kumar Atrey
1
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.
Page 152
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Abstract ID: 1_245
Software Upgradation of PXI Based Data Acquisition for Aditya Experiments
Vipul K Panchal1, Chhaya Chavda
1, Vijay Patel
1, Narendra Patel
1, Joydeep Ghosh
1
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.
Page 153
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Abstract ID: 1_253
Development, Integration and Testing of Automated Triggering Circuit for Hybrid DC Circuit Breaker
Deven Kanabar1, Swati Roy
1, Chiragkumar Dodiya
1, Subrata Pradhan
1
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 Rathod
2, Snehal Jayaswal
2, Pradeep Chauhan
2, Joydeep Ghosh
2, Rakesh L
Tanna2, Prabal K Chattopadhyay
2, Mohan Parmar
3, Jinto Thomas
2, Madan B Kalal
2,
Krishnamachari Sathyanarayana2, Mohsin Malek
3, Pratik Patel
3, Ramkrushna Panchal
2, 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 Das
1, Anees Bano
1, Subrata Pradhan
1
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 Pathak
1
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.
Page 157
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Abstract ID: 1_265
Assembly of Aditya Upgrade Tokamak
Madan B Kalal1, Rakesh L Tanna
1, Joydeep Ghosh
1, Shailesh B Bhatt
1, Dinesh S Varia
1, Sharvil
Patel1, Vaibha Ranjan
1, Devraj H Sadharkiya
1, Ramkrushna Panchal
1, Rohit Kumar
1, Harshita
Raj1, Krishnamachari Sathyanarayana
1, Kulav Rathod
1, Kumarpalsinh A Jadeja
1, Kaushal M
Patel1, Kaushik S Acharya
1, Prabal K Chattopadhyay
1, Ashok V Apte
2, Yogesh C Saxena
1,
Dhiraj Bora1, Shell-N-Tube Team
3
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.
Page 158
10th Asia Plasma & Fusion Association Conference
155 | P a g e
Abstract ID: 1_266
The Refurbishment of Damaged Toroidal Magnetic Field coils for Aditya Upgrade
Devraj H Sadharakiya1, Rakesh L Tanna
1, Joydeep Ghosh
1, Prabal K Chattopadhyay
1, Sharvil
Patel1, Vaibhav Ranjan
1, Rohit Kumar
1, Harshita Raj
1, Krishnamachari Sathayanarayana
1,
Madan B Kalal1, Dinesh S Varia
1, Ramkrushna Panchal
1, Kulav Rathod
1, Shailesh B Bhatt
1, A
Vardharajulu1, Yogesh C Saxena
1, Dhiraj Bora
1, Shell-N-Tube Team
2
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.
Page 159
10th Asia Plasma & Fusion Association Conference
156 | P a g e
Abstract ID: 1_270
Conceptual Design of Dump Resistor for Superconducting CS of SST-1
Swati Roy1, Subrata Pradhan
1, Arun Panchal
1
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).
Page 160
10th Asia Plasma & Fusion Association Conference
157 | P a g e
Abstract ID: 1_272
Safety and Environment Aspects of Tokamak-type Fusion Power Reactor - An Overview
Bharatkumar Doshi1, D Chenna Reddy
1
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.
Page 161
10th Asia Plasma & Fusion Association Conference
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Abstract ID: 1_274
Fusion Blanket Materials Development and Recent R&D Activities
Chandra Sekhar Sasmal1, Shiju Sam
1, Atikkumar N Mistry
1, Atul Prajapati
1, Hardikkumar M
Tailor1, Jignesh P Chauhan
1, Kinkar Laha
2, Arun Kumar Bhaduri
2, Rajendra Kumar E
1
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.
Page 162
10th Asia Plasma & Fusion Association Conference
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Abstract ID: 1_278
Electrical Properties of Nano Li2TiO3 for Fusion Reactors
S K S Parashar1, Kajal Parashar
1, Paritosh Chaudhuri
2
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.
The 10h milled 3 powder of Li2TiO3 by High energy ball milling (HEBM) calcined at 700 0C
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 5
0C 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 bR and grain boundary gbR 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.
Page 163
10th Asia Plasma & Fusion Association Conference
160 | P a g e
Abstract ID: 1_279
Design of New Superconducting Central Solenoid of SST-1 Tokamak
Upendra Prasad1, Subrata Pradhan
1, Mahesh Ghate
1, Piyush Raj
1, Vipul L Tanna
1, Ziauddin
Khan1, Swati Roy
1, Prosenjit Santra
1, Prabal Biswas
1, Aashoo N Sharma
1, Yohan Khirsti
1,
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.
Page 164
10th Asia Plasma & Fusion Association Conference
161 | P a g e
Abstract ID: 1_281
Design of High Resolution Spectroscopic Diagnostics for SST-1 and ADITYA-U Tokamak
Gaurav Shukla1, Kajal Shah
1, Malay Bikas Chowdhuri
1, Ranjana Manchanda
1, Santanu
Banerjee1, Nilam Ramaiya
1, Joydeep Ghosh
1
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.
Page 165
10th Asia Plasma & Fusion Association Conference
162 | P a g e
Abstract ID: 1_295
Conceptual & Engineering Design of Plug-in Cryostat Cylinder for Superconducting Central Solenoid of SST-1
Prabal Biswas1, Prosenjit Santra
1, Kirit R Vasava
1, Snehal Jayswal
1, Tejas Parekh
1, Pradeep
Chauhan1, Hiteshkumar Patel
1, Subrata Pradhan
1
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.
Page 166
10th Asia Plasma & Fusion Association Conference
163 | P a g e
Abstract ID: 2_4
Investigation of Homoclinic Bifurcation of Plasma Fireball in a Double Plasma Device
Arun Sarma1, Vramori Mitrra
1, Bornali Sarma
1
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.
Page 167
10th Asia Plasma & Fusion Association Conference
164 | P a g e
Abstract ID: 2_14
Determination of the Plasma Composition using Blended Stark-Broadened Emission Lines in a Self-Magnetic Pinch Diode
Subir Biswas1, R Doron
1, V Bernshtam
1, Y Maron
1, M D Johnston
2, M L Kiefer
2, M E Cuneo
2
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).
Page 168
10th Asia Plasma & Fusion Association Conference
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Abstract ID: 2_16
Magnetic Probe Diagnostic Tool to Understand the Dynamics in a Non-transferred dc Plasma Torch
Vidhi Goyal1, Ravi Ganesh
1
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)
Page 169
10th Asia Plasma & Fusion Association Conference
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Abstract ID: 2_22
Localized solutions in Laser Plasma Coupled System with Periodic Time Dependence
Deepa Verma1, Amita Das
1, Bhavesh Patel
1, Predhiman Krishan Kaw
1
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).
Page 170
10th Asia Plasma & Fusion Association Conference
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Abstract ID: 2_23
Coupling of Drift Wave with Dust Acoustic Wave
Atul Kumar1, Amita Das
1, Predhiman Krishan Kaw
1
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).
Page 171
10th Asia Plasma & Fusion Association Conference
168 | P a g e
Abstract ID: 2_30
Resolving Issues Associated with Langmuir Probe Measurements in High Pressure Complex (Dusty) Plasmas
Manjit Kaur1, Sayak Bose
1, Prabal K Chattopadhyay
1, Joydeep Ghosh
1, Yogesh C Saxena
1
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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Abstract ID: 2_32
On the Spatial Behavior of Background Plasma in Different Background Pressure in CPS Device
Subrata Samantaray1, Rita Paikaray
2, Gourishankar Sahoo
2, Parthasarathi Das
2, Joydeep Ghosh
3,
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 Das
1, Rita Paikaray
1, Gourishankar Sahoo
1, 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 Sengupta
1
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 Kant
1
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 9.0b
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 Babu
1
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 kumar
1
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 Mukherjee
1, Sudip Sengupta
1, Amita Das
1
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 Inagaki
1, T Kobayashi
2, 3, M Goto
2, 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<f< 3kHz, m=1component and m=3 component) are shown.
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Abstract ID: 2_48
Production of Quiescent Collisionless Plasma over a Wide Operating Range
Sayak Bose1, Manjit Kaur
1, Prabal K Chattopadhyay
1, Joydeep Ghosh
1, Yogesh C Saxena
1
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
( %1nn ) is produced over a wide operating range by operating the system in low mirror ratio
sourcemainm BBR configuration. Demonstrating the effectiveness of this method magnetized
argon plasma with low density fluctuation %1nn have been produced in the pressure range
~ 35 10 to105 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)
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Abstract ID: 2_50
Effect of Fast Drifting Electrons on Electron Temperature Measurement with a Triple Langmuir Probe
Satyajit Chowdhury1, Subir Biswas
2, Rabindranath Pal
1
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.
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Abstract ID: 2_51
Ponderomotive Force and Backward Raman Scattering in Dense Quantum Plasmas
Punit kumar1, Nisha Singh Rathore
1
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).
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Abstract ID: 2_52
Anode Glow and Double Layer in DC Magnetron Anode Plasma
Samir Chauhan1, Mukesh Ranjan
1, Subroto Mukherjee
1
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)
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Abstract ID: 2_53
Effect of Trapped Particle Nonlinearity in IAW Solitary Wave
Debraj Mandal1, Devendra Sharma
1
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)
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Abstract ID: 2_54
Installation of a 100 kJ Pulsed Power System to Drive Pulsed Plasma Devices
Suramoni Borthakur1, Nayan Talukdar
1, Nirod Neog
1, Tridip Borthakur
1, Rajesh Kumar
2, Rishi
Verma3, Anurag Shyam Shyam
3
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.
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Abstract ID: 2_63
Characterization of the Permanent Magnet Based Hydrogen Helicon Plasma Source for Ion Source Application
Arun Pandey1, Dass Sudhir Kumar
2, Arun Kumar Chakraborty
2
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).
Page 189
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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 Shaw
1, Sabuj Ghosh
1, M S Janaki
1, A N S Iyengar
1
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.
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Abstract ID: 2_66
Radio Frequency Emissions from Plasmas due to Laser Induced Breakdown of Materials
P Prem Kiran1, L Vinoth Kumar
1, Elle Manikanta
1
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).
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Abstract ID: 2_69
Effect of Transverse Magnetic Field on the Steady State Solutions of a Bursian Diode
Sourav Pramanik1, Nikhil Chakrabarti
1, Victor Kuznetsov
2
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.
Page 192
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Abstract ID: 2_70
Wave-breaking Amplitudes of Relativistically Strong Electrostatic Waves in Cold Electron-Positron-Ion Plasmas
Mithun Karmakar1, Chandan Maity
2, Nikhil Chakrabarti
1
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.
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Abstract ID: 2_71
Nonlinear Coherent Structures of Alfven Wave in a Collisional Plasma
Sayanee Jana1, Nikhil Chakrabarti
1, Samiran Ghosh
2
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).
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Abstract ID: 2_72
Parallel Connection Length and Flow-fluctuation Cycle in Simple Toroidal Device
Umesh Kumar1, Shekar Goud Thatipamula
1, Rajaraman Ganesh
1, Yogesh C Saxena
1, 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).
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Abstract ID: 2_80
Controllable Location of Polarization Reversal in Nonuniform Helicon Plasma
Sonu Yadav1, Prabal K Chattopadhyay
1, Joydeep Ghosh
1, Soumen Ghosh
1
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).
Page 196
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Abstract ID: 2_82
Hot Tungsten Plate Based Ionizer for Cesium Plasma in a Multi-Cusp Field Experiment
Amitkumar D Patel1, Meenakshee Sharma
1, Narayanan Ramasubramanian
1, 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).
Page 197
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Abstract ID: 2_107
Development of Three Dimensional Magnetic Field Probe with Signal Conditioning Electronics
Kiran Patel1, Narayan Behera
1, Rajesh Kumar Singh
1, Ajai Kumar
1
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).
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Abstract ID: 2_121
State of Art Data Acquisition System for Large Volume Plasma Device
Ritesh Sugandhi1, Pankaj Srivastava
1, Amulya Kumar Sanyasi
1, Prabhakar Srivastav
1, Lalit
Mohan Awasthi1, Shiban Krishna Mattoo
1, Vijay Parmar
2, Keyur Makadia
2, Ishan Patel
2,
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
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Popular Talk
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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.
Page 201
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Oral Session-1
Page 202
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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 Ejiri
1, Tsujii Naoto
1, Takahiro Shinya
1, Hirokazu Furui
1, Hiroto Homma
1,
Kenta Nakamura1, Masateru Sonehara
1, Wataru Takahashi
1, Toshihiro Takeuchi
1, Hiro Togashi
1,
Kazuya Toida1, Satoru Yajima
1, Hibiki Yamazaki
1, Yusuke Yoshida
1
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).
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Abstract ID: 0_152
ELM Control using Low-n RMPs in KSTAR and its Perspective to Beyond-ITER
Youngmu M Jeon1, Jong-kyu Park
2, Yongkyoon In
1, Jayhyun Kim
1, Siwoo W Yoon
1, Gunyoung
Y Park1, Yeong-kook Oh
1, Hyeon Park
1,3, KSTAR Team
1
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)
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Abstract ID: 0_179
Development of Long Pulse Radiofrequency Heating and Current Drive Systems and Scenarios for WEST
Annika Ekedahl1, Clarisse Bourdelle
1, Jean-Francois Artaud
1, Jean-Michel Bernard
1, Laurent
Colas1, Joan Decker
1, Léna Delpech
1, Rémi Dumont
1, Marc Goniche
1, Walid Helou
1, Julien
Hillairet1, Gilles Lombard
1, Roland Magne
1, Patrick Mollard
1, Eric Nardon
1, Yves Peysson
1,
Emmanuelle Tsitrone1, Zhaoxi Chen
2, Bojiang Ding
2, Xianzu Gong
2, Miaohui Li
2, Yuntao Song
2,
Yongsheng Wang2, Qingxi Yang
2, Yanping Zhao
2 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
absorption (n//0=2.0) takes place in the region r/a=0.3-0.6 at pedestal a density of nped=3×1019
m-3
.
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.
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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).
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Abstract ID: 0_184
Behaviors of Impurity in ITER and DEMOs using BALDUR Integrated Predictive Modeling Code
Thawatchai Onjun1, Wannapa Buangam
1, Apiwat Wisitsorasak
2
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.
Page 207
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Abstract ID: 4_284
Rapid Purification of Hydrogen Isotope Gas by Palladium Alloy Membrane Separator
Xiong Yifu1, Song Jiangfeng
1,Jing Wenyong1,He Mingmin
1,Ba Jingwen1, Shiyan
1
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.
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Abstract ID: 0_214
Measurements and Controls Implementation for the WEST Project
Raju Daniel1, P Moreau
2, Manisha Bhandarkar
1, S Brémond
2, J Bucalossi
2, Vishnu K Chaudhari
1,
X Courtois2, Jasraj Dhongde
1, C Gil
2, Aveg Kumar
1, Praveena Kumari
1, M Lewerentz
4, P Lotte
2,
Imran Mansuri1, Harish Masand
1, O Meyer
2, M Missirlian
2, E Nardon
2, R Nouailletas
2,
Kiritkumar B Patel1, Sutapa Ranjan
1, C Rapson
3, G Raupp
3, N Ravenel
2, F Samaille
2, Manika
Sharma1, J Signoret
2, A Spring
4, J M Travere
2, W Treuterrer
3, A Werner
4, WEST team
2
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/m
2 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
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Poster Session-4
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Abstract ID: 2_122
Super Rogue Wave in Plasma
Pallabi Pathak1, Sumita Kumari Sharma
1, Heremba Bailung
1
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).
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Abstract ID: 2_123
Experiment on Dust Acoustic Solitons in Strongly Coupled Dusty Plasma
Abhijit Boruah1, Sumita Kumari Sharma
1, Heremba Bailung
1
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).
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Abstract ID: 2_125
Controllable Transition from Positive Space Charge to Negative Space Charge in an Inverted Cylindrical Magnetron
Ramkrishna Rane1, Mainak Bandyopadhyay
2, Mukesh Ranjan
1, Subroto Mukherjee
1
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.
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Abstract ID: 2_126
Measurement of Electron Energy Probability Function in Weakly Magnetized Plasma
Deiji Kalita1, Bharat Kakati
2, Bipul Kumar Saikia
1, Mainak Bandyopadhyay
3, 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)
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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 Mukherjee
1, Rajaraman Ganesh
1, Abhijeet Sen
1
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).
Page 215
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Abstract ID: 2_140
Investigation of Magnetic Drift on Transport of Plasma across Magnetic Field
Parismita Hazarika1, Bidyut Das
2, Monojit Chakraborty
1, Mainak Bandyopadhyay
2
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.
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Abstract ID: 2_141
High Intensity High Contrast Femtosecond Laser Absorption in Solid
Kamalesh Jana1, Amitava Adak
1, Moniruzzaman Shaikh
1, Deep Sarkar
1, Indranuj Dey
1, Amit D
Lad1, G Ravindra Kumar
1
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.
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Abstract ID: 2_153
Lithium Vapor Density Diagnostics for the PWFA Plasma Source at IPR
Mohandas Kizhupadathu Krishnan1, Sivakumaran Valluvadasan
1, Sneha Singh
1, 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.
Page 218
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Abstract ID: 2_157
Turbulent, Megagauss Magnetic Fields in Intense, Ultrashort Laser Pulse Interaction with Solids
Amit D Lad1, Gourab Chatterjee
1, Kevin Schoffler
2, Prashant Singh
1, Sudip Sengupta
3,
Predhiman Krishan Kaw3, Luis Silva
2, Amita Das
3, G Ravindra Kumar
1
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 – 10
3 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).
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Abstract ID: 2_158
Design and Characterization of Cesium Oven for a Multi-cusp Plasma Device
Meenakshee Sharma1, Amitkumar D Patel
1, Narayan Ramasubramanian
1
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.
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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 G
1, Sijo Sebastian
1, Neethu Jayakumar
1, Anu Varghese
1, Neethu
Theresa Willington2, Venugopal Chandu
1
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.
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Abstract ID: 2_165
Two Dimensional Imaging of Laser Produced Plasma in Magnetic field
Narayan Behera1, Rajesh Kumar Singh
1, Ajai Kumar
1
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.
Page 222
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Abstract ID: 2_172
The Effect of Addition of Lighter Ions in a Five Component Multi-Ion Plasma
Sijo Sebastian1, Manesh Michael
1, Sreekala G
1, Neethu Theresa Willington
2, Anu Varghese
1,
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.
Page 223
10th Asia Plasma & Fusion Association Conference
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Abstract ID: 2_182
Effect of Ablation Geometry on the Formation of Stagnation Layer in Laterally Colliding Plasmas
Alamgir Mondal1, Rajesh K Singh
1, Ajai Kumar
1
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.
Page 224
10th Asia Plasma & Fusion Association Conference
221 | P a g e
Abstract ID: 2_183
Enhanced Confinement by Controlling Instability in Toroidal Electron Plasma of SMARTEX-C
Lavkesh Lachhvani1, Sambaran Pahari
2, Manu Bajpai
1, Prabal K Chattopadhyay
1, 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.
Page 225
10th Asia Plasma & Fusion Association Conference
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Abstract ID: 2_185
Study of Phase Space Structures in Driven 1D Vlasov Poisson Model
Pallavi Trivedi1, Rajaraman Ganesh
1
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)
Page 226
10th Asia Plasma & Fusion Association Conference
223 | P a g e
Abstract ID: 2_186
Synchronization dynamics and Arnold tongues for two coupled glow discharge plasma sources
Neeraj Chaubey1, Subroto Mukherjee
1, A N Sekar Iyengar
2, Abhijeet Sen
1
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)
Page 227
10th Asia Plasma & Fusion Association Conference
224 | P a g e
Abstract ID: 2_194
Optical Kerr Gated Time Resolved Cherenkov Emission Produced during Ultra Intense Laser Solid Interaction
Moniruzzaman Shaikh1, Amit Lad
1, Deep Sarkar
1, Sheroy Tata, Indranuj Dey, Gourab
Chatterjee1, Rajeev P P
2 , G Ravindra Kumar
1
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.
Page 228
10th Asia Plasma & Fusion Association Conference
225 | P a g e
Abstract ID: 2_199
Imaging of Terahertz Emission from Intense High-Contrast Ultrashort-Pulse Laser-Solid Interaction
Indranuj Dey1, Deep Sarkar
1, Moniruzzaman Shaikh
1, Sheroy Tata
1, Amit D Lad
1, G Ravindra
Kumar1
1Tata Institute of Fundamental Research, India
Email: [email protected]
When an intense (Io ~ 1018
- 1019
W/cm2, ~ 800 nm), high-contrast (pedestal/peak ~ 10
-9),
ultrashort pulse (td ~ 30 fs, rep. rate 10 Hz) is obliquely focused on to a solid surface, a very
high density plasma (ne ~ 1022
m-3
) 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).
Page 229
10th Asia Plasma & Fusion Association Conference
226 | P a g e
Abstract ID: 2_218
Pulsed Plasma for the Study of Coherent Structure in the Electron Magnetohydrodyanamic Regime
Garima Joshi1,2
, Ravi Ganesh1, Subroto Mukherjee
1
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
phenomena in this regime, the plasma conditions required are approximately ne ~1011
– 1012
cm-3
,
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)
Page 230
10th Asia Plasma & Fusion Association Conference
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Abstract ID: 2_228
Chaos to Order Transitions in Chaotic Magnetic Fields
A N Sekar Iyengar1, M S Janaki
1, Pankaj Kumar Shaw
1, Subha Samanta
1
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).
Page 231
10th Asia Plasma & Fusion Association Conference
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Abstract ID: 2_229
Investigation of Coherent Modes in the Chaotic Fluctuation in the SINP Tokamak
Pankaj Kumar Shaw1, Debajyoti Saha
1, Sabuj Ghosh
1, Subha Samanta
1, M S Janaki
1, 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).
Page 232
10th Asia Plasma & Fusion Association Conference
229 | P a g e
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 Bose
2, Subroto Mukherjee
1
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.
Page 233
10th Asia Plasma & Fusion Association Conference
230 | P a g e
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 Adak
1, Moniruzzaman Shaikh
1, Indranuj Dey
1, Amit D Lad
1, 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).
Page 234
10th Asia Plasma & Fusion Association Conference
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Abstract ID: 2_244
Laser Heated Emissive Probe for Plasma Potential Measurement in Fusion Plasmas
Vara Prasad Kella1, Payal Mehta
2, Joydeep Ghosh
1, Arun Sarma
2
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).
Page 235
10th Asia Plasma & Fusion Association Conference
232 | P a g e
Abstract ID: 2_246
Study of Fluctuation Induced Particle Flux in the Background of ETG plasma in LVPD
Prabahkar Srivastav1, Lalit Awasthi
1, Amulya Kumar Sanyasi
1, Pankaj Srivastava
1, Ratneshwar
Jha1, Raghvendra Singh
2, Predhiman Krishan Kaw
1
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).
Page 236
10th Asia Plasma & Fusion Association Conference
233 | P a g e
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.
Page 237
10th Asia Plasma & Fusion Association Conference
234 | P a g e
Abstract ID: 2_261
High Energy Neutral Atoms from High Intensity Laser Plasma Interaction
Sheroy Tata1, Malay Dalui
1, T Madhu Trivikram
1, Jagannath Jha
1, M Krishnamurthy
1
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).
Page 238
10th Asia Plasma & Fusion Association Conference
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Abstract ID: 2_262
Role of Magnetic Cusp for Multiple Axial Potential Structures (MAPS) Formation
Soumen Ghosh1, Pabal K Chattopadhyay
1, Joydeep Ghosh
1, Dhiraj Bora
1
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).
Page 239
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Abstract ID: 2_263
Enhanced Proton Acceleration by Ultrashort Laser Pulse Interaction with Nanostructured Thin Films
Angana Mondal1, Malay Dalui
1, Sheroy Tata
1, Subhrangshu Sarkar
1, Jagannath Jha
1, Amit Lad
1,
W -m Wang2, Z m Sheng
3, M Krishnamurthy
1, P Ayyub
1
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.
Page 240
10th Asia Plasma & Fusion Association Conference
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Abstract ID: 2_276
DAQ System for Low Density Plasma Parameters Measurement
Rashmi S Joshi1, Suryakant B Gupta
1
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.
Page 241
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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.
Page 242
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Abstract ID: 3_11
Modeling of Electromagnetic Fields during Plasma Startup in SST-1 Tokamak
Amit Kumar Singh1, Indranil Bandyopadhyay
2, Srinivasan Radhakrishnana
2, SST-1 Team
1
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.
Page 243
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Abstract ID: 3_19
Oscillating Two-stream Instability of a Plasma Wave in Ion-Motion Regime
Pinki Yadav1, Devki Nandan Gupta
1, Avinash Khare
1
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).
Page 244
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Abstract ID: 3_20
Development of a 3D-3V PIC code to study PSI processes in Tokamak Divertor Region
Sayan Adhikari1, Kalyan Sindhu Goswami
1
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
Page 245
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Abstract ID: 3_21
Betatron Radiation from Laser Wakefield Acceleration in a Plasma Channel
Devki Nandan Gupta1, Inhyuk Nam
2, Hyyong Suk
2
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).
Page 246
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Abstract ID: 3_39
Particle in Cell Simulations of Beam Plasma System
Chandrasekhar Shukla1, Atul Kumar
1, Bhavesh Patel
1, Amita Das
1, Kartik Patel
2
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)
Page 247
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Abstract ID: 3_43
PIC Modeling of Negative Ion Extraction from a Dust-Seeded Plasma
Ananya Phukan1, Pranjal Bhuyan
1
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)
Page 248
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Abstract ID: 3_62
Dynamics of dusty fluid in a streaming sheared plasma
Modhuchandra Laishram1, Devendra Sharma
1, Predhiman Krishan Kaw
1
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).
Page 249
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Abstract ID: 3_75
Numerical Analysis on Bandwidth and Growth Rate of Plasma-Filled Gyrotron Devices
Elaheh Allahyari1, Shahrooz Saviz
1
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,0 zB e 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.
Page 250
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Abstract ID: 3_76
Gyro-TWT in a Vane-Loaded Waveguide with Inner Dielectric
Reyhaneh S. Hashemi1, Elaheh Allahyari
1, Shahrooz Saviz
1
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.
Page 251
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Abstract ID: 3_77
Effect of Plasma Column on the Radial Profile of Electric Field of Gyrotron Devices
Elaheh Allahyari1, Shahrooz Saviz
1
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,0 zB e , 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.
Page 252
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Abstract ID: 3_83
Current Gradient Modes of Two Dimensional Electron Magnetohydrodynamics (EMHD)
Gurudatt Gaur1, Predhiman Krishan Kaw
1
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).
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Abstract ID: 3_86
A Poynting like Theorem for Generalized Hydrodynamic Equations
Vikram Singh Dharodi1, Bhavesh Patel
1, Amita Das
1, Predhiman Krishan Kaw
1
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.
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Abstract ID: 3_100
Identification of Nonlinear Resonance Absorption in a Laser Driven Deuterium Cluster using Molecular Dynamics Simulation
Sagar Sekhar Mahalik1, Mrityunjay Kundu
1
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).
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Abstract ID: 3_128
1D PIC simulation of relativistic Buneman instability
Roopendra Singh Rajawat1, Sudip Sengupta
1
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
that breakdown of linear growth (saturation) occurs when |E|2/16πW0 ~ 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.
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Abstract ID: 3_130
Molecular Dynamics Simulation of Dust Particle Levitation in the Presence of Sheath
Sandeep Kumar1, Amita Das
1, Sanat Kumar Tiwari
1, Predhiman Krishan Kaw
1
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).
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Abstract ID: 3_135
Conceptual Study of High-Field LHCD in KSTAR
Young-soon Bae1, S Shiraiwa
2, P Bonoli
2, S J Wang
1, G Wallace
2, J C Wright
2, R Parker
2, Won
Namkung2, M H Cho
3, H. Park
1,4
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.
Page 258
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Abstract ID: 3_138
Integrated Core-SOL Simulations of L-Mode Plasma in ITER and Indian DEMO
Apiwat Wisitsorasak1, Thawatchai Onjun
2, Wittawat Kanjanaput
2
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.
Page 259
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Abstract ID: 3_150
Potential around a dust grain in collisional plasma
Rakesh Moulick1, Kalyan Sindhu Goswami
1
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.
Page 260
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Abstract ID: 3_167
Numerical simulation of a novel non-transferred arc plasma torch operating with nitrogen
Gavisiddayya Hiremath1, Ramachandran Kandasamy
2, Ravi Ganesh
3
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.
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Poster Session-5
Page 262
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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
Page 263
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Abstract ID: 3_188
Sensitivity analysis of upstream plasma condition for SST-1 X-divertor configuration with SOLPS
Himabindu Manthena1, Anil K Tyagi
1, Deepti Sharma
1, Devendra Sharma
1, 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.
Page 264
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Abstract ID: 3_247
Radiation Effects on the Laser Ablative Shockwaves from Aluminum under Atmospheric Conditions
Sai Shiva S1, Leela C H
1, Sijoy C D
2, Shashank Chaturvedi
2, Prem Kiran P
1
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).
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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 Dhabhai
2, Naveen Kumar Jaiman
3
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.
Page 266
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Abstract ID: 3_275
Real-time Horizontal Position Control for Aditya-Upgrade Tokamak
Rohit Kumar1, Joydeep Ghosh
1, Rakesh L Tanna
1, Praveen Lal
1, Prabal K Chattopadhyay
1,
Chhaya Chavda1, Vipul K Panchal
1, Vijay Patel
1, Chet Narayan Gupta
1, Raju Daniel
1, 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.
Page 267
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Abstract ID: 3_294
Prediction of Temperature and Stress Distributions in Substrate and Coating during Plasma Spraying
Raja Mohan1, H Gavisiddayya
1, K Ramachandran
2, P V A Padmanabhan
3, T K Thiyagarajan
3
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.
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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 Kumar
1, Sriprakash Verma
1, Hriday Patel
1, Rajesh Trivedi
1, 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).
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Abstract ID: 4_9
Effect of Geometrical Imperfection on Buckling Failure of ITER VVPSS Tank
Saroj Kumar Jha1, Girish Kumar Gupta
2, Manish Kumar Pandey
1, Avik Bhattacharya
1, Gaurav
Jogi1, Anil Kumar Bhardwaj
1
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).
Page 270
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Abstract ID: 4_17
Nuclear Analyses of Indian LLCB Test Blanket System in ITER
H L Swami1, Akshaya Kumar Shaw
1, Chandan Danani
1, Paritosh Chaudhuri
1
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
Page 271
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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.
Page 272
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References:
[1] T. Wanatabe,“The impact of grain boundary character distribution on fracture in
polycrystals”, Mater. Sci. Eng. A, 176, 39 (1994).
Page 273
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Abstract ID: 4_29
Alternate Design of ITER Cryostat Skirt Support System
Manish Kumar Pandey1, Girish Kumar Gupta
1, Anil Kumar Bhardwaj, Saroj Kumar Jha
1
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).
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Abstract ID: 4_31
Neutronics Analysis, Shielding Optimization and Radiation Waste Analysis for X-Ray Crystal Spectrometer of ITER
P V Subhash1, Gunjan Indauliya
2, Sai Chaitanya Tadepalli
1, Shrichand Jakhar
3, Sanjeev Kumar
Varshney1, Siddharth Kumar
1, K Raja Krishna
4, Nirav Bhaliya
2, Robin Barnsley
3, Bernascolle
Phillipe3, Shrishail B Padasalagi
1, Sapna Mishra
1 and Vinay Kumar
1
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.
Page 275
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Abstract ID: 4_78
Preliminary Optical Design of Polarization Splitter Box for ITER ECE Diagnostic System
Ravinder Kumar1, Suman Danani
1, Hitesh Kumar Pandya
1, Vinay Kumar
1
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.
Page 276
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Abstract ID: 4_87
Development of High Voltage and High Current Test Bed for Transmission Line Components
Akhil Jha1, Manojkumar Patel
1, J V S Harikrishna
1, Ajesh P
1, Rohit Anand
1, Rajesh Trivedi
1,
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.
Page 277
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Abstract ID: 4_94
Development of Control System for Multi-converter High Voltage Power Supply using Programmable SoC
Rasesh Dave1, Jagruti Dharangutti
2, N P Singh
1, Aruna Thakar
1, Hitesh Dhola
1, Sandip Gajjar
1,
Darshan Kumar Parmar1, Tanish Zaveri
2, Ujjwal Kumar Baruah
1
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.
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Abstract ID: 4_95
Development and Validation of I-Activation Analysis Code
Sai Chaitanya Tadepalli1, P V Subhash
1, Gunjan Indauliya
2
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.
Page 279
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Abstract ID: 4_97
Indigenous Manufacturing Realization of Twin Source and its Auxiliary System
Ravi Pandey1, Mainak Bandyopadhyay
2, Deepak Parmar
2, Ratnakar Kumar Yadav
2, Himanshu
Tyagi2, Jignesh Soni
1, Hardik Shishangiya
2, Dass Sudhir Kumar
2, Sejal Shah
2, Gourab Bansal
1,
Kaushal Pandya1, Kanubhai Parmar
1, Mahesh Vuppugalla
1, Agrajit Gahlaut
1, 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.
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Abstract ID: 4_98
Wilkinson Type Lumped Element Combiner-Splitter for Indigenous Amplifier Development
Manojkumar Patel1, Akhil Jha
1, J V S Harikrishna
1, Rajesh Trivedi
1, Aparajita Mukherjee
1
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.
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Abstract ID: 4_102
Preliminary Design Development of ITER X-ray Survey Spectrometer
Sanjeev Kumar Varshney1, Siddharth Kumar
1, Sapna Mishra
1, Subhash Puthenveetil
1, Kuashal
Joshi1, Shivakant Jha
1, Vinay Kumar
1, Robin Barnsley
2, Philippe Bernascolle
2, Gunter
Bertschinger2, Stefan Simrock
2, Jean-Marc Drevon
2, and Michael Walsh
2
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.
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Abstract ID: 4_112
Integration & Validation of LCU with Different Sub-systems for Diacrode Based Amplifier
Rajnish Kumar1, Sriprakash Verma
1, Dipal Soni
1, Hriday Patel
1, Gajendra Suthar
1, Hrushikesh
Dalicha1, Hitesh Dhola
1, Amit Patel
1, Dishang Upadhyay
1, Akhil Jha
1, Manojkumar Patel
1,
Rajesh Trivedi1, Raghuraj Singh
1, Harsha Machchhar
1, Aparajita Mukherjee
1
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.
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Abstract ID: 4_114
Comparative Analysis on Flexibility Requirements of Typical Cryogenic Transfer Lines
Mohit Jadon1, Uday Kumar
1, Ketan Choukekar
1, Nitin Shah
1, Biswanath Sarkar
1
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)
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Abstract ID: 4_116
Dynamics of Cold Helium Flow inside a Cryoline used for Large Cryogenic Distribution System
Uday Kumar1, Mohit Jadon
1, Ketan Choukekar
1, Vinit Shukla
1, Pratik Patel
1, Himanshu Kapoor
1,
Nitin Shah1, Srinivasa Muralidhara
1, Biswanath Sarkar
1
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).
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Abstract ID: 4_137
Final Configuration with Assembly Assessment of the 100kV High Voltage Bushing for the Indian Test Facility
Dheeraj Kumar Sharma1, Sejal Shah
2, M Venkata Nagaraju
2, Mainak Bandyopadhyay
2,
Chandramouli Rotti2, Arun Kumar Chakraborty
2
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
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Abstract ID: 4_143
Preliminary Design of O-mode Radiometer for ITER ECE Diagnostic
Suman Danani1, Hitesh Pandya
1, Ravinder Kumar
1, Max E Austin
2, Victor S Udintsev
3, 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.
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Abstract ID: 4_146
System Upgradation for Surface Mode Negative Ion Beam Extraction Experiments in ROBIN
Kaushal Pandya1, Gourab Bansal
1, Jignesh Soni
1, Agrajit Gahlaut
1, Ratnakar Kumar Yadav
2,
Mahesh Vuppugalla1, Himanshu Tyagi
2, Kanubhai Parmar
1, Hiren Mistri
2, Jignesh Bhagora
1,
Bhavesh K Prajapati1, Kartik J Patel
1, Manar Bhuyan
2, Mainak Bandyopadhyay
1, 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.
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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
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Abstract ID: 4_147
Thermo-mechanical Design Methodology for ITER Cryo-distribution Cold Boxes
Vinit Shukla1, Pratik Patel
1, Hiten Vaghela
1, Jotirmoy Das
1, Nitin Shah
1, Ritendra Bhattacharya
1,
Hyun-sik Chang2, Biswanath Sarkar
1
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.
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Abstract ID: 4_148
Preliminary Design of Bellows for the DNB Beam Source by EJMA & FE Linear Analysis
Shobhit Trapasiya1, Venkata Nagaraju Muvvala
2, Rambilas P
2, Dheeraj Kumar Sharma
3,
Roopesh Gangadharan2, Chandramouli Rotti
2, Arun Kumar Chakraborty
2
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.
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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 Topiwala
3, Kedar Bhope
2, Alpesh Patel
2, Chandramouli
Rotti1, Arun Kumar Chakraborty
1
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.
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Abstract ID: 4_159
Significance of ITER IWS Material Selection and Qualification
Bhoomi K Mehta1, Haresh A Pathak
1, Gurlovleen Singh Phull
1, Rahul Kumar Laad
1, Abha
Maheshwari1, Jigar Raval
1
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.
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Abstract ID: 4_196
ITER ECE Diagnostic: Design Progress of IN-DA and its Role for Physics Study
Hitesh Kumar Pandya1, Ravinder Kumar Jakhmola
1, Suman Danani
1, Shrishail B Padasalagi
1,
Sajal Thomas1, Vinay Kumar
1, G Taylor
2, A Khodak
2, W L Rowan
3, S Houshmandyar
3, V S
Udintsev4, N Casal
4, M Walsh
4
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.
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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.
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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 Yadav
1, Amit Sircar
1
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.
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Abstract ID: 4_254
Seismic Design of ITER Component Cooling Water System-1 Piping
Aditya Prakash Singh1, Mahesh Jadhav
1, Lalit Kumar Sharma
1, Dinesh Kumar Gupta
1, Nirav
Patel1, Rakesh Ranjan
1, Guman Gohil
1, Hirenkumar A Patel
1, Jinendra Dangi
1, Mohit Kumar
1, 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.
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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 Pathak
1, Bhoomi K Mehta
1, Gurlovleen Singh Phull
1, Rahul Laad
1,
Moin Shaikh1, Siju George
2, Kaushal Joshi
2, Ziauddin Khan
2
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.
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Abstract ID: 4_257
Manufacturing and Assembly of IWS Support Rib and Lower Bracket for ITER Vacuum Vessel
Rahul Laad1, Yatin Sarvaiya
1, Haresh A Pathak
1, Raval Jigar
2, Chang-ho Choi
2
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.
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Abstract ID: 4_260
Finite Element Analysis for ITER Ferromagnetic In-wall Shielding Block
Moinuddin Shaikh1, Haresh A Pathak
1, Raval Jigar
2, Tailhardat Oliver
3
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.
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Abstract ID: 4_264
Development of XM-19 Fasteners for the IWS Blocks Assemblies
Sunil Dani1, Gurlovleen Singh
1, Haresh A Pathak
1, Jigar Raval
2, Chang-ho Choi
2
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.
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Abstract ID: 4_299
Present design status of Erosion and Tritium Monitor diagnostics for ITER
Govindarajan Jagannathan1, Nancy Ageorges
2, David Anthoine
3, Sharath Delanthabettu
4, Roger
Reichle1, George Vayakis
1 and Michael Walsh
1
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
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Abstract ID: 5_81
Study of Structures and Stability in Nitrogen Plasma Jet
Nirupama Tiwari1, K C Meher
1, Srikumar Ghorui
1
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.
Page 303
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Abstract ID: 5_84
Pesticides Removal from Cabbage using Array of Atmospheric Pressure Plasma Jet
Akshay Vaid1, Chirayu Patil
1, Ramkrishna Rane
1, Subroto Mukherjee
1, Sudhir Nema
1, Hetal
Bhatt2, R V Prasad
2
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.
Page 304
10th Asia Plasma & Fusion Association Conference
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Abstract ID: 5_85
Comparison of Gas and Plasma Carburizing of AISI 8620 Low Carbon Steel
Alphonsa Joseph1, Ghanshyam Jhala
1, Akshay Vaid
1, Suryakant Gupta
1, Keena Kalaria
1, Naresh
Vaghela1, Subroto Mukherjee
1
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.
Page 305
10th Asia Plasma & Fusion Association Conference
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Abstract ID: 5_92
Experimental Study to Improve Anti-felting Characteristics of Merino Wool Fiber by Atmosphere Pressure Air Plasma
Nisha Chandwani1, Purvi Deva
1, Vishal Jain
1, Sudhir Nema
1, Subroto Mukherjee
2
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)
Page 306
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Abstract ID: 5_103
Surface Chemistry and Wettability Study of Air Plasma Treated Polyethylene by Atmospheric Pressure Dielectric Barrier Discharge
Purvi Deva1, Nisha Chandwani
1, Vishal Jain
1, Sudhir Nema
1, Subroto Mukherjee
2
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).
Page 307
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Abstract ID: 5_227
Electrical Characteristics of a DC Non-transferrerd Arc Plasma Torch Using Theory of Dynamic Similarity
Yugesh V1, Ravi Ganesh
2, K Ramachandran
3
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).
Page 308
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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 Shyam
1,2
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 12
th
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)
Page 309
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Abstract ID: 5_280
Plasma Sterilization for Bio-decontamination
Suryakant Gupta1, Sudhir Nema
1
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.
Page 310
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Oral Session-2
Page 311
10th Asia Plasma & Fusion Association Conference
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Abstract ID: 1_56
Superficial Layer MHD Effect and Full-cover Free Surface Flow Characterization
Zengyu Xu1, Chuanjie Pan
1, Xiujie Zhang
1, Lu Peng
1
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.
Page 312
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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 Jaeger
2, J C Hosea
1, C Lau
3, R J Perkins
1, C K Phillips
1, G Taylor
1
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.
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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/m
2 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.
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Abstract ID: 1_104
Recent Progress of the ECRH System on HL-2A
Shaodong Song1, Mei Huang
1, He Wang
1, Jun Rao
1, Bo Lu
1, Zihua Kang
1, Mingwei Wang
1, Kun
Feng1, Chao Wang
1, Jieqiong Wang
1, Jiruo Ye
1, Feng Zhang
1, Xiaolan Zou
2, Gerardo Giruzzi
2,
Roland Magne2, Zhongbing Shi
1, Qingwei Yang
1, Weimin Xuan
1, Xuru Duan
1
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.
Page 315
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Abstract ID: 1_109
The 3.7GHz LHCD System on HL-2A
Bo Lu1, Roland Magne
2, Xingyu Bai
1, Hao Zeng
1, Yali Chen
1, Chao Wang
1, Emmanuel
Bertrand2, Lena Delpech
2, Annika Ekedahl
2, Julien Hillairet
2, Jun Liang
1, Jieqiong Wang
1,
Zhihua Kang1, Kun Feng
1 and Jun Rao
1
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.
Page 316
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Oral Session-3
Page 317
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Abstract ID: 0_42
Observation of Up-Down Asymmetry in Impurity Line Emissions from the Ergodic Layer of Large Helical Device
Tetsutarou Oishi1, Shigeru Morita
1, Xianli Huang
2, Hongming Zhang
2, Motoshi Goto
1
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).
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Abstract ID: 1_290
Current Status of Safety design and Safety Analysis for China ITER Helium Coolant Ceramic Breeder Test Blanket System
Long Zhang
1, Qixiang Cao
1, Yanling Wang
1, Yanjing Chen
1, Qijie Wang
1, Fengchao Zhao
1, Fen
Wang1, Xinghua Wu
1, Xiaoyu Wang
1, Kaiming Feng
1
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.
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Abstract ID: 4_91
Destructive Analysis on the ITER FW Small Scale Mock-ups
Pinghuai Wang1, Jiming Chen
1, Danhua Liu
1, Fanya Jin
1, Bo Yang
1
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.
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Abstract ID: 1_131
EAST Articulated Maintenance Arm for EAST and WEST
Yong Cheng1, Yuntao T Song
1, Eric Villedieu
2, Vincent Bruno
2
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
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Abstract ID: 0_88
Improvements in a Tracer-Encapsulated Solid Pellet and Its Injector for More Advanced Plasma Diagnostics
Naoki Tamura1, Shigeru Sudo
2, Chihiro Suzuki
1, Hisamichi Funaba
1, Masaru Takagi
3, Nakahiro
Satoh3, Hiromi Hayashi
1, Hiroya Maeno
1, Mitsuhiro Yokota
1, Hideki Ogawa
1
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).
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Oral Session-4
Page 323
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Abstract ID: 3_297
Simulation and Modeling of Magnetic Field Dynamics in Laser Plasma Interaction
Amita Das1, Chandrashekhar Shukla
1, Atul Kumar
1, Bhavesh Patel
1, Predhiman Krishan Kaw
1
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.
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Abstract ID: 2_74
Electrical Transverse Transport in Lorentz Plasma with Strong Magnetic Field and Collision Effect
Baisong Xie1, Chong LV
1, 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).
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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 Kumar
2, Rajesh Kumar Singh
1
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).
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Abstract ID: 0_34
Thermionic Divertors for Tokamaks
Avinash Khare1, Sanjay K Mishra
2, Predhiman Krishan Kaw
2
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.
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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
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Oral Session-4
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Abstract ID: 4_118
Technical Developments and Present Status of the ITER Cryolines and Cryo-distribution Systems
Biswanath Sarkar1, Nitin Shah
1, Hitensinh Vaghela
1, Ketan Choukekar
1, Pratik Patel
1, Himanshu
Kapoor1, Srinivasa Muralidhara
1, Jotirmoy Das
1, Uday Kumar
1, Anuj Garg
1, Vinit Shukla
1,
Mohit Jadon1, Vikas Gaur
1, Bikash Dash
1, Shk Madeenavalli
1
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.
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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.
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Abstract ID: 1_207
Upgradation Plans of SST-1 Cryogenics System at IPR
Vipul L Tanna1, SST-1 Cryo Team
1, Subrata Pradhan
1
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.