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Page 1: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

EPS2005, Session "P-1" Abstracts

Session Author PosterTitle

P-1.001 P.Träskelin Molecular dynamics simulation of erosion of tungsten carbide by deuterium bombardment

P-1.002 B.N.Bazylev Erosion of dome armour after multiple disruptions and ELMs in ITER

P-1.003 I.S.Landman Contamination and radiation losses in post-ELM tokamak plasma

P-1.004 T.Lunt Experimental Investigation on the Plasma-Wall Transition

P-1.005 TilmannLunt Ion temperature measurements by means of a combined force - Mach - Langmuir probe

P-1.006 A.Herrmann Filamentary heat deposition to the first wall in ASDEX Upgrade

P-1.007 B.KurzanFine Structure of Type-I Edge Localized Modes in the Steep Gradient Region in ASDEX Upgrade

P-1.008 D.P.Coster Edge simulations of an ASDEX Upgrade Ohmic shot

P-1.009 H.W.Mueller Plasma flow in the scrape-off layer of ASDEX Upgrade

P-1.010 R.Dux Tungsten Erosion at Auxiliary Limiters in ASDEX Upgrade

P-1.011 V.Rohde Carbon migration at the divertor of ASDEX Upgrade

P-1.012 Y.Feng Role of recycling in W7-AS divertor plasmas

P-1.013 A.Kirschner Modelling of tritium retention and target lifetime of the ITER divertor

P-1.014 A.KreterInvestigation of carbon transport by 13CH4 injection through graphite and tungsten test limiters in TEXTOR

P-1.015 A.LitnovskyCarbon deposition and fuel accumulation in castellated limiters exposed in the SOL of TEXTOR

P-1.016 C.Busch Impact of the DED on ion transport and poloidal rotation in TEXTOR

P-1.017 D.BorodinModelling of hydrocarbon transport and emission after methane injection into the TEXTOR boundary plasma using the ERO code

P-1.018 G.Sergienko High temperature erosion of tungsten exposed to the TEXTOR edge plasma

P-1.019 G.Sergienko Tungsten melting under high power load in the TEXTOR edge plasma

P-1.020 G.Telesca Screening and radiation efficiency of carbon with Dynamic Ergodic Divertor on TEXTOR

P-1.021 M.W.JakubowskiOn the influence of the magnetic resonances on the heat flux structure of the Dynamic Ergodic Divertor

P-1.022 OliverSchmitz Impact of the Dynamic Ergodic Divertor on the Structure of the Plasma Edge at TEXTOR

P-1.023 S.S.Abdullaev Structure of stochastic field lines near the separatrix in poloidal divertor tokamaks

P-1.024 V.Philipps Removal of carbon layers by oxygen treatment of TEXTOR

P-1.025 A.S.Kukushkin Improved Modelling Of Neutrals And Consequences For The Divertor Performance In Iter

P-1.026 O.V.Ogorodnikova Simulation of brittle destruction of different types of graphite using PEGASUS-3D code

P-1.027 O.V.OgorodnikovaParametric investigation of temperature and stress evolution in actively cooled plasma-facing components during high heat fluxes

P-1.028 M.K.Salem The Influence of Resonant Helical Field on The Zeff in IR-T1 Tokamak

P-1.029 M.Kuldkepp Oxygen impurity profile studies in the EXTRAP T2R reversed field pinch

P-1.030 J.JuulRasmussen Turbulent Transport and Mixing of Impurities in the Plasma Edge

P-1.031 M.Priego Clustering and pinch of impurities in plasma edge turbulence

Page 2: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

P-1.032 F.G.RiminiHigh Power ICRH scenarios in Tore-Supra a potential route towards improved core confinement at high density

P-1.033 D.Elbèze Scaling of confinement in the ITPA L-mode database with dimensionless variables

P-1.034 F.ImbeauxGiant Oscillations of Electron Temperature during zero loop voltage discharges on Tore Supra

P-1.035 Jean-FrançoisArtaud Predictive integrated modelling for ITER scenario

P-1.036 P.DevynckThe origin of the long time correlations of the density fluctuations in the Scrape off Layer of the Tore Supra Tokamak

P-1.037 V.S.UdintsevElectron Temperature Fluctuation Studies in Different Confinement Regimes by Means of Correlation ECE on Tore Supra

P-1.038 G.Fuhr Zero Dimensional Model for Transport Barrier Oscillations in Tokamak Edge Plasmas

P-1.039 R.JhaStudy of nonlinear phenomena in a tokamak plasma using a novel Hilbert transform technique

P-1.040 JoyantiChutiaLong range time correlations in the electrostatic fluctuations of a low temperature dc magnetised plasma

P-1.041 M.Aizawa Transport Properties of Low Aspect Ratio L 1 Helical Systems

P-1.042 H.TakenagaTransient electron heat transport and reduced density fluctuation after pellet injection in JT-60U reversed shear plasmas

P-1.043 M.Kikuchi Measurement of local electrical conductivity and thermodynamical coefficients in JT-60U

P-1.044 Y.Idomura Comparisons of gyrokinetic PIC and CIP codes

P-1.045 N.Ohno Intermittent Fluctuation Property of JT-60U Edge Plasmas

P-1.046 Y.YagiFirst results of the Gas Puffing Imaging Diagnostics in a reversed-field pinch plasma First results of the Gas Puffing Imaging Diagnostics in a reversed-field pinch plasma First results of the Gas Puffing Imaging Diagnostics in a reversed-field pinch plasma

P-1.047 J.MiyazawaWeak temperature dependence of the thermal diffusivity in high-collisionality regimes in LHD

P-1.048 M.ElMouden3D Simulation of the Magnetic Shear contribution on the Improvement of the Confinement in Plasma of Tokamak

P-1.049 A.Scarabosio Momentum transport and plasma rotation spin up in TCV

P-1.050 Ch.SchlatterSimulation of the Absolute TCV Compact Neutral Particle Analyser Charge-Exchange Spectrum

P-1.051 E.FableDensity behavior during eITBs in TCV discharges experimental observations and theoretical calculations via transport simulations

P-1.052 Y.CamenenElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas

P-1.053 R.O.Dendy Analysis of dissipation in MHD turbulence simulations in two and three dimensions

P-1.054 OKwon Numerical Plasma Edge MHD Stability Analysis Revisited

P-1.055 S.S.KimEffects of radio frequency waves on dissipative low frequency instabilities in mirror plasmas

P-1.056 R.Jiménez-Gómez Studies of MHD instabilities in TJ-II plasmas

P-1.057 T.S.Pedersen First results from the Columbia Non-neutral Torus

P-1.058 X.Sarasola Field Line Mapping Results in the CNT Stellarator

P-1.059 J.F.Lyon Recent Developments In Quasi-Poloidal Stellarator Physics

P-1.060 B.Stratton Fast soft x-ray camera observation of fast and slow reconnection events on NSTX

P-1.061 EDFredrickson Scaling of kinetic instability induced fast ion losses in NSTX

P-1.062 M.C.Zarnstorff Equilibrium of High-Beta Plasmas in W7-AS

Page 3: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

P-1.063 N.N.Gorelenkov Resonant kinetic ballooning modes in burning plasma

P-1.064 R.Raman Transient CHI Solenoid-free Plasma Startup in NSTX

P-1.065 QingweiYang Investigations of disruption on the HL-2A tokamak

P-1.066 HogunJhangA Toroidal Shell Model for Active Stabilization of Resistive Wall Modes and Its Application to KSTAR Plasmas

P-1.067 Y.M.Jeon Design of Optimal Plasma Position and Shape Controller for KSTAR

P-1.068 E.J.Strait Feedback Stabilization of Resistive Wall Modes in DIII-D

P-1.069 J.R.Ferron Control of DIII-D Advanced Tokamak Discharges

P-1.070 T.A.Casper Operational Enhancements in DIII-D Quiescent H-Mode Plasmas

P-1.071 R.Raman Fueling Requirements for Advanced Tokamak operation

P-1.072 B.E.Chapman Initial exploration of the density limit in the MST RFP

P-1.073 R.Cavazzana Optical Investigation of Edge Turbulence on RFX-mod

P-1.074 E.Gazza Fast optical spectrometer for the charge exchange diagnostic on RFX-mod

P-1.075 C.Mazzotta Study of Plasma density profiles evolution using the new scanning interferometer for FTU

P-1.076 G.DeTemmerman Mirror Test for ITER Optical Characterisation of Metal Mirrors in Divertor Tokamaks

P-1.077 E.Gauthier Design of a wide-angle infrared thermography diagnostic for JET

P-1.078 L.BertalotNeutron energy measurements of Trace Tritium plasmas with NE213 compact spectrometer at JET

P-1.079 A.HjalmarssonDevelopment of new neutron emission spectrometry diagnostics for fusion experiments at JET

P-1.080 M.Gatu-JohnsonDiagnosis of high-energy fuel ions on ITER with neutron emission spectroscopy NES Monte Carlo calculations based on NES measurements on JET DT plasmas

P-1.081 MarcoTardocchiMPR neutron emission spectroscopy of fast tritons from T D ion cyclotron heating in JET plasmas

P-1.082 V.StancalieNew method to calculate the Gaunt factor for the refinement of Zeff evaluation in fusion plasmas

P-1.083 G.BonheureFirst study of 2-D spatial distribution of D-D and D-T neutron emission in JET Elmy H-mode plasmas with Tritium puff

P-1.084 M.E.Notkin Absorption experiments on the CASTOR tokamak

P-1.085 A.A.Lizunov Mse-Diagnostic For Multi-Chord Measurents Of Plasma Beta In Gdt

P-1.086 P.A.Bagryansky Dispersion Interferometer based on CO2 - laser

P-1.087 GusakovE.Z.Investigation of the Upper Hybrid Resonance Cross-Polarization Scattering Effect at the FT-2 Tokamak

P-1.088 A.Popov Spatial Resolution of Poloidal Correlation Reflectometry

P-1.089 I.I.Orlovskiy Hilbert Spectrum Analysis of Mirnov Signals

P-1.090 K.Yu.Vukolov Mitigation of hydrocarbon film deposition on in-vessel mirrors

P-1.091 Yu.V.Gott A Vacuum Photoemission Detector for X-ray Tomography

P-1.092 D.P.Kostomarov Calculation of Plasma Boundary Using Video Images

P-1.093 V.Yu.Sergeev Fast Electron Studies In T-10 Plasmas By Means Of Carbon Pellet Injection

P-1.094 G.VanWassenhove Study of the ICRH antenna coupling at TEXTOR

P-1.095 S.Nowak Electron Cyclotron Current Drive experiments in the FTU tokamak

P-1.096 E.Barbato Interpretation od LHCD efficiency scaling with the electron temperature

P-1.097 E.Giovannozzi Plasmoid drift during vertical pellet injection in FTU discharges

Page 4: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

P-1.098 G.Granucci Quantification of suprathermal current drive on FTU

P-1.099 A.V.Voronin Injection of intense plasma jet in the spherical tokamak Globus-M

P-1.100 IJenkins Off-Axis NBI fast ion dynamics in Trace Tritium Experiment

P-1.101 N.V.Sakharov Behavior of Ions in Auxiliary Heating Experiments in Globus-M Spherical Tokamak

P-1.102 V.A.Kornev First experiments on NBI in the TUMAN-3M tokamak

P-1.103 V.B.Minaev Study of the Beam - Plasma Interaction in the Globus-M Spherical Tokamak

P-1.104 L.N.Khimchenko Radiative power piculiarities during impurity pellet injection into T-10 plasmas

P-1.105 V.G.Kapralov Recent results of hydrogen pellet injection

P-1.106 N.B.Rodionov ICRF Heating together with neutral beams in Volume Neutron Sources JUST-T

P-1.107 T.Bolzonella Overview of global MHD behaviour in the modified RFX Reversed Field Pinch

P-1.108 G.Cenacchi The scientific program of the Ignitor experiment

P-1.109 W.Kernbichler Simple criteria for optimization of trapped particle confinement in stellarators

P-1.110 W.Kernbichler Neoclassical transport for LHD in the 1/ nu regime analyzed by the NEO code

P-1.111 W.KernbichlerCalculation of neoclassical transport in stellarators with finite collisionality using integration along magnetic field lines

P-1.112 K.Schoepf Fast Ion Confinement in Tokamak Current Hole Regimes

P-1.113 A.Nicolai Modelling of Plasma Rotation under the Influence of Helical Perturbations in TEXTOR

P-1.114 Y.KikuchiModelling of the penetration process of externally applied magnetic perturbation of the DED on TEXTOR

P-1.115 R.Preuss Stellarator scaling considering uncertainties in machine parameters

P-1.116 D.Sharma Role of stochasticity in W7-X edge transport

P-1.117 R.Coelho Effect of Alfvén resonances on the penetration of error fields on a rotating viscous plasma

P-1.118 J.-E.Dahlin Advanced Reversed Field-Pinch Confinement Scaling Laws

P-1.119 Y.Q.Liu A Uniform Framework to Study Resistive Wall Modes

P-1.120 A.K.Wang An improved fluid description on toroidal ITG modes

P-1.121 J.Urban Methodology of electron Bernstein wave emission simulations

P-1.122 S.SinmanA Novel ST Configurative Events with Controllable and Reproducible Alternative Self-organization Process

P-1.123 B.Labit Drift waves in the TORPEX toroidal plasma device

P-1.124 M.PodestaExperimental studies of plasma production and transport mechanisms in the toroidal device TORPEX

P-1.125 T.Hiraishi Formation of Very Deep Potential Well with Electrode Biasing in a Toroidal Device

P-1.126 A.Stark Ion dynamics in a collisionless magnetic reconnection experiment

P-1.127 F.M.Aghamir Eigen Modes of a Dielectric Loaded Coaxial Plasma Waveguide

P-1.128 A.R.BabazadehStudy of Gas Admixture Influences On The Pinch Dynamics In A 90 kJ Filippov Type Plasma Focus

P-1.129 V.A.Rantsev-Kartinov Local Destruction of Magnetic Surfaces and Impurity Distributions in Òokamak

P-1.130 E.A.Evangelidis Angular momentum coupling in tokamaks

P-1.131 F.Porcelli Long term evolution of 3D collisionless magnetic reconnection

P-1.132 C.IonitaQualitative similarities between edge localised modes ELMs in fusion plasmas and complex space charge configurations CSCCs in low-temperature plasmas

P-1.133 Z.P.Xu Diagnosis of Wire-Array Z-Pinch Implosion Using X-ray Framing Cameras

Page 5: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

P-1.134 S.Dan'ko Elaboration of High-Current Drivers Aimed at the Inertial Fusion Energy

P-1.135 J.M.PerladoInertial Fusion Reactor Physics effect of Activation and Radiation Damage of Materials, and Tritium emissions.

P-1.136 Ph.Nicolaï A practical nonlocal model for electron transport in magnetized laser-plasmas

P-1.137 WenluZhang Evolution of Rayleigh-Taylor Instability with Arbitrary Density Profiles

P-1.138 M.Kaluza Self-Generated Magnetic Field Distributions in Multiple-Beam Produced Plasmas

P-1.139 N.Ozaki Laser-driven flyer impact experiments on LULI 2000 laser facility

P-1.141 T.PisarczykOptical investigation of flyer disk acceleration and collision with massive target on the PALS laser facility

P-1.142 S.BorodziukNumerical modelling of strong shock waves and craters for the experiments using single and double solid targets irradiated by high power iodine laser PALS

P-1.143 G.GregoriExperimental characterization of a strongly coupled solid density plasma generated in a short-pulse laser target interaction

P-1.144 L.TorrisiIon energy measurements in laser-generated plasmas at INFN-LNS and PALS research centre

P-1.145 K.B.Fournier Absolute x-ray yields from laser-irradiated Ge-doped aerogel targets

P-1.146 B.Sharkov Stopping Power Measurements for 100-keV/u Cu2 Ions

P-1.147 J.Wolowski Interaction of high-energy laser pulses with plasmas of different density gradients

P-1.148 S.DepierreuxThomson scattering of electron plasma waves stimulated by Raman backscattering in gasbag plasmas

P-1.149 S.F.Martins High intensity B field generation in underdense plasmas and the Inverse Faraday Effect

P-1.150 J.E.Santos Stimulated Raman Scattering with broadband effects

P-1.151 M.D.Barriga-Carrasco H2 distributions after traversing plasma targets

P-1.152 R.FedosejevsHeating of Tantalum Plasma for Studies on the Activation of the 6.238 keV Nuclear Level of Ta-181

P-1.153 L.O.Silva Stimulated Brillouin scattering by broadband radiation sources

P-1.154 KevinLewisAnalysis of the propagation of a laser beam through a preformed plasma using imaging diagnostics

P-1.155 F.Girard Experimental multi-keV x-ray conversion efficiencies from laser exploded germanium foil.

P-1.156 JonHowe Periodic features modifying the Heb line profile from an aluminium plasma

P-1.157 N.V.Vvedenskii Generation of Terahertz Radiation during Optical Breakdown of a Gas

Page 6: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Molecular dynamics simulations

of erosion of tungsten carbide by deuterium bombardment

Petra Träskelin, Kai Nordlund, Juhani Keinonen

Association Euratom-TEKES, Accelerator Laboratory,

P.O.B. 43, FI-00014 University of Helsinki, Finland

The selection of plasma facing materials for present and next-generation fusion devices is

still an open question. Refractory metal carbides are interesting candidates due totheir low

sputtering yields. Metal carbides are not only naturally present at the interfaces between the

carbon first wall and the metallic parts underneath in fusion reactors, but also formed when

hydrocarbon molecules which have been eroded under particle bombardment react with metal

parts in other sections of the plasma chamber. Mixed WC layers could therefore be formed due

to redeposition of eroded hydrocarbon molecules.

The most relevant metal carbide to be considered in this context is tungsten carbide. By us-

ing a reactive WCH-potential we have performed molecular dynamics simulations to elucidate

processes occurring under device operation at the reactor first walls. Tungsten carbide is an

extremely hard material and might provide a compromise between the materialswhich are cur-

rently dominating in the plasma chamber, tungsten and carbon, since the erosion yield is small.

The erosion due to low-energy H on these surfaces was investigated in more detail by per-

forming cumulative simulations of deuterium impinging onto WC structures. As a result, we see

that amorphous WC surface layers are formed regardless of the initial WC structure. Loosely

bound hydrocarbons on these surfaces can erode by the swift chemical sputtering mechanisms.

The threshold for C erosion from WC due to D by this mechanism is less than 10 eV,much less

than the threshold of about 60 eV predicted by physical sputtering models. Thismeans that also

mixed WC-like materials can be expected to be subject to chemical erosion of C down to very

low energies of impinging D or T particles, just like C-based divertor materials. The W erosion

yields are not subject to chemical erosion, meaning that there is preferential chemical sputtering

of C. The C content in thin WC layers formed by C redeposition might thus bereduced under

D bombardment.

P-1.001, Monday June 27, 2005

Page 7: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Erosion of dome armour after multiple disruptions and ELMs in ITER

B.N. Bazylev1, G.Janeschitz2, I.S. Landman1, S.E. Pestchanyi1 1Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe, Germany

2Forschungszentrum Karlsruhe, Fusion, P.O. Box 3640, 76021 Karlsruhe, Germany

In the future tokamak ITER a part of confined plasma is dumped onto the divertor

armour, during intense transient events such as disruptions and bursts of the Edge Localized

Mode (ELM). This may result in a surface melting and further evaporation. During one

ITER discharge about 103 ELMs are expected, and during ITER operation several hundred

disruptions, interspaced by ELMs may occur. The heat load of a single giant ELM or a

disruption causes a plasma shield being formed from evaporated material in front of the

target. This shielding layer is a source of intense radiation at GW/m2 level with durations of

0.5 ms for ELMs and up to 10 ms for the disruptions. Intense radiation from the vapour

shield may leads to enhanced erosion of nearby dome armour.

Pure sintered W or tungsten lamellae are foreseen as armour for the dome. In case of

W armour the main mechanisms for damage are surface melting and melt motion. Melt

motion in the thin layer may produce surface roughnesses and droplet splashing thus

causing erosion and determining the lifetime of the armour.

For tungsten dome armour the results of fluid dynamics simulation of the melt motion

erosion after repetitive radiation heat loads caused by multiple disruptions with the energy

deposition Q of 10-30 MJ/m2 and the duration t of 1-10 ms and of multiple ELMs with Q=

1-3 MJ/m2 and t= 0.1-0.5 ms are presented. For different single disruptions and ELMs, the

heat loads at the divertor surface and the radiation at the lateral walls are calculated using

the two-dimensional MHD code FOREV-2D. Reduction of the radiation heat load due to

absorption in the material vaporized from the dome surface is taken into account. The target

melt motion erosion is calculated by the fluid dynamics code MEMOS-1.5D in the ‘shallow

water’ approximation, with the surface tension and viscosity of molten metal taken into

account. The evaporation and melt motion erosion for different types of tungsten armour is

analyzed.

P-1.002, Monday June 27, 2005

Page 8: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Contamination and radiation losses in post-ELM tokamak plasma

I. S. Landman, G. Janeschitz, S.E. Pestchanyi

Forschungszentrum Karlsruhe, IHM, Post box 3640 D-76021, Karlsruhe, Germany

Future tokamaks such as ITER are going to operate in the H-mode regime with

repetitive edge localized instabilities. At each burst of ELM the lost DT plasma comes

from the scrape-off layer at the divertor armour surface. After such a pulse of the size

0.3-0.5 ms and 1-3 MW/m2 the surface emits eroded and then ionized armour materials

(carbon or tungsten) backwards. The resulting contamination of the SOL may cause

enhanced penetration of the impurities into the pedestal and core regions, which reduces

the reactor performance, for instance because of increased radiation losses.

In the last few years the influx of eroded material and accumulation of the impurity rich

plasma in the SOL have been investigated with the MHD code FOREV-2D, in which

radiation transport in a multi-fluid plasma is one of most developed features. In this

work further progress is reported on the behaviour of impurities in the post-ELM

plasma. The penetration of impurities inside the pedestal region of ITER and in the core

is modelled and radiation losses are estimated.

The confined DT plasma is modelled as a fluid in which an impurity of W- or C-ions is

assumed. The impurity concentration at the boundary (the separatrix) is calculated with

FOREV-2D as a function of time. A self-consistent two-dimensional model for pedestal

and core plasma diffusion and equilibrium in ITER magnetic field configuration is

developed. The plasma transport is based on the conductivities corresponding to the

neo-classical theory (ions) and a semi-empirical model for temperature gradient

turbulences.

P-1.003, Monday June 27, 2005

Page 9: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Experimental Investigation on the Plasma-Wall Transition

G.Fußmann1, T.Lunt1, N.Ezumi2

1Institut für Physik, Humboldt-Universität zu Berlin, AG Plasmaphysik

2Nagano National College of Technology, Tokuma, Nagano

The streaming of an argon plasma through a ‘magnetic’ nozzle magnetic field

configuration of the linear plasmagenerator PSI-2 towards an absorbing target plate was

studied experimentally by means of Laser Induced Fluorescence. This non-perturbative

diagnostic allows the measurement of the ion velocity distribution, and thus the streaming

velocity and the ion temperature in particular. Due to the nozzle effect induced by an

inhomogeneous B-field, the transition to supersonic flow velocities can take place far

away from the edge of the electrostatic sheath that builds up at the absorbing target plate.

Two different situations were observed: under standard neutral gas pumping conditions

half-sided Maxwellian ion-distributions as predicted by theory with Mach numbers

around M~0.5 were found. Decreasing the neutral density by maximum pumping affords,

supersonic fluxes with distributions clearly deviating form the Maxwellian case are

finally observed. Emphasis will be put on the interpretation of the half-sided distribution

functions.

P-1.004, Monday June 27, 2005

Page 10: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Ion temperature measurements by means of a combined force - Mach -

Langmuir probe

T.Lunt1, C.Hidalgo2, E.Calderón2

1Institut für Physik, Humboldt-Universität zu Berlin, AG Plasmaphysik

2Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas, CIEMAT

Chavers et al. [1] have proposed recently the use of a force sensor to measure momentum

fluxes in a plasma experiment. It has been shown experimentally that forces as small as

several mN are measurable in a non-fusion plasma discharge. Here we report on a

combined force – Mach – Langmuir probe. Comparing the force on the probe head,

which is proportional to cs2, with the difference of ion saturation currents at the two

opposite collectors (Mach probe), ∆Is ∝ cs, the speed of sound cs can be obtained. If

additionally the electron temperature is known from Langmuir probe measurements, the

ion temperature can be deduced. This could be a valuable diagnostic for the edge layer of

fusion devices which is until now only hardly covered by other methods. We will report

on measurements that are currently performed in the plasmagenerator PSI-2 in Berlin.

[1] D.G.Chavers, et al. Momentum flux measuring instrument for neutral and charged

particle flows, Review of Scientific Instruments, Vol. 73, No. 10, Oct. 2002

P-1.005, Monday June 27, 2005

Page 11: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Filamentary heat deposition to the first wall in ASDEX Upgrade

A. Herrmann1, J. Neuhauser

1, V. Rohde

1, W. Bobkov

1, T. Eich

1, A. Kirk

2, B. Kurzan

1,

H.-W. Müller1, ASDEX Upgrade team

1 Max-Planck-Institut für Plasmaphysik, EURATOM-IPP Association, Garching, Germany

2 EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, UK

The investigation of heat deposition to non-divertor components in present experiments and

its extrapolation to a next step device as ITER is essential because it would effect the design

and the necessary heat receiving capacity of the first wall. Measurements in different

tokamaks reveal that a fraction of up to 50 % of the energy ejected during ELMs from the

plasma is deposited outside the divertor. Heat fluxes of a few tens of MW/m2 with deposition

times in the order of a few hundred microseconds are observed by thermography in ASDEX

Upgrade. Independent investigations of the heat and particle transport in the outer midplane

of ASDEX Upgrade by reciprocating Langmuir probes, fast Thomson scattering system and

thermography reveal a burst like or filamentary structure.

The Thomson scattering system running in burst mode detects well separated density and

temperature blobs in the plasma edge and the SOL moving radially outward. A fast

reciprocating Langmuir probe in the outer midplane measures temporally structured ion

saturation currents during ELMs. These measured bursts may be interpreted as filaments

ejected from the plasma edge rotating toroidally and moving radially outward. Snapshot like

thermographic measurements of the heat load pattern at outside limiters show a poloidally

structured heat deposition which can be interpreted as filamentary heat deposition with

toroidal mode numbers of 10-25 lasting a few hundred microseconds. A subset of filaments

shows a fine structure in the heat deposition pattern on a few millimetre scale. This

filamentary heat deposition contributes with less than 1 % to the overall heat load on non-

divertor components. Nevertheless, they cause the maximum heat flux and as a consequence

the maximum surface temperature which is the main concern for the first wall design in a next

step device. Most of the ELM energy is deposited as ‘background’ with moderate heat fluxes.

Diagnostic improvements at ASDEX Upgrade allow now a combined investigation of the

filamentary transport in the SOL by Langmuir probes and thermography for the same ELM.

The evolution of the ELM structure can be resolved with a time resolution down to 100 µs by

thermography. A further improvement of the temporal resolution can be achieved by ELM

sorting according to the start of the ELM. In addition, the combined measurements allow

proofing the assumption of toroidal rotation and radial movement necessary to interpret the

probe measurements.

The results of these combined measurements and its interpretation in the framework of

existing filamentary transport models is presented in the paper.

P-1.006, Monday June 27, 2005

Page 12: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Fine Structure of Type-I Edge Localized Modes in the Steep Gradient

Region in ASDEX Upgrade

B. Kurzan, H. D. Murmann, J. Neuhauser and the ASDEX Upgrade Team

Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstr. 2, D-85748

Garching

The steep edge pressure gradient characteristic of the H-mode in tokamaks is relaxed by

quasi-periodically occuring edge-localized modes (ELMs). In the following type-I ELMs are

investigated. It is expected that during an ELM field-aligned structures at the plasma edge are

observed experimentally. 1D radial plasma profiles have been obtained so far by Thomson scat-

tering and reflectometry. These profiles are obtained eitheralong a radial line with a resolution

of several millimeters, or are averaged in the poloidal direction, or in time. These profiles only

show a flattening during the ELM. Radially and poloidally localized maxima have been found

in the scrape-off layer by many diagnostics on several tokamaks.

The Thomson scattering diagnostic in ASDEX Upgrade was recently upgraded by a new

data acquisition system. It is now possible to measure the fine structure of an ELM in the

electron density and temperature in the steep gradient and pedestal region, where the ELM

originates, as predicted theoretically. For this the 5 lasers of the Thomson diagnostic, which are

staggered radially by 1:5 mm and which for this investigation are located on the low field side

of the tokamak, are fired within 2µs. 2D images in the poloidal plane of the steep gradient and

pedestal region are thus obtained.

Before the ELM a locked precursor structure with a toroidal mode number of 10 is frequently

observed in the 2D images for the co-injected plasmas investigated so far. During the ELM lo-

calized maxima (‘blobs’) around the separatrix and corresponding minima (‘holes’) towards

the pedestal are observed frequently during an ELM. The deduced toroidal mode numbers of

the blob structures are in the range between 8 and 20. This is in agreement with theoretical

predictions for the most unstable peeling-ballooning modes. This new experimental result con-

firms the physical model that type-I ELMs originate indeed inthe steep gradient region, and

not e. g. in the scrape-off layer where such structures were observed so far. The number of

independent filaments existing in the scrape-off layer was scaled from the Thomson scattering

results to be around 80 during the ELM. The particles lost to the divertor by these filaments is

in the range of the globally lost particles during an ELM of some percent. This is in agreement

with results obtained with Langmuir probes on other tokamaks.

P-1.007, Monday June 27, 2005

Page 13: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Edge simulations of an ASDEX Upgrade Ohmic shot

D.P. Coster, A. Chankin, G. Conway, L. Kaveeva∗, C. Konz, M. Reich, T. Ribeiro, V.

Rozhansky∗, J. Schirmer, B.D. Scott, S. Voskoboynikov∗ and the ASDEX Upgrade Team

Max-Planck-Institut fur Plasmaphysik, EURATOM Association, D-85748 Garching beiMunchen, Germany

∗St. Petersburg State Polytechnical University, St. Petersburg, Russia

Excellent measurements of the upstream edge electron and ion temperatures, electron

density and electric field have been made for the AUG “standard” Ohmic shot. SOLPS

simulations have been performed to produce the best possible match of the upstream

temperature and density measurements under various assumptions, producing estimates

of the radial energy and particle transport coefficients. Under the assumption of equal

power flows via the electrons and ions at the core boundary, the ion thermal diffusivity

was found to be 0.52 m2s−1 and the electron thermal diffusivity 0.44 m2s−1. The ion

neo-classical thermal diffusivity was found to be about 0.1 m2s−1. For slightly lower

density conditions, gyro-fluid turbulence simulations based on the experimentally mea-

sured gradient lengths found the ion thermal diffusivity to be 0.15 m2s−1 and the electron

thermal diffusivity 0.49 m2s−1, but with about 1/6 of the power in the ion channel. In

addition, when run with the drift terms enabled in SOLPS, the calculated radial electric

field compares well with that measured.

P-1.008, Monday June 27, 2005

Page 14: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Plasma flow in the scrape-off layer of ASDEX Upgrade

H.W. Müller1, V. Bobkov1, V. Rohde1, M. Maraschek1, J. Neuhauser1, A. Schmid1,

M. Tsalas2 and ASDEX Upgrade Team1

1 Max-Planck-Institut für Plasmaphysik, EURATOM-Association,

D-85748 Garching, Germany2 NCRS Demokritos, Inst. of Nuclear Technology - Rad. Prot., Attica, Greece

Although scrape-off layer (SOL) physics in divertor tokamaks is relatively well understood, the

situation is worse for in/out asymmetries in the SOL, divertor plasma parameters and the plasma

flows. For example low field side profile and transport analysis reveales fast radial transport to

the outer wall, while an investigation of low-to-high field side hydrogen and carbon fluxes indi-

cate dominant high field side recycling. Global SOL plasma flows, driven e.g. by drift effects,

poloidal pressure asymmetries or plasma rotation, migth reconcile these findings.

In this paper we focus on measurements in ASDEX Upgrade which were performed by

Langmuir probes on a reciprocating manipulator close to the outer midplane in connection

with probes in the divertor. Two different probe arrangements were used with the manipulator

probes; simple pin probes and in-plane probes. The in-plane mounted probes allow for higher

heat fluxes onto the probe and reduce the uncertainty in the probe area for the flow measure-

ments. In addition to well established lower divertor diagnostics and Langmuir probes at the

inner heat shield the Langmuir probe arrays in the upper divertor have been refurbished and re-

arranged, allowing now for detailed measurements there during vertical shift experiments and in

upper single null configurations (including shot to shot reversal of the toroidal magnetic field).

We carried out experiments in ohmic-/L-mode and in H-mode. At low densities in ohmic

plasmas with lower single null (LSN) configuration the upward midplane flow at the magnetic

low field side (LFS) reached velocities ofM ≈ 0.7, showing a maximum about 1−2cm outside

the separatrix. Increasing the plasma density caused a reduction of the Mach number, but the

flow in the outer midplane is still directed towards the inner divertor. First measurements in H-

mode discharges indicate that the Mach numbers in H- and L-mode are similar. In general the

flow stagnation point at the LFS is located below the midplane in LSN discharges. Therefore

the flow with parallel sound velocities near 105ms−1 offers a high potential for mass transport

in the SOL from the outer midplane to the inner divertor in lower single null configurations.

Our measurements will be discussed in relation to the results of other tokamaks. Further

measurements are still in progress at ASDEX Upgrade (e.g. upper single null discharges are

planned) and more details will be reported and discussed at the conference.

P-1.009, Monday June 27, 2005

Page 15: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Tungsten Erosion at Auxiliary Limiters in ASDEX Upgrade

R. Dux1, V. Bobkov1, A. Herrmann1, K. Krieger1, R. Neu1, T. Pütterich1, V. Petrzilka2,

V. Rohde1, ASDEX Upgrade Team1

1 Max-Planck-Institut für Plasmaphysik, EURATOM Association, Garching, Germany

2 Association EURATOM/IPP.CR, Prague, Czech Republic

In order to test the reactor compatibility of high-Z plasma facing components (PFC), a step-

by-step increase of tungsten coated PFCs towards a full tungsten machine is pursued at ASDEX

Upgrade. At present, almost 70% of the total PFC area consists of W-coated graphite tiles.

The enhancements for the 2005 campaign focused on the auxiliary limiters, which receive the

highest load of the main chamber components. ASDEX Upgrade has 12 poloidal limiters on the

low field side: a pair of side limiters for each of the 4 ICRH antennas and a pair of guard limiters

at each side of the 2 neutral beam ducts, which are between the two ICRH antenna doublets. The

toroidal width of the guard limiters was increased from 10 to 20 cm and the radial distance to

the plasma was changed from 12 mm to 6 mm behind the position of the ICRH limiters to allow

for a better power load sharing between the limiters. One of the ICRH limiters and one guard

limiter is equipped with W-coated tiles. The tungsten influx from the limiters was measured on

9 lines-of-sight using a WI spectral line at 400,8 nm.

During the 2004 campaign, measurements of W influx from a guard limiter pointed towards

a dominant fast D

particle contribution to the average deposited energy per deuterium ion

and the sputtering of tungsten. For ICR heating on the close-by antenna doublet, which has a

minimum toroidal distance of 0.8 m to the guard limiter, the W influx per heating power was

observed to increase by a factor of 1.5 compared to pure NBI injection. Post mortem analyses

of the coated tiles by x-ray fluorescence and Rutherford backscattering confirm a net erosion of

several hundreds of nm.

During the present campaign the tungsten influx from the tungsten ICRH side limiter was

measured in dedicated H-mode plasmas identical to the ones performed during 2004. An in-

crease by a factor of 100 was found compared to the W influx density at the old guard limiter,

reflecting the fact that the ICRH limiters are the components closest to the low field side plasma.

Also the ICR induced local W influx from the antenna side limiter has strongly increased, where

the above mentioned enhancement factor compared to NBI heating can be on the order of 20.

Details of the influence of ICRH as well as of the relation of W sputtering by thermal impu-

rity ions and by fast D

from NBI will be investigated during the ongoing campaign and will

presented at the conference.

P-1.010, Monday June 27, 2005

Page 16: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Carbon migration at the divertor of ASDEX UpgradeV. Rohde1, M. Mayer1, J. Likonen2 and ASDEX Upgrade Team1 Max-Plan k-Institut fur Plasmaphysik, EURATOM Asso iation, Gar hing , Germany2 VTT Pro esses, Asso iation EURATOM/TEKES, Esspoo, FIN-02044 VTT, FinlandIn present fusion experiments Carbon is the most ommon rst wall material. Graphiteoers ex ellent thermo me hani al and ele tri al properties. Large type I ELM's expe tedat the ITER divertor require the use of arbon based materials. But Carbon is stronglyeroded, whi h lead to the formation of deposition layers. These a:C-H type layers will ontain a signi ant amount of tritium. As for safety reasons the tritium inventory isrestri ted, the formation of layers at remote areas has to be ontained. To understandthe deposition pro esses laboratory experiments on a:C-H layer growth are not suÆ ient,be ause only experiments in fusion devi es mat h all relevant pro esses at the same time.During the last experimental ampaigns a ombined experiment was realised to inves-tigate the arbon layer formation at the omplete divertor region of ASDEX Upgrade.The deposition and erosion on the target plates had been measured by Re/C markerstripes. Strong deposition is found at the inner divertor target- and bae tiles. At theouter divertor erosion is observed at the bae region, whereas at the strike point moduledeposition and erosion is observed at the same lo ation. Up to 35 Si wafer and 5 avityprobes were mounted as deposition monitors at remote areas. The markers over almostall relevant regions providing high spatial but no temporal resolution. The layers areformed by high sti king spe ies, whi h are identied by the de ay length of the depo-sition thi kness and avity probes. Monitors at the same position, but with dierentorientation with respe t to the magneti eld show strong dieren es of the depositionlayer thi knesses.To investigate the layer formation pro esses at remote areas time resolved measure-ments using quartz mi robalan e monitors, Langmuir probes and residual gas analysisare used. Whereas the inner divertor shows almost onstant layer growth, the pi tureis mu h more ompli ated at the outer divertor. The layer growth varies with the neu-tral density and the strike point position. Low density plasmas with high input powereven ause erosion on the remotely deposited layers. Additionally Langmuir probes atthe target plates and below the divertor omplement the data. A parasiti plasma isobserved below the divertor stru ture, whi h in uen es the layer by surfa e a tivationor layer erosion.

P-1.011, Monday June 27, 2005

Page 17: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Role of recycling in W7-AS divertor plasmas

Y. Fenga, F. Sardeia, P. Grigulla, J. Kisslingerb, K. McCormickb, D.Reiterc

a Max-Planck-Institut für Plasmaphysik, Teilinstitut Greifswald, Euratom Association, Wendelsteinstr. 1, D-17491 Greifswald, Germany b Max-Planck-Institut für Plasmaphysik, Euratom Association, D-85748 Garching, Germany c Institut für Plasmaphysik, Forschungszentrum Jülich Gmbh, Euratom Association, Trilateral Euregio Cluster, D-52425 Jülich, Germany

The island divertor in W7-AS, with respect to previous limiters, made the plasma density

easily controllable even in the presence of a strong NBI-source, showing a significant

improvement of the recycling conditions. Impurity radiation could be kept within the

island SOL to enable a stable partial detachment without remarkable loss of the global

energy content. The divertor operation gave a discovery of a new HDH-regime

characterized by high density and good energy and low impurity confinement. Based on

EMC3/EIRENE simulations, experimental results and simple models, the paper presents

a detailed analysis of neutral transport behavior under different recycling conditions,

aimed at identifying the role of the recycling neutrals in establishing improved

confinement regimes. Special attention will be paid to the HDH-transition and the

bifurcation behavior associated with the strongly non-linear recycling process. Discussion

will be made on the jump of the edge density during the HDH-transition, differences and

similarities in recycling conditions between H* and HDH-modes and the role of the

recycling process in ne,down roll-over effect and detachment transition and stability.

Email: [email protected]

P-1.012, Monday June 27, 2005

Page 18: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Modelling of tritium retention and target lifetime of the ITER divertor

A. Kirschner1, S. Droste

1, D. Borodin

1, V. Philipps

1, G. Federici

2, J. Roth

3

1Institut für Plasmaphysik, Forschungszentrum Jülich GmbH , EURATOM Association,

Trilateral Euergio Cluster, 52425 Jülich, Germany

2ITER JWS Garching Co Center, Boltzmannstr. 2, 85748 Garching, Germany

3Max-Planck Institut für Plasmaphysik, EURATOM Association, Boltzmannstr. 2,

85748 Garching, Germany

The current ITER design comprises three different wall materials: beryllium for the first

wall, tungsten for the baffles and dome structure and carbon fibre composites (CFC) for

the divertor plates [1]. With respect to carbon materials the big advantage of not melting

fronts the addiction to chemical erosion, which leads to tritium retention and reduced

lifetime which will largely determine the number of possible discharges in ITER. Most

of the current experimental results on carbon erosion, transport and tritium retention

origin from full carbon machines wherefore extrapolation to ITER is difficult requiring

predictive modelling for the conditions in ITER.

This contribution presents recent ERO modelling of carbon erosion, transport and re-

deposition in the divertor of ITER. The background plasma is provided by B2-Eirene

calculations. Chemical erosion yields are taken from the recent formula taking into

account the dependence on flux, incident energy and surface temperature. [2]. These

assumptions lead to a reduction of the chemical gross erosion by about one order of

magnitude compared to a fixed yield of 1%. The total amount of carbon species that

escape towards the dome region is reduced by a factor of about 3. It is shown that this

quantity depends largely on the assumption of the re-erosion yield of re-deposited

carbon layers [3] which is assumed to be enhanced compared to that of substrate

material. Also the effect of beryllium and tungsten deposition on the carbon erosion will

be addressed.

Parameter studies of the effect of strike point sweeping along the divertor plates on the

carbon transport will be presented. The results indicate an increase of target lifetime by

a factor of about 1.5.

[1] ITER Technical Basis, ITER EDA Documentation Series No. 24, IAEA, Vienna 2002

[2] J. Roth et al., J. Nucl. Mat. 337-339 (2005), 970

[3] A. Kirschner et al., J. Nucl. Mat. 328 (2004), 62

P-1.013, Monday June 27, 2005

Page 19: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

!

A. Kreter, D. Borodin, S. Brezinsek, S. Droste, T. Hirai1, A. Kirschner, A. Litnovsky, M. Mayer2, Y. Sakawa3, U. Samm, O. Schmitz, G. Sergienko, T. Tanabe3, V. Philipps,

A. Pospieszczyk, Y. Ueda4, P. Wienhold and TEXTOR team !"#$!

" %&'('& )*+ *

!"#$! &'('& $ ,- - !"#$! .&/(.

01 *23 43 (5(-.567 * 2# 4# &5&-6./8

13CH4 was injected through graphite and tungsten spherical limiters in reproducible TEXTOR

discharges. These materials were chosen, as they represent the actual compromise for the

plasma facing components in the divertor chamber of ITER. 13C was used to distinguish

puffed and intrinsic carbon in the layer deposited on the limiter surface.

Shot-by-shot video recordings show a continuous growth of the deposit in the vicinity of the

puffing hole. A pronounced difference in the 13C deposition pattern on the limiters was

observed. Several post mortem analysis techniques were applied to characterize the deposited

film. Nuclear reaction analysis and secondary ion mass spectrometry showed, that the ratios

of the locally deposited to the puffed carbon amount are 4 % for graphite and only 0.3 % for

tungsten. The maximum of the deposit thickness for both limiters is situated near the puffing

hole. In particular, the distributions of 13C and D are strongly peaked at the hole, whereas 12C

is more uniformly distributed. The maximum thickness is only about a factor of 2 larger for

the graphite (2.1 the deposition efficiency is mainly due to the difference in the deposition area on both

limiters. The ratio of 13C to total C varies from 90 % in the vicinity of the puffing hole to 30-

40 % at the deposit edge for both experiments. The D to C ratios are in the range of 10-20 %

for graphite and 5-15 % for tungsten.

The lower 13C deposition efficiency for tungsten can be explained by the effects of kinetic

reflection: the carbon reflection coefficient is <0.01 for carbon surface and ~0.4 for tungsten.

Another possible effect is the enhancement of physical re-erosion by background plasma

atoms for carbon deposited on tungsten surfaces. This effect is based on the lower energy

losses of the plasma particles reflected from the underlying tungsten atoms due to the large

mass difference, leading to a more effective sputtering of C on the top of the tungsten

surface. Results of simulations of these effects with the ERO code will be presented.

P-1.014, Monday June 27, 2005

Page 20: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Carbon deposition and fuel accumulation in castellated limiters exposed

in the SOL of TEXTOR

A. Litnovsky1, V. Philipps

1, P. Wienhold

1, G. Sergienko

1, O. Schmitz

1, A. Kreter

1,

P.Karduck2, M. Blöme

2, B. Emmoth

3, M. Rubel

4

1Institut für Plasmaphysik, Association EURATOM, TEC, FZ- Jülich, Germany 2Herzogenrather Dienstleistungszentrum GbR, Herzogenrath, Germany

3Department of Microelectronics, KTH, Association EURATOM – VR, Kista, Sweden 4Alfvén Laboratory, KTH, Association EURATOM – VR, Stockholm, Sweden

Castellated structures are proposed for the divertor and the first wall of ITER to ensure

the thermo-mechanical durability of the machine [1]. A concern with such a structure is the

possible accumulation of fuel in the gaps. Dedicated investigations of the fuel inventory in

castellated structures are underway on several tokamaks. In TEXTOR metallic limiters with

ITER-like castellation were exposed in the SOL. In a first experiment, the limiter was exposed

in a deposition-dominated area and carbon deposits were found both on the top plasma facing

surfaces and in the gaps. The fuel accumulation in the gaps was estimated to be at least 30% of

the overall fuel retention on this limiter [2].

Recently, a second castellated limiter was exposed in the erosion-dominated area of the

SOL of TEXTOR. An average plasma fluence accumulated by the limiter was 8.5×1019

D/cm2,

which is approximately 4 times higher than in the previous experiment. After the exposure

deposits were found on thin stripe-like zones of the gaps close to the plasma facing side,

similarly with the previous experiment. However, on the stripes of the gaps with direct view to

the plasma flux, narrow shiny erosion zones were observed. Several surface diagnostics were

applied to assess the elemental composition of the deposits and depth distribution of their

constituents. No deposits were detected neither on the plasma facing top surfaces nor on the

plasma open sides of the gaps. On the plasma shadowed areas of the gaps deposits with the

maximal thickness up to 500 nm have been observed, consisted from carbon films enriched

with hydrogen, deuterium, boron and oxygen. The data demonstrate that carbon and fuel is

retained in the gaps of a metallic limiter which is in turn, in the erosion-dominated area. A

significant amount of Mo from the limiter was found intermixed into the deposit layer. The

present contribution provides the detailed analysis of deposits and fuel retention in gaps.

[1] W. Daenner et al., Fusion Eng. Des. 61-62 (2002) 61;

[2] A. Litnovsky et al, J. Nucl. Mater (2005), in press.

P-1.015, Monday June 27, 2005

Page 21: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Impact of the DED on ion transport and poloidal rotation in

TEXTOR

C. Busch1 , K.-H. Finken1, S. Jachmich2, M. Jakubowski1, A. Kramer-Flecken1, M.

Lehnen1, U. Samm1, O. Schmitz1, B. Unterberg1

1Institut fur Plasmaphysik, Forschungszentrum Julich GmbH, EURATOM-Association,Trilateral Euregio Cluster, D-52425 Julich, Germany

2Laboratoire de Physique des Plasmas / Laboratorium voor Plasmafysica, ERM / KMS,

EURATOM Association, B-1000 Brussels, Belgium

The recently installed Dynamic Ergodic Divertor (DED) in the TEXTOR tokamak al-

lows for a distinct ergodisation of the plasma edge. In this contribution the influence of

the DED on poloidal rotation and transport of carbon impurities are presented. The un-

derlying data stems from passive line emission and active charge exchange spectroscopy

with the diagnostic hydrogen beam. The edge observation system (r/a=1.0-0.5) consists

of 20 radial glass fibre channels. The light is transferred to a high resolution spectrometer

(littrow geometry, n=46) and recorded by a 1024x1024 pixel CCD camera at a dispersion

of 0.7 A/mm. In addition there are reference channels looking from the opposite direc-

tion which are projected onto the same detector. This enables a differential measurement

of the poloidal rotation with a theoretical resolution of approximately 0.5 km/s. The

hydrogen beam is pulsed at 10 Hz to separate the passive background signal, however

the time resolution so far is 1 s due to the limited level of the active signal.

The DED 3/1 configuration is characterized by a deep penetration of the perturbing field

up to the q=2 surface. Here the analysis is based on passive CIII emission originating

from a thin radial emission shell just inside the last closed flux surface to be evaluated.

A 2/1 and then a 3/1 tearing mode develop successively with rising perturbation field,

which in turn has influence on the degree of ergodisation. Under these conditions a re-

versal of the rotation with increasing ergodisation at this edge position has been found:

The initial rotation in the unperturbed case compares with earlier measurements where

the rotation had been found to be dominated by the ExB drift (Er negative and pointing

inward). Therefore, we conclude a reversal of the radial electric field with DED. This

has indeed been confirmed by independent probe measurements [1].

For the DED 12/4 configuration, which is characterized by a much more shallow pene-

tration of the perturbation, active CVI spectra and thus radial profiles are at hand which

supplement the CIII data. Similar to the afore mentioned results the rotation changes

compatible with an increase of the radial electric field. However, the ion temperature

profiles have not yet shown to be affected by the DED in the 12/4 configuration.

[1] S. Jachmich et al, this conference

P-1.016, Monday June 27, 2005

Page 22: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Modelling of hydrocarbon transport and emission after methane injection

into the TEXTOR boundary plasma using the ERO code

D.Borodin, A.Kirschner, S.Brezinsek, V.Philipps, A.Pospieszczyk, S.Droste, G.Sergienko

Institut für Plasmaphysik, Forschungszentrum Jülich GmbH, EURATOM-Assoziation,

Trilateral Euregio Cluster, 52425 Jülich, Germany

The erosion of carbon-based materials, which are foreseen for divertors of fusion

devices including ITER, determines their lifetime, and, even more important, the co-

deposition of radioactive tritium which has to be minimised. Therefore, the general

understanding of the carbon erosion-deposition patterns along with the underlying process

database and reliable diagnostics are of large importance. The aim of the present work is to

model the transport and recycling of hydrocarbons injected into the boundary plasma of

TEXTOR and to obtain the so-called D/XB values (inverse photo efficiencies), which are

necessary to measure the fluxes of the molecular species by spectroscopy. For this purpose

the three-dimensional Monte-Carlo code ERO is used.

The modelling is done for experiments at TEXTOR, in which a known amount of

CD4 molecules was injected near the LCFS through cylindrically shaped gas inlets with

different sizes of the surrounding surface. The spectroscopy is based on the observation of

the CD A-X band emission. The D/XB values can be defined by the coefficient between the

intensity of this emission and the number of CD4 molecules injected.

For these studies a number of improvements were made in the ERO code: additional

limiter geometry (and corresponding visualization routines), a new set of molecular data for

the CH4 reaction chain [1] and enhanced re-erosion of deposited carbon. Carbon deposited

from the background plasma, one from the injection and re-deposited is treated separately.

The influence of different parameters was studied: plasma density and temperature, effective

sticking probabilities and (re-)erosion rate, surface size. The modelling results suggest that

the recycling on the gas inlet surface leads to a reduction of the D/XB values, mostly due to

chemical re-erosion of the carbon deposited from the injection.

The same parameters have been varied also in the experiment: plasma parameters by

changing the radial limiter position, influence of recycling by varying the inlet surface size.

A detailed comparison between modelling and experiment is presented.

[1] R.Janev and D.Reiter, Jülich report, Jül-3966, 2002

P-1.017, Monday June 27, 2005

Page 23: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

High temperature erosion of tungsten exposed to the

TEXTOR edge plasma

G. Sergienko1, A. Huber

1, A. Kreter

1, V. Philipps

1, A. Pospieszczyk

1,

M. Rubel2, B. Schweer

1, O. Schmitz

1

1 Institut für Plasmaphysik, Forschungszentrum Jülich GmbH, EURATOM­Association,

Trilateral Euregio Cluster, 52425 Jülich, Germany 2

Alfvén Laboratory, Royal Institute of Technology, Association EURATOM-VR,

100 44 Stockholm, Sweden

Tungsten is foreseen at the full divertor for JET, the upper divertor regions in ITER and a

most promising candidate to replace CFC carbon also on the high heat flux lower divertor

areas. The arguments for this are the low sputtering coefficient and high melting point of

tungsten. The main concern with the use of tungsten is the ability to melt and the associated

melt layer loss. Another open question is the possibility of enhanced erosion of tungsten at

high temperatures, as reported for tungsten exposed to a high current argon plasma arc [1].

Thus, the investigation of tungsten erosion under extreme conditions in the edge plasma of

tokamaks is important to qualify the operational limits for this material.

To do this, the erosion characteristics of tungsten have been investigated at temperatures

extending up to the melting point. A solid tungsten plate (75 x 63 mm2) with a thickness of

2 mm was heated up until melting by the plasma load in TEXTOR. The plate was fixed on

a graphite roof limiter with an angle of 20° to the magnetic field lines. The surface

temperature of the tungsten plate was measured by a single colour pyrometer at the position

of maximum heat flux. The 2D temperature distribution was recorded by a video camera

equipped with an infrared cut-off filter. The released flux of tungsten atoms from the plate

was measured spectroscopically in the near UV spectral region to reduce the influence of

the thermal radiation continuum from the hot surface. It was found that the atomic tungsten

flux from the plate was nearly constant up to a temperature of about 3200 K. With further

temperature rise, the flux grows exponentially with an increase by a factor of 7 at 3600 K

(close to melting point of tungsten). The impact of this behaviour on the performance of

high-Z wall will be discussed.

[1] E.P. Vaulin et al. , Sov. J. Plasma Phys. Vol.7, No 2 (1981) 239 - 242

P-1.018, Monday June 27, 2005

Page 24: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Tungsten melting under high power load

in the TEXTOR edge plasma

G. Sergienko1, A. Huber

1, A. Kreter

1, V. Philipps

1, M. Rubel

2,

B. Schweer1, O. Schmitz

1, M. Tokar

1

1 Institut für Plasmaphysik, Forschungszentrum Jülich GmbH, EURATOM­Association,

Trilateral Euregio Cluster, 52425 Jülich, Germany 2

Alfvén Laboratory, Royal Institute of Technology, Association EURATOM-VR,

100 44 Stockholm, Sweden

Tungsten is a strong candidate for plasma facing components (PFC) in the ITER divertor.

The most critical question with the use of tungsten on the high heat flux areas is the ability

to melt which may occur during high transient heat spikes from ELMs or disruptions.

Possible melt layer loss or melt layer motion can strongly enhance the local erosion of the

divertor tile and may also produce irregular surfaces which may be subject to hot spots in

following discharges. Investigations of the behaviour of the melt layer in a strong magnetic

field under high heat plasma flux are most important to assess the use of tungsten on those

areas. In the TEXTOR edge plasma, a thin solid tungsten plate was heated up by plasma

impact until melting. The plate was thermally isolated from the holder and fixed on a

graphite ruff limiter, which was inserted into the plasma from the top of the torus.

The tungsten was observed to melt poloidally along the plate edge at the region of the

maximum heat flux. The liquid tungsten moved fast perpendicular to the magnetic field

lines along the plate surface in the jxB direction whereby the current direction corresponds

to the thermo-electron current emitted from the hot tungsten surface. A large blob of liquid

tungsten was collected at the plate edge due to surface tension forces, which then moved up

to wards the scrape-off layer plasma along the edge of the underlying graphite limiter. The

motion of liquid tungsten produced a deep furrow all along the plate surface.

Ex-situ surface characterisation of the tungsten plate has been performed with a number of

techniques. The essence of surface studies is following: no indication of blistering effects

are found on the plate, neither in the melt-layer zone nor outside it; island-type inclusions

of carbon particles are detected in the eroded region; re-crystallisation of molten material

has lead to the formation of large grains and distinct grain boundaries; small dust-like

spherical granules of tungsten are observed. The latter indicates a possibility of the

formation of tungsten dust from the molten mass in a magnetic controlled fusion device.

P-1.019, Monday June 27, 2005

Page 25: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Screening and radiation efficiency of carbon with Dynamic Ergodic Divertor on TEXTOR

G.Telesca 1, G.Verdoolaege 1, K.Crombe1, M. Lehnen2, A. Pospieszczyk2 ,B.Unterberg 2,

G. Van Oost 1

1Department of Applied Physics, Ghent University, Rozier 44, B-9000 Gent, Belgium 2 Institut fuer Plasmaphysik, Forschungszentrum Juelich GmbH, EURATOM Association,

D-52425 Juelich, Germany This paper deals with a study made at TEXTOR to evaluate the change of screening and

radiation properties of carbon under the action of the Dynamic Ergodic Divertor operated

in m = 12, n = 4 mode. The results refer to the maximal nominal current on the DED

coils, which has been recently reached in the m/n =12/4 configuration. The diagnostic

tools used in this study are: spectroscopic measurements in the UV for the intensity of

carbon lines, bremsstrahlung in the visible for the determination of Zeff, and bolometric

signals for the total radiated power, Prad. CIII and CV lines are detected simultaneously

along nine lines of sight pointing partly (5 chords) at the graphite divertor tiles and partly

(4 chords) far from any carbon source, so that both local and global effects can be

determined. Their ratio can provide information on the level of carbon screening and on

the radiation properties of carbon. Prad and Zeff, and their normalized ratio, are also used

to characterize transport and radiation of impurities during DED operation. We report on

the behavior of carbon released from the divertor tiles (intrinsic carbon) and also of that

injected as an extrinsic impurity (methane) from sources located at different positions.

Preliminary analysis of the data indicates no significant enhancement in screening and in

cooling efficiency for carbon released (or puffed in) from the divertor region. However, a

clear beneficial effect in carbon transport is seen at relatively high values of the safety

factor (q(a) > 3.2) when methane is injected from a valve located far from the divertor

plate. The correlation of the experimental evidence with the structure of the perturbed

magnetic field is discussed.

P-1.020, Monday June 27, 2005

Page 26: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

On the influence of the magnetic resonances on the heat flux structure of

the Dynamic Ergodic Divertor

M.W. Jakubowski1, S.S. Abdullaev1, K.H. Finken1, M. Lehnen1, U. Samm1,

O. Schmitz1, K.H. Spatschek2, B. Unterberg1, R. Wolf1

1 Institut für Plasmaphysik, Forschungszentrum Jülich GmbH, EURATOM Association,

Trilateral Euregio Cluster, D-52425 Jülich, Germany2 Institut für Theoretische Physik I, Heinrich-Heine-Universität Düsseldorf, D-40225

Düsseldorf, Germany

The Dynamic Ergodic Divertor (DED) in TEXTOR is designed to provide an ergodized

volume in the plasma edge in order to control heat and particle exhaust. The DED coil currents

create magnetic islands; if these islands overlap, the magnetic field lines become ergodic. The

near field of the DED deflects the magnetic field lines such that they intersect the walls. The

region of short connection lengths is called the laminar zone. The structure of the perturbed

volume strongly depends on the safety factor profile and the plasma pressure. At the higher

level of ergodization (i.e. at higher plasma current and lower beta poloidal) the laminar zone is

dominant, while at the lower level of ergodization the ergodic region dominates. The heat and

particles are deposited on the divertor target plate forming a stripe-like pattern. The features of

the ergodized volume produced by the DED has been already discussed (e.g. in [1]).

The temperature distribution over the divertor target plates is measured by a infrared camera.

The incoming heat flux is evaluated from the temperature evolution. To investigate the influence

of theq-profile on the heat flux pattern few series of the discharges were performed, where the

plasma current was ramped in order to vary the edge safety factor (qedge≃ 4⇒ qedge

≃ 2.3).

It is found that the structure of the strike zones is strongly correlated to the value of the edge

safety factor. The general tendency is that the strike zone splits, ifq . 3.25. However, one can

identify substructures, which can be attributed to a certain range of the edge safety factor, i.e.

they appear atqedge1 and disappear atqedge

2 with ∆q/q ∼ 0.07. The topological considerations

performed with the Atlas code allows to identify flux tubes consisting of field lines with short

connection lengths, which appear at a given value of the edge safety factor. Probably these flux

tubes are responsible for the substructures in the heat flux pattern.

References

[1] B. Unterberg, et al., Proceedings of 20th IAEA Fusion Energy Conference, Portugal

(2004) EX/P5-33

P-1.021, Monday June 27, 2005

Page 27: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Impact of the Dynamic Ergodic Divertor on

the Structure of the Plasma Edge at TEXTOR

O. Schmitz, S. Abdullaev, S. Brezinsek, C. Busch, K. H. Finken, M. Jakubowski,

M. Lehnen, A. Pospieszczyk, U. Samm, B. Schweer, G. Sergienko, B. Unterberg

Institut fur Plasmaphysik, Forschungszentrum Julich GmbH, EURATOM-Association,

Trilateral Euregio Cluster, D-52425 Julich, Germany

The Dynamic Ergodic Divertor (DED) has been installed at TEXTOR to investigate the

potential of an ergodised plasma edge region to control the particle and power exhaust

from magnetically confined fusion plasmas. The calculated magnetic topology [1] in the

DED perturbed volume around the q=3 surface consists of three regions. A region with

isolated island chains, a region where these islands overlap - called ergodic region - and

a laminar region, where the field lines are deflected towards the DED target tiles with

short connection lengths. In the m/n = 12/4-mode configuration the pattern of the lam-

inar field lines intersecting the DED target tiles has a characteristic four-fold structure

of 4x2 strike zones and private flux regions in between.

In this contribution the electron density ne and temperature Te measured in the plasma

edge by means of Beam Emission Spectroscopy on thermal He- and Li-Beams will be

compared with the calculated magnetic topology. The particle fluxes on the DED tiles

were observed supplementary with CCD cameras equipped with CIII and Hα

filters.

With increasing DED current a decrease of ne, T

eand of the electron pressure p

eis

detected in the plasma edge at the Low Field Side (LFS) accompanied by a splitting of

the particle and heat flux patterns on the DED target. Comparisons of the calculated

profiles of the connection lengths and the ne, T

eand p

eprofiles conjoin this observation

with the formation of a helical divertor: a flux tube of field lines with short connection

lengths is created in the region in front of the observation volume at the LFS causing a

rapid flow of particles to the DED target. The ne, T

eand p

eprofiles move inward into

this DED Scrape-Off Layer. This confirms the calculated radial extension of this laminar

flux tube of 2-5 cm for 2.8 < qa

< 3.4 into the region of previously closed magnetic flux

surfaces. The poloidal extent of this structure was investigated by sweeping the DED

field. This leads to an alternating appearance of laminar regions and parts of ergodic

regions in front of the atomic beams with the mentioned decrease of ne, T

eand p

ein the

laminar regions.

The experimental findings presented in this contribution prove the formation of an open

helical divertor as a consequence of the DED perturbation. The close agreement with

the prescribed magnetic topology allows to optimize the properties of the DED divertor

by adjusting the position of the resonant surfaces to the DED coils.

[1] M.W. Jakubowski, S.S. Abdullaev and K.H. Finken, Nuclear Fusion 44, (2004) S1-11

P-1.022, Monday June 27, 2005

Page 28: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Structure of stochastic field lines near the separatrix in poloidal divertor

tokamaks

S.S. Abdullaev, K.H. Finken, M. Jakubowski, M. Lehnen, R. Wolf

Institut für Plasmaphysik, Forschungszentrum Jülich GmbH, EURATOM Association,

Trilateral Euregio Cluster, D–52425 Jülich, Germany

The structure of stochastic magnetic field lines at the plasma edge mainly determines the spatial

structure of plasma boundary and deposition patterns of heat and particles on divertor plates [1].

Moreover, recent experimental studies in the DIII-D tokamak show that a stochastic magnetic

boundary created by edge magnetic perturbation suppresses most large edge–localized modes

(ELMs) in high confinement (H-mode) plasmas [2]. In this presentation we demonstrate meth-

ods and tools to study the fine structure of stochastic magnetic field lines at the plasma edge

formed due to effect of external magnetic perturbation in poloidal divertor tokamaks. For this

purpose we have chosen a simplified model of plasma: the equilibrium plasma is modeled by

three–current loops. The external magnetic perturbations are created by N pair of loop coils with

opposite flowing currents. This model is studied using the Hamiltonian formulation of magnetic

field line equations. Two symplectic mapping methods are applied to study magnetic field lines

near the separatrix. The first mapping method is based on the Hamiltonian formulation of field

lines equations in Clebsch coordinates and it has been previously applied to study field line’s

structure in the TEXTOR-DED [3]. The second approach is the method of the canonical separa-

trix mapping. It maps the poloidal flux and toroidal angle to the plane perpendicularly crossing

the poloidal section along the X-line. The structure of stochastic layer is studied not only by

Poincaré sections of field lines but also plotting so-called laminar plots. The latter are contour

plots of open stochastic field lines near the separatrix and on the divertor plates with different

of wall to wall connection lengths. They reveal a fine structure of field lines which cannot be

studied by Poincaré sections.

References

[1] K.H. Finken, S.S. Abdullaev, W. Biel, et.al. Plasmas Phys. Contr. Fusion 46, B 143 (2004).

[2] T. Evans et.al, Phys. Rev. Lett., 92, 253003 (2004).

[3] S.S. Abdullaev, K.H. Finken, K.H. Spatschek, Phys. Plasmas, 6, 153 (1999).

P-1.023, Monday June 27, 2005

Page 29: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Removal of carbon layers by oxygen treatment of TEXTOR

V.Philipps1, G.Sergienko

1, A.Lyssoivan

2, H.G.Esser

1, M. Freisinger

1, H. Reimer

1

1 Institut für Plasmaphysik, Forschungszentrum Jülich GmbH, EURATOM Association,

D-52425 Jülich, Germany 2 Laboratoire de Physique des Plasmas / Laboratorium voor Plasmafysica, ERM / KMS,

EURATOM Association, B-1000 Brussels, Belgium

Long-term tritium retention in re-deposited carbon layers is the major drawback for the use

of graphite as plasma facing material and a major safety concern for ITER. Work is needed

in this area to understand more clearly the basic processes that lead to tritium co-deposition

with carbon and to predict more precisely this behaviour in ITER, as well as to develope in-

situ techniques for tritium removal that are applicable in ITER. Apart from ventilation with

oxygen, plasma-assisted techniques can use oxygen for cleaning at lower wall temperatures,

but with the drawback that hidden areas are difficult to reach.

In order to proof the feasibility of different oxygen cleaning techniques, ICRF discharge

conditioning/cleaning (ICRF-DC) has been tested for the first time and compared with the

standard glow discharge (GDC). The ICRF plasma discharges have been performed with the

toroidal magnetic field on, BT=1.8/2.4 T, in different oxygen/helium mixtures,

O18

/(O18

+He4)…0.13/1.0, in the pressure range ptot=(1.5/3.3)·10

-4 mbar. Reproducible ICRF

plasma generation could be demonstrated with a coupled RF power/pulse about 50/60 kW

(f=32.5 MHz, vRF…1.0 s). About 16 ICRF-DC were done with a total duration of ~32 s.

Residual mass and optical spectroscopy show a rapid consumption of the oxygen molecules

filling the TEXTOR chamber, with a half time of about 24 ms. This is accompanied by the

appearance of hydrogen that is obviously released from the walls by the plasma impact.

However, due to the fast oxygen consumption and the lack of oxygen feedback in this first

experiment, the averaged carbon removal rate remained small while the removal rate

extrapolated from the start of the ICRF plasma (where the O2 pressure was large) is high.

Recovery to normal plasma operation was possible without additional conditioning albeit

with an oxygen content initially about a factor of 3 higher then before, but decreased rapidly

with further plasma discharges.

In the second experiment, GDC was applied in TEXTOR in different mixtures of He/O2 for

about 3 hours. Mass spectroscopy shows the total conversion of the injected O2 in CO and

CO2 with little influence of the He/O2 mixture on the production rate. Probes made from SS,

Si, a-C:H layers and a-B:H layers on Si were inserted at two different locations at 200°C

during the GDC and analysed before and after the oxygen treatment. Fringe analysis show

the complete removal of the a-C:H layers. Analysis of oxygen incorporated in the various

probe is underway and will be shown. Plasma recovery after the GDC treatment was more

difficult in this case and finally achieved only by an additional boronisation. The present

analysis indicates enhanced MHD activity in the start up phase of the plasma while the

radiation level was increased by about a factor of two and thus similarly enhanced as

observed after the previous oxygen treatments of TEXTOR.

P-1.024, Monday June 27, 2005

Page 30: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

IMPROVED MODELLING OF NEUTRALS AND

CONSEQUENCES FOR THE DIVERTOR PERFORMANCE IN ITERA.S. Kukushkin1, H.D. Pacher2, V. Kotov3, D. Reiter3, D. Coster4, G.W. Pacher5

1 ITER International Team, Garching Joint Work Site, Garching, Germany;2 INRS-EMT, Varennes, Québec, Canada; 3 FZ Jülich, Jülich, Germany;

4 Max-Planck IPP, Garching, Germany; 5 Hydro-Québec, Varennes, Québec, Canada

In B2-EIRENE modelling of ITER, the usual, linear Monte-Carlo modelling of neutral

transport is inadequate, since the large dimensions and high neutral density make the neutrals in

the PFR collisional, providing bulk particle scattering. Its neglect in the linear Monte-Carlo

method results in definite artefacts in the divertor plasma behaviour when the dome geometry is

modified. We have developed and implemented a non-linear Monte-Carlo model, taking into

account neutral-neutral and molecule-ion collisions, thereby enabling for the first time

meaningful comparison of various divertor geometries, including those without dome.

Recent improvements include the modelling of carbon neutral-neutral collisions. The

results show that, in comparison with previous modelling for which only the DT and He neutral

models had been updated, there are only modest changes to the plasma parameters of the SOL

and divertor plasma. However, the target erosion is reduced, indicating a strong influence of the

carbon neutral-neutral collisions on the net erosion and redeposition of carbon.

The importance of the transparency of the dome-supporting structures is being re-

examined with the more complete model taking neutral-neutral collisions into account. First

results indicate that the plasma parameters are significantly affected only if the transparency is

reduced by a factor 5 below the design value, i.e. the design is robust in this parameter.

Since the dome noticeably adds to the complexity and cost of the ITER divertor, we are

re-examining the consequences of its removal. Its functions are essentially compression of

neutrals in the PFR to alleviate helium exhaust, reduction of neutral influx to the core plasma

near the X-point, and neutron shielding at the bottom of the divertor (the latter is treated

elsewhere). Initial results (complete series will be shown in the paper) from B2-Eirene

modelling with the improved neutral model indicate that removal of the dome requires higher

pumping speed at the pump duct and leads to some reduction of the divertor power load,

accompanied by an increase of the separatrix density and a decrease of the separatrix

temperature. The consequences on divertor loading as well as on confinement and operating

window (pumping speed requirements, X-point MARFE conditions) will be discussed.

Transport of the Lyman-series radiation in the divertor plasma, which can be relevant for

the dome assessment, is also discussed in the paper. Large dimensions and high neutral density

make the ITER divertor opaque for this radiation, and this changes the balance between the

ionisation and recombination. The dome intercepts this radiation, thus modifying radiation

coupling between the inner and outer divertors.

P-1.025, Monday June 27, 2005

Page 31: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Simulation of brittle destruction of different types of graphite using

PEGASUS-3D code

O. V. Ogorodnikova1, S. Pestchanyi

2, J. Linke

1

1Forschungszentrum Juelich, EURATOM-Association, IWV-2, 52425 Juelich, Germany

2 Forschungszentrum Karlsruhe, EURATOM-Associaton, IHM, 76021 Karlsruhe, Germany

With regard to next generation fusion devices, it is important to investigate the mechanism

of particle emission from carbon based materials in order to reliably estimate the erosion

and the lifetime of the component. Numerical simulation of brittle destruction during high

heat flux load for fine grain graphite with different grain distributions and porosity and

highly ordered pyrolytic graphite HOPG has been performed using the 3-D

thermomechanics code ‘PEGASUS-3D’. The code is based on a crack generation induced

by thermal stresses. Due to breaking bonds between grains caused by thermal stress, solid

particles are easily emitted from a graphite sample during high thermal load. In the paper, it

is shown that brittle destruction is a result of anisotropy of thermal stress in different

directions. The particle erosion is much less for pyrolytic graphite compared with fine grain

graphite. The anisotropy due to the thermal gradient has less influence on the development

of cracks in graphite. A detailed investigation of the structure and grade of the material,

namely porosity, grain size, order and oriented anisotropy, on the particle erosion has been

done. Additionally, influence of surface and volumetric heating and sample pre-heating on

the erosion rate and degradation of different kinds of graphite is studied.

P-1.026, Monday June 27, 2005

Page 32: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Parametric investigation of temperature and stress evolution in actively

cooled plasma-facing components during high heat fluxes

O. V. Ogorodnikova, M. Roedig , J. Linke

Forschungszentrum Juelich, EURATOM-Association, 52425 Juelich, Germany

Materials in contact with plasma in fusion devices should be able to sustain extremely high

heat loads. Modeling of the temperature and stress distributions in plasma-facing

components is important under two points of view: (i) to find reasons influenced the heat

transfer degradation and (ii) to improve the material design and cooling conditions. In the

present work, the thermo-stress analysis for Be and CFC flat tile modules and W macro-

brush modules in the range of power loads between 0.1 and 10 MW/m2 has been done. An

influence of the thicknesses of plasma-facing material and heat sink material, cooling water

temperature and water speed and the effect of the interface geometry on the temperature

and stress distributions has been investigated. For Be and W the presence of the oxide

layers on the surface has also been studied. The influence imperfections of the joint

between the plasma-facing armour and heat sink on the heat removal efficiency has also

been simulated by the suggestion of areas with lower thermal conductivity. Using finite

element methods the temperature on the interface between plasma-facing armour and heat

sink material can be predicted from experimentally measured surface temperatures.

P-1.027, Monday June 27, 2005

Page 33: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

The Influence of Resonance Helical Field on the Ze f f in IR-T1 Tokamak

M. K. Salem1, M. M. Darian1,2, M. Ghoranneviss1, R. Arvin1, A. TalebiTaher1, A.Hojabri2

1Plasma Physics Research Center, Science and Research Campus, I. Azad University,

Tehran 14778, Iran2 Physics Group, I. Azad University, Karaj 31485-333, Iran

The effect of resonant helical field (RHF) on effective ion charge,Ze f f , in IR-T1 tokamak is

studied theoretically and experimentally.

The RHF in tokamak is an external magnetic field which can improve the plasma confine-

ment. This field is produced by conductors wound externally around the tokamak torus with a

given helicity.

The IR-T1 tokamak is a small air-core transformer tokamak with circular cross section and

without conducting shell and divertor. Its aspect ratio is R/a = 45 cm /12 cm. In IR-T1, RHF is

generated by two sets of helical coils installed outside the vacuum vessel. The pulsed dc RHF

configuration (l=2,3) has the optimal current and variable time.

To understand how RHF affects the IR-T1 plasma, the theoretical calculation for magnetic

field components produced by RHF is considered. The influence of RHF components on the

main external field, toroidal, is discussed. Then the results are applied to calculation ofZe f f

value through anomaly factor. Finally the theoretical and experimental results arise from RHF

are compared with our previous results, obtained without RHF.

P-1.028, Monday June 27, 2005

Page 34: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Oxygen impur ity profile studies in the EXTRAP T2R reversed field pinch

M. Kuldkepp 1, E. Rachlew

1, S. Menmuir

1, Y. Corre

2, P. R. Brunsell

3, M. Cecconello

3

1Dept. of Physics, KTH, EURATOM -VR Association, SE-10691 Stockholm, Sweden 2 Association EURATOM-CEA, DSM-DRFC, CEN Cadarache, F-13108 St Paul lez Durance, France 3Alfvén Laboratory, KTH, EURATOM -VR Association, SE-10044 Stockholm, Sweden

The medium sized reversed field pinch (RFP) EXTRAP T2R has an all metal first wall in

contrast to the more common graphite wall found in devices like RFX and the previous

EXTRAP T2. Recent comparisons of bolometric data [1] between EXTRAP T2R and RFX

have suggested very different radial profiles of impurity emission. Oxygen is the main

intrinsic plasma impurity in EXTRAP T2R and the VUV spectrometer data shows strong

emission from OV and OVI.

The oxygen emission profiles have recently been measured with a 5-channel UV-visible

spectrometer. From these results the impurity density profiles have been deduced by using

ADAS data. In addition, the impurity emission and density profiles have been computed with

an OSCR (Onion Skin Collisional Radiative) model constrained with the finite confinement

time of particles. The OSCR code has a small number of free parameters of which all but one

are partially set by measurements or external self-consistent codes. The new measurements

show broad non-hollow radial emission profiles of OV and OVI and confirm the earlier

bolometric measurements. Lower ionisation stages (OIV, OIII, OII) are found emitting close

to the edge but still over an appreciable part of the radius, also in agreement with earlier

results.

These results do not agree with measurements done using the same diagnostic on the

previous EXTRAP T2 [2]. This difference could be caused by a much larger penetration of

neutral particles as predicted by neutral density calculations.

The agreement between the OSCR modelling and the radially resolved spectral data gives

further evidence of the models potential in experiments such as EXTRAP T2R where plasma

measurements are done with a limited number of radial channels.

[1] Y. Corre, et al, Physica Scripta, 71, (2005)

[2] J.Sallander, Plasma Phys. Control. Fusion, 41, 679 (1999)

P-1.029, Monday June 27, 2005

Page 35: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Turbulent Transport and Mixing of Impurities in the Edge Plasma

O.E. Garcia, J. Gavnholt, V. Naulin, A. H. Nielsen and J. Juul Rasmussen

Association EURATOM-Risoe National Laboratory, OPL-128, DK-4000 Roskilde Denmark

Recent experimental observations have revealed that the transport in the edge and scrape-

off-layer (SOL) of toroidal plasmas is strongly intermittent and involves large outbreaks of

hot plasma. These bursts, often referred to as “blobs”, is formed near the last closed flux

surface (LCFS) and penetrate far into the SOL. They have a significant effect on the

profiles of density and temperature. We have investigated turbulent dynamics in the edge

and SOL numerically using the Risø ESEL-model that governs the dynamics of interchange

convection modes at the outboard mid-plane of a toroidal device and includes the self-

consistent evolution of the full pressure as well as potential profiles [1].

The transport of impurities in the edge plasma region is of increasing concern in fusion

research experiments. The impurities are mainly generated at the first wall and plasma

facing components, but are subsequently transported into the edge region and often all the

way to the plasma centre. The transport is found to be strongly anomalous and turbulence is

certainly playing a decisive role.

In the present contribution we investigate the turbulent transport and mixing of impurities

in the SOL by employing a test particle approach in the ESEL-model. The impurity density

is assumed to be low and the impurities are ionised, thus they are described as particles that

are passively convected by the turbulent ExB-velocity. We observe that the impurity

transport cannot be described by as simple diffusion process; it is strongly anomalous with

step length probability distributions having fat non-Gaussian tails. However, the impurities

are rapidly mixed in the SOL region and the impurity density attains an “equilibrium”

profile ranging into the edge of the plasma inside the LCFS. The “mixing” time is found to

be only weakly influenced by the initial position of the impurities, and is only few times the

characteristic period of the bursts. Thus, even particles released far into the SOL are rapidly

transported inside the LCFS.

[1] O.E. Garcia, V. Naulin, A.H Nielsen and J. Juul Rasmussen, Phys. Rev. Lett. (2004)

92, 165003.

P-1.030, Monday June 27, 2005

Page 36: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Clustering and pinch of impurities in plasma edge turbulence

M. Priego, O. E. Garcia, V. Naulin, J. Juul Rasmussen

Association EURATOM–Risø National Laboratory,

OPL-128 Risø, DK-4000 Roskilde, Denmark

Abstract

The turbulent transport of impurity particles in plasma edge turbulence is investigated.

The impurities are modeled as a passive fluid advected by the electric and polarization

drifts, while the ambient plasma turbulence is modeled using the Hasegawa–Wakatani

paradigm for resistive drift-wave turbulence. The features of the turbulent transport of im-

purities are investigated by numerical simulations using a novel code that applies semi-

Lagrangian pseudospectral schemes. In particular, we focus on the compressible effects

that arise as a consequence of impurity-particle inertia. First, the density of inertial impu-

rities is found to correlate with the vorticity of the electric drift velocity. Second, a radial

pinch scaling linearly with the mass–charge ratio of the impurities is discovered. Theoreti-

cal explanation for these observations is obtained by analysis of the model equations.

P-1.031, Monday June 27, 2005

Page 37: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

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P-1.032, Monday June 27, 2005

Page 38: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

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P-1.033, Monday June 27, 2005

Page 39: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

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P-1.034, Monday June 27, 2005

Page 40: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Predictive integrated modelling for ITER scenario

J-F Artaud, T.Aniel, V. Basiuk, L.-G. Eriksson, G.Giruzzi, G.T. Hoang, G. Huysmans,

F. Imbeaux, E. Joffrin, Y. Peysson, M. Schneider, P. Thomas

Association Euratom-CEA, CEA Cadarache, CEA-DSM-DRFC,

F-13108 St Paul lez Durance, France

[email protected]

Integrated modeling of ITER scenarios, e.g. inductive H-modes, steady-state and hybridscenarios, is essential for assessing their viability. Such modeling requires an accuratedescription of the relevant physics involved, in particular for the heat and particle transport.Different transport models are able to reproduce the existing experiments in variousdevices. However, they can yield significantly different extrapolation results for ITER,either the global performance or the profiles of plasma parameters (for example pressureand current density). In this work, the uncertainty on the prediction of the scenario forITER is evaluated. For this purpose, we use two transport models, which have beenintensively validated against various discharges from the muti-machine database, especiallyAUG, JET and DIII-D. The first model is GLF23 [1] linked with pedestal model (using twoterms scaling law and critical pressure gradient compute in equilibrium code HELENA).The second is a model in which the diffusion coefficient profile is a gyroBohm likeanalytical function, and is renormalized so that the resulting plasma profiles are consistentwith a given global energy confinement scaling.

This paper reports, for the first time, 1-D integrated simulations of full ITER discharges

using the CRONOS code [2]. The package of codes CRONOS includes modules for 2D

MHD equilibrium, neoclassical transport (NCLASS) heat, currents and particle sources;

and in particular a new orbit following Monte Carlo code dedicated to the simulation of

fusion-born alpha particles[3]. The CRONOS simulations give access to the dynamics of

the discharge and permit the study of the interplay of heat transport, current diffusion and

sources. In addition, these results are checked on 0-D simulations. We finally evaluated the

effect of transport model on fusion power and current profile evolution.

[1] R.E. Waltz and R.L. Miller, Phys. Plasma, 6,4265 (1999)

[2] V. Basiuk et al., Nucl. Fusion, 43,822 (2003)

[3] M. Schneider et al., Proc. 12th ICPP (2004), Nice, France

P-1.035, Monday June 27, 2005

Page 41: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

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P-1.036, Monday June 27, 2005

Page 42: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

!

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P-1.037, Monday June 27, 2005

Page 43: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Zero Dimensional Model for Transport Barrier Oscillations in Tokamak

Edge Plasmas

G. Fuhr1, S. Benkadda1, P. Beyer1, X. Garbet2, P. Ghendrih2, Y. Sarazin2

1 LPIIM, CNRS – Université de Provence, St. Jérôme, Case 321, 13397 Marseille Cedex 20,

France2 Association Euratom – CEA sur la Fusion, CEA Cadarache, 13108 St-Paul-lez-Durance,

France

Transport barriers at the plasma edge are key elements of high confinement regimes in fusion

devices. In typical configurations, such barriers are not stable but exhibit quasi-periodic re-

laxation oscillations. In this work, a zero-dimensional model for such oscillations is presented

describing the non linear dynamics of mode amplitudes. The relevant modes are determined

by applying a proper orthogonal decomposition[1, 2] to the results from three dimensional tur-

bulence simulations with a transport barrier generated by an imposed shear flow[3]. It is found

that the relevant modes depart from linear modes. This leadsto a zero dimensional model which

reproduces barrier oscillations. Furthermore, an analytic expression for the frequency as a func-

tion of shear flow is obtained.

References

[1] J. L. Lumley, inAtmospheric Turbulence and Radio Wave Propagation, edited by A. M.

Yaglom and V. I. Tatarski (Nauka, Moscow, 1967), p. 166.

[2] P. Beyer, S. Benkadda, X. Garbet,Phys. Rev. E61 813 (2000)

[3] P. Beyer, S. Benkadda, G. Fuhr et al.,“Non linear dynamics of transport barrier relax-

ations in tokamak edge plasmas”, to appear in Physical Review Letters.

P-1.038, Monday June 27, 2005

Page 44: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Study of nonlinear phenomena in a tokamak plasma using a novel

Hilbert transform technique

R. Jha1, D. Raju1, A. Sen1 1 Institute for Plasma Research, Bhat, Gandhinagar, India

A tokamak plasma is rich in nonlinearities of various kinds. The interacting low frequency

long wavelength coherent modes are dominant in the core and the confinement regions

whereas modes in a broad range of frequencies and wavelengths typically characterize the

edge plasma. These interactions have been studied conventionally using a varieties of

techniques including Fourier and wavelet transforms. Recently a new technique, known as

the empirical mode decomposition (EMD) method, has been introduced which allows

extraction of a finite number of intrinsic modes from the data. The Hilbert transform of

such modes help to determine instantaneous frequencies and sharp changes in the

instantaneous frequencies are identified as a signature of nonlinear phenomena in the data.

This method is suitable for studying non-linearity present in the transient events. The

plasma transients during start-up and current termination phases in ADITYA tokamak

have been studied using this technique. The analysis of signals from an array of Mirnov

coils shows that nonlinear interaction among low frequency long wavelength modes plays

an important role in current penetration during the start-up phase. On the other hand,

interaction among low m modes lead to disruption during current termination phase.

Langmuir probe data from the turbulent edge plasma have also been analyzed using this

technique. The data show signatures of intermittency in the form of sporadic bursts of

mode energy. The Hilbert spectrum also allows evaluation of the degree of non-

stationarity. It is observed that only high frequency signals (exceeding 20 kHz) are non-

stationary.

P-1.039, Monday June 27, 2005

Page 45: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Long range time correlations in the electrostatic fluctuations of a low

temperature dc magnetised plasma

Joyanti Chutia, Nirab Chandra Adhikary, Arup Ratan Pal and Heremba Bailung

Plasma Physics Laboratory, Material Sciences Division

Institute of Advanced Study in Science and Technology

Vigyan Path, Paschim Boragaon, Garchuk, Guwahati – 781 035

Assam, India.

The electrostatic fluctuations play an essential role both temporally and spatially in the

dynamics of plasma transport. Study of the dynamics of plasma transport is often needed for

controlling the magnetically confined plasmas. There are many experimental observations

which have already been done for determination of long time correlations present in plasma

fluctuations in devices like Tokamak and Stellarators but no such experimental observations

have yet been reported for low temperature magnetized plasmas. In this work we are trying to

locate the long-range time correlation in the plasma fluctuations of a low temperature dc

magnetized plasma system.

Weakly ionized plasma is susceptible to a number of low-frequency electrostatic

instabilities. Among those, one is the ‘E × B instability’, or some times called the ‘cross field

instability’. The study of this electrostatic instability generated due to the effect of E × B flow

in low-pressure plasmas is important to study since this kind of instability may generate in

space plasmas and also in the fusion plasma devices. The resultant E × B drifts, coupled with

an equilibrium radial density gradient can cause exponential growth of the perturbation,

therefore the instability grows at a particular range of applied magnetic field. Collisions with

the background neutral gas tend to damp out this growth, and hence there is a critical gas

pressure at which the instability sets in.

We have analyzed the ion saturation current taken by a Langmuir probe in the plasma

under the influence of this E × B flow having an instability within the range of 45 MHz to 105

MHz. These data are taken at different positions in the plasma chamber both radially and

axially in order to clarify the possible existence of the long-range time correlations present in

the fluctuations by calculating the Hurst exponent by various method. The results clearly

expose the existence of the long-range time correlations present in fluctuations in the plasma.

P-1.040, Monday June 27, 2005

Page 46: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Transport Properties of Low Aspect Ratio L=1 Helical Systems

M. Aizawa, S. Shimizu, A. Aility and S. Shiina*

Institute of Quantum Science, College of Science and Technology,

Nihon University, Tokyo, 101-8308, JAPAN

E-mail:[email protected]

*National Institute of Advanced Industrial Science and Technology, Tsukuba, JAPAN

The L=1 helical axis systems applying the control of effective toroidal curvature term

Tg defined as the sum of usual toroidal curvature term tg and one of the nearest satellite

harmonics of helical field term 0g [1], have been studied to improve particles confinement

properties. The trapped particle confinement in the L=1 helical system with a large field

period number N is considerable satisfactory by the particle orbits tracing, the longitudinal

adiabatic invariant J method and calculating the neoclassical transport particle and heat

fluxes.

If we consider a compact system, a small N and low aspect ratio system is desirable[2].

The transport properties of this compact system have been studied by the same methods

described above, and we have improved a particle transport by controlling the effective

curvature term.

--------------------------------------------------------------------------------------------------------------

[1] M. Aizawa and S. Shiina ; Phys. Rev. Lett. 84 2638 (2000)

[2] M. Aizawa, H.Uchigashima and S. Shiina ; ECA Vol. 28G P-5.106 (2004)

P-1.041, Monday June 27, 2005

Page 47: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Transient electron heat transport and reduced density fluctuation

after pellet injection in JT-60U reversed shear plasmas

H. Takenaga1, N. Oyama1, A. Isayama1, S. Inagaki2, T. Takizuka1, T. Fujita1

1 Naka Fusion Research Establishment, Japan Atomic Energy Research Institute,

801-1 Mukouyama, Naka, Ibaraki 311-0193, Japan

2 National Institute for Fusion Science, 322-6 Oroshi-cho, Toki, Gifu 509-5292, Japan

Understanding of anomalous turbulent transport is a crucial issue, especially for electron

heat transport, because it remains at an anomalous level even with the ion heat transport

reduced to the neoclassical level. When the pellet was injected into the JT-60U reversed shear

plasma with a strong internal transport barrier (ITB) (Ip=2.2 MA and BT=4 T), the central

density and the stored energy started to increase. A time dependent 2D full wave simulation

based on reflectometer signal indicated that the density fluctuation (kr~3 cm-1) was reduced by

a factor of about 2. Power balance analysis before and after the pellet injection indicated

reduction of ion thermal diffusivity to the neoclassical level, but no reduction of electron

thermal diffusivity (ce). Transient response of the electron heat transport during the reduction

of the density fluctuation was investigated for better understanding of relation between

electron heat transport and density fluctuation. Density fluctuation was reduced 6 ms after the

pellet injection, when a cold pulse induced by a pellet deposition outside the ITB was

propagated into the strong ITB region. The reduction of the electron temperature (Te) was

enhanced at the outer ITB portion. However, the cold pulse propagation was stopped in the

ITB region and Te around the ITB shoulder did not decrease. The value of ce estimated from

the power balance can not explain such time behavior of Te. Cold pulse analysis indicated that

ce decreases by a factor of 3 in the inner ITB portion, and ce increases once by a factor of 1.5

at 20 ms after the pellet injection and then decreases to slightly smaller value than that before

pellet injection in the outer ITB portion. The time scale of the ce change (several tens ms) is

similar as the time scale of Te profile change and is larger than the time scale of the

fluctuation reduction (several ms). The electron heat transport seems to be decoupled with the

measured density fluctuation, although the electron heat transport might be related to the

trigger for the reduction of the measured density fluctuation.

This work was partly supported by JSPS, Grant-in-Aid for Scientific Research (A) No. 16206093.

P-1.042, Monday June 27, 2005

Page 48: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Measurement of local electrical conductivity and thermo-dynamical

coefficients in JT-60U

M. Kikuchi , T. Suzuki, T. Fujita and the JT-60 team

Naka Fusion Research Establishment, Japan Atomic Energy Research Institute,

Naka Ibaraki, 311-0193 Japan.

Tokamak confinement requires poloidal field produced by its plasma current. Existence

of neoclassical trapped particle correction to electrical conductivity and bootstrap current are

confirmed experimentally by a global parameters such as surface voltage in many tokamaks

[1]. But detailed measurement of “local” electrical conductivity and thermo-dynamical

coefficients as a driving force of bootstrap current were not yet done in tokamaks.

The MSE measurement enables us to evaluate spacial distributions of current profile

and toroidal electric field in JT-60U tokamaks. And the many discharges in JT-60U show

good agreement between prediction of neoclassical parallel transport theory and the

measurements. Comparison of theoretical and measured local electrical conductivity and

bootstrap coefficients will be presented.

[1] M. Kikuchi, M. Azumi, "Experimental evidence for the bootstrap current in a tokamak",

[Review Article], Plasma Physics and Controlled Fusion 37(1995)1215.

P-1.043, Monday June 27, 2005

Page 49: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Compar isons of gyrokinetic PIC and CIP codes

Y. Idomura1, Y. Kishimoto

1, 2, and S. Tokuda

1, 3

1 Department of Fusion Plasma Research, Naka Fusion Research Establishment, Japan

Atomic Energy Research Institute, Naka, Ibaraki 311-0193, Japan

2 Graduate School of Energy Science, Kyoto University, Uji, Kyoto 611-0011, Japan

3 Center for Promotion of Computer Science and Engineering, Japan Atomic Energy

Research Institute, Ueno, Tokyo 153-0061, Japan

A 5-dimensional gyrokinetic simulation is an essential tool to study anomalous turbulent

transport in tokamak plasmas. Although several gyrokinetic simulations have been

developed based on particle and mesh approaches, most of full torus global simulations

have adopted a particle approach because of limitations on computational resources. A hf

Particle-In-Cell (PIC) method [1] enabled an accurate calculation of small amplitude

(hn/n~1%) turbulent fluctuations in collisionless plasmas. However, it is difficult to apply

a conventional hf PIC method to more realistic long time turbulence simulations where

non-conservative effects such as heat and particle sources and collisions are important,

because it was designed using a conservation property (Liouville’s theorem) of a

collisionless gyrokinetic equation. On the other hand, a mesh approach, which is much

more flexible about treatments of these non-conservative effects, is likely to become

another solution due to recent advances in computational fluid dynamics (CFD) schemes

and increasing computational resources. In order to examine a possibility of a mesh

approach from a point of view of numerical properties and a computational cost, a new

gyrokinetic Vlasov code has been developed using a Constrained-Interpolation-Profile

(CIP) method [2], which is one of advanced CFD schemes based on a semi-Lagrangian

approach. The new code is tested in 4-dimensional gyrokinetic simulations of the slab Ion

Temperature Gradient driven (ITG) turbulence. In this work, numerical properties of the

new gyrokinetic CIP code are shown, and comparisons of gyrokinetic PIC and CIP codes

are discussed.

[1] S. E. Parker and W. W. Lee, Phys. Fluids B 5 77 (1993).

[2] T. Yabe, F. Xiao, and T. Utsumi, J. Comput. Phys. 169 556 (2001).

P-1.044, Monday June 27, 2005

Page 50: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Intermittent Fluctuation Proper ty of JT-60U Edge Plasmas

H. Miyoshi1, N. Ohno

2, Y. Uesugi

3, N. Asakura

4, S. Takamura

1, Y. Miura

4

1. Department of Energy Engineering and Science, Graduate School of Engineering,

Nagoya University, Nagoya 464-8603, Japan

2. Ecotopia Science Institute, Nagoya University, Nagoya 464-8603, Japan

3. Department of Electrical and Electronic Engineering, Graduate School of Engineering,

Kanazawa University, Kanazawa 920-8667, Japan

4. Japan Atomic Energy Research Institute, Naka, Ibaraki 311-0193, Japan

Recently, intermittent convective plasma transport, so-called "blobs" has been observed in

the edge plasmas of several fusion devices, which is thought to play a key role for cross-field

plasma transport. In this presentation, we will report the statistical analysis of the intermittent

edge plasma fluctuation of ion saturation currents Isat and/or floating potential measured with

probes in JT-60U tokamak device.

The fluctuation property has been analyzed with probability distribution function (p.d.f.) to

obtain a basic property of the intermittent plasma transport. When large positive fluctuations

are much greater than expected values from a random distribution (Gaussian distribution), the

p.d.f. is positively skewed. The deviation from the Gaussian distribution function can be

characterized by skewness. In the JT-60U, the reciprocating Mach probes are installed at the

low field side (LFS) mid-plane and just below the X-point. We have mainly analyzed the time

evolution of Isat with the Mach probe installed in the mid-plane at the low-field side and

divertor probe array. The sampling time of Isat is 2 and/or 5os. Cross- and parallel- transports

of the intermittent density bursts including ELM events[1] are also discussed by comparing

the spatiotemporal behaviour of the fluctuations in Isat.

At the LFS mid-plane, the skewness of Isat increases with the distance from separatrix dsep.

It peaks around dsep=60-80mm, where direction of the parallel SOL flow changes downward

to upward, in both L-mode and ELMy H-mode discharge. It indicates that there is strong

relation between the process of cross-field transport like blobs and parallel SOL flow. From

analysis SOL profiles of Isat in ELMy H-mode discharge, decay length of Isat during ELM is

about three times as long as one of Isat between ELMs. Thus the plasma during ELM

convectively transports in the radial direction much easier in comparison with the one

between ELMs, which corresponds to bulk plasma.

[1] N. Asakura, M. Takechi, N. Oyama, T. Nakano, J. of Nucl. Mater. 337-339 (2005) 712-716.

P-1.045, Monday June 27, 2005

Page 51: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

First results of the Gas Puffing Imaging Diagnostics

in a reversed-field pinch plasma

Y. Yagi, H. Koguchi, S. Kiyama, H. Sakakita, Y. Hirano

AIST, Tsukuba, Ibaraki 305-8568, Japan

R. Cavazzana, P. Scarin, G. Serianni, M. Agostini, N. Vianello

Consorzio RFX, Associazione Euratom-ENEA sulla Fusione,

corso Stati Uniti 4, Padova, Italy

The investigation of electrostatic turbulence in the edge region of fusion plasmas, generally

measured by sets of Langmuir probes, has shown coherent structures emerging from the

background, which are responsible for up to 50% of the particle transport in reversed-field

pinches (RFP).

A Gas Puffing Imaging Diagnostic (GPID) has been developed at Consorzio RFX, aimed at

identifying such structures by a non-invasive method, allowing unperturbed and high plasma

current discharges to be investigated. The system consists of a gas-puffing nozzle, 32 optical

chords measuring the D! radiation emitted from the puffed gas, and an array of Langmuir

probes to compare the turbulent pattern with the optical method at low currents.

The equipment was installed and the first measurement using the GPID in RFP was carried

out in the TPE-RX RFP device at AIST, for plasma currents I∀ = 200-350 kA and various

discharge conditions (# = 1.4-1.7, with and without PPCD), providing the following results.

A toroidal propagation velocity of the fluctuations is found in the range 20-30 km/s from the

correlation of line-integrated signals. The wavenumber vs frequency spectrum shows the

characteristic broadband features of turbulence in the edge of RFPs, typically detected by

electrostatic probes. The probability distribution function of fluctuations displays non-

Gaussian tails, which become more pronounced at the shorter time scales, so that an

intermittent character is observed in the range 3-100 µs. Bursts are detected in the signals

and the correspondence of burst clusters with MHD behaviour is assessed.

Thus, it is proved that the GPID is a useful tool for studying the electrostatic turbulence in

the edge region of fusion plasmas, even at high plasma currents and thermal loads.

P-1.046, Monday June 27, 2005

Page 52: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Weak temperature dependence of the thermal diffusivity in

high-collisionality regimes in LHD

J. Miyazawa1, H. Yamada

1, S. Murakami

2, H Funaba

1, and the LHD experimental group

1 National Institute for Fusion Science, 322-6 Oroshi, Toki, 509-5292, Japan

2 Department of Nuclear Engineering, Kyoto University, Kyoto 606-8501, Japan

Positive density dependence of energy confinement times, vE, as expressed in the international

stellarator scaling 95 (ISS95), where vE ¶ ne_bar0.51

P–0.59

(ne_bar is the line-averaged density and

P is the heating power), declines in high-collisionality regimes in the Large Helical Device

(LHD) experiments. In the low-collisionality regime, where parameter dependences in ISS95

agree well with the experiment, the temperature dependence of thermal diffusivity is as strong

as predicted by the gyro-Bohm model and/or the neo-classical theory. As the collisionality

increases to the plateau and the Pfirsh-Schlüter regimes, the temperature dependence becomes

moderate, where the thermal diffusivity is proportional to the square root of the electron

temperature. Also in the high-collisionality regimes, the thermal diffusivity is inversely

proportional to the magnetic field strength and the electron temperature gradient is proportional

to the electron temperature, while both of the electron temperature and its gradient is

proportional to two-thirds of the heating power normalized by the density. Due to the latter

characteristic, the electron temperature profiles converge to a typical shape, like the profile

stiffness observed in tokamaks. Another similarity to the stiffness is found in the temperature

scale length profile, although these two are not necessarily the same phenomenon. Compared

with ISS95, the energy confinement time expected from these observations has a weaker

density dependence together with a mitigated power degradation as vE ¶ ne_bar1/3

P–1/3

. Since the

main heating is a neutral beam (NB) injection in LHD, it is necessary to take into account the

beam deposition profile, which becomes shallower in the high-density plasmas, for the

effective heating power estimation. In this study, the line-averaged value of the NB heat flux is

adopted as the effective heating power. Then, the new scaling with the weak density

dependence and mitigated power degradation well reproduces the experimental vE, for a wide

range of experimental conditions, i.e. the magnetic field strength of 0.4 – 2.75 T (d < 2%), the

total heating power of 2 – 12 MW, and ne_bar of (0.1 – 1) · 1020

m-3

. Even in the low-

collisionality regime where ISS95 reproduces the parameter dependences well, the new scaling

is applicable within 20% error. This indicates the importance of confinement property in the

outer region, where the temperature is relatively low and in the high-collisionality regimes.

P-1.047, Monday June 27, 2005

Page 53: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

3D Simulation of the Magnetic Shear contribution on the Improvement of

the Confinement in Plasma of Tokamak

M. El Mouden1, D. Saifaoui1, A. Dezairi2,

1. Laboratory of Theoretical Physics, Faculty of Sciences- Ain Chock, Casablanca, Morocco 2. Laboratory of Physics of the Condensed Matter, Faculty of Sciences- Ben M’sik, Casablanca, Morocco Email contact of main author: [email protected]

Anomalous transport observed in tokamaks is known as the result of the electrostatic

and magnetic turbulence. Thus, in the presence of electric perturbation and for the normal

profile of the safety factor q, the stochasticity of the trajectories increases and this is the

principal cause of diffusion of particles through magnetic surfaces. However for the reversed

shear case, the most important result is the impressive formation of a strong transport barrier,

which is localized near of minimum value of q (q is the safety factor). This barrier plays a

very important role in the improvement of the plasma confinement while preventing its radial

diffusion. To evaluate quantitatively the diffusion, we simulate from the Mapping equations,

the diffusion coefficient in each of the two previous cases, and we draw the ratio that shows a

clean reduction in the diffusion observed in the reversed magnetic shear profile. Therefore,

the diffusion decreases, the confinement improves and the control of the fusion reactors to

function in these modes permits the reduction of the anomalous transport in the tokamaks.

Then, we simulate the dynamics of plasmas in the torus of the tokamak for both

normal and reversed shear using the 3D toroidal coordinates system and the mean parameters

of the principle tokamak stations (TEXT, JET, ITER) in order to compare the obtained results

which is going to help us to improve our understanding concerning the production of energy

by the thermonuclear way that will be in many years the mean source of energy in the world.

Key words: Plasma confinement, Tokamak, Anomalous transport, Magnetic shear,

Transport barrier, Particles diffusion.

P-1.048, Monday June 27, 2005

Page 54: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Momentum transport and plasma rotation spin up in TCV

A. Scarabosio, A. Bortolon, A. Karpushov, B. Duval, A. Pochelon.

Centre de Recherches en Physique des Plasmas,

Association EURATOM-Confédération Suisse, Ecole Polytechnique Fédérale de Lausanne,

(EPFL), CH-1015 Lausanne, Switzerland

The transport of plasma momentum in tokamaks has been extensively studied in presence of

strong external source (NBI) which masks the spontaneous self-generated plasma rotation. In

the Tokamak à Configuration Variable (TCV) it is now possible to measure the carbon rotation

profile in absence of any external drive using the new active Charge eXchange Recombination

Spectroscopy system (presented in detail in a companion paper). Typically the plasma rotation

profile goes from values close to zero at the edge to up to 40 km/s at the sawtooth inversion

radiusr inv and is flat or slightly hollow insider inv. While finite size MHD modes only partially

reduce plasma rotation, locked mode minor disruptions can completely stop the rotation over

most of the plasma cross section. Subsequently, when the instability has disappeared, the dis-

charge recovers and reaches again stationary conditions for the main plasma parameters, while

the plasma angular velocity increases (spin up) and evolves on a slower time scale. The fre-

quency of the sawtooth precursor increases in a similar way as can be inferred from magnetic

fluctuation measurements at the plasma edge and core soft X-ray emissivity. While the global

time scale for the rotation spin up is of the order of 150-200 ms (»τE) the temporal evolution of

the toroidal rotation differs from the centre to the edge. The plasma region inside the inversion

radius evolves slowly resulting in an initially very hollow rotation profile which then evolves

to a flat profile while approaching stationary condition.

The momentum transport has been modeled with a simple 1D diffusion equation for the angu-

lar velocity in cylindrical approximation. Density gradients are neglected. The model includes

a momentum diffusivity coefficientDµ and a velocity pinch vp which may assume arbitrary

profiles. The equation is solved numerically and the best-fit values forDµ and vp are estimated.

Source-less or different ad hoc source models are used to test the sensitivity of the results and

to gain insight on the source location of the angular momentum. It turns out that sawtooth ac-

tivity has a strong influence on the momentum evolution and good agreement is found simply

usingDµ and vp profiles with two different values, inside and outside the inversion radius. The

results are compared with classical and neoclassical predictions for perpendicular momentum

transport and the different source models will be discussed in details.

P-1.049, Monday June 27, 2005

Page 55: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Simulation of the Absolute TCV Compact Neutral Particle Analyser

Charge-Exchange Spectrum

Ch. Schlatter, B. P. Duval, A. N. Karpushov

Ecole Polytechnique Fédérale de Lausanne (EPFL),

Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne, Switzerland

Knowledge of plasma fuel neutral profiles is indispensable for particle transport studies in

tokamaks. On TCV, a combination of experiment and simulation is used to recover the profiles.

For this purpose, the absolute Charge-eXchange (CX) particle emission energy spectrum for

hydrogen (H) and deuterium (D) of the Compact Neutral Particle Analyser (CNPA) in ohmic

plasma discharges in limiter configuration has been calculated based on simulations using the

kinetic transport code KN1D [1]. The CNPA is installed at the midplane of TCV, with a hor-

izontal line of sight perpendicular to the magnetic axis. Mass separation permits synchronous

measurements of H and D over a wide range of energies (500 .. 50·103 eV) [2].

KN1D requires accurate input profiles for electron density, electron temperature and ion tem-

perature together with the neutral particle pressure at the wall chamber. Te(r) and ne(r) are

obtained from Thomson scattering measurements with the density profile normalised using a

Far InfraRed interferometer. The ion (carbon) temperature profile is obtained from Charge eX-

change Recombination Spectroscopy (CXRS). The fitted profiles are mapped to the chord of

the CNPA. From the simulated hydrogenic neutral profiles, the radial neutral birth and reab-

sorption rate is determined and the remaining contribution to the escaping flux towards the NPA

is calculated. The neutral edge pressure is iterated in the code to achieve agreement with the

experimental CNPA CX-spectrum. Agreement is better than 10% (H) and 15% (D) for CNPA

channels with satisfactory statistics.

Pseudo chord measurements of identical plasma configurations, displaced along the vertical

coordinate, were used to probe different regions of the plasma cross section. The knowledge of

the birth region of the detected neutrals was used to build an edge hydrogen temperature profile

based on the inferred CNPA effective temperature TCNPA. The resulting profile is in agreement

with the CXRS carbon ion temperature profile for ρ = 0.5 .. 0.9 assuming an accuracy of 10%

of TCNPA.

References

[1] B. LaBombard, KN1D, PSFC/RR-01-9, MIT, Cambridge (2001).

[2] F. V. Chernyshev et al., 30th EPS Conf., St. Petersburg, ECA 27A (2003), P-4.71

P-1.050, Monday June 27, 2005

Page 56: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Density behavior during eITBs in TCV discharges: experimental

observations and theoretical calculations via transport simulations

E. Fable, O. Sauter, A. Zabolotsky, H. Weisen

Ecole Polytechnique Fédérale de Lausanne (EPFL)

Centre de Recherches en Physique des Plasmas (CRPP)

Association EURATOM - Confédération Suisse

CH - 1015 Lausanne, Switzerland

Abstract

Internal transport barriers (ITBs) are observed on most Tokamaks. Their dynamics is

being studied with more and more complete theoretical models benchmarked against ex-

perimental database. Because of modelling problems, most of the relevant works deal with

heat transport and the mechanisms of turbulence stabilization. Particle transport is still a

very complex and yet not well known issue, due to the uncertainties in the models and

the measurements of the anomalous pinch velocity and of the sources, in addition to the

question of the relevant anomalous diffusivities. In this paper we show results of an analy-

sis of the particle transport during electron internal transport barriers (eITBs), based both

on experimental data and on simulations using theASTRA code [1]. This issue has been

stimulated by observations from experiments assessing the steady state performance of

fully non-inductive discharges, which show strong correlation in the formation of an eITB

both for electron temperature and for density. Experiments on the effect of ohmic current

perturbation to probe the barriers have shown that the same response is present in both tem-

perature and density. Therefore the analysis shows that electron density transport barriers

are directly related to the local q profile and to the degree of reverse shear.

References

[1] G. V. Pereverzevet al., ASTRA,An Automatic System for Transport Simulations in a Toka-

mak, IPP Report 5/42 (August 1991).

P-1.051, Monday June 27, 2005

Page 57: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Electron heat transport dependence on plasma shape and collisionalityin EC heated L-mode TCV plasmas

Y. Camenen, A. Pochelon, A. Bottino, L. Curchod, E. Fable, I. Pavlov, R. Behn, S. Coda,

T.P. Goodman, M.A. Henderson, J.-M. Moret, L. Porte, O. Sauter, G. Zhuang

Centre de Recherches en Physique des Plasmas CRPP

Ecole Polytechnique Fédérale de Lausanne EPFL

Association EURATOM-Confédération Suisse, CH-1015 Lausanne

The plasma shaping capabilities and flexible ECH system of TCV are used to investigate elec-

tron heat transport in L-mode plasmas. A large range of plasma triangularities, from negative

to positive values, , is explored. Both the EC power deposition location and the

total EC power are varied, resulting in an extraordinarily wide range of normalized tempera-

ture gradient and electron temperature .

The electron heat diffusivity is shown to depend strongly on and weakly on at the

radius of investigation (mid-radius). Various possible dependences of on and ,

suggested by the experiments, will be tested with the ASTRA transport code.

The electron heat diffusivity , calculated from power balance, is clearly found to depend on

plasma triangularity. Measurements of heat pulse propagation confirm these results. A signifi-

cant reduction of , together with an increase of the electron temperature and confinement

time, is observed towards negativeδ. For example, identical profiles are obtained with half

of the EC power at negative triangularityδ=-0.4, as compared to positive triangularityδ=+0.4.

In addition, with an increase of electron collisionality, the electron heat transport is observed

to decrease strongly, consistent with the expected stabilizing effect of collisions on trapped

electron modes (TEM). Local gyro-fluid (GLF23) and global gyro-kinetic (LORB5) simula-

tions both indicate that TEM are unstable and potentially responsible for the anomalous heat

transport in these experimental conditions. GLF23 simulations show a reduction of the TEM

growth rate at high plasma collisionality.

In the present off-axis EC power deposition experiments, slow central electron temperature

oscillations are occasionally observed, similar to the ones observed in TORE SUPRA. The

dependence of their characteristics on power and power deposition location is described.

0.4 δ 0.4< <–

R LTe⁄ Te

Te R LTe⁄

χe Te R LTe⁄

χe

χe

Te

P-1.052, Monday June 27, 2005

Page 58: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Analysis of dissipation in MHD turbulence simulations in two and three

dimensions

J A Merrifield1, T D Arber1, S C Chapman1, R. O. Dendy2,1, W-C Müller3

1Department of Physics, University of Warwick, Coventry CV4 7AL, United Kingdom2Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxfordshire

OX14 3DB, United Kingdom3Max-Planck Institut für Plasmaphysik, EURATOM-Assoziation,Boltzmannstraße 2,

D-85748 Garching, Germany

The physical nature of the most strongly dissipative structures is central to the behaviour of

large scale numerical simulations of magnetohydrodynamic (MHD) turbulence. These struc-

tures relate to intermittency, which must be accommodated in models that are used to eval-

uate the scaling properties of the MHD turbulent fluctuations. Direct statistical analysis of

the spatial scaling of the dissipation, for example by means of structure functions, therefore

contributes to understanding the turbulence displayed by the velocity and magnetic fields. Nu-

merical and physical constraints require that the measured scaling be analysed by extending

turbulent cascade ideas from the inertial range into the range of lengthscales where dissipation

also affects, but does not dominate, the turbulence spectrum. The inferred dimension of the

most strongly dissipating structures (sheetlike, stringlike, or intermediate) is a key element

of such models. The present work focuses on comparing the statistical properties of dissipa-

tion in MHD turbulence simulations in two and three dimensions. The three dimensional data

is from Biskamp and Müller,Phys. Plasmas7, 4889 (2000)). It is found (Merrifieldet al.,

Phys. Plasmas12, 022301 (2005)) that the ratio of dissipation structure function exponents

obtained is that predicted by the She and Leveque (Phys. Rev. Lett72, 336 (1994)) theory

proposed by Biskamp and Müller. This supplies further evidence that the cascade mechanism

in three dimensional MHD turbulence is nonlinear random eddy scrambling, with the level

of intermittency determined by dissipation through the formation of current sheets. The two

dimensional data is from simulations using an isothermal high order code which encompasses

a more extensive inertial range because of its lesser demand on computational resources. In

these simulations, sheetlike dissipative structures can only appear in projection. Analysis of

this data using the techniques outlined above thus provides an important test of their robustness

and consistency, in addition to quantifying the extent to which MHD turbulence simulations

in two and three dimensions capture the same physics.This work was funded by Euratom and the United Kingdom Engineering and Physical Sciences Research Council.

P-1.053, Monday June 27, 2005

Page 59: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Numerical Plasma Edge MHD Stability Analysis Revisited

O. Kwon and S. Saarelma*Dept. Of Physics, Daegu University, Gyungbuk, Korea

* EURATOM/UKAEA Fusion Association, Culham Science Centre,Abingdon, Oxon. UK OX14 3DB

During the high confinement regime or the H-mode, a regular sequence of periods of

MHD activity including rapid loss of particles and energies from the edge region occurs.

These activities known as edge localized modes (ELMs) can deteriorate the global

confinements but are efficient in removing density and impurities. It is therefore desirable

to understand the physics underlying ELM activity. One of the main results found in a

previous study [1] using the MISHKA-I stability code [2] was that just before an ELM,

the equilibrium lies in the region unstable to low- to intermediate-n peeling ballooning

modes, and second stable to high-n ballooning modes due to low shear. After an ELM

crash, the flattening of the pressure gradient makes the plasma return to the low- to

intermediate-n stable region. We have revisited the plasma edge stability analysis of

several discharges in the diagnostic optimized configurations [1]. In this study, we have

used 2-D linearized ideal MHD stability code, ELITE [3]. The results show that our

results with the ELITE code are in good agreement with previous ones with the MISHKA

code both qualitatively and quantitatively. The computing time can be significantly

reduced and the real time analysis of edge MHD stability can be made possible to control

ELMs actively in future tokamak experiments. ELITE also allows the analysis to be

extended to high toroidal mode numbers without computations becoming too heavy.

Acknowledgement: This work was partly funded by the United Kingdom Engineering

and Physical Sciences Research Council and Euratom.

[1] S. Saarelma et al., accepted for publication in Plasma Phys. and Contr. Fusion

(2005).

[2] A.B. Mihailovskii et al., Plasma Phys. Rep. 23 (1997) 844.

[3] H.R. Wilson , P.B. Snyder, G.T.A. Huysmans and R.L. Miller Phys. Plasmas 9 (2002)

1277.

P-1.054, Monday June 27, 2005

Page 60: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Effects of radio frequency waves on dissipative low frequency instabilities

in mir ror plasmas

S. S. Kim and Hogun Jhang

Korea Basic Science Institute, Daejon 305-333, Korea

A study is presented on the influences of an applied strong radio frequency (rf) waves on the

dissipative low frequency instabilities in mirror plasmas. Both the flute and drift-type modes

are considered. A dispersion relation is derived for the low frequency waves based on the

two-fluid approach. Major rf and plasma parameters are identified giving rise to a significant

modification of the stability boundary.

P-1.055, Monday June 27, 2005

Page 61: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Studies of MHD instabilities in TJ-II plasmas

R. Jiménez-Gómez1, I. García-Cortés1, T. Estrada1, D. Spong2, J. A. Jiménez, B. van Milligen1,

A. López-Fraguas1, I. Pastor1 and E. Ascasíbar1

1 Labo ato o Nacional de Fusión, Asociación Euratom CIEMAT, Madrid Spain r ri -

s2 Oak Ridge National Laboratory, Oak Ridge, Tennes ee, USA

MHD instabilities in TJ-II Stellarator are being experimentally characterized in various

plasma parameter regimes and heating scenarios. Magnetic field fluctuations data are

collected using various Mirnov coil sets distributed at different toroidal sector of the

vacuum vessel. Special analysis is carried out by a new poloidal array of 15 probes

measuring poloidal magnetic field fluctuations with frequency resolution up to 1MHz.

This array only spans a poloidal angle of ±π維/2 mainly due to the complicated TJ-II

vacuum vessel geometry.

Most of the observed MHD activity depends on heating method (ECH or NBI).

In ECH plasmas, the effect of low order rationals inside the rotational transform profile

on MHD and transport properties has been previously described [1,2]. The analysis of

Mirnov coils data by Singular Value Decomposition (SVD) method and correlation

analysis techniques [3] is being used in order to understand the MHD involved in these

phenomena. As preliminary results, in discharges having vacuum rotational transform

1.65 at the edge, a rotating coherent mode has been found and it appears to be a

resonant m = 3, n = 5 mode, moving in the ion diamagnetic drift direction with

frequency in the range 20-25 kHz. Signals from reflectometer are compatible with mode

observation although no rotation can be deduced. On the other hand, high frequency

(200-300 kHz) modes have been found in plasmas with line density range 0.6 – 2.5 x

1019

m-3

and heated with ON/OFF-axis ECH (two gyrotrons, 200 kW each) and NBI

(240 kW). The frequency of these modes decrease with density and species mass and

their appearance seem to depend of density profile shape. Considering the low shear of

TJ-II, they are good candidates for Global Alfvén Eigenmodes related to some of the

main low order resonances n/m, 3/2 and 5/3. Reflectometer results show that the mode

is located at ρ ø 0.5-0.6 and rotates in the ion diamagnetic drift direction. HIBP signals

on these discharges indicate the presence of low order rationals as well [4].

1. I. García-Cortes et al., Nuclear Fusion 40 (2000) 1867-1874

2. T. Estrada et al., Plasma Physics and Controlled Fusion, 44 (2002) 1615-1624

3. M. Anton et al., 24th EPS, Berchtesgaden 1997,1ECA 21A part IV, 1645-1648

4. L. Krupnik et al., Electron ITB, rationals and fluctuations in the TJ-II, this conference.

P-1.056, Monday June 27, 2005

Page 62: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

First results from the Columbia Non-neutral Torus

T. Sunn Pedersen1, A. H. Boozer1, J. P. Kremer1, R. G. Lefrancois1, Q. Marksteiner1, X.

Sarasola1, 2

1 Columbia University, New York, NY, USA

2 Currently at CIEMAT, Madrid, Spain

The Columbia Non-neutral Torus (CNT) is a stellarator of unique design, dedicated to the

study of non-neutral and electron-positron plasmas confined on magnetic surfaces. Such

plasmas have unique properties and have not been studied experimentally before. Theory

predicts the existence of stable pure electron plasma equilibria [1,2,3,4]. In the small Debye

length limit, confinement is predicted to be excellent – CNT may be able to confine pure

electron plasmas for minutes.

CNT is a two-period, ultralow aspect ratio stellarator whose magnetic field is created from

only four circular coils, two internal, interlocked (IL) coils, and two external poloidal field

(PF) coils [5]. The angle between the IL coils can be changed to create rather different

magnetic topologies. At the present angle of 64 degrees, CNT has an aspect ratio of A=1.8

and iota ranging from 0.12 at the magnetic axis to 0.22 at the last closed flux surface. At an

angle of 88 degrees, CNT is predicted to have A=2.5, and a nearly flat iota profile, iota

~0.56. Magnetic field strengths up to 0.33 Tesla on axis can be achieved.

CNT operation started in November 2004. Magnetic surfaces have been successfully

mapped and agree with the calculated magnetic fields [6], verifying that large magnetic

surfaces of high quality are present. First pure electron plasmas are expected in the spring

of 2005. We will report on the first experimental results from CNT as well as on results

from recent three-dimensional numerical calculations of pure electron equilibria in various

toroidal geometries, including the present CNT configuration [4].

[1] T. Sunn Pedersen and A. H. Boozer, Phys. Rev. Letters 88, 205002 (2002)

[2] T. Sunn Pedersen, Phys. Plasmas 10, p. 334 (2003)

[3] A. H. Boozer, Phys. Plasmas 11, p. 4709 (2004)

[4] R. Lefrancois et al., submitted to Phys. Plasmas

[5] T. Sunn Pedersen et al., Fusion Science and Technology 46, p. 200 (2004)

[6] See X. Sarasola et al., this conference, for more details.

The CNT experiment is supported by the United States Department of Energy.

P-1.057, Monday June 27, 2005

Page 63: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Field Line Mapping Results in the CNT Stellarator

X. Sarasola1, 2

, T. Sunn Pedersen1, J. P. Kremer

1, R. G. Lefrancois

1, Q. Marksteiner

1,

N.Ahmad3

1 Columbia University, New York, NY, USA 2 Currently at CIEMAT, Madrid, Spain

3 UC Berkeley, Berkeley, CA, USA

The Columbia Non-neutral Torus (CNT), located at Columbia University, is a toroidal, ultra-

high vacuum stellarator designed to confine pure electron and other nonneutral plasmas. The

configuration is unique and simple: Four circular coils create a two-period ultralow aspect

ratio stellarator. A detailed mapping of the nested magnetic surfaces in CNT is one of the

most relevant results achieved during the first months of operation of the experiment. A 50 eV

electron beam emitted by a small moveable electron gun was used to follow the field lines

around the torus and hit two moveable ZnO coated aluminum rods that emit visible light

when struck by the e-beam. For each position of the e-gun, the phosphor rods scanned the

cross-section of the torus allowing a standard digital camera to record a single magnetic

surface in a five second exposure. Detailed mapping of the nested magnetic surfaces was

completed at 0.08 T in several configurations. These experimental results will be presented

along with details of the field line mapping system. The results agree very well with

numerical predictions. In particular, the baseline configuration has an ultralow aspect ratio

(A<2) with nested magnetic surfaces without any significant island chains.

Experiments were also conducted to visualize the full three-dimensional shape of magnetic

surfaces using electron beam ionization of background gas. The gas density and electron

beam energy were varied in the experiment to create a glowing shell in the shape of the

magnetic surface on which the electron beam is injected. Details of these experiments and

pictures of the magnetic surfaces will be shown.

P-1.058, Monday June 27, 2005

Page 64: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

RECENT DEVELOPMENTS IN QUASI-POLOIDAL STELLARATOR

PHYSICS

J.F. Lyon, D.A. Spong, D.J. Strickler, S.P. Hirshman

Oak Ridge National Laboratory, PO Box 2009, Oak Ridge, TN 37831, USA

The quasi-poloidal stellarator QPS, now in the R&D and prototyping phase, has very low

plasma aspect ratio (<R>/<a> ~ 2.7, 1/4–1/2 that of existing stellarators). Approximate

poloidal symmetry in magnetic coordinates is achieved by the use of a racetrack-shaped

magnetic axis and vertically elongated crescent-shaped cross-sections in the regions of high

toroidal curvature. The quasi-poloidal symmetry and reduced effective field ripple lead to

large reductions in: neoclassical transport at low collisionality; bootstrap current; and

poloidal viscosity, which allows large E x B poloidal flows for suppression of anomalous

transport. The magnetic configuration is stable to finite-n ballooning modes, external kink

modes, and vertical instability to <b> ~ 5%.

Nine independent coil currents allow varying: the neoclassical transport by a factor of 12-

36, degree of quasi-poloidal symmetry by a factor of 9, and poloidal viscosity by a factor of

6-30. Departure from ideal poloidal symmetry is found by following electron beam orbits

since the deviation from a flux surface is a direct measure of the non-poloidally symmetric

components of the magnetic field. Magnetic islands can be suppressed by varying modular

coil currents to minimize the residues of the dominant island chains or by using the more

conventional technique of targeting rotational transform profiles that avoid nearby low-

order resonances.

Plasma flow generation and damping affects enhanced confinement regime access,

impurity transport and magnetic island growth. Unlike tokamaks, stellarators have finite

damping and flow components in both toroidal and poloidal directions. The drive

mechanisms for stellarator flows differ from those of tokamaks due to the presence of the

ambipolar electric field. We have developed a fluid moments approach based on a method

by Sugama and Nishimura that self-consistently evaluates both viscosities and neoclassical

transport coefficients for stellarators of arbitrary magnetic field structure. For fixed

parameters typical of an ICRF-heated plasma, our model predicts poloidal/toroidal flow

ratios at the half radius of -0.04 in HSX, 0.45 in NCSX, and 52 in QPS. Further variations

of this flow ratio and flow shear result from different electric field roots.

Supported by USDOE under Contract DE-AC05-00OR22725 with UT-Battelle, LLC.

P-1.059, Monday June 27, 2005

Page 65: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Fast soft x-ray camera observation of fast and slow

reconnection events on NSTX

B. C. Stratton1, S. von Goeler1, J. Breslau1, E. Fredrickson1, W. Park1,

S. Sabbagh2, D. Stutman3, K. Tritz3, and L. Zakharov1

1Princeton Plasma Physics Laboratory, Princeton, New Jersey, USA

2Columbia University, New York, New York, USA

3Johns Hopkins University, Baltimore, Maryland, USA

Reconnection events on the National Spherical Torus Experiment (NSTX) are studied using

data from a new soft x-ray camera diagnostic1, radially-viewing soft x-ray diode arrays, and

Mirnov coils. The camera has a wide-angle tangential view of the plasma and can capture 300

images per discharge at rates up to 500000 frames per second. Two classes of m=n=1

reconnection events are seen: events such as sawteeth and internal reconnection events (IREs)

characterized by rapid (~200 µs) reconnection, and events in which reconnection occurs on a

much slower time scale. The slow events are characterized by a mode with a frequency of ~2

kHz which grows to saturation in 1-2 ms and decays in 20-100 ms. The mode structure in the

slow events is similar to that in the precursor and postcursor oscillations for a sawtooth crash

but on a much slower time scale. The slow events appear to occur only in relatively low !

discharges with ohmic heating or low-power high harmonic fast wave heating, but not in

neutral beam heated discharges, while the fast events occur in all types of discharges. The

ESC equilibrium and stability code is used to reconstruct the mode evolution from the fast

soft x-ray camera data. Nonlinear resistive MHD modeling with the M3D code and PEST

code stability analysis is used to predict the growth rates and island structures of the fast and

slow events, with the goal of understanding the conditions which lead to the two types of

events.

1B. C. Stratton et al., Rev. Sci. Instrum. 75 (2004) 3959.

P-1.060, Monday June 27, 2005

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Scaling of kinetic instability induced fast ion losses in National Spherical

Torus Experiment

E.D. Fredrickson, D. Darrow, S. Medley, J. Menard, H. Park, L. Roquemore,

Princeton Plasma Physics Laboratory, Princeton, NJ

D. Stutman, K. Tritz, Johns Hopkins University, MD,

S. Kubota, University of California, Los Angelos, CA.

K.C. Lee, University of California, Davis, CA

Losses of fast ions are correlated with bursts of high frequency instabilities on

NSTX. It is important to understand the conditions under which these fast ion

losses occur and to predict whether such losses are to be expected in the desired

operational regimes of NSTX or future fusion reactors. These losses raise the

ignition threshold for fusion reactors, adding cost and uncertainty to the design.

They may also challenge the engineering designs of plasma facing components

with high transient heat loads. In this paper we describe some initial experiments

to unfold the empirical scaling of these losses with dimensionless parameters. We

also describe the wide variety of fast ion instabilities whose presence is correlated

with the losses. We focus, in particular, on high performance NSTX plasmas in

regimes similar to the targeted operating regime. The neutron rate signal, most

sensitive to the density of fast ions close to the full beam injection energy, is used

to measure transient fast ion loss events.

In beam heated plasmas we see transient neutron rate drops, correlated with

fast ion driven instabilities, including modes identified as toroidal Alfvén

eigenmodes and fishbone-like fast frequency chirping energetic particle modes.

The transient fast ion loss events are most often correlated with bursting modes

which exhibit strong frequency chirping. The chirping modes have kink rather

than tearing parity and tend to be localized to the plasma core region. Some

strongly chirping modes occur without measurable neutron rate drops, suggesting

that either the modes caused losses of predominantly lower energy ions, or

redistributed the fast ions within the plasma. The TAE-like modes can be bursting

or quasi-stationary and in many cases also exhibit weak frequency chirping

(change in frequency of 10-20% over < 1msec). The modes are seen over the full

range in beta on NSTX, although when conventional MHD (tearing modes, or

saturated kink modes) are present, burst modes are less common.

* Work supported by U.S. DOE Contract DE-AC02-76CH03073.

P-1.061, Monday June 27, 2005

Page 67: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Equilibrium of High-Beta Plasmas in W7-AS

M.C. Zarnstorff 1, A. Weller

2, J. Geiger

2, A. Reiman

1, A. Dinklage

2, J.P. Knauer

2, L.-P. Ku

1,

D. Monticello1, and the W7–AS Team

2 and NBI-Group

2

1Princeton Plasma Physics Laboratory, Princeton, NJ 08543, USA

2Max-Planck-Institut für Plasmaphysik, Euratom Assoc., D-17491 Greifswald, Germany

Quasi-stationary, MHD-quiescent discharges with volume-averaged d-values up to 3.5%

were sustained in the W7-AS for more than 100 energy confinement times. The achieved d

appears to be limited by confinement, but is sensitive to the magnetic configuration,

including the rotational transform, the vertical field, and perturbations from the divertor

control coils. A stability limit was not observed. The achieved d is much higher than the

observed or calculated threshold for n =1 and 2 ideal-MHD instabilities. Experimentally,

these instabilities typically saturate and do not impede access to higher d"values. The plasma

equilibrium is reconstructed, fitting the magnetic diagnostic measurements and the Thomson-

scattering pressure profile. Principal component analysis indicates that the available

magnetic diagnostics are sensitive to two moments of the current profile and three moments

of the pressure profile. The total plasma toroidal current is nulled using a feedback controlled

ohmic current. The reconstructed equilibria show small local toroidal net-current, from the

combindation of the ohmic, bootstrap and beam currents, which can reduce the central

rotational transform by ~0.1 . Analysis of the free-boundary equilibria by PIES indicates that

the magnetic field near the plasma edge becomes increasingly stochastic as d increases. The

achieved maximum d-value in the configurations examined corresponds to a calculated loss

of the outer ~35% of the minor radius to islands and stochastic fields. Thus, the d-limit and

its variation may be due to confinement degradation due to flux-surface break-up. The

parametric variation of the calculated equilibrium properties and estimates of the expected

transport increase will be discussed.

P-1.062, Monday June 27, 2005

Page 68: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Resonant kinetic ballooning modes inburning plasma∗

N. N. Gorelenkov

Princeton Plasma Physics LaboratoryP.O. Box 451, Princeton, NJ 08543-0451

ABSTRACT

The kinetic ballooning modes (KBMs) in a proposed ITER burning plasmaexperiment are investigated. Nominal normal shear plasma with central ion tem-peratureTi0 = 20keV is considered with fusion alpha particle beta in the centerβα0 = 0.9%. With the use of the fully kinetic local ballooning code HINST itisfound that KBMs are stabilized in ITER by the plasma shaping attotal plasmabeta in the center 7%. Resonant interaction of KBMs (resonant KBMs) with fu-sion alpha particles results in instability with large growthrate and toroidal modenumbern 20. RKBMs are unstable in a relatively narrow radial domain near themaximum of alpha particle equilibrium pressure gradient atthe half of the mi-nor radius. RKBM growth rate strongly depends on the alpha particle population.Mode frequency is close to the thermal ion drift frequency. The combined kineticeffect of trapped electron dynamics and finite thermal ion Larmor radii is includedin simulations and has a strong stabilizing effect on the ballooning modes. Inpresent day experiments modes in the same frequency range have been observedin DIII-D [1] and were called beta-induced Alfvén eigenmodes. Analyses indi-cates that BAEs can be identified as rKBMs [2]. Properties of these instabilitiesare investigated for the ITER-like burning plasma.

References

[1] W. W. Heidbrink, E. J.Strait, M. S.Chu, and A. D. Turnbull, Phys. Rev. Lett.71, (1993) 855.

[2] N.N. Gorelenkov, W. W. Heidbrink, Nucl.Fusion42 (2002) 150.

∗This work is supported by US DoE contract DE-AC02-76CH03073

P-1.063, Monday June 27, 2005

Page 69: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Transient CHI Solenoid-free Plasma Startup in NSTX*

R. Raman1, M.G. Bell2, T.R. Jarboe1, D. Mueller2, B.A. Nelson1, J. Menard2

and the NSTX Research Team

1. University of Washington, Seattle, WA, USA2. Princeton Plasma Physics Laboratory, Princeton, NJ, USA

Elimination of the central solenoid is a consideration for the design of toroidal

confinement devices which will then require alternative methods for initiating the plasma

current. A new method of non-inductive startup, referred to as transient coaxial helicity

injection (CHI), has been successfully developed on the HIT-II experiment to produce

100kA of closed-flux toroidal current [1,2]. In this method a plasma current is rapidly

produced by discharging a capacitor bank between coaxial electrodes in the presence of

toroidal and poloidal magnetic fields. The initial poloidal field configuration is chosen such

that the plasma rapidly expands into the chamber. When the injected current is rapidly

decreased, magnetic reconnection occurs near the injection electrodes, with the toroidal

plasma current forming closed flux surfaces. An initial test of this method was conducted

on NSTX during 2004. Toroidal plasma currents up to 140 kA were produced for injector

currents of only 4.4 kA, representing a multiplication factor over 30. However, an

unambiguous demonstration of closed flux beyond the end of the injection pulse was not

achieved because the electron temperature, measured by Thomson scattering to be about

16eV peak, was too low for the L/R decay time of the toroidal plasma current to exceed the

RC decay time of the injector current. Three areas for improvement have been identified:

(1) doubling the injector voltage to 2kV and improving the gas preionization to allow

breakdown at lower gas pressure, thereby increasing the overall energy input per particle,

(2) reducing the separation of the injector flux footprints on the electrodes to promote

reconnection and detachment of the plasma, and (3) improving equilibrium control of the

evolving discharge. In the forthcoming experiments on NSTX, preionization will be

improved by injecting both neutral gas and 10kW of 18GHz ECH power into the chamber

below the lower divertor plates which are used as the injector electrodes. Results from these

new experiments will be reported.

[1] Raman, R., Jarboe, T.R., Nelson, B.A., et al., Phys. Rev. Lett., 90, 075005-1 (2003)[2] Jarboe, T.R., Hamp, W.T., Izzo, V.A., et al., Proceedings of the 20th IAEA Fusion

Energy Conference, Vilamoura, Portugal, IAEA-IC/P 42 (2004)

*Work supported by US DOE contracts DE-FG03-99ER54519 and DE-AC02-76CH03073.

P-1.064, Monday June 27, 2005

Page 70: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Investigations of disruption on the HL-2A tokamak

Qingwei Yang, Xuantong Ding, Zhongbing Shi, Yudong Pan, and HL-2A team

Southwestern Institute of Physics, P.O. Box 432, Chengdu SICHUAN 610041, China

Major disruption is a serious problem for tokamak operation. When the major disruption

occurs, it can not only generate great heat loads on the first wall (and divertor plates) and high

voltage on the devise, but also leads to the large electromagnetic force because of the halo

current. Therefore, how to predict, control and mitigate the major disruption is an important

issue on the tokamak physics studies. In this paper, the characters of major disruption on

HL-2A Ohmic plasma are presented. Furthermore, the prediction methods of the major

disruption are investigated as well.

The HL-2A tokamak (with major radius of R = 1.65m and minor radius of a = 0.4m) has

a close, symmetric and double-null divertor. It is operated in the parameters of plasma current

IP à 200~300kA, toroidal field BT à 2.2T and discharge duration k à 1.0s in the disruption

experiments.

In the HL-2A experiments, the major disruption is always led to by the low-q discharge,

mode locking and MHD instability, displacement events and the high density operation. To

understand and predict the disruptions, the Hugill diagram is utilized to describe the discharge

regimes. Usually, the disruption occurs when limitation boundary, for example, the low-q

limit and the Greenwald limit, are approach or exceed.

In the case of high density operation, the density limit disruptions always undergoes two

stages. In the first stage, the soft X ray emission decreases, and the profile of electron

temperature begins to shrink and collapse. In the s

lose, and the plasma current quenches. Sometimes

kink-like plasma radiation appears in the central

region of plasma. The contraction of plasma

channel maybe plays a key role in major

disruptions. In the low-q discharge, the fast

growth of the MHD instability being considered

that it is the main reason of the disruption. The

disruption what caused by the large Mirnov

perturbations and plasma displacement events

are investigated as well.

econd stage, the huge of energy begins to

m = 1 kink-like radiation

r, mm

395

475

-382472

t, ms

Fig. Kink-like formation radiation

P-1.065, Monday June 27, 2005

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A study is conducted on the active stabilization of resistive wall modes (RWM) in the

toroidal geometry. For the sake of analytical simplicity, a toroidal shell model is

employed for the description of a tokamak plasma. A tractable form of RWM dispersion

relation is derived in the presence of a set of discrete feedback coil currents which is

modeled by a surface current density. The mode coupling arises as a consequence of the

discreteness of the feedback currents. The impact of the mode coupling on the

controllability of the RWM is investigated. The formalism is then applied to the

proposed KSTAR plasmas to evaluate the maximum plasma beta and feedback current

requirements for the FEC/RWM coil system of the KSTAR device.

P-1.066, Monday June 27, 2005

Page 72: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Design of Optimal Plasma Position and Shape Controller for KSTAR

Y.M. Jeon, Y.S. Park, Y.S. Hwang

Nuclear Plasma Experiment Laboratory, Seoul National University, Seoul, Korea

Non-rigid plasma equilibrium response model, which can predict perturbed responses of

plasma equilibrium with conducting structures by external magnetic perturbations, is

developed and applied to the design of optimal plasma position and shape controller for

KSTAR. Plasma equilibrium response model for KSTAR (KPERM) is formulated by

coordinating a perturbed Grad-Shafranov equation and perturbed plasma evolution

equations based on reference equilibrium [1]. KPERM is validated with nonlinear MHD

evolution models for vertical growth rate estimations and vertical displacement control

simulations. In addition, a methodology to identify eddy current spectrums in real-time

is introduced, which is a crucial kernel to directly relate magnetic measurements to

perturbed plasma responses. Instead of Fourier spectrum analysis, coefficients of a

simple distribution functional for eddy currents are mapped to match the magnetic

perturbations of magnetic probes and flux loops. Plasma shape control algorithm is

greatly improved to be reliable by incorporating identified eddy current contributions to

the shape identification and prediction matrix. Designed optimal plasma position and

shape controller using a linear quadratic regulator (LQR) technique with the real-time

eddy current identification methodology shows enhanced control performance through

the improvement of performance index, especially Q matrix, and confirms the flexibility

of KPERM model for the design of KSTAR plasma control system.

[1] Y. M. Jeon, J. K. Park, Y. S. Park, and Y. S. Hwang, “Development of Plasma

Equilibrium Response Model for Optimized Plasma Control of KSTAR tokamak”,

Bulletin of the American Physical Society 49, 8 Savannah, Georgia, USA, November

2004

P-1.067, Monday June 27, 2005

Page 73: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Feedback Stabilization of Resistive Wall Modes in DIII-D*

E.J. Strait1, J. Bialek2, M.S. Chu1, A.M. Garofalo2, G.L. Jackson1, R.J. La Haye1,G.A. Navratil2, M.Okabayashi3, H. Reimerdes2, and J.T. Scoville1

1General Atomics, P.O. Box 85608, San Diego, California, USA2Columbia University, New York, New York, USA

3Princeton Plasma Physics Laboratory, Princeton, New Jersey, USA

Advanced tokamak scenarios often require beta values in the regime where ideal MHDkink instabilities are wall-stabilized. To sustain such discharges with a resistive wall requireseither rapid rotation of the plasma [A.M. Garofalo, Phys. Rev. Lett. 89, 35001 (2002)] ordirect feedback control of the slowly growing resistive wall mode (RWM). Experiments inDIII-D have investigated feedback stabilization with external control coils, and more recentlywith internal control coils [E.J. Strait, Phys. Plasmas 11, 2505 (2004)] that have less couplingto the wall and better matching to the helical mode structure. Feedback stabilization has beendemonstrated at higher beta and lower rotation than was possible with the external coils.DIII-D results will be compared to modeling from the MARS-F code with combined rotationand feedback control.

The performance of the feedback system depends strongly on the characteristics of thecoils and sensors. Analytic modeling can explain the qualitative effects of external vs.internal coils, radial field vs. poloidal field sensors, and of decoupling the sensors from thecontrol coils. For example, the performance of external control coils can be improved bypoloidal field sensors without direct coupling to the coils, consistent with DIII-Dexperimental results. Nonlinear effects can also improve feedback control under someconditions: VALEN modeling shows that with saturation of the coil current, the feedbacksystem can continue to stabilize the RWM in regimes where linear models would predictinstability, again consistent with DIII-D results.

Feedback modeling also predicts that the maximum stable beta without rotation can beincreased, and the stable range of operation at lower beta widened, by improving thebandwidth of the amplifier. This year, a prototype system of high-bandwidth audio amplifiersfor the internal coils has successfully stabilized the RWM. These initial results will becompared to modeling predictions.*Work supported by the US Department of Energy under DE-FC02-04ER54698, DE-FG02-89ER53297, andDE-AC02-76CH03073.

P-1.068, Monday June 27, 2005

Page 74: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Control of DIII-D Advanced Tokamak Discharges*

J.R. Ferron1, T.A. Casper2, E.J. Doyle3, A.M. Garofalo4, P. Gohil1, C.M. Greenfield1,A.W. Hyatt1, R.J. Jayakumar2, C. Kessel5, J.Y. Kim6, R.J. La Haye1, J. Lohr1,

T.C. Luce1, M.A. Makowski2, D. Mazon7, J. Menard5, M. Murakami8, C.C. Petty1,P.A. Politzer1, R. Prater1, T.S. Taylor1, and M.R. Wade8

1General Atomics, P.O. Box 85608, San Diego, California 92186-5608, USA2Lawrence Livermore National Laboratory, Livermore, California, USA3University of California, Los Angeles, California, USA4Columbia University, New York, New York, USA5Princeton Plasma Physics Laboratory, Princeton, New Jersey, USA6Korea Basic Sciences Institute, Daejeon, South Korea7Association Euratom-CEA, CEA-Cadarache, St Paul lez Durance, France8Oak Ridge National Laboratory, Oak Ridge, Tennessee, USA

A key goal in control of an advanced tokamak (AT) discharge is to maintain safety factor(q) and pressure profiles that are consistent with both MHD stability at high beta and a highfraction of the self-generated bootstrap current. This will enable noninductive sustainment of100% of the plasma current, as has been demonstrated at high beta (β = 3.6%, βΝ = 3.4) inDIII-D for up to 1 s [1]. The aim is to create the desired q profile during the dischargeformation and sustain it using electron cyclotron current drive (ECCD), bootstrap current andneutral beam current drive. The time evolution of the q profile during the formation ismodified through feedback control of β. Other techniques for control of the q profile havebeen tested in L-mode cases where the effect of the available gyrotron power is relativelylarge. Control of the time evolution of q(0) during the current ramp-up has been demonstratedusing off-axis ECH to modify the electron temperature and thus the rate of currentpenetration. Avoidance of an increase in q(0) when high power ECCD is applied has beeninvestigated using feedback controlled modification of the rate of increase of the ECCDpower to account for the inductive response on axis. Control of the pressure profile shape isaimed at maintaining broad profiles. Modeling has demonstrated that with qmin > 2, βΝ = 5 ispossible with a sufficiently broad pressure profile, while in the experiment βΝ = 4 with qmin =2 has been achieved with P(0)/⟨P⟩ = 2.3 [2]. Tools for pressure profile control are beingimplemented including real time acquisition of the Te, ne , Ti and rotation profiles andmodification of two neutral beam sources to counter-injection. Simultaneous feedbackcontrol of Te at two spatial locations has been demonstrated using off-axis and on-axis ECH.[1] M. Murakami, et al., “100% Noninductive Operation at High Beta Using Off-Axis ECCD,” submitted to

Nucl. Fusion (2004).[2] J.R. Ferron, et al., “Optimization of DIII-D Advanced Tokamak Discharges With Respect to the Beta

Limit,” to be published in Phys.Plasmas (2005).

*Work supported by the U.S. Department of Energy under DE-FC02-04ER54698, W-7405-ENG-48, DE-FG03-01ER54615, DE-FG02-89ER53297, DE-AC02-76CH03073, and DE-AC05-00OR22725.

P-1.069, Monday June 27, 2005

Page 75: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Operational Enhancements in DIII-D Quiescent H-Mode Plasmas*

T.A. Casper1, K.H. Burrell2, E.J. Doyle3, P. Gohil2, C.J. Lasnier1, A.W. Leonard2,T.H. Osborne2, P.B. Snyder2, D.M. Thomas2, and W.P. West2

1Lawrence Livermore National Laboratory, Livermore, California, USA2General Atomics, P.O. Box 85608, San Diego, California, USA3University of California, Los Angeles, California, USA

In recent experiments performed on DIII-D, we concentrated on extending the operatingrange and improving the overall performance of quiescent H-mode (QH) plasmas. The QH-mode offers an attractive, high-performance operating mode for burning plasmas due to theabsence of pulsed edge-localized-mode-driven losses to the divertor (ELMs). Using counterneutral-beam injection (NBI), we achieve steady plasma conditions with the presence of anedge harmonic oscillation (EHO) replacing the ELMs and providing control of the edgepedestal density. As shown in the figure, by carefully controlling the startup conditions, weare able to access the QH regime directly,without first encountering an extended,detrimental ELMing phase. Employing tri-angularity ramps, we have increased theoperating range of both the pedestal densityand pressure. We include these pedestalconditions in the equilibrium calculationsby incorporation of the self-consistentbootstrap-current. The resulting calculatededge current density is consistent with mea-surements from the lithium beam Zeemanpolarimetry diagnostic. Previously, we had

PNBI (MW)

10

5

ne (1020m-3)

0.4

0.2

Fueling (torr-l/s)

100

50

0

121397

106919

Ha(1015/s)

2

1

0 1 2 3 4

ELMs

Shot 121397 enters QH phase without ELMs as compared with an

earlier QH shot, 106919, ELMing prior to the QH phase.

Time (s)

121397 QH phase

106919QH phase

observed that injection of electron cyclotron (EC) power in the core region provides an abilityto control density profile peaking. Using a combination of EC injection for density profilecontrol and NBI ramps, we increased the overall stored energy achieving βN ~ 3. Thiscombination of EC and NBI also modifies the q profile and achieves a long duration (~3 s)where the on-axis value of q remains stationary and near 1.5. QH-mode plasmas remainmarkedly resilient to changes in auxiliary heating power where up to 3 MW of EC power and15 MW of NBI have been injected without loss of the desirable pedestal conditions. Weinclude these pedestal conditions in the equilibrium calculations by incorporation of the self-consistent bootstrap-current. The resulting calculated edge current density is consistent withmeasurements from the lithium beam Zeeman polarimetry diagnostic. We will discuss detailsof experiments on DIII-D that lead to an expanded range of operation.*Work supported by the US DOE under W-7405-ENG-48, DE-FC02-04ER54698, and DE-FG03-01ER54615.

P-1.070, Monday June 27, 2005

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Fueling Requirements for Advanced Tokamak operation*

Roger Raman

University of Washington, Seattle, WA, USA

Steady-state Advanced Tokamak (AT) scenarios rely on optimized density and

pressure profiles to maximize the bootstrap current fraction. Under this mode of operation,

the fueling system must deposit small amounts of fuel where it is needed, and as often as

needed, so as to compensate for fuel losses, but not to adversely alter the established

density and pressure profiles. Conventional fueling methods have not demonstrated

successful fueling of AT-type discharges and may be incapable of deep fueling long pulse

ELM-free discharges in ITER. Compact Toroid (CT) fueling has the potential to meet these

needs, while simultaneously providing a source of toroidal momentum input.

A fueling system that provides a source of toroidal momentum input, while fueling

the discharge as needed for maintaining plasma stability limits and current drive would

increase the operational window of ITER. The requirements for advanced fueling are

particularly well suited for a CT injection system. A CT is a plasmoid with embedded

magnetic fields. It is a robust structure capable of withstanding large acceleration forces [1].

A fueling system based on CTs would inject on the order of about 5x1021 particles per

second at a velocity of about 300 km/s to provide the required core fueling. The resulting

particle inventory perturbation would be about 0.3% per pulse. For a tangentially mounted

CT injector, the imparted toroidal momentum to the reactor plasma would be the same as

that provided by a 500 keV, 40 MW neutral beam system. Such a neutral beam, however,

would provide only 2x1020 particles per second for fueling. CT systems are also fully

electrical, with the only moving part being the high reliability gas valve. Electrical systems

are generally more reliable than mechanical systems. In addition, in a CT injector, because

of the electrical nature of the injector, it is relatively easy to alter the fuel mass and

deposition location. Altering the accelerator voltage alters the CT kinetic energy density,

thereby changing the depth of penetration and the fuel deposition location. Changing the

amount of gas puffed into the injector region alters the mass of the CT. The injector pulse

recycle time can be as short as several tens of ms, resulting in an operating frequency

capability of over 100 Hz. Because of the electrical nature of the injector, it would be

possible to alter the CT mass and velocity on the tens of ms time scale, giving the reactor

fuel control system full feedback control capability of the density profile, while imparting

toroidal momentum. The physics of CT injection, experimental and theoretical progress to

date, and a conceptual CT injector design for ITER will be described.[1] L.J. Perkins, S.K. Ho, J.H. Hammer, Nucl. Fusion 28, 1365 (1988).

*Work supported by US DOE grant DE-FG02-04ER54779.

P-1.071, Monday June 27, 2005

Page 77: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Initial exploration of the density limit in the MST RFP

M. D. Wyman1, B. E. Chapman1, D. L. Brower2, S. K. Combs3, B. H. Deng2, W. X. Ding2, D.

T. Fehling3, C. R. Foust3, S. P. Oliva1, S. C. Prager1

1University of Wisconsin-Madison, Madison, Wisconsin USA

2University of California-Los Angeles, Los Angeles, California USA

3Oak Ridge National Laboratory, Oak Ridge, Tennessee USA

The density limit and its underlying physics in modern, larger-scale RFP plasmas has

only begun to be explored. Establishing the density limit is important in part since there are

as yet few known fundamental operational limits in the RFP. In tokamak plasmas without

pellet injection, the central line-averaged electron density, <ne>, is generally limited to the

Greenwald value, nG = Ip/!a2. This limit applies as well to plasmas in the RFX RFP, and we

observe it to play at least some role in MST plasmas. This tokamak-RFP commonality

suggests that perhaps additional light can be shed on the physics of the density limit by

measurements made in the RFP.

We report here initial measurements made in the MST, which produces toroidal

deuterium RFP plasmas with major and minor radii of 1.5 m and 0.51 m, respectively. By

injecting deuterium pellets into standard, low-confinement plasmas, one is easily able to

exceed nG, but not without consequences. As <ne> exceeds nG, the toroidal plasma current

begins to ramp down. If <ne> is sustained above nG for a sufficiently long time, the discharge

terminates. If <ne> drops back below nG, the plasma current can recover. This decay of the

current with pellet injection is in contrast to what has been achieved in at least some tokamak

plasmas, where little or no decay of the current is observed. The difference may be linked to

the relatively rapid global particle transport time scale (1 ms) in MST standard plasmas which

allows the quick transfer of core-deposited particles to the edge.

Without pellet injection, some low current (Ip < 200 kA) MST plasmas are observed

with <ne> apparently exceeding nG for the duration of the discharge. These discharges exhibit

a flattop in the plasma current, as usual, but the discharge length is substantially shorter than

usual. During the first and last few milliseconds of more normal discharges, when the current

is ramping up and ramping down, respectively, <ne> routinely exceeds nG. We are working to

understand this phenomenology.

This work was supported by the U.S. Department of Energy.

P-1.072, Monday June 27, 2005

Page 78: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

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P-1.073, Monday June 27, 2005

Page 79: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Fast optical spectrometer for the charge exchange diagnostic on RFX-mod

E. Gazza1, M. Valisa1, L. Carraro1, M.E. Puiatti1, P. Scarin1, B. Zaniol1

1 Consorzio RFX, Associazione Euratom-ENEA sulla fusione, 35127 Padova, Italy

To face the need of better space and time resolved measurements of the relevant plasma

parameters a new diagnostic neutral beam injector (NBI) is to be installed on RFX in spring

2005. The main purpose is to study the radial distribution of the flow fields, of the ion

temperature and of the impurity densities, via the analysis of the charge exchange emission

lines. The detection of charge exchange radiation is known to be a challenging exercise

when diagnostic beam injectors are involved, whose sources have by definition relatively

low ion currents; about 5 A in the case of the RFX NBI. To tackle the problem large

aperture spectrometers are to be used. Further requirements of such spectrometers are good

imaging quality and, in the case of RFX where relatively low temperature and flow

velocities are to be measured, high spectral resolution. High speed, little aberration and

good spectral resolution are colliding requirements and some compromise has to be chosen.

Considering that most of the charge emission lines of interest are in the visible between 450

and 540 nm, at RFX it has been decided to design and build a spectrometer based on a

large, high resolution plane reflecting grating and a long focal length photographic

objective lens. The grating (TYDEX) of 3000 g/mm and a clear aperture of 143 x 180 mm2,

is placed on a high precision rotating turret equipped with a stepper motor. The mount is in

the Littrow configuration. A miniaturized aluminum coated mirror has been used to

minimize the distance between entrance slit and detector on the focal plane so as to

preserve the aperture of the system. The objective lens is a commercial 400 mm f/2.8

NIKON telephoto lens with 83% transmission efficiency (at the 500 nm). The detector is a

bi-dimensional back-illuminated CCD camera (Micromax 512 EBFT- Roper Scientific).

The effective aperture of the spectrometer in the wavelength range of interest is f/3.

Besides the full description of the spectrometer performance, a critical list of the

motivations of the specific choices is given and sample results from the experiment are

presented.

P-1.074, Monday June 27, 2005

Page 80: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Study of Plasma density profiles evolution using the new scanning

interferometer for FTU

C.Mazzotta1, O.Tudisco1, A.Canton2, P.Innocente2, D. Marocco1, P. Micozzi1, G.Monari1,

G.Rocchi1

1Centro Ricerche Energia Frascati, Euratom-ENEA Association, Frascati, Italy

2Consorzio RFX, Euratom-ENEA Association, Padova, Italy

Performance and first results of density profile measurements by a new scanning

interferometer in FTU are described. The diagnostic has been developed by the “Consorzio

RFX” for the Frascati Tokamak Upgrade (FTU).

A resonant tilting mirror placed at the focus of a fixed parabolic one is used to scan the

laser beams within the vertical port. The deflection is cancelled with a second reflection on

the tilting mirror. A CO2 laser (10!W, _=10.6!µm) is used for the measurement, while a CO

laser (1!W, _ =5.4!µm) is used to compensate vibrations. The number of independent line-

average density data depends on the ratio between scan amplitude and beam diameter

(~!1!cm); the number of equivalent chords typically varies from 28 to 34. The wavelength

choice was dictated by the attainment of very high densities (> of 1021!m-3) with multiple

pellet injection. A full profile is scanned in 42 µs. If plasma is steady state within the scan

time, the line-average data can be supposed to be simultaneous and an inversion of the line

integrated profile can be performed.

In this paper we will present results, obtained from the new diagnostic, in different FTU

scenarios where density peaking is an important feature, in particular plasmas with internal

transport barriers obtained with LHCD and ECRH, and PEP regimes sustained by multiple

pellet injection. The comparison with other

diagnostics, as Thomson scattering, will also be

reported. Particular care has been paid for analysis of

pellet fuelled discharges, where very peaked profiles

are obtained. During the penetration of relatively slow

pellets injected from the high field side, in which

density variations are slower than the scan time, the

inversion of the profiles can be performed. Results and

problems of this inversion will be reported. In the figure the 3D evolution of inverted

density profile during a pellet injection is shown.

P-1.075, Monday June 27, 2005

Page 81: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Mirror Test for ITER: Optical Characterisation of Metal Mirrors in

Divertor Tokamaks

G. De Temmerman1, M.J. Rubel

2, J.P. Coad

3, R. Pitts

4, J.R. Drake

2 and P. Oelhafen

1 and

contributors to the JET-EFDA workprogramme*

1Institute of Physics, University of Basel, CH-4056 Basel, Switzerland

2Alfvén Laboratory, KTH, Association EURATOM – VR, 100-44 Stockholm, Sweden

3Culham Science Centre, EURATOM-UKAEA Fusion Association, Oxon OX14 3DB, UK

4Centre de Recherches en Physique des Plasmas, Association EURATOM, Conférédation

Suisse, EPFL, 1015 Lausanne, Switzerland

All optical systems in ITER will be based on first mirrors. The mirrors, as plasma

facing components, may undergo erosion and re-deposition processes eventually influencing

mirror optical properties, i.e. reflectivity, which would have a negative impact on

spectroscopy signals. Therefore, tests of first mirrors have been initiated at several machines,

including JET, where the project is included in the framework of Tritium Retention Studies

(TRS). The choice of JET is related to several unique features of this machine: (a) a large

divertor tokamak with an ITER relevant configuration, (b) plasma pulses of 20 s, (c) carbon

and beryllium environment. A dedicated programme is also carried out at TCV where the

mirrors face a variety of divertor plasma configurations.

The aim of this paper is to give a comprehensive overview of the projects with

particular emphasis on optical characterization, an essential step in the qualification of

mirror components. At JET, molybdenum and stainless steel mirrors (flat front and angled at

45o) have been manufactured and installed in cassettes of pan-pipe shape placed in several

locations of interest to ITER: two on the main chamber wall and three in the divertor (inner,

outer and base). At TCV, a number of mirrors have been installed on a specially designed

manipulator operated from the bottom section of the torus. In all cases the installation was

preceded by very detailed optical studies: total, specular, diffuse reflectivity of all mirrors

was measured by means of a UV-Vis-NIR spectrophotometer and a spectroscopic

ellipsometer. The results of experiments at TCV and optical characterisation of mirrors for

the exposure at JET will be presented in detail.

*See the Appendix of J. Pamela et al., Fusion Energy 2004 (Proc. 20th Int. Fusion Energy

Conference, Vilamoura, 2004), IAEA, Vienna (2004).

P-1.076, Monday June 27, 2005

Page 82: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

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P-1.077, Monday June 27, 2005

Page 83: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Neutron energy measurements of Trace Tritium plasmas

with NE213 compact spectrometer at JET

L.Bertalot1, B Esposito

1, S.Conroy

2, P. Lamalle

3, A.Murari

4, S.Popovichev

5, M.Reginatto

6,

H.Schuhmacher6, A.Zimbal

6 and contributors to the EFDA-JET workprogramme **

1 Associazione EURATOM-ENEA Fusione, v. E. Fermi 45, I-00044 Frascati, Italy

2 Department of Neutron Research, Uppsala University, BOX 525, SE-75120 Uppsala,

Sweden

3 LPP-ERM/KMS, Association EURATOM-Belgium State, Brussels, Belgium

4 Consorzio RFX Assoc. EURATOM ENEA Fusione, Corso Stati Uniti 4, I-35127 Padova,

Italy

5 Euratom/UKAEA Fusion Assoc., Culham Science Centre, Abingdon, OX14 3DB, UK

6 Physikalisch-Technische Bundesanstalt, Bundesallee 100, D-38116 Braunschweig,

Germany

Properties of the energy distribution functions of thermal- and high energy- fuel ions in

tokamak plasmas can be obtained by measuring the energy spectra of the neutron emission.

The Trace Tritium Experimental (TTE) campaign aimed mainly to particle transport studies.

A compact broadband neutron spectrometer, fully characterized for neutron detection (1.5

MeV <En< 20 MeV) at the Physikalisch-Technische Bundesanstalt accelerator facility,

based on a liquid scintillator (NE213) with neutron/gamma discrimination features was

operated successfully during TTE [1,2] with good energy resolution ("FE/E<4% at En =2.5

MeV and FE/E <2% at En =14 MeV). Pulse height spectra of the neutron emission from

different TTE plasma scenarios with Neutral Beam (NB), RadioFrequency (RF) and

combined NB+ RF heating schemes were acquired. Simultaneous spectral acquisition of the

DD (at 2.5 MeV) and DT (at 14 MeV) emissions was performed, due to the broadband

energy feature of the spectrometer. The present paper will report on the comparison between

the TTE measured neutron spectra, obtained with the MAXED unfolding code, and

theoretical spectra evaluated by means of the FPS Monte Carlo kinematics code which takes

into account the various parameters of the investigated plasma scenarios. Aim of this

analysis is the determination of the energy distribution functions of the deuterons and tritons

and their dependence from the different heating scenarios, in order to distinguish the

contribution of the suprathermal ion component of the neutron emissions. Particular

attention will be devoted to the possible RF effects on the deuteron population in combined

heating TTE ELMy H mode plasmas as indicated by recent PION analysis.

[1] A. Zimbal et al., Rev. Sci.Instrum. 75 (2004) 3553 [2] B. Esposito et al., Rev. Sci.Instrum. 75 (2004)

3550

**See Appendix of J. Pamela et al., Fusion Energy 2004 (Proc. 20th Int. Conf. Vilamoura 2004),

IAEA, Vienna 2004).

P-1.078, Monday June 27, 2005

Page 84: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Development of new neutron emission spectrometry diagnostics for fusion experiments at JET

J.Källne, S.Conroy, G.Ericsson, M.Gatu Johnson, L.Giacomelli, G.Gorini1), C.Hellesen,

A.Hjalmarsson, A.Murari2), S.Popovichev3), E.Ronchi, E.Sanden Andersson, H.Sjöstrand, J.Sousa4), M.Tardocchi1), J.Thun, M.Weiszflog, and contributors to the EFDA-

JET workprogramme*

INF, Uppsala University, EURATOM-VR Association, Uppsala, Sweden 1) Istituto di Fisica del Plasma, EURATOM-ENEA-CNR Association, Milan, Italy

2) EURATOM-ENEA-CNR Association, Padova, Italy 3)JET, Culham Science Centre, ABINGDON, UK, EURATOM-UKAEA Association

4) Associação EURATOM/IST, Centro de Fusão Nuclear, Instituto Superior Técnico, Av. Rovisco Pais 1, 1049-001 Lisboa, Portugal

Fusion experiments will be possible at the start-up of JET later this year with access to two new neutron spectrometers. One is the upgrade of the magnetic proton recoil (MPR) spectrometer, which has earlier been used for 14-MeV dt neutron measurements during the DTE1 campaign of 1997. These represented a break-through in neutron emission spectroscopy (NES), partly, because of the possibility to operate at high count rates (up to 0.7 MHz achieved). The MPR played also an important role in the TTE campaign of 2003 as it then was used as a NES control room diagnostic for the first time. The upgraded instrument, MPRu, can also be used for 2.5-MeV dd neutrons.

The other instrument is a neutron time-of- flight (TOF) spectrometer, which has been designed for optimized rate (TOFOR). With an expected TOFOR maximum count rate of about 0.4 MHz, JET will have equipment for fusion experiments with NES diagnostics that has never existed before. TOFOR and MPRu will view the plasma along direction perpendicular and semi-tangentially, respectively. This will give increased ability to use non-isotropies in the neutron emission arising from fast ion velocity components generated by NB and ICRH power injection. In this contribution we will describe the key features in the technical design of MPRu and TOFOR. Especially the rate handling capability of TOFOR will be discussed with reference to extensive component tests performed before installation on JET. Also the projected background rejection capability of the new MPRu focal plane hodoscope will be assessed as based on tests of pulse shape response to different kinds of radiation. Finally the NES diagnostic capabilities afforded by the new instruments will be highlighted in the context of the envisaged research programme on JET in the coming experimental campaigns. This includes enhanced NB injection power and ICRH power with the new ITER-like antenna. Of particular interest, in this context, is the use of NES for direct measurement of the fusion performance of the antenna in terms of the neutron yield rates attained and its distribution on thermal and supra-thermal ion reactions. Another aspect is that, since advanced NES measurements can be performed in D plasmas, there will be ample opportunities for developing NES diagnostic methods enhanced with the added capability offered by dual sight lines. Thus, the use of the TOFOR and MPRu spectrometers to observe D plasmas is the best opportunity that will exist for the coming few years to explore and develop NES for its central role in burning plasma fusion experiments as will be the mission of ITER. *See the Appendix of J.Pamela et al., Fusion Energy 2004 (Proc. 20th Int. Conf. Vilamoura, 2004) IAEA, Vienna (2004).

P-1.079, Monday June 27, 2005

Page 85: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Diagnosis of high-energy fuel ions on ITER with neutron emission spectroscopy (NES): Monte Carlo calculations based on NES measurements on JET DT plasmas

L.Ballabio1), S.Conroy2), G.Ericsson2), M.Gatu Johnson2), L.Giacomelli2), W.Glasser2), G.Gorini1), A.Hjalmarsson2), J.Källne2), A.Murari3), E.Sanden Andersson2), H.Sjöstrand2),

M.Tardocchi1), M.Weiszflog2), and contributors to the EFDA-JET workprogramme*

1) Istituto di Fisica del Plasma, EURATOM-ENEA-CNR Association, Milan, Italy

2) INF, Uppsala University, EURATOM-VR Association, Uppsala, Sweden 3) EURATOM-ENEA-CNR Association, Padova, Italy

High-energy fuel ion (HEFI) populations are created in plasmas subjected to neutral

beam (NB) injection and ion cyclotron resonance heating (ICRH). These heating schemes were tested at JET in the DTE1 and TTE campaigns and diagnosed with the help of neutron emission spectroscopy (NES) [1]. The JET experience can be transferred to burning plasma studies on ITER but for suitable rescaling including difference in plasma size and conditions, and machine operating parameters. This means, e.g., that ITER NB will use 1-MeV tangential D beams (compared with 0.15 MeV at oblique angle in JET) leading to d deposition into circulating orbits with a pitch angle about 30°. ICRH will use different heating schemes including some resonating with T or D as tested on JET and shown to produce HEFI populations with pitch angle ˜90° and “tail” temperatures T⊥=100 keV depending on power density conditions. Another source of HEFI populations of both d and t of up to 3 MeV is α+d and α+t knock-on collisions, which give rise to a so-called alpha knock-on neutron (AKN) signature in the emission spectrum. With ITER temperatures in the 20-keV range, the AKN would make about 10-3 of the total emission compared to 10-5 for JET.

NB, ICRH and AKN induced HEFI components have all been the object of NES measurements on JET and paradigms have been worked out for the analysis/interpretation of the data and projection to ITER; this includes a “bulk” (B) component mostly due to thermal fuel ions. Synergies between NB and ICRH have also been observed but are not considered here. The relative intensities of the HEFI components depend on plasma and heating conditions and were often found to dominate at JET for both NB and ICRH in high performance discharges. This is different from high performance ITER H-mode, which is estimated to be 99% thermal, or, Qth/Q=0.99. Lower performance and transient conditions would give higher HEFI fractions and lower Qth/Q ratio; not to forget, the experiments with lower Qth/Q will have to pave the way to reach and optimize high performance conditions.

NES diagnostics benefit from optimised separation of the signatures in the neutron spectrum. This has been studied for the ITER conditions in new Monte Carlo calculations of the neutron emission spectrum for different heating scenarios. It is found that the AKN, NB and ICRH signatures can be distinguished under most conditions, especially because of the strong anisotropy of the NB and ICRH components. An interesting aspect in this context is the dual-sight line measurements now planned with the new JET instrumentation. A similar sight line arrangement can be considered for ITER. Results from the simulation studies will be presented including the diagnostic implications for burning plasma experiments. [1] See e.g. S.Conroy et al, this conference. *See the Appendix of J.Pamela et al., Fusion Energy 2004 (Proc. 20th Int. Conf. Vilamoura, 2004) IAEA, Vienna (2004).

P-1.080, Monday June 27, 2005

Page 86: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

MPR neutron emission spectroscopy of fast tritons from (T)D ion cyclotron heating in JET plasmas

S.Conroy2), G.Ericsson2), M. Gatu Johnson2), L.Giacomelli2), W.Glasser2), G.Gorini1), A.Hjalmarsson2), T. Johnson4), J.Källne2), P.U.Lamalle3), H.Sjöstrand2), E.Sundén

Andersson2), M.Tardocchi1), M.Weiszflog2), and contributors to the EFDA-JET workprogramme*

1) Istituto di Fisica del Plasma, EURATOM-ENEA-CNR Association, Milan, Italy 2) INF, Uppsala University, EURATOM-VR Association, Uppsala, Sweden 3) LPP-ERM/KMS, Association EURATOM-Belgian State, partner in TEC, Brussels,Belgium 4) Alfven Laboratory, KTH, Euratom-VR Association, Sweden

The 2003 JET Trace Tritium campaign provided an opportunity for ion cyclotron resonance heating (ICRH) of tritium at low concentrations in deuterium plasmas. A favourable heating scheme in these conditions is (T)D where triton are accelerated at their fundamental cyclotron frequency. Up to 1.5 MW ICRH was coupled to the plasma, mostly by direct deposition on the puffed tritium by cyclotron damping. The resulting neutron yield is dominated by supra-thermal emission from energetic tritons, which provides good conditions for testing of models used in the analysis of neutron emission spectroscopy (NES) from ICRH plasmas. NES measurements were carried out with the Magnetic Proton Recoil (MPR) neutron spectrometer viewing the plasma horizontally at an angle of about 47° relative to the magnetic field on-axis. The viewing volume is representative of the core plasma where fast triton reactions with deuterons take place. For the NES analysis it is assumed that plasma conditions are uniform within this volume and the neutron emission has two contributions referred to as “bulk” (B) and “high energy” (HE). The bulk contribution is described by reactions between thermal deuterons of temperature Td and tritons with an effective temperature TB; the resulting neutron energy spectrum shape is nearly Gaussian and is exactly so if TB=Td as is sometimes assumed for the analysis. This NES component is relatively weak for (T)D heating, its intensity IB being typically 20% of the intensity IHE. The latter is modelled by reactions between thermal deuterons and a “cut Maxwellian” triton distribution in velocity space consisting of Maxwellian tritons of temperature THE, with velocities in the angular range 90°±10° relative to the magnetic field. This is a simple model prescription providing a strongly anisotropic distribution that can be used for routine best- fit analysis of ICRH neutron spectra to provide “tail temperature” values THE. The extent to which the THE values are model dependent is addressed here by comparing the “cut Maxwellian” results with a two-temperature (bi-) Maxwellian model featuring parallel (T//) and perpendicular (T⊥) temperatures with T⊥>>T//. This model is strongly anisotropic and frequently used for ICRH theory. Detailed comparison with the results of Fokker-Plank simulations using the SELFO code is also underway. Comparison of the models shows that the slope (in log scale) of the high energy tail of the neutron spectrum is directly related to T⊥ and only weakly dependent on the T// value. T⊥=THE provides good fit to the NES spectrum for reasonable values of T//. The two fitted spectra differ only near the peak. This has some effect on the IB values being generally higher with bi-Maxwellian model. It is thus found that the high energy tail of the spectrum can be used to determine THE and IHE almost model independently, while extraction of IB and, especially, TB are more sensitive to model assumptions.

Another aspect of the NES measurements concerns toroidal rotation. Experiments during TTE revealed that ICRH directive antenna phasing (co- and counter-current) resulted in (±300 km/s) triton velocities; this is another model- independent parameter accessible to NES diagnosis. The above results will be discussed in the perspective of relevance for advanced NES instrumentation on ITER, where IHE will be typically 1-2 orders of magnitude smaller than IB and yet will remain an essential source of information about fast fuel ions [1]. [1] L. Ballabio et al, this conference. *See the Appendix of J.Pamela et al., Fusion Energy 2004 (Proc. 20th Int. Conf. Vilamoura, 2004) IAEA, Vienna (2004).

P-1.081, Monday June 27, 2005

Page 87: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

New method to calculate the Gaunt factor for the refinement ofZeff evaluation in fusion plasmas

V.Stancalie1 and contributors to the EFDA-JET workprogramme*

1Laser Department, National Institute for Laser, Plasma and Radiation Physics,

P.O.Box MG-36, Bucharest, 077125 ROMANIA, Association EURATOM MEdC

The present paper’ s main concern is with ensuring the completeness of the contribution

of the non-fully stripped ions to Zeff and not with the consequential modeling of this

quantity. The key issue then is the starting point of uncertainties in the fundamental

component Gaunt factor.

Our proposal is to consider the possibility of using the Coulomb Green’ s function and its

Sturmian representation1, to calculate positions of bound and excited Rydberg states. We

have applied this method for a system with 4 electrons, Be-like C ion, as an example. The

proper description of such system is a pair-coupling scheme. The pair-coupling scheme

requires to include as CIII symmetries: 1Se, 3Pe,5De for J=0e; and 1P0,3S0,3P0,3D0,5P0,5D0,5F0 for

J=10. After recoupling for J=0e and J=10, there are 28 and 72 channels, respectively. The

needed dipole radial matrix elements between hydrogenic and Sturmian wave function have

structure similar to the well-known Gordon formula for dipole matrix elements between

hydrogenic bound states.

Comparisons between effective quantum numbers calculated on the basis of Coulomb-

Green’s function and those reported as output from the R-matrix code2 are shown in Table

1. Finally, bound-free Gaunt factors for 1s22p3/23p → 1s22p3/2 ns series in CIII are given in

Fig.1.

Table1. Effective quantum numbers for 1s22sns(1S0) and 1s22p1/2 np(3P0) states

* See the Appendix of J. Pamela et al., Fusion Energy 2004 (Proc. 20th Int. Conf. Vilamoura, 2004) IAEA,Vienna (2004).

state this work Ref.2.2s 3s (1S0) 4s 5s 10s2p1/25p(3P0) 6p 7p

2.644833.641484.522359.653814.860435.851086.86591

2.66493.64114.56499.64444.86095.86146.8631

Fig.1. Bound-free Gaunt factor for 1s22p3/23p-1s22p3/2 ns series

1M.Poirier, Phys.Rev.A38(1998)3484; 2K.Berrington, J.Pelan, L.Quigley, Phys.Scr.57(1998)549

P-1.082, Monday June 27, 2005

Page 88: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

First study of 2-D spatial distribution of D-D and D-T

neutron emission in JET Elmy H-mode plasmas with

Tritium puff

G. Bonheure1, J. Mlynar

2, L. Bertalot

3, S. Conroy

4, A. Murari

5, S. Popovichev

6,

L. Zabeo6 and EFDA-JET Contributors

1. ERM - KMS, B 1000 Brussels, Belgium, Partner in the Trilateral Euregio Cluster

2. Association EURATOM-IPP.CR, CZ-182 21 Prague 8, Czech Republic

3. Euratom/ENEA Association, Frascati, Italy

4. INF, Uppsala University, EURATOM-VR Association, Uppsala, Sweden

5. Consorzio RFX, Associazione ENEA-Euratom per la Fusione, Padova, Italy

6. Euratom/UKAEA Association, Culham Science Centre, Abingdon, Oxon, UK

JET neutron cameras detect both 2.45 and the 14 MeV neutrons along 19 lines of

sight, 9 vertical and 10 horizontal. In the past this unique diagnostic provided very

useful information about various plasma phenomena but in general the data analysis

was limited to the study of the line-integrated measurements. More recently the

tomographic reconstructions have become more frequently used [1-6] but have never

been applied systematically to the investigation of the fuel mixture. In this paper, the

first two dimensional (2D) spatial distributions of 14 MeV and 2.5 MeV neutrons,

obtained with a tomographic algorithm based on the Minimum Fisher Regularisation,

are reported. From the ratio of these tomographic reconstructions, the 2D spatial

distribution of the tritium concentration n(T)/n(D) is derived for the first time for a set

of 30 ELMy H-mode plasmas from the last Trace Tritium experiments at JET. This

approach is interesting essentially because it does not require modelling of the tritium

source and it does not depend on the beam deposition and the beam slowing down,

reducing significantly the uncertainty in the final estimate of the tritium concentration.

These profiles can be used for particle transport studies and provide for unique 2-D

pictures of tritium puffing. With the described method, asymmetries in the 14 MeV

D-T neutron yield were detected with the vertical camera, showing a higher emission

towards the outboard side of the vacuum vessel, in the phase immediately following

the tritium puff and which can be explained by orbit effects of fast particles[6].

Preliminary observations with the horizontal camera show evidence of further

asymmetries which were not reported before and for which fast particles could

possibly play a role. The impact of these new results on the interpretation of neutron

emission for transport studies will also be discussed.

[1] FB Marcus et al JET internal Report

[2] J.Pamela, IAEA - 20th Fusion Energy Conference, Vilamoura, Portugal, 2004

[3] D. Stork IAEA - 20th Fusion Energy Conference, Vilamoura, Portugal, 2004

[4] J.Pamela, D. Stork.46th Annual Meeting of the APS Division of Plasma Physics

15-19 November

2004, Savannah, Georgia, USA

[5] D. Stork.46th Annual Meeting of the APS Division of Plasma Physics 15-19

November 2004,

Savannah, Georgia, USA

[6] K-D Zastrow et al PPCF 46(2004) B255-B265

P-1.083, Monday June 27, 2005

Page 89: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Absorption experiments on the CASTOR tokamak

M.E. Notkin1, A.I. Livshits

1, M. Hron

2, J. Stockel

2

1 Bonch Bruyevich StateUniversity, Saint Petersburg, Russia

2 Institute of Plasma Physics, Association EURATOM-IPP.CR, Prague, Czech Republic

The present work is undertaken to investigate the extra-equilibrium absorption in

tokamak environment with the purposes to better understand the role of nonmetallic

coatings at plasma facing materials in the D/T inventory and recycling, and to develop a

method of neutral flux diagnostic.

The probability of absorption, c. of low-energy hydrogen particles (~1 to tens of

eV) in metals radically depends on the thickness of a nonmetallic film typically covering

the metal surface. In the case of a monolayer film, this probability is close to that for a

clean metallic surface. But it decreases dramatically, if the film thickness exceeds one

monolayer. In reality, the thickness of nonmetallic coating depends on temperature: at high

enough temperatures, typically only one nonmetallic monolayer exists at the surface, and

the probability of absorption of supra-thermal hydrogen is very high. At lower

temperatures, both monolayer and polyatomic coatings are possible, and, correspondingly,

c may vary over a wide range.

A movable plasma facing absorption probe (AP) of 0.02 mm Nb foil is installed into

the CASTOR tokamak to investigate the dependence of the probability of absorption of

supra-thermal hydrogen particles upon the type and thickness of nonmetallic coating, and

on metal temperature. The AP can be exposed to plasma at various distances, and then can

be moved into a chamber separated from the torus to analyze the amount of absorbed

hydrogen by thermal desorption. The type and thickness of the coating can be controllably

varied in situ.

The absorption of supra-thermal particles (~7 eV Franck-Condon atoms) from

plasma discharge was investigated as a function of AP temperature, of AP distance from

plasma, and of plasma discharge duration. A reliable registration of H atoms was

demonstrated to be possible in spite of a short plasma pulse duration and a relatively high

H2 pressure background. H atom absorption probability was found to weakly depend on

metal temperature. Most of supra-thermal hydrogen atoms are absorbed by the probe during

the start-up phase of discharge.

P-1.084, Monday June 27, 2005

Page 90: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

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P-1.085, Monday June 27, 2005

Page 91: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

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P-1.086, Monday June 27, 2005

Page 92: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Investigation of the Upper Hybrid Resonance

Cross-Polarization Scattering Effect at the FT-2 Tokamak

Altukhov A.B., Esipov L.A., Gurchenko A.D., Gusakov E.Z., Stepanov A.Yu.

Ioffe Institute, St.Petersburg, Russia

Magnetic component of small-scale plasma turbulence can play an important role in

electron transport disturbing the system of nested magnetic surfaces and leading to huge

energy losses along the field lines. The cross-polarization scattering (CPS) diagnostics

utilizing microwave probing perpendicular to the tokamak magnetic field provides a unique

opportunity for measuring relatively low magnetic turbulence level in the hot plasma core

because intensive density fluctuations do not contribute to the CPS signal in this

experimental geometry [1]. The CPS effect was used for diagnostic development on Tore

Supra [2], where the poor localized extraordinary to ordinary mode (X›O) conversion was

studied in the presence of probing wave cut off protecting the O-mode receiving antenna

from the higher level X-mode radiation scattered from the density fluctuations. The

alternative scheme of the experiment utilizing the CPS effect in the Upper Hybrid Resonance

(UHR) of the probing microwave was investigated recently at the FT-1 tokamak, where the

RADAR scheme was used to confirm the UHR origin of the CPS signal [3].

In the present paper the first measurements of the CPS spectra performed at the FT-2

tokamak where a double antenna set was installed at the low magnetic field side in the same

poloidal cross-section, but opposite to the steerable focusing antennae used for UHR

microwave back scattering investigation are reported. The plasma is probed by X-mode from

the high field side and both O-mode and X-mode spectra are studied with the new antennae

set for different values of plasma density, current and probing antenna vertical position.

Dependence of the CPS spectra on the UHR and antenna position is investigated. The

experimental conditions at which the CPS spectrum is most likely associated with the UHR

are determined. The first radial correlation measurements to confirm the UHR origin of the

CPS signal are carried out.

1. Lehner T., Gresillon D., et al. Proc. 12th EPS Conf. on Control Fusion and Plasma

Physics, Budapest, 1985, pt.II, p.664.

2. X.L. Zou, L. Colas, M. Paume et al., Phys. Review Lett., 1995, V.75, p.1090.

3. D.G. Bulyiginskiy, A.D. Gurchenko, E.Z. Gusakov et al., Phys. Plasmas, 2001, V.8,

p.2224.

P-1.087, Monday June 27, 2005

Page 93: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Spatial Resolution of Poloidal Correlation Reflectometry

E.Z.Gusakov, A.Yu.Popov

Ioffe Physico-Technical Institute, St.Petersburg, Russia

Poloidal correlation reflectometry utilizing microwave plasma probing by several poloidally

separated antennae is often used nowadays for plasma rotation diagnostics and turbulence anal-

ysis [1]. The poloidal rotation velocity is determined in this technique from the temporal shift

of the maximum of the cross correlation function of scattering signals in two poloidally sepa-

rated channels. The localization of measurements is based on the assumption that the microwave

scattering off long wave-length fluctuations dominating in the turbulence spectra occurs in the

cut-off layer.

In the present paper the described experimental scheme is analyzed theoretically in the frame

of the 2D WKB approximation valid both in linear and nonlinear regime of scattering off long

scale density fluctuations. The analysis performed in the cylinder geometry accounts for the

plasma curvature effects onto the poor localized forward scattering along the incident wave

trajectory, produced by the long scale turbulence, which is dominant in the fluctuation reflec-

tometry signal [2]. The explicit expressions for both the cross and auto correlation functions

of signals in two poloidally separated channels are obtained for arbitrary profiles of plasma

density, rotation velocity, turbu lence spatial distribution and spectra. The conditions at which

the fluctuation poloidal velocity measurement is possible are determined and its localization is

estimated. The derived explicit expressions are valid in linear and strongly nonlinear regimes of

fluctuation reflectometry. They are convenient for determination of measurement accuracy and

for justification of the reliability of obtained rotation velocity profiles, which is illustrated for

the T-10 tokamak experimental conditions [1].

The deterioration of the diagnostics performance in the nonlinear fluctuation reflectometry

regime, caused by suppression of correlations is discussed for different probing schemes.

References

[1] V.A.Vershkov, S.V.Soldatov, D.A.Shelukhin et al, 30th Conf. Control. Fusion Plasma

Phys. ECA 27A, P-2.56 (2003)

[2] E.Z.Gusakov, B.O.Yakovlev, Plasma Phys. Control Fusion 44, 2525(2002)

P-1.088, Monday June 27, 2005

Page 94: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Fig. 1. Perturbation of poloidal field Bs with m/n=2/1; its

amplitude Amp(Bs) and instantaneous

frequency fs obtained by HSA/EMD.

Hilber t Spectrum Analysis of Mirnov Signals

I.I. Orlovskiy, A.M. Kakurin

Russian Research Center “Kurchatov Institute”, Moscow, Russia

e-mail: [email protected]

Hilbert spectrum analysis (HSA) is rapidly developing technique for analysis of non-

stationary signals. It is based on the Hilbert transform which associates any real data with

corresponding complex (analytical) signal. For such signals phase and instantaneous

frequency are uniquely defined that allows representing analyzed data in a time-frequency

domain with resolution limited only by the sampling rate of the original signal. However,

the results of HSA are meaningful only for monocomponent signals, i.e. for the signals

which contain single oscillation. Since experimental data usually contains various

oscillations, experimental signal should be decomposed to a set of monocomponent signals

which are suitable for further processing by HSA. Such pre-processing is performed by

recently developed empirical mode decomposition algorithm (EMD). The method is fully

complete and adaptive since the decomposition is performed without predetermined basis

and the result depends on the local properties of the signal.

HSA has been applied to experimental signals of magnetic probes (Mirnov signals) in T-10

tokamak. The method provides information on dynamics of MHD perturbations including

instantaneous frequency deviation during the period of oscillation (fig. 1). Such deviation is

associated with the influence of error field on MHD mode. In case of multimode

perturbations the signals are preprocessed by EMD algorithm for mode separating and de-

noising. Combination of HSA and EMD provides qualitatively new and higher level of

investigation of large-scaled MHD perturbations. The method can be also applied to the

signals of any diagnostics whose amplitude and frequency vary in time.

450 460 470 480 490 500

0

1

2

time [ms]

f [

kHz]

0

3

6

Am

p(B

) [1

0-4 T

] -8

0

8

B [

10-4 T

]

P-1.089, Monday June 27, 2005

Page 95: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Mitigation of hydrocarbon film deposition on in-vessel mir rors

K.Yu. Vukolov, A.A. Medvedev, S.N. Zvonkov

RRC “Kurchatov Institute”, 123182 Moscow, Russia

It is well known that erosion and redeposition of plasma-facing components in fusion

devices lead to the creation of carbon based compositions on their surface. The

deterioration of diagnostic mirrors is one of consequences of this process. Depositions on

the mirrors not only reduce the intensity of reflected radiation, but also strongly change its

spectrum. So, the mitigation of the deposition is necessary for the normal operation of

ITER optical diagnostics where a high number of in-vessel mirrors will be used. In order to

develop techniques for mitigation of deposition a through study of corresponding processes

in modern fusion devices is necessary.

First experiments on exposure of mirrors have been carrying out on JET, TEXTOR, Tore-

Supra and T-10. In particular, stainless still mirrors were undergone of a long-term

exposure in upper diagnostic port near carbon limiters of T-10. Part of the mirrors was

screened from plasma during exposure. The hydrocarbon films were found on all mirrors as

a result of the exposure, but the thickness of deposits on screened mirrors was less than 100

nm and their reflectivity not changed practically. So it is possible to protect mirror by

means a location behind screen with small pupil as proposed for ITER H Alfa Spectroscopy

system.

Last experiments on T-10 were conducted without ring limiter and were characterized by

high erosion of movable carbon based limiter. In result at the mirrors located in upper

diagnostic port in front of movable limiter the deposition rate increased about 10 times (up

to 4 nm/s) as compared with plasma operation mode with ring limiter. A few mirrors were

exposed only during vacuum vessel conditioning. These mirrors were deposited with

opaque films about 1 om thickness. It means that diagnostic mirrors in ITER should be

protected by shutter.

The paper also presented results of the first experiment on hydrocarbon film deposition on

heated metallic mirrors by means magnetron sputtering of graphite cathode in deuterium

discharge. It is shown that the mitigation of hydrocarbon film deposition take place at the

temperature of mirrors higher than 300flC.

P-1.090, Monday June 27, 2005

Page 96: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

A Vacuum Photoemission Detector for X-ray Tomography

on the ITER.

Yu.V.Gott, M.M.Stepanenko

Russian Research Centre Kurchatov Institute,

Kurchatov sq., Moscow, 123182 Russia

A Vacuum Photoemission Detector (VPD) designed for ITER plasma tomography

with help of plasma X-ray thermal radiation is described. Such detector allows us to detect

X-ray thermal radiation in the presence of intense neutron and gamma fluxes. The results

of the VPD tests with help of X-ray tube radiation and with help of 60

Co gamma radiation

are presented. It is shown that for ITER parameters the noise signal will be about 100

times less than signal from X-ray radiation. The signal value, about 10 oA, give us

possibility to transport it to control room without using a preamplifier.

P-1.091, Monday June 27, 2005

Page 97: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Calculation of Plasma Boundary Using Video Images

D.P. Kostomarov1, A.A. Lukianitsa1, F.S. Zaitsev1,a,

V.V. Zlobin1, R.J. Akers2, L.C. Appel2,b, D. Taylor2,

1Moscow State University, Faculty of Computational Mathematics and Cybernetics, RF2Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, UK

ae-mail: [email protected], be-mail: [email protected]

An important direction of plasma diagnostics in toroidal devices is reconstruction of

plasma boundary shape and position using experimental measurements. In the recent

past the techniques were mainly based on magnetic measurements and X-rays. Usually

reconstruction of the plasma boundary for the whole discharge with traditional tech-

niques requires time and it is not easy to calculate the boundary between discharges

during the experimental campaign. In some cases plasma boundary reconstruction is

inaccurate, since mathematically the problem is deeply ill-posed.

In this paper we propose a new fast and relatively accurate technique for reconstruct-

ing plasma shape and position. It is based on video image processing obtained by fast

camera. Such kind of images are routinely available on a number of tokamaks. The main

physical effect exploited is higher brightness of the plasma boundary.

The complexity of the problem is determined by several factors. Due to high dynamics

of the process it is impossible to change exposition synchronously. So, some frames

become indistinct and/or over exposed. The internal surface of the toroidal chamber has

usually mirror-like properties and light, reflected from different technological ledges and

apertures, mixes with the plasma fluorescence. Several areas with similar brightness can

be present. The image is distorted by optical properties of the camera which should be

removed for correct reconstruction of the plasma boundary coordinates.

The authors managed to create fast and reliable algorithm for accurate extracting

cylindrical coordinates (R, Z) of the plasma boundary from video image. The algorithm

is based on the method of dynamic programming. Formulation of the mathematical

problem and details of the algorithm, including the method for camera calibration, will

be presented in the contribution.

The algorithm is implemented in code VIP (Video Image Processing), which can

evolve (R, Z) coordinates of the boundary synchronously with selected graphs of mea-

sured plasma parameters. Results of MAST discharges processing will be presented and

compared with magnetic reconstructions, given by code EFIT.

VIP results can be used as an additional constraint for the flux surfaces reconstruc-

tion procedures, e.g. presented in Ref. [1] or EFIT. Other possible applications of the

algorithm will be discussed. High speed of the algorithm allows to hope for creation of

real-time feed-back plasma shape and position control system based on image processing.

Several cameras can allow to reconstruct 3D plasma shape and obtain information about

plasma rotation and axial asymmetry.

Acknowledgement. The MSU work was partly funded by the Russian Foundation

for Basic Research, grants No. 02-01-00299 and SS-1349.2003.1. The UKAEA work was

jointly funded by the UK Department of Trade and Industry and by Euratom.

References. [1] F.S. Zaitsev, A.B. Trefilov, R.J. Akers. An Algorithm for Recon-

struction of Plasma Parameters Using Indirect Measurements. 30th EPS Conf. on

Contr. Fus. and Plasma Phys., St. Petersburg, 2003, p-2.70.

P-1.092, Monday June 27, 2005

Page 98: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

FAST ELECTRON STUDIES IN T-10 PLASMAS BY MEANS OF

CARBON PELLET INJECTION

V.Yu. Sergeev1, V.M. Timokhin

1, V.G. Skokov

1, S.V. Krylov

2, V.I. Poznyak

2,

P.V. Savrukhin2, L.N. Khimchenko

2 and B.V. Kuteev

2

1 State Polytechnical University, St.Petersburg, Russia

2 Nuclear Fusion Institute, Russian Research Center “Kurchatov Institute”, Moscow, Russia

The new impurity pellet injection system of T-10 has a wide spectrum of applications

for diagnostics and high temperature plasma discharge control. The fast electron studies by

means of carbon pellet injection were carried out and results of these experiments are

presented in the paper.

In the experiments, spherical carbon pellets of 0.4-0.6 mm in size were accelerated up

to 500 m/s velocities in the direction of plasma core. The pellet ablation was observed in CII

(723 nm) line emission by the CCD camera, the wide-view photodetector and the set of the

narrow collimated photodetectors. The data set obtained by these diagnostics allowed us to

calculate the pellet ablation rate profile versus plasma minor radii with accuracy of about

1 cm. The experiments were carried out in wide plasma and injection parameters range.

Results of the pellet ablation rate measurements are compared with those simulated

using the NGS pellet ablation model [1]. In OH discharge, the enhanced ablation zone of

about cm widths starts to appear when the plasma density decreases below 1.5·1013

cm-3

.

The peaks on the pellet ablation rate radial profiles are more pronounced at lower plasma

densities. One can suppose that the reason of the enhanced ablation might be runaways

generated at the beginning stage of the discharge.

More narrow peaks of the enhanced ablation rate with less than cm width might be

distinguished on the ablation rate profiles of the pellet injected during ECR additional

heating stage. It might appear due to the suprathermal ECR driven electrons similar to those

reported in Ref. [2]. Data of ECE and HXR diagnostics are analyzed and compared with

pellet ablation features. Possible mechanisms of the pellet ablation enhancement are

discussed.

References

[1] Kuteev B.V., Sergeev V.Yu., Tsendin L.D., Plasma Phys. Rep. 10 (1984) 572.

[2] Timokhin V.M., et al., Techn. Phys. Letters 30 (2004) 298.

P-1.093, Monday June 27, 2005

Page 99: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Study of the ICRH antenna coupling at TEXTOR

G. Van Wassenhove1, P. Dumortier1, F. Louche1, A. Lyssoivan1, A. Messiaen1, O.

Schmitz2, M. Vervier1

Partners in the Trilateral Euregio Cluster: 1 LPP-ERM/KMS, Euratom-Belgian State Association, Brussels, Belgium

2 IPP, Forschungszentrum Jülich GmbH, EURATOM Association, D-52425 Jülich, Germany

Measurements of the impact of plasma conditions on the antenna distributed loading

resistance have been restarted at TEXTOR to study the coupling in case of increased

plasma antenna distance during, for instance, use of dynamic ergodic divertor and to

study the possible improvement of coupling with gas injection in the vicinity of the

antenna. The measurement of the antenna resistance at TEXTOR is based on the

determination of the standing wave pattern in the transmission line between the ICRH

generator and the RF antennae with voltage probes and directional couplers. Those

measurements are now available routinely with a time response of 10-4 s. The RF

heating conditions during this study were: r phasing, p=32.5 MHz, BT0= 2.25T, H/D

ratio ~= 10-20%. The main factor determining the antenna impedance is the distance

between the antenna and the plasma. This is for instance seen during a programmed

displacement of the plasma during a shot. A good correlation is found between the

measured antenna impedance and the position of the cut-off density layer (ne~=2 1018

m-3) measured with the Li beam diagnostic for many experimental conditions with

various plasma conditions (position of the plasma, heating power, injection of Neon

inducing a transition in an improved confinement regime…). The experimental results

of RA versus position of ne at cut-off density are well fitted by an exponential decay law

with a decreasing length of ~3.7 cm. The dependence of the antenna impedance on the

other plasma parameters (BT ,H/D ratio, crash of the saw-tooth...) and increase of the

coupling resistance in case of injection of gas in the vicinity of the antenna are

systematically studied.

P-1.094, Monday June 27, 2005

Page 100: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Electron Cyclotron Current Drive experiments in the FTU tokamak

S.Nowak1, G.Granucci1, C.Sozzi1, A.Bruschi1, F.Gandini1, L.Panaccione2, P.Buratti2,

O.Tudisco2, C.Mazzotta2, E. Giovannozzi2, ECRH1 and FTU2 team

1IFP CNR EURATOM Association, Milano, Italy

2 ENEA EURATOM Association, Frascati, Italy

Electron Cyclotron Current Drive (ECCD) experiments were performed in the FTU

tokamak with EC power up to 1.6 MW delivered by 4 gyrotrons at 140 Ghz. The EC

launching system /1/, steerable in poloidal and in discrete toroidal angles, allows to

localize along the minor radius the non-inductive current generated by high collimated

beams, injected from the low field side in O-mode fundamental harmonic. The aim of

the ECCD experiments was to explore the full range of the injection parameters, to

assess the calculation models and the efficiency of the ECCD at ITER relevant plasma

density and toroidal magnetic field.

EC driven currents were generated during up to 400 ms in up-shifted scheme in target

plasmas with Ip=360 kA, line electron density 0.6<ne<0.7 1020 m-3 , 4.5<Bt<5.2 T,

3<Te<5 keV and 2<Zeff<3. EC current (IECCD) was evaluated using two different

techniques: from the comparison of loop voltage measurements in co and counter cases

and from the determination of plasma resistance using the neoclassical resistivity. We

found a good agreement (within 10%) with the theoretical calculations performed by

using the ECWGB beam tracing code/2/. The estimations of IECCD for all the allowed

injection toroidal angles are presented; values up to 30 kA were obtained for injection

toroidal angles of ±20°. Even with this low overall IECCD, significant effects due to the

local re-shaping of the plasma current density were observed, including modifications

of the sawtooth activity in discharges with co or counter EC injection and eased access

to the ITB regime.

/1/ Granucci G., et al, Fusion Science & Tech. 45 (2004) 387

/2/ S.Nowak , E. Lazzaro , G. Ramponi, Phys. Plasmas 3, (1996) 4140

P-1.095, Monday June 27, 2005

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Interpretation of the LHCD efficiency scaling with the electron

temperature

E. Barbato

Associazione EURATOM-ENEA sulla Fusione, CR Frascati (Roma), Italy

In the present paper, the increase of the current drive (CD) efficiency by

Lower Hybrid waves (LH), as a function of the electron temperature, is interpreted

as due to a temperature dependence of both power absorption and n|| power

spectrum in the plasma, as they result from a numerical calculation based on

standard ray-tracing Fokker Planck code package. Such a code is applied to

simulate one FTU shot at several temperature levels. This calculation shows that,

according to the experimental findings, there is an increase of the LHCD

efficiency as a function of the volume average electron temperature, <TE>VOL.

Such an increase is linear up to <TE>VOL=0.6KeV, for the chosen FTU parameters,

and shows, as expected, a sign of saturation at <TE>VOL>1KeV. From the

numerical calculations it results that at low temperature, when multiple pass occur,

absorption in the electron tail is lower, due to the collisional absorption-taking

place in the periphery; furthermore the n|| spectrum in the plasma, broadened by the

toroidal geometry up to the value need for the absorption, is larger in high value

side, also affecting the LHCD efficiency. On the contrary at higher temperature

both these effects tend to disappear and the LHCD efficiency increases.

P-1.096, Monday June 27, 2005

Page 102: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Plasmoid drift during vertical pellet injection in FTU discharges

E. Giovannozzi1, S.V. Annibaldi1, M.L. Apicella1, L.R. Baylor2, P. Buratti1, M. De Benedetti1, B. Esposito1, D. Frigione1, L. Garzotti3, G. Granucci1, O. Kroegler1,

D. Marocco1, S. Martini3, C. Mazzotta1, G. Monari1, P.B. Parks4, L. Pieroni1, P. Smeulders1, M. Romanelli1, D. Terranova3, O. Tudisco1 and the FTU Team

1) Associazione EURATOM-ENEA sulla fusione, Centro Ricerche Frascati, c.p. 65, 00044 Frascati, Roma, Italy.

2) Oak Ridge National Laboratory, Oak Ridge, Tennessee, USA 3)Associazione EURATOM-ENEA-CNR sulla Fusione, Consorzio RFX,

Corso Stati Uniti 4, I-35100, Padova, Italy 4) General Atomics, San Diego, California 92186-5608, USA

Fuelling the central plasma region is a key issue in tokamak experiments. FTU tokamak

(major radius R = 0.935 m, minor radius a = 0.3 m, maximum magnetic field Bt = 8 T,

maximum plasma current Ip = 1.6 MA) allows the study of pellet ablation at high field and

density typical of a fusion reactor. A vertical injector has been used to study pellet ablation

and plasmoid drift. Pellets with ~1.5x1020 particles and a speed of 500 m/s are injected along

a vertical chord on the high field side. Two main mechanisms contribute to transporting the

pellet material to the plasma center on a fast time scale, namely plasmoid drift, and MHD

advection [1,2]. Plasmoids formed during pellet ablation drift along the radial direction and

take the pellet particles near the q=1 surface. Then MHD events (basically m=1 instabilities),

advect the density to the plasma center, resulting in very peaked profiles. Thomson

scattering density measurements were available in some discharges just after pellet ablation,

before any MHD event. As expected the density profile was hollow at that stage.

The measured density profile has been compared with the results of a pellet ablation and

relocation code. The code is based on a description of plasmoids as they cross the magnetic

field lines including effects due to the pressure, curvature and safety factor profiles [3].

Results of this comparison will be discussed in detail. During pellet ablation broad band

MHD activity at very high frequency has been observed. This MHD activity disappears as

soon as the pellet ablation is completed, as shown by comparison between magnetic

fluctuations and Dα signals. A possible explanation is the formation of Alfven waves during

plasmoid drift [4]. These experimental results will be compared with model predictions.

References

[1] Giovannozzi, E. submitted to Nuclear Fusion (2004)

[2] Annibaldi S.V., et al. Nucl. Fusion 44, (2004) 12

[3] Parks, P.B. and Baylor, L.R: accepted by Phys. Rev. Lett. (2004)

[4] Parks, P.B. Nuclear Fusion 32,12 (1992) 2137

P-1.097, Monday June 27, 2005

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Quantification of suprathermal current drive on FTU

G.Granucci1, A.Bruschi1, P.Buratti2, G.Calabrò2, R.Cesario2, D.Farina1, F.Gandini1,

E.Giovannozzi2, C. Gormezano3, C.Mazzotta2, S.Nowak1, L.Panaccione2, V.Pericoli-

Ridolfini2, S.Podda2, G.Regnoli2, A.Simonetto1, C.Sozzi1, O.Tudisco2, ECRH1 and FTU2 team

1IFP CNR EURATOM Association, Milano, Italy

2 ENEA EURATOM Association, Frascati, Italy

3 retired in Paris

The suprathermal absorption of EC wave on plasma with LHCD generated fast electrons is

widely used on FTU. The availability of 1.1 MW of EC power at 140 GHz opened the

possibility of new experiments after the pioneering proof of principle obtained at lower

power level (0.4 MW). The presence of fast electrons in the plasma allows the resonant

interaction to occur at frequencies shifted up or down with respect to the cold resonance,

depending on the launched N//EC and on v// (parallel speed of electrons). The fast electron

population, which directly absorbs the EC power, is generated and sustained by LHCD (8

GHz, 1.5 MW). The mostly used suprathermal EC absorption scheme on FTU is the one

based on resonance at down-shifted frequencies, in which the EC wave is injected in a

plasma with toroidal field (BT=7T) well above the resonant one (5T). The presented results

refer to a plasma with line density in the range 0.6 - 1.0 1020m-3 and current between 400 and

800 kA. The measured overall CD efficiency is well above that due to the simple sum of the

expected current drive due to EC and LH, indicating the existence of a synergy between the

two waves. A comparison of two different injected polarizations (O-mode, X-mode) is

presented at different toroidal EC injection angles. The suprathermal ECCD interaction with

LHCD is applied, as presented in this work, also for the ITBs formation at high density

(ne0=0.9x1020m-3) and high magnetic field (BT=7.2T). The suprathermal ECCD, exhibiting the

same efficiency of LHCD (up to 0.3 1020 WA-1m-2), could be used to generate a substantial

amount of non-inductive current for sustaining steady state plasmas in the future reactors.

P-1.098, Monday June 27, 2005

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Injection of intense plasma jet in the spher ical tokamak Globus-M

K.B. Abramova, V.K. Gusev, Yu.V. Petrov, N.V. Sakharov, I.P. Scherbakov, A.V. Voronin

A.F.Ioffe Physico-Technical Institute, 194021 St. Petersburg, Politechnicheskaya st. 26

Russia

Plasma fuelling and density profile control are significant problems for any magnetic trap

with high performance operation. Further investigations of intense plasma jet at the test stand

and its injection in the tokamak Globus-M are presented. Results are discussed.

Injection was performed both at a small angle (15 degrees) to the vertical axis and at

equatorial plane (perpendicular to the vertical axis). High-kinetic energy jet generated with

already developed two stage plasma source (highly ionised hydrogen plasma jet with density

1022

m–3

, total number of accelerated particles @1019

, flow velocity @100 km/s) was injected

into Globus-M. During some experiments the plasma source was moved away from the

tokamak for about 1 m distance. This might help for the transformation of dense plasma jet

into dense neutral jet due to time-of-flight recombination process and improve penetration

into tokamak magnetic field. Comparison of injection efficiencies from different poloidal

position of the plasma gun is made.

Super fast dense gas stream injection (@10 km/s), produced only by the first (gas generating)

stage of the two-stage source, into tokamak Globus-M was done. The efficiency of gas and

plasma injection was compared.

On the course of preparation for the next set of the experiments plasma source characteristic

upgrade at the test stand were done. Plasma injection of the upgraded jet source with velocity

@200 km/s or density @1022

m–3

in the Globus-M is planned for the 2005-year spring

campaign.

The work is supported by IAEA, Research Contract No 12408 and RFBR grant No 04-02-

17606.

P-1.099, Monday June 27, 2005

Page 105: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Off-Axis NBI fast ion dynamics in Trace Tritium Experiment

I Jenkins1, C D Challis1, J Hobirk2, Yu F Baranov1, L Bertalot3, D L Keeling1,

V Kiptily 1, S E Sharapov1 and contributors to the EFDA-JET workprogramme *

1 EURATOM-UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxon, OX14 3DB, UK2 Max-Planck-Institut für Plasmaphysik, EURATOM Association, D-85748 Garching, Germany

3 Associazone EURATOM-ENEA sulla Fusione, C R Frascati, Frascati, Italy*see the Appendix of J Pamela et al, Proc 20th IAEA Fusion Energy Conference 2004 (Vilamoura, 2004)

The understanding of fast ion physics is important for modelling and

interpretation of neutral beam injection (NBI) in tokamaks. It is required for the

derivation of transport coefficients and the simulation of heating and current drive in

beam heated plasmas. However, such simulations do not always reproduce experimental

effects, as is the case for off-axis NBI current drive on ASDEX Upgrade#.

The JET tokamak possesses unique tools to diagnose NBI fast ion behaviour in

the ability to inject tritium beams and in a 2-D 14MeV neutron camera, which can be

used to measure the neutron profiles from DT reactions caused by fast tritons in a

deuterium plasma. Experiments have been performed with the injection of short tritium

beam blips (~300ms) with both on- and off-axis beam trajectories. Plasma conditions

were chosen so that the thermal neutron yield was negligible compared with beam-

target and, in the case where deuterium beams were also used, beam-beam interactions.

Data was obtained at two values of toriodal magnetic field, 1.2T and 3T, corresponding

to q95≈3.3 and q95≈8.5 respectively. At high field the neutron profiles indicate a peaked

fast ion distribution for on-axis tritium injection and a hollow profile for off-axis beams

with an inboard-outboard asymmetry, in agreement with Monte Carlo simulations using

the TRANSP code. In the low field cases the neutron profile was again peaked for on-

axis injection but markedly less hollow in the off-axis cases. This effect is not explained

by the presence of sawtooth oscillations or other MHD phenomena detectable with

Mirnov coils. Initial simulations of the high and low field plasmas do not reproduce the

measured change in the neutron profile suggesting a radial redistribution of the fast ions

in the low field plasmas. Detailed comparison of simulation with measurement will be

presented along with discussion of the degree to which a classical picture of fast ion

behaviour can be reconciled with these observations.

This work was performed under the European Fusion Development Agreement,

and funded partly by the UK Engineering and Physical Sciences Council and by

EURATOM.

#A Stäbler et al Fusion Science and Technology 44 (2003) 730

P-1.100, Monday June 27, 2005

Page 106: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Behavior of Ions in Auxiliary Heating Experiments in Globus-M Spherical

Tokamak.

N.V.Sakharov1, B.B.Ayushin

1, A.G.Barsukov

2, F.V.Chernyshev

1, V.V.D’yachenko

1,

V.K.Gusev1, R.G.Levin

1, V.B.Minaev

1, A.B. Mineev

3, M.I. Mironov

1, M.I.Patrov

1,

Yu.V. Petrov1, O.N.Scherbinin

1, G.N. Tilinin

2, S.Yu. Tolstyakov

1.

1A.F.Ioffe Physico-Technical Institute, St.Petersburg, Russia

2 Nuclear Fusion Institute, RRC “Kurchatov Institute”, Moscow, Russia

3 D.V. Efremov Institute of Electrophysical Apparatus, St. Petersburg, Russia

Plasma auxiliary heating was studied in low aspect ratio plasmas in spherical

tokamak Globus-M (major radius 0.36 m, minor radius 0.24 m, toroidal magnetic field 0.3-

0.4 T, vertical elongation 1.2-2). We used a tangential neutral beam injection (NBI) with the

beam power up to 0.7 MW and the beam energy changed in the experiments in the range 20-

30 keV. The ion cyclotron resonance heating (ICRH) at the frequency of 7.5 MHz and the

power of 0.2-0.3 MW was performed on the fundamental harmonic of the hydrogen minority

in deuterium plasma. The ion temperature in the plasma core was studied by means of the

12-channels neutral particle analyzer ACORD-12. The auxiliary heating experiments were

carried out in various limiter and divertor magnetic configurations with upper and lower X-

point positions. The ion confinement is satisfactory described by the neoclassical transport

coefficients in a low collisionality regime. For both methods of auxiliary heating the power

absorbed by ions appeared to be comparable or exceeded the electron-ion heat flux. This led

to a strong increase of the ion temperature in the plasma bulk. At the same time no

significant increase of the electron temperature measured by Thomson scattering diagnostic

was observed. The NBI led to the increase of the plasma toroidal rotation up to 30 km/s. The

NBI was also accompanied by sawtooth oscillations and sometimes by MHD modes. These

phenomena can explain the early saturation of the ion temperature rise observed during the

heating pulse. The ICRH experiments were carried out in a wide range of the hydrogen

percentage in deuterium plasma. The charge exchange energy spectra of deuterium and

hydrogen are presented. The estimates of total energy confinement time values were derived

from EFIT analysis of magnetic measurements. The MHD modes structure was studied by

Mirnov coils and SXR pin-hole camera. First results of combined NBI and ICRH heating are

described

P-1.101, Monday June 27, 2005

Page 107: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

First experiments on NBI in the TUMAN-3M tokamak

L.G. Askinazi1, A.G. Barsukov

2, F.V. Chernyshev

1, V.E. Golant

1, V.A. Kornev

1, S.V.

Krikunov1, V.V. Kuznetsov

2, A.D. Lebedev

1, S.V. Lebedev

1, A.D. Melnik

1, A.A.

Panasenkov2, A.R. Polevoi

2, S.A. Ponaev

1, D.V. Razumenko

1, V.V. Rozhdestvensky

1, A.I.

Smirnov1, G.N. Tilinin

2, A.S. Tukachinsky

1, M.I. Vildjunas

1, N.A. Zhubr

1

1Ioffe Institute RAS, St.-Petersburg, RF

2RRC “ Kurchatov Institute”, Moscow, RF

Neutral Beam Injection (NBI) heating on the TUMAN-3M is aimed on increasing the

experimental resources of the tokamak [1, 2]. Preparations for the NBI experiments on

TUMAN-3M have been completed in spring 2004. In the NBI system tests the ion source

current 30 A, beam energy 28 keV and NB power 500 kW in 20 ms pulse have been obtained.

In order to provide full beam absorption the following setup were chosen for the first

experiments: tangential co-injection with beam energy 22 keV.

The NBI heating on a small tokamak is a complicated task because of relatively small

characteristic confinement times as compared with beam slowing down time. In the TUMAN-

3M the typical energy confinement time is 5-7 ms, whereas slowing down time is 15-18 ms.

Other feature of the experiment is small relative portion of the power delivered from beam to

plasma ions. Because of low electron temperature in the target plasma ~ 70% of the beam

energy is absorbed by electrons and only 30% goes to ion component.

In the first experimental runs the NBI power was 330 kW. The target plasma

parameters were as follows: Ip=130 kA, Bt=0.8 T, nav=(1.3-3.0)©1019

m-3

, Te(0)<0.6 keV,

Ti(0)<0.2 keV. NBI heating resulted in 2-3 fold increase in the stored energy, increase of Ti(0)

from 190 to 330 eV. After boronization a tendency of FTi(0) increase was observed,

indicating essential role of charge-exchange losses in ion heat balance.

The first NBI experiments have revealed peaking of the density, electron temperature

and current density profiles. These effects could be attributed to increasing Ware pinch and

some current drive in the plasma core. Current drive efficiency will be tested in the planned

experiments with counter-injection. Fast ion confinement and its effect on ion/electron

heating will be studied in further experiments.

This work was supported by Russian Foundation for Basic Researches (Grant ヽ 03-

02-17417) and by Department of Education and Science (Project TUMAN-3M ヽ 01-06,

Grant “Leading Scientific School” ヽ 2216.2003.2).

References

1. Vorobjev G.ぜ. et al. “Plasma Physics”, v 9, 1983, p. 105.

2. Askinazi L.G., et al, “Plasma Devices and Operations”, v 11, 2003, pp. 211-218.

P-1.102, Monday June 27, 2005

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Study of the Beam - Plasma Interaction

in the Globus-M Spherical Tokamak

V.B. Minaev 1)

, B.B. Ayushin 1)

, A.G. Barsukov 2)

, F.V. Chernyshev 1)

, L.A. Esipov 1)

,

V.K. Gusev 1)

, V.G. Kapralov 3)

, S.V. Krikunov 1)

, V.M. Leonov 2)

, R.G. Levin 1)

,

A.N. Novokhatskii 1)

, M.I. Patrov 1)

, Yu.V. Petrov 1)

, K.A. Podushnikova 1)

,

V.V. Rozhdestvenskii 1)

, N.V. Sakharov 1)

, G.N. Tilinin 2)

, S.Yu. Tolstyakov 1)

1) Ioffe Physico-Technical Institute, RAS, St. Petersburg, Russia

2) NFI RRC “Kurchatov Institute”, Moscow, Russia

3) St. Petersburg State Polytechnical University, St. Petersburg, Russia

Results of the neutral beam injection experiment in the spherical tokamak Globus-M [1, 2] at

the Ioffe Institute are presented. Co-injection scheme was chosen. Two types of ion sources

with different ultimate ion currents were applied. The injector construction made possible to

change the beam energy step by step in the range of 20 – 30 keV. The output power varied

from 0.3 to 1.0 MW and depended on the beam energy and the kind of ion source. We

compared the efficiency of plasma heating by means of hydrogen and deuterium beams with

the same energy and power in one experimental session. The parametric dependence of beam

absorption was studied in the range of plasma densities (1 – 7)©1019

m-3

, plasma currents 150

– 250 kA and toroidal magnetic field 0.3 – 0.4 T in limiter and divertor configurations. The

heating of ion components was investigated by means of 12-channel neutral particle

analyzer. The charge exchange hydrogen and deuterium energy spectra were studied both for

thermal and high energy ranges. Thermalization rates of high and low energy particles were

investigated and compared to neoclassical model. The beam absorption modeling by ASTRA

code and by simple 1D code using neoclassical ion transport coefficients was performed.

[1] Gusev V.K., Golant V.E., Gusakov E.Z., et al., Technical Physics, Vol.44 (1999) No. 9,

pp. 1054-1057

[2] Askinazi L.G., Barsukov A.G., V.E.Golant, et al., Plasma Devices and Operations,

Vol.11 (2003) pp.211-218

P-1.103, Monday June 27, 2005

Page 109: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Study of MHD events initiated by pellet injection into T-10 plasmas

B. Kuteev1, L. Khimchenko

1, S. Krylov

1, Yu. Pavlov

1, V. Pustovitov

1, D. Sarychev

1,

V. Sergeev2, V. Skokov

2, V. Timokhin

2

1 NFI RRC “Kurchatov Institute”, Moscow, Russia2 State Polytechnical University, St. Petersburg, Russia

There are several events which might be responsible for ultra fast transport [1]

of heat and particles during pellet ablation stage in a tokamak. Those are jumps of

transport coefficients, plasma drifts in the pellet vicinity and MHD events with time

scale significantly shorter than the pellet ablation time. The role of the latter is still not

very well understood due to a lack of studies. This paper is devoted to detailed study of

the effects during the pellet ablation phase (~one millisecond) with main objective to

determine the relation between pellet (material Li, C, KCl, size and velocity) and

plasma parameters (q-value a the pellet position, plasma density and temperature) which

initiate microsecond MHD events in plasma.

The pellets were injected into both into Ohmic and ECE heated plasmas (up to 3

MW) in the T-10 tokamak at various stages of the plasma discharge, in a wide range

from the very beginning up to the post-disruption stage.

It is observed that at some conditions a pellet ablates in the plasma without

accompanying MHD events. This occurs at the highest plasma densities even if a pellet

penetrates through q=1 magnetic surface. The ablation rate corresponds to NGSM in

this case.

Small scale events may occur near rational magnetic surfaces and the ablation

rate fluctuations may be explained by reconnection. Both increase of the longitudinal

heat flow due to plasma convection from higher temperature region and growth of the

electric field generating supra-thermal electrons may be responsible for the enhanced

ablation. Large scale MHD events envelop a region inside q<3. It is observed that the

MHD-cooled area is not poloidally symmetric.

Mechanisms of the phenomena observed and their consequences on tokamak

operation are discussed.

1. M Sakamoto et al 1991 Plasma Phys. Control. Fusion 33 583-594.

P-1.104, Monday June 27, 2005

Page 110: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Recent results of hydrogen pellet injection

V.G.Kapralov1, B.V.Kuteev

1, G.A.Baranov

2, V.K.Gusev

3, V.S.Koidan

4, S.V.Lebedev

3,

L.G.Askinazi3, V.Yu.Bakharev

1, S.M.Egorov

1, V.V.Elagin

5, P.G.Gabdullin

5,

S.A.Perfiljev2, V.V.Postupayev

4, V.N.Skripunov

2, S.V.Sergeev

1, S.Salem

1, I.V.Klotchkov

5,

D.S.Moseev5, I.A.Sharov

5, A.N.Kuznetsov

5

1State Polytechnical University, St.Petersburg, Russia

2Efremov Institute, St.Petersburg, Russia

3Ioffe Physico-Technical Institute, St.Petersburg, Russia

4Budker Institute of Nuclear Physics, Novosibirsk, Russia

5TUAP Ltd., St. Petersburg, Russia

The contribution presents recent results in the filed of hydrogen pellet injection

physics and technologies for tokamaks and multimirror traps.

Current status of the pellet injection system for Globus-M spherical tokamak are

discussed. Results of testing of ITER relevant technical solutions on this centrifuge injector

are presented, including design of interface unit with curved guide tube and usual design

based on the stop cylinder.

Current status of pellet injector for GOL-3 multimirror machine is described. this

simple injector allows to produce main plasma discharge in GOL-3 by pellet evaporation

with powerful electron beam. The injector produces small solid hydrogen pellets (<1 mm)

with extremely low velocity (<10 m/s). Slow pellet is a target for an electron beam that

heats plasma in GOL-3.

Current status of joining of pellet injection with NBI injection is presented as well.

The system includes a pellet injector based on in-situ technology and NBI injector both

connected to the same tangential port of TUMAN-3M tokamak. The geometry allows

comparison the different regimes with co- and counter- NBI, co- and counter- pellet

injection and pellet injection directly in the region of beam-plasma interaction.

The work was supported by RFBR grants 04-01-16911, 05-02-17160 and 05-02-17269.

P-1.105, Monday June 27, 2005

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ICRF Heating together with neutral beams in Volume Neutron Sources

JUST-T.

E.ん. Azizov, A.A. Chernov, V.N. Dokuka, A.V. Krasilnikov, R.R. Khayrutdinov,

N.B. Rodionov

State Research Center of Russia TRINITI, Troitsk, Moscow region

At the present time the conception [1] of Volume Neutron Sources (VNS) for

transmutation of minor actinides based on JUST-T tokamak project with the aspect ratio

A=2 is considered. Such value of aspect allows us to use the advantages of both: spherical

tokamaks (large elongation, increased values of safety factor q95 and normalized beta dN)

and standard fusion devices (SN configuration with elongation k95…1.7). Injection of neutral

beams with energy 140 KeV and 200„400 KeV are suggested as an auxiliary plasma heating

with full power Paux~ 40 MW. In this work the tokamak plasma heating by waves of

cyclotron frequency range (ICRF) in parallel with neutral beams is proposed. With the help

of 2D full wave code a simulation of wave excitation and dissipation in Ion Cyclotron

(ICRF) Range of Frequencies at JUST-T in the regime of minority atoms heating has been

carried out. The calculations are performed for 3He minority in DT plasmas with almost

equal values of D and T concentration. The problem of effective formation of the traveling

fast waves in tokamak JUST-T by phasing of several loop antennas located at certain

distance in toroidal direction to provide the current drive is considered. The possibility of

plasma current generation due to trapping of ions, created during neutral injection, into

magnetic wells of fast sonic waves traveling in the direction of beam injection is studied.

ICRF Heating together with neutral beams in Volume Neutron Sources JUST-T let us to

decrease energy and power of the injected neutral beams.

[1]. E.A.Azizov et al., "The Concept of the Volumetric Neutron Source of The Basis of the

JUST-T Tokamak for Minor Actinides Transmutation", Plasma Device and Operations, Vol.

11. No. 4, 279 (2003).

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Overview of global MHD behaviour in the modified RFX Reversed

Field Pinch

T. Bolzonella, E. Martines, D. Terranova, P. Zanca, R. Cavazzana, L. Grando,

N. Pomaro, G. Serianni, N. Vianello, M. Zuin

Consorzio RFX, Associazione Euratom-ENEA sulla Fusione,

Corso Stati Uniti 4, I-35127, Padova, Italy

The RFX Reversed Field Pinch device has recently undergone important modifications

of its magnetic boundary, the more relevant being the substitution of the thick shell

surrounding the vacuum chamber (450 ms Bv penetration time, to be compared to a

100-150 ms typical duration of a plasma discharge) with a resistive one (50 ms Bv

penetration time). As in other toroidal confinement devices, global MHD instabilities

are deeply influenced by magnetic boundary characteristics and by the related mean

equilibrium current and field profiles.

In this work we present a first overview of MHD instabilities behaviour in the modified

RFX with the main aim of comparing the last results to the situation found in the past

RFX experiments. New magnetic diagnostics help the study of magnetic fluctuations: 4

toroidal arrays of 48 coils measuring the 3 components of B are placed outside the

vessel and give a detailed description of low frequency (0-5 kHz) fluctuations for modes

with toroidal and poloidal mode numbers n=0-24 and m=0-2 respectively. In the paper

number and periodicity of the main global instabilities will be shown. Of particular

interest is the description of the occurrence of the phase- and wall-locking phenomenon

of m=1 tearing modes and of its relation with main plasma parameters such as plasma

current and density. The presence of other instabilities related to the short penetration

time of the new shell (Resistive Wall Modes, RWM) will be addressed as well.

A further new set of coils installed for the first time inside the vacuum vessel (2 toroidal

arrays of 48 coils each measuring Bt fluctuations) allows for the first time in RFX the

characterisation of the fast behaviour (>5 kHz) of global MHD instabilities. Internally

resonant tearing modes show fast (10-20 kHz) rotations even in presence of the wall

locking phenomenon. Dynamics of dynamo relaxation events can be studied in detail at

fast frequencies as well. The relation between external and internal measurements is

finally addressed.

P-1.107, Monday June 27, 2005

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The Scientific Program of the Ignitor Experiment

G. Cenacchi1, B. Coppi2, A. Airoldi3, F. Bombarda1, P. Detragiache1 and M. Romanelli1,

1Ignitor Project Group, ENEA,Italy, 2MIT, Cambridge, MA (US),3IFP, CNR.Milano, Italy

Demonstration of ignition, the study of the physics of the ignition process, and the

heating and control methods for a magnetically confined burning plasma are the most

pressing issues in present day research on nuclear fusion and they are specifically addressed

by the Ignitor experiment [B. Coppi et al., Nucl. Fusion 41, 1253 (2001)]. The Ignitor

machine has been designed to produce up to 11 MA of toroidal plasma current with about 9

MA of poloidal current within relatively small dimensions (R0 @ 1.32 m, a ¥ b @ 0.47 ¥ 0.86

m2) and with reasonable safety factors against macroscopic plasma instabilities. The main

heating process is Ohmic heating through the high plasma current, although an ICRH system

is adopted as auxiliary heating. Particle fuelling is provided through gas and high speed

pellet injection. The experiment is the first that has been proposed and designed to achieve

fusion ignition conditions in well confined deuterium-tritium plasmas.

On the way to reach its main goals, and after, Ignitor will produce intermediate

results and will explore plasma regimes that are not accessible to other existing or planned

experiments, providing needed information on some of the most critical extrapolations

involved in designing burning plasma experiments. High magnetic field experiments can

overlap with the envisioned operational regimes of large-scale devices in terms of the

relevant dimensionless plasma parameters but at the same time they can explore the only

available path to approach ignition and thus open the way for realistic fusion reactors. It is

important to note that, while tritium is the necessary step forward for any new fusion facility

of conventional concept, even with H, He, and D Ignitor will provide results that can justify,

in themselves, the construction of the machine. For example, the unique feature of having an

elongated cross section but no divertor, and a high Z first wall acting as an extended toroidal

limiter can test an alternative, considerably simpler, solution for plasma facing components

in a burning plasma experiment. Perhaps even more important is the level of internal power

density available in Ignitor. The experimental life of the Ignitor device will follow three

stages, characterized mainly by different plasma components: phase I in H and He, phase II

in D and phase III in D-T, where the use of T will allow to carry out the most ambitious part

of the program. In this work we present examples of the experiments and studies that can be

carried out in each of the three phases.

P-1.108, Monday June 27, 2005

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Simple criteria for optimization of trapped particle confinement in

stellarators ∗

V.V. Nemov1, S.V. Kasilov1, W. Kernbichler2, G.O. Leitold2

1 Institute of Plasma Physics, National Science Center “Kharkov Institute of Physics and

Technology”, Akademicheskaya Str. 1, 61108 Kharkov, Ukraine2 Institut für Theoretische Physik - Computational Physics, Technische Universität Graz,

Petersgasse 16, A–8010 Graz, Austria

Improving the collisionlessα-particle containment in stellarators is one of the key issues

in stellarator optimization problems. The most consequent approaches for the investigation of

this problem are realized in codes which follow particle orbits and, therefore, allow for direct

computation of particle losses. To increase computing efficiency, of course, also simple criteria

which address this problem properly are of big interest (e.g., minimization of the geodesic

curvature of the magnetic field line, residuals in the magnetic spectrum of quasi-symmetric

systems, effective ripple, WATER parameter) .

In the present work, new simplified criteria are proposed which are based on the computa-

tion of the bounce averaged∇ B-drift velocity of trapped particles across magnetic surfaces.

For a given stellarator magnetic field, the pertinent optimization parameters are numerically

calculated using a field line following code. With such optimization parameters being zero,

an absolute confinement of reflected particles is guaranteed. Comparisons between results for

different simplified criteria as well as for direct computations ofα-particle losses reveal the

applicability of the method.

The proposed criteria are applied to some magnetic configurations for which the neoclassi-

cal confinement properties were studied formerly by different methods. In particular, a bench-

mark with effective ripple results is performed. From those results follows that the considered

technique provides a convenient auxiliary approach for the investigation of collisionless con-

tainment of trapped particles in stellarators.

∗This work has been carried out within the Association EURATOM-ÖAW and with funding from the AustrianScience Fund (FWF) under contract P16797-N08. The content of the publication is the sole responsibility of itspublishers and it does not necessarily represent the views of the Commission or its services.

P-1.109, Monday June 27, 2005

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Neoclassical transport for LHD in the 1/ν regime analyzed by the NEO

code ∗

V.V. Nemov1, M. Isobe2, S.V. Kasilov1, W. Kernbichler3, K. Matsuoka2, S. Okamura2

1 Institute of Plasma Physics, National Science Center “Kharkov Institute of Physics and

Technology”, Akademicheskaya Str. 1, 61108 Kharkov, Ukraine2 National Institute for Fusion Science, Oroshi-Cho 322-6, Toki-city, Gifu-Pref. 509-5292,

Japan3 Institut für Theoretische Physik - Computational Physics, Technische Universität Graz,

Petersgasse 16, A–8010 Graz, Austria

For LHD (Large Helical Device) the diffusion coefficients in the 1/ν regime are frequently

calculated using the codes GIOTA [1] and DCOM [2]. For a number of optimized stellarator

configurations the NEO code [3] was used for such calculations. Here, this code is applied to

compute the 1/ν transport also in LHD. In particular, these calculations allow for a benchmark

between NEO and GIOTA and, in addition, reveal the pertinent advantages of both codes when

compared to each other. This is of interest because both codes are much faster than codes based

on Monte-Carlo techniques (e.g. DCOM).

Computations are performed for magnetic configurations corresponding to fixed boundary

VMEC equilibria. Especially inward shifted LHD configurations are considered. Calculations

for such configurations are of interest since recent experimental findings for LHD [4] show

that good MHD stability and favorable transport are compatible in the inward shifted configu-

ration (Rax=3.6 m). A rather wide range of radiiRax of the magnetic axes is considered to find

an optimum value ofRax from the viewpoint of 1/ν transport. The results obtained are bench-

marked with the corresponding results obtained recently with the GIOTA code [5] as well as

with the Monte-Carlo calculations from the DCOM code [2].

Acknowledgments

The authors are indebted to Dr. M. Yokoyama (NIFS) for providing GIOTA results and Dr. K.Y.

Watanabe (NIFS) who had prepared the corresponding VMEC boundaries.

References

[1] M. Yokoyama, L. Hedrick, K.Y. Watanabe et al., submitted to NIFS Report, (2005)

[2] S. Murakami, A. Wakasa, H. Maaßberg et al., Nucl. Fusion42, L19 (2002)

[3] V.V. Nemov, S.V. Kasilov, W. Kernbichler and M.F. Heyn, Phys. Plasmas6, 4622 (1999)

[4] O. Motojima, N. Ohyabu, A. Komori et al., Nucl. Fusion43, 1674 (2003)

[5] M. Yokoyama, K.Y. Watanabe et al., J. Plasma Fusion Res., Rapid Communication 0095

∗This work was partly supported by the Association EURATOM-ÖAW. The content of the publication is thesole responsibility of its publishers and it does not necessarily represent the views of the Commission or itsservices.

P-1.110, Monday June 27, 2005

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Calculation of neoclassical transport in stellarators with finite

collisionality using integration along magnetic field lines∗

W. Kernbichler1, S.V. Kasilov2, G.O. Leitold1, V.V. Nemov2

1 Institut für Theoretische Physik - Computational Physics, Technische Universität Graz,

Petersgasse 16, A–8010 Graz, Austria2 Institute of Plasma Physics, National Science Center “Kharkov Institute of Physics and

Technology”, Akademicheskaya Str. 1, 61108 Kharkov, Ukraine

A new numerical method is presented which allows for an efficient evaluation of neoclas-

sical transport coefficients and of the bootstrap coefficient in stellarators for the case there is

no radial electric field. In this method, the approach [1] used in code NEO to compute the 1/ν

transport coefficient during integration along the magnetic field line is generalized to arbitrary

collisionality regimes. In more detail, the linearized steady-state drift kinetic equation (DKE)

is solved by a finite-difference method. The solution of the DKE is described in terms of the

phase space flux density throughs= constcuts, wheres is the distance measured along the

magnetic field line. The phase space is split into "ripples" which cover finite intervals overs

and extend into the velocity space. Within such a ripple, the problem is discretized by intro-

ducing levels over the perpendicular action. The distribution of these levels is specific for the

ripple. The DKE is approximated by a coupled set of ordinary differential equations overs for

the integrals of the phase space flux density over bands between the levels. The general solution

of the kinetic equation for a single ripple is then expressed in terms of a set of matrix relations

between the discretized phase space flux densities of particles entering and leaving the ripple

trough its boundaries. The whole set of these matrices is called a "propagator". The final solu-

tion is obtained after subsequent joining of these propagators using their group property. The

method has similar advantages as the NEO code, such as high speed, good convergence in low

collisionality regimes as well as the possibility of computations for magnetic fields given in

magnetic and real space coordinates, in particular, for magnetic fields resulting directly from

the Biot-Savart law and from new equilibrium codes such as PIES and HINT.

References

[1] V.V. Nemov, S.V. Kasilov, W. Kernbichler and M.F. Heyn, Phys. Plasmas6, 4622 (1999)

∗This work has been carried out within the Association EURATOM-ÖAW and with funding from the AustrianScience Fund (FWF) under contract P16797-N08. The content of the publication is the sole responsibility of itspublishers and it does not necessarily represent the views of the Commission or its services.

P-1.111, Monday June 27, 2005

Page 117: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Fast Ion Confinement in Tokamak Current Hole Regimes

K. Schoepf 1, P. Neururer

1, V. Yavorskij

1,2, V. Goloborod’ko

1,2

1Association EURATOM-OEAW, Institute for Theoretical Physics, University of Innsbruck, Austria

2Institute for Nuclear Research, Ukrainian Academy of Sciences, Kiev, Ukraine

While a tokamak current hole scenario featuring a central plasma region with nearly

zero toroidal current density and hence with no poloidal magnetic field is known to provide

an improved confinement of the bulk plasma, it may enhance the loss of fast ions such as

NBI ions and charged fusion products [1,2]. However, in a fusion reactor it is the retention

of these energetic particles, which is crucial for plasma heating and sustaining the reaction

conditions required. Hence the objective of the present study is to elucidate the influence of a

hole in the toroidal current on the transport behaviour of fast ions in a tokamak. Based on a

simple analytical current profile model [3] for axisymmetric tokamak equilibria, we

characterize completely the possible orbit topologies of fast ions and determine the

confinement domains for the different types of ion orbits. The trajectorial alterations induced

by the presence of a current hole as well as the consequences for fast ion transport are

calculated and illustrated. Further the fast ion distribution function is computed in the

constants-of-motion space using a Fokker-Planck collision operator. Finally, for a specific

JET current hole scenario where beam ions are injected on axis into near stagnation orbits,

we can derive analytically the stationary distribution fb of NBI ions [4] in satisfactory

agreement with numerical Fokker-Planck simulations.

[1] V.A. YAVORSKIJ, et al., Nucl. Fusion 43, (2003) 1077

[2] V.A. YAVORSKIJ, et al., Nucl. Fusion 44, (2004) L5-L9

[3] K. SCHOEPF, et al., 31st EPS Conf. Pl. Ph. Contr. Fus., London, June/July 2004,

ECA Vol. 28B, P-5.124 (2004)

[4] P. NEURURER, Fast ion confinement in a current hole tokamak, Diploma Thesis,

Institute for Theoretical Physics, University of Innsbruck, Austria (2004)

P-1.112, Monday June 27, 2005

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Modelling of Plasma Rotation under the Influenceof Helical Perturbations in TEXTOR

A. Nicolai1, U. Daybelge

2, C. Yarim

2

1Institut fur Plasmaphysik, Forschungszentrum Julich GmbH, Euratom

Association, Trilateral Euregio Cluster, D-52425 Julich, Germany

2Istanbul Technical University, Faculty of Aeronautics and Astronautics,

80626 Maslak, Istanbul, Turkey

ExB shear flow may be one of the possible reasons for the formation of a transport

barrier leading to the H - mode. Therefore it has stimulated a widespread interest

and is under investigation, both experimentally and theoretically.

The ambipolarity constraint and the parallel momentum equation of the revisited

neoclassical theory allow to predict the parallel and poloidal flow speeds and therefore

the radial electric field via the usual radial momentum balance equation.

The theory also accounts for the friction with the recycled neutral gas due to charge

exchange.

In contrast to the ohmically heated ALCATOR with a very short gradient length of

the temperature profile, in the L - mode TEXTOR and JET plasmas an analogous

temperature pedestal is not observed and the momentum input is dominated by NBI

rather than by the temperature gradient term as in the ALCATOR - case.

To account for the turbulence prevailing in the L - mode, the perpendicular viscosity

can be replaced by an anomalous one.

The shearing rate of the velocity field can be strongly influenced by a localized braking

or accelerating force. Thus this force can be a possible mean for an ITB.

In particular, the friction force due to helical perturbations with the mode - numbers

m, n is dominant if this perturbation is resonant along the field lines (q=m/n).

The DED - coils /1/ of TEXTOR provide revolving helical perturbations predomi-

nantly at the q=3 surface in the standard (m=12, n=4) - configuration but also at

the q=2 surface in the (m=3, n=1) configuration due to the larger penetration depth

of the (m=2, n=1) mode. Since according to /2/ the relative velocity between the

magnetic field structure and the plasma enters in the momentum source term, a large

braking or accelerating effect in the plasma edge can be expected, if the plasma and

the rotating field are synchronized. In the momentum balance also the friction due

to the eddy currents in the wall is taken into account.

The main results can be summarized as follows:

Due to the braking term /2/ a local minimum can be formed with a large velocity

shearing rate S at both sides of the minimum. However, since the maximum rotation

speed at the plasma center is reduced, the necessary shearing rate may not be reached.

Accelerating the plasma with CO - NBI (720 kW) and a corotating DED (frequency

velocity gradient ofdvt

dr= 1.3 10

6 1

secmay be generated which should be sufficient to

suppress the ITG - instability.

/ 1/ S. S. Abdullaev, K. H. Finken, K. H. Spatschek Plas. 6 (1999) 153

/ 2/ R. Fitzpatrick, Nucl. Fusion 33, 1049 (1993)

P-1.113, Monday June 27, 2005

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Modelling of the penetration process of externally applied magnetic

perturbation of the DED on TEXTOR

Y. Kikuchi1,a, K.H. Finken1, D. Reiser1, G. Sewell2, M. Jakubowski1, M. Lehnen1 1 Institut fuer Plasmaphysik,Forschungszentrum Juelich GmbH,

D-52425 Juelich, Germany 2 Institute of Mathematics, University of Texas at El Paso, El Paso, USA

a JSPS Postdoctoral Fellowships for Research Abroad

The dynamic ergodic divertor (DED) experiment has been started on TEXTOR tokamak

[1]. The DED can apply not only static but also rotating magnetic perturbation fields with

a frequency of up to 10 kHz. In this paper, the penetration process of the static and

rotating magnetic perturbation fields of the DED into tokamak plasmas has been

investigated by numerical simulations based on the reduced set of one-fluid, resistive and

viscous MHD equations in a cylindrical geometry. The computational domain is divided

into 4 sections: Section #1 represents the plasma, #2 is the vacuum between the plasma

and the DED coil, #3 represents the DED coil and #4 is the vacuum outside the DED coil.

The equations are linearized and expanded in Fourier form in toroidal and poloidal

directions. Here the perturbation fields are assumed to be a single mode. In addition, an

equilibrium poloidal field and a toroidal rotation velocity are prescribed which is iterated

by a quasi-linear approach. In the present time-dependent and one-dimensional problem,

the differential equations were numerically solved by PDE2D code (finite element solver)

with boundary conditions.

The stability analysis of the model shows a good agreement with conventional

stability index of tearing mode F’ criterion and the dependence of the perturbed fields on

the resistivity and viscosity of the plasma. When the critical magnitude of the magnetic

perturbation is exceeded, the plasma rotation velocity drops and large magnetic islands

appear in the quasi-linear model. This is one candidate to explain that the m/n = 2/1

tearing type mode was triggered in the case of the m/n = 3/1 DED configuration in the real

experiment when the DED coil current was above the critical threshold [2].

[1] K.H. Finken, et al., Phys. Rev. Lett. 94 (2005) 015003.

[2] H.R. Koslowski, et al., Proc. 31st EPS Conf., London (2004) P1. 126.

P-1.114, Monday June 27, 2005

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Stellarator scaling considering uncertainties in machine parameters

R. Preuss1, E. Ascasibar2, A. Dinklage1, V. Dose1, J.H. Harris3, A. Kus1,

S. Okamura4, F. Sano5, U. Stroth6, J. Talmadge7 and H. Yamada4

1 Max-Planck-Institut für Plasmaphysik, EURATOM Association, Germany

2 CIEMAT, Madrid, Spain

3 Australian National University, Canberra, Australia

4 National Institute for Fusion Science, Toki, Japan

5 Institute of Advanced Energy, Kyoto University, Uji, Kyoto, Japan

6 University of Kiel, Kiel, Germany

7 University of Wisconsin, Madison, USA

The International Stellarator Confinement Database (ISCDB(1)) is examined in order to de-

rive scaling expressions for the confinement time. We present a thorough discussion of the

uncertainties of entries to the scaling expressions, i.e. the machine parameters. Uncertainties

in the machine parameters not only increase by virtue of the error propagation law the mea-

surement uncertainty of the quantity of interest but should be incorporated in the whole data

analysis process right from the beginning. To achieve this we employ two methods: an error in

variables technique and a probability theoretical approach, i.e. Bayesian inference. Both meth-

ods are compared and the evolving scaling expressions are discussed with respect to former

results.

(1): URL of ISCDB: http://www.ipp.mpg.de/ISS and http://iscdb.nifs.ac.jp/

P-1.115, Monday June 27, 2005

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Role of sto hasti ity in W7-X edge transportD. Sharma, Y. Feng, F. SardeiMax-Plan k-Institut fur Plasmaphysik, IPP-Euratom Asso iationTeilinstitut Greifswald, Wendelsteinstr. 1, D-17491 Greifswald, GermanyThe energy transport in the edge of W7-X for topologies with oexisting reg-ular ux surfa es, losed islands and sto hasti regions is simulated with the3D Monte-Carlo plasma edge transport ode EMC3-EIRENE using realisti geometries for wall, divertor targets and baes. The radial plasma tempera-ture proles are obtained for a large range of anomalous ross-eld heat diu-sion oeÆ ient. It is observed that the widths of both the radial temperatureproles and the deposition patterns on the divertor target elements shrinksteadily with redu ing the ross-eld ondu tivity and no residual diusionfrom eld-line sto hasti ity appears in the limit of small diusion oeÆ ient.The energy transport seems to behave in a manner that the sto hasti zone lose to the main separatrix would onsist of regular magneti surfa es. Theanalysis learly indi ates that the sto hasti energy transport to the targetsin typi al W7-X ongurations remains marginal ompared to the dominant ollisional pro esses.

P-1.116, Monday June 27, 2005

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Effect of Alfvén resonances on the penetration of error fields on a rotating

viscous plasma

R. Coelho1 and E. Lazzaro

2

1Associação EURATOM/IST, Centro de Fusão Nuclear, 1049-001 Lisbon,CODEX, Portugal

2Istituto di Fisica del Plasma del CNR, Assoc. EURATOM-ENEA-CNR per la Fusione,

Via R.Cozzi 53, Milan,Italy

Abstract

The penetration of the intrinsic magnetic error field of a tokamak in the confined

plasma (and subsequent amplification) may eventually lead to the plasma disruption. When

the plasma rotates, the plasma response to the static error field is influenced by the presence

of a pair of Alfvén resonances located around the rational q-surface where magnetic

reconnection is to take place. These two Alfvén resonances may potentially shield the external

magnetic perturbations when the plasma is highly conducting (Reynolds number S>>107-8

)

and rotating at speeds above 4kHz. While extremely relevant for safe ITER plasma operation

(avoiding amplified locked modes), experimental evidence, however, suggests that this

resonance pair has little effect in preventing reconnection at the q=2 and the subsequent onset

of the hazardous locked mode (that may disrupt the plasma). In this work we investigate the

plasma response to a static m=2,n=1 error field component in both inviscid and viscous

rotating plasmas. Both inertial and electromagnetic induced forces are essential to account for

the overall plasma response to the static fields. A plasma rotation threshold is found

(depending on plasma viscosity), separating two different regimes: one where the Alfvén

resonances shield the penetration of the external magnetic field and there is negligible

reconnection and the other where forced reconnection dominates and an island is formed and

grows with time.

P-1.117, Monday June 27, 2005

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Advanced Reversed Field-Pinch Confinement Scaling Laws J.-E. Dahlin, J. Scheffel

Alfvén Laboratory, Royal Institute of Technology, Stockholm, Sweden

A series of resistive magnetohydrodynamic numerical simulations are performed to

generate scaling laws for magnetic fluctuations, energy confinement time τE and poloidal

beta βθ for the advanced reversed field-pinch (RFP). Strongly improved scaling with basic

initial parameters is obtained as compared to the conventional RFP1. Early results indicate

an improved scaling of τE with initial temperature T0 compared to the conventional RFP on

the order of τE (adv.) / τE (conv.) ∝ T0. The improved behaviour of the advanced RFP stems

from the introduction of current profile control (CPC)2,3. In the present numerical

simulations, CPC is performed by implementation of a parameter free automatic feedback

algorithm, optimised to reduce the fluctuation caused ×v B electric field3. The scheme

introduces an ad-hoc electric field within the plasma volume, automatically adjusted to

dynamically control the plasma into more quiescent behaviour by eliminating current

driven tearing mode instabilities and reducing resistive interchange modes.

[1] J. Scheffel and D. D. Schnack, Phys. Rev. Lett. 85 (2000) 322.

[2] C. R. Sovinec and S. C. Prager, Nucl. Fusion 39 (1999) 777.

[3] J.-E. Dahlin et al, 31st EPS Conference on Plasma Physics 28G (2004) P-5.193.

P-1.118, Monday June 27, 2005

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A Uniform Framework to Study Resistive Wall Modes

Yueqiang Liu

Department of Applied Mechanics, EURATOM/VR Fusion Association,

Chalmers University of Technology, Goteborg, Sweden

Toroidal simulations using the MARS-F code have shown that the plasma response, due

to the linear stability of resistive wall modes (RWM), can be well described by frequency-

dependent low order rational functions. These transfer functions, defined as the ratio of the

sensor signals with and without the plasma, fully describe the linear plasma response to an

externally applied magnetic field. We introduce the transfer functions in the same way for

both unstable and stable plasmas. This enable us to study, in a uniform framework, both

feedback control of the RWM and the resonant field amplification (RFA), for rotating and

non-rotating plasmas. For feedback study, the obvious advantage of introducing transfer

functions is that one decouples the plasma dynamic from the controller design, making the

controller design more flexible. In RFA experiments, these transfer functions are easily

constructed from the measured signals with traveling waves excitation.

We demonstrate how these transfer functions can be obtained and used for the RWM study,

in both analytical theory and toroidal calculations. Three cases are considered.

1. From the Fitzpatrick-Aydemir model (or similarly, Chu’s model), one can construct

transfer functions describing unstable plasma response to feedback signals, as well as

(rotationally stabilized) stable plasma response to external error fields. These effec-

tively single mode models give qualitative understanding of the combined effect of

plasma rotation and feedback for the RWM stabilization.

2. Cylindrical theory, developed by Bondeson and Liu, gives transfer functions for mul-

tiple modes. The poles and residues distribution of transfer functions leads to physical

interpretation on why internal poloidal sensors are superior than radial sensors for

feedback control of the RWM.

3. MARS-F computations result in low order transfer functions, which can be viewed

as lumped model of the RWM dynamics. We show that multiple modes (at least two

or three poles) are necessary to describe the response of unstable plasmas, whereas

stable plasmas (e.g. RFA) can often be well described by single mode, especially at

low frequency ranges.

P-1.119, Monday June 27, 2005

Page 125: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

An Improved Fluid Description on Toroidal ITG Modes

A.K.Wang

Southwestern Institute of Physics, Po.Box 432, Chengdu 610041, P.R.China

Abstract

In this paper, the toroidal ion temperature gradient (ITG) modes are studied from a set

of model equations,

,])(

[

])(

)()([)(

2//

2//

*2

**

i

isDi

e

spisDee

i

ipi

P

pck

T

eckk

n

n

fvy

fhytyyfy

Y--

Y-/Y//?-Y `

(1)

,3

5]

)(2[

3

5

])(

3

5)()(

3

5[)(

2//

2//

*2

*1

*

i

iDi

i

isDi

e

spisDepi

i

ipi

n

n

P

pck

T

eckk

P

p

fyfvy

fhytyyvfy

/Y

--

Y//Y--/?-Y `

/

(2)

where , , and . In addition, we have

, , , , and

. Compared with conventional fluid approach, the present includes the

background drift, . The diamagnetic drift frequency in the left

hand sides of Eqs.(1) and (2) originates from the background drift retained,

2/1)/( ies mTc ?

iVk ©/ Di ?y

ne eBLT /`

iV ?

eBmT ies /)( 2/1?t

eBRTk i /2 ` Dey

idiE //VV --

ei TT /?v

eBRe /?Y y

e k* ?y

/

V

Tk2 `? eipi ** )1( yjvy -/?

pi*y

pidiiiE *// )( yyy -Y?©-©/?-©/ VkVkVVk , (3)

which is the main different point of the present work from the traditional approach. Here the

induces only a Doppller shift but the will acts on the ITG instability. Eqs.(1) and

(2) reduce to the Eqs.(A.15) and (A.16) in the appendix of Ref.(1), respectively, if the

introduced in this paper and the perturbed parallel motion of ions are deleted and meanwhile

is taken. Based on the present model, the properties of ITG modes and the critical

stability thresholds are investigated numerically and compared with the previous fluid and

kinetic results.

iVk ©

y?Y

pi*y

pi*y

[1]. M. Fröjdh, P. Strand, J. Weiland et al., Plasma Phys.Control.Fusion 38,325 (1996).

P-1.120, Monday June 27, 2005

Page 126: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Methodology of electron Bernstein wave emission simulations

J. Urban1, J. Preinhaelter

1, V. Shevchenko

2, G. Taylor

3, M. Valovic

2, P. Pavlo

1, L. Vahala

4,

G. Vahala5

1 EURATOM/IPP.CR Association, Institute of Plasma Physics, 182 21 Prague, Czech Rep.

2 EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon, OX14 3DB UK

3 Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543, USA

4 Old Dominion University, Norfolk, VA 23529, USA

5 College of William & Mary, Williamsburg, VA 23185, USA

Electron Bernstein wave (EBW) emission, allowed by mode conversion to ordinary and

extraordinary waves, is typical for devices with high-density plasmas and low magnetic field,

such as spherical tokamaks (ST) and stellarators.

Our studies are primary focused on EBW emission from spherical tokamaks. To interpret

experimental results we have developed a simulation code, which consists of an antenna

model, an EBW-X-O conversion efficiency computation and a ray tracing. A realistic 3D

model of ST plasma is used in the simulations. The instantaneous magnetic field and its

spatial derivatives are reconstructed from a 2D splining of two potentials determined by an

equilibrium reconstruction code (usually EFIT). The plasma density and temperature profiles

in the whole poloidal cross-section of the plasma are obtained from a mapping of the

Thomson scattering measurements to the magnetic surfaces. The antenna model consists of a

horn, a thin lens, mirrors and a vacuum window. The Gaussian beam formalism is used to

solve the wave propagation between the radiometer and the plasma.

Simulation results for MAST and NSTX spherical tokamaks, compared to the experiments,

are presented at the end of the paper. Possible utilizations of EBW emission, e.g. electron

temperature measurement or magnetic field reconstruction, are discussed.

P-1.121, Monday June 27, 2005

Page 127: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

A Novel ST Configurative Events with Controllable and Reproducible

Alternative Self-organization Process

S. Sinman1 and A. Sinman2

1Middle East Technical University, Electrical and Electronics Engineering Department,2Turkish Atomic Energy Authority, Nuclear Fusion Laboratory, Ankara, Turkey

The aim of this study is to identify the physical bases of an alternative self-organization

mechanism that exists on the STPC-EX machine [1, 2] and to determine complementary

features with respect to present compact toroid concepts. The operational properties of last

version of STPC-EX is modified with new the dense plasma creation method such as the

stepping discharge (STPD). The conventional pulsed discharge, the gas pressure, the pulse

height, the pulse duration and the repetition rate are the basic collective dischage parameters

of a gas breakdown, whereas in the STPD procedure Fig.1(a)), these collective discharge

parameters are not necessary. The demonstration of spherical tokamak plasma (STP)

creation using the spherical snowplough (SSP) by dual-axial z-pinch (DAZP) and/or self-

reversed field pinch combined with DAZP (Fig.1(b)) are presented. The spherical tokamak

plasma in the envelope of SSP is shaped relating to the m = 0 mode of DAZP. In this

procedure, the basic objects to be characterised at the conventional STP are controlled by the

principal structural geometry of the STPC-EX setup and previously selected reference data

of the current-launcher. The main points achieved in this study are: aspect ratio = 1.2 - 1.6;

averaged beta = 0.46 - 0.62; elongation = 4 - 6; triangularity = 0.42 - 0.58; sustainment time

= 4.3 - 6.5 ms; energy confinement time = 45 - 136 ms; plasma temperature = 118-177 eV .

(a) (b)

Time (1) (2)

FIG.1 (a) Typical stepping discharge oscillogram taken from the STPC-EX set-up at the final phase,showing the toroidal magnetic field versus time, BT(t). Time-scale: 3.5 ms/div. and vertical-scale:0.07 T/div. (b) Typical photographic result of stepping discharge taken from the STPC-EX set-up atthe final phase, showing the perfectly self-created spherical tokamak plasma (1: the formedmagnetic piston, (2): the compressed plasma current channel)).

References: S.Sinman and A.Sinman [1] Sorrento, IC/P-04], [2] Lyon, IC/P-01.

BT versus Time

P-1.122, Monday June 27, 2005

Page 128: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Drift waves in the TORPEX toroidal plasma device

B. Labit, A. Fasoli, M. McGrath, S. Müller, G. Plyushchev, M.Podestà and F.M. Poli

CRPP - EPFL, Association EURATOM-Confédération Suisse, 1015 Lausanne, Switzerland.

In a toroidal plasma, a wide variety of gradient driven instabilities, generally referred to as

drift waves, can be linearly unstable. Their nonlinear evolution can lead to electrostatic tur-

bulence and cross-field particle and energy transport. It istherefore important to identify the

conditions under which drift waves occur and to reconstructthe spatio-temporal evolution of

the related fluctuations.

Local measurements of plasma density and floating potentialfluctuations are performed us-

ing Langmuir probes across the whole plasma cross section ofthe TORPEX toroidal device.

The microwave power controls the density gradient. The ion mass and the neutral gas density,

varied by acting on the gas injection rate, determine the relative importance of collisional pro-

cesses: Coulomb collisions or ion and electron collisions with neutrals. The parallel connection

length can also be varied by changing the vertical magnetic field. A peak in the frequency spec-

trum of density fluctuations around 4 8kHz, in the range of frequencies expected for drift

waves driven by the density gradient, is observed in the plasma region where the gradients of

density and magnetic field are co-linear. The parallel and perpendicular wavenumbers and the

form of the dispersion relationω(k) are evaluated experimentally together with the local tur-

bulence induced particle flux. These observations are interpreted on the basis of a linearised

two-fluid model[1], which includes the main ingredients fordrift wave turbulence: a density

gradient, the magnetic curvature, the parallel dynamics and collisions with neutral particle. The

flute limit, corresponding to a vanishing parallel wavenumber, can also be studied. The effect

of the neutral collision frequency on the mode stability andthe phase shift between density and

potential fluctuations, predicted to vary between 0 andπ=2 as the collisionality is increased, is

studied experimentally by varying the neutral gas density.As a complement to the wave char-

acterisation, an attempt of imaging the plasma fluctuationsis performed by applying the Con-

ditional Average Sampling technique to probe measurementsover the whole TORPEX cross

section. Spatio-temporal patterns of the density fluctuations can be reconstructed, including

wave fronts and possible turbulent macroscopic structures.

This work is partly funded by theFonds National Suisse pour la Recherche Scientifique.

References

[1] O.E. Garcia, J. Plasma Physics,65, pp 81-96 (2001)

P-1.123, Monday June 27, 2005

Page 129: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Experimental studies of plasma production and transport mechanisms in

the toroidal device TORPEX

M.Podestà, A.Fasoli, B.Labit, M.McGrath, S.Müller, G.Plyushchev and F.M.Poli

CRPP-EPFL, Association Euratom-Confédération Suisse, Lausanne – Switzerland

The mechanisms of plasma production and transport are studied in TORPEX, a toroidal

device with major and minor radii R=1m and a=0.2m. Currentless plasmas of Argon,

Hydrogen and other noble gases are obtained using microwaves at f=2.45GHz in the

electron-cyclotron range of frequency, injected from the low-field side with O-mode

polarisation. Typical values of plasma densities and electron temperatures are n~1016–

1017m-3 and Te~5-10eV. A small vertical field, BZ~1mT, is superimposed to the dominant

toroidal magnetic field, Bφ~0.1T, to optimise the confinement time and the symmetry of the

plasma profiles.

The roles played by the electron-cyclotron and upper-hybrid resonances in the plasma

production mechanisms are investigated in discharges with modulation of the injected

microwave power. A fast modulation of the power allows one to separate the phenomena

directly related to the ionisation of the neutrals, characterised by a fast time-scale, from the

slower relaxation leading to the stationary profiles. The spatial profile of the particle source

can be recovered from the measurements and modelled, for example to use it as input for

numerical simulations of the plasma dynamics. Moreover, the dependence of the upper-

hybrid frequency on the density leads to a tight coupling between the plasma dynamics and

the absorbed microwave power, which in some experimental conditions manifests as large

amplitude coupled oscillations in the density and the absorbed power.

Along with the characterisation of the plasma production mechanisms and the resulting

plasma profiles, the transport properties of TORPEX plasmas will be investigated. The

response of the plasma to a modulation of the injected power can be analysed using Fourier

and System Identification analysis techniques to estimate the transport coefficients. The

results will be compared with the local properties of particle fluxes measured from a set of

electrostatic probes, including Mach probes and a four-tip probe configured to extract the

turbulent component of the particle flux.

This work is partly funded by the Fonds National Suisse pour la Recherche Sciéntifique

P-1.124, Monday June 27, 2005

Page 130: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Formation of Very Deep Potential Well with Electrode Biasing

in a Toroidal Device

T. Hiraishi1, Y. Fukuzawa

1, N. Ohno

2, S. Takamura

1

1Graduate School of Engineering, Nagoya University, Nagoya 464-8603, Japan 2Eco Topia Science Institute, Nagoya University, Nagoya 464-8603, Japan

Radial electric field Er and associated ExB plasma poloidal rotation are well known to

have an important role in tokamak plasma confinement. In addition, it has become clear

from both sides of experiment and theory that the formation of radial electric field is

deeply related to the L-H mode transition. The radial electric field is induced by a

cross-field current from an electrode located in the interior of the plasma. Biasing to a

cold electrode is not so efficient to generate the large Er with negative biasing. However,

a large amount of electron current driven by an arc discharge between hot

electron-emissive electrode inserted into the plasma center and the vacuum vessel have a

great effect on the formation of very deep electrostatic potential well [1,2]. In this

research, we aim to investigate the physical mechanism of the formation of deep

electrostatic potential well and its dynamics under a variety of experimental conditions.

The outline of the A.C. tokamak device, CSTN-IV is as follows: the major radius R =

0.4 m, the minor radius a = 0.1 m, the plasma current Ip < 1.5 kA, the toroidal magnetic

field BT < 0.13 T, the plasma density ne > 1.0x1018

m-3

, and the electron temperature Te <

15 eV. A small disc electrode made of LaB6 with the diameter of 6 mm and the thickness

of 0.5 mm is inserted into the plasma center. The negative biasing voltage is applied

between the electrode and the vacuum chamber at the flattop of Ip during 250"os. The

floating potential (near the center) was found to drop down to about –1.0 kV when the

radial arcing current flows. This value corresponds to more than 500 times as deep as the

electron temperature in CSTN-IV. The potential structure depends on the radial

resistance and the intensities of radial current. Therefore, we attempt to evaluate

experimentally the radial resistance under a variety of experimental conditions, and to

study the dependence of the resistance on several discharge parameters. Consequently,

we found that the radial resistance becomes large when the plasma current is small (< 500

A) and the toroidal magnetic field is strong (> 0.12 T), with a kind of bifurcation nature.

A large radial electric field may provide a strong poloidal ExB rotation, which would give

a good confinement instead of the poloidal magnetic field associated by the plasma

current to cancel the vertical charge separation due to the toroidal drift. That is, it is said

as an electrostatic potential confinement of the toroidal plasma.

In the conference, the relation among the radial resistance, the toroidal and the poloidal

magnetic field intensities will be reported in addition to the discussion on the electrostatic

confinement of toroidal plasma with a small poloidal magnetic field.

[1] S. Takamura, Y. Shen et al., Jpn. J. Appl. 25 (1986) 103.

[2] H. Kojima, Y. Fukuzawa, T. Manabe, S. Takamura, et al., Czech. J. Phys. 53 (2003) 895.

[3] Y. Fukuzawa, S. Takamura, et al., 31st EPS conference on Plasma Physics (2004).

P-1.125, Monday June 27, 2005

Page 131: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Ion dynamics in a collisionless magnetic reconnection experiment

A. Stark1, W. Fox2, J. Egedal2, O. Grulke1, T. Klinger1

1 Max-Planck Institute, Greifswald, Germany2 Massachusetts Institute of Technology, Cambridge, US

Recently, softX -ray emissions were observed in the solar corona during the occurrence of a

solar flare [1], indicating strong ion acceleration in certain regions of the flare. It is speculated

that such ion beams occur as a result of magnetic reconnection, the breaking and recombina-

tion of field lines. For a better understanding of the response of ions to magnetic reconnection,

controlled laboratory experiments are necessary. An experiment designed for studies of recon-

nection under (collisionless) conditions close to those found in astrophysical plasmas is the

Versatile Toroidal Facility (VTF) at the MIT Plasma Scienceand Fusion Center [2]. Poloidal

and toroidal magnetic field coils form a poloidal cusp-field with a toroidal guiding field. Recon-

nection is driven via a third toroidal solenoid. In this paper measurements of the ion velocity

distribution function (IVDF) parallel to theX -line obtained with laser-induced fluorescence dur-

ing magnetic reconnection are presented. It is demonstrated that the ion temperature increases

significantly if reconnection is driven. Furthermore it is shown that the ion temperature is pro-

portional to the reconnection rate. A time resolved analysis yields the evolution of the IVDF

within a reconnection cycle and reveals strong variations of the ion temperature during a recon-

nection cycle. Furthermore, a large non-thermal (beam) ionpopulation occurs at the maximum

reconnection rate, supposedly due to an inflow of plasma fromouter regions of the cusp field.

These findings are supported by measurements of the plasma flow with Mach probes.

References

[1] S. Masuda, T. Kosugi, H. Hara and Y. Ogawara Nature371, 495 (1994).

[2] J. Egedal at al., Rev. of Sci. Instrum.71, 3351 (2000).

P-1.126, Monday June 27, 2005

Page 132: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Eigen Modes of a Dielectr ic Loaded

Coaxial Plasma Waveguide

F.M.Aghamir1, 2

and M. N. Zandieh1

1) Dept. of Physics, University of Tehran,, Tehran, Iran

2) Plasma Physics research center, IAU, Tehran, Iran

Abstract

Electromagnetic radiation from a dielectric loaded coaxial plasma waveguide is studied

theoretically. High frequency Eigen modes of a dielectric loaded coaxial waveguide in the

presence of an annular plasma column is presented. The plasma column is assumed to be

under the influence of a uniform axial magnetic field so as to maintain its position inside the

waveguide. The dispersion equation is derived through the application of the appropriate

boundary conditions, which results in an eighth order determinant. The Eigen modes are

determined by equating this determinant to zero. In the presence of dielectric layer on the

conducting surfaces, the azimuthally symmetric modes have been identified as perturbed TM,

perturbed TE, cyclotron, and space charge modes for coaxial waveguide. Numerical solutions

are obtained for these four families of electromagnetic and electrostatic modes.

P-1.127, Monday June 27, 2005

Page 133: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Study of Gas Admixture Influences On The Pinch Dynamics In A 90 kJ

Filippov Type Plasma Focus

A.R. Babazadeh 1,2

, S.M. Sadat Kiai 2, M.V. Roshan

2

1 Faculty of science, Qom campus, Azad Islamic University, Qom, Iran

2 Dept. of physics, Qom University, P.O. Box 37165, Qom, Iran

Abstract

In this paper we present an experimental work concerning the effect of gas admixture on

the pinch dynamics in a Filippov type (25 kV, 288µF) plasma focus, DENA. Deuterium

pressure of 0.3 – 1.5 torr and krypton admixture of 0.5% - 3% by volume, have been used as

working gases. The main results have been obtained for the optimum pressure of deuterium

and deuterium + krypton. A study of the time-resolved pulsed neutron signals by the time of

flight technique made at angles of 0 and ヾ/2 radians show that the contribution of non-

thermal neutron production in the quiet phase of deuterium discharges is not considerable;

this is inconsistent for the Mather type plasma focuses.

Furthermore, the addition of krypton admixture to the deuterium working gas causes a sudden

increase in the non thermal neutron production. A survey of the experimental results presents

that the probabilities of multi-spike discharges are 75%and 20% with and without krypton

admixture, respectively. The dip of negative spike in the current derivative signal for the case

of krypton admixture is four times more than deuterium only gas discharges .The life time of

the pinches, measured in terms of current derivative, were 60-180 ns with the gas admixture

and 180 –200 ns without admixture discharges.

P-1.128, Monday June 27, 2005

Page 134: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Local Destruction of Magnetic Surfaces and

Impurity Distributions in Tokamak.

D.Kh.Morozov, V.A.Rantsev-Kartinov

INF RRC "Kurchatov Institute", Moscow, Russia, [email protected]

In this paper the actual problem of control by a radial profile (RP) of impurity dis-

tribution (ID) in tokamak is considered. As it has been shown, dynamics of magnetic sur-

faces and their structure may influence the RP of the ID significantly. The influence of

saw-tooth fluctuations (STF) on the dynamics of impurity carrying out from the tokamak

plasma core was considered in Ref. [1a,b]. It has been shown that the periodic reconnec-

tion of magnetic field lines leads to "washing away" (WA) of the impurity. It occurs that

efficiency of this process depends both on the STF frequency and the impurity atomic

number (the WA of heavy impurity is essentially higher). In this paper the stimulation of

the STF by the periodic (at time) electron cyclotron heating (ECH) of plasma near the cer-

tain resonance surfaces is suggested. Also, the influence of a magnetic field stochastiza-

tion near a separatrix on the WA of impurities out of the closed magnetic surface region is

considered. It has been shown [3], that the heavy impurity diffusion inside the plasma col-

umn may be decreased significantly (by the order of magnitude) with the magnetic field

stochastization near the separatrix. The latest may be realized by the toroidal symmetry

breakdown related to the discontinuity of the toroidal magnetic field coils. Joint considera-

tion of these effects can enable to find a method of an impurity profile operation which is

based on external influence on magnetic surfaces in any point on radius of the plasma col-

umn. The width of a reconnection zone as well as the process frequency may be con-

trolled. Especially, this method may be applied effectively for the fusion reactor where

some heavy elements (Mo, W, Re) are used as a construction elements. Effects considered

can be useful also for magnetic field structure and a current profile researches in tokamaks

by means of impurity spectroscopy.

REFERENCES

1. D.Kh. Morozov, V.A. Rantsev-Kartinov, a) Fizika Plasmy (Rep. Plasma Phys.), 20, No 12, p. 1051, (1994); b) Rev. Sci. Instrum., 66, No1, p. 505, (1995).

2. D.Kh.Morozov, V.A.Rantsev-Kartinov and J.J. Herrera, Phys. Plasmas, 2, No 5, p. 1540, (1995)

P-1.129, Monday June 27, 2005

Page 135: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Angular momentum coupling in tokamaks

E.A. Evangelidis1, G.J.J. Botha2

1 Demokritos University of Thrace, Xanthi, Greece2 University of Leeds, Leeds, United Kingdom

In the analysis of motion of a charged particle on a magnetic field line, Alfvén showed the

existence of a force

f = −

mu2⊥

κ

2N, (1)

P

b_

N_CCΩ

T_

B=BT__

with N the first normal of the co-moving trihedral given by (T,N,b)

andκ the curvature. HereT is the unit vector along the magnetic

field andb the binormal of the orthogonal system. In a reference

frame located at the centre of curvature (CC) and rotating with an

angular velocityΩ, a particle at pointP and moving in a circular

orbit, develops a centrifugal force−mΩ2ρN = −(mu2

‖/ρ)N with

ρ the distance fromP to the origin of the reference frame. When

combined with the force described by equation (1), this gives the

total force

ft = −

mu2⊥

κ

2N−

mu2‖

ρN = −mκ

(

u2⊥

2+u2

)

N. (2)

In a rotating reference system there exists also the Coriolis force, which is of no importance

here. The force acting on a charge at pointP produces the well known drift velocity

vd =

ft ×BeB2 =

1eB

ft ×T =

eB

(

u2⊥

2+u2

)

b. (3)

In a tokamak configuration the magnetic field lines lie on nested flux surfaces. With the toroidal

component the dominant field, one can consider a reference system at the origin of the major

radius rotating with an angular velocityΩ = Ω(R). Moreover,Ω decreases in such a way as

to accommodate the decrease ofu‖(R) asR increases. The dynamical problem of motion of a

particle in a rotating reference system with variable angular velocity gives the total force

ftot = −

(

u2‖+

u2⊥

2

)

N+

2Mz

m(∇Ω ) , (4)

where in this expression∇Ω = −|∂Ω/∂ρ| eρ , with eρ = −N. Hence the last term in equation

(4) shows the existence of an inwardly directed force due to the coupling of the differential

rotation (∇Ω ) with the angular momentumMz = mρ2Ω of the particle. The same considerations

for the plane of the poloidal cross section lead to the existence of a similar coupling for the

poloidal rotation of the plasma.

P-1.130, Monday June 27, 2005

Page 136: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Long term evolution of 3D

collisionless magnetic reconnection

D. Borgogno1, D. Grasso1, F. Porcelli1, F. Califano2, F. Pegoraro2

1 Burning Plasma Research Group, INFM, Politecnico di Torino, Italy2 Dipartimento di Fisica, Universita di Pisa, Italy

Abstract

The nonlinear behavior of reconnecting modes in three spatial dimensions (3D)

is investigated, on the basis of a collisionless fluid model in slab geometry, assuming

a strong constant guide field [1]. Unstable modes in the so-called large ∆′ regime are

considered. The nonlinear coupling of initial perturbations with different helicities

introduces additional helicities that evolve in time in agreement with quasilinear

estimates, as long as their amplitudes remain relatively small. Magnetic field lines

become stochastic when islands with different helicities are present [2]. In this paper

we present new results obtained simulating the reconnection process starting with a

Harris Pinch magnetic equlibrium configuration. We confirm the results concerning

the first nonlinear phase, obtained in Ref.[2] with a periodic equilibrium configu-

ration. The new equilibrium adoptded here allows us to extend the investigation

to the long term evolution phase. We show the spatial distribution and the time

evolution of the current density and vorticity structures that typically form in col-

lisionless regimes. On the basis of the definiton of the reconnection rate presented

in Ref. [2], we also present some speculations about the tendency of the system to

reach a saturated state.

References

[1] T.J. Schep, F. Pegoraro, B.N. Kuvshinov, Phys. Plasmas 1, 2843 (1994).

[2] D. Borgogno, D. Grasso et al., Phys. Plasmas 12,032309 (2005).

P-1.131, Monday June 27, 2005

Page 137: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Qualitative similar ities between edge localised modes (ELMs) in fusion

plasmas and complex space charge configurations (CSCCs) in

low-temperature plasmas

D. G. Dimitriu1, C. Ionita

2, R. Schrittwieser

2

1 Department of Plasma Physics, “Al. I. Cuza” University, Iasi, Romania 2 Institute for Ion Physics, Leopold-Franzens University, Innsbruck, Austria

The high confinement mode (H-mode) offers a promising regime of operation for a

tokamak plasma. H-mode operation is characterized by the formation of an edge transport

barrier (ETB), a thin layer with suppressed anomalous transport near the magnetic separatrix,

resulting in a steep edge density gradient (the so-called pedestal) and improved confinement.

The ETB generally features strong periodic bursts of particles and energy, referred to as edge

localized modes (ELMs) [1,2]. The energy impact on the plasma-facing components may lead

to an unacceptable heat load on the divertor. However, ELMs provide a mechanism by which

He ash and impurities can be removed from the plasma and the plasma can be regulated,

enabling stationary H-mode operation. For this reason, understanding and control of ELMs are

critical for the operation of next step devices such as ITER.

In low-temperature plasma it is well-known [3] that, under certain experimental con-

ditions, in front of a positively biased electrode immersed in plasma a complex space charge

configuration (CSCC) appears in form of an ion-rich plasma region confined by an electrical

double layer (DL). At a certain threshold value of the potential on this electrode, the CSCC

transits into a dynamic state, in which periodic disruptions and re-aggregations of the DL

occur, during which particles and energy are released into the surrounding plasma.

Here, we would like to present additional support for our thesis by emphasizing some

further obvious qualitative similarities between the behaviour of ELMs (especially dithering

ELMs and type I ELMS) and a low-temperature CSCC in the dynamic state. The experimental

observations shed new light on the complex physical mechanisms of these two phenomena.

References

[1] H. Zohm, Plasma Phys. Control. Fusion 38 (1996) 105;

[2] J. W. Connor, Plasma Phys. Control. Fusion 40 (1998) 191;

[3] B. Song, N. D’Angelo and R. L. Merlino, J. Phys. D: Appl. Phys. 24 (1991) 1789.

P-1.132, Monday June 27, 2005

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Diagnosis of Wire-Array Z-Pinch Implosion Using X-ray Framing Cameras

Z.P.Xu1, Z.H.Li1, R.K.Xu1, G.X.Xia1, F.Q.Zhang1, J.C.Chen1, J.L.Yang1, C.Guo1, J.M.Ning1,

L.B.Li1, F.J.Song1, K.N.Mitrofanov2 and E.V.Grabovski2

1 Institute of Nuclear Physics and Chemistry, P. O. Box 919–212, Mianyang 621900, China2 Troitsk Institute for Innovation and Thermonuclear Researches, Troitsk 142190, Russia

In the Sino-Russian joint Z-Pinch experiment on Angara-5-1 facility(3MA, 60ns) and in the

experiment carried out recently on QiangGuang-1 facility(1.5MA, 80ns), two x-ray framing

cameras, with gate time of about 2ns and 80ps, respectively, are employed to observe x-ray

distribution with rough energy resolution in the early stage and final stage of various wire-array

implosions. The frame photographs obtained by the nano-second gated framing camera indicate

no uniform plasma sheath is formed in the process. At early times, X-ray framing images show

that the foremost radiation comes from central part of array, and double well-defined radiation

rings, drifting to the anode and the cathode at 65 10 cm/s× and 72.4 10 cm/s× respectively, are often

produced near the electrodes. The frame photographs obtained by the pico-second gated

framing camera reflect the fast compression process around the x-ray peak emission and the

double-region compression process, and this provides an experimental clue to explain the

double-region compression phenomenon.

P-1.133, Monday June 27, 2005

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Elaboration of High-Current Dr ivers Aimed at the Iner tial Fusion Energy

Yu.Bakshaev, A.Bartov, P.Blinov, A.Chernenko, K.Chukbar, S.Dan’ko, G.Dolgachev,

L.Dubas, F.Fedotkin, Yu.Kalinin, A.Kingsep, A.Korelskiy, V.Korolev, D.Maslennikov,

V.Mizhiritsky, A.Shashkov, V.Smirnov, G.Ustroev

Russian Research Centre “Kurchatov Institute”, Moscow, Russia

The results of experiments on the S-300 pulsed power machine are presented, devoted to

the study of operation of co-axial magnetically self-insulated transporting line, by the linear

current flow density on the inner electrode surface up to j à 0.5 MA/mm. The specific

parameters of this current-carrying line correspond to those of the conceptual project of IFE

reactor based on the fast Z-pinch. The duration of efficient functioning for such a line has

been measured and possible reasons for broken isolation have been studied. Experiments

are gone on devoted to the energy transfer into the high-current tiny wire array with the

typical radius R Ã 1 mm, as well as study of its dynamics and analysis of the tiny Hohlraum

heating. The output device is based on the principle of the plasma flow switch operating in

the nanosecond range. Typical parameters of the experiment are as follows: I ~ 1 MA, k ~ 5

ns. The experimental activity is gone on aimed at the study of Plasma Opening Switch

(POS) use as the output unit sharpening the pulse of the next generation pulsed power

machines, in particular, of the MOL machine (4-6 MV, 3 MA, 100 ns), the test bed of the

“Baikal” IFE generator on the base of inductive energy storage. The multi-module POS

scheme with close packing of modules has been elaborated, the principles and conditions of

its synchronization have been checked experimentally [2, 3]. The results of subsequent

experiments are presented devoted to the POS operation ability in the conditions typical of

MOL or “Baikal”, to wit, 1) elimination of POS re-closing cutting off the load from the

inductive storage; 2) expansion of the conductivity phase before breaking the circuit in the

conditions of powerful pulse (~2 os, ~80% of energy) and long pre-pulse (~38 os, ~80% of

charge); 3) suppression of the intense axial plasma motion and thereby the conservation of

minimal POS length. The work was supported by the contract # 346778 “Sandia

Laboratories – Kurchatov Institute”, by the contract # 860 of the Russian Agency of

Atomic Energy, and, partially, by the Russian Foundation for Basic Research, grant 03-02-

16766, and by the President’s of Russian grant “Scientific school” NSH-2292-2003-2.

[1] S.A.Dan’ko et al, Proc. 31st EPS Conf. on Plasma Phys., London, 28 June – 2 July, ECA, V. 28B, O-1.17. [2] A.Kingsep et al, Proc. Int. Conf. “BEAMS’04”, St. Petersburg, July 2004, WE-O7-13. [3] A.Kingsep et al, Proc. 12th Int. Congress on Plasma Phys., Nice, France, 2004, O-D4-1.

P-1.134, Monday June 27, 2005

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Inertial Fusion Reactor Physics: effect of Activation and Radiation Damageof Materials, and Tritium emissions.

J.M. Perlado1, J. Sanz1,2, M. Velarde1, O. Cabellos1, C. Arévalo1, N. García-Herranz4,E. Martínez1, F. Mota1, S. Reyes3, M.J. Caturla5, J. Marian3, G. Velarde1, M. Victoria1,P. Cepas1, M.L. Gámez6

1. Instituto de Fusión Nuclear, Universidad Politécnica de Madrid, Madrid, Spain2. Department of Power Engineering, Univ. Nacional Educación a Distancia, Madrid3. Lawrence Livermore National Laboratory, Livermore CA, USA4. Department of Nuclear Engineering, ETSII, Univ. Politécnica de Madrid, Madrid5. Department of Applied Physics, Universidad de Alicante6. Department of Applied Physics, ETSII, Universidad Politécnica de Madrid, Madrid

Waste management assessment of different types of steels for the inertial fusion HYLIFE-II

reactor is performed. Hands-on and Remote recycling are unacceptable for steels in general.

304SS has a very good performance for shallow land burial (SLB), and both Cr-W ferritic

steels (FS) and, particularly, Oxide-Dispersion-Strengthened FS are very likely to be

acceptable. Two impurity elements question to obtain reduced activation (RA) steels

acceptable for SLB: Nb, Mo. Uncertainties in neutron induced long-lived activities in natural

elements from H to Bi due to activation cross section uncertainties are estimated for

HYLIFE-II and SOMBRERO vessel structures.

Data source terms emitted to the atmosphere in the current reactor designs indicate that the

elementary tritium (HT) can overcome up to one order of magnitude the effect of potential

releases of tritiated water vapour (HTO). For this reason, the analysis and evaluation of the

unknown chronic dose becomes indispensable. The behaviour of the tritium should be

simulated using two well differentiated studies: deterministic and probabilistic. Our

conclusion is that probabilistic studies provide the real dynamics of the process, and a

detailed study of each climatic variables becomes indispensable because it modifies the

concentration of HT.

Multiscale modelling of microstructure evolution of self-ion irradiated Fe will be compared

with experiments using TEM diagnosis, and effect of impurities, temperature and dose will

be reported. The stress-strain curve of FeCr steels under irradiation is calculated using

Molecular Dynamics (MD), and simple analytical models. A neutron source of enough

intensity and adequate energy spectrum is needed (IFMIF) which will be very specific in the

case of pulsed Inertial Fusion, as we claimed in past years. New experiments and modelling

(MD and Tight-binding MD) of radiation damage in SiC, C, and Silica amorphous glass will

be presented, and MonteCarlo diffusion of defects in hcp materials.

P-1.135, Monday June 27, 2005

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A practical nonlocal model for electron transport in magnetized

laser-plasmas

Ph. Nicolai, J.-L. Feugeas, G. Schurtz

Centre Lasers Intenses et Applications (UMR 5107)

Université Bordeaux 1, 33405 Talence cedex, France

In laser produced plasmas, the heat conduction plays a crucial role. Various processes, like

laser absorption, energy redistribution, ablation rate, parametric and hydrodynamic instabilities

could be directly or indirectly modified by the electron transport. The classical Spitzer-H ¨arm’s

flux does not allow to reproduce experimental data except forvery low laser intensities. It is

believed now that this problem is mainly due to the nonlocal nature of the heat conduction.

Nevertheless other mechanisms as self generated magnetic fields may modify and reduce elec-

tron transport too. The existent models are often 1D, which is not sufficient for interpretations

of many experiments. Therefore a nonlocal model has to be at least 2- even 3D. Last, this model

needs to be fast enough to be implemented into large multi-dimensional hydrodynamic codes.

The model described in this work aims at extending the formula of G. Schurtz, Ph. Nicolai and

M. Busquet1 to magnetized plasmas. A complete system of nonlocal equations is derived from

kinetic equations with self-consistent E and B fields. This equations are analyzed and simplified

in order to be implemented into large laser fusion codes and coupled to other relevant physics.

The model is applied to two problems. A simple one which demontrates the main features of the

model. A second one more realistic, which concerns the energy transfer in a laser configuration.

References

[1] G. Schurtz, Ph. Nicolai and M. Busquet, Phys. Plasmas7, 4238 (2000)

P-1.136, Monday June 27, 2005

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Evolution of Rayleigh-Taylor Instability with Arbitrary Density Profiles

Wenlu Zhang, Ding Li, and Huisan Cai

Department of Modern Physics, University of Science and Technology of China, Hefei

230026, China

A new analytical approach has been developed to investigate the evolution of

Rayleigh-Taylor (RT) instability by employing the intuitive time-expanding method. An

analytical criterion of RT instability, which is actually the squared value of the growth rate,

has been obtained valid for arbitrary density profiles and magnetic shears. It is shown that the

dependence of growth rate on the wavelength and density-gradient scale length is quite

different for varied density profiles. A steeper density distribution is accompanied with a

higher growth rate. As an example, a comparison between instability with a power-law

density distribution and that with exponential distribution has been made, and conclusions are

for small exponent sign, growth rate i of power-law distribution is greater than that of

exponent distribution when wave number is small or very large whereas for large exponent

sign, i of power-law is larger than that of the exponential for all the perturbation wave

numbers.

P-1.137, Monday June 27, 2005

Page 143: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Self-Generated Magnetic Field Distributions in Multiple-Beam

Produced Plasmas

M. Kaluza1,2

, P. Nilson1, L. Willingale

1, C. Kamberidis

1, M. S. Wei

1,

A. E. Dangor1, R. G. Evans

1, R. Kingham

1, M. Tatarakis

1, and K. Krushelnick

1

1Plasma Physics Group, The Blackett Laboratory, Imperial College, London, UK

2e-mail address: [email protected]

The importance of self-generated magnetic fields and heat-transport inhibition in ignition-

scale hohlraums is currently receiving much theoretical attention. In particular, the spatio-

temporal evolution of the self-generated magnetic fields and their effect on the plasma

evolution inside the hohlraum are not well understood. Megagauss-level magnetic fields,

attributable to the ee nT ∇×∇ mechanism, may be sufficiently large inside gas-filled

hohlraums to affect the electron energy distribution by magnetizing the electrons ( )1>ecτω

and reducing the thermal conductivity 221/1 ecτωκ +≈ , altering the x-ray emission and

uniformity inside the hohlraum, laser-beam propagation and pointing to the inner wall

surfaces, parametric instabilities, and beam filamentation.

Here, we report on recent measurements taken using the VULCAN laser facility at the

Rutherford Appleton Laboratory, wherein the blow-off plasma generated from planar Au and

Al solid targets was characterized. The targets were irradiated by single- and double-beam

configurations. The pulses at 1053 nm had a duration of 1 ns and were focused by f/10-

optics to an intensity of 1014-15

W/cm2.

X-ray emission from the interaction region was monitored using a filtered pinhole camera. A

synchronized, frequency-quadrupled (263 nm) probe beam of 10 ps duration was passed

transverse to the target surface. With this beam, the plasma density could be measured using

a modified Nomarski interferometer. Simultaneously, the spatial distribution of the magnetic

field in the plasma was obtained by looking at the Faraday rotation of the probe pulse. By

varying the delay of the probe pulse also the temporal evolution of the magnetic field

structures could be observed. Significant differences in the x-ray emission and the magnetic

field distribution between single- and double-beam configurations were observed.

In an upcoming experiment, the temperature distribution will be measured inside the blow-

off plasma by means of Thomson scattering of a focused 263 nm, 1 ns pulse-duration probe

beam. The scattered light is spectrally dispersed using a high-dispersion spectrometer and

temporally resolved using an optical streak camera.

P-1.138, Monday June 27, 2005

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Laser-driven flyer impact experiments on LULI 2000 laser facility

N. Ozaki,1

M. Koenig,1

A. Benuzzi-Mounaix,1

K. A. Tanaka,2, 3

W. Nazarov,4

T. Vinci,1

A. Ravasio,1

M. Esposito,5

S. Lepape,1

E. Henry,1, ∗

G. Huser,1, ∗

K. Nagai,2

and M. Yoshida6

1Laboratoire pour l’Utilisation de Lasers Intenses (LULI),

Ecole Polytechnique, 91128 Palaiseau Cedex, France

2Institute of Laser Engineering, Osaka University, Suita, Osaka 565-0871, Japan

3Faculty of Engineering, Osaka University, Suita, Osaka 565-0871, Japan

4Department of Chemistry, University of Dundee, Dundee DD14HN, United Kingdom

5Dipartimento di Fisica “G. Occhialini” and INFM, Universita di Milano-Bicocca, Italy

6National Institute of Advanced Industrial Science and Technology, Tsukuba, Ibaraki 305-8565, Japan

(Dated: February 18, 2005)

Flyer impact experiments have been performed using laser-driven shock waves at the Laboratoire

pour l’Utilisation de Lasers Intenses (LULI), Ecole Polytechnique. Laser-accelerated flyer technique

had been studied to access extremely high-pressures in materials due to the impact[1]. Additionally,

recent experiment has demonstrated very smooth pressure loading like isentropic compression with

a density-graded projectile (expanding plasma)[2]. However, the conditions of flyer and impacted

materials have not been sufficiently investigated.

In this experiments, three types of flyer targets; (i) simple metal flyer (aluminum single foil),

(ii) the multi-layered one[3], and (iii) high-Z metal buffered by low-density plastic foam[4], were

investigated. Typical shock-loaded material was fused-quartz plate. All diagnostics were optical:

the rear-side ones were two velocity interferometers and a self-emission measurements calibrated for

brightness temperature, on the transverse side we had a shadowgraphy diagnostic.

In the foam-buffered flyer targets, tantalum foils travelled 100 µm distance for ∼ 2 ns, the highest

averaged velocity reaching 55 km/s. Shock wave gradually accelerated in quartz by the flyer impact

was generated, and then the shock wave passing a distinct boundary to a conductive state was

in-time/in-situ observed. This flyer impact method is a way to produce very unique conditions in

equation-of-state (EOS) diagram of material.

[1] R. Cauble et al., Phys. Rev. Lett. 70, 2102 (1993).

[2] J. Edwards et al., Phys. Rev. Lett. 92, 075002 (2004).

[3] K. A. Tanaka et al., Phys. Plasmas 7, 676 (2000).

[4] M. Koenig et al., Appl. Phys. Lett. 75, 3026 (1999).

∗Present address: Commissariat a l’Energie Atomique (CEA), 91680, Bruyeres-leChatel, France

P-1.139, Monday June 27, 2005

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Optical investigation of flyer disk acceleration and collision

with massive target on the PALS laser facility

T. Pisarczyk1, S. Borodziuk1, N. N. Demchenko2, S. Yu. Gus’kov2, K. Jungwirth3,

M. Kalal4, A. Kasperczuk1, V. N. Kondrashov5, B. Kralikova3, E. Krousky3,

J. Limpouch3,4, K. Masek3, M. Pfeifer3, P. Pisarczyk6, K. Rohlena3, V. B. Rozanov2,

J. Skala3, and J. Ullschmied3

1 Institute of Plasma Physics and Laser Microfusion, 23 Hery St., 00-908 Warsaw, Poland 2 P.N. Lebedev Physical Institute of RAS, Leninskyi Ave. 53, 117 924 Moscow, Russia 3 PALS Research Center, AS CR, Na Slovance 3, 182 21 Prague 8, Czech Republic 4 Czech Technical University, FNSPE, Brehova 7, 115 19 Prague 1, Czech Republic 5 Troitsk Institute of Innovation and Thermonuclear Research, 142 190 Troitsk, Russia 6 Warsaw University of Technology, ICS, 15/19 Nowowiejska St., 00-665 Warsaw, Poland

To continue our investigation on crater formation [1] in different conditions, we

have carried out experiments with double targets consisted of a disk placed in front of a

massive target with spacing of 200 µm between them. Both elements of the targets were

made of Al. The 6 µm thick disks with a diameter of 300 µm were covered by thin

polyethylene foil (2.5 µm thick) to reduce X-ray radiation. The disks were supported by 10

µm diameter carbon fibers. The following disk irradiation conditions were used: laser

energy of 100 J, laser wavelength of 1.315 µm, pulse duration of 0.4 ns, and laser spot

diameter of 250 µm. To measure some plasma parameters and accelerated disk velocity a

three frame interferometric system was used. Efficiency of crater creation by a disk impact

related to that for a direct laser action was determined using crater parameters, which were

obtained by means of a crater replica technique.

The experimental results concern the two main stages: (a) ablative plasma

generation and disk acceleration and (b) disk impact and crater creation. Spatial density

distributions at different moments of plasma generation and expansion, flyer disk motion,

as well as shapes and dimensions of craters are shown. Discussion of the experimental

results on the basis of the 2-D theoretical model of a laser-solid target interaction is carried

out.

[1] S.Yu. Gus’kov, S. Borodziuk, M. Kalal, A. Kasperczuk, B. Kralikova, E. Krousky,

J. Limpouch, K. Masek, P. Pisarczyk, M. Pfeifer, K. Rohlena, J. Skala, J. Ullschmied: Quantum Electronics 34 (2004) 989

P-1.141, Monday June 27, 2005

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Numerical modelling of strong shock waves and craters for the experiments using single and double solid targets irradiated by high

power iodine laser (PALS)

S. Borodziuk1, N. N. Demchenko2, S. Yu. Gus’kov2, K. Jungwirth3, M. Kalal4,

A. Kasperczuk1, B. Kralikova3, E. Krousky3, V. N. Kondrashov5, J. Limpouch3,4,

K. Masek3, M. Pfeifer3, P. Pisarczyk6, T. Pisarczyk1, K. Rohlena3, V. B. Rozanov2,

J. Skala3, and J. Ullschmied3

1 Institute of Plasma Physics and Laser Microfusion, 23 Hery St., 00-908 Warsaw, Poland 2 P.N. Lebedev Physical Institute of RAS, Leninskyi Ave. 53, 117 924 Moscow, Russia 3 PALS Research Center, AS CR, Na Slovance 3, 182 21 Prague 8, Czech Republic 4 Czech Technical University, FNSPE, Brehova 7, 115 19 Prague 1, Czech Republic 5 Troitsk Institute of Innovation and Thermonuclear Research, 142 190 Troitsk, Russia 6 Warsaw University of Technology, ICS, 15/19 Nowowiejska St., 00-665 Warsaw, Poland Numerical modelling was aimed at simulation of successive events resulting from

interaction of laser beam – single and double targets. It was performed by means of the 2D

Lagrangian hydrodynamics code ATLANT-HE [1]. This code is based on one-fluid and

two-temperature model of plasma with electron and ion heat conductivity consideration.

The code has an advanced treatment of laser light propagation and absorption.

This numerical modelling corresponds to the experiment which was carried out with

the use of the PALS facility. Two types of planar solid targets, i.e. single massive Al slabs

and double targets consisting of 6 µm thick Al foil and Al slab were applied. These targets

were irradiated by the iodine laser pulses of two wavelengths: 1.315 and 0.438 µm. The

pulse duration of 0.4 ns and a focal spot diameter of 250 µm at a laser energy of 130 J were

used.

The numerical modelling allowed us to obtain more detailed description of shock

wave propagation and crater formation.

[1] A.B. Isakov, N.N. Demchenko, I.G. Lebo, V.B. Rozanov, & V.F. Tishkin, (2003).

2D Lagrangian code “ATLANT-HE” for simulation of plasma interaction with allowance for hot electron generation and transport. ECLIM 2002, Proc. SPIE 5228, 143-150.

P-1.142, Monday June 27, 2005

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Experimental characterization of a strongly coupled solid density

plasma generated in a short-pulse laser target interaction

G. Gregori, S. B. Hansen, H.-K. Chung , A. J. Mackinnon, M. H. Key, N. Izumi, J.

King, P. K. Patel, R. Shepherd, R. A. Snavely, S. C. Wilk, and S. H. Glenzer

University of California, Lawrence Livermore National Laboratory

We report high resolution K! spectra from 5 µm thick buried Cu foils illuminated at

laser intensities of 1018-1019 W/cm2 with 10-400 J in 0.4-10 ps pulse duration. In order

to keep the copper foil at solid density, a 1 µm Al protective layer was deposited on

both sides of the Cu foil. A high reflectivity Bragg crystal coupled to an image plate

detector was used to spectrally resolve the time integrated K! fluorescence induced

by the relativistic electrons generated by collective laser-plasma absorption at the

front surface of the target. By fitting the width of the experimental line spectra with

an average atom model which includes self-consistent solution for bound and free

electron wavefunctions and all the relevant line shifts from multiply ionized atoms,

we are able to infer time and spatially averaged electron temperatures (Te) and

ionization state (Z) in the foil. Our results show increasing values for Te and Z when

the overall mass of the target is reduced, indicating increased heating due to electron

reflection from the Debye sheath, which leads to enhanced coupling of the laser

energy into the target. In particular, we measure peak temperatures in excess of 200

eV with Z~13-14. For these conditions the ion-ion coupling constant is ∀ii~8-9, thus

suggesting the achievement of a strongly coupled plasma regime. Comparison with

emission features calculated with a fully relativistic multi-configuration atomic

structure code is used to assess the accuracy of our measurements to less than 20-40

eV.

This work was performed under the auspices of the U.S. Department of Energy by

University of California Lawrence Livermore National Laboratory under contract No.

W-7405-Eng-48.

P-1.143, Monday June 27, 2005

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Ion energy measurements in laser-generated plasmas at INFN-LNS and

PALS research centre

L. Torrisi1, S. Gammino1, L. Andò1, A.M. Mezzasalma1, A. Picciotto1, L. Laska2,

J. Krasa2, K. Rohlena2, J. Badziak3, P. Parys3, J. Wolowski3

1INFN-LNS, Catania, Italy and Università di Messina, Italy

2IP-ASCR, Prague

3IPPLM, Warsaw

Temperatures of pulsed laser-generated plasma have been measured at INFN-LNS of Catania

and PALS of Prague. In the first case the laser intensity was of the order of 1010 W/cm2 while

in the second case it reaches about 1015 W/cm2.

In both cases an ion energy analyser, using a controllable electrostatic deflection to measure

the energy-to-charge ratio, was employed in time-of-flight configuration. Ion energy

distributions and charge state distributions were measured along the direction normal to the

irradiated target.

The energy distributions depend on the laser intensity and on the ion charge state. At high

laser intensity different ion groups are emitted from the hot plasma due to different

mechanisms of production in the non-equilibrium phenomena investigated, such as self-

focusing, ionisation and recombination effects and hydrodynamic processes.

Experimental data show Boltzmann distributions which are shifted towards high energy

increasing the charge state. A so called Boltzmann-Coulomb-shifted function was employed

to fit the experimental data and to calculate the temperature-like parameters characterising a

mean energy of different ion groups ("ion temperatures") and the components of the ion

velocity due to thermal effects, hydrodynamic expansion and Coulomb interactions.

At INFN-LNS temperatures of the order of hundreds eV and charge states up to about 10+

were measured. At PALS the "ion temperatures" from 1 keV up to 80 keV and charge states

up to about 50+ were measured.

P-1.144, Monday June 27, 2005

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A b s o l u t e x r a y y i e l d s f r o m l a s e r i r r a d i a t e d G e d o p e d a e r o g e l t a r g e t sK . B . F o u r n i e r 1 , M . T o b i n 1 , J . F . P o c o 1 , K . B r a d l e y 1 ,C . A . C o v e r d a l e 2 , D . B e u t l e r 2 , M . S e v e r s o n 21 L a w r e n c e L i v e r m o r e N a t i o n a l L a b o r a t o r y , 7 0 0 0 E a s t A v e n u e , L i v e r m o r e , C A 9 4 5 5 02 S a n d i a N a t i o n a l L a b o r a t o r y , A l b u q u e r q u e , N MW e h a v e m e a s u r e d t h e p r o d u c t i o n o f h X Y 1 0 k e V x _ r a y s f r o m l o w _ d e n s i t y G e _ d o p e d a e r o g e lt a r g e t s a t t h e O M E G A l a s e r . T h e t a r g e t s w e r e 1 . 2 m m l o n g b y 1 . 5 m m d i a m e t e r b e r y l l i u mc y l i n d e r s f i l l e d w i t h G e _ d o p e d ( 2 0 a t o m i c p e r c e n t ) S i O 2 f o a m . T h e d o p e d _ f o a m d e n s i t y w a s 5o r 7 m g / c c . T h e s e t a r g e t s a r e a m a j o r a d v a n c e o v e r p r e v i o u s d o p e d a e r o g e l s [ 1 ] : i n s t e a d o fs u s p e n d i n g t h e d o p a n t i n t h e S i O 2 m a t r i x , t h e G e a t o m s , w i t h c h e m i s t r y s i m i l a r t o S i , a r ei n c o r p o r a t e d d i r e c t l y i n t h e m a t r i x . T h u s , t h e l e v e l o f d o p a n t i s i n c r e a s e d b y m o r e t h a n af a c t o r o f s i x . F o r t y b e a m s o f t h e O M E G A l a s e r ( o = 3 5 1 n m ) i l l u m i n a t e d t h e t w o c y l i n d r i c a lf a c e s o f t h e t a r g e t w i t h a t o t a l p o w e r t h a t a p p r o a c h e d 2 0 T W . T h e l a s e r i n t e r a c t i o n s t r o n g l yi o n i z e s t h e t a r g e t ( n e / n c r i t x 0 . 1 5 – 0 . 2 0 ) , a n d a l l o w s t h e l a s e r _ b l e a c h i n g w a v e t os u p e r s o n i c a l l y i o n i z e t h e h i g h _ Z e m i t t e r i o n s i n t h e s a m p l e . T h e h e a t i n g o f t h e t a r g e t w a si m a g e d w i t h a g a t e d ( 2 0 0 p s t i m e r e s o l u t i o n ) x _ r a y f r a m i n g c a m e r a , f i l t e r e d t o o b s e r v e > 8k e V . 2 _ D r a d i a t i v e _ h y d r o d y n a m i c c a l c u l a t i o n s p r e d i c t r a p i d a n d u n i f o r m h e a t i n g o v e r t h ew h o l e t a r g e t v o l u m e w i t h m i n i m a l e n e r g y l o s s e s i n t o h y d r o d y n a m i c m o t i o n . G e K _ s h e l l x _ r a ye m i s s i o n w a s s p e c t r a l l y r e s o l v e d w i t h a t w o _ c h a n n e l c r y s t a l s p e c t r o m e t e r a n d r e c o r d e d w i t ht e m p o r a l r e s o l u t i o n w i t h a s e t o f c a l i b r a t e d p h o t o c o n d u c t i v e d e v i c e s ( P C D s ) . T h ec a l c u l a t i o n s p r e d i c t 1 5 0 – 2 0 0 J o f x _ r a y e n e r g y o u t p u t w i t h h X Y 1 0 k e V . T h e e f f e c t o fs h a p i n g a n d d e l a y i n g t h e l a s e r p u l s e i s s t u d i e d . A f u l l d e s c r i p t i o n o f t h e e x p e r i m e n t a n d t h ep r e l i m i n a r y r e s u l t s o f o u r a n a l y s i s w i l l b e p r e s e n t e d . T h i s w o r k w a s p e r f o r m e d u n d e r t h ea u s p i c e s o f t h e U . S . D e p a r t m e n t o f E n e r g y b y U n i v e r s i t y o f C a l i f o r n i a L a w r e n c e L i v e r m o r eN a t i o n a l L a b o r a t o r y u n d e r c o n t r a c t N o . W _ 7 4 0 5 _ E n g _ 4 8 . S a n d i a i s a m u l t i p r o g r a ml a b o r a t o r y o p e r a t e d b y S a n d i a C o r p o r a t i o n , a L o c k h e e d M a r t i n C o m p a n y , f o r t h e U n i t e dS t a t e s D e p a r t m e n t o f E n e r g y u n d e r C o n t r a c t D E _ A C 0 4 _ 9 4 A L 8 5 0 0 0 .[ 1 ] K . B . F o u r n i e r e t a l . , P h y s . R e v . L e t t . 9 2 , 1 6 5 0 0 5 ( 2 0 0 4 )

P-1.145, Monday June 27, 2005

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Stopping Power Measurements for 100-keV/u Cu2+ Ions

in Ionized Matter

M. Basko1, G. Belyaev1, A. Fertman1, A. Golubev1, A. Kantsyrev1, V. Koshelev1,

A. Kuznecov2, R. Kuibeda1, T. Kulevoy1, T. Mutin1, V. Pershin1, I. Roudskoy1,

B. Sharkov1, G. Smirnov1, V. Turtikov1, S. Vysotskiy1 1 ITEP, Moscow, Russia

2 MEPhI, Moscow, Russia

Reliable data on stopping powers and energy losses for different ion-target

combinations is a hot topic for a wide variety of experiments in plasma physics,

atomic and nuclear physics as well as in target design for inertial confinement

fusion. The new experimental results on the low energy ion beam interaction

with hydrogen plasma are presented.

A new beam transport line for multi-charged heavy ions from the 27 MHz

ITEP RFQ accelerator to the target area has been designed and assembled. The

plasma generated by igniting an electric discharge in two collinear quartz tubes

of 6 mm in diameter and 79 mm long. The capacitor bank of 3 mF, discharged at

voltages 3 kV, produces the electric current of 3 kA in two opposite directions in

either of the two quartz tubes. Arial density of free electrons has been measured

by using the method of time resolved two-wavelength Mach-Zehnder

interferometry in axial direction ( 172 10maxfen l ≈ ⋅ cm-2).

The stopping powers of Nitrogen and Hydrogen gases for 100 keV/u Cu2+

ions have been experimentally determined:

( )NS 9 9 0 6exp . .= ± , ( )HydS 27 1 5 2exp . .= ± MeV/(mg/cm2).

A time-of-flight measurements is presented to determine the energy loss of

Cu2+ ions in partially ionized hydrogen plasma: E 3 4 0 7max . .= ±∆∆∆∆ keV/u for the

initial gas pressure 0.95 mbar. A significant increase of investigated plasma

stopping power compared with cold matter has been demonstrated.

This work is supported by RFBR-03-02-17226, CRDF BRHE Y2-P-11-07 and

IAEA Research Contarct No: 11637.

P-1.146, Monday June 27, 2005

Page 151: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Interaction of high-energy laser pulses with plasmas of different density

gradients

J. Wolowski1 , J. Badziak

1, S. Gammino

2, H. Hora

3, J. Krása

4, L. Láska

4, A. Mezzasalma

5,

P. Parys1, M. Pfeifer

4, K. Rohlena

4, M. Rosinski

1, L. Ryc

1, L. Torrisi

2,5, J. Ullschmied

4

1Institute of Plasma Physics and Laser Microfusion, Warsaw, Poland

2INFN-Laboratori Nazionali del Sud, Catania, Italy

3University of New South Wales, Sydney

4Institute of Physics and PALS RC, ASCR, Prague, Czech Republic

5Università di Messina, Messina, Italy

The characteristics of the laser-produced plasma depend, among other factors, on

distribution of the electron density during interaction of high-energy laser radiation with

expanding plasma. First of all, different interaction mechanisms depend on electron density

gradient in plasma, particularly, in the vicinity of a critical density (ncr). But at densities

lower than ncr the efficiency of collisional absorption, as well as stimulated Brillouin and

Raman scatterings (SBS and SRS) increase when the density gradient decreases. The SBS

process is dangerous for the efficiency of indirect laser fusion because it generates hot

electrons in the high-Z plasma produced by laser beams on the inner surface of the

Hohlraum capsule.

In this contribution we describe study of the influence of the electron density

gradient on laser-plasma interaction processes on the basis of measurements of

characteristics of ion streams and x-rays emitted from the plasma produced by a high-

energy PALS iodine laser system (operating at 438 nm wavelength). The change of the

electron density distribution was realized by generation of a pre-plasma in the front of the

target by a pre-pulse (~10 J in a 0.4-ns pulse) preceding the main heating pulse (~140 J in a

0.4-ns pulse) by 0 – 4.6 ns. The time-of-flight methods were used for diagnosis of ion

stream emitted from plasma. The x-ray emission was investigated with the use of

semiconductor detectors. It has been demonstrated that the maximum and mean energy of

the fast ions as well as the yields of both hard and soft x-rays attain highest values for the

delay times in the range of 0-1.2 ns and decrease for longer delay times. But for time delays

of 3.5-4.6 ns intense streams of fast ions expanding close to the target normal with lower

mean energy and energy spread were observed. The possible laser-plasma interaction

mechanisms responsible for these effects were analysed.

P-1.147, Monday June 27, 2005

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Thomson scattering of electron plasma waves stimulated by Ramanbackscattering in gasbag plasmas

S. Depierreux1, H.A. Baldis2, W. Seka3, J.D. Moody4,S.H. Glenzer4, R.K. Kirkwood4, R. Bahr3

1 Département de Conception et de Réalisation des Expérimentations, CEA-DIF, BP12,91680 Bruyères-le-Châtel, France

2 Department of Applied Science, University of California, Davis, CA 956163 Laboratory for Laser Energetics, University of Rochester, Rochester, NY 146274 Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, CA 94551

The stimulated Raman backscattering (SRS) instability is the decay of the incident

laser into an electron plasma wave (EPW) and a backscattered electromagnetic wave. It is

detrimental in the context of laser driven inertial confinement fusion as it can potentially

reflect a large amount of the incident laser energy and drive high amplitude EPWs that can

saturate by producing hot electrons.

The growth and saturation of SRS in the specific regime of high temperature

homogeneous plasmas as will be produced with the coming Laser MégaJoule (LMJ) and

National Ignition Facility (NIF) laser facilities has been subject of much study and still lacks a

complete description. It was indeed expected from linear theory that the high Landau damping

of the stimulated EPWs would prevent significant levels of SRS in this regime, but

experiments performed with similar levels of EPW’s Landau damping have measured large

amount of backscattered SRS light. These experiments were performed in gasbags or small

holhraum targets in order to reach high temperatures with the presently available laser

facilities. Due to the smaller sizes targets, the plasmas are less homogeneous and comprise

large hydrodynamic fluctuations that will not be present in the LMJ and NIF plasmas.

Previous experiments had no spatial resolution and were therefore unable to identify the

regions of the plasma that contribute to the measured global SRS backscattered light.

We have implemented a new Thomson scattering (TS) geometry for probing SRS-

driven EPWs on the 351 nm Omega laser facility at University of Rochester. This TS

diagnostic has been used to probe SRS EPWs with time, space and wavelength resolutions in

gasbags of Te ~ 1.5 keV and ne/nc ~ 5 % corresponding to kλDe between 0.35 and 0.5. We will

present experimental results obtained with this TS diagnostic and more especially discuss (i)

the spatial distribution of SRS EPWs in the gasbag and (ii) the features of EPWs driven at

high intensity.

P-1.148, Monday June 27, 2005

Page 153: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

High intensity B field generation in underdense plasmas

and the Inverse Faraday Effect

S. F. Martins1, R. A. Fonseca1, L. O. Silva1, F. Tsung2, W. B. Mori2

1 GoLP/CFP, Instituto Superior Técnico, Portugal

2 University of California Los Angeles, CA 90095, U.S.A.

Several physical mechanisms can generate high intensity magnetic fields in underdense plas-

mas. The diversity of these mechanisms and the difficulty in identifying the different phenomena

responsible for the measured fields have been source of strong controversy. One of the possible

mechanisms is the Inverse Faraday Effect (IFE), a magneto-optical effect, where a longitudinal

B field is generated by a circularly polarized beam propagating in a medium with free electrons.

In this work, we perform a comprehensive study of the IFE both with theory and simulations.

The model of [1] is extended to include, in the quasi-static approximation, the role played by

the longitudinal profile of the laser. We show that shorter laser pulse durations enhance the IFE

by factors of the order of unity for common lasers. The role of the ionization is also addressed.

Ionization leads to a different physical scenario that can generate stronger density gradients.

A new model for IFE is thus developed to study the influence of the ionization in the B field

generation. To test the model, an ionization module has been implemented in osiris 2.0 [2], for

different ionization models and with several ionization levels. Three-dimensional particle-in-

cell simulations confirm that IFE is stronger in the self-generated case than in the pre-ionized

case.

References

[1] Z. M. Sheng and J. Meyer-ter-Vehn, Phys. Rev. E 54, 1833 (1996).

[2] R. A. Fonseca et al., LNCS 2331, 342-351, (Springer, Heidelberg, 2002).

P-1.149, Monday June 27, 2005

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Stimulated Raman scattering with broadband effects

J. E. Santos, L. O. Silva

GoLP/CFP, Instituto Superior Técnico, Lisbon, Portugal

The Wigner formalism of quantum mechanics provides an alternative formulation to de-

scribe waves propagating in an inhomogeneous, dispersive and anisotropic medium, based on

the phase-space representation of wave fields [1]. However the wave equation describes a two

mode problem (incident and reflected waves), and all previous theoretical models only deal with

the single mode problem, where propagation is assumed to obey a Schrödinger-like equation.

A generalisation would allow for a direct connection with kinetic theory and would provide a

unique way to describe forward and backward scattering instabilities of broadband radiation

sources, with implications in laboratory and astrophysical scenarios.

We first build an alternative formulation to describe the laser propagation in a cold plasma

based on the Wigner formalism generalised to Klein-Gordon like-fields. We constructed a 2×2

Wigner matrix [2] on the basis of the Hamiltonian form of the Klein-Gordon equation of a

charged scalar particle field introduced by Feshbach and Villars [3]. The diagonal elements

describe forward and backward photon densities, and the off-diagonal elements correspond to

cross-densities in phase-space. In the corresponding quantum problem the mass is assumed to

be fixed, here a further generalisation is required to study a variable mass problem, since the

electron plasma frequency exhibits spacial and time dependences. The system of coupled trans-

port equations governing the evolution of the photon densities in phase-space is then derived.

The system of transport equations for the photons is coupled with the relativistic fluid equa-

tions for the plasma. The resulting dispersion relation holds for all values of a0. All results

regarding forward and backward stimulated Raman scattering are recovered. We then employ

the general dispersion relation to determine, from first principles and for the first time, the effect

of a broadband radiation spectrum on these instabilities.

References

[1] I. M. Besieris and F. D. Tappert, J. Math. Phys. 44, 2119 (1973); 14, 704 (1973); 14, 1829

(1973).

[2] O. T. Serimaa, J. Javanainen and S. Varró, Phys. Rev. A 33, 2913 (1986).

[3] H. Feshbach and F. Villars, Rev. Mod. Phys., 30, 24 (1958).

P-1.150, Monday June 27, 2005

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H2+ distributions after traversing plasma targets

M. D. Barriga-Carrasco

Universidad de Castilla-La Mancha, Ciudad Real, Spain

The energy loss of ion beams in plasmas is an important quantity for the ICF. For

atomic ions moving in plasmas, the energy loss is well understood based on various

theoretical models, such as the linear Vlasov-Poisson theory [1], the binary collision theory

[2], and the nonlinear Vlasov-Poisson theory [3]. For the slowing-down processes of

molecular ions in plasmas, however, it has been shown that the energy loss of an molecular

ion is strongly influenced by the interference resulting from spatial correlation among the

molecule constituent particles. This so-called vicinage effect on the energy loss of

molecular ions in plasma targets has been described theoretically by several authors [4]

within the framework of the linearized Vlasov-Poisson theory. But to date there is no

studies about these vicinage effects considering a full simulation of the transport of the

molecular ion.

Here we have performed computer simulations of the trajectory followed by the

protons resulting from the dissociation of H2+ molecules after traversing plasma targets. We

use dielectric formalism to describe the forces due to electronic excitations in the medium;

the self-retarding proton force and the wake force created by its partner proton. We also

consider the Coulomb repulsion between the pair of protons. Nuclear collisions with target

plasma nuclei are incorporated through a Monte Carlo code. The distributions of the energy

loss, exit angle, dwell time and internuclear separations of the proton fragments are

discussed for several target plasma densities and temperatures.

[1] T. Peter and J. Meyer-ter-Vehn, Phys. Rev. A 43, 1998 (1991).

[2] H. B. Nersisyan et al., Phys. Rev. E 67, 026411 (2003).

[3] G. Zwicknagel, Nucl. Instrum. Methods Phys. Res. B 197, 22 (2002).

[4] C. Deutsch and P. Fromy , Phys. Rev. E 51, 632 (1995).

P-1.151, Monday June 27, 2005

Page 156: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Heating of Tantalum Plasma for Studies on the Activation of the

6.238 keV Nuclear Level of Ta-181

R. Fedosejevs*,1, F. Gobet2, F. Dorchies1 , C. Fourment1, M.M. Aléonard2, G. Claverie2,

M.Gerbaux2, G. Malka2, J.N. Scheurer2, M. Tarisien2, F. Hannachi2, F. Blasco1, D.

Descamps1, G. Schurtz1, Ph. Nicolai1 and V. Tikhonchuk1

1 Centre Lasers Intenses et Applications, Université Bordeaux I, France

2 Centre d’Études Nucléaires de Bordeaux Gradignan, Université Bordeaux I, France

Previous reports [1] have indicated that the activation and decay of the 6.238 keV

nuclear level of 181Ta can be enhanced significantly in femtosecond laser heated tantalum

plasmas. The modifications are attributed to the high density plasma environment and high

ionization of the tantalum ions. Thus, an accurate understanding of the detailed plasma

conditions present in such an experiment are required to properly assess any expected

changes in activation and decay rates.

An experiment has been carried out to characterize a similar femtosecond heated

plasma and to estimate the isomeric excitation yield using the femtosecond laser system at

CELIA. The tantalum target was heated at 45 degrees angle of incidence using p-polarized 45

fs Ti:sapphire laser pulses at intensities of 1 to 6 x1016 W/cm2 . Measurements were carried

out of the ion emission using Langmuir probes and x-ray emission using both CCD and

NaI(Tl) pulse height detection systems. Detailed measurements were made of the preplasma

levels present in the experiment using the Langmuir probes. The deposition and implantation

of the escaping tantalum ions and atoms from the plasma onto a plastic collector foil was also

characterized using Rutherford Backscattering Spectrometry.

The experimental measurements of plasma conditions will be presented and compared

to analytical models and hydrodynamical calculations of the preplasma, plasma and

expansion. Details will then be presented of the deduced heating, ionization and plasma

expansion conditions for the heating of tantalum targets at these intensities and pulse length

and implications for the activation measurements.

1. A.V. Andreev et al., JETP 91, 1163-1175 (2000) * on leave from the University of Alberta, Edmonton, AB T6G2V4, Canada

P-1.152, Monday June 27, 2005

Page 157: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Stimulated Brillouin scattering driven by broadband radiation sources

L.O. Silva, J.E. Santos

GoLP/CFP, Instituto Superior Técnico, 1049-001 Lisboa, Portugal

The interaction of intense radiation with plasmas is a problem of paramount importance in a

wide range of scenarios. When the radiation pulse length is comparable or larger than the

typical time scale of the ion dynamics, not only stimulated Raman scattering (SRS) can

occur, but also stimulated Brillouin scattering plays an important role. This is even more

critical for conditions near the critical surface, relevant for ICF; for densities above nc/4,

SRS is suppressed and SBS is crucial to understand laser-plasma coupling.

In this work, we employ the formalism based on the Wigner description of the Klein-

Gordon equation (see J. E. Santos and L. O. Silva, this conference) to understand how the

broadband features of the pump laser determine the growth rate of SBS. This formalism is

based on a statistical description of the electromagnetic field, in the photon phase-space,

thus allowing for the description of arbitrary fields, with random statistics or not. We

explore the role played by a broadband pump field. We use the term “broadband” in a

general sense, to describe fields with a wide spectral content, and fields with an arbitrary

transverse wavenumber distribution.

For a monochromatic pump we recover the standard growth rates for SBS. Our model also

yields the generalized dispersion relation for SBS with an arbitrary statistics of the field.

The generalized dispersion relation is analyzed for simple photon distribution functions for

which analytical results can be derived. The consequences of our results for ICF and, in

particular, for the design of radiation beams capable of avoiding SBS are also outlined.

Work partially supported by FCT (Portugal) and DoE.

P-1.153, Monday June 27, 2005

Page 158: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Analysis of the propagation of a laser beam

through a preformed plasma using imaging diagnostics

K.Lewis1, 2, G.Riazuelo2, C.Labaune1

1 Laboratoire pour l'Utilisation des Lasers Intenses, UMR 7605 CNRS-Ecole Polytechnique-CEA-Université Paris VI, Ecole Polytechnique 91128 Palaiseau Cedex, France. 2 CEA DAM Ile-de-France, BP 12, 91680 Bruyères-Le-Châtel, France.

Propagation of an intense laser beam through an underdense CH plasma and

diagnostics used to analyze underlying experimental observations have been thoroughly

modeled using the laser plasma interaction code PARAX. Intensity distribution computed

in the plasma by the code cannot be directly compared to the observed intensity in the

experimental diagnostics. Before arriving on a diagnostic, light scattered by the plasma

undergoes nonlinear processes such as autofocalisation and filamentation, and propagates

through non-ideal optical components. Numerical simulations progressively include

propagation in the plasma, diagnostic’s modeling, and are finally compared with

experimental data. The convolution of spatially and temporally localized computed

magnitudes by the plasma and diagnostic’s transfer response enables a fruitful comparison

between simulations and recent measurements based on far field images of the transmitted

laser light through a preformed plasma.

The numerical code was developed at Commissariat à l’Energie Atomique. The

experimental results were obtained with the six beam facility of the Laboratoire pour

l’Utilisation des Lasers Intenses (LULI). The interaction beam which impinged on the

preformed underdense plasma was smoothed with a random phase plate. The intensity

distribution was observed with a streak camera (time resolved 1D images), and with gated

optical imagers (2D images with good temporal resolution).

P-1.154, Monday June 27, 2005

Page 159: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Exper imental multi-keV x-ray conversion efficiencies from laser

exploded germanium foil.

F. Girard1, J.P. Jadaud

1, M. Naudy

1, B. Villette

1, D. Babonneau

1, M. Primout

1,

M.C. Miller2, L.J. Suter

2, C. Constantin

2, J. Grun

3, J.F. Davis

4

1 CEA / DAM Ile de France, Bruyères le Châtel, France 2 LLNL, Livermore, USA

3 NRL, Washington DC, USA 3 Alme & Assoc., Alexandria, USA

Experiments with a thin foil irradiated with 2 laser pulses (one delayed in time) lead to

hot and underdense plasmas, which are efficient to produce multi-keV K-shell emission.

Previous works with prepulsed foils of titanium (Hec at 4.7 keV) and copper (Hec at

8.3 keV) showed high multi-keV x-ray conversion efficiencies. They are increased by a

factor of more than 2 in comparison with thick foils and are close to efficiencies

obtained with gas targets.

Exploded thin foil experiments have been performed on the OMEGA laser facility in

Rochester to quantify the multi-keV x-ray conversion from germanium targets. X-ray

power was measured by filtered diodes (DMX broadband spectrometer), which was fit

to the germanium K-shell emission around 10.3 keV. A conversion efficiency

enhancement by a factor of 2.2 is measured relatively to the case without pre-pulse

within the spectral bandwidth of 8 < hp < 10 keV.

P-1.155, Monday June 27, 2005

Page 160: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Periodic features modifying the He line profile from an aluminium plasma

Jon Howe, D. M. Chambers1, C. Courtois, E. Förster2, C. D. Gregory, I. M. Hall, J.

Howe, O. Renner3, I. Uschmann2 and N. C. Woolsey

Department of Physics, University of York, Heslington, York, YO10 5DD

1 AWE Ltd., Aldermaston, Berkshire, RG7 4PR

2Institute of Optics and Quantum Electronics, University of Jena, 07743 Jena,

Germany

3Institute of Physics, Academy of Sciences CR, 18221 Prague, Czech Republic

High-density laboratory based laser produced plasmas offer a wealth of interesting

phenomenon. X-ray line shapes emitted by highly ionised atoms give an insight into

the processes that occur in these plasmas. Using highly dispersive toroidal crystal

spectrometers (HDTS) it is possible spectrally resolve and spatially resolve these X-

ray spectral line shapes. The Al11+ Heβ (1s3p – 1s2), Heγ (1s4p – 1s2), and Heδ (1s5p

– 1s2) emission from a plasma created with 200 mJ, 800 nm laser pulse, stretched to

3.4 ps and focussed between 1014 W/cm2 and 1016 W/cm2 is studied in detail. Data

analysis, coupled with hydrodynamic simulations, is used to extract the electron

densities and temperatures of the plasma and to unfold the time integrated nature of

the spectroscopic measurements. In addition, the high resolution and high luminosity

spectrometer has enabled the measurement of unusual intensity modulations on Heβ

transitions. These modulations and their possible interpretation will be discussed.

P-1.156, Monday June 27, 2005

Page 161: EPS2005, Session P-1 AbstractsElectron heat transport dependence on plasma shape and collisionality in EC heated L-mode TCV plasmas P-1.053 R.O.Dendy Analysis of dissipation in MHD

Generation of Terahertz Radiation during Optical Breakdown of a Gas

V.B. Gildenburg, N.V. Vvedenskii

Institute of Applied Physics, Russian Academy of Sciences, Nizhny Novgorod, Russia

We consider the new method of generation of terahertz radiation (~1-10 THz)

based on the phenomenon of parametric conversion of more long-wavelength radiation in

time-varying plasma. As a concrete example we consider the interaction of millimeter

radiation with a long plasma column created during optical breakdown of a gas inside the

dielectric capillary tube or in a caustic of axicon lens. Based on the results of analytical

and numerical studies of the excitation and radiation of the free Langmuir oscillations in

the inhomogeneous laser-created plasma [1, 2] we define the range of optimal parameter

values of the scheme proposed (pressure of ionized gas, energy, duration and focusing

angle of an ionizing laser pulse, amplitude, frequency and polarization of the radiation

transformed) satisfying the conditions of the most effective generation of THz radiation.

The comparison of the results obtained with the results of Refs. [3-6], in which the

radiation of plasma oscillations excited in static electric field was considered, shows that

the conversion of the high-frequency wave can provide much more THz radiation

intensity and allows the wide control of its directivity.

This work was supported by RFBR (Grant No. 04-02-16684).

[1] V.B. Gildenburg, N.V. Vvedenskii, Phys. Plasmas, v. 8, p. 1953 (2001).

[2] N.V. Vvedenskii, V.B. Gildenburg, JETP Lett., v. 76, p. 380 (2002).

[3] W.B. Mori, T. Katsouleas, J.M. Dawson, C.H. Lai, Phys. Rev. Lett., v. 74, p. 542

(1995).

[4] D. Hashimshony, A. Zigler, K. Papadopoulos, Phys. Rev. Lett., v. 86, p. 2806 (2001).

[5] T. Loffler, H.G. Roskos, Journal of Applied Physics, v. 91, p. 2611 (2002).

[6] S.V. Golubev, E.V. Suvorov, A.G. Shalashov, JETP Lett., v. 79, p. 361 (2004).

P-1.157, Monday June 27, 2005