eps2005, session p-1 abstractselectron heat transport dependence on plasma shape and collisionality...
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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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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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 !"#$!
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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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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P-1.034, Monday June 27, 2005
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
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
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P-1.037, Monday June 27, 2005
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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P-1.073, Monday June 27, 2005
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
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
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
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P-1.077, Monday June 27, 2005
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
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
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
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
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
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
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
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P-1.086, Monday June 27, 2005
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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).
P-1.106, Monday June 27, 2005
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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 =
mκ
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 = −
mρ
(
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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
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).
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