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10th Asia Plasma & Fusion Association Conference

APFA 2015


Monday, 14 December 2015 – Friday, 18 December 2015

Hosted by

Institute for Plasma Research

Gandhinagar, India

Book of Abstracts


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


1 | P a g e

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


2 | P a g e

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


3 | P a g e

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


4 | P a g e

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


5 | P a g e

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


6 | P a g e

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


7 | P a g e

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


8 | P a g e

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


9 | P a g e

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.


12 | P a g e

Invited Talk (Session-1)


13 | P a g e

Abstract ID: 1_285

Japanese Strategy of Fusion Roadmap

Akio Komori1, Hiroshi Yamada

1, Satoru Sakakibara

1

1National Institute for Fusion Science, Japan


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)


14 | P a g e

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


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


15 | P a g e



16 | P a g e

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


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


17 | P a g e

Abstract ID: 0_291

Fusion Research in India

Dhiraj Bora1

1Institute for Plasma Research, India


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.


18 | P a g e

Abstract ID: 4_301

Status of ITER Project Activities in KO-DA

Kijung J Jung1



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.


19 | P a g e

Abstract ID: 4_296

ITER India R & D and ITER Package Progress

Shishir Deshpande1

1Institute of Plasma Research, India


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.


20 | P a g e

Poster Session-1


21 | P a g e

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



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.


22 | P a g e

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


2Ecole Polytechnique Federale de Lausanne, France


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)


23 | P a g e

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


2Max-Planck-instite fur Plasmaphysik, Germany


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)


24 | P a g e

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



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).


25 | P a g e

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


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).


26 | P a g e

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


2Institute of Plasma Physics Chinese Academy of Sciences, China



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.


27 | P a g e

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




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.


28 | P a g e

Abstract ID: 0_89

Neutral Particle Profiles during ICRH Experiments in Aditya Tokamak

Nilam Ramaiya1, Ritu Dey



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



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).


29 | P a g e

Abstract ID: 0_96

Understanding of Impurity Behavior in SST-1 Plasmas Using Visible Spectroscopy

Ranjana Manchanda1, Nilam Ramaiya


1, Santanu Banerjee

1, Joydeep

Ghosh1, SST-1 Team

1



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.


30 | P a g e

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



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.


31 | P a g e

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, Y

Shankar Joisa1, Joydeep Ghosh

1, Rakesh L Tanna

1, Aditya Team

1



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).


32 | P a g e

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


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).


33 | P a g e

Abstract ID: 0_133

Study of Neutral Particle Transport in Aditya Tokamak Plasma using DEGAS2 Code

Ritu Dey1, Joydeep Ghosh



1, Santanu Banerjee

1,

Nilam Ramaiya1, Aditya Team

1



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).


34 | P a g e

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, Surya Kumar Pathak

1, SST-1 Team

1


2ITER-India, Institute for Plasma Research, India


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)


35 | P a g e

Abstract ID: 0_144

Equilibrium Reconstruction of Plasma Discharges in the Aditya Tokamak

Deepti Sharma1, Santanu Banerjee

1, Amit Kumar Singh


1, Raju

Daniel1, Rakesh L Tanna

1, Joydeep Ghosh

1, Y Shankara Joisa


1, Surya

Kumar Pathak1, Aditya Team

1


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)


36 | P a g e

Abstract ID: 0_149

Ohmic Discharges with Improved Confinement in Tokamak Aditya

Rakesh L Tanna1, Harshita Raj

1, Joydeep Ghosh


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,

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,

Ratneshwar Jha1, Amita Das

1, Dhiraj Bora

1


2Gujarat University, India


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.


37 | P a g e

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



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.


38 | P a g e

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, 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, Ranjana

Manchanda1, Yogesh C Saxena

1, Rabindranath Pal

2 and Aditya Team

1


2Saha Institute of Nuclear Physics, 1ndia


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.


39 | P a g e

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


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.


40 | P a g e

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


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.


41 | P a g e

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




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.


42 | P a g e

Abstract ID: 0_190

Divertor Coil Power Supply in Aditya Tokamak for improved Plasma Operation

Vaibhav Ranjan1, Kunal Shah

1, Motibhai N Makawana


1, A

Varadharajulu1, Joydeep Ghosh

1, Rakesh L Tanna


1, Raju Daniel,

Srinivasan Radhakrishnana1, Yogesh C Saxena

1



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.


43 | P a g e

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



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.


44 | P a g e

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,

Sameer Kumar Jha1, Santanu Banerjee

1, Santosh P Pandya

1, Shishir Purohit

1, Shwetang N

Pandya1, Snehlata Gupta


1, Umeshkumar C Nagora

1, Varsha Siju

1, Vishnu

K Chaudhari1, Y Shankar Joisa

1



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.


45 | P a g e

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



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.


46 | P a g e

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


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.


47 | P a g e

Abstract ID: 0_211

Observation of MHD Phenomenon for SST-1 Superconducting Tokamak

Manisha Bhandarkar1, Jasraj Dhongde

1, Subrata Pradhan

1



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.


48 | P a g e

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


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).


49 | P a g e

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


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.


50 | P a g e

Abstract ID: 0_220

Observation on Runaway Discharges in SST-1 Experiments

Kiritkumar B Patel1, Subrata Pradhan

1



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).


51 | P a g e

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



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.


52 | P a g e

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



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)


53 | P a g e

Abstract ID: 0_230

A Fixed Frequency Reflectometer to Measure Density Fluctuations at Aditya Tokamak

Parveen Kumar Atrey1,2

, Dhaval Pujara2, Subroto Mukherjee

1


2Nirma University, India


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).


54 | P a g e

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



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


55 | P a g e

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



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.


56 | P a g e

Abstract ID: 0_240

Installation and Commissioning of SST-1 Thomson scattering system

Jinto Thomas1, Vishal Pillai

1, Neha Singh

1, Kiran Patel


1, Ajai Kumar

1



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.


57 | P a g e

Abstract ID: 0_249

Limiter and Divertor Systems – Conceptual and Mechanical Design for Aditya Tokamak Upgrade

Kaushal Patel1, Kulav Rathod


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



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.


58 | P a g e

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


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.


59 | P a g e

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,

Kaushal M Patel1



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.


60 | P a g e

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



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).


61 | P a g e

Abstract ID: 0_268

Thermal Imaging of SST-1 Limiters

Shwetang N Pandya1, Santosh P Pandya

1, Kumar Ajay

1



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.


62 | P a g e

Abstract ID: 0_271

The Upgradation of Aditya Tokamak

Shailesh B Bhatt1, Joydeep Ghosh

1, Rakesh L Tanna

1, Chhaya Chavda


1,

Prabal K Chattopadhyay1, Raju Daniel


1, Kaushik S Acharya

1,

Kalpesh M Chaudhary1, Someswar Dutta


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



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


63 | P a g e

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.


64 | P a g e

Abstract ID: 0_277

Development of Non-circular Metal Seal for Aditya Tokamak Upgrade Vacuum Vessel

Kaushik S Acharya1, Kaushal M Patel


1, Kulav Rathod

1, Nilesh D Patel

1,

Kalpesh M Chaudhary1, Shailesh B Bhatt

1, Aditya Team

1



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.


65 | P a g e

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



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.


66 | P a g e

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,



1, Subrat Jana, Chet Narayan

Gupta, Subrata Pradhan1


[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.


67 | P a g e

Abstract ID: 0_304

Implementation of SST-1 plasma position control using vertical field

Kirti Mahajan1, Jasraj Dhongde



1, Aveg Kumar

1,



1, Subrat Jana, Chet Narayan

Gupta, Subrata Pradhan1


[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.


68 | P a g e

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



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


69 | P a g e

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



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).


70 | P a g e

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



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.


71 | P a g e

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



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.


72 | P a g e

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



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.


73 | P a g e



74 | P a g e

Abstract ID: 0_151

Progress of JT-60SA Construction and R&D of its Heating Systems

Yoshitaka Ikeda1

1Japan Atomic Energy Agency, Japan


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.


75 | P a g e

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


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


76 | P a g e

Abstract ID: 1_206

Status of the WEST Project

Jérôme Bucalossi1, WEST Team

1

1CEA, DSM, IRFM, France


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.


77 | P a g e

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


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)


78 | P a g e



79 | P a g e

Abstract ID: 0_287

Progress of Experiment on HL-2A

Xuru Duan1, HL-2A Team

1



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.


80 | P a g e

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


2Institute of Advanced Energy, Kyoto University, Japan

3Plasma Research Center, University of Tsukuba, Japan

4Research Institute for Applied Mechanics, Kyushu University, Japan


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.


81 | P a g e

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, Manisha Bhandarkar

1, Jasraj Dhongde

1, Braj

Kishore Shukla1, Imran Mansuri

1, Yohan Khristi

1, Yuvakiran Paravastu


1,

Dinesh Sharma1, Kalpeshkumar R Dhanani

1, Pratibha Semwal

1, Siju George

1, Subrata Jana

1,

Pradip Panchal1, Rohitkumar Panchal

1, Rakeshkumar Patel


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, 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, 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



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 èddy

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.


82 | P a g e

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


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


83 | P a g e

Poster Session-2


84 | P a g e

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


2I-Design Engineering Solutions Ltd., India


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.


85 | P a g e

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



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 %.


86 | P a g e

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


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.


87 | P a g e

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


88 | P a g e

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


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)


89 | P a g e

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



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/


90 | P a g e

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



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.


91 | P a g e

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


2I-DESIGN Engineering Solutions Ltd., Waghoili, Pune


“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.


92 | P a g e

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



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.


93 | P a g e

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



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.


94 | P a g e

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,

Pratibha Semwal1, Prashant Thankey

1, Mohammad Shoaib Khan

1, Subrata Pradhan

1



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.


95 | P a g e

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, Prashant Thankey

1, Pratibha

Semwal1, Arun Prakash A


1, Snehal Jaiswal

1, Pradeep Chauhan

1,

Subrata Pradhan1



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.


96 | P a g e

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, Siju George

1,

Yuvakiran Paravastu1, Arun Prakash A

1, Prashant Thankey


1, Mohammad

Shoaib Khan1, Partha Saikia

1, Subrata Pradhan

1



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.


97 | P a g e

Abstract ID: 1_79

Design and Integration of SMBI System for SST 1

Siju George1, Yuvakiran Paravastu



1,

Dilip Raval1, Ziauddin Khan

1, Santanu Banerjee

1, Subrata Pradhan

1



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.


98 | P a g e

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


2M. S. University, India


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.


99 | P a g e

Abstract ID: 1_93

Electron Beam Welding: Study of Process Capabilities and Limitations towards Development of Nuclear Components

Gautam Vadolia1, Kongkham Premjit Singh

1



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.


100 | P a g e

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



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.


101 | P a g e

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.


102 | P a g e

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


2DIII-D, General Atomics, USA


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.


103 | P a g e

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



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.


104 | P a g e

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


2DMSRD establishment, Kanpur, India


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).


105 | P a g e

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



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.


106 | P a g e

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,

Joydeep Ghosh2, J. Sharma

3

1Shri Vaishnav Institute of Technology and Science, India


3 M. B. Khalsa College, India


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).


107 | P a g e

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


2Bhabha Atomic Research Centre-Visakhapatnam, India



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


108 | P a g e

[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


109 | P a g e

Abstract ID: 1_117

Role of ECRH in SST-1 Tokamak Plasma

Braj Kishore Shukla1, Dhiraj Bora

1, Ratneshwar Jha

1, Subrata Pradhan

1



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.


110 | P a g e

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



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)


111 | P a g e

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


2G. B. Pant University of Agriculture & Technology, India


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).


112 | P a g e

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


2ITER-India, Institute for Plasma Research


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.


113 | P a g e

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



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.


114 | P a g e

Abstract ID: 1_132

Optimization of Geometrical Parameters for High Heat Flux Components (Vapotrons)

Sajal Thomas1, Shrishail B Padasalagi

1



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.


115 | P a g e

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




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


116 | P a g e

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



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.


117 | P a g e

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



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).


118 | P a g e

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, Subrata Pradhan

1



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.


119 | P a g e

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



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).


120 | P a g e

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



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.


121 | P a g e

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



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.


122 | P a g e

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,

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



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.


123 | P a g e

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


3Mahatma Gandhi Institute of Technology, India


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).


124 | P a g e

Abstract ID: 1_176

Assembly & Metrology of First Wall Components of SST-1

Tejas Parekh1, Prosenjit Santra

1, Prabal Biswas

1, Hiteshkumar Patel


1,

Snehal Jaiswal1, Pradeep Chauhan


1, Arun Prakash A

1, Dhaval Bhavsar

1,

Dilip Raval1, Ziauddin Khan

1, Subrata Pradhan

1



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.


125 | P a g e

Abstract ID: 1_189

Trap Site Formation and their Distribution Studies in Porous Lithium Titanate

Chandan Danani1, Manoj Warrier

2, Paritosh Chaudhuri

1




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.


126 | P a g e

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


3Vidyavardhini College of Engineering and Technology, India


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%.


127 | P a g e

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



3Vidyavardhini College of Engineering and Technology, India


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.


128 | P a g e

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




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).


129 | P a g e

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




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).


130 | P a g e

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



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.


131 | P a g e

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



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.


132 | P a g e

Abstract ID: 1_208

Overall Behaviour of PFC Integrated SST-1 Vacuum System

Ziauddin Khan1, Dilip Raval


1, Pratibha Semwal

1, Kalpeshkumar R

Dhanani1, Siju George


1, Arun Prakash A

1, Gattu Ramesh

1, Prashant

Thankey1, Firozkhan Pathan

1, Subrata Pradhan

1



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.


133 | P a g e

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



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.


134 | P a g e

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



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.


135 | P a g e

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



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


136 | P a g e

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



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.


137 | P a g e

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,

Upendra Prasad1



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.


138 | P a g e



139 | P a g e

Abstract ID: 4_200

ITER and its Diagnostics- the Way Ahead

Michael Walsh1, ITER Team

1

1ITER Organization, France


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.


140 | P a g e

Abstract ID: 4_292

Status of the Realization of the Neutral Beam Test Facility

Vanni Toigo1

1Consorzio RFX, Italy


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.


141 | P a g e

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


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.


142 | P a g e

Poster Session-3


143 | P a g e

Abstract ID: 1_225

Precision Electronics and Measurement Techniques for the Superconducting Joint Resistance

Yohan Khristi1, Subrata Pradhan

1, Kalpesh Doshi

1



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.


144 | P a g e

Abstract ID: 1_226

Preliminary Results from Electron Cyclotron Measurements at SST-1

Varsha Siju1, Praveena Kumari


1



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.


145 | P a g e

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


2CEA-IRFM, France


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:



146 | P a g e

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,

Arun Kumar Chakraborty1

1 ITER-India, Institute for Plasma Research, India

2Airframe Aerodesigns Pvt. Ltd., India 3Vacuum Techniques Pvt. Ltd, India


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.


147 | P a g e

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



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).


148 | P a g e

Abstract ID: 1_242

Density Measurement Systems at SST Tokamak

Umeshkumar C Nagora1, Surya Kumar Pathak


1



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.


149 | P a g e

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



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.


150 | P a g e

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



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).


151 | P a g e

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



3Shell-N-Tube Pvt. Ltd., India


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.


152 | P a g e

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



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).


153 | P a g e

Abstract ID: 1_259

Michelson Interferometer Diagnostics for Broadband ECE Measurement

Abhishek Sinha1, Surya Kumar Pathak

1



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.


154 | P a g e

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, Rohit Kumar

1, Harshita

Raj1, Krishnamachari Sathyanarayana

1, Kulav Rathod


1, Kaushal M

Patel1, Kaushik S Acharya


1, Ashok V Apte

2, Yogesh C Saxena

1,

Dhiraj Bora1, Shell-N-Tube Team

3


2Space Application Center, India 3Shell-N-Tube Pvt. Ltd., India


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.


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, Sharvil

Patel1, Vaibhav Ranjan

1, Rohit Kumar

1, Harshita Raj

1, Krishnamachari Sathayanarayana

1,

Madan B Kalal1, Dinesh S Varia


1, Kulav Rathod

1, Shailesh B Bhatt

1, A

Vardharajulu1, Yogesh C Saxena

1, Dhiraj Bora

1, Shell-N-Tube Team

2


2Shell-N-Tube Pvt. Ltd., India


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.


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



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).


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



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.


158 | P a g e

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


2Indira Gandhi Center for Atomic Research (IGCAR), India


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.


159 | P a g e

Abstract ID: 1_278

Electrical Properties of Nano Li2TiO3 for Fusion Reactors

S K S Parashar1, Kajal Parashar


2

1KIIT University, Bhubaneswar, India 2Institute for Plasma Research, India


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.


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



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.


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, Santanu

Banerjee1, Nilam Ramaiya

1, Joydeep Ghosh

1



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.


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



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.


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


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.


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


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).


165 | P a g e

Abstract ID: 2_16

Magnetic Probe Diagnostic Tool to Understand the Dynamics in a Non-transferred dc Plasma Torch

Vidhi Goyal1, Ravi Ganesh

1



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)


166 | P a g e

Abstract ID: 2_22

Localized solutions in Laser Plasma Coupled System with Periodic Time Dependence

Deepa Verma1, Amita Das

1, Bhavesh Patel


1



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).


167 | P a g e

Abstract ID: 2_23

Coupling of Drift Wave with Dust Acoustic Wave

Atul Kumar1, Amita Das


1



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).


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, Joydeep Ghosh

1, Yogesh C Saxena

1



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.


169 | P a g e

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



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)


170 | P a g e

[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.


171 | P a g e

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


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).


172 | P a g e

Abstract ID: 2_38

Relativistic Cylindrical and Spherical Plasma Waves

Arghya Mukherjee1, Sudip Sengupta

1



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)


173 | P a g e

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


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.


174 | P a g e

Abstract ID: 2_44

Electric Field Assisted Sintering (EFAST): Plasma?

Prasad Mattipally1, Devarasetty Suresh Babu

1

1Osmania University, India


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.


175 | P a g e

Abstract ID: 2_45

Dispersion of Linearly Polarized Electromagnetic Wave in Magnetized Quantum Plasma

Abhisek Kumar Singh1, Punit kumar

1

1University of Lucknow, India


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).


176 | P a g e

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



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).


177 | P a g e

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


3SOKENDAI (The Graduate University for Advanced Studies), Japan


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).


178 | P a g e

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.


179 | P a g e

Abstract ID: 2_48

Production of Quiescent Collisionless Plasma over a Wide Operating Range

Sayak Bose1, Manjit Kaur


1, Joydeep Ghosh

1, Yogesh C Saxena

1



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)


180 | P a g e

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


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.


181 | P a g e

Abstract ID: 2_51

Ponderomotive Force and Backward Raman Scattering in Dense Quantum Plasmas

Punit kumar1, Nisha Singh Rathore

1

1University of Lucknow, India


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).


182 | P a g e

Abstract ID: 2_52

Anode Glow and Double Layer in DC Magnetron Anode Plasma

Samir Chauhan1, Mukesh Ranjan

1, Subroto Mukherjee

1



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)


183 | P a g e

Abstract ID: 2_53

Effect of Trapped Particle Nonlinearity in IAW Solitary Wave

Debraj Mandal1, Devendra Sharma

1



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)


184 | P a g e

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




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.


185 | P a g e

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




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).


186 | P a g e

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


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.


187 | P a g e

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



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).


188 | P a g e

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


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.


189 | P a g e

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


2Government General Degree College Singur, Hooghly, India


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.


190 | P a g e

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


Low-frequency Magneto Hydrodynamic waves [1] in general and Alfvén 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 Alfvén 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 Alfvén 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 Alfvén 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 Alvénic 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).


191 | P a g e

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



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).


192 | P a g e

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



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).


193 | P a g e

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



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).


194 | P a g e

Abstract ID: 2_107

Development of Three Dimensional Magnetic Field Probe with Signal Conditioning Electronics

Kiran Patel1, Narayan Behera


1, Ajai Kumar

1



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).


195 | P a g e

Abstract ID: 2_121

State of Art Data Acquisition System for Large Volume Plasma Device

Ritesh Sugandhi1, Pankaj Srivastava


1, Prabhakar Srivastav

1, Lalit

Mohan Awasthi1, Shiban Krishna Mattoo

1, Vijay Parmar

2, Keyur Makadia

2, Ishan Patel

2,

Sandeep Shah2


2Optimized Solutions Private Limited, India


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


196 | P a g e

Popular Talk


197 | P a g e

Exploration of the Solar System & Beyond: The Indian Scene

Jitendra Nath Goswami1

1Physical Research Laboratory, Ahmedabad


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.


198 | P a g e

Oral Session-1


199 | P a g e

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


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).


200 | P a g e

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


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)


201 | P a g e

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


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.


202 | P a g e

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).


203 | P a g e

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


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.


204 | P a g e

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


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.


205 | P a g e

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


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


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:



206 | P a g e

Poster Session-4


207 | P a g e

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


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).


208 | P a g e

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


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).


209 | P a g e

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


2ITER-India, Institute for Plasma Research,India


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.


210 | P a g e

Abstract ID: 2_126

Measurement of Electron Energy Probability Function in Weakly Magnetized Plasma

Deiji Kalita1, Bharat Kakati

2, Bipul Kumar Saikia


3, Siddhartha Sankar

Kausik1





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)


211 | P a g e

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



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).


212 | P a g e

Abstract ID: 2_140

Investigation of Magnetic Drift on Transport of Plasma across Magnetic Field

Parismita Hazarika1, Bidyut Das

2, Monojit Chakraborty


2


2 ITER-India, Institute for Plasma Research, India


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.


213 | P a g e

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.


214 | P a g e

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



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.


215 | P a g e

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


2Instituto Superior Tecnico, Universidade De Lisboa



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).


216 | P a g e

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



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.


217 | P a g e

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


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.


218 | P a g e

Abstract ID: 2_165

Two Dimensional Imaging of Laser Produced Plasma in Magnetic field

Narayan Behera1, Rajesh Kumar Singh

1, Ajai Kumar

1



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.


219 | P a g e

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


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.


220 | P a g e

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



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.


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, Yogesh

Yeole1




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.


222 | P a g e

Abstract ID: 2_185

Study of Phase Space Structures in Driven 1D Vlasov Poisson Model

Pallavi Trivedi1, Rajaraman Ganesh

1



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)


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




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)


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


2Rutherford Appleton Laboratory, United Kingdom


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.


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, Sheroy Tata

1, Amit D Lad

1, G Ravindra

Kumar1



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).


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




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)


227 | P a g e

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



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).


228 | P a g e

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



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).


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


1


2Department of Physics, Jadavpur University, Kolkata, India


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.


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, Indranuj Dey

1, Amit D Lad

1, G Ravindra

Kumar1



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).


231 | P a g e

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


2Venus International College of Technology, India


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).


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, Pankaj Srivastava

1, Ratneshwar

Jha1, Raghvendra Singh


1


2WCI Center for Fusion Theory, National Fusion Research Institute, Korea


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).


233 | P a g e

Abstract ID: 2_252

Exhibiting Electrons in Nanoplasmas: An Estimate

Jagannath Jha1



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.


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



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).


235 | P a g e

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



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).


236 | P a g e

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


2Chinese Academy of Sciences, China

3Shanghai Jiao Tong University, China


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.


237 | P a g e

Abstract ID: 2_276

DAQ System for Low Density Plasma Parameters Measurement

Rashmi S Joshi1, Suryakant B Gupta

1



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.


238 | P a g e

Abstract ID: 3_0

Numerical Study of Instabilities in Magnetized Inhomogeneous Plasmas

Jyoti Chaudhary1

1Manipal University Jaipur, Rajasthan, India


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.


239 | P a g e

Abstract ID: 3_11

Modeling of Electromagnetic Fields during Plasma Startup in SST-1 Tokamak

Amit Kumar Singh1, Indranil Bandyopadhyay


2, SST-1 Team

1




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.


240 | P a g e

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


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).


241 | P a g e

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


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


242 | P a g e

Abstract ID: 3_21

Betatron Radiation from Laser Wakefield Acceleration in a Plasma Channel

Devki Nandan Gupta1, Inhyuk Nam

2, Hyyong Suk

2


2Gwangju Institute of Science and Technology, South Korea


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).


243 | P a g e

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


2Bhabha Atomic Research Centre, India


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)


244 | P a g e

Abstract ID: 3_43

PIC Modeling of Negative Ion Extraction from a Dust-Seeded Plasma

Ananya Phukan1, Pranjal Bhuyan

1


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)


245 | P a g e

Abstract ID: 3_62

Dynamics of dusty fluid in a streaming sheared plasma

Modhuchandra Laishram1, Devendra Sharma


1



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).


246 | P a g e

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


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.


247 | P a g e

Abstract ID: 3_76

Gyro-TWT in a Vane-Loaded Waveguide with Inner Dielectric

Reyhaneh S. Hashemi1, Elaheh Allahyari

1, Shahrooz Saviz

1



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.


248 | P a g e

Abstract ID: 3_77

Effect of Plasma Column on the Radial Profile of Electric Field of Gyrotron Devices

Elaheh Allahyari1, Shahrooz Saviz

1



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.


249 | P a g e

Abstract ID: 3_83

Current Gradient Modes of Two Dimensional Electron Magnetohydrodynamics (EMHD)

Gurudatt Gaur1, Predhiman Krishan Kaw

1



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).


250 | P a g e

Abstract ID: 3_86

A Poynting like Theorem for Generalized Hydrodynamic Equations

Vikram Singh Dharodi1, Bhavesh Patel

1, Amita Das


1



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.


251 | P a g e

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



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).


252 | P a g e

Abstract ID: 3_128

1D PIC simulation of relativistic Buneman instability

Roopendra Singh Rajawat1, Sudip Sengupta

1



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.


253 | P a g e

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



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).


254 | P a g e

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


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.


255 | P a g e

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


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.


256 | P a g e

Abstract ID: 3_150

Potential around a dust grain in collisional plasma

Rakesh Moulick1, Kalyan Sindhu Goswami

1



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.


257 | P a g e

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



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.


258 | P a g e

Poster Session-5


259 | P a g e

Abstract ID: 3_180

Nonlinear MHD modeling in LHD plasmas with peaked pressure profiles

Yasuhiro Suzuki1,2


2SOKENDAI, The Graduate University for Advanced Studies, Japan


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


260 | P a g e

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



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.


261 | P a g e

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




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).


262 | P a g e

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


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.


263 | P a g e

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,

Chhaya Chavda1, Vipul K Panchal

1, Vijay Patel


1, Raju Daniel

1, Aditya

Team1



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.


264 | P a g e

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


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.


265 | P a g e

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



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).


266 | P a g e

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




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).


267 | P a g e

Abstract ID: 4_17

Nuclear Analyses of Indian LLCB Test Blanket System in ITER

H L Swami1, Akshaya Kumar Shaw

1, Chandan Danani


1



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


268 | P a g e

Abstract ID: 4_28

Preferential Binding of Self-interstitial Atoms over Vacancies to Grain Boundaries of Tungsten: A Lattice Statics Study

Prithwish K Nandi1



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.


269 | P a g e

References:

[1] T. Wanatabe,“The impact of grain boundary character distribution on fracture in

polycrystals”, Mater. Sci. Eng. A, 176, 39 (1994).


270 | P a g e

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



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).


271 | P a g e

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


2Bhakti Consultants, India

3ITER-Organization, France

4UPES Dehradun, India


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.


272 | P a g e

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



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.


273 | P a g e

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



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.


274 | P a g e

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


1




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.


275 | P a g e

Abstract ID: 4_95

Development and Validation of I-Activation Analysis Code

Sai Chaitanya Tadepalli1, P V Subhash

1, Gunjan Indauliya

2


2Bhakti Consultants, India


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.


276 | P a g e

Abstract ID: 4_97

Indigenous Manufacturing Realization of Twin Source and its Auxiliary System

Ravi Pandey1, Mainak Bandyopadhyay

2, Deepak Parmar


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




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.


277 | P a g e

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



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.


278 | P a g e

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




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.


279 | P a g e

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



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.


280 | P a g e

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



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)


281 | P a g e

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



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).


282 | P a g e

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,

Chandramouli Rotti2, Arun Kumar Chakraborty

2




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


283 | P a g e

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


2The University of Texas at Austin, USA



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.


284 | P a g e

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


2,

Mahesh Vuppugalla1, Himanshu Tyagi

2, Kanubhai Parmar

1, Hiren Mistri

2, Jignesh Bhagora

1,

Bhavesh K Prajapati1, Kartik J Patel

1, Manar Bhuyan


1, Arun Kumar

Chakraborty2,




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.


285 | P a g e

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


286 | P a g e

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


2ITER-Organisation, France


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.


287 | P a g e

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





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.


288 | P a g e

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





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.


289 | P a g e

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



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.


290 | P a g e

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


2Princeton Plasma Physics Laboratory, USA,

3Institute for Fusion Studies,The University of Texas at Austin, USA,



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.


291 | P a g e

Abstract ID: 4_234

Manufacturing Experience of an ‘Angled’ Accelerator Grid for DNB Beam Source

Jaydeep Joshi1



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.


292 | P a g e

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



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.


293 | P a g e

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


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.


294 | P a g e

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




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.


295 | P a g e

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




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.


296 | P a g e

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


3Assystem EOS, France


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.


297 | P a g e

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



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.


298 | P a g e

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


2Kampf Telescope Optics GmbH, Munich, Germany

3Bertin Technologies, France

4Indira Gandhi Center for Atomic Research (IGCAR), India


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


299 | P a g e

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


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.


300 | P a g e

Abstract ID: 5_84

Pesticides Removal from Cabbage using Array of Atmospheric Pressure Plasma Jet

Akshay Vaid1, Chirayu Patil

1, Ramkrishna Rane


1, Sudhir Nema

1, Hetal

Bhatt2, R V Prasad

2


2Anand Agriculture University, India


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.


301 | P a g e

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



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.


302 | P a g e

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


2




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)


303 | P a g e

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


2

1FCIPT, Institute for Plasma Research, India



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).


304 | P a g e

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


3Bharathiar University, India


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).


305 | P a g e

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


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)


306 | P a g e

Abstract ID: 5_280

Plasma Sterilization for Bio-decontamination

Suryakant Gupta1, Sudhir Nema

1



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.


307 | P a g e

Oral Session-2


308 | P a g e

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



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.


309 | P a g e

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


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.


310 | P a g e

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


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.


311 | P a g e

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


2 CEA, IRFM, France


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.


312 | P a g e

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


2CEA, IRFM, France


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.


313 | P a g e

Oral Session-3


314 | P a g e

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


2Graduate University for Advanced Studies, Japan


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).


315 | P a g e

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


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.


316 | P a g e

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



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.


317 | P a g e

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


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


318 | P a g e

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


2Chubu University, Japan

3Hamamatsu Photonics Inc., Japan


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).


319 | P a g e

Oral Session-4


320 | P a g e

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

1Institute for Plasma Research, Inida


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.


321 | P a g e

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


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).


322 | P a g e

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


1


2Weizmann Institute of Science, Israel


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).


323 | P a g e

Abstract ID: 0_34

Thermionic Divertors for Tokamaks

Avinash Khare1, Sanjay K Mishra


2


2Institute or Plasma Research, India


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.


324 | P a g e

Abstract ID: 3_181

Modeling of ITER Disruption scenarios using TSC

Indranil Bandyopadhyay1, Amit K Singh



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


325 | P a g e

Oral Session-4


326 | P a g e

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



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.


327 | P a g e

Abstract ID: 1_289

Cryogenic Technology of the New Millennium – Competence of DH Industries

Ronald Den Heijer1

1DH Industries, The Netherlands


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.


328 | P a g e

Abstract ID: 1_207

Upgradation Plans of SST-1 Cryogenics System at IPR

Vipul L Tanna1, SST-1 Cryo Team

1, Subrata Pradhan

1



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.

Documents

10th Asia Plasma & Fusion Association Conference Book APFA-2015.pdf · 10th Asia Plasma & Fusion Association Conference . APFA 2015 . 10th Asia Plasma & Fusion Association Conference