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Sample records for asdex tokamak

  1. Investigation of magnetic modes in the ASDEX tokamak

    International Nuclear Information System (INIS)

    Properties of MHD-modes in the ASDEX Tokamak have been investigated by application and further development of the MIRNOV-diagnostics, i.e. measurement of magnetic field fluctuations. In addition to evaluation methods supported by models, also a model-independent statistical data analysis makes sense. The very important physics of mode locking, i.e. the slowing-down of rotating modes is examined. An elaborated theoretical model allows an interpretation of experimental results. Especially interesting is the loss of the angular momentum of rotating plasmas by mode locking. Experiments for mode stabilisation and prevention of electric current breakdown are discussed. Additional MHD-processes under different plasma conditions are treated on the fundament of the devloped model ideas. The author shows that the main tokamak plasma is described very well by one-dimensional models with cylindrical geometry, while the boundary zone of the plasma demands a more complex analysis. In the appendix a concept for the investigation of the MHD-activity in ASDEX-Upgrade is discussed. (AH)

  2. Investigation of limiter recycling in the divertor tokamak ASDEX

    International Nuclear Information System (INIS)

    A divertor experiment like the ASDEX tokamak is especially suited for studying ion recycling at a material limiter, because the plasma can alternatively be limited by a magnetic limiter (separatrix) or by a material limiter. The role of the material limiter in ion recycling is documented by observing the increase in charge exchange flux emitted at the limiter position, and the decrease in external gas input necessary to keep the plasma line density invariant, when the material limiter is moved to the plasma. Ion recycling occurs predominantly at the outside section of a ring limiter. The limiter material saturates shortly after the start of the discharge. About 60% of the total recycling occurs at the limiter, which is nearly 100% of the ion recycling. The remaining 40% of the total recycling is carried by charge exchange neutrals. Due to saturation, the recycling coefficient at the limiter is 1; the recycling coefficient of the charge exchange neutrals at the wall is approximately 0.5 giving rise to a total recycling coefficient of limiter discharges of 0.8-0.9. It is observed that the plasma resistivity increases when the material limiter is moved toward the separatrix. The increase in Zsub(eff) can tentatively be explained by proton sputtering. (orig.)

  3. Fast-ion losses induced by ACs and TAEs in the ASDEX Upgrade tokamak

    NARCIS (Netherlands)

    Garcia-Munoz, M.; Hicks, N.; van Voornveld, R.; Classen, I.G.J.; Bilato, R.; Bobkov, V.; Brambilla, M.; Bruedgam, M.; Fahrbach, H. U.; Igochine, V.; Jaemsae, S.; Maraschek, M.; Sassenberg, K.

    2010-01-01

    The phase-space of convective and diffusive fast-ion losses induced by shear Alfven eigenmodes has been characterized in the ASDEX Upgrade tokamak. Time-resolved energy and pitch-angle measurements of fast-ion losses correlated in frequency and phase with toroidal Alfven eigenmodes (TAEs) and Alfven

  4. Fast-ion transport induced by Alfvén eigenmodes in the ASDEX Upgrade tokamak

    DEFF Research Database (Denmark)

    Garcia-Munoz, M.; Classen, I.G.J.; Geiger, B.;

    2011-01-01

    A comprehensive suite of diagnostics has allowed detailed measurements of the Alfvén eigenmode (AE) spatial structure and subsequent fast-ion transport in the ASDEX Upgrade (AUG) tokamak [1]. Reversed shear Alfvén eigenmodes (RSAEs) and toroidal induced Alfvén eigenmodes (TAEs) have been driven u...

  5. An operational test of a time-of-flight analyser at the ASDEX tokamak

    International Nuclear Information System (INIS)

    The performance of a time-of-flight energy analyser, used for the investigation of the slowing-down spectrum of the heating beams at the ASDEX tokamak, is described. The time-of-flight analyser has a short flight path (15 cm). Its energy resolution amounts to a few per cent. The analyser was equipped with a preselecting, achromatic magnet system to separate the fast neutrals from thermal plasma particles and light

  6. Design of magnetic probes for MHD measurements in ASDEX tokamak

    International Nuclear Information System (INIS)

    The design of magnetic probes (Mirnov coils) is described in this report. These probes are used in ASDEX to investigate MHD modes and measure the plasma displacement together with magnetic flux loops. Concerning the high temperature rise during a plasma shot proper material for the coil form of the magnetic probes and the suitable wire and cable in the high vacuum chamber in conjunction with special geometrical construction have been selected. The electrical circuit updated to operate in a high noise environment is shown and first MHD mode signals demonstrate the effeciency of the system. (orig.)

  7. ASDEX upgrade - definition of a tokamak experiment with a reactor compatible polaoidal divertor

    International Nuclear Information System (INIS)

    ASDEX Upgrade is intended as the next experimental step after ASDEX. It is designed to investigate the physics of a divertor tokamak as closely as possible to fusion reactor requirements, without thermonuclear heating. It is characterized by a poloidal divertor configuration with divertor coils located outside the toroidal field coils, by machine parameters which allow a line density within the plasma boundary sufficient to screen fast CX particles from the plasma core, by a scrape-off layer essentially opaque to neutrals produced at the target plates, and, finally, by an auxiliary heating power high enough for producing a reactor-like power flux density through the plasma boundary. Design considerations on the basis of physical and technical constraints yielded the tokamak system optimized with respect to effort and costs as described in the following. It uses normal-conducting coil systems, is the size of ASDEX, and has a field of 3.9 T, a plasma current of up to 1.5 MA, and a pulse duration of 10 s. To provide the required power flux density, an ICRH power of 10 MW is needed. For comparison, a superconducting version is under investigation. (orig.)

  8. An equipment protection and safety system for the ASDEX tokamak

    International Nuclear Information System (INIS)

    Our compromise between safety requirements and costs is a hybrid of relay, solid-state and computer-controlled protection systems used for ASDEX. The toroidal field coils, ohmic heating coils, vertical field coils, divertor coils, radial field coils, stainless-steel vacuum vessel and structure are protected by measuring the water flow (131 channels), temperature (142 channels), mechanical displacements (141 channels), voltage symmetry (28 channels), current symmetry (6 channels), weight of the vessel (8 channels) and the overvoltage. To detect flow, temperature, displacement, voltage, current and weight, we use the following devices: Venturi tubes (self-made), RTD thermoresistors (Pt-100), linear potentiometers (1 kΩ), voltage dividers (self-made), Rogowski coils (self-made) and straing gauges. (orig.)

  9. Conceptual design of a Langmuir probe system for the tokamak ASDEX-UPGRADE

    International Nuclear Information System (INIS)

    The conceptual design of a Langmuir probe system for the tokamak ASDEX-UPG is presented. This system is intended to carry out electrostatic measurements, in space and time, on the boundary layer plasma over the largest possible volume of the divertor plasma during discharges. Conducted by preset design requirements a fast probe system is proposed. During discharges signal measurements will be performed by means of a data-acquisition system and the motion will be controlled by a real-time computer. The desired information concerning plasma parameters and the motion of the probe system will be available to the diagnostician via a video display unit. (author)

  10. Viewgraphs presented at the ASDEX/DOE workshop on disruptions in divertor tokamaks

    International Nuclear Information System (INIS)

    The emphasis of this year's ASDEX/DOE workshop was on disruptions in diverted tokamaks. The meeting was held here at MIT on 14--15 March. It is particularly appropriate that MIT hosted the workshop this year, since Alcator C-Mod had just recently completed its very first run campaign, and disruptions are one of the key areas of research in our program. There were a total of 14 speakers, with participants from IPP (Garching), CRPP (Lausanne), Culham, General Atomics, PPPL, Sandia, ORNL, the ITER JCT, and MIT. The subjects addressed included statistical analysis of disruption probabilities in ASDEX, modelling of the vertical axisymmetric plasma motion in DIII-D, impact of disruptions on the design of the ITER divertors, modelling of runaway electrons, and TSC calculations of disruption-induced currents and forces in TPX, etc. One item of particular interest to us was the experimental correlation of halo current magnitude with plasma current on ASDEX-Upgrade. The data indicates at least a linear, and possibly even a quadractic dependence. This has important implications for Alcator C-Mod, since it would predict halo currents of order 1 MA or more at full performance. At the conclusion of the talks, an informal discussion of disruption databases was held, primarily for the purpose of helping us develop a useful one for C-Mod

  11. Statistical analysis of the global energy confinement time in ohmic discharges in the ASDEX tokamak

    International Nuclear Information System (INIS)

    In ohmic discharges in all tokamaks at low plasma densities the global energy confinement time, τE, increases almost linearly with the density (LOC, linear ohmic confinement). In tokamaks with sufficiently large dimensions, τE saturates at a critical density (ASDEX bar ne- ≅ 3 x 1019 m-3) and is nearly constant at higher densities (SOC, saturated ohmic confinement). In the same density region some experiments report a further confinement regime for deuterium discharges in which τE exceeds the saturated value and is further increased (IOC, improved ohmic confinement). There the global energy confinement time roughly behaves as in the LOC regime. For both the LOC and the SOC regimes an isotope effect, i.e. the dependence of τ on the ion mass, is reported as an additional aspect of the ohmic energy confinement. A statistical analysis is performed to identify the parameters which are responsible for the properties of the energy confinement in these discharges in ASDEX. In contrast to earlier reports on confinement time scalings in ASDEX OH, only discharges with a full experimental description of kinetic electron and ion parameters, i.e. profiles of densities, temperatures and Zeff, are used to evaluate the energy contents of both species. By means of statistics it is shown that the characteristics of τE are mainly caused by the behaviour of the electron energy flux and the ohmic input power. The ion energy flux, does not play a significant role. Furthermore, the IOC regime is explained as a continuation of the low-density LOC regime. Both the isotope effect and the density dependence of τE are caused by features of the electron energy transport. (Author)

  12. Axisymmetric disruption dynamics including current profile changes in the ASDEX-Upgrade tokamak

    International Nuclear Information System (INIS)

    Axisymmetric MHD simulations have revealed a new driving mechanism that governs the vertical displacement event (VDE) dynamics in tokamak disruptions. A rapid flattening of the plasma current profile during the disruption plays a substantial role in dragging a single null-diverted plasma vertically towards the divertor. As a consequence, the occurrence of downward-going VDEs predominates over the upward-going ones in bottom-diverted discharges. This dragging effect, due to an abrupt change in the current profile, is absent in up-down symmetric limiter discharges. These simulation results are consistent with experiments in ASDEX-Upgrade. Together with the attractive force that arises from passive shell currents induced by the plasma current quench, the dragging effect explains many details of the VDE dynamics over the whole period of the disruptive termination. (author)

  13. Application of radial correlation doppler reflectometry on the ASDEX Upgrade tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Pinzon, J.R.; Stroth, U. [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstr. 2, D-85748 Garching (Germany); Physik-Department E28, TUM, D-85748 Garching (Germany); Happel, T. [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstr. 2, D-85748 Garching (Germany); Hennequin, P. [Laboratoire de Physique des Plasmas, Ecole Polytechnique (France); Collaboration: The ASDEX Upgrade Team

    2015-05-01

    Doppler Reflectometry (DR) is a diagnostic used for the characterization of plasma density turbulence in magnetic confinement devices. It allows to measure the perpendicular propagation velocity of density fluctuations and their perpendicular wavenumber spectrum with good spatial resolution. By studying the correlation between signals of two reflectometers probing at different radial positions (Radial Correlation DR), it is possible to evaluate the radial correlation length L{sub r} of the plasma turbulence by scanning the radial separation Δr. However, results from analytical calculations and two-dimensional full-wave simulations indicate that the L{sub r} measurement by RCDR is not straightforward and might depend on factors such as plasma velocity, fluctuation amplitudes and probing beam angle. Experimental data from the ASDEX Upgrade tokamak are studied. An assessment of the viability of the use of different signals and analysis methods, including an evaluation of potential caveats, is given.

  14. Poloidal asymmetries of the heavy ions in the ASDEX Upgrade tokamak

    International Nuclear Information System (INIS)

    Poloidal asymmetries of heavy ions in the tokamak plasma are caused by the presence of forces parallel with field-lines which have comparable magnitude to the thermal pressure. The most important examples are the centrifugal force (CF) and the electric force (EF). The CF is caused by fast toroidal rotation of the plasma column which is pushing impurity ions, that have a substantially higher mass than the main ions, on the outer-side of the plasma. And the EF can be produced by ion cyclotron heated fast particles with high pitch angle that are trapped by the mirror force on the low field side of the plasma. The excessive charge produced by these particles is affecting highly charged impurities and pushing them to the high field side of the plasma. From predictions based on neoclassical and turbulent theory, it follows that the radial flux of heavy ions will be significantly changed by the presence of these asymmetries. The purpose of this study is to investigate the presence of these asymmetries in ASDEX Upgrade and verify the predicted consequences on the particles flux. High intrinsic content of the tungsten in AUG plasma makes this device well suitable for such studies. Precise measurement of the SXR (soft-X-ray) radiation profiles has identified a presence of CF generated asymmetries in every NBI heated Asdex discharge. Poloidal asymmetry should than lead to the significant change in the neoclassical and turbulent radial transport of these heavy ions. High intrinsic content of the tungsten in Asdex plasma makes this device well suitable for studying these asymmetries. Precise measurement of the SXR (soft-X-ray) radiation profiles has identified a presence of CF generated asymmetries in every NBI heated Asdex discharge. For heavy and highly charged impurities multiple mechanisms exist that produce non-constant impurities densities on the flux surfaces. As for neoclassical and turbulent transport models such an asymmetry is of highly importance an effort is

  15. Poloidal asymmetries of the heavy ions in the ASDEX Upgrade tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Odstrcil, Tomas [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Physik-Department E28, Technische Universitaet Muenchen, Garching (Germany); Puetterich, Thomas; Angioni, Clemente; Bilato, Roberto; Gude, Anja; Vezinet, Didier [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Mazon, Didier [CEA, IRFM, Saint Paul-lez-Durance (France); Collaboration: ASDEX Upgrade Team

    2015-05-01

    Poloidal asymmetries of heavy ions in the tokamak plasma are caused by the presence of forces parallel with field-lines which have comparable magnitude to the thermal pressure. The most important examples are the centrifugal force (CF) and the electric force (EF). The CF is caused by fast toroidal rotation of the plasma column which is pushing impurity ions, that have a substantially higher mass than the main ions, on the outer-side of the plasma. And the EF can be produced by ion cyclotron heated fast particles with high pitch angle that are trapped by the mirror force on the low field side of the plasma. The excessive charge produced by these particles is affecting highly charged impurities and pushing them to the high field side of the plasma. From predictions based on neoclassical and turbulent theory, it follows that the radial flux of heavy ions will be significantly changed by the presence of these asymmetries. The purpose of this study is to investigate the presence of these asymmetries in ASDEX Upgrade and verify the predicted consequences on the particles flux. High intrinsic content of the tungsten in AUG plasma makes this device well suitable for such studies. Precise measurement of the SXR (soft-X-ray) radiation profiles has identified a presence of CF generated asymmetries in every NBI heated Asdex discharge. Poloidal asymmetry should than lead to the significant change in the neoclassical and turbulent radial transport of these heavy ions. High intrinsic content of the tungsten in Asdex plasma makes this device well suitable for studying these asymmetries. Precise measurement of the SXR (soft-X-ray) radiation profiles has identified a presence of CF generated asymmetries in every NBI heated Asdex discharge. For heavy and highly charged impurities multiple mechanisms exist that produce non-constant impurities densities on the flux surfaces. As for neoclassical and turbulent transport models such an asymmetry is of highly importance an effort is

  16. A compact lithium pellet injector for tokamak pedestal studies in ASDEX Upgrade.

    Science.gov (United States)

    Arredondo Parra, R; Moreno Quicios, R; Ploeckl, B; Birkenmeier, G; Herrmann, A; Kocsis, G; Laggner, F M; Lang, P T; Lunt, T; Macian-Juan, R; Rohde, V; Sellmair, G; Szepesi, T; Wolfrum, E; Zeidner, W; Neu, R

    2016-02-01

    Experiments have been performed at ASDEX Upgrade, aiming to investigate the impact of lithium in an all-metal-wall tokamak and attempting to enhance the pedestal operational space. For this purpose, a lithium pellet injector has been developed, capable of injecting pellets carrying a particle content ranging from 1.82 × 10(19) atoms (0.21 mg) to 1.64 × 10(20) atoms (1.89 mg). The maximum repetition rate is about 2 Hz. Free flight launch from the torus outboard side without a guiding tube was realized. In such a configuration, angular dispersion and speed scatter are low, and a transfer efficiency exceeding 90% was achieved in the test bed. Pellets are accelerated in a gas gun; hence special care was taken to avoid deleterious effects by the propellant gas pulse. Therefore, the main plasma gas species was applied as propellant gas, leading to speeds ranging from 420 m/s to 700 m/s. In order to minimize the residual amount of gas to be introduced into the plasma vessel, a large expansion volume equipped with a cryopump was added into the flight path. In view of the experiments, an optimal propellant gas pressure of 50 bars was chosen for operation, since at this pressure maximum efficiency and low propellant gas flux coincide. This led to pellet speeds of 585 m/s ± 32 m/s. Lithium injection has been achieved at ASDEX Upgrade, showing deep pellet penetration into the plasma, though pedestal broadening has not been observed yet. PMID:26931850

  17. A compact lithium pellet injector for tokamak pedestal studies in ASDEX Upgrade

    Science.gov (United States)

    Arredondo Parra, R.; Moreno Quicios, R.; Ploeckl, B.; Birkenmeier, G.; Herrmann, A.; Kocsis, G.; Laggner, F. M.; Lang, P. T.; Lunt, T.; Macian-Juan, R.; Rohde, V.; Sellmair, G.; Szepesi, T.; Wolfrum, E.; Zeidner, W.; Neu, R.

    2016-02-01

    Experiments have been performed at ASDEX Upgrade, aiming to investigate the impact of lithium in an all-metal-wall tokamak and attempting to enhance the pedestal operational space. For this purpose, a lithium pellet injector has been developed, capable of injecting pellets carrying a particle content ranging from 1.82 × 1019 atoms (0.21 mg) to 1.64 × 1020 atoms (1.89 mg). The maximum repetition rate is about 2 Hz. Free flight launch from the torus outboard side without a guiding tube was realized. In such a configuration, angular dispersion and speed scatter are low, and a transfer efficiency exceeding 90% was achieved in the test bed. Pellets are accelerated in a gas gun; hence special care was taken to avoid deleterious effects by the propellant gas pulse. Therefore, the main plasma gas species was applied as propellant gas, leading to speeds ranging from 420 m/s to 700 m/s. In order to minimize the residual amount of gas to be introduced into the plasma vessel, a large expansion volume equipped with a cryopump was added into the flight path. In view of the experiments, an optimal propellant gas pressure of 50 bars was chosen for operation, since at this pressure maximum efficiency and low propellant gas flux coincide. This led to pellet speeds of 585 m/s ± 32 m/s. Lithium injection has been achieved at ASDEX Upgrade, showing deep pellet penetration into the plasma, though pedestal broadening has not been observed yet.

  18. Recharging of the ohmic-heating transformer by means of lower-hybrid current drive in the ASDEX tokamak

    Science.gov (United States)

    Leuterer, F.; Eckhartt, D.; Söldner, F.; Becker, G.; Bernhardi, K.; Brambilla, M.; Brinkschulte, H.; Derfler, H.; Ditte, U.; Eberhagen, A.; Fussman, G.; Gehre, O.; Gernhardt, J.; Gierke, G. V.; Glock, E.; Gruber, O.; Haas, G.; Hesse, M.; Janeschitz, G.; Karger, F.; Keilhacker, M.; Kissel, S.; Klüber, O.; Kornherr, M.; Lisitano, G.; Magne, R.; Mayer, H. M.; McCormick, K.; Meisel, D.; Mertens, V.; Müller, E. R.; Münich, M.; Murmann, H.; Poschenrieder, W.; Rapp, H.; Ryter, F.; Schmitter, K. H.; Schneider, F.; Siller, G.; Smeulders, P.; Steuer, K. H.; Vien, T.; Wagner, F.; Woyna, F. V.; Zouhar, M.

    1985-07-01

    Recharging of the Ohmic-heating transformer of a tokamak by means of lower-hybrid waves is demonstrated experimentally in ASDEX. The results are analyzed on the basis of a simple transformer circuit. A recharging efficiency is defined and found to depend on rf power, plasma density, and plasma resistivity modified by the applied rf power. Up to now, we achieved in our recharging experiments in ASDEX a flux swing of FİOHMdt=0.24 V sec, at an rf power of PRF=690 kW, with a pulse duration of 1 sec, while maintaining a plasma with n¯e=4×1012 cm-3 and Ip=290 kA.

  19. Inter-ELM evolution of the edge current density profile on the ASDEX Upgrade tokamak

    International Nuclear Information System (INIS)

    The sudden decrease of plasma stored energy and subsequent power deposition on the first wall of a tokamak device due to edge localised modes (ELMs) is potentially detrimental to the success of a future fusion reactor. Understanding and control of ELMs is critical for the longevity of these devices and also to maximise their performance. The commonly accepted picture of ELMs posits a critical pressure gradient and current density in the plasma edge, above which coupled magnetohydrodynamic (MHD) peeling-ballooning modes are driven unstable. Much analysis has been presented in recent years on the spatial and temporal evolution of the edge pressure gradient. However, the edge current density has typically been overlooked due to the difficulties in measuring this quantity. In this thesis, a novel method of current density recovery is presented, using the equilibrium solver CLISTE to reconstruct a high resolution equilibrium utilising both external magnetic and internal edge kinetic data measured on the ASDEX Upgrade (AUG) tokamak. The evolution of the edge current density relative to an ELM crash is presented, showing that a resistive delay in the buildup of the current density is unlikely. An uncertainty analysis shows that the edge current density can be determined with an accuracy consistent with that of the kinetic data used. A comparison with neoclassical theory demonstrates excellent agreement between the current density determined by CLISTE and the calculated profiles. Three ELM mitigation regimes are investigated: Type-II ELMs, ELMs suppressed by external magnetic perturbations (MPs), and Nitrogen seeded ELMs. In the first two cases, the current density is found to decrease as mitigation onsets, indicating a more ballooning-like plasma behaviour. In the latter case, the flux surface averaged current density can decrease while the local current density increases, thus providing a mechanism to suppress both the peeling and ballooning modes.

  20. Adjoint Monte Carlo Simulation of Fusion Product Activation Probe Experiment in ASDEX Upgrade tokamak

    CERN Document Server

    Äkäslompolo, Simppa; Tardini, Giovanni; Kurki-Suonio, Taina

    2015-01-01

    The activation probe is a robust tool to measure flux of fusion products from a magnetically confined plasma. A carefully chosen solid sample is exposed to the flux, and the impinging ions transmute the material makig it radioactive. Ultra-low level gamma-ray spectroscopy is used post mortem to measure the activity and, thus, the number of fusion products. This contribution presents the numerical analysis of the first measurement in the ASDEX Upgrade tokamak, which was also the first experiment to measure a single discharge. The ASCOT suite of codes was used to perform adjoint/reverse Monte-Carlo calculations of the fusion products. The analysis facilitated, for the first time, a comparison of numerical and experimental values for absolutely calibrated flux. The results agree to within 40%, which can be considered remarkable considering the fact that all features of the plasma cannot be accounted in the simulations. Also an alternative probe orientation was studied. The results suggest that a better optimized...

  1. Transport analysis of high radiation and high density plasmas in the ASDEX Upgrade tokamak

    Directory of Open Access Journals (Sweden)

    Casali L.

    2014-01-01

    Full Text Available Future fusion reactors, foreseen in the “European road map” such as DEMO, will operate under more demanding conditions compared to present devices. They will require high divertor and core radiation by impurity seeding to reduce heat loads on divertor target plates. In addition, DEMO will have to work at high core densities to reach adequate fusion performance. The performance of fusion reactors depends on three essential parameters: temperature, density and energy confinement time. The latter characterizes the loss rate due to both radiation and transport processes. The DEMO foreseen scenarios described above were not investigated so far, but are now addressed at the ASDEX Upgrade tokamak. In this work we present the transport analysis of such scenarios. Plasma with high radiation by impurity seeding: transport analysis taking into account the radiation distribution shows no change in transport during impurity seeding. The observed confinement improvement is an effect of higher pedestal temperatures which extend to the core via stiffness. A non coronal radiation model was developed and compared to the bolometric measurements in order to provide a reliable radiation profile for transport calculations. High density plasmas with pellets: the analysis of kinetic profiles reveals a transient phase at the start of the pellet fuelling due to a slower density build up compared to the temperature decrease. The low particle diffusion can explain the confinement behaviour.

  2. Adjoint Monte Carlo simulation of fusion product activation probe experiment in ASDEX Upgrade tokamak

    International Nuclear Information System (INIS)

    The activation probe is a robust tool to measure flux of fusion products from a magnetically confined plasma. A carefully chosen solid sample is exposed to the flux, and the impinging ions transmute the material making it radioactive. Ultra-low level gamma-ray spectroscopy is used post mortem to measure the activity and, thus, the number of fusion products. This contribution presents the numerical analysis of the first measurement in the ASDEX Upgrade tokamak, which was also the first experiment to measure a single discharge. The ASCOT suite of codes was used to perform adjoint/reverse Monte Carlo calculations of the fusion products. The analysis facilitates, for the first time, a comparison of numerical and experimental values for absolutely calibrated flux. The results agree to within a factor of about two, which can be considered a quite good result considering the fact that all features of the plasma cannot be accounted in the simulations.Also an alternative to the present probe orientation was studied. The results suggest that a better optimized orientation could measure the flux from a significantly larger part of the plasma. A shorter version of this contribution is due to be published in PoS at: 1st EPS conference on Plasma Diagnostics

  3. Pellet refuelling of particle loss due to ELM mitigation with RMPs in the ASDEX Upgrade tokamak at low collisionality

    CERN Document Server

    Valovič, M; Kirk, A; Suttrop, W; Cavedon, M; Fischer, L R; Garzotti, L; Guimarais, L; Kocsis, G; Cseh, G; Plőckl, B; Szepesi, T; Thornton, A; Mlynek, A; Tardini, G; Viezzer, E; Scannell, R; Wolfrum, E

    2015-01-01

    The complete refuelling of the plasma density loss (pump-out) caused by mitigation of Edge Localised Modes (ELMs) is demonstrated on the ASDEX Upgrade tokamak. The plasma is refuelled by injection of frozen deuterium pellets and ELMs are mitigated by external resonant magnetic perturbations (RMPs). In this experiment relevant dimensionless parameters, such as relative pellet size, relative RMP amplitude and pedestal collisionality are kept at the ITER like values. Refuelling of density pump out requires a factor of two increase of nominal fuelling rate. Energy confinement and pedestal temperatures are not restored to pre-RMP values by pellet refuelling.

  4. Pellet refuelling of particle loss due to ELM mitigation with RMPs in the ASDEX Upgrade tokamak at low collisionality

    Science.gov (United States)

    Valovič, M.; Lang, P. T.; Kirk, A.; Suttrop, W.; Cavedon, M.; Cseh, G.; Dunne, M.; Fischer, L. R.; Garzotti, L.; Guimarais, L.; Kocsis, G.; Mlynek, A.; Plőckl, B.; Scannell, R.; Szepesi, T.; Tardini, G.; Thornton, A.; Viezzer, E.; Wolfrum, E.; the ASDEX Upgrade Team; the EUROfusion MST1 Team

    2016-06-01

    The complete refuelling of the plasma density loss (pump-out) caused by mitigation of edge localised modes (ELMs) is demonstrated on the ASDEX Upgrade tokamak. The plasma is refuelled by injection of frozen deuterium pellets and ELMs are mitigated by external resonant magnetic perturbations (RMPs). In this experiment relevant dimensionless parameters, such as relative pellet size, relative RMP amplitude and pedestal collisionality are kept at the ITER like values. Refuelling of density pump out of the size of Δ n/n∼ 30% requires a factor of two increase of nominal fuelling rate. Energy confinement and pedestal temperatures are not restored to pre-RMP values by pellet refuelling.

  5. The MHD stability analysis of type I ELMS in ASDEX Upgrade Tokamak

    International Nuclear Information System (INIS)

    The ELMs or edge localized modes are plasma instabilities localized in the edge region of a tokamak plasma. They cause periodic expulsions of particles and energy. The ELMs play a significant role in the confinement of the plasma, helium exhaust and diverter erosion. These are crucial issues in tokamak operation and, thus, understanding the underlying physical mechanism behind the ELM phenomenon is very important. The ELMs are classified into three different types based on the plasma conditions, where they are observed, and, on the ELM frequency response to the heating power. In this thesis, type I ELMs which are the most intense and the most damaging to the diverters, are studied. A model for the ELMs presented by Connor et al. is tested in experimental ASDEX Upgrade plasmas. In the Connor model, the ELMs are explained as a result of two instabilities, ballooning and peeling modes. Also a phenomenon called the bootstrap current plays a significant role by being the destabilising trigger to the peeling modes. The method used to study the model is MHD or magnetohydrodynamics. The theory of the ideal MHD equilibrium and the linear stability analysis is described. Inclusion of the bootstrap current to the equilibrium construction is introduced. The equilibria are created using experimental data from plasma shots that display type I ELMs. The stability analysis indicates that the investigated ELM model is a feasible explanation for type I ELMs. The pressure gradient near the plasma edge was found to be close to the ballooning stability boundary as predicted by the model. The peeling mode stability analysis confirms the prediction of the model that as the bootstrap current increases, the plasma becomes unstable for peeling modes with low to intermediate toroidal mode numbers. The mode numbers agree with the experimental results. In the experiments with high triangularity, low ELM frequency and ELM-free periods were observed. This indicates better stability of the plasma

  6. Report on the low-RF-power and data-acquisition-systems of the 2.45 GHz Lower-Hybrid Transmitter at the ASDEX tokamak

    International Nuclear Information System (INIS)

    This report relates to the low-power section of the 2.45 GHz transmitter used for current drive experiments at the ASDEX tokamak. The high-RF-power section is dealt with elsewhere. Data acquisition and evaluation of quantities pertaining to the Lower Hybrid experiment are also treated here. As in the previous report on the 1.3 GHz system (M. Zouhar: 'Beschreibung des Niederleistungsteils des HF-Systems fuer die LH-Experimente in ASDEX.', IPP-report 4/218, February 1984), most space is spent in commenting upon the amplitude (power)- and phase-feedback control loops. (orig.)

  7. Analysis of the H-mode density limit in the ASDEX upgrade tokamak using bolometry

    Energy Technology Data Exchange (ETDEWEB)

    Bernert, Matthias

    2013-10-23

    The high confinement mode (H-mode) is the operational scenario foreseen for ITER, DEMO and future fusion power plants. At high densities, which are favourable in order to maximize the fusion power, a back transition from the H-mode to the low confinement mode (L-mode) is observed. This H-mode density limit (HDL) occurs at densities on the order of, but below, the Greenwald density. In this thesis, the HDL is revisited in the fully tungsten walled ASDEX Upgrade tokamak (AUG). In AUG discharges, four distinct operational phases were identified in the approach towards the HDL. First, there is a stable H-mode, where the plasma density increases at steady confinement, followed by a degrading H-mode, where the core electron density is fixed and the confinement, expressed as the energy confinement time, reduces. In the third phase, the breakdown of the H-mode and transition to the L-mode, the overall electron density is fixed and the confinement decreases further, leading, finally, to an L-mode, where the density increases again at a steady confinement at typical L-mode values until the disruptive Greenwald limit is reached. These four phases are reproducible, quasi-stable plasma regimes and provide a framework in which the HDL can be further analysed. Radiation losses and several other mechanisms, that were proposed as explanations for the HDL, are ruled out for the current set of AUG experiments with tungsten walls. In addition, a threshold of the radial electric field or of the power flux into the divertor appears to be responsible for the final transition back to L-mode, however, it does not determine the onset of the HDL. The observation of the four phases is explained by the combination of two mechanisms: a fueling limit due to an outward shift of the ionization profile and an additional energy loss channel, which decreases the confinement. The latter is most likely created by an increased radial convective transport at the edge of the plasma. It is shown that the

  8. Analysis of the H-mode density limit in the ASDEX upgrade tokamak using bolometry

    International Nuclear Information System (INIS)

    The high confinement mode (H-mode) is the operational scenario foreseen for ITER, DEMO and future fusion power plants. At high densities, which are favourable in order to maximize the fusion power, a back transition from the H-mode to the low confinement mode (L-mode) is observed. This H-mode density limit (HDL) occurs at densities on the order of, but below, the Greenwald density. In this thesis, the HDL is revisited in the fully tungsten walled ASDEX Upgrade tokamak (AUG). In AUG discharges, four distinct operational phases were identified in the approach towards the HDL. First, there is a stable H-mode, where the plasma density increases at steady confinement, followed by a degrading H-mode, where the core electron density is fixed and the confinement, expressed as the energy confinement time, reduces. In the third phase, the breakdown of the H-mode and transition to the L-mode, the overall electron density is fixed and the confinement decreases further, leading, finally, to an L-mode, where the density increases again at a steady confinement at typical L-mode values until the disruptive Greenwald limit is reached. These four phases are reproducible, quasi-stable plasma regimes and provide a framework in which the HDL can be further analysed. Radiation losses and several other mechanisms, that were proposed as explanations for the HDL, are ruled out for the current set of AUG experiments with tungsten walls. In addition, a threshold of the radial electric field or of the power flux into the divertor appears to be responsible for the final transition back to L-mode, however, it does not determine the onset of the HDL. The observation of the four phases is explained by the combination of two mechanisms: a fueling limit due to an outward shift of the ionization profile and an additional energy loss channel, which decreases the confinement. The latter is most likely created by an increased radial convective transport at the edge of the plasma. It is shown that the

  9. Influence of Alfven eigenmodes and ion cyclotron heating on the fast-ion distribution in the ASDEX Upgrade tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Weiland, Markus; Geiger, Benedikt; Bilato, Roberto; Schneider, Philip; Tardini, Giovanni; Lauber, Philipp; Ryter, Francois; Schneller, Mirjam [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Collaboration: ASDEX Upgrade Team

    2015-05-01

    Fast, supra-thermal ions are created in the tokamak ASDEX Upgrade by neutral beam injection and ion cyclotron resonance heating (ICRH) and they are needed for plasma heating and current drive. A possibility to study them is the spectroscopic observation of line radiation (fast-ion D-alpha, FIDA), which emerges from charge exchange reactions. Here, the fast ions can be distinguished from the thermal particles through there strong Doppler-shift, and their radial density profile can be measured and compared to theoretical models. An analysis of the whole Doppler spectrum yields information about the 2D velocity distribution f(v {sub parallel}, v {sub perpendicular} {sub to}). Observation from different viewing angles allows consequently a tomographic reconstruction of f(v {sub parallel}, v {sub perpendicular} {sub to}). For this purpose, the FIDA diagnostic at ASDEX Upgrade has been extended from two to five views, and the spectrometer setup was improved to allow a simultaneous measurement of blue and red Doppler shifts. These recently developed diagnostic capabilities are used to study changes in the fast-ion distribution, which are caused by Alfven eigenmodes. Moreover, the further acceleration of fast ions through 2{sup nd} harmonic ICRH is investigated and compared to theoretical predictions.

  10. An optical scanning system for spectroscopic impurity flux investigations inside the ASDEX tokamak

    International Nuclear Information System (INIS)

    A scanning mirror system was developed to resolve impurity flux sources spatially across about 2/3 of the ASDEX surface by using visible spectroscopy. A totally computer-controlled layout allows wide-range spatial scanning during a discharge. Spectra over a range of ∝ 150 A are recorded with an integration time down to 20 ms. The versatility of this new system is illustrated by means of first observations of ASDEX discharges with additional heating (NI, LH, ICRH) and modulated gas puffing experiments. (orig.)

  11. Comparison of wall/divertor deuterium retention and plasma fueling requirements on the DIII-D, TdeV and ASDEX Upgrade tokamaks

    International Nuclear Information System (INIS)

    We present a comparison of the wall deuterium retention and plasma fueling requirements of three diverted tokamaks, DIII-D, TdeV and ASDEX Upgrade, with different fractions of graphite coverage of stainless steel or Inconel outer walls and different heating modes. Data from particle balance experiments on each tokamak demonstrate well-defined differences in wall retention of deuterium gas, even though all three tokamaks have complete graphite coverage of divertor components and all three are routinely boronized. This paper compares the evolution of the change in wall loading and net fueling efficiency for gas during dedicated experiments without helium glow discharge cleaning on the DIII-D and TdeV tokamaks. On the DIII-D tokamak, it was demonstrated that the wall loading could be increased by >1250 Torr l (equivalent to 150 x plasma particle content) plasma inventories resulting in an increase in fueling efficiency from 0.08 to 0.25, whereas the wall loading on the TdeV tokamak could only be increased by <35 Torr l (equivalent to 50 x plasma particle content) plasma inventories at a maximum fueling efficiency ∝1. Data from the ASDEX Upgrade tokamak suggests qualitative behavior of wall retention and fueling efficiency similar to DIII-D. (orig.)

  12. Interpretation of the effects of electron cyclotron power absorption in pre-disruptive tokamak discharges in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Tokamak disruptions are events of fatal collapse of the magnetohydrodynamic (MHD) confinement configuration, which cause a rapid loss of the plasma thermal energy and the impulsive release of magnetic energy and heat on the tokamak first wall components. The physics of the disruptions is very complex and non-linear, strictly associated with the dynamics of magnetic tearing perturbations. The crucial problem of the response to the effects of localized heat deposition and current driven by external (rf) sources to avoid or quench the MHD tearing instabilities has been investigated both experimentally and theoretically on the ASDEX Upgrade tokamak. The analysis of the conditions under which a disruption can be prevented by injection of electron cyclotron (EC) rf power, or, alternatively, may be caused by it, shows that the local EC heating can be more significant than EC current drive in ensuring neoclassical tearing modes (NTMs) stability, due to two main reasons: first, the drop of temperature associated with the island thermal short circuit tends to reduce the neoclassical character of the instability and to limit the EC current drive generation; second, the different effects on the mode evolution of both the location of the power deposition relative to the island separatrix and the island shape deformation lead to less strict requirements of precise power deposition focussing. A contribution to the validation of theoretical models of the events associated with NTM is given and can be used to develop concepts for their control, relevant also for ITER-like scenarios.

  13. Influence of plasma rotation on tearing mode stability on the ASDEX upgrade tokamak

    International Nuclear Information System (INIS)

    Neoclassical tearing modes (NTM) are one of the most serious performance limiting instabilities in next-step fusion devices like ITER. NTMs are destabilised as a consequence of a seed perturbation (trigger) and are driven by a loss of helical bootstrap current inside the island. The appearance of these instabilities is accompanied with a loss of confined plasma energy. Additionally, these modes can stop the plasma rotation, lock to the vessel wall, flush out all plasma energy and terminate a discharge via a disruption. In ITER the confinement reduction will limit the achievable fusion power, whereas a disruption is likely to damage the vessel wall. In order to mitigate and control NTMs in ITER, extrapolations based on the present understanding and observations must be made. One key issue is the rotation dependence of NTMs, especially at the NTM onset. ITER will be operated at low plasma rotation, which is different from most present day experiments. No theory is currently available to describe this dependence. Experiments are therefore required to provide a basis for the theory to describe the physics. Additionally from the experiments scalings can be developed and extrapolated in order to predict the NTM behaviour in the parameter range relevant for ITER. Another important issue is the influence of externally applied magnetic perturbation (MP) fields on the NTM stability and frequency. These fields will be used in ITER primarily for the mitigation of edge instabilities. As a side effect they can slow down an NTM and the plasma rotation, which supports the appearance of locked modes. Additionally, they can also influence the stability of an NTM. This interaction has to be predicted for ITER, based on models validated at present day devices. In this work the influence of plasma rotation on the NTM onset at the ASDEX Upgrade tokamak (AUG) is investigated. An onset database has been created in which the different trigger mechanisms have been identified. Based on this

  14. Untersuchung der Struktur und Dynamik magnetischer Inseln im Tokamak ASDEX Upgrade

    OpenAIRE

    Meskat, John Patrick

    2001-01-01

    Neoklassische Tearing Moden begrenzen das maximale beta in magnetisch eingeschlossenen Fusionsplasmen. In dieser Arbeit werden die Struktur und Dynamik von Tearing Moden und magnetischen Inseln in ASDEX Upgrade theoretisch und experimentell untersucht. Die magnetische Struktur wird mit realistischen helikalen magnetischen Flüssen modelliert. Als Störfluß dient eine analytische Anpassungsfunktion an Lösungen der Tearing-Mode-Gleichung. Das resultierende Temperaturprofil kann mit der Wärmele...

  15. Analysis of the ion energy transport in ohmic discharges in the ASDEX tokamak

    International Nuclear Information System (INIS)

    An analysis of the local ion energy transport is performed for more than one hundred well documented ohmic ASDEX discharges. These are characterized by three different confinement regimes: the linear ohmic confinement (LOC), the saturated ohmic confinement (SOC) and the improved ohmic confinement (IOC). All three are covered by this study. To identify the most important local transport mechanism of the ion heat, the ion power balance equation is analyzed. Two methods are used: straightforward calculation with experimental data only, and a comparison of measured and calculated profiles of the ion temperature and the ion heat conductivity, respectively. A discussion of the power balance shows that conductive losses dominate the ion energy transport in all ohmic discharges of ASDEX. Only inside the q=1-surface losses due to sawtooth activity play a role, while at the edge convective fluxes and CX-losses influence the ion energy transport. Both methods lead to the result that both the ion temperature and the ion heat conductivity are consistent with predictions of the neoclassical theory. Enhanced heat losses as suggested by theories eg. on the basis of ηi modes can be excluded. (orig.)

  16. Scintillator based detector for fast-ion losses induced by magnetohydrodynamic instabilities in the ASDEX upgrade tokamak

    International Nuclear Information System (INIS)

    A scintillator based detector for fast-ion losses has been designed and installed on the ASDEX upgrade (AUG) tokamak [A. Herrmann and O. Gruber, Fusion Sci. Technol. 44, 569 (2003)]. The detector resolves in time the energy and pitch angle of fast-ion losses induced by magnetohydrodynamics (MHD) fluctuations. The use of a novel scintillator material with a very short decay time and high quantum efficiency allows to identify the MHD fluctuations responsible for the ion losses through Fourier analysis. A Faraday cup (secondary scintillator plate) has been embedded behind the scintillator plate for an absolute calibration of the detector. The detector is mounted on a manipulator to vary its radial position with respect to the plasma. A thermocouple on the inner side of the graphite protection enables the safety search for the most adequate radial position. To align the scintillator light pattern with the light detectors a system composed by a lens and a vacuum-compatible halogen lamp has been allocated within the detector head. In this paper, the design of the scintillator probe, as well as the new technique used to analyze the data through spectrograms will be described. A last section is devoted to discuss the diagnosis prospects of this method for ITER [M. Shimada et al., Nucl. Fusion 47, S1 (2007)].

  17. Fast-ion redistribution and loss due to edge perturbations in the ASDEX Upgrade, DIII-D and KSTAR tokamaks

    International Nuclear Information System (INIS)

    The impact of edge localized modes (ELMs) and externally applied resonant and non-resonant magnetic perturbations (MPs) on fast-ion confinement/transport have been investigated in the ASDEX Upgrade (AUG), DIII-D and KSTAR tokamaks. Two phases with respect to the ELM cycle can be clearly distinguished in ELM-induced fast-ion losses. Inter-ELM losses are characterized by a coherent modulation of the plasma density around the separatrix while intra-ELM losses appear as well-defined bursts. In high collisionality plasmas with mitigated ELMs, externally applied MPs have little effect on kinetic profiles, including fast-ions, while a strong impact on kinetic profiles is observed in low-collisionality, low q95 plasmas with resonant and non-resonant MPs. In low-collisionality H-mode plasmas, the large fast-ion filaments observed during ELMs are replaced by a loss of fast-ions with a broad-band frequency and an amplitude of up to an order of magnitude higher than the neutral beam injection prompt loss signal without MPs. A clear synergy in the overall fast-ion transport is observed between MPs and neoclassical tearing modes. Measured fast-ion losses are typically on banana orbits that explore the entire pedestal/scrape-off layer. The fast-ion response to externally applied MPs presented here may be of general interest for the community to better understand the MP field penetration and overall plasma response. (paper)

  18. ASDEX-UG. ASDEX upgrade project proposal. Phase 2

    International Nuclear Information System (INIS)

    The objective of ASDEX UG is to investigate the problems relating to tokamak divertor physics and the boundary layer of hot plasmas which cannot be covered otherwise by either ASDEX or other EUROPEAN tokamaks, including JET, but whose investigation is indispensable for NET and INTOR. The configuration of ASDEX UG is changed as compared with ASDEX due to the requirement that all poloidal field coils are located outside the toroidal field magnet. This leads to a highly elongated D-shaped plasma with an ''open'' divertor, which does not allow to close the divertor chamber by such simple means as in ASDEX. In section 2, the aims of ASDEX UG are repeated briefly and the essential features and parameters of the tokamak system are summarized. The summary includes an overview of the tokamak design, the time schedule of design and construction concluding with the estimated investment cost and manpower required. In section 3 the tokamak system components are treated. The circuits and energy supply for the different electrical components are described in section 4. Auxiliary heating requirements and methods are discussed in section 5. Section 6 presents a survey over the periphery of the tokamak system including preparation of the building and radiation shielding. Section 7 outlines the physical programme. Section 8 is devoted to diagnostics. Finally, the principal concepts for control, data acquisition and handling are outlined in section 9. (orig./AH)

  19. Profile measurements of the electron temperature on the ASDEX Upgrade, COMPASS, and ISTTOK tokamak using Thomson scattering, triple, and ball-pen probes

    Science.gov (United States)

    Adamek, J.; Müller, H. W.; Silva, C.; Schrittwieser, R.; Ionita, C.; Mehlmann, F.; Costea, S.; Horacek, J.; Kurzan, B.; Bilkova, P.; Böhm, P.; Aftanas, M.; Vondracek, P.; Stöckel, J.; Panek, R.; Fernandes, H.; Figueiredo, H.

    2016-04-01

    The ball-pen probe (BPP) technique is used successfully to make profile measurements of the electron temperature on the ASDEX Upgrade (Axially Symmetric Divertor Experiment), COMPASS (COMPact ASSembly), and ISTTOK (Instituto Superior Tecnico TOKamak) tokamak. The electron temperature is provided by a combination of the BPP potential (ΦBPP) and the floating potential (Vfl) of the Langmuir probe (LP), which is compared with the Thomson scattering diagnostic on ASDEX Upgrade and COMPASS. Excellent agreement between the two diagnostics is obtained for circular and diverted plasmas and different heating mechanisms (Ohmic, NBI, ECRH) in deuterium discharges with the same formula Te = (ΦBPP - Vfl)/2.2. The comparative measurements of the electron temperature using BPP/LP and triple probe (TP) techniques on the ISTTOK tokamak show good agreement of averaged values only inside the separatrix. It was also found that the TP provides the electron temperature with significantly higher standard deviation than BPP/LP. However, the resulting values of both techniques are well in the phase with the maximum of cross-correlation function being 0.8.

  20. Profile measurements of the electron temperature on the ASDEX Upgrade, COMPASS, and ISTTOK tokamak using Thomson scattering, triple, and ball-pen probes

    Czech Academy of Sciences Publication Activity Database

    Adámek, Jiří; Müller, H.W.; Silva, C.; Schrittwieser, R.; Ionita, C.; Mehlmann, F.; Costea, S.; Horáček, Jan; Kurzan, B.; Bílková, Petra; Böhm, Petr; Aftanas, Milan; Vondráček, Petr; Stöckel, Jan; Pánek, Radomír; Fernandes, H.; Figueiredo, H.

    2016-01-01

    Roč. 87, č. 4 (2016), s. 043510. ISSN 0034-6748 R&D Projects: GA ČR(CZ) GA15-10723S; GA ČR(CZ) GAP205/12/2327; GA ČR(CZ) GA14-35260S; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : ball-pen probe (BPP) * ASDEX Upgrade * Langmuir probe (LP) * ISTTOK (Instituto Superior Tecnico TOKamak) * COMPASS (COMPact ASSembly), Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.614, year: 2014 http://scitation.aip.org/content/aip/journal/rsi/87/4/10.1063/1.4945797

  1. Migration and deposition of 13C in the full-tungsten ASDEX Upgrade tokamak

    Science.gov (United States)

    Hakola, A.; Likonen, J.; Aho-Mantila, L.; Groth, M.; Koivuranta, S.; Krieger, K.; Kurki-Suonio, T.; Makkonen, T.; Mayer, M.; Müller, H. W.; Neu, R.; Rohde, V.; ASDEX Upgrade Team

    2010-06-01

    The migration of carbon in low-density, low-confinement plasmas of ASDEX Upgrade was studied by injecting 13C into the main chamber of the torus at the end of the 2007 experimental campaign. A selection of standard tungsten-coated lower-divertor and main-chamber tiles as well as a complete set of lower-divertor tiles with an uncoated poloidal marker stripe were removed from one poloidal cross section and analysed using secondary ion mass spectrometry. The poloidal deposition profiles of 13C on both the tungsten-coated tiles and on the uncoated graphite areas of the marker tiles were measured and compared. For the W-coated lower-divertor tiles, 13C was deposited mainly on the high-field side tiles, while barely detectable amounts of 13C were observed on low-field side samples. In contrast, on the uncoated marker stripes the deposition was equally pronounced in the high-field and low-field side divertor. The marker-tile results are in agreement with those obtained from graphite tiles after the 2003 and 2005 13C experiments in ASDEX Upgrade. In the case of W-coated tiles, the 13C measurements were complemented by determining the total amount of deposited carbon (12C) on the tiles, which also shows strong deposition at the inner parts of the lower divertor. The estimated deposition of 13C on W at the divertor areas was less than 1.5% of the injected amount of 13C atoms. The 13C analyses of the main-chamber tiles and small silicon samples mounted in remote areas revealed significant deposition in the upper divertor, in upper parts of the heat shield, in the limiter region close to the injection valve, and below the roof baffle. Approximately 8% of the injected 13C is estimated to have accumulated in these regions. Possible reasons for the different deposition patterns on W and on graphite in different regions of the torus are discussed.

  2. Migration and deposition of 13C in the full-tungsten ASDEX Upgrade tokamak

    International Nuclear Information System (INIS)

    The migration of carbon in low-density, low-confinement plasmas of ASDEX Upgrade was studied by injecting 13C into the main chamber of the torus at the end of the 2007 experimental campaign. A selection of standard tungsten-coated lower-divertor and main-chamber tiles as well as a complete set of lower-divertor tiles with an uncoated poloidal marker stripe were removed from one poloidal cross section and analysed using secondary ion mass spectrometry. The poloidal deposition profiles of 13C on both the tungsten-coated tiles and on the uncoated graphite areas of the marker tiles were measured and compared. For the W-coated lower-divertor tiles, 13C was deposited mainly on the high-field side tiles, while barely detectable amounts of 13C were observed on low-field side samples. In contrast, on the uncoated marker stripes the deposition was equally pronounced in the high-field and low-field side divertor. The marker-tile results are in agreement with those obtained from graphite tiles after the 2003 and 2005 13C experiments in ASDEX Upgrade. In the case of W-coated tiles, the 13C measurements were complemented by determining the total amount of deposited carbon (12C) on the tiles, which also shows strong deposition at the inner parts of the lower divertor. The estimated deposition of 13C on W at the divertor areas was less than 1.5% of the injected amount of 13C atoms. The 13C analyses of the main-chamber tiles and small silicon samples mounted in remote areas revealed significant deposition in the upper divertor, in upper parts of the heat shield, in the limiter region close to the injection valve, and below the roof baffle. Approximately 8% of the injected 13C is estimated to have accumulated in these regions. Possible reasons for the different deposition patterns on W and on graphite in different regions of the torus are discussed.

  3. Improvement of the divertor bolometer diagnostic in the ASDEX Upgrade tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Sehmer, Till; Meister, Hans; Bernert, Matthias; Koll, Juergen; Reimold, Felix; Wischmeier, Marco; Fantz, Ursel [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Collaboration: ASDEX Upgrade Team

    2015-05-01

    For future fusion devices such as ITER, the radiation balance in the divertor region will have a significant impact on the power exhaust balance. Therefore, scenarios with strongly localized radiation, like radiation in the high field side high density (HFSHD) region, X-Point radiation or radiation in the divertor legs during detachment, will be investigated in the next ASDEX Upgrade (AUG) operation campaign 2015. To obtain accurately the absolute divertor radiation out of these measurements, the AUG foil bolometer diagnostic system in the divertor region has been enhanced; two new cameras have been designed and manufactured. One will be mounted below the roof baffle and contains 28 lines of sight (LOS), which will observe the mentioned regions of particular physical interest. The second camera consists of 4 LOS and will be mounted at the high field side above the inner divertor nose. It will observe radiation arising from the X-Point region and from the outer divertor. The data will be analysed with a tomographic reconstruction algorithm to localize and quantify the divertor radiation.

  4. Interpretation of D_alpha Imaging Diagnostics Data on the ASDEX Upgrade Tokamak

    OpenAIRE

    Harhausen, Jens

    2009-01-01

    The Tokamak configuration is a promising concept for magnetic confinement fusion. Cross-field transport in the plasma core leads to a plasma flux across the separatrix into the scrape-off layer, where it is guided along field lines towards the divertor targets. A return flux of neutral particles after plasma-wall interaction is directed towards the plasma chamber. Each discharge scenario is accompanied by a characteristic recycling pattern. The dominant mechanisms of neutralplasma...

  5. Investigation of the impurity transport in the ASDEX tokamak by spectroscopical methods

    International Nuclear Information System (INIS)

    Plasma impurities: a central problem of controlled thermonuclear fusion; magnetic plasma confinement in a Tokamak; methods to the determination of plasma impurity transport coefficients - by temporally modulated gas admission; the transport equation for impurities; neoclassical and anomalous transport; harmonic analysis of time-dependent signals; solutions of the transport equation; experimental equipment and measurements; measuring results - consistency of simple transport models with radial phase measurements; linearity of the transport processes; plasma disturbance by impurity injection; determination of the diffusion coefficient by simplified transport models; comparison of transport models for impurities and background plasma; measurements of the impurity transport at the plasma edge by high modulation frequencies. (AH)

  6. Suppression of sawtooth oscillations by lower-hybrid current drive in the ASDEX tokamak

    Science.gov (United States)

    Söldner, F. X.; McCormick, K.; Eckhartt, D.; Kornherr, M.; Leuterer, F.; Bartiromo, R.; Becker, G.; Bosch, H. S.; Brocken, H.; Derfler, H.; Eberhagen, A.; Fussmann, G.; Gehre, O.; Gernhardt, J.; Gierke, G. V.; Giuliana, A.; Glock, E.; Gruber, O.; Haas, G.; Hesse, M.; Hofmann, J.; Izvozchikov, A.; Janeschitz, G.; Karger, F.; Keilhacker, M.; Klüber, O.; Lackner, K.; Lenoci, M.; Lisitano, G.; Mast, F.; Mayer, H. M.; Meisel, D.; Mertens, V.; Müller, E. R.; Münich, M.; Murmann, H.; Niedermeyer, H.; Pietrzyk, A.; Poschenrieder, W.; Rapp, H.; Riedler, H.; Röhr, H.; Ryter, F.; Schmitter, K. H.; Schneider, F.; Setzensack, C.; Siller, G.; Smeulders, P.; Speth, E.; Steuer, K.-H.; Vien, T.; Vollmer, O.; Wagner, F.; Woyna, F. V.; Zasche, D.

    1986-09-01

    The sawtooth oscillations in tokamak discharges with Ohmic and neutral-beam heating could be suppressed when a large part of the plasma current was driven by lower-hybrid waves (IHF/Ip~=0.5). The stabilization is due to a flattening of the current profile j(r) and an increase of q(0) above 1. Higher central electron temperatures are obtained with neutral-beam heating if the sawteeth are stabilized. The increase in total energy content in this case was 30% higher than in the presence of sawteeth.

  7. Experimental studies and modelling of high radiation and high density plasmas in the ASDEX upgrade tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Casali, Livia

    2015-11-24

    Fusion plasmas contain impurities, either intrinsic originating from the wall, or injected willfully with the aim of reducing power loads on machine components by converting heat flux into radiation. The understanding and the prediction of the effects of these impurities and their radiation on plasma performances is crucial in order to retain good confinement. In addition, it is important to understand the impact of pellet injection on plasma performance since this technique allows higher core densities which are required to maximise the fusion power. This thesis contributes to these efforts through both experimental investigations and modelling. Experiments were conducted at ASDEX Upgrade which has a full-W wall. Impurity seeding was applied to H-modes by injecting nitrogen and also medium-Z impurities such as Kr and Ar to assess the impact of both edge and central radiation on confinement. A database of about 25 discharges has been collected and analysed. A wide range of plasma parameters was achieved up to ITER relevant values such as high Greenwald and high radiation fractions. Transport analyses taking into account the radiation distribution reveal that edge localised radiation losses do not significantly impact confinement as long as the H-mode pedestal is sustained. N seeding induces higher pedestal pressure which is propagated to the core via profile stiffness. Central radiation must be limited and controlled to avoid confinement degradation. This requires reliable control of the impurity concentration but also possibilities to act on the ELM frequency which must be kept high enough to avoid an irreversible impurity accumulation in the centre and the consequent radiation collapse. The key role of the f{sub ELM} is confirmed also by the analysis of N+He discharges. Non-coronal effects affect the radiation of low-Z impurities at the plasma edge. Due to the radial transport, the steep temperature gradients and the ELM flush out, a local equilibrium cannot be

  8. Experimental studies and modelling of high radiation and high density plasmas in the ASDEX upgrade tokamak

    International Nuclear Information System (INIS)

    Fusion plasmas contain impurities, either intrinsic originating from the wall, or injected willfully with the aim of reducing power loads on machine components by converting heat flux into radiation. The understanding and the prediction of the effects of these impurities and their radiation on plasma performances is crucial in order to retain good confinement. In addition, it is important to understand the impact of pellet injection on plasma performance since this technique allows higher core densities which are required to maximise the fusion power. This thesis contributes to these efforts through both experimental investigations and modelling. Experiments were conducted at ASDEX Upgrade which has a full-W wall. Impurity seeding was applied to H-modes by injecting nitrogen and also medium-Z impurities such as Kr and Ar to assess the impact of both edge and central radiation on confinement. A database of about 25 discharges has been collected and analysed. A wide range of plasma parameters was achieved up to ITER relevant values such as high Greenwald and high radiation fractions. Transport analyses taking into account the radiation distribution reveal that edge localised radiation losses do not significantly impact confinement as long as the H-mode pedestal is sustained. N seeding induces higher pedestal pressure which is propagated to the core via profile stiffness. Central radiation must be limited and controlled to avoid confinement degradation. This requires reliable control of the impurity concentration but also possibilities to act on the ELM frequency which must be kept high enough to avoid an irreversible impurity accumulation in the centre and the consequent radiation collapse. The key role of the fELM is confirmed also by the analysis of N+He discharges. Non-coronal effects affect the radiation of low-Z impurities at the plasma edge. Due to the radial transport, the steep temperature gradients and the ELM flush out, a local equilibrium cannot be

  9. Confinement studies on ASDEX

    International Nuclear Information System (INIS)

    This report deals with investigations of plasma confinement carried out on the ASDEX divertor tokamak. It is comprised of two sections: The first section deals with the study of the condinement degradation in auxiliary heated sicharges; in the second section a regime will be described where the severe deterioration of the confinement quality at high heating power is avoided. This regime is called the H-mode because of the high confinement characteristics in contrast to the low confinement L-mode. There is evidence that the L-mode characteristics will not lead to sufficient confinement quality for successful plasma burning. The H-mode was observed for the first time on ASDEX. At present it is considered as the confinement regime with the best prospects for future tokamak operation. But both the study of L- and H-mode confinement physics have increased our understanding on energy transport in tokamaks. (orig./GG)

  10. First Studies of ITER Diagnostic Mirrors in a Tokamak with All-metal Interior: Results of First Mirror Test in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Full text: In ITER, mirrors will be used as plasma-viewing elements in all optical and laser diagnostics. In the harsh environment mirror performance will degrade hampering the operation of respective diagnostics. The most adverse effect on mirrors is caused by the deposition of impurities and it is expected that the most challenging situation will occur in the divertor. With envisaged changes to all-metal plasma-facing components (PFCs) in ITER, an assessment of mirror performance in the existing divertor tokamak with all-metal PFCs is urgently needed. Such an experiment was made in the ASDEX Upgrade with all-tungsten PFCs as proposed by the International Tokamak Physics Activity (ITPA) Topical Group on Diagnostics, supported by the Specialists Working Group on First Mirrors and carried out in the frame of collaboration between Forschungszentrum Juelich and IPP Garching. Four molybdenum and four copper mirrors were mounted at the inner wall, in the dome facing the inner and outer divertor targets and in the pump-duct and exposed for seven months in ASDEX Upgrade. After exposure, degradation of the reflectivity was detected on all mirrors. The mirrors in the pump duct almost preserved their reflectivity unlike the mirrors in the dome facing the outer divertor which suffered from highest deposition and the strongest reflectivity degradation. Remarkably, only on the mirror facing the inner divertor and having very thin deposition layer of 15 nm, the carbon fraction was about 50 at.%. On all other mirrors this fraction did not exceed 20 at.%. The exposure of diagnostic mirrors in the tokamak with all-metal PFCs demonstrated a positive trend to a reduction of net deposition and minor changes in the reflectivity of mirrors located in the pump-duct far away from divertor plasmas. However, the degradation of all exposed mirrors underlines the necessity of an active mirror recovery. Urgent R&D is needed to address the lifetime issues of mirrors in ITER divertor. (author)

  11. Fast-ion losses induced by ELMs and externally applied magnetic perturbations in the ASDEX Upgrade tokamak

    International Nuclear Information System (INIS)

    Phase-space time-resolved measurements of fast-ion losses induced by edge localized modes (ELMs) and ELM mitigation coils have been obtained in the ASDEX Upgrade tokamak by means of multiple fast-ion loss detectors (FILDs). Filament-like bursts of fast-ion losses are measured during ELMs by several FILDs at different toroidal and poloidal positions. Externally applied magnetic perturbations (MPs) have little effect on plasma profiles, including fast-ions, in high collisionality plasmas with mitigated ELMs. A strong impact on plasma density, rotation and fast-ions is observed, however, in low density/collisionality and q95 plasmas with externally applied MPs. During the mitigation/suppression of type-I ELMs by externally applied MPs, the large fast-ion bursts observed during ELMs are replaced by a steady loss of fast-ions with a broad-band frequency and an amplitude of up to an order of magnitude higher than the neutral beam injection (NBI) prompt loss signal without MPs. Multiple FILD measurements at different positions, indicate that the fast-ion losses due to static 3D fields are localized on certain parts of the first wall rather than being toroidally/poloidally homogeneously distributed. Measured fast-ion losses show a broad energy and pitch-angle range and are typically on banana orbits that explore the entire pedestal/scrape-off-layer (SOL). Infra-red measurements are used to estimate the heat load associated with the MP-induced fast-ion losses. The heat load on the FILD detector head and surrounding wall can be up to six times higher with MPs than without 3D fields. When 3D fields are applied and density pump-out is observed, an enhancement of the fast-ion content in the plasma is typically measured by fast-ion D-alpha (FIDA) spectroscopy. The lower density during the MP phase also leads to a deeper beam deposition with an inward radial displacement of ≈2 cm in the maximum of the beam emission. Orbit simulations are used to test different models for 3D

  12. Development of a flexible Doppler reflectometry system and its application to turbulence characterization in the ASDEX Upgrade tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Troester, Carolin Helma

    2008-04-15

    An essential challenge in present fusion plasma research is the study of plasma turbulence. The turbulence behavior is investigated experimentally on the ASDEX Upgrade tokamak using Doppler reflectometry, a diagnostic technique sensitive to density fluctuations at a specific wavenumber k {sub perpendicular} {sub to}. This microwave radar diagnostic utilizes localized Bragg backscattering of the launched beam (k{sub 0}) by the density fluctuations at the plasma cutoff layer. The incident angle {theta} selects the probed k {sub perpendicular} {sub to} via the Bragg condition k {sub perpendicular} {sub to} {approx} 2k{sub 0}sin{theta}. The measured Doppler shifted frequency spectrum allows the determination of the perpendicular plasma rotation velocity, u {sub perpendicular} {sub to} =v{sub E} {sub x} {sub B}+v{sub turb}, directly from the Doppler frequency shift(f{sub D} = u {sub perpendicular} {sub to} k {sub perpendicular} {sub to} /2{pi}), and the turbulence amplitude from the backscattered power level. This thesis work presents a survey of u {sub perpendicular} {sub to} radial profiles and k {sub perpendicular} {sub to} spectrum measurements for a variety of plasma conditions obtained by scanning the antenna tilt angle. This was achieved by extending the existing V-band Doppler reflectometry system (50 - 75 GHz) with a new W-band system (75 - 110 GHz), which was especially designed for measuring the k {sub perpendicular} {sub to} spectrum and additionally expands the radial coverage into the plasma core region. It consists of a remote steerable antenna with an adjustable line of sight allowing for dynamic wavenumber selection up to 25 cm {sup -1} and a reflectometer with a 'phase locked loop' stabilized transmitter allowing for the precise determination of the instrument response function. The proper system functionality was demonstrated by laboratory testing and benckmarking against the V-band system. The new profile measurements obtained show a

  13. High-speed lithium pellet injector commissioning in ASDEX Upgrade to investigate impact of Li in an all-metal wall tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Arredondo Parra, Rodrigo; Lang, Peter Thomas; Ploeckl, Bernhard [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Cardella, Antonino [Technische Universitaet Muenchen, Garching (Germany); Fusion for Energy, Garching (Germany); Macian Juan, Rafael [Technische Universitaet Muenchen, Garching (Germany); Neu, Rudolf [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Technische Universitaet Muenchen, Garching (Germany)

    2015-05-01

    Encouraging results with respect to plasma performance have been observed in several tokamak devices (TFTR, NSTX, etc) when injecting Lithium. Recently, a pedestal broadening resulting in an enhanced energy content during transient ELM-free H-mode phases was achieved in DIII-D. Experiments are planned at ASDEX Upgrade, aiming to investigate the impact of Li in an all-metal wall tokamak and to enhance the pedestal operational space. For this purpose, a Lithium pellet injector has been developed, capable of injecting pellets with a particle content up to 1.64 . 10{sup 20} atoms (1.89 mg) at a foreseen maximum repetition rate of 3 Hz. Free flight launch from the torus outboard side without a guiding tube is envisaged. A transfer efficiency exceeding 90 % was achieved in the test bed. Pellets will be accelerated in a gas gun; hence special care must be taken to avoid deleterious effects by the propellant gas pulse, this being the main plasma gas, leading to speeds ranging from 500 (m)/(s) to 800 (m)/(s). Additionally, a large expansion volume equipped with a cryopump is added in to the flight path. The injector is expected to commence operation by May 2015.

  14. Overview of ASDEX Upgrade results

    DEFF Research Database (Denmark)

    Stroth, U.; Adamek, J.; Aho-Mantila, L.;

    2013-01-01

    The medium size divertor tokamak ASDEX Upgrade (major and minor radii 1.65 m and 0.5 m, respectively, magnetic-field strength 2.5 T) possesses flexible shaping and versatile heating and current drive systems. Recently the technical capabilities were extended by increasing the electron cyclotron r...

  15. Regime of Improved Confinement and High Beta in Neutral-Beam-Heated Divertor Discharges of the ASDEX Tokamak

    Science.gov (United States)

    Wagner, F.; Becker, G.; Behringer, K.; Campbell, D.; Eberhagen, A.; Engelhardt, W.; Fussmann, G.; Gehre, O.; Gernhardt, J.; Gierke, G. V.; Haas, G.; Huang, M.; Karger, F.; Keilhacker, M.; Klüber, O.; Kornherr, M.; Lackner, K.; Lisitano, G.; Lister, G. G.; Mayer, H. M.; Meisel, D.; Müller, E. R.; Murmann, H.; Niedermeyer, H.; Poschenrieder, W.; Rapp, H.; Röhr, H.; Schneider, F.; Siller, G.; Speth, E.; Stäbler, A.; Steuer, K. H.; Venus, G.; Vollmer, O.; Yü, Z.

    1982-11-01

    A new operational regime has been observed in neutral-injection-heated ASDEX divertor discharges. This regime is characterized by high βp values comparable to the aspect ratio A (βp=1.9 MW, a mean density n¯e>=3×1013 cm-3, and a q(a) value >=2.6. Beyond these limits or in discharges with material limiter, low βp values and reduced particle and energy confinement times are obtained compared to the Ohmic heating phase.

  16. Supplement to 'ASDEX Upgrade, definition of a tokamak experiment with a reactor-compatible poloidal divertor' (IPP-report 1/197, March 1982)

    International Nuclear Information System (INIS)

    Since March 1982 the better understanding of the divertor physics, both by theory and experiments, and the development of the ASDEX Upgrade concept have considerably improved and simplified the ASDEX Upgrade design. Single null poloidal divertor configurations were calculated, which can well compete with elongated limiter configurations in reduced poloidal field effort. The role of recycling and its limitation set by the available energy flux, observed experimentally and explained by a plasma boundary flow model, led to a refined formulation of the line density requirements. Finally, a discussion of the attainable temperature and densities allowed clearly to distinguish between ASDEX and ASDEX Upgrade and pointed out the dominant role of the plasma current. The ASDEX Upgrade basic data are summarized as presented to the EURATOM advisory board. (orig.)

  17. Time and space-resolved energy flux measurements in the divertor of the ASDEX tokamak by computerized infrared thermography

    International Nuclear Information System (INIS)

    A new, fully computerized and automatic thermographic system has been developed. Its two central components are an AGA THV 780 infrared camera and a PDP-11/34 computer. A combined analytical-numerical method of solving the 1-dimensional heat diffusion equation for a solid of finite thickness bounded by two parallel planes was developed. In high-density (anti nsub(e) = 8 x 1013 cm-3) neutral-beam-heated (L-mode) divertor discharges in ASDEX, the power deposition on the neutralizer plates is reduced to about 10-15% of the total heating power, owing to the inelastic scattering of the divertor plasma from a neutral gas target. Between 30% and 40% of the power is missing in the global balance. The power flow inside the divertor chambers is restricted to an approximately 1-cm-thick plasma scrape-off layer. This width depends only weakly on the density and heating power. During H-phases free of Edge Localized Mode (ELM) activity the energy flow into the divertor is blocked. During H-phases with ELM activity the energy is expelled into the divertor in very short intense pulses (several MW for about one hundred μs). Sawtooth events are able to transport significant amounts of energy from the plasma core to the peripheral zones and the scrape-off layer, and they are frequently correlated with transitions from the L to the H mode. (orig./AH)

  18. ASDEX contributions to the 19th European conference on controlled fusion and plasma heating (Innsbruck, June 29 to July 3, 1992). - ASDEX contributions to the 10th PSI conference (Monterey, USA, March 30 to April 3, 1992)

    International Nuclear Information System (INIS)

    This paper contains 10 contributions to the following topics: Characteristic features of density fluctuations associated with the L-H-transition in the ASDEX tokamak; change of internal inductance and anisotropy during lower hybrid current drive in ASDEX; a study of the SOL density profile behavior in ASDEX; attempt to model the edge turbulence of a tokamak as a random superposition of eddies; H-mode power threshold in ASDEX; influence of divertor geometry and boronization on elm-free H-mode confinement in ASDEX; ICRF power limitation relation to density limit in ASDEX; reflectometry measurements of the m=1 satellite mode in L- and H-mode plasmas in ASDEX; confiment scaling for the ASDEX L-mode in different divertor configurations; particle and energy transport scalings in the ASDEX scrape-off layer. (orig./MM)

  19. Overview of ASDEX Upgrade results

    Czech Academy of Sciences Publication Activity Database

    Stroth, U.; Adámek, Jiří; Aho-Mantila, L.; Äkäslompolo, S.; Amdor, C.; Angioni, C.; Balden, M.; Bardin, S.; Barrera Orte, L.; Behler, K.; Belonohy, E.; Bergmann, A.; Bernert, M.; Bilato, R.; Birkenmeier, G.; Bobkov, V.; Boom, J.; Bottereau, C.; Bottino, A.; Braun, F.; Brezinsek, S.; Brochard, T.; Brüdgam, M.; Buhler, A.; Burckhart, A.; Casson, F.J.; Chankin, A.; Chapman, I.; Clairet, F.; Classen, I.G.J.; Coenen, J.W.; Conway, G.D.; Coster, D.P.; Curran, D.; da Silva, F.; de Marné, P.; D’Inca, R.; Douai, D.; Drube, R.; Dunne, M.; Dux, R.; Eich, T.; Eixenberger, H.; Endstrasser, N.; Engelhardt, K.; Esposito, B.; Fable, E.; Fischer, R.; Fünfgelder, H.; Fuchs, J.C.; Gál, K.; García Munoz, M.; Geiger, B.; Giannone, L.; Görler, T.; da Graca, S.; Greuner, H.; Gruber, O.; Gude, A.; Guimarais, L.; Günter, S.; Haas, G.; Hakola, A.H.; Hangan, D.; Happel, T.; Härtl, T.; Hauff, T.; Heinemann, B.; Herrmann, A.; Hobirk, J.; Höhnle, H.; Hölzl, M.; Hopf, C.; Igochine, V.; Ionita, C.; Janzer, A.; Jenko, F.; Käsemann, C.-P.; Kallenbach, A.; Kálvin, S.; Kantor, M.; Kappatou, A.; Kardaun, O.; Kasparek, W.; Kaufmann, M.; Kirk, A.; Klingshirn, H.-J.; Kocan, M.; Kocsis, G.; Konz, C.; Koslowski, R.; Krieger, K.; Kubic, M.; Kurki-Suonio, T.; Kurzan, B.; Lackner, K.; Lang, P.T.; Lauber, P.; Laux, M.; Lazaros, A.; Leipold, F.; Leuterer, F.; Lindig, S.; Lisgo, S.; Lohs, A.; Lunt, T.; Maier, H.; Makkonen, T.; Mank, K.; Manso, M.-E.; Maraschek, M.; Mayer, M.; McCarthy, P.J.; McDermott, R.; Mehlmann, F.; Meister, H.; Menchero, L.; Meo, F.; Merkel, P.; Merkel, R.; Mertens, V.; Merz, F.; Mlynek, A.; Monaco, F.; Müller, S.; Müller, H.W.; Münich, M.; Neu, G.; Neu, R.; Neuwirth, D.; Nocente, M.; Nold, B.; Noterdaeme, J.-M.; Pautasso, G.; Pereverzev, G.; Plöckl, B.; Podoba, Y.; Pompon, F.; Poli, E.; Polozhiy, K.; Potzel, S.; Püschel, M.J.; Pütterich, T.; Rathgeber, S.K.; Raupp, G.; Reich, M.; Reimold, M.; Ribeiro, T.; Riedl, R.; Rohde, V.; v.Rooij, G.; Roth, J.; Rott, M.; Ryter, F.; Salewski, M.; Santos, J.; Sauter, P.; Scarabosio, A.; Schall, G.; Schmid, K.; Schneider, P.A.; Schneider, W.; Schrittwieser, R.; Schubert, M.; Schweinzer, J.; Scott, B.; Sempf, M.; Sertoli, M.; Siccinio, M.; Sieglin, B.; Sigalov, A.; Silva, A.; Sommer, F.; Stäbler, A.; Stober, J.; Streibl, B.; Strumberger, E.; Sugiyama, K.; Suttrop, W.; Tala, T.; Tardini, G.; Teschner, M.; Tichmann, C.; Told, D.; Treutterer, W.; Tsalas, M.; Van Zeeland, M.A.; Varela, P.; Véres, G.; Vicente, J.; Vianello, N.; Vierle, T.; Viezzer, E.; Viola, B.; Vorpahl, C.; Wachowski, M.; Wagner, D.; Wauters, T.; Weller, A.; Wenninger, R.; Wieland, B.; Willensdorfer, M.; Wischmeier, M.; Wolfrum, E.; Würsching, E.; Yu, Q.; Zammuto, I.; Zasche, D.; Zehetbauer, T.; Zhang, Y.; Zilker, M.; Zohm, H.

    2013-01-01

    Roč. 53, č. 10 (2013), s. 104003-104003. ISSN 0029-5515. [IAEA Fusion Energy Conference/24./. San Diego, 08.10.2012-13.10.2012] Institutional support: RVO:61389021 Keywords : tokamak * ASDEX * ITER * ICRH system Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.243, year: 2013 http://m.iopscience.iop.org/0029-5515/53/10/104003/pdf/0029-5515_53_10_104003.pdf

  20. Entropy relaxation of ASDEX plasmas

    International Nuclear Information System (INIS)

    In tokamak discharges with improved ohmic confinement (IOC) in ASDEX a transition is observed from flat density profiles towards more peaked ones, while the normalized temperature profile is preserved. For this behaviour of the radial profiles it is shown that the entropy of the plasma increases during the IOC phase. Hence IOC and entropy relaxation are closely related. If the IOC phase is long enough, one finds stationary plasma states, which are compared with the relaxed state described in theory. (orig.)

  1. Pellet imaging techniques on ASDEX

    International Nuclear Information System (INIS)

    As part of a USDOE/ASDEX collaboration, a detailed examination of pellet ablation in ASDEX with a variety of diagnostics has allowed a better understanding of a number of features of hydrogen ice pellet ablation in a plasma. In particular, fast gated photos with an intensified Xybion CCD video camera allow in-situ velocity measurements of the pellet as it penetrates the plasma. With time resolution of typically 100 nanoseconds and exposures every 50 microseconds, the evolution of each pellet in a multi-pellet ASDEX tokamak plasma discharge can be followed. When the pellet cloud track has striations, the light intensity profile through the cloud is hollow (dark near the pellet), whereas at the beginning or near the end of the pellet trajectory the track is typically smooth (without striations) and has a gaussian-peaked light emission profile. New, single pellet Stark broadened DαDβ, and Dγ spectra, obtained with a tangentially viewing scanning mirror/spectrometer with Reticon array readout, are consistent with cloud densities of 2 x 1017cm-3 or higher in the regions of strongest light emission. A spatially resolved array of Dα detectors shows that the light variations during the pellet ablation are not caused solely by a modulation of the incoming energy flux as the pellet crosses rational q-surfaces, but instead are a result of a dynamic, non-stationary, ablation process. 20 refs., 4 figs

  2. Direct measurements of the plasma potential in ELMy H-mode plasma with ball-pen probes on ASDEX Upgrade tokamak

    Czech Academy of Sciences Publication Activity Database

    Adámek, Jiří; Stöckel, Jan; Brotánková, Jana; Horáček, Jan; Rohde, V.; Müller, H. W.; Herrmann, A.; Schrittwieser, R.; Mehlmann, F.; Ionita, C.

    390-391, - (2009), s. 1114-1117. ISSN 0022-3115. [International Conference on Plasma-Surface Interactions in Controlled Fusion Device/18th./. Toledo, 26.05.2008-30.05.2008] R&D Projects: GA AV ČR KJB100430601 Institutional research plan: CEZ:AV0Z20430508 Keywords : Edge plasma * Electric field * ELMs * H-mode * ASDEX-Upgrade Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.933, year: 2009 http://dx.doi.org/10.1016/j.jnucmat.2009.01.286

  3. Fusion reaction product diagnostics in ASDEX

    International Nuclear Information System (INIS)

    A diagnostic method was developed to look for the charged fusion products from the D(D,p)T-reactions in the divertor tokamak ASDEX. With a semi-conductor detector it was possible to evaluate the ion temperature in thermal plasmas from the proton energy spectra as well as from the triton spectra. In lower-hybrid wave heated plasmas non-thermal (fast) ions were observed. These ions create fusion products with a characteristically different energy spectrum. (orig.)

  4. Disruption studies on ASDEX upgrade

    International Nuclear Information System (INIS)

    Disruptions generate large thermal and mechanical stresses on the tokamak components and are occasionally responsible for damages to the machine. For a future reactor disruptions have a significant impact on the design since all loading conditions must be analyzed in accordance with stricter design criteria (due to safety or difficult maintenance). Therefore the uncertainties affecting the predicted stresses must be reduced as much as possible with a more comprehensive set of measurements and analyses in this generation of experimental machines, and avoidance/predictive methods must be developed further. Disruption studies on ASDEX Upgrade are focused on these subjects, namely on: (1) understanding the physical mechanisms leading to this phenomenon in order to learn to avoid it or to predict its occurrence and to mitigate its effects; (2) analyzing the effects of disruptions on the machine to determine the functional dependence of the thermal and mechanical loads upon the discharge parameters. This allows, firstly, to dimension or reinforce the machine components to withstand these loads and, secondly, to extrapolate them to tokamaks still in the design phase; (3) learning to mitigate the consequence of disruptions, i.e. thermal loads, mechanical forces and runaways with injection of impurity pellets or gas. This paper is focused on most recent results concerning points, i.e. on the analysis of the degree of asymmetry of the forces and on the use of impurity puff for mitigation

  5. Disruption studies in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Disruption generate large thermal and mechanical stresses on the tokamak components. For a future reactor disruptions have a significant impact on the design since all loading conditions must be analyzed in accordance with stricter design criteria (due to safety or difficult maintenance). Therefore the uncertainties affecting the predicted stresses must be reduced as much as possible with a more comprehensive set of measurements and analyses in this generation of experimental machines, and avoidance/ predictive methods must be developed further. The study of disruptions on ASDEX Upgrade is focused on these subjects, namely on: (1) understanding the physical mechanisms leading to this phenomenon and learning to avoid it or to predict its occurrence (with neural networks, for example) and to mitigate its effects; (2) analyzing the effects of disruptions on the machine to determine the functional dependence of the thermal and mechanical loads upon the discharge parameters. This allows to dimension or reinforce the machine components to withstand these loads and to extrapolate them to tokamaks still in the design phase; (3) learning to mitigate the consequence of disruptions. (author)

  6. Laser-blow-off experiments on ASDEX

    International Nuclear Information System (INIS)

    In 1985 laser blow-off system was installed on the ASDEX tokamak and in the following period a number of experiments were performed. The idea of these investigations was to study the impurity behaviour under different plasma conditions. New and unexpected results were obtained: 1. Considerable improvement of the impurity confinement time in OH plasmas near to the density limit, 2. influence on ELMs by impurities, 3. strong dependence of the impurity confinement time on the gradient of the plasma current, and 4. different impurity confinement under various heating conditions. (orig.)

  7. Periodic multichannel Thomson scattering in ASDEX

    International Nuclear Information System (INIS)

    The optical and electronic design of the Thomson scattering experiment in the ASDEX-Tokamak is described. This Thomson scattering system is employed as a standard diagnostic for the evaluation of electron temperature and density simultaneously at 16 spatial points in ASDEX. The light source is a Nd-YAG laser emitting at 1.06 μm wavelength, which is capable of delivering 60 pulses per second for a period of about 7 sec. This period includes the whole ASDEX plasma discharge. The scattered light is detected by Si-avalanche diodes. Density calibration is carried out by rotational anti-Stokes Raman scattering from molecular hydrogen. The system is capable of measuring densities as low as 5x1012 cm-3 and electron temperatures in the range from 150 eV to 5 keV. The data-processing system and the calculations which lead to the final output of Te/Ne-profiles are discussed. Examples of profile measurements are given showing the possibilities of the system under various plasma conditions. Technical details of the system are described in tables listed in the appendix. (orig.)

  8. Visible spectroscopy on ASDEX

    International Nuclear Information System (INIS)

    In this report visible spectroscopy and impurity investigations on ASDEX are reviewed and several sets of visible spectra are presented. As a basis for identification of metallic impurity lines during plasma discharges spectra from a stainless steel - Cu arc have been recorded. In a next step a spectrum overview of ASDEX discharges is shown which reveals the dominating role of lines from light impurities like carbon and oxygen throughout the UV and visible range (2000 A ≤ λ ≤ 8000 A). Metallic impurity lines of neutrals or single ionized atoms are observed near localized surfaces. The dramatic effect of impurity reduction by boronization of the vessel walls is demonstrated in a few examples. In extension to some ivesti-gations already published, further diagnostic applications of visible spectroscopy are presented. Finally, the hardware and software system used on ASDEX are described in detail. (orig.)

  9. Spectroscopy as a major programme in ASDEX - a discussion study

    International Nuclear Information System (INIS)

    This report deals with the objectives and possibilities of a spectroscopy programme in ASDEX and provides some basic information on the relevant processes of atomic physics in tokamaks. The spectroscopic analogies found in observation of astrophysical objects are also briefly treated. In addition, the possibilities for conducting investigations in alternative high-Z ion sources are discussed. A first proposal for an appropriate programme is then formulated. (orig.)

  10. Data acquisition and real time signal processing of plasma diagnostics on ASDEX Upgrade using LabVIEW RT

    International Nuclear Information System (INIS)

    LabVIEW is a valuable tool for building multi-core multi-threaded application using commercial off the shelf components, FPGAs and third party hardware for real time diagnostics on ASDEX tokamak upgrade. 5 real time labVIEW RT diagnostics in operation with connection to UDP/XML framework of the ASDEX control system are presented in this poster: 1) real time flux surfaces, 2) vacuum field calculations, 3) halo currents, 4) bolometer and 5) manometer

  11. Stability investigations of the ASDEX feedback system with filters for reducing thyristor noise

    International Nuclear Information System (INIS)

    A computer program for analysing the absolute and relative stabilities of any complex system by the root-locus method was developed. It is used to reanalyse the present horizontal position feed-back control in the ASDEX tokamak and to select the optimum parameters for this system with RCL filters for reducing thyristor noise. (orig.)

  12. Fast-ion transport and neutral beam current drive in ASDEX upgrade

    DEFF Research Database (Denmark)

    Geiger, B.; Weiland, M.; Jacobsen, Asger Schou;

    2015-01-01

    The neutral beam current drive efficiency has been investigated in the ASDEX Upgrade tokamak by replacing on-axis neutral beams with tangential off-axis beams. A clear modification of the radial fast-ion profiles is observed with a fast-ion D-alpha diagnostic that measures centrally peaked profil...

  13. Interpretation of fast measurements of plasma potential, temperature and density in SOL of ASDEX Upgrade

    DEFF Research Database (Denmark)

    Horacek, J.; Adamek, J.; Müller, H.W.;

    2010-01-01

    This paper focuses on interpretation of fast (1 µs) and local (2–4 mm) measurements of plasma density, potential and electron temperature in the edge plasma of tokamak ASDEX Upgrade. Steady-state radial profiles demonstrate the credibility of the ball-pen probe. We demonstrate that floating poten...

  14. Divertor efficiency in ASDEX

    Science.gov (United States)

    Engelhardt, W.; Becker, G.; Behringer, K.; Campbell, D.; Eberhagen, A.; Fussmann, G.; Gehre, O.; Gierke, G. V.; Glock, E.; Haas, G.; Huang, M.; Karger, F.; Keilhacker, M.; KlÜber, O.; Kornherr, M.; Lisitano, G.; Mayer, H.-M.; Meisel, D.; Müller, E. R.; Murmann, H.; Niedermeyer, H.; Poschenrieder, W.; Rapp, H.; Schneider, F.; Siller, G.; Steuer, K.-H.; Venus, G.; Vernickel, H.; Wagner, F.

    1982-12-01

    The divertor efficiency in ASDEX is discussed for ohmically heated plasmas. The parameters of the boundary layer both in the torus midplane and the divertor chamber have been measured. The results are reasonably well understood in terms of parallel and perpendicular transport. A high pressure of neutral hydrogen builds up in the divertor chamber and Franck-Condon particles recycle back through the divertor throat. Due to dissociation processes the boundary plasma is effectively cooled before it reaches the neutralizer plates. The shielding property of the boundary layer against impurity influx is comparable to that of a limiter plasma. The transport of iron is numerically simulated for an iron influx produced by sputtering of charge exchange neutrals at the wall. The results are consistent with the measured iron concentration. First results from a comparison of the poloidal divertor with toroidally closed limiters (stainless steel, carbon) are given. Diverted discharges are considerably cleaner and easier to create.

  15. A data bank of disruptive discharges in ASDEX

    International Nuclear Information System (INIS)

    The compilation of data banks relating to plasma disruptions is important for the design of next-step devices and tokamak reactors, as a means of establishing safe operation regimes and assessing the residual risk from such events. ASDEX has an operational history of 33509 plasma shots covering an exceptionally wide range of machine conditions: Divertor/limiter configurations; Ohmic, NBI, ICRH and LH heating; carbonization, boronization wall-conditioning, gas-puff and pellet refuelling. We have compiled a data base of the Disruptive Operationl Regimes in ASDEX (DORA), which contains the relevant information for all ASDEX-discharges and is available on tape and readable by different data bank systems for further evaluation. We first describe the criteria applied to recognize and classify disruptions and the information about them stored in the file. In a second part we use the DORA file for some sample applications of physical or engineering interest. In an appendix we give the data and format information necessary to read the DORA file. (orig.)

  16. Fast-ion transport in the presence of magnetic reconnection induced by sawtooth oscillations in ASDEX Upgrade

    NARCIS (Netherlands)

    Geiger, B.; Garcia-Munoz, M.; Dux, R.; Ryter, F.; Tardini, G.; Orte, L. B.; Classen, I.G.J.; Fable, E.; Fischer, R.; Igochine, V.; McDermott, R. M.

    2014-01-01

    The transport of beam-generated fast ions has been investigated experimentally at the ASDEX Upgrade tokamak in the presence of sawtooth crashes. After sawtooth crashes, phase space resolved fast-ion D-alpha measurements show a significant reduction of the central fast-ion density-more than 50%-toget

  17. Destabilisation of TAE modes using ICRH in ASDEX Upgrade

    International Nuclear Information System (INIS)

    In ASDEX Upgrade, toroidicity induced Alfven eigenmodes (TAEs) are destabilised by ICRH in conventional and advanced scenarios, at low density. Most unstable TAEs have toroidal modes numbers (n=3,4,5,6) and experiments with reversed current and magnetic field showed that the TAE propagate in the current direction. On one hand, the analysis of the unstable TAE in ASDEX Upgrade shows that the data is consistent with the results from previous studies performed in other tokamaks. In particular, the measured TAE frequency, in the range (150-200kHz), is consistent with theoretical TAE frequency calculated for the parameters of the discharges performed in these experiments. On the other hand, some interesting new features have also been observed in the ASDEX Upgrade data. TAE (n=-1) was observed, which propagates in the opposite direction to the plasma current. It was also concluded that plasma rotation is insufficient to explain the experimentally observed frequency differences between two adjacent toroidal mode numbers. The measured relative fluctuation amplitude of the TAE eigenfunction in the soft X-rays channels increases towards the plasma edge. These results are consistent with the ideal MHD calculations and show that the TAE are of a global nature and not core localised TAE modes. These results are particularly important, because the radial extent of AE is a key factor in the redistribution of the energetic ions in the presence of unstable TAE. In advanced Tokamak scenarious, no evidence of Alfven Cascades was observed in experiments with ICRH minority heating in the early phase of the discharge. The effect of Electron Cyclotron Current Drive (ECCD) on the TAE amplitude was studied by applying different levels of ECCD in similar plasma configurations, with equivalent ICRH power. It was observed that ECCD has a slight destabilizing effect on the TAE. (author)

  18. Nitrogen implantation in tungsten and migration in the fusion experiment ASDEX upgrade

    International Nuclear Information System (INIS)

    The implantation of nitrogen ions into tungsten was studied in laboratory experiments to understand the interaction of nitrogen containing fusion plasmas with tungsten walls. The resulting model of W-N interaction was tested by experiments in the tokamak ASDEX Upgrade. Using the measurements from these experiments as boundary condition, nitrogen transport and re-distribution in the plasma were modeled by self-consistent WallDYN-DIVIMP simulations.

  19. Nitrogen implantation in tungsten and migration in the fusion experiment ASDEX upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Meisl, Gerd Korbinian

    2015-01-12

    The implantation of nitrogen ions into tungsten was studied in laboratory experiments to understand the interaction of nitrogen containing fusion plasmas with tungsten walls. The resulting model of W-N interaction was tested by experiments in the tokamak ASDEX Upgrade. Using the measurements from these experiments as boundary condition, nitrogen transport and re-distribution in the plasma were modeled by self-consistent WallDYN-DIVIMP simulations.

  20. Overview of ASDEX Upgrade results

    DEFF Research Database (Denmark)

    Kallenbach, A.; Adamek, J.; Aho-Mantila, L.;

    2011-01-01

    The ASDEX Upgrade programme is directed towards physics input to critical elements of the ITER design and the preparation of ITER operation, as well as addressing physics issues for a future DEMO design. After the finalization of the tungsten coating of the plasma facing components, the re-availa...

  1. ICRF heating analysis on ASDEX plasmas

    International Nuclear Information System (INIS)

    ICRF (ion cyclotron range of frequencies) waves heating in ASDEX tokamak is analysed. The excitation, propagation and absorption are studied by using a global wave code. This analysis is combined with a Fokker-Planck code, and the generation of fast ions and thermalization of the absorbed power are obtained theoretically. The wave form in the plasma, the loading resistance and reactance of the antenna are calculated for both the minority ion heating and the second harmonic resonance heating. Attention is given to the change of the antenna loading associated with the L/H transition. Optimum conditions for the loading are discussed. In the minority heating case, the tail generation and thermalization are analyzed. Spatial profiles of the tail-ion temperature and the power transferred to the bulk electrons and ions are obtained. Central as well as off-central heating cases are investigated. The ratio of the electron heating power is obtained. Finally, the effect of the reactive electric field is discussed in connection with rf losses and impurity production. (orig.)

  2. The tungsten divertor experiment at ASDEX Upgrade

    Science.gov (United States)

    Neu, R.; Asmussen, K.; Krieger, K.; Thoma, A.; Bosch, H.-S.; Deschka, S.; Dux, R.; Engelhardt, W.; García-Rosales, C.; Gruber, O.; Herrmann, A.; Kallenbach, A.; Kaufmann, M.; Mertens, V.; Ryter, F.; Rohde, V.; Roth, J.; Sokoll, M.; Stäbler, A.; Suttrop, W.; Weinlich, M.; Zohm, H.; Alexander, M.; Becker, G.; Behler, K.; Behringer, K.; Behrisch, R.; Bergmann, A.; Bessenrodt-Weberpals, M.; Brambilla, M.; Brinkschulte, H.; Büchl, K.; Carlson, A.; Chodura, R.; Coster, D.; Cupido, L.; de Blank, H. J.; de Peña Hempel, S.; Drube, R.; Fahrbach, H.-U.; Feist, J.-H.; Feneberg, W.; Fiedler, S.; Franzen, P.; Fuchs, J. C.; Fußmann, G.; Gafert, J.; Gehre, O.; Gernhardt, J.; Haas, G.; Herppich, G.; Herrmann, W.; Hirsch, S.; Hoek, M.; Hoenen, F.; Hofmeister, F.; Hohenöcker, H.; Jacobi, D.; Junker, W.; Kardaun, O.; Kass, T.; Kollotzek, H.; Köppendörfer, W.; Kurzan, B.; Lackner, K.; Lang, P. T.; Lang, R. S.; Laux, M.; Lengyel, L. L.; Leuterer, F.; Manso, M. E.; Maraschek, M.; Mast, K.-F.; McCarthy, P.; Meisel, D.; Merkel, R.; Müller, H. W.; Münich, M.; Murmann, H.; Napiontek, B.; Neu, G.; Neuhauser, J.; Niethammer, M.; Noterdaeme, J.-M.; Pasch, E.; Pautasso, G.; Peeters, A. G.; Pereverzev, G.; Pitcher, C. S.; Poschenrieder, W.; Raupp, G.; Reinmüller, K.; Riedl, R.; Röhr, H.; Salzmann, H.; Sandmann, W.; Schilling, H.-B.; Schlögl, D.; Schneider, H.; Schneider, R.; Schneider, W.; Schramm, G.; Schweinzer, J.; Scott, B. D.; Seidel, U.; Serra, F.; Speth, E.; Silva, A.; Steuer, K.-H.; Stober, J.; Streibl, B.; Treutterer, W.; Troppmann, M.; Tsois, N.; Ulrich, M.; Varela, P.; Verbeek, H.; Verplancke, Ph; Vollmer, O.; Wedler, H.; Wenzel, U.; Wesner, F.; Wolf, R.; Wunderlich, R.; Zasche, D.; Zehetbauer, T.; Zehrfeld, H.-P.

    1996-12-01

    Tungsten-coated tiles, manufactured by plasma spray on graphite, were mounted in the divertor of the ASDEX Upgrade tokamak and cover almost 90% of the surface facing the plasma in the strike zone. Over 600 plasma discharges have been performed to date, around 300 of which were auxiliary heated with heating powers up to 10 MW. The production of tungsten in the divertor was monitored by a W I line at 400.8 nm. In the plasma centre an array of spectral lines at 5 nm emitted by ionization states around W XXX was measured. From the intensity of these lines the W content was derived. Under normal discharge conditions W-concentrations around 0741-3335/38/12A/013/img12 or even lower were found. The influence on the main plasma parameters was found to be negligible. The maximum concentrations observed decrease with increasing heating power. In several low power discharges accumulation of tungsten occurred and the temperature profile was flattened. The concentrations of the intrinsic impurities carbon and oxygen were comparable to the discharges with the graphite divertor. Furthermore, the density and the 0741-3335/38/12A/013/img13 limits remained unchanged and no negative influence on the energy confinement or on the H-mode threshold was found. Discharges with neon radiative cooling showed the same behaviour as in the graphite divertor case.

  3. Comparison between 2D turbulence model ESEL and experimental data from AUG and COMPASS tokamaks

    DEFF Research Database (Denmark)

    Ondac, Peter; Horacek, Jan; Seidl, Jakub;

    2015-01-01

    In this article we have used the 2D fluid turbulence numerical model, ESEL, to simulate turbulent transport in edge tokamak plasma. Basic plasma parameters from the ASDEX Upgrade and COMPASS tokamaks are used as input for the model, and the output is compared with experimental observations obtained...

  4. Nitrogen retention in ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Meisl, G., E-mail: gmeisl@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, 85748 Garching (Germany); Physik-Department E28, Technische Universität München, 85747 Garching (Germany); Schmid, K.; Oberkofler, M.; Krieger, K. [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, 85748 Garching (Germany); Lisgo, S.W. [ITER Organization, FST, Route de Vinon, CS 90 046, 13067 Saint Paul Lez Durance Cedex (France); Aho-Mantila, L. [VTT, FI-02044 VTT (Finland); Reimold, F. [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, 85748 Garching (Germany)

    2015-08-15

    We investigated the transport of nitrogen through the plasma and the interaction of nitrogen with tungsten under divertor exposure conditions during nitrogen-seeding experiments in ASDEX Upgrade. Using the divertor manipulator system, tungsten samples were exposed to well-characterized L-mode plasmas with and without nitrogen seeding. We also simulated nitrogen transport and re-distribution in these discharges by self-consistent WallDYN–DIVIMP modeling. For these simulations we applied a W–N surface model based on laboratory experiments and plasma backgrounds from SOLPS. In contrast to the conclusion from Kallenbach and Dux (2010) [5] we find that the N retention in ASDEX Upgrade is in agreement with results from laboratory experiments.

  5. Operation of ASDEX Upgrade with high-Z wall coatings

    International Nuclear Information System (INIS)

    The material for plasma facing components of a future fusion device is still not decided. At present most experiments use graphite, because of its good thermo mechanical properties and the low radiation potential of carbon. Due to the high erosion yield and, especially, due to the codeposition with tritium, its use in a fusion reactor is still questionable. Based on the good experience using tungsten as divertor material in ASDEX Upgrade, which demonstrated that a divertor tokamak can be operated with a tungsten divertor without reduction of the performance, a step by step strategy was followed. Main sources of the carbon are predicted at the inner heat shield, which covers the central column. Tungsten test tiles confirm the erosion at this position due to charge exchange neutral, but also a non negligible ion sputtering component. A first step was done by siliconisation. In ASDEX Upgrade the maximal silicon concentration was 0.002. Consequently the performance of the experiment was not influenced by silicon radiation. A second step was done by tungsten coating of 1.2m2 of the inner heat shield. Experiments are done without subsequent wall coating, which would cover the tungsten. Spectroscopically measured central tungsten densities are always below ∼5*10-6 and mostly below the detection limit. Again no influence on the plasma performance parameters are found. Extrapolation to ITER conditions yields concentrations, which will not prohibit successful operation. The next step in ASDEX Upgrade will be a mostly tungsten covered inner heat shield at the next experimental campaign. (author)

  6. Ammonia production in nitrogen seeded plasma discharges in ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Rohde, V., E-mail: Volker.Rohde@ipp.mpg.de; Oberkofler, M.

    2015-08-15

    In present tokamaks nitrogen seeding is used to reduce the power load onto the divertor tiles. Some fraction of the seeded nitrogen reacts with hydrogen to form ammonia. The behaviour of ammonia in ASDEX Upgrade is studied by mass spectrometry. Injection without plasma shows strong absorption at the inner walls of the vessel and isotope exchange reactions. During nitrogen seeding in H-mode discharges the onset of a saturation of the nitrogen retention is observed. The residual gas consists of strongly deuterated methane and ammonia with almost equal amounts of deuterium and protium. This confirms the role of surface reactions in the ammonia formation. The results are consistent with findings in previous investigations. A numerical decomposition of mass spectra is under development and will be needed for quantitative evaluation of the results obtained.

  7. Experiment planning and execution workflow at ASDEX Upgrade

    International Nuclear Information System (INIS)

    We present the current workflow from experiment proposals to the actual execution and evaluation of discharges at the ASDEX Upgrade tokamak. Requests for experiments are solicited from both within the IPP and from external collaborators in the yearly call-for-proposals, checked for feasibility and compliance with the project's research goals and collected in a proposal database. During the campaign shot requests are derived from the proposals and in weekly operation meetings the requests are mapped to a schedule (shot list). Before the execution of discharges a complete set of configuration data needs to be assembled. After the execution follows the analysis (including the evaluation of the discharge as to its usefulness for the underlying proposal) and logging of the attained parameters in a physics logbook. The paper describes processes, software tools, and information management showing how they ultimately lead to an improved scientific productivity.

  8. Prediction of disruptions on ASDEX Upgrade using discriminant analysis

    International Nuclear Information System (INIS)

    In this paper, a set of simple predictive criteria, each optimized for a given type of disruption, is explored. Disruptions that occurred in the years from 2005 to 2009 in the ASDEX Upgrade tokamak are classified into several types in a first step. Then, discriminant analysis is used as the main approach to the disruption prediction and a log-linear discriminant function, constructed with five global plasma parameters that have been selected from an initial group of ten variables, is derived for the edge cooling disruptions. The function is tested off-line over 308 discharges and is shown to work reliably. It describes a clear dependence of the disruption boundary on the plasma parameters.

  9. Supersonic molecular beam fuelling at ASDEX upgrade

    International Nuclear Information System (INIS)

    The supersonic molecular beam injection (SBMI), is a fuelling technique that has been developed at Tore-Supra. This technique is based on the fast expansion of a small volume of deuterium at high pressure (typically 2-5 bars) through a nozzle into the plasma chamber. Results obtained in limiter configuration look promising: enhanced fuelling efficiency (defined as the increase in plasma particle content divided by the number of injected particles) compared to gas puff, in the range 30-50% range, have been observed. Recently one SBMI has been implemented in the divertor tokamak ASDEX upgrade in order to assess the fuelling potential of the technique. The comparison of these results with those obtained in Tore-Supra with the same hardware leads to the following conclusion: the impact of the SBMI on the plasma is much weaker. The main reason comes from the difference in the amount of particle involved per pulse that is in fact limited for technical reasons. The fuelling efficiency reaches 30% in L-mode and low density H-mode plasmas and is reduced to less than 15% in high density H-mode discharges. The amount of particle per pulse was about 10-15% of the plasma content leading to a small edge perturbation, smaller than a typical type-I ELM. A significant increase in the beam flux could improve the beam penetration and consequently the fuelling efficiency by a stronger edge cooling but the impact of such a perturbation on the confinement is still to assess. (A.C.)

  10. Overview of ASDEX Upgrade results

    Science.gov (United States)

    Zohm, H.; Angioni, C.; Arslanbekov, R.; Atanasiu, C.; Becker, G.; Becker, W.; Behler, K.; Behringer, K.; Bergmann, A.; Bilato, R.; Bobkov, V.; Bolshukhin, D.; Bolzonella, T.; Borrass, K.; Brambilla, M.; Braun, F.; Buhler, A.; Carlson, A.; Conway, G. D.; Coster, D. P.; Drube, R.; Dux, R.; Egorov, S.; Eich, T.; Engelhardt, K.; Fahrbach, H.-U.; Fantz, U.; Faugel, H.; Finken, K. H.; Foley, M.; Franzen, P.; Fuchs, J. C.; Gafert, J.; Fournier, K. B.; Gantenbein, G.; Gehre, O.; Geier, A.; Gernhardt, J.; Goodman, T.; Gruber, O.; Gude, A.; Günter, S.; Haas, G.; Hartmann, D.; Heger, B.; Heinemann, B.; Herrmann, A.; Hobirk, J.; Hofmeister, F.; Hohenöcker, H.; Horton, L. D.; Igochine, V.; Jacchia, A.; Jakobi, M.; Jenko, F.; Kallenbach, A.; Kardaun, O.; Kaufmann, M.; Keller, A.; Kendl, A.; Kim, J.-W.; Kirov, K.; Kochergov, R.; Kollotzek, H.; Kraus, W.; Krieger, K.; Kurki-Suonio, T.; Kurzan, B.; Lang, P. T.; Lasnier, C.; Lauber, P.; Laux, M.; Leonard, A. W.; Leuterer, F.; Lohs, A.; Lorenz, A.; Lorenzini, R.; Maggi, C.; Maier, H.; Mank, K.; Manso, M.-E.; Mantica, P.; Maraschek, M.; Martines, E.; Mast, K.-F.; McCarthy, P.; Meisel, D.; Meister, H.; Meo, F.; Merkel, P.; Merkel, R.; Merkl, D.; Mertens, V.; Monaco, F.; Mück, A.; Müller, H. W.; Münich, M.; Murmann, H.; Na, Y.-S.; Neu, G.; Neu, R.; Neuhauser, J.; Nguyen, F.; Nishijima, D.; Nishimura, Y.; Noterdaeme, J.-M.; Nunes, I.; Pautasso, G.; Peeters, A. G.; Pereverzev, G.; Pinches, S. D.; Poli, E.; Proschek, M.; Pugno, R.; Quigley, E.; Raupp, G.; Reich, M.; Ribeiro, T.; Riedl, R.; Rohde, V.; Roth, J.; Ryter, F.; Saarelma, S.; Sandmann, W.; Savtchkov, A.; Sauter, O.; Schade, S.; Schilling, H.-B.; Schneider, W.; Schramm, G.; Schwarz, E.; Schweinzer, J.; Schweizer, S.; Scott, B. D.; Seidel, U.; Serra, F.; Sesnic, S.; Sihler, C.; Silva, A.; Sips, A. C. C.; Speth, E.; Stäbler, A.; Steuer, K.-H.; Stober, J.; Streibl, B.; Strumberger, E.; Suttrop, W.; Tabasso, A.; Tanga, A.; Tardini, G.; Tichmann, C.; Treutterer, W.; Troppmann, M.; Urano, H.; Varela, P.; Vollmer, O.; Wagner, D.; Wenzel, U.; Wesner, F.; Westerhof, E.; Wolf, R.; Wolfrum, E.; Würsching, E.; Yoon, S.-W.; Yu, Q.; Zasche, D.; Zehetbauer, T.; Zehrfeld, H.-P.

    2003-12-01

    Recent results from the ASDEX Upgrade experimental campaigns 2001 and 2002 are presented. An improved understanding of energy and particle transport emerges in terms of a 'critical gradient' model for the temperature gradients. Coupling this to particle diffusion explains most of the observed behaviour of the density profiles, in particular, the finding that strong central heating reduces the tendency for density profile peaking. Internal transport barriers (ITBs) with electron and ion temperatures in excess of 20 keV (but not simultaneously) have been achieved. By shaping the plasma, a regime with small type II edge localized modes (ELMs) has been established. Here, the maximum power deposited on the target plates was greatly reduced at constant average power. Also, an increase of the ELM frequency by injection of shallow pellets was demonstrated. ELM free operation is possible in the quiescent H-mode regime previously found in DIII-D which has also been established on ASDEX Upgrade. Regarding stability, a regime with benign neoclassical tearing modes (NTMs) was found. During electron cyclotron current drive (ECCD) stabilization of NTMs, bgrN could be increased well above the usual onset level without a reappearance of the NTM. Electron cyclotron resonance heating and ECCD have also been used to control the sawtooth repetition frequency at a moderate fraction of the total heating power. The inner wall of the ASDEX Upgrade vessel has increasingly been covered with tungsten without causing detrimental effects on the plasma performance. Regarding scenario integration, a scenario with a large fraction of noninductively driven current (geq50%), but without ITB has been established. It combines improved confinement (tgrE/tgrITER98 ap 1.2) and stability (bgrN les 3.5) at high Greenwald fraction (ne/nGW ap 0.85) in steady state and with type II ELMy edge and would offer the possibility for long pulses with high fusion power at reduced current in ITER.

  11. Overview of ASDEX Upgrade results

    DEFF Research Database (Denmark)

    Zohm, H.; Adamek, J.; Angioni, C.;

    2009-01-01

    ASDEX Upgrade was operated with a fully W-covered wall in 2007 and 2008. Stationary H-modes at the ITER target values and improved H-modes with H up to 1.2 were run without any boronization. The boundary conditions set by the full W wall (high enough ELM frequency, high enough central heating and...... threshold in He are more favourable than in H, suggesting that He operation could allow us to assess H-mode operation in the non-nuclear phase of ITER operation....

  12. Propagation of cold pulses and heat pulses in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Experiments on electron heat transport were performed in the tokamak ASDEX Upgrade, mainly in ohmically heated plasmas, applying either edge cooling by impurity injection or edge heat pulses with ECH. Repetitive pulses within one plasma discharge were made allowing Fourier transformation of the temperature perturbation. This yields a good signal to noise ratio up to high harmonics and allows a detailed investigation of the pulse propagation. For densities lower than 1.8x1019m-3, an increase of the central electron temperature was found as the response to the edge cooling via impurity injection similar to observations made in other tokamaks. The inversion does not appear instantaneously, but with a time delay roughly compatible with diffusion. Modeling of the propagation of the cold pulses in the framework of the IFS-PPPL model yields qualitative agreement. However the predicted increase of the ion temperature is not observed experimentally on the fast time scale. The response to ECH heat pulses is not perfectly symmetrical to cold pulse experiments, but the similarities suggest a common underlying physical mechanism. No inversion of the heat pulse is found, instead the initial pulse from the edge is associated with a second, much slower heat pulse in the centre which is similar (and not symmetrical) to that of the cold pulses. It is found that the central increase is related to the arrival of the pulse close to the inversion radius and not to the initial pulse. (author)

  13. On-line prediction and mitigation of disruptions in ASDEX Upgrade

    International Nuclear Information System (INIS)

    An on-line predictor of the time to disruption has been installed on the ASDEX Upgrade tokamak. It is suitable either for avoidance of disruptions or for mitigation of those that are unavoidable. The prediction uses a neural network trained on eight plasma parameters and their time derivatives extracted from 99 disruptive discharges. The network was tested off-line over 500 discharges and was found to work reliably and to be able to predict the majority of the disruptions. The trained network was installed on-line, tested over 128 discharges and used to inject killer pellets to mitigate the disruption loads. (author)

  14. Flux surface shaping effects on tokamak edge turbulence and flows

    International Nuclear Information System (INIS)

    The influence of shaping of magnetic flux surfaces in tokamaks on gyro-fluid edge turbulence is studied numerically. Magnetic field shaping in tokamaks is mainly due to elongation, triangularity, shift and the presence of a divertor X-point. A series of tokamak configurations with varying elongation 1 ≤ κ ≥ 2 and triangularity 0 ≤ δ ≤ 0.4, and an actual ASDEX Upgrade divertor configuration are obtained with the equilibrium code HELENA and implemented into the gyro-fluid turbulence code GEM. The study finds minimal impact on the zonal flow physics itself, but strong impact on the turbulence and transport. (authors)

  15. Recent ASDEX Upgrade research in support of ITER and DEMO

    Science.gov (United States)

    H. Zohmthe ASDEX Upgrade Team; the EUROfusion MST1 Team

    2015-10-01

    Recent experiments on the ASDEX Upgrade tokamak aim at improving the physics base for ITER and DEMO to aid the machine design and prepare efficient operation. Type I edge localized mode (ELM) mitigation using resonant magnetic perturbations (RMPs) has been shown at low pedestal collisionality (νped\\ast discharge. Disruption mitigation studies using massive gas injection (MGI) can show an increased fuelling efficiency with high field side injection, but a saturation of the fuelling efficiency is observed at high injected mass as needed for runaway electron suppression. Large locked modes can significantly decrease the fuelling efficiency and increase the asymmetry of radiated power during MGI mitigation. Concerning power exhaust, the partially detached ITER divertor scenario has been demonstrated at Psep/R = 10 MW m-1 in ASDEX Upgrade, with a peak time averaged target load around 5 MW m-2, well consistent with the component limits for ITER. Developing this towards DEMO, full detachment was achieved at Psep/R = 7 MW m-1 and stationary discharges with core radiation fraction of the order of DEMO requirements (70% instead of the 30% needed for ITER) were demonstrated. Finally, it remains difficult to establish the standard ITER Q = 10 scenario at low q95 = 3 in the all-tungsten (all-W) ASDEX Upgrade due to the observed poor confinement at low βN. This is mainly due to a degraded pedestal performance and hence investigations at shifting the operational point to higher βN by lowering the current have been started. At higher q95, pedestal performance can be recovered by seeding N2 as well as CD4, which is interpreted as improved pedestal stability due to the decrease of bootstrap current with increasing Zeff. Concerning advanced scenarios, the upgrade of ECRH power has allowed experiments with central ctr-ECCD to modify the q-profile in improved H-mode scenarios, showing an increase in confinement at still good MHD stability with flat elevated q-profiles at values

  16. Overview of ASDEX Upgrade results

    International Nuclear Information System (INIS)

    Significant progress has been made on ASDEX Upgrade during the last two years in the basic understanding of transport, in the extension of the improved H-mode in parameter space and towards an integrated operating scenario, and in the development of control methods for major performance limiting instabilities. Highlights were the understanding of particle transport and the control of impurity accumulation based on it, the satisfactory operation with predominantly tungsten-clad walls, the improved H-mode operation over density ranges and for temperature ratios covering the ITER requirements on ν*, n/nGW and Te/Ti, the ELM frequency control by pellet injection, and the optimisation of NTM suppression by DC-ECCD through variation of the launching angle. From these experiments an integrated scenario has emerged which extrapolates to a 50 % improvement in nTτ or a 30 % reduction of the required current compared to the ITER base-line assumptions, with moderately peaked electron and controllable high-Z density profiles. (author)

  17. Characterization and interpretation of the Edge Snake in between type-I edge localized modes at ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Sommer, F; Guenter, S; Kallenbach, A; Maraschek, M; Boom, J; Fischer, R; Hicks, N; Reiter, B; Wolfrum, E [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2, D-85748 Garching, EURATOM Association (Germany); Luhmann, N C Jr [University of California at Davis, Davis, CA 95616 (United States); Park, H K [POSTECH, Pahang, Gyeongbuk 790-784 (Korea, Republic of); Wenninger, R, E-mail: fabian.sommer@ipp.mpg.de [Universitaetssternwarte der Ludwig-Maximilians-Universitaet, D-81679 Muenchen (Germany)

    2011-08-15

    A new magnetohydrodynamic instability called the 'Edge Snake', which was found in 2006 at the tokamak ASDEX Upgrade during type-I ELMy H-modes, is investigated. It is located within the separatrix in the region of high temperature and density gradients and has a toroidal mode number of n = 1. The Edge Snake consists of a radially and poloidally strongly localized current wire, in which the temperature and density profiles flatten. This significant reduction in pressure gradient leads to a reduction in the neoclassical Bootstrap current and can plausibly explain the drive of the instability. The experimental observations point towards a magnetic island with a defect current inside the O-point of the island. The Edge Snake is compared with similar instabilities at JET, DIII-D and ASDEX Upgrade.

  18. Laboratory astrophysics on ASDEX Upgrade: Measurements and analysis of K-shell O, F, and Ne spectra in the 9 - 20 A region

    Science.gov (United States)

    Hansen, S. B.; Fournier, K. B.; Finkenthal, M. J.; Smith, R.; Puetterich, T.; Neu, R.

    2006-01-01

    High-resolution measurements of K-shell emission from O, F, and Ne have been performed at the ASDEX Upgrade tokamak in Garching, Germany. Independently measured temperature and density profiles of the plasma provide a unique test bed for model validation. We present comparisons of measured spectra with calculations based on transport and collisional-radiative models and discuss the reliability of commonly used diagnostic line ratios.

  19. Comparison of a radial fractional transport model with tokamak experiments

    Science.gov (United States)

    Kullberg, A.; Morales, G. J.; Maggs, J. E.

    2014-03-01

    A radial fractional transport model [Kullberg et al., Phys. Rev. E 87, 052115 (2013)], that correctly incorporates the geometric effects of the domain near the origin and removes the singular behavior at the outer boundary, is compared to results of off-axis heating experiments performed in the Rijnhuizen Tokamak Project (RTP), ASDEX Upgrade, JET, and DIII-D tokamak devices. This comparative study provides an initial assessment of the presence of fractional transport phenomena in magnetic confinement experiments. It is found that the nonlocal radial model is robust in describing the steady-state temperature profiles from RTP, but for the propagation of heat waves in ASDEX Upgrade, JET, and DIII-D the model is not clearly superior to predictions based on Fick's law. However, this comparative study does indicate that the order of the fractional derivative, α, is likely a function of radial position in the devices surveyed.

  20. Comparison of a radial fractional transport model with tokamak experiments

    Energy Technology Data Exchange (ETDEWEB)

    Kullberg, A., E-mail: kulladam@ucla.edu; Morales, G. J.; Maggs, J. E. [Department of Physics and Astronomy, University of California, Los Angeles, Los Angeles, California 90095 (United States)

    2014-03-15

    A radial fractional transport model [Kullberg et al., Phys. Rev. E 87, 052115 (2013)], that correctly incorporates the geometric effects of the domain near the origin and removes the singular behavior at the outer boundary, is compared to results of off-axis heating experiments performed in the Rijnhuizen Tokamak Project (RTP), ASDEX Upgrade, JET, and DIII-D tokamak devices. This comparative study provides an initial assessment of the presence of fractional transport phenomena in magnetic confinement experiments. It is found that the nonlocal radial model is robust in describing the steady-state temperature profiles from RTP, but for the propagation of heat waves in ASDEX Upgrade, JET, and DIII-D the model is not clearly superior to predictions based on Fick's law. However, this comparative study does indicate that the order of the fractional derivative, α, is likely a function of radial position in the devices surveyed.

  1. Statistical analyses of local transport coefficients in ohmic ASDEX discharges

    International Nuclear Information System (INIS)

    Tokamak energy transport is still an unsolved problem. Many theoretical models have been developed, which try to explain the anomalous high energy-transport coefficients. Up to now these models have been applied to global plasma parameters. A comparison of transport coefficients with global confinement time is only conclusive if the transport is dominated by one process across the plasma diameter. This, however, is not the case in most of the Ohmic confinement regimes, where at least three different transport mechanisms play an important role. Sawtooth activity leads to an increase in energy transport in the plasma centre. In the intermediate region turbulent transport is expected. Candidates here are drift waves and resistive fluid turbulences. At the edge, ballooning modes or rippling modes could dominate the transport. For the intermediate region, one can deduce theoretical scaling laws for τE from turbulent theories. Predicted scalings reproduce the experimentally found density dependence of τE in the linear Ohmic confinement regime (LOC) and the saturated regime (SOC), but they do not show the correct dependence on the isotope mass. The relevance of these transport theories can only be tested in comparing them to experimental local transport coefficients. To this purpose we have performed transport calculations on more than a hundred Ohmic ASDEX discharges. By Principal Component Analysis we determine the dimensionless components which dominate the transport coefficients and we compare the results to the predictions of various theories. (orig.)

  2. Simulations of global electrostatic microinstabilities in ASDEX Upgrade discharges

    Science.gov (United States)

    Bottino, A.; Peeters, A. G.; Sauter, O.; Vaclavik, J.; Villard, L.

    2004-01-01

    Electrostatic microinstabilities in ion internal barrier (ITB) and H-mode discharges of the ASDEX Upgrade tokamak [O. Gruber, R. Arslanbekov, C. Atanasiu et al., Nucl. Fusion 41, 1369 (2001)] have been investigated with a full radius gyrokinetic code. The code models linear stability and includes the effect of an equilibrium radial electric field and trapped electrons. In order to simulate plasmas in experimental conditions [k⊥ρL˜O(1)], the long wavelength approximation in the quasineutrality equation has been replaced by a Padé expansion of the modified Bessel function. Results show that the E×B flow, induced by the radial electric field, changes the linear stability of the dominant ion temperature gradient modes. The electrostatic potential eddies are tilted by the sheared flow thus reducing the radial extent and the growth rate of modes. However, the finite value of the flow has a stabilizing effect too; the most unstable modes are shifted away from the unfavorable curvature region leading to lower linear growth rates. In addition to this at least two other mechanisms give an important contribution to the stabilization in the ITB region; the reverse shear profile itself and, to a lesser degree, the local value of the temperature ratio, τ=Te/Ti.

  3. Simulations of global electrostatic microinstabilities in ASDEX Upgrade discharges

    International Nuclear Information System (INIS)

    Electrostatic microinstabilities in ion internal barrier (ITB) and H-mode discharges of the ASDEX Upgrade tokamak [O. Gruber, R. Arslanbekov, C. Atanasiu et al., Nucl. Fusion 41, 1369 (2001)] have been investigated with a full radius gyrokinetic code. The code models linear stability and includes the effect of an equilibrium radial electric field and trapped electrons. In order to simulate plasmas in experimental conditions [kperpendicularρL∼O(1)], the long wavelength approximation in the quasineutrality equation has been replaced by a Pade expansion of the modified Bessel function. Results show that the ExB flow, induced by the radial electric field, changes the linear stability of the dominant ion temperature gradient modes. The electrostatic potential eddies are tilted by the sheared flow thus reducing the radial extent and the growth rate of modes. However, the finite value of the flow has a stabilizing effect too; the most unstable modes are shifted away from the unfavorable curvature region leading to lower linear growth rates. In addition to this at least two other mechanisms give an important contribution to the stabilization in the ITB region; the reverse shear profile itself and, to a lesser degree, the local value of the temperature ratio, τ=Te/Ti

  4. Electron heat transport studies using transient phenomena in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Experiments in tokamaks suggest that a critical gradient length may cause the resilient behavior of Te profiles, in the absence of ITBs. This agrees in general with ITG/TEM turbulence physics. Experiments in ASDEX Upgrade using modulation techniques with ECH and/or cold pulses demonstrate the existence of a threshold in R/LTe when Te>Ti and Te≤Ti. For Te>Ti linear stability analyses indicate that electron heat transport is dominated by TEM modes. They agree in the value of the threshold (both Te and ne) and for the electron heat transport increase above the threshold. The stabilization of TEM modes by collisions yielded by gyro-kinetic calculations, which suggests a transition from TEM to ITG dominated transport at high collisionality, is experimentally demonstrated by comparing heat pulse and steady-state diffusivities. For the Te∼Ti discharges above the threshold the resilience, normalized by Te3/2 , is similar to that of the TEM dominated cases, despite very different conditions. The heat pinch predicted by fluid modeling of ITG/TEM turbulence is investigated by perturbative transport in off-axis ECH-heated discharges. (author)

  5. The multiple facets of ohmic confinement in ASDEX

    International Nuclear Information System (INIS)

    The scaling of the energy confinement time with plasma density and current has been investigated for Ohmically heated tokamak discharges in ASDEX. The linear dependence τE ∝ anti ne is maintained in the high density Improved Ohmic Confinement (IOC) regime with peaked density profiles. The peaking of the radial density profile can be brought about by reducing the net power flow through the plasma surface thereby effecting a reduction of the edge density. Tailoring of the radiation profile with the addition of low-Z impurities e.g. neon gives access to the IOC regime under conditions where otherwise the degraded Saturated Ohmic Confinement (SOC) behavior prevails. The energy confinement time increases with current and decreases with heating poweer also in Ohmic discharges as shown by a statistical analysis. But with the intrinsic coupling between power and current, both relationships cancel and τE becomes independent of POH and Ip. The two most prominent features of Ohmic confinement can therefore be explained on the basis of simple physical models. (orig.)

  6. Poloidal asymmetries of heavy impurities in the ASDEX upgrade plasma

    International Nuclear Information System (INIS)

    For heavy and highly charged impurities multiple mechanisms exist that produce non-constant impurities densities on the flux surfaces. As for neoclassical and turbulent transport models such an asymmetry is highly importance an effort is launched to experimentally characterize the asymmetries comparing them with theoretical expectations. In the ASDEX upgrade tokamak (AUG) is routinely observed increase of outboard tungsten density in fast rotating plasma. This asymmetry is caused by the centrifugal force pushing tungsten ions outward due to its high mass. Furthermore, the high charge makes heavy impurities sensitive to poloidal variations of the plasma potential. The variation can be generated by magnetic trapped ions heated by RF heating. In such a case, the presence of an inboard asymmetry or at least the absence of an outboard asymmetry due to the centrifugal force can be observed. Finally, ion-impurity friction enhanced by the large charge of the impurity ions may cause a relatively weak up-down asymmetry of the impurity density. The aim of this poster is to show first evidence of these asymmetries in the AUG plasmas, the description of the used methodology, and to compare with theoretical models based on the parallel force balance.

  7. First 50 pps Thomson scattering diagnostics in a tokamak

    International Nuclear Information System (INIS)

    Electron temperature and density measurements by Thomson scattering were performed for the first time for the whole duration of a tokamak discharge. A 50 pps Nd:YAG laser at 1.06 μm was used in ASDEX in combination with Si avalanche photodiode detectors. Density calibration was done by rotational anti-Stokes Raman scattering from hydrogen. The system is used for measurements at electron densities of as low as 2 x 1012 cm-3. (orig.)

  8. Fast ion temperature measurements using ball-pen probes in the SOL of ASDEX Upgrade during L-mode

    Czech Academy of Sciences Publication Activity Database

    Adámek, Jiří; Horáček, Jan; Müller, H. W.; Schrittwieser, R.; Tichý, M.; Nielsen, A.H.

    Mulhouse : European Physical Society, 2011 - (Becoulet, A.; Hoang, T.; Stroth, U.), P1.059-P1.059 ISBN 2-914771-68-1. - (EPS. 35G). [European Physical Society Conference on Plasma Physics /38th./. Strasbourg (FR), 27.06.2011-01.07.2011] R&D Projects: GA AV ČR KJB100430901; GA MŠk 7G09042 Institutional research plan: CEZ:AV0Z20430508 Keywords : ASDEX Upgrade * ball-pen probes * L-mode * SOL * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics http://ocs.ciemat.es/EPS2011PAP/pdf/P1.059.pdf

  9. Overview of nitrogen-15 application as a tracer gas for material migration and retention studies in tokamaks

    International Nuclear Information System (INIS)

    Experimental and analytical procedures related to the application of nitrogen-15 isotope for material migration studies have been developed and used for tracer experiments in the TEXTOR and ASDEX-Upgrade tokamaks in order to assess the retention of nitrogen in plasma-facing components made of graphite and tungsten. The surface study was performed by time-of-flight heavy ion elastic recoil detection analysis and by means of nuclear reaction analysis based on the 15N(p,γα)12C process. In both tokamaks nitrogen retention has exceeded 10% of the injected gas. In ASDEX-Upgrade the largest fraction of 15N has been detected on protruding parts near the injection port, while around 4% has been found in the divertor. The ASDEX-Upgrade results have also been modeled. Helium trapping has been measured in deposits containing tungsten and nitrogen. (paper)

  10. High speed cine film studies of plasma behaviour and plasma surface interactions in tokamaks

    International Nuclear Information System (INIS)

    High speed cine photography is a useful diagnostic aid for studying plasma behaviour and plasma surface interactions. Several workers have filmed discharges in tokamaks including ASDEX, DITE, DIVA, ISX, JFT2, TFR and PLT. These films are discussed and examples given of the observed phenomena which include plasma limiter interactions, diverted discharges, disruptions, magnetic islands and moving glowing objects often known as 'UFOs'. Examples of plasma structures in ASDEX and DITE not previously published are also given. The paper also reports experiments in DITE to determine the origin of UFOs. (orig.)

  11. High speed cine film studies of plasma behaviour and plasma surface interactions in tokamaks

    Science.gov (United States)

    Goodall, D. H. J.

    1982-12-01

    High speed cine photography is a useful diagnostic aid for studying plasma behaviour and plasma surface interactions. Several workers have filmed discharges in tokamaks including ASDEX, DITE, DIVA, ISX, JFT2, TFR and PLT. These films are discussed and examples given of the observed phenomena which include plasma limiter interactions, diverted discharges, disruptions, magnetic islands and moving glowing objects often known as 'UFOs'. Examples of plasma structures in ASDEX and DITE not previously published are also given. The paper also reports experiments in DITE to determine the origin of UFOs.

  12. High speed cine film studies of plasma behaviour and plasma surface interactions in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Goodall, D.H.J. (Euratom/UKAEA Fusion Association, Abingdon (UK). Culham Lab.)

    High speed cine photography is a useful diagnostic aid for studying plasma behaviour and plasma surface interactions. Several workers have filmed discharges in tokamaks including ASDEX, DITE, DIVA, ISX, JFT2, TFR and PLT. These films are discussed and examples given of the observed phenomena which include plasma limiter interactions, diverted discharges, disruptions, magnetic islands and moving glowing objects often known as 'UFOs'. Examples of plasma structures in ASDEX and DITE not previously published are also given. The paper also reports experiments in DITE to determine the origin of UFOs.

  13. Consistency between real and synthetic fast-ion measurements at ASDEX Upgrade

    Science.gov (United States)

    Rasmussen, J.; Nielsen, S. K.; Stejner, M.; Geiger, B.; Salewski, M.; Jacobsen, A. S.; Korsholm, S. B.; Leipold, F.; Michelsen, P. K.; Moseev, D.; Schubert, M.; Stober, J.; Tardini, G.; Wagner, D.; The ASDEX Upgrade Team

    2015-07-01

    Internally consistent characterization of the properties of the fast-ion distribution from multiple diagnostics is a prerequisite for obtaining a full understanding of fast-ion behavior in tokamak plasmas. Here we benchmark several absolutely-calibrated core fast-ion diagnostics at ASDEX Upgrade by comparing fast-ion measurements from collective Thomson scattering, fast-ion {{\\text{D}}α} spectroscopy, and neutron rate detectors with numerical predictions from the TRANSP/NUBEAM transport code. We also study the sensitivity of the theoretical predictions to uncertainties in the plasma kinetic profiles. We find that theory and measurements generally agree within these uncertainties for all three diagnostics during heating phases with either one or two neutral beam injection sources. This suggests that the measurements can be described by the same model assuming classical slowing down of fast ions. Since the three diagnostics in the adopted configurations probe partially overlapping regions in fast-ion velocity space, this is also consistent with good internal agreement among the measurements themselves. Hence, our results support the feasibility of combining multiple diagnostics at ASDEX Upgrade to reconstruct the fast-ion distribution function in 2D velocity space.

  14. Cooling water calorimetry measuring results from the first years of ASDEX Upgrade operation

    International Nuclear Information System (INIS)

    At the tokamak ASDEX Upgrade an extensive cooling water calorimetry system was installed. This system has measured the toroidal and poloidal distributions of the energy deposition by monitoring the temperature rise of the cooling water in 80 separate cooling units in the divertor plates and the central heat shield. The measurements show, that there exist no toroidal asymmetries in the energy deposition on the divertor plates for all kinds of ohmic discharges and for ICRH discharges with a toroidal magnetic field directed opposite to the plasma current. However, Neutral Beam Injection causes a toroidal asymmetric energy deposition profile. Furthermore the reduction of the poloidal in-out asymmetry of the energy load at the divertor plates due to magnetic field reversion was detected. Making up the general energy balance of ASDEX Upgrade, adding the energy detected by the cooling water calorimetry system and the radiation loss energy measured by the bolometry diagnostic, one gets 92%-97% of the energy input. (orig./HD)

  15. Consistency between real and synthetic fast-ion measurements at ASDEX Upgrade

    DEFF Research Database (Denmark)

    Rasmussen, Jesper; Nielsen, Stefan Kragh; Pedersen, Morten Stejner;

    2015-01-01

    by comparing fast-ion measurements from collective Thomson scattering, fast-ion spectroscopy, and neutron rate detectors with numerical predictions from the TRANSP/NUBEAM transport code. We also study the sensitivity of the theoretical predictions to uncertainties in the plasma kinetic profiles. We find......Internally consistent characterization of the properties of the fast-ion distribution from multiple diagnostics is a prerequisite for obtaining a full understanding of fast-ion behavior in tokamak plasmas. Here we benchmark several absolutely-calibrated core fast-ion diagnostics at ASDEX Upgrade...... that theory and measurements generally agree within these uncertainties for all three diagnostics during heating phases with either one or two neutral beam injection sources. This suggests that the measurements can be described by the same model assuming classical slowing down of fast ions. Since the three...

  16. Design and performance of the collective Thomson scattering receiver at ASDEX Upgrade

    DEFF Research Database (Denmark)

    Furtula, Vedran; Salewski, Mirko; Leipold, Frank;

    2012-01-01

    Here we present the design of the fast-ion collective Thomson scattering receiver for millimeter wave radiation installed at ASDEX Upgrade, a tokamak for fusion plasma experiments. The receiver can detect spectral power densities of a few eV against the electron cyclotron emission background...... is divided into 50 IF channels tightly spaced in frequency space. The channels are terminated by square-law detector diodes that convert the signal power into DC voltages. We present measurements of the transmission characteristics and performance of the main receiver components operating at mm......-wave frequencies (notch, bandpass, and lowpass filters, a voltage-controlled variable attenuator, and an isolator), the down-converter unit, and the IF components (amplifiers, bandpass filters, and detector diodes). Furthermore, we determine the performance of the receiver as a unit through spectral response...

  17. Poloidal asymmetric flow and current relaxation of ballooned transport during I-phase in ASDEX Upgrade

    Science.gov (United States)

    Manz, P.; Birkenmeier, G.; Fuchert, G.; Cavedon, M.; Conway, G. D.; Maraschek, M.; Medvedeva, A.; Mink, F.; Scott, B. D.; Shao, L. M.; Stroth, U.

    2016-05-01

    Turbulence driven poloidal asymmetric parallel flow and current perturbations are studied for tokamak plasmas of circular geometry. Whereas zonal flows can lead to in-out asymmetry of parallel flows and currents via the Pfirsch-Schlüter mechanism, ballooned transport can result in an up-down asymmetry due to the Stringer spin-up mechanism. Measurements of up-down asymmetric parallel current fluctuations occurring during the I-phase in ASDEX Upgrade are not responses to the equilibrium by the Pfirsch-Schlüter current, but can be interpreted as a response to strongly ballooned plasma transport coupled with the Stringer spin-up mechanism. A good agreement of the experimental measured limit-cycle frequencies during I-phase with the Stringer spin-up relaxation frequency is found.

  18. Erosion of tungsten and steel in the main chamber of ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Hakola, A., E-mail: antti.hakola@vtt.fi [VTT, P.O. Box 1000, 02044 VTT (Finland); Koivuranta, S.; Likonen, J. [VTT, P.O. Box 1000, 02044 VTT (Finland); Herrmann, A.; Maier, H.; Mayer, M. [Max-Planck-Institut für Plasmaphysik, Boltzmannstrasse 2, 85748 Garching (Germany); Neu, R. [Max-Planck-Institut für Plasmaphysik, Boltzmannstrasse 2, 85748 Garching (Germany); Technische Universität München, 85747 Garching (Germany); Rohde, V. [Max-Planck-Institut für Plasmaphysik, Boltzmannstrasse 2, 85748 Garching (Germany)

    2015-08-15

    We have investigated net erosion and deposition of W and P92 steel in ASDEX Upgrade during its full-W operational phase. The outer divertor and the outer midplane are the strongest net erosion region for W, with rates up to 0.12 nm/s and 0.05 nm/s, respectively. The eroded W is transported via the scrape-off layer plasma and predominantly deposited in the upper (20–30%) and inner divertors (40–60%). The inner midplane does not contribute significantly to the erosion–deposition balance such that the remaining W is deposited in shadowed areas of the tokamak. Steel is eroded 3–10 times faster than W but could be used at the top and inner parts of the main chamber where the erosion rate is ∼0.01 nm/s.

  19. Validation of gyrokinetic modelling of light impurity transport including rotation in ASDEX Upgrade

    CERN Document Server

    Casson, F J; Angioni, C; Camenen, Y; Dux, R; Fable, E; Fischer, R; Geiger, B; Manas, P; Menchero, L; Tardini, G

    2013-01-01

    Upgraded spectroscopic hardware and an improved impurity concentration calculation allow accurate determination of boron density in the ASDEX Upgrade tokamak. A database of boron measurements is compared to quasilinear and nonlinear gyrokinetic simulations including Coriolis and centrifugal rotational effects over a range of H-mode plasma regimes. The peaking of the measured boron profiles shows a strong anti-correlation with the plasma rotation gradient, via a relationship explained and reproduced by the theory. It is demonstrated that the rotodiffusive impurity flux driven by the rotation gradient is required for the modelling to reproduce the hollow boron profiles at higher rotation gradients. The nonlinear simulations validate the quasilinear approach, and, with the addition of perpendicular flow shear, demonstrate that each symmetry breaking mechanism that causes momentum transport also couples to rotodiffusion. At lower rotation gradients, the parallel compressive convection is required to match the mos...

  20. Global migration of impurities in tokamaks

    International Nuclear Information System (INIS)

    The migration of impurities in tokamaks has been studied with the help of tracer-injection (13C and 15N) experiments in JET and ASDEX Upgrade since 2001. We have identified a common pattern for the migrating particles: scrape-off layer flows drive impurities from the low-field side towards the high-field side of the vessel. Migration is also sensitive to the density and magnetic configuration of the plasma, and strong local variations in the resulting deposition patterns require 3D treatment of the migration process. Moreover, re-erosion of the deposited particles has to be taken into account to properly describe the migration process during steady-state operation of the tokamak. (paper)

  1. Automated in situ line of sight calibration of ASDEX Upgrade bolometers

    International Nuclear Information System (INIS)

    The ITER Bolometer Robot Test Rig (IBOROB) is a robot-based diagnostic tool, which allows the measurement of the lines of sight (LOS) of the ITER bolometer prototypes. Up to now, it was only used as a LOS characterization device for the ITER collimator development. IBOROB was further developed and can now be operated in ASDEX Upgrade during a regular maintenance shutdown. At present, once a diagnostic like the bolometry is mounted inside the vessel, the actual LOS orientations are not measured, they are derived from CAD. The new procedure allows the fully automatic three-dimensional in situ measurement of bolometer LOS. The spatial distribution, the poloidal and toroidal alignment in the experiment coordinate system (CS), can be determined. The absolute accuracy, in reference to the tokamak CS, is provided by an additional calibration performed with a measurement arm by FARO Technologies Inc. Therefore, the amount of misalignment from the theoretical expectations can be quantified. In addition specific camera type dependencies such as internal camera reflections can be identified. Due to the high position accuracy of the robot, the LOS can be resolved with a spatial resolution of up to 0.1°. The method is explained in detail and results from two exemplary bolometer foil cameras obtained in a first set-up in ASDEX Upgrade are presented. The different steps and components needed to apply the measurements in the vessel are described with a focus on the constraints, e.g. geometrical, for an application of this method in a tokamak. Finally the consequences of the results are extrapolated to ITER and evaluated

  2. Automated in situ line of sight calibration of ASDEX Upgrade bolometers

    Energy Technology Data Exchange (ETDEWEB)

    Penzel, F., E-mail: florian.penzel@ipp.mpg.de [Max-Planck-Institute for Plasmaphysics, EURATOM Association, Garching (Germany); Meister, H.; Bernert, M.; Sehmer, T.; Trautmann, T.; Kannamüller, M.; Koll, J. [Max-Planck-Institute for Plasmaphysics, EURATOM Association, Garching (Germany); Koch, A.W. [Institute for Measurement Systems and Sensor Technology, Technische Universität München (Germany)

    2014-10-15

    The ITER Bolometer Robot Test Rig (IBOROB) is a robot-based diagnostic tool, which allows the measurement of the lines of sight (LOS) of the ITER bolometer prototypes. Up to now, it was only used as a LOS characterization device for the ITER collimator development. IBOROB was further developed and can now be operated in ASDEX Upgrade during a regular maintenance shutdown. At present, once a diagnostic like the bolometry is mounted inside the vessel, the actual LOS orientations are not measured, they are derived from CAD. The new procedure allows the fully automatic three-dimensional in situ measurement of bolometer LOS. The spatial distribution, the poloidal and toroidal alignment in the experiment coordinate system (CS), can be determined. The absolute accuracy, in reference to the tokamak CS, is provided by an additional calibration performed with a measurement arm by FARO Technologies Inc. Therefore, the amount of misalignment from the theoretical expectations can be quantified. In addition specific camera type dependencies such as internal camera reflections can be identified. Due to the high position accuracy of the robot, the LOS can be resolved with a spatial resolution of up to 0.1°. The method is explained in detail and results from two exemplary bolometer foil cameras obtained in a first set-up in ASDEX Upgrade are presented. The different steps and components needed to apply the measurements in the vessel are described with a focus on the constraints, e.g. geometrical, for an application of this method in a tokamak. Finally the consequences of the results are extrapolated to ITER and evaluated.

  3. Recent ECRH results in ASDEX Upgrade

    International Nuclear Information System (INIS)

    We report about experiments in ASDEX Upgrade using our ECRH system with f = 140 GHz, P = 4 x 0.5 MW, and T = 2 sec. The following topics are covered: studies of modulated power deposition, studies of the electron heat transport via power balance analysis and heat wave analysis and a comparison with turbulent transport theory, generation of an internal transport barrier for the electron heat flux, impact of ECRH on particle and impurity transport, and studies related to neoclassical tearing modes and to sawteeth. (authors)

  4. Nitrogen migration and retention in ASDEX upgrade

    International Nuclear Information System (INIS)

    To limit the power load in high-power plasma operation, impurity seeding is mandatory. Nitrogen has been established as optimal choice in ASDEX upgrade. However, as N is subject to wall pumping, a self-consistent model of the N source flux distribution is required. N retention in tungsten was studied in laboratory experiments under well-defined exposure conditions. The applicability of the so established model of W-N interaction was tested by experiments in ASDEX upgrade. W samples were exposed to plasmas with and without N seeding and analyzed by ion beam analysis. Using these data as boundary condition, N transport and re-distribution in the plasma were studied by self-consistent WallDYN-DIVIMP modelling. The dynamic change of the N erosion source at plasma exposed W surfaces was then computed by WallDYN using an improved W-N surface model. First simulations show, in agreement with the experiment, a strong rise of the N re-erosion flux within the first second. By this approach the experimental results from sample analysis, spectroscopy and N pumped by the vacuum system can be interpreted for the first time within a unified self-consistent model.

  5. Divertor geometry optimization for ASDEX Upgrade

    International Nuclear Information System (INIS)

    One of the critical questions to be solved for ITER (or any other reactor) is the power exhaust problem (compatible with particle exhaust). Optimized divertors have to be tested in existing geometries based mainly on the idea of closing them very efficiently to the main chamber and, by the choice of the plate and baffle geometry, positively influencing the flow pattern of hydrogen favoring good impurity entrainment. Also, for ASDEX Upgrade there is an experimental necessity for an improved divertor due to the increased heating power (24 MW will be available from 1997 on compared to the present 18 MW). We present the optimization strategy for the divertor II of ASDEX Upgrade, using elaborate numerical models and codes (B2-Eirene) as well as simple models. We start with the choice of a proper target plate geometry, and then further discuss how main chamber and private flux baffling will be done, and how this affects neutral recirculation pattern and pumping properties. For the final configuration the impurity entrainment properties are analyzed. (orig.)

  6. The H-mode of ASDEX

    International Nuclear Information System (INIS)

    This paper is a review of the work on the H-mode done on ASDEX since its discovery in 1982. In detail, it presents (1) the development of the plasma profiles - steep edge gradients and flat bulk profiles, (2) the MHD properties resulting from the profile changes, including an extensive stability analysis, (3) the impurity development with special emphasis on the MHD aspects and on neoclassical impurity transport effects in quiescent H-phases, (4) a detailed study of the edge properties including the evidence of 3-dimensional distortions at the edge. The part on confinement encompasses scaling studies and the results of transport analysis. The power threshold of the H-mode is found to depend slightly on the density but hardly on the toroidal field or current. The operational range of the H-mode includes new results on the limiter H-mode of ASDEX and on the development of the H-mode under beam current drive conditions. Several experiments are described which demonstrate the crucial role of the edge electron temperature in the H-mode transition. New material on magnetic and density fluctuation studies at the plasma edge within the edge transport barrier is presented. Finally, the findings on ADSEX are compared with those on other machines and are used to test various H-mode theories. (orig.)

  7. New operational spaces for the electron cyclotron resonance heating at ASDEX upgrade

    International Nuclear Information System (INIS)

    In this thesis, new electron cyclotron resonance heating (ECRH) scenarios were developed for an extension of the operational space at the tokamak ASDEX Upgrade in view of ITER compatibility. In the last years, the first wall material at ASDEX Upgrade was changed from graphite to tungsten, and the ECRH is needed to control the tungsten concentration in the plasma core. But, in ITER-like plasma discharges at ASDEX Upgrade, the usage of the ECRH in the typically used second harmonic extraordinary polarised mode (X2 mode) is limited. In these ITER-scenarios a small safety factor should be achieved, which implements an increase of the plasma current at ASDEX Upgrade. A higher plasma current and a high confinement lead to a raised density and for the ITER scenario to an electron density above the cutoff of the X2 mode at ASDEX Upgrade. Therefore, the X2 mode is reflected at the cutoff layer and cannot be used for central heating and the control of the tungsten concentration. One possibility to overcome this problem is to apply the third harmonic mode at reduced magnetic field. Here the cutoff is increased by 33% due to the dependence on the magnetic field. However, at the reachable plasma parameters at the reduced field the absorption of the X3 mode is incomplete (60-70 %) and the shine-trough power can destroy microwave sensitive components in ASDEX Upgrade. To solve this problem the magnetic field has to be optimized. A slightly increased magnetic field from 1.7 T to 1.8 T moves the second harmonic resonance in the region of confined plasma with high temperatures and density, so that this resonance can act as beam dump. The deposition in the plasma core is still central enough for the tungsten control ability of the ECRH. The benefit of the beam dump was verified in experiments with two different magnetic fields (1.7 T and 1.8 T). In case of the higher magnetic field, the stray radiation was reduced; simultaneously the electron temperature was increased. In addition

  8. Multi-channel Langmuir-probe and Hα-measurements of edge fluctuations on ASDEX

    International Nuclear Information System (INIS)

    The anomalous transport observed in tokamaks is caused by turbulent fluctuations, the nature of which is still poorly understood. Diagnostic difficulties are one major reason for this lack of understanding, at least with respect to the bulk plasma. The plasma edge, however, is accessible by several diagnostics permitting localized measurements of different parameters with good spatial and temporal resolution. For this reason one can hope to obtain enough information about edge fluctuations to permit the development of theoretical models. Different ranges of plasma parameters and the lack of closed magnetic surfaces distinguish this plasma zone from the bulk plasma. Edge turbulence might well involve other mechanisms than the turbulence in the bulk. Like in many limiter tokamaks Langmuir probes were used in the ASDEX divertor device for measurements of the floating potential and of the ion saturation current. Under certain assumptions the electron density and the plasma potential can be derived from these data. Observation of the Hα-light emitted from the edge in the vicinity of a neutral gas source yields information about the electron density. While probe measurements are more suitable for quantitative evaluations including the calculation of the local particle flux the Hα-method is not calibrated and integrates radially over the edge. It is however applicable in situations where probes fail because of excessive heat load. With 16-channel arrays both methods permit spatial correlations and wavenumber spectra to be determined without any further assumptions. (orig./AH)

  9. Fueling efficiency of gas puffing in ASDEX

    Science.gov (United States)

    Mayer, H.-M.; Wagner, F.; Becker, G.; Behringer, K.; Campbell, D.; Eberhagen, A.; Engelhardt, W.; Fussman, G.; Gehre, O.; Gierke, G. v.; Glock, E.; Haas, G.; Huang, M.; Karger, F.; Keilhacker, M.; Klüber, O.; Niedermeyer, H.; Poschenrieder, W.; Rapp, H.; Schneider, F.; Siller, G.; Steuer, K.-H.; Venus, G.

    1982-12-01

    The fueling efficiency for gas puffing, i.e. the fraction of the external gas flux that is ionized inside the separatrix, is reduced in divertor discharges since part of it is ionized in the scrape-off layer and pumped off by the divertor. The fueling efficiency is determined by switching-off the gas feed during the stationary phase of a discharge and dividing the time derivative of the total number of particles inside the separatrix by the external gas flux. The determination of this time derivative must take into account profile changes. In ASDEX the fueling efficiency ranges from close to 1.0 for discharges with a stainless steel poloidal limiter and decreases to about 0.2 at high densities ( 6 × 10 13 cm -3 line average) for diverted discharges. These results are compared with estimates of the fueling efficiency which include molecular disintegration, plasma albedo for neutral atoms and imperfect wall reflection.

  10. The ASDEX integrated data analysis system AIDA

    International Nuclear Information System (INIS)

    Since about two years, the ASDEX integrated data analysis system (AIDA), which combines the database (DABA) and the statistical analysis system (SAS), is successfully in operation. Besides a considerable, but meaningful, reduction of the 'raw' shot data, it offers the advantage of carefully selected and precisely defined datasets, which are easily accessible for informative tabular data overviews (DABA), and multi-shot analysis (SAS). Even rather complicated, statistical analyses can be performed efficiently within this system. In this report, we want to summarise AIDA's main features, give some details on its set-up and on the physical models which have been used for the derivation of the processed data. We also give short introduction how to use DABA and SAS. (orig.)

  11. Experimental study of the principles governing tokamak transport

    Science.gov (United States)

    Wagner, F.; Gruber, O.; Lackner, K.; Murmann, H. D.; Speth, E.; Becker, G.; Bosch, H. S.; Brocken, H.; Cattanei, G.; Dorst, D.; Eberhagen, A.; Elsner, A.; Erckmann, V.; Fussmann, G.; Gehre, O.; Gernhardt, J.; Gierke, G. V.; Glock, E.; Grieger, G.; Grigull, P.; Haas, G.; Hacker, H.; Hartfuss, H. J.; Jäckel, H.; Jaenicke, R.; Janeschitz, G.; Junker, J.; Karger, F.; Kasparek, W.; Keilhacker, M.; Kick, M.; Klüber, O.; Kornherr, M.; Kroiss, H.; Kuehner, M.; Lenoci, M.; Lisitano, G.; Maassberg, M.; Mahn, C.; Marlier, S.; Mayer, H. M.; McCormick, K.; Meisel, D.; Mertens, V.; Müller, E. R.; Müller, .; Müller, G.; Niedermeyer, H.; Ohlendorf, W.; Poschenrieder, W.; Rapp, H.; Rau, F.; Renner, H.; Riedler, H.; Ringler, H.; Sardei, F.; Schüller, P. G.; Schwörer, K.; Siller, G.; Söldner, F.; Steuer, K.-H.; Thumm, M.; Tutter, M.; Vollmer, O.; Weller, A.; Wilhelm, R.; Wobig, H.; Würsching, E.; Zippe, M.

    1986-05-01

    Both in ohmically and beam-heated L-mode discharges of ASDEX, the electron-temperature (Te) profile shape can be varied over a wide range by the choice of the safety factor qa. The power-deposition profile, on the contrary, has no effect on the Te-profile shape. In current-free WVII-A stellarator plasmas, no such invariance property is found. An independent constraint seems to fix the current distribution j(r) of the tokamak, which defines the conditions of electron heat transport.

  12. Multi-channel Langmuir-probe and Hα-measurements of edge fluctuations on ASDEX

    International Nuclear Information System (INIS)

    The anomalous transport observed in tokamaks is caused by turbulent fluctuations, the nature of which is still poorly understood. Diagnostic difficulties are one major reason for this lack of understanding, at least with respect to the bulk plasma. The plasma edge, however, is accessible by several diagnostics permitting localized measurements of different parameters with good spatial and temporal resolution. For this reason one can hope to obtain enough information about edge fluctuations to permit the development of theoretical models. Different ranges of plasma parameters and the lack of closed magnetic surfaces distinguish this plasma zone from the bulk plasma. Edge turbulence might well involve other mechanisms than the turbulence in the bulk. Although transport in the bulk plasma receives more attention transport in the edge plasma and edge physics are very relevant for reactor design. The realistic hope to find a solution and the importance of the problem for the next step in fusion research are reasons for the strong effort in this field on many tokamaks. Like in many limiter tokamaks Langmuir probes were used in the ASDEX divertor device for measurements of the floating potential and of the ion saturation current. Under certain assumptions the electron density and the plasma potential can be derived from these data. Observation of the Hα-light emitted from the edge in the vicinity of a neutral gas source yields information about the electron density. While probe measurements are more suitable for quantitative evaluations including the calculation of the local particle flux the Hα-method is not calibrated and integrates radially over the edge. It is however applicable in situations where probes fail because of excessive heat load. With 16-channel arrays both methods permit spatial correlations and wavenumber spectra to be determined without any further assumptions. (author) 4 refs., 2 figs

  13. Characterization of type-I ELM induced filaments in the far scrape-off layer of ASDEX upgrade

    International Nuclear Information System (INIS)

    This thesis focuses on the characterization of filaments and their propagation in the ASDEX Upgrade tokamak. The aim is to provide experimental measurements for understanding the filament formation process and their temporal evolution, and to provide a comprehensive database for an extrapolation to future fusion devices. For this purpose, a new magnetically driven probe for filament measurements has been developed and installed in ASDEX Upgrade. The probe carries several Langmuir probes and a magnetic coil in between. The Langmuir probes allow for measurements of the radial and poloidal/toroidal propagation of filaments as well as for measurements of filament size, density, and their radial (or temporal) evolution. The magnetic coil on the filament probe allows for measurements of currents in the filaments. A set of 7 coils, measuring 3 field components at different positions along the filament, has been used to measure the magnetic signature during an ELM. The aim was, on the one hand, to study which role filaments play for the magnetic structure, and on the other hand if the parallel currents predicted by the sheath damped model could be verified. Filament temperatures have been derived and the corresponding heat transport mechanisms have been studied. (orig.)

  14. Characterization of type-I ELM induced filaments in the far scrape-off layer of ASDEX upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Schmid, Andreas

    2008-03-18

    This thesis focuses on the characterization of filaments and their propagation in the ASDEX Upgrade tokamak. The aim is to provide experimental measurements for understanding the filament formation process and their temporal evolution, and to provide a comprehensive database for an extrapolation to future fusion devices. For this purpose, a new magnetically driven probe for filament measurements has been developed and installed in ASDEX Upgrade. The probe carries several Langmuir probes and a magnetic coil in between. The Langmuir probes allow for measurements of the radial and poloidal/toroidal propagation of filaments as well as for measurements of filament size, density, and their radial (or temporal) evolution. The magnetic coil on the filament probe allows for measurements of currents in the filaments. A set of 7 coils, measuring 3 field components at different positions along the filament, has been used to measure the magnetic signature during an ELM. The aim was, on the one hand, to study which role filaments play for the magnetic structure, and on the other hand if the parallel currents predicted by the sheath damped model could be verified. Filament temperatures have been derived and the corresponding heat transport mechanisms have been studied. (orig.)

  15. Experimental investigation of heat transport and divertor loads of fusion plasmas in all metal ASDEX upgrade and JET

    International Nuclear Information System (INIS)

    This work presents divertor heat load studies conducted at two of the largest tokamaks currently in operation, ASDEX Upgrade and the Joint European Torus (JET). A commonly agreed empirical scaling for the power fall-off length in H-mode obtained in carbon devices is validated in JET with the ILW. Bohm and Gyro-Bohm like models are identified as possible candidates describing the divertor broadening. Quantities for the assessment of the thermal load induced by transient heat loads are defined. JET with the ILW exhibits an on average longer ELM duration as compared to the carbon wall. For identical pedestal conditions the ELM durations in both cases are found to be the same within error bars. The energy fluency is found to depend mainly on the pedestal pressure with a weak dependence on the relative loss in stored energy. This is noteworthy since the current extrapolation to ITER assumes a linear dependence on the relative ELM size.

  16. Collisionless microtearing modes in standard tokamak configurations

    International Nuclear Information System (INIS)

    Microtearing Modes (MTM) are electromagnetic microinstabilities occurring in magnetically confined fusion plasmas driven by parallel electron current and collisions in the presence of electron temperature gradient. MTMs were first predicted to occur in such plasmas in early 70s. Collisional MTMs have recently gathered attention in Spherical Tokamak configurations and RFPs. Very recently collisional MTMs have been reported in configuration relevant to standard tokamak, namely ASDEX-U. Perhaps for the first time, we show the existence of MTMs in purely collisionless limit and in large aspect ratio tokamak configurations using fully gyrokinetic full radius linear calculations. The physics of both electron scale as well as minor radius scale are resolved in the studies. Results of the studies, such as the 2-D structure of the mode and the dependence of growth rates on plasma pressure, perpendicular (to B0) wavelength spectrum and the effect of Landau damping and magnetic drift resonance will be presented. A comparison with another electromagnetic mode, namely Kinetic Ballooning Mode, which is driven by ion temperature gradient will also be shown. (author)

  17. Pellet injection into ASDEX upgrade plasmas

    International Nuclear Information System (INIS)

    This work comprises results obtained using the new centrifuge injection system for the two first years of pellet injection experiments at Asdex Upgrade until the end of the 1995 experimental campaign. The main aim of the pellet injection investigation is to develop scenarios allowing for a more flexible plasma density control means of injection of cryogenic solid hydrogen pellets. Efforts have been made to develop scenarios allowing more flexible plasma density control by injecting cryogenic solid hydrogen pellets. While the injection of pellets during ohmic discharges was found to be most efficient and also improves the plasma performance, increasing the auxiliary heating power causes a detoriation of the pellet fuelling efficiency. A further strong reduction of the pellet fuelling efficiency by an additional process was observed for the more reactor-relevant conditions of shallow particle deposition during H-mode phases. With injection during type I ELMy H-mode phases, each pellet was found to trigger the release of an ELM and therefore cause particle losses mainly from the edge region. In the type I ELMy H-mode, only sufficient pellet penetration allowed noticeable, persistent particle deposition in the plasma by the pellets. Applying adequate pellet injection conditions and favourable scenarios using combined pellet/gas puff refuelling, significant density ramp-up to densities exceeding the empirical Greenwald limit by up to a factor of two was achieved even for strongly heated H-mode plasmas. (orig.)

  18. L-H transport barrier formation: Monte Carlo simulation of the sheared E x B flow dynamics in tokamaks

    International Nuclear Information System (INIS)

    Monte Carlo ion simulation and semi-analytical calculation of the neoclassical radial current balance in a tokamak plasma edge show no spontaneous bifurcation of radial electric field Er in contrast with earlier orbit loss models, but bifurcation and a solitary Er generation by electrode polarization are seen. Strong orbit loss shear is found for H-mode threshold conditions in ASDEX Upgrade, indicating that orbit loss is the major source of the shear for L-H transition. (orig.)

  19. Overview of the EUROfusion Medium Size Tokamak program

    Science.gov (United States)

    Martin, Piero; Beurskens, Marc; Coda, Stefano; Eich, Thomas; Meyer, Hendrik; the EUROfusion MST1 Team

    2015-11-01

    As a result of the new organization of the European fusion programme, now under the umbrella of the EUROfusion Consortium, the MST (Medium Size Tokamaks) task force is in charge of executing the European science programme in the ASDEX Upgrade, TCV and MAST-U tokamaks. This paper will present an overview of the main results obtained in the 2014 campaign-where only ASDEX upgrade was operating-and the preliminary achievements of the recently started 2015/16 campaign, where also TCV will contribute. The main subjects of the experimental campaigns are (i) the development of scenarios relevant for the ITER Q=10 goal, in an all metal wall device (ii) the understanding of ELM mitigation/suppression with pellets and resonant magnetic perturbations, and in particular the effect of density versus collisionality, (iii) the understanding and optimization of methods for disruption mitigation or avoidance and runaway electrons control and (iv) the exploration of ITER and DEMO relevant scenarios with high normalized separatrix power flux, Psep / R , (Psep is the power through the separatrix, R the major radius) and tolerable target heat loads. The overview of the future programs in MST will be given. http://www.euro-fusionscipub.org/mst1

  20. Deuterium depth profile quantification in a ASDEX Upgrade divertor tile using secondary ion mass spectrometry

    Science.gov (United States)

    Ghezzi, F.; Caniello, R.; Giubertoni, D.; Bersani, M.; Hakola, A.; Mayer, M.; Rohde, V.; Anderle, M.

    2014-10-01

    We present the results of a study where secondary ion mass spectrometry (SIMS) has been used to obtain depth profiles of deuterium concentration on plasma facing components of the first wall of the ASDEX Upgrade tokamak. The method uses primary and secondary standards to quantify the amount of deuterium retained. Samples of bulk graphite coated with tungsten or tantalum-doped tungsten are independently profiled with three different SIMS instruments. Their deuterium concentration profiles are compared showing good agreement. In order to assess the validity of the method, the integrated deuterium concentrations in the coatings given by one of the SIMS devices is compared with nuclear reaction analysis (NRA) data. Although in the case of tungsten the agreement between NRA and SIMS is satisfactory, for tantalum-doped tungsten samples the discrepancy is significant because of matrix effect induced by tantalum and differently eroded surface (W + Ta always exposed to plasma, W largely shadowed). A further comparison where the SIMS deuterium concentration is obtained by calibrating the measurements against NRA values is also presented. For the tungsten samples, where no Ta induced matrix effects are present, the two methods are almost equivalent.The results presented show the potential of the method provided that the standards used for the calibration reproduce faithfully the matrix nature of the samples.

  1. Impurity production and plasma performance in ASDEX discharges with ohmic and auxiliary heating

    Science.gov (United States)

    Fussmann, G.; ASDEX Team; NI Team; Icrh Team; Hofmann, J.; Janeschitz, G.; Lenoci, M.; Mast, F.; McCormick, K.; Murmann, H.; Poschenrieder, W.; Roth, J.; Setzensack, C.; Staudenmaier, G.; Steuer, K.-H.; Taglauer, E.; Verbeek, H.; Wagner, F.; Becker, G.; Bosch, H. S.; Brocken, H.; Eberhagen, A.; Gehre, O.; Gernhardt, J.; Gierke, G. V.; Clock, E.; Gruber, O.; Haas, G.; Izvozchikov, A.; Karger, F.; Kaufmann, M.; Keilhacker, M.; Klüber, O.; Kornherr, M.; Lackner, K.; Lisitano, G.; Mayer, H. M.; Meisel, D.; Mertens, V.; Müller, E. R.; Neuhauser, J.; Niedermeyer, H.; Noterdaeme, J.-M.; Pietrzyk, Z. A.; Rapp, H.; Riedler, H.; Röhr, H.; Ryter, F.; Schneider, F.; Siller, G.; Smeulders, P.; Söldner, F. X.; Speth, E.; Steinmetz, K.; Tsois, N.; Ugniewski, S.; Vollmer, O.; Wesner, F.; Zasche, D.

    1987-02-01

    A review is given on investigations in the ASDEX Tokamak on impurities in ohmically, NI and ICRH heated plasmas. For ohmic discharges in H 2 and D 2 it is found that iron release from the wall can be explained by sputtering due to neutral charge exchange (CX) atoms. In the case of He, however, significant contributions caused by ion sputtering are inferred. Comparing discharges with C limiters in He and D 2 suggests that in the case of hydrogen chemical processes are involved in C sputtering. By means of wall carbonization the concentrations of metal ions in the plasma could be substantially reduced. This achievement is of particular importance for NI counter-injection and ICRH, where under non-carbonized conditions severe impurity problems occur. We studied impurity confinement in the case of various heating scenarios by means of the laser injection technique. The poorest confinement is found for the L-phase of NI. Metal injection into the high confinement H-phase generally causes temporary suppression of the edge localized modes (ELM's). With respect to ICRH we conclude that enhanced wall erosion — probably due to the production of high energy ions in the boundary — together with a slightly increased impurity confinement is the dominant reason for the increase of the metallic concentrations. Impurity sputtering as an alternative strong erosion process was experimentally ruled out.

  2. Development of tungsten coated first wall and high heat flux components for application in ASDEX Upgrade

    International Nuclear Information System (INIS)

    In the tokamak experiment ASDEX Upgrade, the investigation of tungsten as a first wall material is an ongoing research project. In a step-by-step strategy, the tungsten covered surface area is increased from campaign to campaign. For this purpose an industrial-scale method for coating graphite with micrometer tungsten films had to be identified. Test coatings deposited by magnetron sputtering and by plasma-arc deposition were compared. By X-ray analysis it was found that sputter-deposited coatings suffer from high compressive stress (1.7 GPa). This leads to delamination when a film thickness of about 3 μm is exceeded. For arc-deposited coatings, a compressive stress value of 0.5 GPa was determined and no delamination occurred up to the maximum film thicknesses investigated, i.e. 10 μm. Upon thermal loading, none of the arc-deposited coatings failed up to the melting condition, while one sputter-coating delaminated. First results on similar investigations employing CFC substrates are presented

  3. Estimation of sheath potentials in front of ASDEX upgrade ICRF antenna with SSWICH asymptotic code

    Energy Technology Data Exchange (ETDEWEB)

    Křivská, A., E-mail: alena.krivska@rma.ac.be [LPP-ERM/KMS, Royal Military Academy, 30 Avenue de la Renaissance B-1000, Brussels (Belgium); Bobkov, V.; Jacquot, J.; Ochoukov, R. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Colas, L. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Milanesio, D. [Politecnico di Torino, Corso Duca degli Abruzzi 24, I-10129 Torino (Italy)

    2015-12-10

    Multi-megawatt Ion Cyclotron Range of Frequencies (ICRF) heating became problematic in ASDEX Upgrade (AUG) tokamak after coating of ICRF antenna limiters and other plasma facing components by tungsten. Strong impurity influx was indeed produced at levels of injected power markedly lower than in the previous experiments. It is assumed that the impurity production is mainly driven by parallel component of Radio-Frequency (RF) antenna electric near-field E// that is rectified in sheaths. In this contribution we estimate poloidal distribution of sheath Direct Current (DC) potential in front of the ICRF antenna and simulate its relative variations over the parametric scans performed during experiments, trying to reproduce some of the experimental observations. In addition, relative comparison between two types of AUG ICRF antenna configurations, used for experiments in 2014, has been performed. For this purpose we use the Torino Polytechnic Ion Cyclotron Antenna (TOPICA) code and asymptotic version of the Self-consistent Sheaths and Waves for Ion Cyclotron Heating (SSWICH) code. Further, we investigate correlation between amplitudes of the calculated oscillating sheath voltages and the E// fields computed at the lateral side of the antenna box, in relation with a heuristic antenna design strategy at IPP Garching to mitigate RF sheaths.

  4. Local effects of ECRH on argon transport at ASDEX upgrade

    International Nuclear Information System (INIS)

    Future deuterium-tritium magnetically confined fusion power plants will most probably rely an high-Z Plasma Facing Components (PFCs) such as tungsten. This choice is determined by the necessity of low erosion of the first wall materials (to guarantee a long lifetime of the wall components) and by the need to avoid the too high tritium wall retention of typical carbon based PFCs. The experience gathered at the ASDEX Upgrade (AUG) tokamak has demonstrated the possibility of reliable and high performance plasma operation with a full tungsten-coated first wall. The observed accumulation of tungsten which can lead to excessive radiation losses is mitigated with the use of Electron Cyclotron Resonance Heating (ECRH). Although this impurity control method is routinely performed at AUG, the underlying physics principles are still not clear. This thesis aims an providing further knowledge an the effects of ECRH an the transport of impurities inside the core plasma. The transport of argon has been therefore investigated in-depth in purely ECR heated L-mode (low-confinement) discharges. Studies an impurity transport in centrally ECR heated nitrogen-seeded H-mode (high-confinement) discharges have also been performed. To this scope, a new crystal X-ray spectrometer of the Johann type has been installed an AUG for argon concentration and ion temperature measurements. New methods for the experimental determination of the total argon density through the integrated use of this diagnostic and of the Soft X-Ray (SXR) diode arrays have been developed. This gives the possibility of evaluating the full profiles of the argon transport coefficients from the linear flux-gradient dependency of local argon density. In comparison to classical χ2-minimization methods, the approach proposed here delivers transport coefficients intrinsically independent of the modelling of periodic relaxation mechanisms such as those Lied to sawtooth MHD (Magneto-Hydro-Dynamic) activity. Moreover, the good

  5. Ignition in near term D-3He tokamak reactors: Appendix B

    International Nuclear Information System (INIS)

    The prospects for achieving breakeven and ignition in near term ETR type tokamaks under D/He-3 relevant conditions are considered. Using present scaling laws for beta in the first stability regime, it is found that CIT may be close to breakeven with the presently planned toroidal magnetic field system, if the ASDEX H-mode scaling law is used. With Kaye-Goldston scaling, Q = .22 can be attained, but this requires an excessive amount of rf heating power. Larger devices, such as NET/INTOR, can ignite with ASDEX H-mode scaling with an increase of the toroidal field by 20% and removal of the blanket and reduction of the inboard shield to that required for D/He-3. 5 refs., 4 figs., 2 tabs

  6. A review of ELMs in divertor tokamaks

    International Nuclear Information System (INIS)

    This paper reviews what is known about edge localized modes (ELMs), with an emphasis on their effect on the scrape-off layer and divertor plasmas. ELM effects have been measured in the ASDEX-U, C-Mod, COMPASS-D, DIII-D, JET, JFT-2M,JT-60U, and TCV tokamaks and are reported here. At least three types of ELMs have been identified and their salient features determined. Type-1 giant ELMs can cause the sudden loss of up to 10-15% of the plasma stored energy but their amplitude (ΔW/W) does not increase with increasing power. Type- 3 ELMs are observed near the H-mode power threshold and produce small energy dumps (1-3% of the stored energy). All ELMs increase the scrape- off layer plasma and produce particle fluxes on the divertor targets which are as much as ten times larger that the quiescent phase between ELMs. The divertor heat pulse is largest on the inner target, unlike that of L-Mode or quiescent H-mode; some tokamaks report radial structure in the heat flux profile which is suggestive of islands or helical structures. The power scaling of Type-1 ELM amplitude and frequency have been measured in several tokamaks and has recently been applied to predictions of the ELM Size in ITER. Concern over the expected ELM amplitude has led to a number of experiments aimed at demonstrating active control of ELMs. Impurity gas injection with feedback control on the radiation loss in ASDEX-U suggests that a promising mode of operation (the CDH-mode) with a very small type-3 ELMs can be maintained with heating power sell above the H-mode threshold, where giant type-1 ELMs can be maintained with heating power well above the H-mode threshold, where Giant type-1 ELMs are normally observed. While ELMs have many potential negative effects, the beneficial effect of ELMs in providing density control and limiting the core plasma impurity content in high confinement H- mode discharges should not be overlooked

  7. Ion temperature in SOC and IOC discharges in ASDEX

    International Nuclear Information System (INIS)

    Active and passive charge exchange measurements were made to investigate the behaviour of the central ion temperature and the temperature profile for SOC and IOC discharges in ASDEX. Both methods show an increase in the central ion temperature during transition from SOC to IOC. Both methods also show a wider temperature profile for ions than for electrons. Peaking of the ion temperature profile during IOC cannot be definitely concluded from the measurements. (author) 7 refs., 4 figs

  8. Control Processes and Machine Protection on ASDEX Upgrade

    International Nuclear Information System (INIS)

    ASDEX Upgrade's new real-time discharge control system connects to all coils, heating and fuelling actuators, to protection systems, and to a large number of real-time diagnostics. It evaluates and monitors many physics quantities and provides an overall plant and plasma state, to execute coordinated feedback processes, and operate actuators with advanced strategies in high performance plasma scenarios. The control system can react to specific plasma and plant states with new control goals, expressed through alternate sets of discharge schedule references, or through new references computed in real-time. The method provides extreme flexibility to respond with context-specific control strategies to optimize plasma performance, reduce the criticality of a discharge or minimize machine stress. This ability gives the control system a key role in ASDEX Upgrade's three-layer protection hierarchy, where the most flexible protection level is fully computer-based. We will sketch the control process model's features to support safety critical applications, such as self-monitoring, automatic alarm propagation and watchdog monitoring, give an overview of the plant and plasma monitor processes, show how these are mapped to the protection layers of ASDEX Upgrade, and explain the control system's activities within these. (author)

  9. Feedback-controlled NTM stabilization on ASDEX Upgrade

    Directory of Open Access Journals (Sweden)

    Stober J.

    2015-01-01

    Full Text Available On ASDEX Upgrade a concept for real-time stabilization of NTMs has been realized and successfully applied to (3,2- and (2,1-NTMs. Since most of the work has meanwhile been published elsewhere, a short summary with the appropriate references is given. Limitations, deficits and future extensions of the system are discussed. In a second part the recent work on using modulated ECCD for NTM stabilisation is described in some detail. In these experiments ECCD power is modulated according to a magnetic footprint of the rotating NTM. In agreement with earlier results it could be shown that O-point heating reduces the necessary average power for stabilisation whereas X-point heating hampers stabilisation. Although this modulated scheme is not relevant for routine NTM stabilisation on ASDEX Upgrade it may be mandatory for ITER or DEMO. On ASDEX Upgrade it has been re-developed to demonstrate the usage of a FAst DIrectional Switch to continously heat the O-point of the rotating island with only one gyrotron switching between two launchers which target the mode at locations separated in phase by 180 degrees as described in [1].

  10. Varennes Tokamak

    International Nuclear Information System (INIS)

    A consortium of five organizations under the leadership of IREQ, the Institute de Recherche d'Hydro-Quebec has completed a conceptual design study for a tokamak device, and in January 1981 its construction was authorized with funding being provided principally by Hydro-Quebec and the National Research Council, as well as by the Ministre d'Education du Quebec and Natural Sciences and Engineering Research Council of Canada (NSERC). The device will form the focus of Canada's magnetic-fusion program and will be located in IREQ's laboratories in Varennes. Presently the machine layout is being finalized from the physics point of view and work has started on equipment design and specification. The Tokamak de Varennes will be an experimental device, the purpose of which is to study plasma and other fusion related phenomena. In particular it will study: 1. Plasma impurities and plasma/liner interaction; 2. Long pulse or quasi-continuous operation using plasma rampdown and eventually plasma current reversal in order to maintain the plasma; and 3. Advanced diagnostics

  11. Role of runaway electrons in LHCD regimes with improved confinement on the CASTOR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Voitsekhovich, I. [Kurchatov Institute, Moscow (Russian Federation); Stoeckel, J.; Zacek, F. [Akademie Ved Ceske Republiky, Prague (Czech Republic). Ustav Fyziky Plazmatu

    1993-12-31

    Lower hybrid current drive (LHCD) experiments in low density plasmas on ASDEX, CASTOR, WT-3, VERSATOR and HT-6B tokamaks demonstrated an improvement of the particle confinement at moderate lower hybrid powers (P{sub LH}). Moreover, the experiments have shown that a reduction of edge electrostatic fluctuations is probably responsible for this effect. However, the mechanism behind the reduction of fluctuations has remained unclear. Here we try to explain the reduction of fluctuations by enhanced population and non-ambipolar losses of runaway electrons with LHCD. (author) 8 refs., 3 figs.

  12. M3D-C1 simulations of plasma response in ELM-mitigated ASDEX Upgrade and DIII-D discharges

    Science.gov (United States)

    Lyons, B. C.; Ferraro, N. M.; Haskey, S. R.; Logan, N. C.

    2015-11-01

    The extended magnetohydrodynamics (MHD) code M3D-C1 is used to study the time-independent, linear response of tokamak equilibria to applied, 3D magnetic perturbations. In doing so, we seek to develop a more complete understanding of what MHD phenomena are responsible for the mitigation and suppression of edge-localized modes (ELMs) and to explain why the success of ELM suppression experiments differs both within a single tokamak and across different tokamaks. We consider such experiments on ASDEX Upgrade and DIII-D. We examine how resonant and non-resonant plasma responses are affected by varying the relative magnitude and phase of sets of magnetic coils. The importance of two-fluid effects, rotation profiles, plasma β, collisionality, bootstrap current profiles, and various numerical parameters are explored. The results are verified against other MHD codes (e.g., IPEC, MARS), correlated to observations of ELM mitigation or suppression, and validated against observed magnetic responses. Work supported in part by US DOE under DE-FC02-04ER54698, DE-AC02-09CH11466, and the FES Postdoctoral Research Program.

  13. A new diagnostic for ASDEX upgrade edge ion temperatures by lithium-beam charge exchange recombination spectroscopy

    International Nuclear Information System (INIS)

    This thesis work investigates the measurement of ion temperatures at the edge of a magnetically confined plasma used for fusion research at the ASDEX Upgrade tokamak operated by Max-Planck-Institut fuer Plasmaphysik in Garching. The H-mode plasma regime, default scenario of the next step experiment ITER, is characterized by an edge transport barrier, which is not yet fully explained by theory. Experimentally measured edge ion temperature profiles will help to test and develop models for these barriers. Transport theory on a basic level is introduced as background and motivation for the new diagnostic. The standard model for an edge plasma instability named ''edge localized mode'' (ELM) observed in H-mode is described. The implementation of a new diagnostic for ion temperature measurements with high spatial resolution in the plasma edge region, its commissioning and the validation of the measurements comprises the main part of this work. The emission of line radiation induced by charge exchange processes between lithium atoms injected by a beam source and fully ionized impurities (of C and He) is observed with a detection system consisting of spectrometers and fast cameras. Due to the narrow beam (1 cm) and closely staggered optical fibers (6 mm), unprecedented spatial resolution of edge ion temperatures in all major plasma regimes of the ASDEX Upgrade tokamak was achieved. The spectral width of the line radiation (He II at 468.5 nm and C VI at 529.0 nm) contains information about the local ion temperature from thermal Doppler-broadening, which is the dominant broadening mechanism for these lines. The charge-exchange contribution to the total line radiation locally generated by the lithium is determined by gating the beam. Fitting a Gaussian model function to the local line radiation results in absolute line widths which can be directly converted into a temperature. The equilibration of impurities with the main plasma is fast enough that the assumption of nearly

  14. Deuterium depth profile quantification in a ASDEX Upgrade divertor tile using secondary ion mass spectrometry

    International Nuclear Information System (INIS)

    Highlights: • We measured absolute local concentration of D in W samples by three Secondary Ions Mass Spectrometry (SIMS) apparatus. • Implanted primary standard and special secondary standards were prepared to calibrate the measurements. • D concentrations integrated along the depth were compared with absolute NRA measurements. - Abstract: We present the results of a study where secondary ion mass spectrometry (SIMS) has been used to obtain depth profiles of deuterium concentration on plasma facing components of the first wall of the ASDEX Upgrade tokamak. The method uses primary and secondary standards to quantify the amount of deuterium retained. Samples of bulk graphite coated with tungsten or tantalum-doped tungsten are independently profiled with three different SIMS instruments. Their deuterium concentration profiles are compared showing good agreement. In order to assess the validity of the method, the integrated deuterium concentrations in the coatings given by one of the SIMS devices is compared with nuclear reaction analysis (NRA) data. Although in the case of tungsten the agreement between NRA and SIMS is satisfactory, for tantalum-doped tungsten samples the discrepancy is significant because of matrix effect induced by tantalum and differently eroded surface (W + Ta always exposed to plasma, W largely shadowed). A further comparison where the SIMS deuterium concentration is obtained by calibrating the measurements against NRA values is also presented. For the tungsten samples, where no Ta induced matrix effects are present, the two methods are almost equivalent.The results presented show the potential of the method provided that the standards used for the calibration reproduce faithfully the matrix nature of the samples

  15. H-mode pedestal scaling in DIII-D, ASDEX Upgrade, and JET

    International Nuclear Information System (INIS)

    Multidevice pedestal scaling experiments in the DIII-D, ASDEX Upgrade (AUG), and JET tokamaks are presented in order to test two plasma physics pedestal width models. The first model proposes a scaling of the pedestal width Δ/a ∝ ρ*1/2 to ρ* based on the radial extent of the pedestal being set by the point where the linear turbulence growth rate exceeds the ExB velocity. In the multidevice experiment where ρ* at the pedestal top was varied by a factor of four while other dimensionless parameters where kept fixed, it has been observed that the temperature pedestal width in real space coordinates scales with machine size, and that therefore the gyroradius scaling suggested by the model is not supported by the experiments. The density pedestal width is not invariant with ρ* which after comparison with a simple neutral fuelling model may be attributed to variations in the neutral fuelling patterns. The second model, EPED1, is based on kinetic ballooning modes setting the limit of the radial extent of the pedestal region and leads to Δψ ∝ βp1/2. All three devices show a scaling of the pedestal width in normalised poloidal flux as Δψ ∝ βp1/2, as described by the kinetic ballooning model; however, on JET and AUG, this could not be distinguished from an interpretation where the pedestal is fixed in real space. Pedestal data from all three devices have been compared with the predictive pedestal model EPED1 and the model produces pedestal height values that match the experimental data well.

  16. Deuterium depth profile quantification in a ASDEX Upgrade divertor tile using secondary ion mass spectrometry

    Energy Technology Data Exchange (ETDEWEB)

    Ghezzi, F., E-mail: ghezzi@ifp.cnr.it [Istituto di Fisica del Plasma “Piero Caldirola” IFP Euratom-ENEA-CNR Association, Via R. Cozzi 53, 20125 Milan (Italy); Caniello, R. [Istituto di Fisica del Plasma “Piero Caldirola” IFP Euratom-ENEA-CNR Association, Via R. Cozzi 53, 20125 Milan (Italy); Giubertoni, D.; Bersani, M. [FBK, Via Sommarive 18, 38123 Povo, TN (Italy); Hakola, A. [VTT, Association Euratom-Tekes, P.O. Box 1000, 02044 VTT (Finland); Mayer, M.; Rohde, V. [Max-Planck-Institut für Plasmaphysik, 85748 Garching (Germany); Anderle, M. [Knowledge Department, Autonomous Province of Trento, 38123 Trento (Italy)

    2014-10-01

    Highlights: • We measured absolute local concentration of D in W samples by three Secondary Ions Mass Spectrometry (SIMS) apparatus. • Implanted primary standard and special secondary standards were prepared to calibrate the measurements. • D concentrations integrated along the depth were compared with absolute NRA measurements. - Abstract: We present the results of a study where secondary ion mass spectrometry (SIMS) has been used to obtain depth profiles of deuterium concentration on plasma facing components of the first wall of the ASDEX Upgrade tokamak. The method uses primary and secondary standards to quantify the amount of deuterium retained. Samples of bulk graphite coated with tungsten or tantalum-doped tungsten are independently profiled with three different SIMS instruments. Their deuterium concentration profiles are compared showing good agreement. In order to assess the validity of the method, the integrated deuterium concentrations in the coatings given by one of the SIMS devices is compared with nuclear reaction analysis (NRA) data. Although in the case of tungsten the agreement between NRA and SIMS is satisfactory, for tantalum-doped tungsten samples the discrepancy is significant because of matrix effect induced by tantalum and differently eroded surface (W + Ta always exposed to plasma, W largely shadowed). A further comparison where the SIMS deuterium concentration is obtained by calibrating the measurements against NRA values is also presented. For the tungsten samples, where no Ta induced matrix effects are present, the two methods are almost equivalent.The results presented show the potential of the method provided that the standards used for the calibration reproduce faithfully the matrix nature of the samples.

  17. Influence of externally applied magnetic perturbations on neoclassical tearing modes at ASDEX Upgrade

    International Nuclear Information System (INIS)

    The influence of externally applied magnetic perturbations (MPs) on neoclassical tearing modes (NTM) and the plasma rotation in general is investigated at the ASDEX Upgrade tokamak (AUG). The low n resonant components of the applied field exert local torques and influence the stability of NTMs. The non-resonant components of the error field do not influence MHD modes directly but slow down the plasma rotation globally due to a neoclassical toroidal viscous torque (NTV). Both components slow down the plasma rotation, which in consequence increases the probability for the appearance of locked modes. To investigate the impact of externally applied MPs on already existing modes and the influence on the rotation profile, experimental observations are compared to modelling results. The model used here solves a coupled equation system that includes the Rutherford equation and the equation of motion, taking into account the resonant effects and the resistive wall. It is shown that the NTV torque can be neglected in this modelling. To match the experimental frequency evolution of the mode the MP field strength at the resonant surface has to be increased compared to the vacuum approximation. This leads to an overestimation of the stabilizing effect on the NTMs. The reconstruction of the entire rotation profile via the equation of motion including radial dependencies, confirms that the NTV is negligibly small and that small resonant torques at different resonant surfaces have the same effect as one large one. This modelling suggests that in the experiment resonant torques at different surfaces are acting and slowing down the plasma rotation requiring a smaller torque at the specific resonant surface of the NTM. This additionally removes the overestimated influence on the island stability, whereas the braking of the island's rotation is caused by the sum of all torques. Consequently, to describe the effect of MPs on the evolution of one island, all other islands and the

  18. Management of complex data flows in the ASDEX Upgrade plasma control system

    International Nuclear Information System (INIS)

    Highlights: ► Control system architectures with data-driven workflows are efficient, flexible and maintainable. ► Signal groups provide coherence of interrelated signals and increase the efficiency of process synchronisation. ► Sample tags indicating sample quality form the fundament of a local event handling strategy. ► A self-organising workflow benefits from sample tags consisting of time stamp and stream activity. - Abstract: Establishing adequate technical and physical boundary conditions for a sustained nuclear fusion reaction is a challenging task. Phased feedback control and monitoring for heating, fuelling and magnetic shaping is mandatory, especially for fusion devices aiming at high performance plasmas. Technical and physical interrelations require close collaboration of many components in sequential as well as in parallel processing flows. Moreover, handling of asynchronous, off-normal events has become a key element of modern plasma performance optimisation and machine protection recipes. The manifoldness of plasma states and events, the variety of plant system operation states and the diversity in diagnostic data sampling rates can hardly be mastered with a rigid control scheme. Rather, an adaptive system topology in combination with sophisticated synchronisation and process scheduling mechanisms is suited for such an environment. Moreover, the system is subject to real-time control constraints: response times must be deterministic and adequately short. Therefore, the experimental tokamak device ASDEX Upgrade employs a discharge control system DCS, whose core has been designed to meet these requirements. In the paper we will compare the scheduling schemes for the parallelised realisation of a control workflow and show the advantage of a data-driven workflow over a managed workflow. The data-driven workflow as used in DCS is based on signals connecting process outputs and inputs. These are implemented as real-time streams of data samples

  19. Recent ASDEX Upgrade research in support of ITER and DEMO

    DEFF Research Database (Denmark)

    Zohm, H.; Ahn, J.; Aho-Mantila, L.;

    2015-01-01

    decisive element for the L–H power threshold. A physics based scaling of the density at which the minimum PLH occurs indicates that ITER could take advantage of it to initiate H-mode at lower density than that of the final Q = 10 operational point. Core density fluctuation measurements resolved in radius...... in the all-tungsten (all-W) ASDEX Upgrade due to the observed poor confinement at low βN. This is mainly due to a degraded pedestal performance and hence investigations at shifting the operational point to higher βN by lowering the current have been started. At higher q95, pedestal performance can be...

  20. Recent results of reflectometry on ASDEX-upgrade

    International Nuclear Information System (INIS)

    Reflectometry is well known to be very sensitive to plasma density fluctuations. The study of plasma response in broadband frequency operation is concentrated on the obtention of the main peak and many techniques have been developed to filter the unwanted components. In comparison little work has been done to understand the remaining part of the signal. This paper presents some recent results about plasma fluctuations obtained with FM-reflectometry on ASDEX-Upgrade. They demonstrate the rich content information of both the fixed frequency and broadband signals and suggest that they can be used in a complementary way. (A.L.B.)

  1. 2D electron cyclotron emission imaging at ASDEX Upgrade (invited)

    Energy Technology Data Exchange (ETDEWEB)

    Classen, I. G. J. [Max Planck Institut fuer Plasmaphysik, 85748 Garching (Germany); FOM-Institute for Plasma Physics, Rijnhuizen, 3430 BE Nieuwegein (Netherlands); Boom, J. E.; Vries, P. C. de [FOM-Institute for Plasma Physics, Rijnhuizen, 3430 BE Nieuwegein (Netherlands); Suttrop, W.; Schmid, E.; Garcia-Munoz, M.; Schneider, P. A. [Max Planck Institut fuer Plasmaphysik, 85748 Garching (Germany); Tobias, B.; Domier, C. W.; Luhmann, N. C. Jr. [University of California at Davis, Davis, California 95616 (United States); Donne, A. J. H. [FOM-Institute for Plasma Physics, Rijnhuizen, 3430 BE Nieuwegein (Netherlands); Eindhoven University of Technology, 5600 MB Eindhoven (Netherlands); Jaspers, R. J. E. [Eindhoven University of Technology, 5600 MB Eindhoven (Netherlands); Park, H. K. [POSTECH, Pohang, Gyeongbuk, 790-784 (Korea, Republic of); Munsat, T. [University of Colorado, Boulder, Colorado 80309 (United States)

    2010-10-15

    The newly installed electron cyclotron emission imaging diagnostic on ASDEX Upgrade provides measurements of the 2D electron temperature dynamics with high spatial and temporal resolution. An overview of the technical and experimental properties of the system is presented. These properties are illustrated by the measurements of the edge localized mode and the reversed shear Alfven eigenmode, showing both the advantage of having a two-dimensional (2D) measurement, as well as some of the limitations of electron cyclotron emission measurements. Furthermore, the application of singular value decomposition as a powerful tool for analyzing and filtering 2D data is presented.

  2. Neutron calibration techniques for comparison of tokamak results

    International Nuclear Information System (INIS)

    A workshop on 1--3 August 1989 reviewed the techniques, uncertainties, and experiences of neutron calibration on PLT, TFTR, JET, Tore Supra, JT-60, JIPPT-IIU, Alcator C-Mod, ATF, FT, ASDEX, Textor, and DIII-D. In the summary session, the workshop participants discussed possible consensus neutron calibration techniques appropriate to D-D plasmas in tokamaks. The application of such techniques would facilitate a more accurate comparison of neutron yields from different devices, and also allow new calibration techniques to relate their precision to a reference value. General agreement was reached on the suitability of two techniques: (1) a 252Cf source calibration of epithermal neutron detectors, and (2) threshold neutron activation of Ni foils placed vertically above or below the plasma. This paper will present details on detector positioning, neutron transport calculations, and interlab normalization needed to accomplish the standardized calibration using a Cf neutron source

  3. GPEC, a real-time capable Tokamak equilibrium code

    CERN Document Server

    Rampp, Markus; Fischer, Rainer

    2015-01-01

    A new parallel equilibrium reconstruction code for tokamak plasmas is presented. GPEC allows to compute equilibrium flux distributions sufficiently accurate to derive parameters for plasma control within 1 ms of runtime which enables real-time applications at the ASDEX Upgrade experiment (AUG) and other machines with a control cycle of at least this size. The underlying algorithms are based on the well-established offline-analysis code CLISTE, following the classical concept of iteratively solving the Grad-Shafranov equation and feeding in diagnostic signals from the experiment. The new code adopts a hybrid parallelization scheme for computing the equilibrium flux distribution and extends the fast, shared-memory-parallel Poisson solver which we have described previously by a distributed computation of the individual Poisson problems corresponding to different basis functions. The code is based entirely on open-source software components and runs on standard server hardware and software environments. The real-...

  4. Termoska pro tokamak

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan

    2014-01-01

    Roč. 7, prosinec (2014), s. 16-17 Institutional support: RVO:61389021 Keywords : fusion * tokamak * cryostat * ITER Subject RIV: BL - Plasma and Gas Discharge Physics http://3pol.cz/1604-termoska-pro-tokamak

  5. PPPL tokamak program

    International Nuclear Information System (INIS)

    The economic prospects of the tokamak are reviewed briefly and found to be favorable - if the size of ignited tokamak plasmas can be kept small and appropriate auxiliary systems can be developed. The main objectives of the Princeton Plasma Physics Laboratory tokamak program are: (1) exploration of the physics of high-temperature toroidal confinement, in TFTR; (2) maximization of the tokamak beta value, in PBX; (3) development of reactor-relevant rf techniques, in PLT

  6. Edge density measurements with a fast Li-beam probe on tokamak and stellarator experiments

    International Nuclear Information System (INIS)

    High-energy neutral Li-beam probes have advanced to the point where they are a standard diagnostic on W7-AS and ASDEX-Upgrade, both in terms of the Li-beam injector and the reconstruction algorithm to arrive at densities along the beam ne(z) from the Li[2p-2s] resonance line profile. With beam energies in the range 30-70 keV and neutral equivalent currents of >1mA, it is possible to produce ne(z) profiles for line densities nez14 cm-2 with a radial resolution of ∝0.5 cm and time response ≤0.2 msec. The IPP Li-gun is described. By way of example, the diagnostic layout on W7-AS is sketched and salient results from experiment presented which serve to explore diagnostic limits and to underline the viability of the technique. Densities over the range 12-1014 cm-3 are accessible, permitting full coverage of the core density gradient region on W7-AS. Examples from ASDEX involving the H-mode and pellet injection are presented to exemplify time response. Scaling of SOL density e-folding lengths are introduced to point out possible differences between tokamak (ASDEX, ASDEX-Up) and stellarator (W7-AS) transport behavior perpendicular to field lines. A neutral lithium beam can also be employed to measure (a) impurity ion temperatures and densities via CXRS, and (b) neutral pressure outside the plasma column. These aspects lies outside the scope of the present paper. (orig.)

  7. Status of tokamak research

    International Nuclear Information System (INIS)

    An overall review of the tokamak program is given with particular emphasis upon developments over the past five years in the theoretical and experimental elements of the program. A summary of the key operating parameters for the principal tokamaks throughout the world is given. Also discussed are key issues in plasma confinement, plasma heating, and tokamak design

  8. Control processes and machine protection on ASDEX Upgrade

    International Nuclear Information System (INIS)

    Safe operation of ASDEX Upgrade is guaranteed by a conventional hierarchy of simple and robust hard-wired systems for personnel and machine protection featuring standardized switch-off procedures. Machine protection and handling of off-normal events is further enhanced and peak and lifetime stress minimized through the plasma control system. Based on a real-time process model supporting safety critical applications with data quality tagging, process self-monitoring, watchdog monitoring and alarm propagation, processes detect complex and critical failures and reliably perform case-sensitive counter measures. Intelligent real-time failure handling is done with hardware or software redundancy and performance degradation, or modification of reference values to continue or terminate discharges with reduced machine stress. Examples implemented so far on ASDEX Upgrade are given, such as recovery from measurement failures, switch-over of redundant actuators, handling of actuator limitations, detection of plasma instabilities, plasma state dependent soft landing, or handling of failed switch-off procedures through breakers disconnecting the machine from grid

  9. Tolerable ELMs in conventional and advanced scenarios at ASDEX upgrade

    International Nuclear Information System (INIS)

    Recent ASDEX Upgrade experiments have integrated benign type II ELMs exhibiting tolerable peak heat loads on the target plates with high plasma performance in terms of confinement, beta, and density. In both conventional and advanced H-modes, the operation window with type II ELMs was extended down to q95 ∼ 3.5 in close to double null configurations at sufficiently high edge pedestal density above 50% of the Greenwald density. Type I ELMs are suppressed at almost constant pedestal parameters and confinement levels, presumably due to a change in edge stability provided by higher edge magnetic shear. Since conventional reactor designs are optimised at q95 ∼ 3, operation with type II ELMs has to compensate the required higher q-value by enhanced performance. This was achieved in H-mode scenarios with high βN > 3.5, improved confinement (H98-P ∼ 1.3), and averaged densities of 90% of Greenwald density allowing for type II ELMs. In this integrated scenario with a high triangular plasma shape the confinement is improved via peaked density profiles. Active triggering and mitigation of type I ELMs by means of hydrogen pellet injection was demonstrated with enhancement of the ELM frequency to the pellet rate of 20 Hz. The enhanced Dα (EDA) mode was not reproducible at ASDEX Upgrade. (author)

  10. COMPARISON BETWEEN 2D TURBULENCE MODEL ESEL AND EXPERIMENTAL DATA FROM AUG AND COMPASS TOKAMAKS

    Directory of Open Access Journals (Sweden)

    Peter Ondac

    2015-04-01

    Full Text Available In this article we have used the 2D fluid turbulence numerical model, ESEL, to simulate turbulent transport in edge tokamak plasma. Basic plasma parameters from the ASDEX Upgrade and COMPASS tokamaks are used as input for the model, and the output is compared with experimental observations obtained by reciprocating probe measurements from the two machines. Agreements were found in radial profiles of mean plasma potential and temperature, and in a level of density fluctuations. Disagreements, however, were found in the level of plasma potential and temperature fluctuations. This implicates a need for an extension of the ESEL model from 2D to 3D to fully resolve the parallel dynamics, and the coupling from the plasma to the sheath.

  11. Tokamak operation with high-Z plasma facing components

    International Nuclear Information System (INIS)

    Due to wall lifetime requirements and the problem of tritium co-deposition in hydrocarbon layers, a future burning plasma will most probably have a full high-Z wall. The prime candidate material is tungsten, which exhibits good thermo-mechanical properties and has a high energy threshold for physical sputtering. To investigate the reactor-compatibility of this wall material, ASDEX Upgrade is being converted into a full-tungsten coated tokamak in a step-by-step approach, with presently almost 70 % of W wall coverage. The effect of the reduction of primary carbon coverage on the plasma is so far moderate. Under tokamak conditions, carbon behaves like a recycling impurity, due to the deposition and re-erosion of soft hydrocarbon layers on the tungsten surface. During high density H-mode operation, the central tungsten concentrations remain typically low, i.e. well under 10-5. The situation is more critical in the improved H-mode or hybrid scenario. Here, the combination of hot edge conditions and peaked central density profiles result in high central tungsten concentrations of up to 10-4, which would be critical in a reactor. However, core electron density peaking is reduced by use of central ICR or ECR heating and thus in turn suppresses central tungsten accumulation. For extrapolation to reactor conditions, we need to separate the effects of the tungsten wall source, the penetration over the edge transport barrier (ETB) and the core transport with its strong neoclassical contribution. These issues are addressed by inspecting the tungsten behaviour in various discharge scenarios and parameters in ASDEX Upgrade. These include radiative cooling by medium-Z seed impurities and ELM frequency control by pellet injection to simulate a reactor plasma with small edge and divertor impurity radiation levels and a separatrix power flux close to the H-L threshold. Fast ions produced by NBI and ICR heating at the low field side appear to be an important tungsten erosion mechanism

  12. Quantification of the impact of large and small-scale instabilities on the fast-ion confinement in ASDEX Upgrade

    Science.gov (United States)

    Geiger, B.; Weiland, M.; Mlynek, A.; Reich, M.; Bock, A.; Dunne, M.; Dux, R.; Fable, E.; Fischer, R.; Garcia-Munoz, M.; Hobirk, J.; Hopf, C.; Nielsen, S.; Odstrcil, T.; Rapson, C.; Rittich, D.; Ryter, F.; Salewski, M.; Schneider, P. A.; Tardini, G.; Willensdorfer, M.

    2015-01-01

    The confinement fast ions, generated by neutral beam injection (NBI), has been investigated at the ASDEX Upgrade tokamak. In plasmas that exhibit strong sawtooth crashes, a significant sawtooth-induced internal redistribution of mainly passing fast ions is observed, which is in very good agreement with the theoretical predictions based on the Kadomtsev model. Between the sawtooth crashes, the fishbone modes are excited which, however, do not cause measurable changes in the global fast-ion population. During experiments with on- and off-axis NBI and without strong magnetohydrodynamic (MHD) modes, the fast-ion measurements agree very well with the neo-classical predictions. This shows that the MHD-induced (large-scale), as well as a possible turbulence-induced (small-scale) fast-ion transport is negligible under these conditions. However, in discharges performed to study the off-axis NBI current drive efficiency with up to 10 MW of heating power, the fast-ion measurements agree best with the theoretical predictions that assume a weak level anomalous fast-ion transport. This is also in agreement with measurements of the internal inductance, a Motional Stark Effect diagnostic and a novel polarimetry diagnostic: the fast-ion driven current profile is clearly modified when changing the NBI injection geometry and the measurements agree best with the predictions that assume weak anomalous fast-ion diffusion.

  13. Tokamak Systems Code

    International Nuclear Information System (INIS)

    The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged

  14. Characterization of dust particles produced in an all-tungsten wall tokamak and potentially mobilized by airflow

    Energy Technology Data Exchange (ETDEWEB)

    Rondeau, A., E-mail: anthony.rondeau@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES, SCA, 91192 Gif-sur-Yvette (France); Peillon, S.; Roynette, A.; Sabroux, J.-C.; Gelain, T.; Gensdarmes, F. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES, SCA, 91192 Gif-sur-Yvette (France); Rohde, V. [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, 85748 Garching (Germany); Grisolia, C. [CEA, IRFM, 13108 Saint-Paul-lez-Durance (France); Chassefière, E. [Laboratoire Géosciences Paris Sud (GEOPS), UMR 8148, Université Paris Sud, 91403 Orsay Cedex (France)

    2015-08-15

    At the starting of the shutdown of the AUG (ASDEX Upgrade: Axially Symmetric Divertor EXperiment) German tokamak, we collected particles deposited on the divertor surfaces by means of a dedicated device called “Duster Box”. This device allows to collect the particles using a controlled airflow with a defined shear stress. Consequently, the particles collected correspond to a potentially mobilizable fraction, by an airflow, of deposited dust. A total of more than 70,000 tungsten particles was, analysed showing a bimodal particle size distribution with a mode composed of flakes at 0.6 μm and a mode composed of spherical particles at 1.8 μm.

  15. Characterization of dust particles produced in an all-tungsten wall tokamak and potentially mobilized by airflow

    International Nuclear Information System (INIS)

    At the starting of the shutdown of the AUG (ASDEX Upgrade: Axially Symmetric Divertor EXperiment) German tokamak, we collected particles deposited on the divertor surfaces by means of a dedicated device called “Duster Box”. This device allows to collect the particles using a controlled airflow with a defined shear stress. Consequently, the particles collected correspond to a potentially mobilizable fraction, by an airflow, of deposited dust. A total of more than 70,000 tungsten particles was, analysed showing a bimodal particle size distribution with a mode composed of flakes at 0.6 μm and a mode composed of spherical particles at 1.8 μm

  16. Disruption prediction with adaptive neural networks for ASDEX Upgrade

    International Nuclear Information System (INIS)

    In this paper, an adaptive neural system has been built to predict the risk of disruption at ASDEX Upgrade. The system contains a Self Organizing Map, which determines the 'novelty' of the input of a Multi Layer Perceptron predictor module. The answer of the MLP predictor will be inhibited whenever a novel sample is detected. Furthermore, it is possible that the predictor produces a wrong answer although it is fed with known samples. In this case, a retraining procedure will be performed to update the MLP predictor in an incremental fashion using data coming from both the novelty detection, and from wrong predictions. In particular, a new update is performed whenever a missed alarm is triggered by the predictor. The performance of the adaptive predictor during the more recent experimental campaigns until November 2009 has been evaluated.

  17. Real time capable infrared thermography for ASDEX Upgrade

    Science.gov (United States)

    Sieglin, B.; Faitsch, M.; Herrmann, A.; Brucker, B.; Eich, T.; Kammerloher, L.; Martinov, S.

    2015-11-01

    Infrared (IR) thermography is widely used in fusion research to study power exhaust and incident heat load onto the plasma facing components. Due to the short pulse duration of today's fusion experiments, IR systems have mostly been designed for off-line data analysis. For future long pulse devices (e.g., Wendelstein 7-X, ITER), a real time evaluation of the target temperature and heat flux is mandatory. This paper shows the development of a real time capable IR system for ASDEX Upgrade. A compact IR camera has been designed incorporating the necessary magnetic and electric shielding for the detector, cooler assembly. The camera communication is based on the Camera Link industry standard. The data acquisition hardware is based on National Instruments hardware, consisting of a PXIe chassis inside and a fibre optical connected industry computer outside the torus hall. Image processing and data evaluation are performed using real time LabVIEW.

  18. Real time capable infrared thermography for ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Sieglin, B., E-mail: Bernhard.Sieglin@ipp.mpg.de; Faitsch, M.; Herrmann, A.; Brucker, B.; Eich, T.; Kammerloher, L.; Martinov, S. [Max-Planck Institute for Plasma Physics, Boltzmannstr. 2, D-85748 Garching (Germany)

    2015-11-15

    Infrared (IR) thermography is widely used in fusion research to study power exhaust and incident heat load onto the plasma facing components. Due to the short pulse duration of today’s fusion experiments, IR systems have mostly been designed for off-line data analysis. For future long pulse devices (e.g., Wendelstein 7-X, ITER), a real time evaluation of the target temperature and heat flux is mandatory. This paper shows the development of a real time capable IR system for ASDEX Upgrade. A compact IR camera has been designed incorporating the necessary magnetic and electric shielding for the detector, cooler assembly. The camera communication is based on the Camera Link industry standard. The data acquisition hardware is based on National Instruments hardware, consisting of a PXIe chassis inside and a fibre optical connected industry computer outside the torus hall. Image processing and data evaluation are performed using real time LabVIEW.

  19. A new time constant in ASDEX determining the OH confinement

    International Nuclear Information System (INIS)

    The transient response of the stored energy to density variations is studied in ASDEX ohmic discharges. It is found that the phase delay between the stored energy to the density variations is much smaller than the energy confinement time, τE, in the density regime where τE scales like the Alcator scaling (anti ne c). The phase delay increases dramatically in the high density regime where τE saturates with density (anti ne > anti nc). The phase delay associated with density increase by pellet injection is small for operation at both high and low density. The value observed with pellet injection is as short as that seen in the low density gas puffing regime. (orig./GG)

  20. Disruption prediction with adaptive neural networks for ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Cannas, B.; Fanni, A. [Electrical and Electronic Engineering Dept., University of Cagliari, Piazza D' Armi, 09123 Cagliari (Italy); Pautasso, G. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany); Sias, G., E-mail: giuliana.sias@diee.unica.it [Electrical and Electronic Engineering Dept., University of Cagliari, Piazza D' Armi, 09123 Cagliari (Italy)

    2011-10-15

    In this paper, an adaptive neural system has been built to predict the risk of disruption at ASDEX Upgrade. The system contains a Self Organizing Map, which determines the 'novelty' of the input of a Multi Layer Perceptron predictor module. The answer of the MLP predictor will be inhibited whenever a novel sample is detected. Furthermore, it is possible that the predictor produces a wrong answer although it is fed with known samples. In this case, a retraining procedure will be performed to update the MLP predictor in an incremental fashion using data coming from both the novelty detection, and from wrong predictions. In particular, a new update is performed whenever a missed alarm is triggered by the predictor. The performance of the adaptive predictor during the more recent experimental campaigns until November 2009 has been evaluated.

  1. Real time capable infrared thermography for ASDEX Upgrade

    International Nuclear Information System (INIS)

    Infrared (IR) thermography is widely used in fusion research to study power exhaust and incident heat load onto the plasma facing components. Due to the short pulse duration of today’s fusion experiments, IR systems have mostly been designed for off-line data analysis. For future long pulse devices (e.g., Wendelstein 7-X, ITER), a real time evaluation of the target temperature and heat flux is mandatory. This paper shows the development of a real time capable IR system for ASDEX Upgrade. A compact IR camera has been designed incorporating the necessary magnetic and electric shielding for the detector, cooler assembly. The camera communication is based on the Camera Link industry standard. The data acquisition hardware is based on National Instruments hardware, consisting of a PXIe chassis inside and a fibre optical connected industry computer outside the torus hall. Image processing and data evaluation are performed using real time LabVIEW

  2. Carbon deposition and deuterium inventory in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Carbon erosion and deposition in the ASDEX Upgrade divertor was investigated using a poloidal section of marked divertor tiles and silicon samples below the divertor structure. The whole inner divertor is a net carbon deposition area, while a large fraction of the outer divertor is erosion dominated and the roof baffle tiles show a complicated distribution of erosion and deposition areas. In total, 43.7 g B+C were redeposited, of which 88% were deposited on tiles and 9% in remote areas (below roof baffle, on vessel wall structures). Identified carbon sources in the main chamber are too low by a factor of ten to explain the observed carbon divertor deposition, but carbon erosion is observed at the outer divertor tiles. Deuterium is trapped mainly on the surfaces of the inner divertor tiles. The long term retention in codeposited hydrocarbon layers is about 3% of the total deuterium fuel input. (author)

  3. First results of ion cyclotron resonance heating on ASDEX upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Noterdaeme, J.; Hoffmann, C.; Brambilla, M.; Buechl, K.; Eberhagen, A.; Field, A.; Fuchs, C.; Gehre, O.; Gernhardt, J.; Gruber, O.; Haas, G.; Hermann, A.; Hofmeister, F.; Kallenbach, A.; Lieder, G.; Mertens, V.; Murmann, H.; de Pena Hempel, S.; Poschenrieder, W.; Richter, T.; Ryter, F.; Salmon, N.; Salzmann, H.; Schneider, W.; Wesner, F.; Zehrfeld, H.; Zohm, H. (Max-Planck-Institute for Plasmaphysics, D-8046 Garching (Germany)); ASDEX Upgrade Team

    1994-10-15

    ASDEX Upgrade is equipped with an ICRH system consisting of 4 generators of 2 MW power each and 4 double loop antennas. The generators, tuneable in frequency from 30 to 120 MHz, cover several heating scenarios over a wide range of magnetic fields (1 T[lt]B[sub t][lt]3.9 T): minority heating of H and He[sub 3] and second harmonic heating of H and D. ICRH-heated discharges in ASDEX Upgrade were so far carried out mainly at 30 MHz and a magnetic field of 2 T (H minority in D and He). Peak powers of 2.4 MW and pulse length up to 2.5 s were achieved (total energy 3.75 MJ). In L-mode, the density on turn-on of the ICRH stays constant, or even decreases. The ratio of radiated power to total input power is unchanged (60% in an unboronized machine, 30% in a freshly boronized machine) between Ohmic and ICRH phases. The electron temperature increases with 0.9 MW from 1 to 1.25 keV, the loop voltage drops. Transitions to the H-mode were easily and reliably achieved with ICRH alone (necessary ICRH power as low as 0.9 MW) and the length of the ELMy H-mode phases was limited only by the applied ICRH pulse length (ELMy H-mode phases of up to 2 s were achieved). The paper presents further results on heating and confinement in L and H-mode, antenna and edge studies and on divertor measurements. Preliminary experiments, performed with a combination of H minority heating (30 MHz) and H second harmonic (60 MHz) in 600 kA He and D discharges (H minority in the 5 to 20% range) at 2 T, and with non-resonant heating (30 MHz and 60 MHz at 1.35 T) are briefly discussed.

  4. Tokamak engineering mechanics

    International Nuclear Information System (INIS)

    Provides a systematic introduction to tokamaks in engineering mechanics. Includes design guides based on full mechanical analysis, which makes it possible to accurately predict load capacity and temperature increases. Presents comprehensive information on important design factors involving materials. Covers the latest advances in and up-to-date references on tokamak devices. Numerous examples reinforce the understanding of concepts and provide procedures for design. Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study of mechanical/fusion engineering with a general understanding of tokamak engineering mechanics.

  5. Tokamak engineering mechanics

    CERN Document Server

    Song, Yuntao; Du, Shijun

    2013-01-01

    Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study

  6. Tokamak concept innovations

    International Nuclear Information System (INIS)

    This document contains the results of the IAEA Specialists' Meeting on Tokamak Concept Innovations held 13-17 January 1986 in Vienna. Although it is the most advanced fusion reactor concept the tokamak is not without its problems. Most of these problems should be solved within the ongoing R and D studies for the next generation of tokamaks. Emphasis for this meeting was placed on innovations that would lead to substantial improvements in a tokamak reactor, even if they involved a radical departure from present thinking

  7. Avoidance of Disruptions at High βN in ASDEX Upgrade with off-Axis ECRH

    International Nuclear Information System (INIS)

    Full text: The control of disruptions in tokamak plasmas by means of electron cyclotron resonance heating (ECRH) has been addressed in several machines. The technique is based on the stabilization of magnetohydrodynamic (MHD) modes occurring at a disruption through the localized injection of ECRH on a resonant surface. A delay in the occurrence of disruption has been achieved in some cases and complete avoidance in other cases. So far, these experiments have dealt with L-mode plasmas. The first experiments of this type carried out in H-mode plasmas are described in this paper. Delay and/or complete avoidance of disruptions has been achieved in ASDEX Upgrade using localized injection of ECRH (1.5 MW) in a high βN scenario (1 MA, 2.2 T, with NBI (7.5 MW)). In these discharges (at relatively low q95 and low density) neoclassical tearing modes (NTMs) are excited and when they lock a disruption occurs. The same scheme used in previous disruption avoidance experiments in FTU and AUG has been applied: as soon as the disruption precursor signal (loop voltage and/or locked mode detector) reaches the preset threshold, the ECRH power is triggered by real time control. A poloidal scan in deposition location has been performed by varying the poloidal angles of the launching mirrors in different discharges. Complete avoidance is achieved when the power is injected close to the q = 3/2 surface: in this case multiple unlocking of MHD modes occurs after ECRH application. When ECRH is injected at more external locations, the discharges, although not disrupting immediately as in the reference case, show no mode unlocking and eventually disrupt. For injection at inner locations the discharge disrupts as in the reference case. The absorption of ECRH is therefore found to modify the sequence of events occurring at a disruption by acting locally (through a change in resistivity) on the gradient of plasma current which is supposed to be the main drive of the island amplitude growth rate

  8. Novel free-boundary equilibrium and transport solver with theory-based models and its validation against ASDEX Upgrade current ramp scenarios

    International Nuclear Information System (INIS)

    Tokamak scenario development requires an understanding of the properties that determine the kinetic profiles in non-steady plasma phases and of the self-consistent evolution of the magnetic equilibrium. Current ramps are of particular interest since many transport-relevant parameters explore a large range of values and their impact on transport mechanisms has to be assessed. To this purpose, a novel full-discharge modelling tool has been developed, which couples the transport code ASTRA (Pereverzev et al 1991 IPP Report 5/42) and the free boundary equilibrium code SPIDER (Ivanov et al 2005 32nd EPS Conf. on Plasma Physics vol 29C (ECA) P-5.063 and http://epsppd.epfl.ch/Tarragona/pdf/P5_063.pdf), utilizing a specifically designed coupling scheme. The current ramp-up phase can be accurately and reliably simulated using this scheme, where a plasma shape, position and current controller is applied, which mimics the one of ASDEX Upgrade. Transport of energy is provided by theory-based models (e.g. TGLF (Staebler et al 2007 Phys. Plasmas 14 055909)). A recipe based on edge-relevant parameters (Scott 2000 Phys. Plasmas 7 1845) is proposed to resolve the low current phase of the current ramps, where the impact of the safety factor on micro-instabilities could make quasi-linear approaches questionable in the plasma outer region. Current ramp scenarios, selected from ASDEX Upgrade discharges, are then simulated to validate both the coupling with the free-boundary evolution and the prediction of profiles. Analysis of the underlying transport mechanisms is presented, to clarify the possible physics origin of the observed L-mode empirical energy confinement scaling. The role of toroidal micro-instabilities (ITG, TEM) and of non-linear effects is discussed. (paper)

  9. Impact of magnetic perturbation fields on tokamak plasmas

    International Nuclear Information System (INIS)

    Non-axisymmetric external magnetic perturbation (MP) fields arise in every tokamak e.g. due to not perfectly positioned external coils. Additionally many tokamaks, like ASDEX Upgrade (AUG), are equipped with a set of external coils, which produce a 3D MP field in addition to the equilibrium field. This field is used to either compensate for the intrinsic MP field or to influence MHD instabilities such as Edge Localised Modes (ELMs) or Neoclassical Tearing Modes (NTMs). But these MP fields can also give rise to a more global plasma response. The resonant components can penetrate the plasma and influence the stability of existing NTMs or even lead to their formation via magnetic reconnection. In addition they exert a local torque on the plasma. These effects are less pronounced at high plasma rotation where the resonant field components are screened. The non-resonant components do not influence NTMs directly but slow down the plasma rotation globally via the neoclassical toroidal viscous torque. The island formation caused by the MP field as well as the interaction of pre-existing islands with the MP field at AUG is presented. It is shown that these effects can be modelled using a simple forced reconnection theory. Also the effect of resonant and non-resonant MPs on the plasma rotation at AUG is discussed.

  10. Electrical conductivity in tokamaks and extended neoclassical theory

    International Nuclear Information System (INIS)

    The electrical conductivity measurements reported from various tokamaks (D-III, PLT, TEXT, ASDEX, JT-60, TEXTOR, JET, TFTR) and compared with the usual neoclassical theory are here also compared with the extended neoclassical theory where the electron-electron collision rate is anomalous while the electron-ion collision rate remains Coulombian. It is found that, out of the 14 experiments considered, three are consistent with both the neoclassical and the extended neoclassical theories, four are consistent only with the extended neoclassical theory, and four are consistent with the neoclassical theory and also, within the experimental errors, not inconsistent with the extended neoclassical theory; the remaining three experiments appear to be incompatible with both theories. It is concluded that the extended neoclassical theory is in better agreement with conductivity experiments than the conventional neoclassical theory and, indeed, the extended theory is a serious candidate for explaining tokamak behaviour, since it accommodates naturally an anomalous electron thermal transport, which the conventional neoclassical theory is unable to do. (author). 31 refs, 1 fig

  11. Analysis of fast ion induced instabilities in tokamak plasmas

    CERN Document Server

    Horváth, László

    2015-01-01

    In magnetic confinement fusion devices like tokamaks, it is crucial to confine the high energy fusion-born helium nuclei ($\\alpha$-particles) to maintain the energy equilibrium of the plasma. However, energetic ions can excite various instabilities which can lead to their enhanced radial transport. Consequently, these instabilities may degrade the heating efficiency and they can also cause harmful power loads on the plasma-facing components of the device. Therefore, the understanding of these modes is a key issue regarding future burning plasma experiments. One of the main open questions concerning energetic particle (EP) driven instabilities is the non-linear evolution of the mode structure. In this thesis, I present my results on the investigation of $\\beta$-induced Alfv\\'{e}n eigenmodes (BAEs) and EP-driven geodesic acoustic modes (EGAMs) observed in the ramp-up phase of off-axis NBI heated plasmas in the ASDEX Upgrade tokamak. These modes were well visible on several line-of-sights (LOSs) of the soft X-ra...

  12. Real-time control of the plasma density profile on ASDEX upgrade

    International Nuclear Information System (INIS)

    The tokamak concept currently is the most promising approach to future power generation by controlled thermonuclear fusion. The spatial distribution of the particle density in the toroidally confined fusion plasma is of particular importance. This thesis work therefore focuses on the question as to what extent the shape of the density profile can be actively controlled by a feedback loop in the fusion experiment ASDEX Upgrade. There are basically two essential requirements for such feedback control of the density profile, which has been experimentally demonstrated within the scope of this thesis work: On the one hand, for this purpose the density profile must be continuously calculated under real-time constraints during a plasma discharge. The calculation of the density profile is based on the measurements of a sub-millimeter interferometer, which provides the line-integrated electron density along 5 chords through the plasma. Interferometric density measurements can suffer from counting errors by integer multiples of 2π when detecting the phase difference between a probing and a reference beam. As such measurement errors have severe impact on the reconstructed density profile, one major part of this work consists in the development of new readout electronics for the interferometer, which allows for detection of such measurement errors in real-time with high reliability. A further part of this work is the design of a computer algorithm which reconstructs the spatial distribution of the plasma density from the line-integrated measurements. This algorithm has to be implemented on a computer which communicates the measured data to other computers in real-time, especially to the tokamak control system. On the other hand, a second fundamental requirement for the successful implementation of a feedback controller is the identification of at least one actuator which enables a modification of the density profile. Here, electron cyclotron resonance heating (ECRH) has been

  13. 3D Fokker-Planck calculation of combined fast wave/lower hybrid and electron cyclotron current drive in tokamaks

    International Nuclear Information System (INIS)

    In a non-reactor tokamak environment, lower hybrid current drive can be combined with electron cyclotron waves, both (1) to control the radial profile of LH current deposition, and (2) to enhance the current drive efficiency. A related rf synergy is the use of multiple LH spectra for radial profile control as demonstrated in the ASDEX tokamak. In a reactor environment, fast waves provide an appropriate primary current drive system which can penetrate radially to the plasma core, and can be combined with ECCD. We use the CQL3D Fokker-Planck code to study these processes. Modelings of LHCD radial profile control by ''filling the spectral gap'' with EC or with additional LH power are presented. In the reactor environment, a range of cases with combined fast wave and electron cyclotron waves are examined, but no useful synergisms are found

  14. Effect of sonic poloidal flows in determining flow and density asymmetries for trace impurities in the tokamak edge pedestal

    CERN Document Server

    Fable, E; Viezzer, E

    2013-01-01

    The structure of poloidal and toroidal flows of trace impurities in the edge pedestal of tokamak plasmas is studied analytically and numerically. Parallel momentum balance is analysed upon retaining the following terms: poloidal and toroidal centrifugal forces (inertia), pressure force, electric force, and the friction force. It is shown that, when the poloidal flow is such to produce a properly defined Mach number of order unity somewhere on the flux surface, shock fronts can form. The shock fronts can modify the predicted asymmetry structures in both the flow and the density profile along the poloidal arc. Predictions of the theory are shown against experimental observations in the ASDEX Upgrade tokamak, showing good qualitative and quantitative agreement if the inertia term associated with the poloidal flow is retained.

  15. Modelling and experiments on NTM stabilisation at ASDEX upgrade

    International Nuclear Information System (INIS)

    In the next fusion device ITER the so-called neoclassical tearing modes (NTMs) are foreseen as being extremely detrimental to plasma confinement. This type of resistive instability is related to the presence in the plasma of magnetic islands. These are experimentally controlled with local electron cyclotron current drive (ECCD) and the island width decay during NTM stabilisation is modelled using the so-called Modified Rutherford equation. In this thesis, a modelling of the Modified Rutherford equation is carried out and simulations of the island width decay are compared with the experimentally observed ones in order to fit the two free machine-independent parameters present in the equation. A systematic study on a database of NTM stabilisation discharges from ASDEX Upgrade and JT-60U is done within the context of a multi-machine benchmark for extrapolating the ECCD power requirements for ITER. The experimental measurements in both devices are discussed by means of consistency checks and sensitivity analysis and used to evaluate the two fitting parameters present in the Modified Rutherford equation. The influence of the asymmetry of the magnetic island on stabilisation is for the first time included in the model and the effect of ECCD on the marginal island after which the mode naturally decays is quantified. The effect of radial misalignment and over-stabilisation during the experiment are found to be the key quantities affecting the NTM stabilisation. As a main result of this thesis, the extrapolation to ITER of the NTM stabilisation results from ASDEX Upgrade and JT-60U shows that 10MW of ECCD power are enough to stabilise large NTMs as long as the O-point of the island and the ECCD beam are perfectly aligned. In fact, the high ratio between the island size at saturation and the deposition width of the ECCD beam foreseen for ITER is found to imply a maximum allowable radial misalignment of 2-3 cm and little difference in terms of gained performance between

  16. Modelling and experiments on NTM stabilisation at ASDEX upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Urso, Laura

    2009-07-27

    In the next fusion device ITER the so-called neoclassical tearing modes (NTMs) are foreseen as being extremely detrimental to plasma confinement. This type of resistive instability is related to the presence in the plasma of magnetic islands. These are experimentally controlled with local electron cyclotron current drive (ECCD) and the island width decay during NTM stabilisation is modelled using the so-called Modified Rutherford equation. In this thesis, a modelling of the Modified Rutherford equation is carried out and simulations of the island width decay are compared with the experimentally observed ones in order to fit the two free machine-independent parameters present in the equation. A systematic study on a database of NTM stabilisation discharges from ASDEX Upgrade and JT-60U is done within the context of a multi-machine benchmark for extrapolating the ECCD power requirements for ITER. The experimental measurements in both devices are discussed by means of consistency checks and sensitivity analysis and used to evaluate the two fitting parameters present in the Modified Rutherford equation. The influence of the asymmetry of the magnetic island on stabilisation is for the first time included in the model and the effect of ECCD on the marginal island after which the mode naturally decays is quantified. The effect of radial misalignment and over-stabilisation during the experiment are found to be the key quantities affecting the NTM stabilisation. As a main result of this thesis, the extrapolation to ITER of the NTM stabilisation results from ASDEX Upgrade and JT-60U shows that 10MW of ECCD power are enough to stabilise large NTMs as long as the O-point of the island and the ECCD beam are perfectly aligned. In fact, the high ratio between the island size at saturation and the deposition width of the ECCD beam foreseen for ITER is found to imply a maximum allowable radial misalignment of 2-3 cm and little difference in terms of gained performance between

  17. A new B-dot probe-based diagnostic for amplitude, polarization, and wavenumber measurements of ion cyclotron range-of frequency fields on ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Ochoukov, R.; Bobkov, V.; Faugel, H.; Fünfgelder, H. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Noterdaeme, J.-M. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Applied Physics Department, University Gent, 9000 Gent (Belgium)

    2015-11-15

    A new B-dot probe-based diagnostic has been installed on an ASDEX Upgrade tokamak to characterize ion cyclotron range-of frequency (ICRF) wave generation and interaction with magnetized plasma. The diagnostic consists of a field-aligned array of B-dot probes, oriented to measure fast and slow ICRF wave fields and their field-aligned wavenumber (k{sub //}) spectrum on the low field side of ASDEX Upgrade. A thorough description of the diagnostic and the supporting electronics is provided. In order to compare the measured dominant wavenumber of the local ICRF fields with the expected spectrum of the launched ICRF waves, in-air near-field measurements were performed on the newly installed 3-strap ICRF antenna to reconstruct the dominant launched toroidal wavenumbers (k{sub tor}). Measurements during a strap current phasing scan in tokamak discharges reveal an upshift in k{sub //} as strap phasing is moved away from the dipole configuration. This result is the opposite of the k{sub tor} trend expected from in-air near-field measurements; however, the near-field based reconstruction routine does not account for the effect of induced radiofrequency (RF) currents in the passive antenna structures. The measured exponential increase in the local ICRF wave field amplitude is in agreement with the upshifted k{sub //}, as strap phasing moves away from the dipole configuration. An examination of discharges heated with two ICRF antennas simultaneously reveals the existence of beat waves at 1 kHz, as expected from the difference of the two antennas’ operating frequencies. Beats are observed on both the fast and the slow wave probes suggesting that the two waves are coupled outside the active antennas. Although the new diagnostic shows consistent trends between the amplitude and the phase measurements in response to changes applied by the ICRF antennas, the disagreement with the in-air near-field measurements remains. An electromagnetic model is currently under development to

  18. Commissioning activities and first results from the collective Thomson scattering diagnostic on ASDEX Upgrade (invited)

    DEFF Research Database (Denmark)

    Meo, Fernando; Bindslev, Henrik; Korsholm, Søren Bang; Furtula, Vedran; Leuterer, F.; Leipold, Frank; Michelsen, Poul; Nielsen, Stefan Kragh; Salewski, Mirko; Stober, J.; Wagner, D.; Woskov, P.

    2008-01-01

    The collective Thomson scattering (CTS) diagnostic installed on ASDEX Upgrade uses millimeter waves generated by the newly installed 1 MW dual frequency gyrotron as probing radiation at 105 GHz. It measures backscattered radiation with a heterodyne receiver having 50 channels (between 100 and 110...... alignment of the system. First results in near perpendicular of scattered spectra in a neutral beam injection (NBI) and ion cyclotron resonance heating (ICRH) plasma (minority hydrogen) on ASDEX Upgrade have shown evidence of ICRH heating phase of hydrogen. ©2008 American Institute of Physics...

  19. Multivariate statistical models for disruption prediction at ASDEX Upgrade

    International Nuclear Information System (INIS)

    In this paper, a disruption prediction system for ASDEX Upgrade has been proposed that does not require disruption terminated experiments to be implemented. The system consists of a data-based model, which is built using only few input signals coming from successfully terminated pulses. A fault detection and isolation approach has been used, where the prediction is based on the analysis of the residuals of an auto regressive exogenous input model. The prediction performance of the proposed system is encouraging when it is applied to the same set of campaigns used to implement the model. However, the false alarms significantly increase when we tested the system on discharges coming from experimental campaigns temporally far from those used to train the model. This is due to the well know aging effect inherent in the data-based models. The main advantage of the proposed method, with respect to other data-based approaches in literature, is that it does not need data on experiments terminated with a disruption, as it uses a normal operating conditions model. This is a big advantage in the prospective of a prediction system for ITER, where a limited number of disruptions can be allowed

  20. Power inverter design for ASDEX Upgrade saddle coils

    Energy Technology Data Exchange (ETDEWEB)

    Teschke, M., E-mail: teschke@ipp.mpg.de [Max Planck Institute for Plasma Physics, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany); Suttrop, W.; Rott, M. [Max Planck Institute for Plasma Physics, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany)

    2013-10-15

    Highlights: ► A cost effective inverter topology for AUG's 16 in-vessel saddle coils has been found. ► Use of commercially available power modules is possible. ► Exchange of reactive power between multiple inverters is possible. ► Influence of electromagnetic noise to AUG's diagnostics was measured. ► Gas insulation of electric feed through significantly depends on magnetic fields. It is protected by fast turn-off circuit. -- Abstract: A set of 16 in-vessel saddle coils has been installed in the ASDEX Upgrade (AUG) experiment since the end of 2011 [1]. To achieve full performance, it is necessary to operate them with alternating current (AC) of arbitrary waveforms. To generate spatially resolved magnetic fields, it is required to allocate separate power inverters to every single coil. Therefore, different topologies are analyzed and compared. Studies of the commutation behavior of power stages, different pulse width modulation (PWM) schemes and single-phase-to-earth fault detection are executed. Experiments to evaluate the electromagnetic interference (EMI) of possible inverter topologies on the AUG diagnostics are done as well. A special focus is put on the feasibility of analyzed topologies using industrially available and fully assembled “power modules” to minimize development effort and costs.

  1. Progress in controlling ICRF-edge interactions in ASDEX upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Bobkov, Vl., E-mail: bobkov@ipp.mpg.de; Ochoukov, R.; Bilato, R.; Braun, F.; Carralero, D.; Dux, R.; Faugel, H.; Fünfgelder, H.; Jacquot, J.; Lunt, T.; Potzel, S.; Pütterich, Th. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr.2, D-85748 (Germany); Jacquet, Ph.; Monakhov, I. [CCFE, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Zhang, W.; Noterdaeme, J.-M.; Stepanov, I. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr.2, D-85748 (Germany); Department of Applied Physics, Gent University, 9000 Gent (Belgium); Colas, L.; Meyer, O. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Czarnecka, A. [Institute of Plasma Physics and Laser Microfusion, Hery 23 Str., 01-497 Warsaw (Poland); and others

    2015-12-10

    RF measurements during variation of the strap voltage balance of the original 2-strap ICRF antenna in ASDEX Upgrade at constant power are consistent with electromagnetic calculations by HFSS and TOPICA, more so for the latter. RF image current compensation is observed at the antenna limiters in the experiment at a local strap voltage of about half of the value of the remote strap, albeit with a non-negligible uncertainty in phasing. The RF-specific tungsten (W) source at the broad-limiter 2-strap antenna correlates strongly with the RF voltage at the local strap at the locations not connected to opposite side of the antenna along magnetic field lines. The trends of the observed increase of the RF loading with injection of local gas are well described by a combined EMC3-Eirene – FELICE calculations, with the most efficient improvement confirmed for the outer-midplane valves, but underestimated by about 1/3. The corresponding deuterium density tailoring is also likely responsible for the decrease of local W sources observed in the experiment.

  2. Tokamak ARC damage

    International Nuclear Information System (INIS)

    Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage

  3. International tokamak reactor

    International Nuclear Information System (INIS)

    Since 1978, the US, the European Communities, Japan, and the Soviet Union have collaborated on the definition, conceptual design, data base assessment, and analysis of critical technical issues for a tokamak engineering test reactor, called the International Tokamak Reactor (INTOR). During 1985-1986, this activity has been expanded in scope to include evaluation of concept innovations that could significantly improve the tokamak as a commercial reactor. The purposes of this paper are to summarize the present INTOR design concept and to summarize the work on concept innovations

  4. Survey of Tokamak experiments

    International Nuclear Information System (INIS)

    The survey covers the following topics:- Introduction and history of tokamak research; review of tokamak apparatus, existing and planned; remarks on measurement techniques and their limitations; main results in terms of electron and ion temperatures, plasma density, containment times, etc. Empirical scaling; range of operating densities; impurities, origin, behaviour and control (including divertors); data on fluctuations and instabilities in tokamak plasmas; data on disruptive instabilities; experiments on shaped cross-sections; present experimental evidence on β limits; auxiliary heating; experimental and theoretical problems for the future. (author)

  5. The Thor tokamak experiment

    International Nuclear Information System (INIS)

    The main characteristics of the plasma produced in Thor tokamak discharges are described. The machine performances are outlined and the experimental results relevant to the equilibrium, the stability and the control of the discharge regimes are discussed in detail. (author)

  6. Modular tokamak magnetic system

    Science.gov (United States)

    Yang, Tien-Fang

    1988-01-01

    A modular tokamak system comprised of a plurality of interlocking moldules. Each module is comprised of a vacuum vessel section, a toroidal field coil, moldular saddle coils which generate a poloidal magnetic field and ohmic heating coils.

  7. Research using small tokamaks

    International Nuclear Information System (INIS)

    These proceedings of the IAEA-sponsored meeting held in Nice, France 10-11 October, 1988, contain the manuscripts of the 21 reports dealing with research using small tokamaks. The purpose of this meeting was to highlight some of the achievements of small tokamaks and alternative magnetic confinement concepts and assess the suitability of starting new programs, particularly in developing countries. Papers presented were either review papers, or were detailed descriptions of particular experiments or concepts. Refs, figs and tabs

  8. Tokamak simulation code manual

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Moon Kyoo; Oh, Byung Hoon; Hong, Bong Keun; Lee, Kwang Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-01-01

    The method to use TSC (Tokamak Simulation Code) developed by Princeton plasma physics laboratory is illustrated. In KT-2 tokamak, time dependent simulation of axisymmetric toroidal plasma and vertical stability have to be taken into account in design phase using TSC. In this report physical modelling of TSC are described and examples of application in JAERI and SERI are illustrated, which will be useful when TSC is installed KAERI computer system. (Author) 15 refs., 6 figs., 3 tabs.

  9. Tokamak simulation code manual

    International Nuclear Information System (INIS)

    The method to use TSC (Tokamak Simulation Code) developed by Princeton plasma physics laboratory is illustrated. In KT-2 tokamak, time dependent simulation of axisymmetric toroidal plasma and vertical stability have to be taken into account in design phase using TSC. In this report physical modelling of TSC are described and examples of application in JAERI and SERI are illustrated, which will be useful when TSC is installed KAERI computer system. (Author) 15 refs., 6 figs., 3 tabs

  10. Edge density measurements with a fast Li beam probe in tokamak and stellarator experiments

    International Nuclear Information System (INIS)

    High energy neutral Li beam probes have advanced to the point where they are a standard diagnostic tool on W7-AS and ASDEX-Upgrade, both in terms of the Li beam injector and the reconstruction algorithm to arrive at densities ne(z) along the beam from the Li[2p-2s] resonance line profile. With beam energies in the range 30-70 keV and neutral equivalent currents greater than 1 mA, it is possible to derive ne(z) profiles for line densities nez14 cm-2, with a radial resolution of about 0.5 cm and time response of less than 0.2 ms. By way of example, the diagnostic layout on the W7-AS stellarator is sketched and salient results from experiments discussed. Densities over the range 1012-1014 cm-3 are accessible, permitting full coverage of the core density gradient region on W7-AS. Examples from the ASDEX tokamak that involve the H-mode and pellet injection are presented to exemplify the time response. (orig.)

  11. ASDEX-upgrade poloidal field coils. Specifications, load, stress. Pt. 1

    International Nuclear Information System (INIS)

    The pulsed and irregular load of the ASDEX-Upgrade poloidal field coils postulates a careful analysis of the maximum stresses in order to find out the criteria for specifications and supervision. For operating conditions known at the time of coil design, maximum stresses and loads are compiled and the resulting features and limitations of coil operation are explained. (orig.)

  12. Fast ion millimeter wave collective Thomson scattering diagnostics on TEXTOR and ASDEX upgrades

    DEFF Research Database (Denmark)

    Michelsen, S.; Korsholm, Søren Bang; Bindslev, H.; Meo, F.; Michelsen, Poul; Tsakadze, E.L.; Egedal, J.; Woskov, P.; Hoekzema, J.A.; Leuterer, F.; Westerhof, E.

    2004-01-01

    Collective Thomson scattering (CTS) diagnostic systems for measuring fast ions in TEXTOR and ASDEX Upgrade are described in this article. Both systems use millimeter waves generated by gyrotrons as probing radiation and the scattered radiation is detected with heterodyne receivers having 40...

  13. ASDEX papers at the 13th European conference on controlled fusion and plasma physics

    International Nuclear Information System (INIS)

    This report provides 29 ASDEX papers concerning pellet refuelling, confinement, high-beta plasma and MHD-equilibrium, heating by ICR, lower hybrid and current-drive, impurity studies and plasma diagnostics. All of these papers have been indexed separately. (GG)

  14. Enhancement of the FIDA diagnostic at ASDEX Upgrade for velocity space tomography

    DEFF Research Database (Denmark)

    Weiland, M.; Geiger, B.; Jacobsen, Asger Schou;

    2016-01-01

    Recent upgrades to the FIDA (fast-ion D-alpha) diagnostic at ASDEX Upgrade are discussed. The diagnostic has been extended from three to five line of sight arrays with different angles to the magnetic field, and a spectrometer redesign allows the simultaneous measurement of red- and blue-shifted ...

  15. Commissioning of inline ECE system within waveguide based ECRH transmission systems on ASDEX upgrade

    NARCIS (Netherlands)

    Bongers, W.A.; Kasparek, W.; Doelman, N. J.; Braber, R. van den; Brand, H. van den; Meo, F.; Baar, M.R. de; Amerongen, F.J.; Donné, A.J.H.; Elzendoorn, B.S.Q.; Erckmann, V.; Goede, A.P.H.; Giannone, L.; Grünwald, G.; Hollman, F.; Kaas, G.; Krijger, B.; Michel, G.; Lubyako, L.; Monaco, F.; Noke, F.; Petelin, M.; Plaum, B.; Purps, F.; Pierik, J.G.W. ten; Schüller, C.; Slob, J.W.; Stober, J.K.; Schütz, H.; Wagner, D.; Westerhof, E.; Ronden, D.M.S.

    2012-01-01

    A CW capable inline electron cyclotron emission (ECE) separation system for feedback control, featuring oversized corrugated waveguides, is commissioned on ASDEX upgrade (AUG). The system is based on a combination of a polarization independent, non-resonant, Mach-Zehnder diplexer equipped with diele

  16. Comparison of fast ion collective Thomson scattering measurements at ASDEX Upgrade with numerical simulations

    DEFF Research Database (Denmark)

    Salewski, Mirko; Meo, Fernando; Stejner Pedersen, Morten;

    2010-01-01

    Collective Thomson scattering (CTS) experiments were carried out at ASDEX Upgrade to measure the one-dimensional velocity distribution functions of fast ion populations. These measurements are compared with simulations using the codes TRANSP/NUBEAM and ASCOT for two different neutral beam injecti...

  17. Carbon influx studies in the main chamber of ASDEX Upgrade

    International Nuclear Information System (INIS)

    Carbon sources in the main chamber of ASDEX Upgrade, especially the 12 guard limiters at the low field side (LFS), were determined spectroscopically using recently installed lines of sight. Absolute photon fluxes were measured for spectral lines in the visible wavelength range referring to all spin systems of C+1 and C+2. A simple transport model for carbon enabled the simulation of the radial distribution of carbon radiation and the determination of the effective inverse photon efficiency, which was used for the evaluation of ion fluxes. The model also predicts the fraction of eroded particles that are transported out of the plasma before further ionization occurs. Comparison of the calculated losses with measurements showed good agreement in L-mode cases, whereas in H-mode cases the CIII/CII radiation ratio was too high by a factor 1.5. The contribution of each spin system to the ion flux was independently measured. For C+1 and C+2 the spin system distribution was found to be close to equilibrium. The line-of-sight-integrated photon fluxes were spatially separated for many lines of sight by Zeeman-analysis and differential measurements. This allowed us to determine the total influx from the high field side and LFS. Surprisingly, the carbon source at the inner heat shield was larger than the carbon influx from the limiter source at the LFS. This is very pronounced for the H-mode case investigated, where 60-80% of the carbon atoms emerge from the heat shield. This source is due to recycling or re-erosion of carbon, which probably originates from the limiters, because approximately 85% of the heat shield area consisted of tungsten coated tiles

  18. Carbon influx studies in the main chamber of ASDEX Upgrade

    Science.gov (United States)

    Pütterich, T.; Dux, R.; Gafert, J.; Kallenbach, A.; Neu, R.; Pugno, R.; Yoon, S. W.; ASDEX Upgrade Team

    2003-10-01

    Carbon sources in the main chamber of ASDEX Upgrade, especially the 12 guard limiters at the low field side (LFS), were determined spectroscopically using recently installed lines of sight. Absolute photon fluxes were measured for spectral lines in the visible wavelength range referring to all spin systems of C+1 and C+2. A simple transport model for carbon enabled the simulation of the radial distribution of carbon radiation and the determination of the effective inverse photon efficiency, which was used for the evaluation of ion fluxes. The model also predicts the fraction of eroded particles that are transported out of the plasma before further ionization occurs. Comparison of the calculated losses with measurements showed good agreement in L-mode cases, whereas in H-mode cases the CIII/CII radiation ratio was too high by a factor 1.5. The contribution of each spin system to the ion flux was independently measured. For C+1 and C+2 the spin system distribution was found to be close to equilibrium. The line-of-sight-integrated photon fluxes were spatially separated for many lines of sight by Zeeman-analysis and differential measurements. This allowed us to determine the total influx from the high field side and LFS. Surprisingly, the carbon source at the inner heatshield was larger than the carbon influx from the limiter source at the LFS. This is very pronounced for the H-mode case investigated, where 60-80% of the carbon atoms emerge from the heatshield. This source is due to recycling or re-erosion of carbon, which probably originates from the limiters, because ap85% of the heatshield area consisted of tungsten coated tiles.

  19. CURRIN - a mathematical model of the electricity increase phase of a tokamak-plasma

    International Nuclear Information System (INIS)

    The zero-dimensional time-dependent mathematical model CURRIN for the current initiation phase in a large tokamak is presented. It describes the ionization of the neutral gas in terms of coupled mass and energy conservation equations for five components - H02, H+2, H01, H+1 and electrons. For studying impurity effects nine equations for O01 to O81 have been added. These equations are coupled to an equivalent external OH circuit to simulate the tokamak start-up phase. All rate coefficients for hydrogen and oxygen needed for the above equation are compiled. The system of equations is prepared for numerical integration by normalization. Its numerical solution forms the basis of the computer model CURRIN. The particle and energy balances of the system are continuously checked and monitored during the solution. Finally, some sample calculations are presented which refer to the current initiation phase in the ASDEX divertor tokamak. Special regard is devoted to the computer time needed for the various options of CURRIN. (orig.)

  20. Recent advances in gyrokinetic full-f particle simulation of medium sized Tokamaks with ELMFIRE

    International Nuclear Information System (INIS)

    Large-scale kinetic simulations of toroidal plasmas based on first principles are called for in studies of transition from low to high confinement mode and internal transport barrier formation in the core plasma. Such processes are best observed and diagnosed in detached plasma conditions in mid-sized tokamaks, so gyrokinetic simulations for these conditions are warranted. A first principles test-particle based kinetic model ELMFIRE[1] has been developed and used in interpretation[1,2] of FT-2 and DIII-D experiments. In this work we summarize progress in Cyclone (DIII-D core) and ASDEX Upgrade pedestal region simulations, and show that in simulations the choice of adiabatic electrons results in quenching of turbulence (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  1. Recent advances in gyrokinetic full-f particle simulation of medium sized Tokamaks with ELMFIRE

    Energy Technology Data Exchange (ETDEWEB)

    Janhunen, S.J.; Kiviniemi, T.P.; Korpio, T.; Leerink, S.; Nora, M. [Helsinki University of Technology, Euratom-Tekes Association, Espoo (Finland); Heikkinen, J.A. [VTT, Euratom-Tekes Association, Espoo (Finland); Ogando, F. [Helsinki University of Technology, Euratom-Tekes Association, Espoo (Finland); Universidad Nacional de Educacion a Distancia, Madrid (Spain)

    2010-05-15

    Large-scale kinetic simulations of toroidal plasmas based on first principles are called for in studies of transition from low to high confinement mode and internal transport barrier formation in the core plasma. Such processes are best observed and diagnosed in detached plasma conditions in mid-sized tokamaks, so gyrokinetic simulations for these conditions are warranted. A first principles test-particle based kinetic model ELMFIRE[1] has been developed and used in interpretation[1,2] of FT-2 and DIII-D experiments. In this work we summarize progress in Cyclone (DIII-D core) and ASDEX Upgrade pedestal region simulations, and show that in simulations the choice of adiabatic electrons results in quenching of turbulence (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  2. Joint research using small tokamaks

    International Nuclear Information System (INIS)

    Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive coordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Coordinated Research Project, is presented

  3. Joint research using small tokamaks

    International Nuclear Information System (INIS)

    Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive co-ordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Co-ordinated Research Project is presented. (author)

  4. Radial electric field studies in the plasma edge of ASDEX upgrade

    International Nuclear Information System (INIS)

    In magnetically confined fusion plasmas, edge transport barriers (ETBs) are formed during the transition from a highly turbulent state (low confinement regime, L-mode) to a high energy confinement regime (H-mode) with reduced turbulence and transport. The performance of an H-mode fusion plasma is highly dependent on the strength of the ETB which extends typically over the outermost 5% of the confined plasma. The formation of the ETB is strongly connected to the existence of a sheared plasma flow perpendicular to the magnetic field caused by a local radial electric field Er. The gradients in Er and the accompanying E x B velocity shear play a fundamental role in edge turbulence suppression, transport barrier formation and the transition to H-mode. Thus, the interplay between macroscopic flows and transport at the plasma edge is of crucial importance to understanding plasma confinement and stability. The work presented in this thesis is based on charge exchange recombination spectroscopy (CXRS) measurements performed at the plasma edge of the ASDEX Upgrade (AUG) tokamak. During this thesis new high-resolution CXRS diagnostics were installed at the outboard and inboard miplane of AUG, which provide measurements of the temperature, density and flows of the observed species. From these measurements the radial electric field can be directly determined via the radial force balance equation. The new CXRS measurements, combined with the other edge diagnostics available at AUG, allow for an unprecedented, high-accuracy localization (2-3 mm) of the Er profile. The radial electric field has been derived from charge exchange spectra measured on different impurity species including He2+, B5+, C6+ and Ne10+. The resulting Er profiles are found to be identical within the uncertainties regardless of the impurity species used, thus demonstrating the validity of the diagnostic technique. Inside the ETB the Er profile forms a deep, negative (i.e. directed towards the plasma center

  5. Texas Experimental Tokamak

    International Nuclear Information System (INIS)

    This progress report covers the period from November 1, 1990 to April 30, 1993. During that period, TEXT was operated as a circular tokamak with a material limiter. It was devoted to the study of basic plasma physics, in particular to study of fluctuations, turbulence, and transport. The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics, specifically to conduct a research program under the following main headings: (1) to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks, in particular to understand the role of turbulence; (2) to study physics of the edge plasma, in particular the turbulence; (3) to study the physics or resonant magnetic fields (ergodic magnetic divertors, intra island pumping); and (4) to study the physics of electron cyclotron heating (ECRH). Results of studies in each of these areas are reported

  6. Microwave Tokamak Experiment

    International Nuclear Information System (INIS)

    The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. The experiment, soon to be operational, provides an opportunity to study dense plasmas heated by powers unprecedented in the electron-cyclotron frequency range required by the especially high magnetic fields used with the MTX and needed for reactors. 1 references, 5 figures, 3 tables

  7. Bulk ion heating with ICRF waves in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Mantsinen, M. J., E-mail: mervi.mantsinen@bsc.es [Catalan Institution for Research and Advanced Studies, Barcelona (Spain); Barcelona Supercomputing Center, Barcelona (Spain); Bilato, R.; Bobkov, V. V.; Kappatou, A.; McDermott, R. M.; Odstrčil, T.; Tardini, G.; Bernert, M.; Dux, R.; Maraschek, M.; Noterdaeme, J.-M.; Ryter, F.; Stober, J. [Max-Planck-Institut für Plasmaphysik, Garching (Germany); Nocente, M. [Dipartimento di Fisica “G. Occhialini”, Università degli Studi di Milano-Bicocca, Milano (Italy); Istituto di Fisica del Plasma “P. Caldirola”, CNR, Milano (Italy); Hellsten, T. [Dept. of Fusion Plasma Physics, EES, KTH, Stockholm (Sweden); Mantica, P.; Tardocchi, M. [Istituto di Fisica del Plasma “P. Caldirola”, CNR, Milano (Italy); Nielsen, S. K.; Rasmussen, J.; Stejner, M. [Technical University of Denmark, Department of Physics, Lyngby (Denmark); and others

    2015-12-10

    Heating with ICRF waves is a well-established method on present-day tokamaks and one of the heating systems foreseen for ITER. However, further work is still needed to test and optimize its performance in fusion devices with metallic high-Z plasma facing components (PFCs) in preparation of ITER and DEMO operation. This is of particular importance for the bulk ion heating capabilities of ICRF waves. Efficient bulk ion heating with the standard ITER ICRF scheme, i.e. the second harmonic heating of tritium with or without {sup 3}He minority, was demonstrated in experiments carried out in deuterium-tritium plasmas on JET and TFTR and is confirmed by ICRF modelling. This paper focuses on recent experiments with {sup 3}He minority heating for bulk ion heating on the ASDEX Upgrade (AUG) tokamak with ITER-relevant all-tungsten PFCs. An increase of 80% in the central ion temperature T{sub i} from 3 to 5.5 keV was achieved when 3 MW of ICRF power tuned to the central {sup 3}He ion cyclotron resonance was added to 4.5 MW of deuterium NBI. The radial gradient of the T{sub i} profile reached locally values up to about 50 keV/m and the normalized logarithmic ion temperature gradients R/LT{sub i} of about 20, which are unusually large for AUG plasmas. The large changes in the T{sub i} profiles were accompanied by significant changes in measured plasma toroidal rotation, plasma impurity profiles and MHD activity, which indicate concomitant changes in plasma properties with the application of ICRF waves. When the {sup 3}He concentration was increased above the optimum range for bulk ion heating, a weaker peaking of the ion temperature profile was observed, in line with theoretical expectations.

  8. High speed gas valve for massive gas injection in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Dibon, Mathias; Neu, Rudolf [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Technische Universitaet Muenchen, Boltzmannstr. 15, 85748 Garching (Germany); Herrmann, Albrecht; Mank, Klaus; Mertens, Vitus; Pautasso, Gabriella; Ploeckl, Bernhard [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany)

    2015-05-01

    For the purpose of mitigating the forces on the vessel during disruptions, a system for massive gas injection is used at the tokamak ASDEX Upgrade. Three gas valves are mounted inside the vacuum vessel with a distance of 10 cm between the nozzle exit (diameter 14 mm) and the plasma edge. This requires the valves to be insensitive to strong magnetic fields, especially on the magnetic high field side, to high temperatures and ionizing radiation. The new High Speed Gas Valve is meant to replace an old electromagnetic valve in order to increase the gas delivery efficiency and the operational reliability. The valve is closed by compressed air (20 bar) acting on through a piston on the stem and valve plate, pushing the seal onto the sealing edge and closing the gas chamber. Piezoelectric clamps secure the stem while the 80 cm{sup 3} gas chamber is filled with neon or argon at a pressure of up to 50 bar. The valve opens up when the piezoelectric actuators release the stem and a stack of disk springs accelerates the valve plate until it reaches its maximum stroke of 4.5 mm after 4 ms. This allows a mass flow rate of the gas up to 10{sup 5} (Pa*m{sup 3})/(s). A characterisation of the valve is presented in the contribution.

  9. Global energy confinement scaling for neutral-beam-heated tokamaks

    International Nuclear Information System (INIS)

    A total of 677 representative discharges from seven neutral-beam-heated tokamaks has been used to study the parametric scaling of global energy confinement time. Contributions to this data base were from ASDEX, DITE, D-III, ISX-B, PDX, PLT, and TFR, and were taken from results of gettered, L-mode type discharges. Assuming a power law dependence of tau/sub E/ on discharge parameters kappa, I/sub p/, B/sub t/, anti n/sub e/ P/sub tot/, a, and R/a, standard multiple linear regression techniques were used in two steps to determine the scaling. The results indicate that the discharges used in the study are well described by the scaling tau/sub E/ α kappa/sup 0.28/ B/sub T//sup -0.09/ I/sub p//sup 1.24/anti n/sub e//sup -0.26/ P/sub tot//sup -0.58/ a/sup 1.16/ (R/a)/sup 1.65/

  10. Global energy confinement scaling for neural-beam-heated tokamaks

    International Nuclear Information System (INIS)

    A total of 677 representative discharges from seven neutral-beam-heated tokamaks have been used to study the parametric scaling of global energy confinement time. Contributions to this data base were from Asdex, DITE, D-III, ISX-B, PDX, PLT and TFR, and were taken from results of gettered, L-mode type discharges. Assuming a power law dependence of tausub(E) on the discharge parameters kappa, Isub(p), Bsub(t), n-barsub(e)Psub(tot), a and R, standard multiple linear regression techniques were used in two steps to determine the scaling. The results indicate that the discharges used in the study are well described by the scaling tausub(E) is proportional to kappasup(0.28)Bsub(T)sup(-0.09)Isub(p)sup(1.24)n-barsub(e)sup(0.26) Psub(tot)sup(-0.58)asup(-0.49)Rsup(1.65). (author)

  11. Research using small tokamaks

    International Nuclear Information System (INIS)

    The technical reports contained in this collection of papers on research using small tokamaks fall into four main categories, i.e., (i) experimental work (heating, stability, plasma radial profiles, fluctuations and transport, confinement, ultra-low-q tokamaks, wall physics, a.o.), (ii) diagnostics (beam probes, laser scattering, X-ray tomography, laser interferometry, electron-cyclotron absorption and emission systems), (iii) theory (strong turbulence, effects of heating on stability, plasma beta limits, wave absorption, macrostability, low-q tokamak configurations and bootstrap currents, turbulent heating, stability of vortex flows, nonlinear islands growth, plasma-drift-induced anomalous transport, ergodic divertor design, a.o.), and (iv) new technical facilities (varistors applied to establish constant current and loop voltage in HT-6M), lower-hybrid-current-drive systems for HT-6B and HT-6M, radio-frequency systems for HT-6M ICR heating experimentation, and applications of fiber optics for visible and vacuum ultraviolet radiation detection as applied to tokamaks and reversed-field pinches. A total number of 51 papers are included in the collection. Refs, figs and tabs

  12. Reconnection in tokamaks

    International Nuclear Information System (INIS)

    Calculations with several different computer codes based on the resistive MHD equations have shown that (m = 1, n = 1) tearing modes in tokamak plasmas grow by magnetic reconnection. The observable behavior predicted by the codes has been confirmed in detail from the waveforms of signals from x-ray detectors and recently by x-ray tomographic imaging

  13. Advanced tokamak concepts

    NARCIS (Netherlands)

    Oomens, A. A. M.

    1996-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described and the main e

  14. Advanced tokamak concepts

    NARCIS (Netherlands)

    Oomens, A. A. M.

    1998-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described and the main e

  15. Research using small tokamaks

    International Nuclear Information System (INIS)

    This document consists of a collection of papers presented at the IAEA Technical Committee Meeting on Research Using Small Tokamaks. It contains 22 papers on a wide variety of research aspects, including diagnostics, design, transport, equilibrium, stability, and confinement. Some of these papers are devoted to other concepts (stellarators, compact tori). Refs, figs and tabs

  16. Sawtooth phenomena in tokamaks

    International Nuclear Information System (INIS)

    A review of experimental and theoretical investigaions of sawtooth phenomena in tokamaks is presented. Different types of sawtooth oscillations, scaling laws and methods of interanl disruption stabilization are described. Theoretical models of the sawtooth instability are discussed. 122 refs.; 4 tabs

  17. IFS Numerical Laboratory Tokamak

    International Nuclear Information System (INIS)

    A numerical laboratory of a tokamak plasma is being developed. This consists of the backbone (the overall manager in terms of the MPPL programming language), and the modularized components that can be plugged in or out for a particular run and their hierarchical arrangement. The components include various metrics for overall geometry various dynamics, field calculations, and diagnoses. 2 refs

  18. Energy and particle control characteristics of the ASDEX Upgrade 'LYRA' divertor

    International Nuclear Information System (INIS)

    In 1997 the new 'LYRA' divertor went into operation at ASDEX Upgrade and the neutral beam heating power was increased to 20 MW by installation of a second injector. This leads to the relatively high value of P/R of 12 MW/m. It has been shown that the ASDEX Upgrade LYRA divertor is capable of handling such high heating powers. Measurements presented in this paper reveal a reduction of the maximum heat flux in the LYRA divertor by more than a factor of two compared to the open Divertor I. This reduction is caused by radiative losses inside the divertor region. Carbon radiation cools the divertor plasma down to a few eV where hydrogen radiation losses become significant. They are increased due to an effective reflection of neutrals into the hot separatrix region. B2-Eirene modelling of the performed experiments supports the experimental findings and refines the understanding of loss processes in the divertor region. (and others)

  19. Tungsten erosion and redeposition in the all-tungsten divertor of ASDEX Upgrade

    International Nuclear Information System (INIS)

    Net erosion and deposition of tungsten (W) in the ASDEX Upgrade divertor were determined after the 2007 campaign by using thin W marker stripes. ASDEX Upgrade had full-W plasma-facing components during this campaign. The inner divertor and the roof baffle were net W deposition areas with a maximum deposition of about 1x1018 W-atoms cm-2 in the private flux region below the inner strike point. Net erosion of W was observed in the whole outer divertor, with the largest erosion close to the outer strike point. Only a small fraction of the W eroded in the main chamber and in the outer divertor was found in redeposits in the inner divertor, while a large fraction was either redeposited at unidentified places in the main chamber or has formed dust.

  20. Tungsten erosion and redeposition in the all-tungsten divertor of ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Mayer, M; Krieger, K; Matern, G; Neu, R; Rasinski, M; Rohde, V; Sugiyama, K; Wiltner, A [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Boltzmannstrasse 2, 85748 Garching (Germany); Andrzejczuk, M; Fortuna-Zalesna, E; Kurzydlowski, K J; Zielinski, W [Faculty of Materials Science and Engineering, Warsaw University of Technology, Association EURATOM-IPPLM, 02-507 Warsaw (Poland); Hakola, A; Koivuranta, S; Likonen, J [VTT Materials for Power Engineering, EURATOM Association, PO Box 1000, FI-02044 VTT (Finland); Ramos, G [CICATA-Qro, Instituto Politecnico Nacional, Queretaro (Mexico); Dux, R, E-mail: matej.mayer@ipp.mpg.de

    2009-12-15

    Net erosion and deposition of tungsten (W) in the ASDEX Upgrade divertor were determined after the 2007 campaign by using thin W marker stripes. ASDEX Upgrade had full-W plasma-facing components during this campaign. The inner divertor and the roof baffle were net W deposition areas with a maximum deposition of about 1x10{sup 18} W-atoms cm{sup -2} in the private flux region below the inner strike point. Net erosion of W was observed in the whole outer divertor, with the largest erosion close to the outer strike point. Only a small fraction of the W eroded in the main chamber and in the outer divertor was found in redeposits in the inner divertor, while a large fraction was either redeposited at unidentified places in the main chamber or has formed dust.

  1. Update on the ASDEX Upgrade data acquisition and data management environment

    International Nuclear Information System (INIS)

    Highlights: • An exponential growth of data amount was managed over more than twenty years of experiment operation. • Continuous adaptation of the diagnostic software and configuration keeps track with actual experiment demands. • A great number of distributed, varying diagnostics is centrally managed. - Abstract: It has been a while since it had been reported on the status of ASDEX Upgrade data acquisition (DAQ) and data management environment. An update on changes, expansions, and enhancements applied in the last years will be given. The acquired amount of data per shot increased from 4 GiB to 40 GiB in eight years. Network, storage, and archive challenges have been managed by stepwise improvements. New DAQ techniques have been introduced to replace outdated technologies. Real-time diagnostics speed-up data provisioning and contribute to feedback control. Information technology applied to ASDEX Upgrade is under permanent change. Recent and future steps are outlined

  2. Tokamaks (Second Edition)

    International Nuclear Information System (INIS)

    The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

  3. Transport in gyrokinetic tokamaks

    International Nuclear Information System (INIS)

    A comprehensive study of transport in full-volume gyrokinetic (gk) simulations of ion temperature gradient driven turbulence in core tokamak plasmas is presented. Though this ''gyrokinetic tokamak'' is much simpler than experimental tokamaks, such simplicity is an asset, because a dependable nonlinear transport theory for such systems should be more attainable. Toward this end, we pursue two related lines of inquiry. (1) We study the scalings of gk tokamaks with respect to important system parameters. In contrast to real machines, the scalings of larger gk systems (a/ρs approx-gt 64) with minor radius, with current, and with a/ρs are roughly consistent with the approximate theoretical expectations for electrostatic turbulent transport which exist as yet. Smaller systems manifest quite different scalings, which aids in interpreting differing mass-scaling results in other work. (2) With the goal of developing a first-principles theory of gk transport, we use the gk data to infer the underlying transport physics. The data indicate that, of the many modes k present in the simulation, only a modest number (Nk ∼ 10) of k dominate the transport, and for each, only a handful (Np ∼ 5) of couplings to other modes p appear to be significant, implying that the essential transport physics may be described by a far simpler system than would have been expected on the basis of earlier nonlinear theory alone. Part of this analysis is the inference of the coupling coefficients Mkpq governing the nonlinear mode interactions, whose measurement from tokamak simulation data is presented here for the first time

  4. A Probe Head for Simultaneous Measurements of Electrostatic and Magnetic Fluctuations in ASDEX Upgrade Edge Plasma

    DEFF Research Database (Denmark)

    Schrittwieser, R W; Ionita, C; Vianello, N;

    2010-01-01

    For ASDEX Upgrade (AUG) a new probe head was developed for simultaneous measurements of electric and magnetic fluctuations in the edge plasma region. The probe head consists of a cylindrical graphite case. On the front side six graphite pins are mounted. With this arrangement the poloidal and...... is inserted up to three times for 100 ms each by the midplane manipulator into the scrape-off layer. © 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim....

  5. Plasma rotation and ion temperature measurements by collective Thomson scattering at ASDEX Upgrade

    DEFF Research Database (Denmark)

    Stejner Pedersen, Morten; Nielsen, Stefan Kragh; Jacobsen, Asger Schou; Korsholm, Søren Bang; Leipold, Frank; McDermott, R. M.; Michelsen, Poul; Rasmussen, Jesper; Salewski, Mirko; Schubert, Martin; Stober, J.; Wagner, D. H.

    2015-01-01

    We present the first deuterium ion temperature and rotation measurements by collective Thomson scattering at ASDEX Upgrade. The results are in general agreement with boron-based charge exchange recombination spectroscopy measurements and consistent with neoclassical simulations for the plasma...... scenario studied here. This demonstration opens the prospect for direct non-perturbative measurements of the properties of the main ion species in the plasma core with applications in plasma transport and confinement studies....

  6. Characterization of dust collected after plasma operation of all-tungsten ASDEX Upgrade

    International Nuclear Information System (INIS)

    This work presents results of the characterization of dust collected in ASDEX Upgrade, with special emphasis on the size, morphology, structure and composition of the dust particles. The dust particles were collected after the 2009 campaign using the filtered vacuum technique. The structure and composition of the particles were examined by scanning electron microscopy combined with energy-dispersive x-ray spectroscopy, focused ion beam and scanning transmission electron microscopy with special interest in the tungsten particles. (paper)

  7. Improved Collective Thomson Scattering measurements of fast ions at ASDEX Upgrade

    DEFF Research Database (Denmark)

    Rasmussen, Jesper; Nielsen, Stefan Kragh; Stejner Pedersen, Morten;

    2014-01-01

    Understanding the behaviour of the confined fast ions is important in both current and future fusion experiments. These ions play a key role in heating the plasma and will be crucial for achieving conditions for burning plasma in next-step fusion devices. Microwave-based Collective Thomson Scatte...... ASDEX Upgrade are now feasible. The new background subtraction technique could be important for the design of CTS systems in other fusion experiments....

  8. Radial transport of poloidal momentum in ASDEX Upgrade in L-mode and H-mode

    DEFF Research Database (Denmark)

    Mehlmann, F.; Schrittwieser, R.; Naulin, Volker;

    2012-01-01

    A reciprocating probe was used for localized measurements of the radial transport of poloidal momentum in the scrape-off layer (SOL) of ASDEX Upgrade (AUG). The probe measured poloidal and radial electric field components and density. We concentrate on three components of the momentum transport: ......: Reynolds stress, convective momentum flux and triple product of the fluctuating components of density, radial and poloidal electric field. For the evaluation we draw mainly on the probability density functions (PDFs)....

  9. Tungsten transport in the plasma edge at ASDEX upgrade

    International Nuclear Information System (INIS)

    The Plasma Facing Components (PFC) will play a crucial role in future deuterium-tritium magnetically confined fusion power plants, since they will be subject to high energy and particle loads, but at the same time have to ensure long lifetimes and a low tritium retention. These requirements will most probably necessitate the use of high-Z materials such as tungsten for the wall materials, since their erosion properties are very benign and, unlike carbon, capture only little tritium. The drawback with high-Z materials is, that they emit strong line radiation in the core plasma, which acts as a powerful energy loss mechanism. Thus, the concentration of these high-Z materials has to be controlled and kept at low levels in order to achieve a burning plasma. Understanding the transport processes in the plasma edge is essential for applying the proper impurity control mechanisms. This control can be exerted either by enhancing the outflux, e.g. by Edge Localized Modes (ELM), since they are known to expel impurities from the main plasma, or by reducing the influx, e.g. minimizing the tungsten erosion or increasing the shielding effect of the Scrape Off Layer (SOL). ASDEX Upgrade (AUG) has been successfully operating with a full tungsten wall for several years now and offers the possibility to investigate these edge transport processes for tungsten. This study focused on the disentanglement of the frequency of type-I ELMs and the main chamber gas injection rate, two parameters which are usually linked in H-mode discharges. Such a separation allowed for the first time the direct assessment of the impact of each parameter on the tungsten concentration. The control of the ELM frequency was performed by adjusting the shape of the plasma, i.e. the upper triangularity. The radial tungsten transport was investigated by implementing a modulated tungsten source. To create this modulated source, the linear dependence of the tungsten erosion rate at the Ion Cyclotron Resonance

  10. Tungsten transport in the plasma edge at ASDEX upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Janzer, Michael Arthur

    2015-04-30

    The Plasma Facing Components (PFC) will play a crucial role in future deuterium-tritium magnetically confined fusion power plants, since they will be subject to high energy and particle loads, but at the same time have to ensure long lifetimes and a low tritium retention. These requirements will most probably necessitate the use of high-Z materials such as tungsten for the wall materials, since their erosion properties are very benign and, unlike carbon, capture only little tritium. The drawback with high-Z materials is, that they emit strong line radiation in the core plasma, which acts as a powerful energy loss mechanism. Thus, the concentration of these high-Z materials has to be controlled and kept at low levels in order to achieve a burning plasma. Understanding the transport processes in the plasma edge is essential for applying the proper impurity control mechanisms. This control can be exerted either by enhancing the outflux, e.g. by Edge Localized Modes (ELM), since they are known to expel impurities from the main plasma, or by reducing the influx, e.g. minimizing the tungsten erosion or increasing the shielding effect of the Scrape Off Layer (SOL). ASDEX Upgrade (AUG) has been successfully operating with a full tungsten wall for several years now and offers the possibility to investigate these edge transport processes for tungsten. This study focused on the disentanglement of the frequency of type-I ELMs and the main chamber gas injection rate, two parameters which are usually linked in H-mode discharges. Such a separation allowed for the first time the direct assessment of the impact of each parameter on the tungsten concentration. The control of the ELM frequency was performed by adjusting the shape of the plasma, i.e. the upper triangularity. The radial tungsten transport was investigated by implementing a modulated tungsten source. To create this modulated source, the linear dependence of the tungsten erosion rate at the Ion Cyclotron Resonance

  11. Modelling of carbon transport in the outer divertor plasma of ASDEX upgrade

    International Nuclear Information System (INIS)

    Carbon transport in the ASDEX Upgrade outer divertor plasma is investigated in numerical simulations. The SOLPS5.0 code package is used to model the scrape-off layer plasma in a set of repeated lower-single-null L-mode discharges. Special emphasis is given to replicate the plasma conditions measured in the full tungsten, vertical outer target of ASDEX Upgrade, and solutions with and without the effect of drifts are presented. First ERO simulations of hydrocarbon transport in a SOLPS plasma background including drifts are carried out, and significantly closer match to the experimental 13C deposition pattern is obtained than with the solution without drifts. The 2D divertor electric field predicted by SOLPS is applied to the ERO modelling, and it is observed to result in a poloidal hydrocarbon drift that agrees well with the experiment. An increased carbon deposition efficiency, particularly upstream from the source, is obtained in the normal ASDEX Upgrade field configuration (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  12. The ASDEX upgrade digital video processing system for real-time machine protection

    International Nuclear Information System (INIS)

    Highlights: • We present the Real-Time Video diagnostic system of ASDEX Upgrade. • We show the implemented image processing algorithms for machine protection. • The way to achieve a robust operating multi-threading Real-Time system is described. -- Abstract: This paper describes the design, implementation, and operation of the Video Real-Time (VRT) diagnostic system of the ASDEX Upgrade plasma experiment and its integration with the ASDEX Upgrade Discharge Control System (DCS). Hot spots produced by heating systems erroneously or accidentally hitting the vessel walls, or from objects in the vessel reaching into the plasma outer border, show up as bright areas in the videos during and after the reaction. A system to prevent damage to the machine by allowing for intervention in a running discharge of the experiment was proposed and implemented. The VRT was implemented on a multi-core real-time Linux system. Up to 16 analog video channels (color and b/w) are acquired and multiple regions of interest (ROI) are processed on each video frame. Detected critical states can be used to initiate appropriate reactions – e.g. gracefully terminate the discharge. The system has been in routine operation since 2007

  13. Direct observation of current in type-I edge-localized-mode filaments on the ASDEX upgrade tokamak

    DEFF Research Database (Denmark)

    Vianello, N.; Zuin, M.; Cavazzana, R.;

    2011-01-01

    Magnetically confined plasmas in the high confinement regime are regularly subjected to relaxation oscillations, termed edge localized modes (ELMs), leading to large transport events. Present ELM theories rely on a combined effect of edge current and the edge pressure gradients which result...... in intermediate mode number (n≅10-15) structures (filaments) localized in the perpendicular plane and extended along the field lines. By detailed localized measurements of the magnetic field perturbation associated to type-I ELM filaments, it is shown that these filaments carry a substantial current. © 2011...

  14. Large Aspect Ratio Tokamak Study

    International Nuclear Information System (INIS)

    The Large Aspect Ratio Tokamak Study (LARTS) at Oak Ridge National Laboratory (ORNL) investigated the potential for producing a viable longburn tokamak reactor by enhancing the volt-second capability of the ohmic heating transformer through the use of high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were assessed in the context of extended burn operation. Using a one-dimensional transport code plasma startup and burn parameters were addressed. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the startup and shutdown portions of the tokamak cycle. A representative large aspect ratio tokamak with an aspect ratio of 8 was found to achieve a burn time of 3.5 h at capital cost only approx. 25% greater than that of a moderate aspect ratio design tokamak

  15. Next tokamak facility

    International Nuclear Information System (INIS)

    Design studies on a superconducting, long-pulse, current-driven, ignited tokamak, called the Toroidal Fusion Core Demonstration (TFCD), are being conducted by the Fusion Engineering Design Center (FEDC) and Princeton Plasma Physics Laboratory (PPPL) with additional broad community involvement. Options include the use of all-superconducting toroidal field (TF) coils, a superconducting-copper hybrid arrangement of TF coils, or all-copper TF coils. Only the first two options have been considered to date. The general feasibility of these approaches has been established with the goal of high performance (ignition, approx. 390 MW; wall loading approx. 2.2 MW/m2) at minimum capital cost. The preconceptual effort will be completed in early FY 1984 and a selection made from the indicated options. The TFCD is judged to represent a reasonable necessary step between the Tokamak Fusion Test Reactor (TFTR) and the Engineering Test Reactor

  16. Tokamak fusion reactor exhaust

    International Nuclear Information System (INIS)

    This report presents a compilation of papers dealing with reactor exhaust which were produced as part of the TIGER Tokamak Installation for Generating Electricity study at Culham. The papers are entitled: (1) Exhaust impurity control and refuelling. (2) Consideration of the physical problems of a self-consistent exhaust and divertor system for a long burn Tokamak. (3) Possible bundle divertors for INTOR and TIGER. (4) Consideration of various magnetic divertor configurations for INTOR and TIGER. (5) A appraisal of divertor experiments. (6) Hybrid divertors on INTOR. (7) Refuelling and the scrape-off layer of INTOR. (8) Simple modelling of the scrape-off layer. (9) Power flow in the scrape-off layer. (10) A model of particle transport within the scrape-off plasma and divertor. (11) Controlled recirculation of exhaust gas from the divertor into the scrape-off plasma. (U.K.)

  17. Tritium catalyzed deuterium tokamaks

    International Nuclear Information System (INIS)

    A preliminary assessment of the promise of the Tritium Catalyzed Deuterium (TCD) tokamak power reactors relative to that of deuterium-tritium (D-T) and catalyzed deuterium (Cat-D) tokamaks is undertaken. The TCD mode of operation is arrived at by converting the 3He from the D(D,n)3He reaction into tritium, by neutron capture in the blanket; the tritium thus produced is fed into the plasma. There are three main parts to the assessment: blanket study, reactor design and economic analysis and an assessment of the prospects for improvements in the performance of TCD reactors (and in the promise of the TCD mode of operation, in general)

  18. Tokamak pump limiters

    International Nuclear Information System (INIS)

    Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6MW of auxiliary neutral beam heating. Experiments have also been done with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a region may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this Z-mode of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described

  19. [High beta tokamak research

    International Nuclear Information System (INIS)

    Our activities on High Beta Tokamak Research during the past 20 months of the present grant period can be divided into six areas: reconstruction and modeling of high beta equilibria in HBT; measurement and analysis of MHD instabilities observed in HBT; measurements of impurity transport; diagnostic development on HBT; numerical parameterization of the second stability regime; and conceptual design and assembly of HBT-EP. Each of these is described in some detail in the sections of this progress report

  20. Validation of transport models in ASDEX Upgrade current ramps

    International Nuclear Information System (INIS)

    In order to prepare adequate ramp up and down scenarios for ITER, understanding the physics of transport during the current ramps is essential. The aim of the work was to assess the capability of several transport models to reproduce the experimental data during the current ramps. For this purpose, the calculated temperature profiles from different transport models, i.e. Coppi-Tang, Neo-Alcator, Bohm-Gyrobohm, critical gradient model and H98/2 scaling-based are compared to experimental temperature profiles under different conditions. The strong variation of the experimental electron temperature profiles are partly reproduced by the models. The importance of central and edge radiation will be emphasized, as well as the main transport properties of the models, especially in the case of strong local electron heating (ECRH). To investigate the control capabilities of a Tokamak, particularly with regard to ITER, the impact on global plasma parameters like the internal inductance and the stored energy is also investigated.

  1. SUPPRESSION OF TEARING MODES BY MEANS OF LOCALIZED ELECTRON CYCLOTRON CURRENT DRIVE IN THE DIII-D TOKAMAK

    International Nuclear Information System (INIS)

    The onset of tearing modes and the resulting negative effects on plasma performance set significant limits on the operational domain of tokamaks. Modes with toroidal mode number (n) larger than two cause only a minor reduction in energy confinement (<10%). Modes which have a dominant poloidal mode number (m) of three and n=2 lead to a significant reduction in confinement (<30%) at fixed power. The plasma pressure β (normalized to the magnetic field pressure) can be raised further, albeit with very small incremental confinement. Pushing to higher β often destabilizes the m=2/n=1 tearing mode which can lock to the wall and lead to a complete and rapid disruption of the plasma with potentially serious consequences for the tokamak. The β values at which these modes usually appear in conventional tokamak discharges are well below the limits calculated using ideal MHD theory. Therefore, the tearing modes can set effective upper limits on energy confinement and pressure. Significant progress has been made in stabilizing these modes by local current generation using electron cyclotron waves. The tearing mode is essentially a deficit in current flowing helically, resonant with the spatial structure of the local magnetic field. This forms an ''island'' where the magnetic flux is no longer monotonic. It was predicted theoretically [1,2] that replacement of this ''missing'' current would return the plasma to the state prior to the instability. Experiments on the ASDEX-Upgrade [3], JT-60U [4], and DIII-D [5] tokamaks have demonstrated stabilization of m=3/n=2 modes using electron cyclotron current drive (ECCD) to replace the current in the island. Following these initial experiments, recent work on the DIII-D tokamak has demonstrated two significant advances in application of this technique--extending the operational domain stable to m=3/n=2 modes to higher β and the first suppression of the more dangerous m=2/n=1 mode

  2. Status of and prospects for advanced Tokamak regimes from multi-machine comparisons

    International Nuclear Information System (INIS)

    In this series of 21 slides the author presents an assessment of the present fusion performance of the advanced tokamaks (AT) regimes for non-inductive operation. These AT regimes include data from ASDEX Upgrade, DIII-D, FT-U, JET, JT-60U and Tore-Supra. Only data from both the 'hybrid' without necessarily an ITB (internal transport barrier) or the 'steady-state' scenario have been considered because these scenarios are the 2 candidates for the ITER non inductive current drive operation. A new operational diagram is proposed: the figure of merit for fusion performance and confinement H(ITER-89P).βN/q295 versus the bootstrap current fraction e1/2.βP. In this diagram there is a continuous progression from the 'inductive' to the 'hybrid' and 'steady-state' tokamak operating mode. The following range of performance: H(ITER-89P).βN/q295 ∼ 0.3-0.4 at βP ∼ 1, q95 ∼ 5, is expected for Q = 5 non inductive current drive operation for ITER. Fusion performances tend to decrease with the pulse duration, so extending the plasma performances achieved on a short time scale requires operating safely far from the operational limits. Other conclusions concerning the operating domain of dimensionless parameters such as Larmor radius, collisionality, Mach number and ratio of ion to electron temperature are also presented. (A.C.)

  3. Mode coupling structure in Tokamaks

    International Nuclear Information System (INIS)

    A m=1, helically displaced current channel was identified in the ASDEX plasma interior during m=2 mode activity. This was achieved by means of simultaneous data obtained from a new gradient sensitive schlieren diagnostic and BP measurements. They clearly show a rotational-transform-dependent coupling mechanism between the driver m=1 current helix and the m=2 perturbation of the bulk current surrounding it. The mechanism is of central importance for the development of the instability and for the theoretical understanding of mode coupling, mode locking and other varieties of mode structures in plasma. (orig.)

  4. Magnetic confinement experiment -- 1: Tokamaks

    International Nuclear Information System (INIS)

    This report reviews presentations made at the 15th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion on experimental tokamak physics, particularly on advances in core plasma physics, divertor and edge physics, heating and current drive, and tokamak concept optimization

  5. Polarization spectroscopy of tokamak plasmas

    International Nuclear Information System (INIS)

    Measurements of polarization of spectral lines emitted by tokamak plasmas provide information about the plasma internal magnetic field and the current density profile. The methods of polarization spectroscopy, as applied to the tokamak diagnostic, are reviewed with emphasis on the polarimetry of motional Stark effect in hydrogenic neutral beam emissions. 25 refs., 7 figs

  6. Outer divertor of ASDEX Upgrade in low-density L-mode discharges in forward and reversed magnetic field: II. Analysis of local impurity migration

    International Nuclear Information System (INIS)

    Part I (Aho-Mantila L. et al 2012 Nucl. Fusion 52 103006) presented a detailed analysis of outer divertor plasma conditions in low-density L-mode discharges in ASDEX Upgrade. In this paper, we analyse the local migration of carbon that originates from 13CH4 injected into these plasmas from the vertical outer target. Notable changes are observed in the local carbon deposition patterns when reversing the magnetic field in the experiments. Kinetic impurity-following simulations are performed using the 3D ERO code package with 2D background plasma solutions calculated with the SOLPS5.0 code package. The modelling shows that the measured changes are due to the changes in plasma collisionality, dissociation and ionization rates, and E × B drift of the impurities. These conditions affect the direction and rate of impurity migration inside and out of the divertor, having wider consequences on the global migration of impurities in a divertor tokamak. It is further shown that the migration pathways are largely determined by carbon ions and, hence, relevant for impurities in general. Neutral carbon and hydrocarbons are deposited only in the near vicinity of the injection, where they affect the local re-deposition efficiency. In this limited region, a perturbation of the local plasma conditions by the methane puff appears likely, yielding a significant uncertainty for interpreting the deposition efficiencies. The local deposition is largely influenced by the magnetic presheath electric field, the structure of which is the main uncertainty in the SOLPS5.0-ERO simulations. (paper)

  7. Radial electric field studies in the plasma edge of ASDEX upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Viezzer, Eleonora

    2012-12-18

    In magnetically confined fusion plasmas, edge transport barriers (ETBs) are formed during the transition from a highly turbulent state (low confinement regime, L-mode) to a high energy confinement regime (H-mode) with reduced turbulence and transport. The performance of an H-mode fusion plasma is highly dependent on the strength of the ETB which extends typically over the outermost 5% of the confined plasma. The formation of the ETB is strongly connected to the existence of a sheared plasma flow perpendicular to the magnetic field caused by a local radial electric field E{sub r}. The gradients in E{sub r} and the accompanying E x B velocity shear play a fundamental role in edge turbulence suppression, transport barrier formation and the transition to H-mode. Thus, the interplay between macroscopic flows and transport at the plasma edge is of crucial importance to understanding plasma confinement and stability. The work presented in this thesis is based on charge exchange recombination spectroscopy (CXRS) measurements performed at the plasma edge of the ASDEX Upgrade (AUG) tokamak. During this thesis new high-resolution CXRS diagnostics were installed at the outboard and inboard miplane of AUG, which provide measurements of the temperature, density and flows of the observed species. From these measurements the radial electric field can be directly determined via the radial force balance equation. The new CXRS measurements, combined with the other edge diagnostics available at AUG, allow for an unprecedented, high-accuracy localization (2-3 mm) of the E{sub r} profile. The radial electric field has been derived from charge exchange spectra measured on different impurity species including He{sup 2+}, B{sup 5+}, C{sup 6+} and Ne{sup 10+}. The resulting E{sub r} profiles are found to be identical within the uncertainties regardless of the impurity species used, thus demonstrating the validity of the diagnostic technique. Inside the ETB the E{sub r} profile forms a deep

  8. The tokamak as a neutron source

    International Nuclear Information System (INIS)

    This paper describes the tokamak in its role as a neutron source, with emphasis on experimental results for D-D neutron production. The sections summarize tokamak operation, sources of fusion and non-fusion neutrons, principal neutron detection methods and their calibration, neutron energy spectra and fluxes outside the tokamak plasma chamber, history of neutron production in tokamaks, neutron emission and fusion power gain from JET and TFTR (the largest present-day tokamaks), and D-T neutron production from burnup of D-D tritons. This paper also discusses the prospects for future tokamak neutron production and potential applications of tokamak neutron sources. 100 refs., 16 figs., 4 tabs

  9. Tore Supra tokamak

    International Nuclear Information System (INIS)

    This part of the electricity uses chapter of the Engineers Techniques collection is entirely devoted to the technical description of Tore Supra tokamak. A thermonuclear fusion device with magnetic confinement control such as Tore Supra concentrates a huge amount of high power electro-technical and electronic equipments. These power systems play a major role and are sometimes boosted to their extreme limits. From these equipments we can find: big superconducting magnets, big cooled copper magnets, high-voltage power supplies with thyristors (320 MVA installed), several MW hyper-frequency sources, several MW accelerated atom injectors, cryogenic, heat extraction, high-vacuum pumping systems, etc.. The components developed for these applications are numerous and frequently original: superconductor for variable magnetic field, DC static circuit breaker with high switch-off capability (0.7 GVA), 2 MW tetrodes, 500 kW klystrons, 500 kW gyrotrons, very low temperature (3 deg. K) electromechanical pumps, etc.. Tore Supra is a good example of the various applications of electricity and a testimony of the constant progress of the techniques mastered by electricians. This chapter is divided in 5 parts. Part 1 gives some general informations about thermonuclear fusion research, tokamak principles and electrotechnical systems of fusion research devices. Part 2 describes the Tore Supra tokamak, its aims and specificities, its internal components, the poloidal field system and the plasma heating systems. Part 3 concerns the power pulse sources: distribution network, poloidal field power supply, plasma heating systems, and ergodic divertor power supply. Part 4 describes the permanent electric power supplies for the auxiliary systems: toroidal field, cryogenic installation, cooling-drying loops. The last chapter briefly summarizes the perspectives of nuclear fusion research. (J.S.)

  10. Tokamak burn control

    International Nuclear Information System (INIS)

    Research of the fusion plasma thermal instability and its control is reviewed. General models of the thermonuclear plasma are developed. Techniques of stability analysis commonly employed in burn control research are discussed. Methods for controlling the plasma against the thermal instability are reviewed. Emphasis is placed on applications to tokamak confinement concepts. Additional research which extends the results of previous research is suggested. Issues specific to the development of control strategies for mid-term engineering test reactors are identified and addressed. 100 refs., 24 figs., 10 tabs

  11. Maximum entropy tokamak configurations

    International Nuclear Information System (INIS)

    The new entropy concept for the collective magnetic equilibria is applied to the description of the states of a tokamak subject to ohmic and auxiliary heating. The condition for the existence of steady state plasma states with vanishing entropy production implies, on one hand, the resilience of specific current density profiles and, on the other, severe restrictions on the scaling of the confinement time with power and current. These restrictions are consistent with Goldston scaling and with the existence of a heat pinch. (author)

  12. Understanding disruptions in tokamaks

    International Nuclear Information System (INIS)

    This paper describes progress achieved since 2007 in understanding disruptions in tokamaks, when the effect of plasma current sharing with the wall was introduced into theory. As a result, the toroidal asymmetry of the plasma current measurements during vertical disruption event (VDE) on the Joint European Torus was explained. A new kind of plasma equilibria and mode coupling was introduced into theory, which can explain the duration of the external kink 1/1 mode during VDE. The paper presents first results of numerical simulations using a free boundary plasma model, relevant to disruptions.

  13. Tokamak instrumentation and controls

    Energy Technology Data Exchange (ETDEWEB)

    Becraft, W. R.; Bettis, E. S.; Houlberg, W. A.; Onega, R. J.; Stone, R. S.

    1979-02-01

    The three areas of study emphasis to date are: (1) Physics implications for controls, (2) Computer simulation, and (3) Shutdown/aborts. This document reports on the FY 78 efforts (the first year of these studies) to address these problems. Transient scenario options for the startup of a tokamak are developed, and the implications for the control system are discussed. This document also presents a hybrid computer simulation (analog and digital) of the Impurity Study Experiment (ISX-B) which is now being used for corroborative controls investigations. The simulation will be expanded to represent a TNS/ETF machine.

  14. Demonstration tokamak power plant

    Energy Technology Data Exchange (ETDEWEB)

    Abdou, M.; Baker, C.; Brooks, J.; Ehst, D.; Mattas, R.; Smith, D.L.; DeFreece, D.; Morgan, G.D.; Trachsel, C.

    1983-01-01

    A conceptual design for a tokamak demonstration power plant (DEMO) was developed. A large part of the study focused on examining the key issues and identifying the R and D needs for: (1) current drive for steady-state operation, (2) impurity control and exhaust, (3) tritium breeding blanket, and (4) reactor configuration and maintenance. Impurity control and exhaust will not be covered in this paper but is discussed in another paper in these proceedings, entitled Key Issues of FED/INTOR Impurity Control System.

  15. ELM behaviour and linear MHD stability of edge ECRH heated ASDEX Upgrade plasmas

    Science.gov (United States)

    Burckhart, A.; Dunne, M.; Wolfrum, E.; Fischer, R.; McDermott, R.; Viezzer, E.; Willensdorfer, M.; the ASDEX Upgrade Team

    2016-05-01

    In order to test the peeling–ballooning ELM model, ECRH heating was applied to the edge of ASDEX Upgrade type-I ELMy H-mode plasmas to alter the pedestal pressure and current density profiles. The discharges were analysed with respect to ideal MHD stability. While the ELM frequency increased and the pedestal gradients relaxed with edge ECRH, the MHD stability boundary did not change. The results indicate that the peeling–ballooning model is insufficient to fully explain the triggering of ELM instabilities in the presence of edge ECRH heating.

  16. Magnetic diagnostic of SOL-filaments generated by type I ELMs on JET and ASDEX Upgrade

    DEFF Research Database (Denmark)

    Naulin, Volker; Vianello, N.; Schrittwieser, R.;

    2011-01-01

    This contribution is focused on the magnetic signatures of type I ELM filaments. On JET a limited number of high time resolution magnetic coils were used to derive essential ELM filament parameters. The method uses forward modelling and simultaneous fitting of magnetic pickup coil signals to a...... simple model, motivated by observations. A new diagnostic in the form of a reciprocating probe with three magnetic pickup loops was developed for ASDEX Upgrade (AUG). Measurements during the passage of type-I ELM filaments determine the filaments to be in the scrape off layer (SOL) and to carry currents...

  17. Commissioning of inline ECE system within waveguide based ECRH transmission systems on ASDEX upgrade

    DEFF Research Database (Denmark)

    Bongers, W. A.; Kasparek, W.; Doelman, N.;

    2012-01-01

    A CW capable inline electron cyclotron emission (ECE) separation system for feedback control, featuring oversized corrugated waveguides, is commissioned on ASDEX upgrade (AUG). The system is based on a combination of a polarization independent, non-resonant, Mach-Zehnder diplexer equipped with...... dielectric plate beam splitters [2, 3] employed as corrugated oversized waveguide filter, and a resonant Fast Directional Switch, FADIS [4, 5, 6, 7] as ECE/ECCD separation system. This paper presents an overview of the system, the low power characterisation tests and first high power commissioning on AUG....

  18. Commissioning of inline ECE system within waveguide based ECRH transmission systems on ASDEX upgrade

    Directory of Open Access Journals (Sweden)

    Donné A.J.H.

    2012-09-01

    Full Text Available A CW capable inline electron cyclotron emission (ECE separation system for feedback control, featuring oversized corrugated waveguides, is commissioned on ASDEX upgrade (AUG. The system is based on a combination of a polarization independent, non-resonant, Mach-Zehnder diplexer equipped with dielectric plate beam splitters [2, 3] employed as corrugated oversized waveguide filter, and a resonant Fast Directional Switch, FADIS [4, 5, 6, 7] as ECE/ECCD separation system. This paper presents an overview of the system, the low power characterisation tests and first high power commissioning on AUG.

  19. Effect of 3D magnetic perturbations on the plasma rotation in ASDEX Upgrade

    Science.gov (United States)

    Martitsch, A. F.; Kasilov, S. V.; Kernbichler, W.; Kapper, G.; Albert, C. G.; Heyn, M. F.; Smith, H. M.; Strumberger, E.; Fietz, S.; Suttrop, W.; Landreman, M.; The ASDEX Upgrade Team; the EUROfusion MST1 Team

    2016-07-01

    The toroidal torque due to the non-resonant interaction with external magnetic perturbations (TF ripple and perturbations from ELM mitigation coils) in ASDEX Upgrade is modelled with help of the NEO-2 and SFINCS codes and compared to semi-analytical models. It is shown that almost all non-axisymmetric transport regimes contributing to neoclassical toroidal viscosity (NTV) are realized within a single discharge at different radial positions. The NTV torque is obtained to be roughly a quarter of the NBI torque. This indicates the presence of other important momentum sources. The role of these momentum sources and possible integral torque balance measurements are briefly discussed.

  20. Particle influx measurements with the ASDEX-upgrade multichord visible spectroscopy system

    International Nuclear Information System (INIS)

    This report describes the hardware and software components of the ASDEX-Upgrade multichord visible spectroscopy system. Main emphasis is laid on a detailed description of the detector, a free programmable charge-coupled device intensified by a microchannel plate. As an experimental application, flux measurements of different impurity species from the inner heat shield are presented. Poloidal profiles of the released impurity amount obtained for various experimental situations are used to check the plasma position which is derived by the function parametrization analysis. (orig.)

  1. Destabilization of fast particle stabilized sawteeth in ASDEX Upgrade with electron cyclotron current drive

    DEFF Research Database (Denmark)

    Igochine, V.; Chapman, I.T.; Bobkov, V.; Günter, S.; Maraschek, M.; Moseev, Dmitry; Pereversev, G.; Reich, M.; Stober, J.

    2011-01-01

    It is often observed that large sawteeth trigger the neoclassical tearing mode well below the usual threshold for this instability. At the same time, fast particles in the plasma core stabilize sawteeth and provide these large crashes. The paper presents results of first experiments in ASDEX...... Upgrade for destabilization of fast particle stabilized sawteeth with electron cyclotron current drive (ECCD). It is shown that moderate ECCD from a single gyrotron is able to destabilize the fast particle stabilized sawteeth. A reduction in sawtooth period by about 40% was achieved in first experiments...

  2. The particle fluxes in the edge plasma during discharges with improved ohmic confinement in ASDEX

    International Nuclear Information System (INIS)

    In the regime of Improved Ohmic Confinement (IOC) in ASDEX the energy confinement time τE increases linearly with increasing line-averaged density n-bare up to the density limit. The establishment of the IOC is accompanied by a substantial reduction of the external gas feed, concomitant with large decreases of all plasma edge fluxes. However, the data do not supply conclusive evidence that the IOC is primarily connected with the recycling conditions. More recent observations with very clean machine conditions seem to indicate that the impurity radiation plays a significant role. (author)

  3. Concepts for improving the accuracy of gas balance measurement at ASDEX Upgrade

    International Nuclear Information System (INIS)

    The ITER fusion reactor which is under construction will use a deuterium–tritium gas mixture for operation. A fraction of this fusion fuel remains inside of the machine due to various mechanisms. The evaluation of this retention in present fusion experiments is of crucial importance to estimate the expected tritium inventory in ITER which shall be limited due to safety considerations. At ASDEX Upgrade (AUG) sufficiently time-resolved measurements should take place to extrapolate from current 10 s discharges to the at least intended 400 s ones of ITER. To achieve this, a new measurement system has been designed that enables accuracy of better than one per cent

  4. Resolving the bulk ion region of millimeter-wave collective Thomson scattering spectra at ASDEX Upgrade

    DEFF Research Database (Denmark)

    Stejner Pedersen, Morten; Nielsen, Stefan Kragh; Jacobsen, Asger Schou; Korsholm, Søren Bang; Leipold, Frank; Meo, Fernando; Michelsen, Poul; Moseev, Dmitry; Rasmussen, Jesper; Salewski, Mirko; Schubert, M.; Stober, J.; Wagner, D. H.

    2014-01-01

    Collective Thomson scattering (CTS) measurements provide information about the composition and velocity distribution of confined ion populations in fusion plasmas. The bulk ion part of the CTS spectrum is dominated by scattering off fluctuations driven by the motion of thermalized ion populations....... It thus contains information about the ion temperature, rotation velocity, and plasma composition. To resolve the bulk ion region and access this information, we installed a fast acquisition system capable of sampling rates up to 12.5 GS/s in the CTS system at ASDEX Upgrade. CTS spectra with...... temperature, rotation velocity, and plasma composition....

  5. Dual array 3D electron cyclotron emission imaging at ASDEX Upgrade.

    Science.gov (United States)

    Classen, I G J; Domier, C W; Luhmann, N C; Bogomolov, A V; Suttrop, W; Boom, J E; Tobias, B J; Donné, A J H

    2014-11-01

    In a major upgrade, the (2D) electron cyclotron emission imaging diagnostic (ECEI) at ASDEX Upgrade has been equipped with a second detector array, observing a different toroidal position in the plasma, to enable quasi-3D measurements of the electron temperature. The new system will measure a total of 288 channels, in two 2D arrays, toroidally separated by 40 cm. The two detector arrays observe the plasma through the same vacuum window, both under a slight toroidal angle. The majority of the field lines are observed by both arrays simultaneously, thereby enabling a direct measurement of the 3D properties of plasma instabilities like edge localized mode filaments. PMID:25430246

  6. ASDEX contributions to the 17th European conference on controlled fusion and plasma heating

    International Nuclear Information System (INIS)

    The 'ASDEX contributions to the 17th European conference on controlled fusion and plasma heating' (Amsterdam, June 25-29, 1990) hold one invited paper (Physics of enhanced confinement with peaked and board density profiles) and 12 chapters containing 44 contributed papers dealing with the following topics: Lower hybrid current drive and heating; Ion cyclotron heating; General confinement studies; Fluctuation studies; Direct measurement of transport coefficients; H-mode studies; Pellet studies; Divertor and SOL-studies; Impurity and impurity transport studies; Density limit studies; MHD studies; Diagnostic development. (orig./AH)

  7. Dual array 3D electron cyclotron emission imaging at ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Classen, I. G. J., E-mail: I.G.J.Classen@differ.nl; Bogomolov, A. V. [FOM-Institute DIFFER, Dutch Institute for Fundamental Energy Research, 3430 BE Nieuwegein (Netherlands); Domier, C. W.; Luhmann, N. C. [Department of Applied Science, University of California at Davis, Davis, California 95616 (United States); Suttrop, W.; Boom, J. E. [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, 85748 Garching (Germany); Tobias, B. J. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Donné, A. J. H. [FOM-Institute DIFFER, Dutch Institute for Fundamental Energy Research, 3430 BE Nieuwegein (Netherlands); Department of Applied Physics, Eindhoven University of Technology, 5600 MB Eindhoven (Netherlands)

    2014-11-15

    In a major upgrade, the (2D) electron cyclotron emission imaging diagnostic (ECEI) at ASDEX Upgrade has been equipped with a second detector array, observing a different toroidal position in the plasma, to enable quasi-3D measurements of the electron temperature. The new system will measure a total of 288 channels, in two 2D arrays, toroidally separated by 40 cm. The two detector arrays observe the plasma through the same vacuum window, both under a slight toroidal angle. The majority of the field lines are observed by both arrays simultaneously, thereby enabling a direct measurement of the 3D properties of plasma instabilities like edge localized mode filaments.

  8. ITER tokamak device

    Science.gov (United States)

    Doggett, J.; Salpietro, E.; Shatalov, G.

    1991-07-01

    The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER, a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fueling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (1) magnet systems (toroidal and poloidal field coils and cryogenic systems), (2) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (3) first wall, (4) divertor plate (design and materials, performance and lifetime, a.o.), (5) blanket/shield system, (6) maintenance equipment, (7) current drive and heating, (8) fuel cycle system, and (9) diagnostics.

  9. Axisymmetric control in tokamaks

    International Nuclear Information System (INIS)

    Vertically elongated tokamak plasmas are intrinsically susceptible to vertical axisymmetric instabilities as a result of the quadrupole field which must be applied to produce the elongation. The present work analyzes the axisymmetric control necessary to stabilize elongated equilibria, with special application to the Alcator C-MOD tokamak. A rigid current-conserving filamentary plasma model is applied to Alcator C-MOD stability analysis, and limitations of the model are addressed. A more physically accurate nonrigid plasma model is developed using a perturbed equilibrium approach to estimate linearized plasma response to conductor current variations. This model includes novel flux conservation and vacuum vessel stabilization effects. It is found that the nonrigid model predicts significantly higher growth rates than predicted by the rigid model applied to the same equilibria. The nonrigid model is then applied to active control system design. Multivariable pole placement techniques are used to determine performance optimized control laws. Formalisms are developed for implementing and improving nominal feedback laws using the C-MOD digital-analog hybrid control system architecture. A proportional-derivative output observer which does not require solution of the nonlinear Ricatti equation is developed to help accomplish this implementation. The nonrigid flux conserving perturbed equilibrium plasma model indicates that equilibria with separatrix elongation of at least sep = 1.85 can be stabilized robustly with the present control architecture and conductor/sensor configuration

  10. Time-resolved spectroscopy in the Rijnhuizen Tokamak Project tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Box, F.M.A.; Kolk, E. van de [Associatie Euratom-FOM, Nieuwegein (Netherlands). FOM-Instituut voor Plasmafysica; Howard, J. [Plasma Research Laboratory, Research School of Physical Science and Engineering, Australian National University, Canberra 0200 (Australia); Meijer, F.G. [Physics Faculty, University of Amsterdam, Amsterdam (Netherlands)

    1997-03-01

    At the Rijnhuizen Tokamak Project tokamak spectrometers are used to diagnose the velocity distribution and abundances of impurity ions. Quantities can be measured as a function of time, and the temporal resolution depends on the line emissivity and can be as good as 0.2 ms for the strongest lines. Several spectrometers, equipped with a charge-coupled device array, are being used with spectral ranges in the visible, the vacuum UV and the extreme UV. (orig.)

  11. The ASDEX 100 keV neutral lithium beam diagnostic gun

    International Nuclear Information System (INIS)

    The neutral lithium beam gun intended for measurement of the poloidal magnetic field and of the density gradient in the scrape-off layer of ASDEX is described, and test results over a beam energy range of 27-100 keV are presented. In the gun, lithium ions are extracted from a solid emitter (#betta#-Eurcryptite) in a Pierce-type configuration, accelerated and focused in a two-tube immersion lens, and neutralized in a charge-exchange cell using sodium. The beam can be pulsed from less than one to several seconds, depending on experimental needs. At a distance of 165 cm from the gun the neutral beam equivalent current is typically greater than 1 mA (0.16 mA) for a beam energy of 100 keV (27 keV), the beam FWHM being about 8-9 mm. It is found that to produce a particular beam with a certain ratio must be maintained between the extraction and total beam voltages, this relationship depending in turn on the emitter-extractor separation. The principal features which distinguish the ASDEX gun from that employed on W7a are the greater compactness - all the active elements, i.e. emitter, extractor, lens, deflection plates and neutralizer, are contained with 57 cm - and the vacuum vessel, which simultaneously serves as the magnetic shielding. (orig.)

  12. Improved Collective Thomson Scattering measurements of fast ions at ASDEX upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Rasmussen, J.; Nielsen, S. K.; Stejner, M.; Salewski, M.; Jacobsen, A. S.; Korsholm, S. B.; Leipold, F.; Meo, F.; Michelsen, P. K. [Association Euratom-DTU, Technical University of Denmark, Department of Physics, DTU Riso/ Campus, DK-4000 Roskilde (Denmark); Moseev, D. [Association Euratom-FOM Institute DIFFER, 3430 BE Nieuwegein (Netherlands); Schubert, M.; Stober, J.; Tardini, G.; Wagner, D.; Collaboration: ASDEX Upgrade Team

    2014-08-21

    Understanding the behaviour of the confined fast ions is important in both current and future fusion experiments. These ions play a key role in heating the plasma and will be crucial for achieving conditions for burning plasma in next-step fusion devices. Microwave-based Collective Thomson Scattering (CTS) is well suited for reactor conditions and offers such an opportunity by providing measurements of the confined fast-ion distribution function resolved in space, time and 1D velocity space. We currently operate a CTS system at ASDEX Upgrade using a gyrotron which generates probing radiation at 105 GHz. A new setup using two independent receiver systems has enabled improved subtraction of the background signal, and hence the first accurate characterization of fast-ion properties. Here we review this new dual-receiver CTS setup and present results on fast-ion measurements based on the improved background characterization. These results have been obtained both with and without NBI heating, and with the measurement volume located close to the centre of the plasma. The measurements agree quantitatively with predictions of numerical simulations. Hence, CTS studies of fast-ion dynamics at ASDEX Upgrade are now feasible. The new background subtraction technique could be important for the design of CTS systems in other fusion experiments.

  13. The particle fluxes in the edge plasma during discharges with improved ohmic confinement in ASDEX

    International Nuclear Information System (INIS)

    In the recent experimental period of ASDEX a new regime of improved ohmic confinement (IOC) was discovered. So far the energy confinement time τE increased linearly with increasing line averaged density ne up to ne = 3·1013 cm-3 saturated, however, at higher densities. In the new IOC regime τE increases further with increasing ne up to ∼5·1013 cm-3. The IOC regime is achieved for D2 discharges only since the last modification of the ASDEX divertor which substantially increased the recycling from the divertor through the divertor slits. It also led to a reduction in gas consumption for a discharge by a factor of about 2. As it appears, the high fuelling rate required during a fast ramp-up of the plasma density leads to a transition into the Saturated Ohmic Confinememt (SOC) regime. Vice versa, the strong reduction in the external gas feed when the preprogrammed density plateau is reached seems to be essential for establishing the IOC. It is characterized by a pronounced peaking of the density profile. During the transition from the SOC to the IOC regime large variations in the signals of all edge and divertor related diagnostics are observed. In this paper we concentrate on the results of the Low Energy Neutral Particle Analyser (LENA), the sniffer probe, on the mass spectrometers measuring the divertor exhaust pressure. (author) 7 refs., 2 figs

  14. Edge physics and its impact on the improved ohmic confinement in ASDEX

    International Nuclear Information System (INIS)

    The edge conditions play a crucial role in achieving and maintaining the improved ohmic confinement (IOC) regime in ASDEX as has been stated by Haas et al. (1988) and Soeldner et al. (1988). This new regime is obtained after divertor reconstruction in deuterium discharges when the gas puffing is substantially reduced. IOC is then characterized by peaked density profiles and the linear scaling of the energy confinement time τE with the line-averaged density ne is recovered up to the density limit. In this paper, we discuss the evolution of the edge parameters in the transition from the linear (LOC) and then saturated (SOC) to the improved (IOC) ohmic regime. In addition, we describe the edge plasma mainly in terms of edge parameters like the separatrix density instead of bulk parameters such as the line-averaged density. This gives us the opportunity to identify and separate edge effects from the central behaviour. The data in the vicinity of the separatrix stem mainly from the single-pulse multipoint Thomson scattering system, the lithium beam spectroscopy, the Langmuir probe, and the time-of-flight spectrometer in ASDEX. For comparison, we will sometimes use measurements in the divertor chamber by electric triple probes and ionization gauges. (author) 6 refs., 3 figs

  15. Characterizing the edge plasma of different ohmic confinement regimes in ASDEX

    International Nuclear Information System (INIS)

    To compare different ohmic confinement regimes in ASDEX, the edge conditions are analyzed in detail. The results show that the improved ohmic confinement comes along with a drop of the separatrix density. This drop allows density profile to peak and seems to be the trigger of a change in the transport. Simultaneously, a universal scaling between the electron temperature and the electron density at the separatrix prevails for all ohmic scenarios. In addition, the total particle flux across the separatrix is evaluated and found to be strongly correlated to the separatrix density. Thus, the associated convective energy loss contributes less to the total energy losses when the confinement is improved. Since the correlations between edge parameters do not change in different ohmic confinement regimes of ASDEX, the edge physics remains about the same. Improved ohmic confinement is then characterized by an optimum separatrix density which provides a sufficient high edge temperature together with low particle fluxes. These optimum conditions then yield the maximum particle confinement. (orig.)

  16. Research using small tokamaks

    International Nuclear Information System (INIS)

    The technical reports in this document were presented at the IAEA Technical Committee Meeting ''Research on Small Tokamaks'', September 1990, in three sessions, viz., (1) Plasma Modes, Control, and Internal Phenomena, (2) Edge Phenomena, and (3) Advanced Configurations and New Facilities. In Section (1) experiments at controlling low mode number modes, feedback control using external coils, lower-hybrid current drive for the stabilization of sawtooth activity and continuous (1,1) mode, and unmodulated and fast modulated ECRH mode stabilization experiments were reported, as well as the relation to disruptions and transport of low m,n modes and magnetic island growth; static magnetic perturbations by helical windings causing mode locking and sawtooth suppression; island widths and frequency of the m=2 tearing mode; ultra-fast cooling due to pellet injection; and, finally, some papers on advanced diagnostics, i.e., lithium-beam activated charge-exchange spectroscopy, and detection through laser scattering of discrete Alfven waves. In Section (2), experimental edge physics results from a number of machines were presented (positive biasing on HYBTOK II enhancing the radial electric field and improving confinement; lower hybrid current drive on CASTOR improving global particle confinement, good current drive efficiency in HT-6B showing stabilization of sawteeth and Mirnov oscillations), as well as diagnostic developments (multi-chord time resolved soft and ultra-soft X-ray plasma radiation detection on MT-1; measurements on electron capture cross sections in multi-charged ion-atom collisions; development of a diagnostic neutral beam on Phaedrus-T). Theoretical papers discussed the influence of sheared flow and/or active feedback on edge microstability, large edge electric fields, and two-fluid modelling of non-ambipolar scrape-off layers. Section (3) contained (i) a proposal to construct a spherical tokamak ''Proto-Eta'', (ii) an analysis of ultra-low-q and runaway

  17. Spherical tokamak development in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma] (and others)

    2003-07-01

    The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  18. Bootstrap current in a tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Kessel, C.E.

    1994-03-01

    The bootstrap current in a tokamak is examined by implementing the Hirshman-Sigmar model and comparing the predicted current profiles with those from two popular approximations. The dependences of the bootstrap current profile on the plasma properties are illustrated. The implications for steady state tokamaks are presented through two constraints; the pressure profile must be peaked and {beta}{sub p} must be kept below a critical value.

  19. Bootstrap current in a tokamak

    International Nuclear Information System (INIS)

    The bootstrap current in a tokamak is examined by implementing the Hirshman-Sigmar model and comparing the predicted current profiles with those from two popular approximations. The dependences of the bootstrap current profile on the plasma properties are illustrated. The implications for steady state tokamaks are presented through two constraints; the pressure profile must be peaked and βp must be kept below a critical value

  20. The ETE spherical Tokamak project

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Gerson Otto; Andrade, Maria Celia Ramos de; Barbosa, Luis Filipe Wiltgen [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma] [and others]. E-mail: ludwig@plasma.inpe.br

    1999-07-01

    This paper describes the general characteristics of spherical tokamaks, with a brief overview of work in the area of spherical torus already performed or in progress at several institutions. The paper presents also the historical development of the ETE (Spherical Tokamak Experiment) project, its research program, technical characteristics and status of construction in September, 1998 at the Associated plasma Laboratory (LAP) of the National Institute for Space Research (INPE) in Brazil. (author)

  1. Spherical tokamak development in Brazil

    International Nuclear Information System (INIS)

    The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  2. Spheromak injection into a tokamak

    OpenAIRE

    Brown, M R; Bellan, P. M.

    1990-01-01

    Recent results from the Caltech spheromak injection experiment [to appear in Phys. Rev. Lett.] are reported. First, current drive by spheromak injection into the ENCORE tokamak as a result of the process of magnetic helicity injection is observed. An initial 30% increase in plasma current is observed followed by a drop by a factor of 3 because of sudden plasma cooling. Second, spheromak injection results in an increase of tokamak central density by a factor of 6. The high-current/high-density...

  3. Confinement and diffusion in tokamaks

    International Nuclear Information System (INIS)

    The effect of electric field fluctuations on confinement and diffusion in tokamak is discussed. Based on the experimentally determined cross-field turbolent diffusion coefficient, D∼3.7*cTe/eB(δni/ni)rms which is also derived by a simple theory, the cross-field diffusion time, tp=a2/D, is calculated and compared to experimental results from 51 tokamak for standard Ohmic operation

  4. Stellarator - tokamak configurations

    International Nuclear Information System (INIS)

    The stellarator configuration and tokamak configuration with helical fields have been studied both from an equilibrium and stability point of view. The model was restricted to a surface current model with a sharp boundary between plasma and vacuum. A general derivation of equilibrium and stability based on the Energy Principle is given. Physically the unstable modes are identified as external global modes. Detailed numerical results in different parameter regimes are presented and discussed. Critical β-limits for equilibrium and stability are obtained and in particular it is shown that in certain parameter ranges there exist a high-β as well as a low-β-region of stability. 7 refs., 14 figs

  5. The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D3He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions

  6. The ARIES tokamak reactor study

    Energy Technology Data Exchange (ETDEWEB)

    1989-10-01

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D{sup 3}He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions.

  7. Magnetohydrodynamic instability, feedback stabilization, and disruption study for the Korea superconducting tokamak advanced research tokamak

    International Nuclear Information System (INIS)

    Passive and active feedback stabilization schemes being considered in Korea Superconducting Tokamak Advanced Research (KSTAR) device for the stabilization of the resistive magnetohydrodynamic modes such as the resistive wall and the neoclassical tearing are briefly introduced. A short summary is also presented on the tokamak simulation results of disruption dynamics and load in the KSTAR tokamak obtained using the tokamak simulation code (TSC)

  8. The upgradation of Aditya Tokamak

    International Nuclear Information System (INIS)

    Aditya Tokamak is the first Indian tokamak, indigenously built and commissioned at the Institute for Plasma Research, Gandhinagar, Gujarat, India, in September, 1989. Aditya Tokamak has been in operation since more than 25 years. More than 30,000 discharges are taken and a large number of experiments are carried out, with plasma current ranging from 50 KA to 150 KA, lasting for 100 to 250 milliseconds. Various types of wall conditioning techniques and different hot plasma diagnostics are tested and operated on Aditya Tokamak. The experiments for turbulent particle transport and turbulence in the edge plasma, gas puffing, lithium coating, mitigation, plasma disruption, limiter and electron biasing, runaway discharges etc. led to many interesting results contributing immensely to the world of thermonuclear fusion. Experiments on Pre-ionization and Plasma heating by ICRH and ECRH are also worked out. The scientific objectives of Aditya tokamak Upgrade include Low loop voltage plasma start-up with strong pre-ionization having a good plasma control system. The upgrade is designed keeping in mind the experiments, disruption mitigation studies relevant to future fusion devices, runway mitigation studies, demonstration of Radio-frequency heating and current drive etc. This upgraded Aditya tokamak will be used for basic studies on plasma confinement and scaling to larger devices, development and testing of new diagnostics etc. This machine will be easily accessible compared to SST-1 and will be very useful for generation of technical and scientific expertise for future fusion devices. In this paper, especial features of the upgrade including various aspects of designing of new components for Aditya Upgrade tokamak is presented

  9. Bibliography of fusion product physics in tokamaks

    International Nuclear Information System (INIS)

    Almost 700 citations have been compiled as the first step in reviewing the recent research on tokamak fusion product effects in tokamaks. The publications are listed alphabetically by the last name of the first author and by subject category

  10. ASDEX Upgrade Discharge Control System—A real-time plasma control framework

    International Nuclear Information System (INIS)

    Highlights: • The ASDEX Upgrade Discharge Control System (DCS) is a comprehensive control system to conduct fusion experiments. • DCS supports real-time diagnostic integration, adaptable feedback schemes, actuator management and exception handling. • DCS offers workflow management, logging and archiving, self-monitoring and inter-process communication. • DCS is based on a distributed, modular software framework architecture designed for real-time operation. • DCS is composed of re-usable generic but highly customisable components. - Abstract: ASDEX Upgrade is a fusion experiment with a size and complexity to allow extrapolation of technical and physical conditions and requirements to devices like ITER and even beyond. In addressing advanced physics topics it makes extensive use of sophisticated real-time control methods. It comprises real-time diagnostic integration, dynamically adaptable multivariable feedback schemes, actuator management including load distribution schemes and a powerful monitoring and pulse supervision concept based on segment scheduling and exception handling. The Discharge Control System (DCS) supplies all this functionality on base of a modular software framework architecture designed for real-time operation. It provides system-wide services like workflow management, logging and archiving, self-monitoring and inter-process communication on Linux, VxWorks and Solaris operating systems. By default DCS supports distributed computing, and a communication layer allows multi-directional signal transfer and data-driven process synchronisation over shared memory as well as over a number of real-time networks. The entire system is built following the same common design concept combining a rich set of re-usable generic but highly customisable components with a configuration-driven component deployment method. We will give an overview on the architectural concepts as well as on the outstanding capabilities of DCS in the domains of inter

  11. ASDEX Upgrade Discharge Control System—A real-time plasma control framework

    Energy Technology Data Exchange (ETDEWEB)

    Treutterer, W., E-mail: Wolfgang.Treutterer@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstraße 2, 85748 Garching (Germany); Cole, R.; Lüddecke, K. [Unlimited Computer Systems GmbH, Iffeldorf (Germany); Neu, G.; Rapson, C.; Raupp, G.; Zasche, D.; Zehetbauer, T. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstraße 2, 85748 Garching (Germany)

    2014-03-15

    Highlights: • The ASDEX Upgrade Discharge Control System (DCS) is a comprehensive control system to conduct fusion experiments. • DCS supports real-time diagnostic integration, adaptable feedback schemes, actuator management and exception handling. • DCS offers workflow management, logging and archiving, self-monitoring and inter-process communication. • DCS is based on a distributed, modular software framework architecture designed for real-time operation. • DCS is composed of re-usable generic but highly customisable components. - Abstract: ASDEX Upgrade is a fusion experiment with a size and complexity to allow extrapolation of technical and physical conditions and requirements to devices like ITER and even beyond. In addressing advanced physics topics it makes extensive use of sophisticated real-time control methods. It comprises real-time diagnostic integration, dynamically adaptable multivariable feedback schemes, actuator management including load distribution schemes and a powerful monitoring and pulse supervision concept based on segment scheduling and exception handling. The Discharge Control System (DCS) supplies all this functionality on base of a modular software framework architecture designed for real-time operation. It provides system-wide services like workflow management, logging and archiving, self-monitoring and inter-process communication on Linux, VxWorks and Solaris operating systems. By default DCS supports distributed computing, and a communication layer allows multi-directional signal transfer and data-driven process synchronisation over shared memory as well as over a number of real-time networks. The entire system is built following the same common design concept combining a rich set of re-usable generic but highly customisable components with a configuration-driven component deployment method. We will give an overview on the architectural concepts as well as on the outstanding capabilities of DCS in the domains of inter

  12. Moving Divertor Plates in a Tokamak

    International Nuclear Information System (INIS)

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions

  13. Fusion potential for spherical and compact tokamaks

    International Nuclear Information System (INIS)

    The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high β-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect

  14. Moving Divertor Plates in a Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S.J. Zweben, H. Zhang

    2009-02-12

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions.

  15. Fusion potential for spherical and compact tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Sandzelius, Mikael

    2003-02-01

    The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high {beta}-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect.

  16. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars

  17. Test of the predictive capability of B2-Eirene on ASDEX-Upgrade

    International Nuclear Information System (INIS)

    Based on validated B2-Eirene results for the previous divertor of ASDEX Upgrade, the modelling predictions for the new divertor are compared with the actual experimental results. For the same experimental scenarios (L-mode) in both divertors the predictions are robust and in agreement with experimental results. For a full quantitative agreement in H-mode both the carbon chemical sputtering yield and the radial transport had to be adjusted. The new divertor has a reduced power load due to larger radiation losses. These are caused by larger hydrogen losses, enhancement of carbon radiation due to radial transport and convective energy transport into the radiation zone, and larger radial energy transport in the divertor. (author)

  18. Enhancement of the FIDA diagnostic at ASDEX Upgrade for velocity space tomography

    Science.gov (United States)

    Weiland, M.; Geiger, B.; Jacobsen, A. S.; Reich, M.; Salewski, M.; Odstrčil, T.; the ASDEX Upgrade Team

    2016-02-01

    Recent upgrades to the FIDA (fast-ion D-alpha) diagnostic at ASDEX Upgrade are discussed. The diagnostic has been extended from three to five line of sight arrays with different angles to the magnetic field, and a spectrometer redesign allows the simultaneous measurement of red- and blue-shifted parts of the Doppler spectrum. These improvements make it possible to reconstruct the 2D fast-ion velocity distribution f≤ft(E,{{v}\\parallel}/v\\right) from the FIDA measurements by tomographic inversion under a wide range of plasma parameters. Two applications of the tomography are presented: a comparison between the distributions resulting from 60 keV and 93 keV neutral beam injection and a velocity-space resolved study of fast-ion redistribution induced by a sawtooth crash inside and outside the sawtooth inversion radius.

  19. Performance measurements of the collective Thomson scattering receiver at ASDEX Upgrade

    DEFF Research Database (Denmark)

    Furtula, Vedran; Leipold, Frank; Salewski, Mirko;

    2012-01-01

    The fast-ion collective Thomson scattering (CTS) receiver at ASDEX Upgrade can detect spectral power densities of a few eV in the millimeter-wave range against the electron cyclotron emission (ECE) background on the order of 100 eV under presence of gyrotron stray radiation that is several orders...... detector diodes. The performance of the entire receiver is determined by the main receiver components operating at mm-wave frequencies (notch-, bandpass- and lowpass filters, a voltage-controlled variable attenuator, and an isolator), a mixer, and the IF components (amplifiers, band-pass filters......, and detector diodes). We discuss here the design of the entire receiver, focussing on its performance as a unit. The receiver has been disassembled, and the performance of its individual components has been characterized. Based on these individual component measurements we predict the spectral response...

  20. Electric Probe Measurements of the Poloidal Velocity in the Scrape-Off Layer of ASDEX Upgrade

    DEFF Research Database (Denmark)

    Mehlmann, F.; Costea, S.; Schrittwieser, R..; Naulin, Volker; Juul Rasmussen, Jens; Müller, H.W.; Nielsen, Anders Henry; Vianello, N.; Carralero, D.; Rohde, V.; Lux, C.; Ionita, C.

    2014-01-01

    A reciprocating probe head with six pins was used for localized measurements of electric fields and densities in the scrape-off layer (SOL) of ASDEX Upgrade (AUG) up to the edge shear layer (SL) near the Last Closed Flux Surface (LCFS). The edge SL is characterized by a strong sudden change in the...... poloidal velocity v close to the separatrix. The probes were used to determine this velocity by different methods which are critically compared to each other concerning their reliability. By the first method the poloidal velocity was deduced from the radial electric field E-r measured by two radially...... staggered probe pins, with v being due to the E-r x B-phi-drift (B-phi is the toroidal field). The two other methods utilized the cross correlation of two poloidally staggered ion-biased probes and two poloidally staggered floating probes, respectively. In this case the time lags with maximum cross...

  1. Injection of nitrogen-15 tracer into ASDEX-Upgrade: New technique in material migration studies

    Energy Technology Data Exchange (ETDEWEB)

    Petersson, P., E-mail: Per.Petersson@ee.kth.sen [Fusion Plasma Physics, Royal Institute of Technology, Association EURATOM-VR, SE-100 44 Stockholm (Sweden); Hakola, A.; Likonen, J. [VTT, Association EURATOM-TEKES, 02044 Espoo, VTT (Finland); Mayer, M. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Miettunen, J. [Aalto University, Association EURATOM-TEKES, 00076 Aalto (Finland); Neu, R.; Rohde, V. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Rubel, M. [Fusion Plasma Physics, Royal Institute of Technology, Association EURATOM-VR, SE-100 44 Stockholm (Sweden)

    2013-07-15

    For the first time nitrogen-15 gas was used as a tracer for determining of global nitrogen retention in ASDEX-Upgrade. The injection done from the midplane gas inlet on the last day of the experimental campaign was followed by retrieval and ex situ analyses of many limiter and divertor tiles. The study was done by nuclear reaction analysis using the {sup 15}N({sup 1}H, γ{sup 4}He){sup 12}C process and detecting both γ radiation and {sup 4}He. The highest and peaked concentrations of {sup 15}N, 8 × 10{sup 16} cm{sup 2}, were found on limiters close to the injection point, while fairly homogeneous flat profiles were measured on most of the divertor plates. The measured concentrations are compared to an ASCOT simulation of the injection.

  2. Towards non-linear simulations of full ELM crashes in ASDEX upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Lessig, Alexander; Hoelzl, Matthias; Lackner, Karl; Guenter, Sibylle [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Boltzmannstrasse 2, 85748 Garching (Germany)

    2014-07-01

    Edge localized modes (ELMs) of large size are a severe concern for the operation of ITER due to the large transient heat loads on divertor targets and wall structures. Using the non-linear MHD code JOREK, we have performed first simulations of full ELM crashes in ASDEX upgrade, taking into account a large number of toroidal Fourier harmonics. The evolution of the toroidal Fourier spectrum and the drop of pedestal gradients are studied. In particular, we confirm a previously introduced quadratic mode coupling model for the early non-linear evolution of the mode structure and present first results concerning the evolution in the fully non-linear phase. Eventually, we aim to identify different ELM types in our simulations as observed in experiments and to compare the results to experimental observations, e.g., regarding the pedestal evolution and the heat deposition patterns. Work is ongoing to increase poloidal resolution and include diamagnetic stabilization of high mode numbers.

  3. Experimental studies and modeling of complete H-mode divertor detachment in ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Reimold, F., E-mail: Felix.Reimold@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, Boltzmannstraß e 2, D-85748 Garching (Germany); Wischmeier, M.; Bernert, M.; Potzel, S.; Coster, D. [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, D-85748 Garching (Germany); Bonnin, X. [CNRS-LSPM, Université Paris 13, Sorbonne Paris Cité, F-93430 Villetaneuse (France); Reiter, D. [Institut für Energie- und Klimaforschung – Plasmaphysik, Forschungszentrum Jülich GmbH (Germany); Meisl, G.; Kallenbach, A. [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, D-85748 Garching (Germany); Aho-Mantila, L. [VTT, FI-02044 VTT (Finland); Stroth, U. [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, D-85748 Garching (Germany)

    2015-08-15

    Power exhaust in future fusion devices is critical and operation with a detached divertor is foreseen for ITER and DEMO. The evolution of detachment in nitrogen seeded H-mode discharges at ASDEX Upgrade is categorized in four phases. Complete detachment of the outer target is found to be correlated with a strongly localized radiation at the X-point and a pressure loss at the pedestal top at almost constant core plasma pressure. SOLPS modeling shows that enhanced radial transport in the divertor region is necessary to reconcile the experimental profiles with the simulations. The modeling supports the experimental observation of the correlation of complete detachment with an X-point radiation and a reduction of the pedestal top pressure. A remaining discrepancy are significantly lower neutral densities in the divertor compared to experiment. The effects of wall pumping, the particle reflection model and the boundary conditions on the plasma solution are discussed.

  4. Model-based radiation scalings for the ITER-like divertors of JET and ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Aho-Mantila, L., E-mail: leena.aho-mantila@vtt.fi [VTT Technical Research Centre of Finland, FI-02044 VTT (Finland); Bonnin, X. [LSPM – CNRS, Université Paris 13, Sorbonne Paris Cité, F-93430 Villetaneuse (France); Coster, D.P. [Max-Planck Institut für Plasmaphysik, D-85748 Garching (Germany); Lowry, C. [EFDA JET CSU, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Wischmeier, M. [Max-Planck Institut für Plasmaphysik, D-85748 Garching (Germany); Brezinsek, S. [Forschungszentrum Jülich, Institut für Energie- und Klimaforschung Plasmaphysik, 52425 Jülich (Germany); Federici, G. [EFDA PPP& T Department, D-85748 Garching (Germany)

    2015-08-15

    Effects of N-seeding in L-mode experiments in ASDEX Upgrade and JET are analysed numerically with the SOLPS5.0 code package. The modelling yields 3 qualitatively different radiative regimes with increasing N concentration, when initially attached outer divertor conditions are studied. The radiation pattern is observed to evolve asymmetrically, with radiation increasing first in the inner divertor, then in the outer divertor, and finally on closed field lines above the X-point. The properties of these radiative regimes are observed to be sensitive to cross-field drifts and they differ between the two devices. The modelled scaling of the divertor radiated power with the divertor neutral pressure is similar to an experimental scaling law for H-mode radiation. The same parametric dependencies are not observed in simulations without drifts.

  5. Resolving the bulk ion region of millimeter-wave collective Thomson scattering spectra at ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Stejner, M., E-mail: mspe@fysik.dtu.dk; Nielsen, S.; Jacobsen, A. S.; Korsholm, S. B.; Leipold, F.; Meo, F.; Michelsen, P. K.; Rasmussen, J.; Salewski, M. [Department of Physics, Association EURATOM-DTU, Technical University of Denmark, DK-2800 Kgs. Lyngby (Denmark); Moseev, D. [Max-Planck-Institut für Plasmaphysik, EURATOM-Association, Boltzmannstr. 2, 85748 Garching (Germany); Association Euratom-FOM Institute DIFFER, 3430 BE Nieuwegein (Netherlands); Schubert, M.; Stober, J.; Wagner, D. H. [Max-Planck-Institut für Plasmaphysik, EURATOM-Association, Boltzmannstr. 2, 85748 Garching (Germany)

    2014-09-15

    Collective Thomson scattering (CTS) measurements provide information about the composition and velocity distribution of confined ion populations in fusion plasmas. The bulk ion part of the CTS spectrum is dominated by scattering off fluctuations driven by the motion of thermalized ion populations. It thus contains information about the ion temperature, rotation velocity, and plasma composition. To resolve the bulk ion region and access this information, we installed a fast acquisition system capable of sampling rates up to 12.5 GS/s in the CTS system at ASDEX Upgrade. CTS spectra with frequency resolution in the range of 1 MHz are then obtained through direct digitization and Fourier analysis of the CTS signal. We here describe the design, calibration, and operation of the fast receiver system and give examples of measured bulk ion CTS spectra showing the effects of changing ion temperature, rotation velocity, and plasma composition.

  6. Determination of impurity concentrations and Zeff by VUV spectroscopy on ASDEX

    International Nuclear Information System (INIS)

    The impurity concentrations and corresponding Zeff contributions as well as the dilution of the deuterium background plasma in ASDEX are determined by VUV spectroscopy. The methods used are described in detail. We describe the absolute calibration of our VUV survey spectrometer with two different calibration sources, as well as our ZEDIFF time-dependent transport code, used for interpreting the spectroscopic measurements. The assessed spectroscopic Zeff compares quite well with the bremsstrahlung Zeff as demonstrated for a number of representative ohmically and additionally heated discharges. In order to obtain these results readily on a shot-to-shot basis at the end of each discharge, a simplified fast evaluation method is introduced. This fast analysis method yields the central impurity concentrations, the central Zeff contributions, and the dilution of the deuterons. Again, the results from the fast analysis method agree well with those from our extended transport code treatment and with the bremsstrahlung Zeff. (orig.)

  7. Non-linear simulations of ELMs in ASDEX Upgrade including diamagnetic drift effects

    Energy Technology Data Exchange (ETDEWEB)

    Lessig, Alexander; Hoelzl, Matthias; Krebs, Isabel; Franck, Emmanuel; Guenter, Sibylle [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2, 85748 Garching (Germany); Orain, Francois; Morales, Jorge; Becoulet, Marina [CEA-IRFM, Cadarache, 13108 Saint-Paul-Lez-Durance (France); Huysmans, Guido [ITER Organization, 13067 Saint-Paul-Lez-Durance (France)

    2015-05-01

    Large edge localized modes (ELMs) are a severe concern for ITER due to high transient heat loads on divertor targets and wall structures. Using the non-linear MHD code JOREK, we have performed ELM simulations for ASDEX Upgrade (AUG) including diamagnetic drift effects. The influence of diamagnetic terms onto the evolution of the toroidal mode spectrum for different AUG equilibria and the non-linear interaction of the toroidal harmonics are investigated. In particular, we confirm the diamagnetic stabilization of high mode numbers and present new features of a previously introduced quadratic mode coupling model for the early non-linear evolution of the mode structure. Preliminary comparisons of full ELM crashes with experimental observations are shown aiming at code validation and the understanding of different ELM types. Work is ongoing to include toroidal and neoclassical poloidal rotation in our simulations.

  8. Shear strength of the ASDEX upgrade TF coil insulation: Rupture, fatigue and creep behaviour

    International Nuclear Information System (INIS)

    This report is concerned with the interlaminar shear strength of the insulation system for the 16 toroidal field (TF) coils of ASDEX upgrade. The interlaminar shear properties of the glass-epoxy insulation are primarily determined by the resin system (ARALDIT-F, HT 907, DZ 40) and its curing procedure. The pure resin was therefore tested first in tension. The results were taken into account for setting up the method of curing the TF coils. Shear tests of the complete glass-epopxy system were then conducted with tubular torque specimens providing a nearly homogeneous stress distribution. In particular, the influence of the amount of flexibilizer (5, 10, 15 parts of resin weight = PoW) on the rupture and fatigue strengths was assessed at a temperature T=60 C, as also was the temperature dependence of the creep rate (40 C, 60 C, 80 C). The results obtained are not based on safe statistics. Nevertheless, they show clear trends. (orig.)

  9. Application for EURATOM priority support of additional heating for ASDEX Upgrade, phase I and phase II

    International Nuclear Information System (INIS)

    In order to reach the full performance plasma parameters of ASDEX Upgrade as provided by the machine technique a heating power of 12 to 15 MW is required. For the minimum required power the appropriate choice for the basic heating system are 6 MW ICRH and 6 MW neutral injection, both with a long pulse capability of up to 10 seconds. ICRH in a frequency range of 30 to 120 MHz shall cover He3 minority, hydrogen fundamental and 2nd harmonic and deuterium 2nd harmonic heating. For neutral injection four JET sources with 60 keV H0 and 80 A combined in one injection box were chosen. The averaged injection angle is 240 to perpendicular at Rsub(O) = 1.7 m. Both systems shall be installed during 1988. The costs are 57.4 MDM for both. (orig./GG)

  10. Non-linear modeling of the plasma response to RMPs in ASDEX Upgrade

    CERN Document Server

    Orain, F; Viezzer, E; Dunne, M; Becoulet, M; Cahyna, P; Huijsmans, G T A; Morales, J; Willensdorfer, M; Suttrop, W; Kirk, A; Pamela, S; Strumberger, E; Guenter, S; Lessig, A

    2016-01-01

    The plasma response to Resonant Magnetic Perturbations (RMPs) in ASDEX Upgrade is modeled with the non-linear resistive MHD code JOREK, using input profiles that match those of the experiments as closely as possible. The RMP configuration for which Edge Localized Modes are best mitigated in experiments is related to the largest edge kink response observed near the X-point in modeling. On the edge resonant surfaces q = m=n, the coupling between the m + 2 kink component and the m resonant component is found to induce the amplification of the resonant magnetic perturbation. The ergodicity and the 3D-displacement near the X-point induced by the resonant ampli?cation can only partly explain the density pumpout observed in experiments.

  11. Status, Operation, and Extension of the ECRH System at ASDEX Upgrade

    Science.gov (United States)

    Wagner, D.; Stober, J.; Leuterer, F.; Monaco, F.; Müller, S.; Münich, M.; Rapson, C. J.; Reich, M.; Schubert, M.; Schütz, H.; Treutterer, W.; Zohm, H.; Thumm, M.; Scherer, T.; Meier, A.; Gantenbein, G.; Jelonnek, J.; Kasparek, W.; Lechte, C.; Plaum, B.; Goodman, T.; Litvak, A. G.; Denisov, G. G.; Chirkov, A.; Zapevalov, V.; Malygin, V.; Popov, L. G.; Nichiporenko, V. O.; Myasnikov, V. E.; Tai, E. M.; Solyanova, E. A.; Malygin, S. A.

    2016-01-01

    The upgraded electron cyclotron resonance heating (ECRH) system at ASDEX Upgrade (AUG) has been routinely used with eight gyrotrons during the last experimental campaign. A further upgrade will replace the existing system of four short-pulse (140 GHz, 2 s, 500 kW) gyrotrons. The final goal is to have around 6.5-7 MW at 140 GHz (or 5.5 MW at 105 GHz) from eight units available in the plasma during the whole AUG discharge (10 s). The system operates at 140 and 105 GHz with X2, O2 and X3 schemes. For B > 3 T also an ITER-like O1-scenario can be run using the 105 GHz option. Four of the eight launching antennas are capable of fast poloidal movements necessary for real-time control of the location of power deposition.

  12. Event detection and exception handling strategies in the ASDEX Upgrade discharge control system

    Energy Technology Data Exchange (ETDEWEB)

    Treutterer, W., E-mail: Wolfgang.Treutterer@ipp.mpg.de; Neu, G.; Rapson, C.; Raupp, G.; Zasche, D.; Zehetbauer, T.

    2013-10-15

    Highlights: •Event detection and exception handling is integrated in control system architecture. •Pulse control with local exception handling and pulse supervision with central exception handling are strictly separated. •Local exception handling limits the effect of an exception to a minimal part of the controlled system. •Central Exception Handling solves problems requiring coordinated action of multiple control components. -- Abstract: Thermonuclear plasmas are governed by nonlinear characteristics: plasma operation can be classified into scenarios with pronounced features like L and H-mode, ELMs or MHD activity. Transitions between them may be treated as events. Similarly, technical systems are also subject to events such as failure of measurement sensors, actuator saturation or violation of machine and plant operation limits. Such situations often are handled with a mixture of pulse abortion and iteratively improved pulse schedule reference programming. In case of protection-relevant events, however, the complexity of even a medium-sized device as ASDEX Upgrade requires a sophisticated and coordinated shutdown procedure rather than a simple stop of the pulse. The detection of events and their intelligent handling by the control system has been shown to be valuable also in terms of saving experiment time and cost. This paper outlines how ASDEX Upgrade's discharge control system (DCS) detects events and handles exceptions in two stages: locally and centrally. The goal of local exception handling is to limit the effect of an unexpected or asynchronous event to a minimal part of the controlled system. Thus, local exception handling facilitates robustness to failures but keeps the decision structures lean. A central state machine deals with exceptions requiring coordinated action of multiple control components. DCS implements the state machine by means of pulse schedule segments containing pre-programmed waveforms to define discharge goal and control

  13. Parametric investigation of the density profile in the scrape-off layer of ASDEX

    Science.gov (United States)

    McCormick, K.; Pietrzyk, Z. A.; Murmann, H.; Lenoci, M.; ASDEX Team; Becker, G.; Bosch, H. S.; Brocken, H.; Bühl, K.; Eberhagen, A.; Eckhartt, D.; Fussmann, G.; Gehre, O.; Gernhardt, J.; Gierke, G. V.; Glock, E.; Gruber, O.; Haas, G.; Hofmann, J.; Izvozchikov, A.; Janeschitz, G.; Karger, F.; Kaufmann, M.; Keilhacker, M.; Klüber, O.; Kornherr, M.; Lackner, K.; Lang, R. S.; Leuterer, F.; Lisitano, G.; Mast, F.; Mayer, H. M.; Meisel, D.; Mertens, V.; Müller, E. R.; Neuhauser, J.; Noterdaeme, J.-M.; Niedermeyer, H.; Poschenrieder, W.; Rapp, H.; Riedler, H.; Röhr, H.; Roth, J.; Ryter, F.; Sandmann, W.; Schneider, F.; Setzensack, C.; Siller, G.; Smeulders, P.; Söldner, F. X.; Speth, E.; Steinmetz, K.; Steuer, K.-H.; Tsois, N.; Ugniewski, S.; Vlases, G.; Vollmer, O.; Wagner, F.; Wesner, F.; Zasche, D.

    1987-02-01

    Systematic investigations of the scrape-off layer (SOL) in the midplane of ASDEX have been carried out in He, D 2 and H 2 for diverted ohmic discharges over a wide range of plasma conditions: overlinene ˜ 0.5-4.7 × 10 13 cm -3, Ip = 200-450 kA, BT = 16-23 kG, qa˜ 2.4-4.4 and POH = 200-480 kW. For the first two cm outside the separatix, ne is found to decay exponentially with an e-folding length λn given by λn = kqα (He, k = 1.32 cm, α = 0.52; D 2, k =1.29 cm, α = 0.35; H 2, k = 1.18 cm, α = 0.4) when from which follows for qa = 3: λn( D2) ˜ λn( H2) ˜ 0.8 λn( He). The qα scaling is roughly predicted by the simple formula λ n = {D ⊥ L }/{υ ∥} under the assumption D⊥ ∝ mi-0.5 (as has been observed on ASDEX for H 2 and D 2). There appears to be no explicit λn dependence on heating power. λn varies strongly with overlinene in the range overlinene ≤ 1 × 10 13 cm -3, decreasing for example (D 2,H 2; qa = 3.0), from λn ≥ 3 cm at overlinene ˜ 0.5 × 10 13 cm -3 to λn ˜ 1.9 cm for overlinene ≥ 1.5 × 10 13 cm -3, ne at the separatrix is primarily a function of overlinene.

  14. Event detection and exception handling strategies in the ASDEX Upgrade discharge control system

    International Nuclear Information System (INIS)

    Highlights: •Event detection and exception handling is integrated in control system architecture. •Pulse control with local exception handling and pulse supervision with central exception handling are strictly separated. •Local exception handling limits the effect of an exception to a minimal part of the controlled system. •Central Exception Handling solves problems requiring coordinated action of multiple control components. -- Abstract: Thermonuclear plasmas are governed by nonlinear characteristics: plasma operation can be classified into scenarios with pronounced features like L and H-mode, ELMs or MHD activity. Transitions between them may be treated as events. Similarly, technical systems are also subject to events such as failure of measurement sensors, actuator saturation or violation of machine and plant operation limits. Such situations often are handled with a mixture of pulse abortion and iteratively improved pulse schedule reference programming. In case of protection-relevant events, however, the complexity of even a medium-sized device as ASDEX Upgrade requires a sophisticated and coordinated shutdown procedure rather than a simple stop of the pulse. The detection of events and their intelligent handling by the control system has been shown to be valuable also in terms of saving experiment time and cost. This paper outlines how ASDEX Upgrade's discharge control system (DCS) detects events and handles exceptions in two stages: locally and centrally. The goal of local exception handling is to limit the effect of an unexpected or asynchronous event to a minimal part of the controlled system. Thus, local exception handling facilitates robustness to failures but keeps the decision structures lean. A central state machine deals with exceptions requiring coordinated action of multiple control components. DCS implements the state machine by means of pulse schedule segments containing pre-programmed waveforms to define discharge goal and control

  15. Advanced tokamak burning plasma experiment

    International Nuclear Information System (INIS)

    A new reduced size ITER-RC superconducting tokamak concept is proposed with the goals of studying burn physics either in an inductively driven standard tokamak (ST) mode of operation, or in a quasi-steady state advanced tokamak (AT) mode sustained by non-inductive means. This is achieved by reducing the radiation shield thickness protecting the superconducting magnet by 0.34 m relative to ITER and limiting the burn mode of operation to pulse lengths as allowed by the TF coil warming up to the current sharing temperature. High gain (Q≅10) burn physics studies in a reversed shear equilibrium, sustained by RF and NB current drive techniques, may be obtained. (author)

  16. Plasma boundary phenomena in tokamaks

    International Nuclear Information System (INIS)

    The focus of this review is on processes occurring at the edge, and on the connection between boundary plasma - the scrape-off layer (SOL) and the radiating layer - and central plasma processes. Techniques used for edge diagnosis are reviewed and basic experimental information (ne and Te) is summarized. Simple models of the SOL are summarized, and the most important effects of the boundary plasma - the influence on the fuel particles, impurities, and energy - on tokamak operation dealt with. Methods of manipulating and controlling edge conditions in tokamaks and the experimental data base for the edge during auxiliary heating of tokamaks are reviewed. Fluctuations and asymmetries at the edge are also covered. (9 tabs., 134 figs., 879 refs.)

  17. Computational studies of tokamak plasmas

    International Nuclear Information System (INIS)

    Computational studies of tokamak plasmas are extensively advanced. Many computational codes have been developed by using several kinds of models, i.e., the finite element formulation of MHD equations, the time dependent multidimensional fluid model, and the particle model with the Monte-Carlo method. These codes are applied to the analyses of the equilibrium of an axisymmetric toroidal plasma (SELENE), the time evolution of the high-beta tokamak plasma (APOLLO), the low-n MHD stability (ERATO-J) and high-n ballooning mode stability (BOREAS) in the INTOR tokamak, the nonlinear MHD stability, such as the positional instability (AEOLUS-P), resistive internal mode (AEOLUS-I) etc., and the divertor functions. (author)

  18. Experimental study of the radial structure of turbulence with a ultra-fast sweeping reflectometer in ASDEX upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Medvedeva, Anna [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2, D-85748 Garching (Germany); Commissariat a l' Energie Atomique et aux Energies Alternatives, 13108 Saint Paul Lez Durance (France); Universite de Lorraine, 34 cours Leopold, 54000 Nancy (France); Technische Universitat Munchen, James-Franck-Strasse1, D-85748 Garching (Germany); Bottereau, Christine; Clairet, Frederic; Molina, Diego [Universite de Lorraine, 34 cours Leopold, 54000 Nancy (France); Conway, Garrard [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2, D-85748 Garching (Germany); Heuraux, Stephane [Commissariat a l' Energie Atomique et aux Energies Alternatives, 13108 Saint Paul Lez Durance (France); Stroth, Ulrich [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2, D-85748 Garching (Germany); Technische Universitat Munchen, James-Franck-Strasse1, D-85748 Garching (Germany)

    2014-07-01

    Confinement of fusion plasmas is restricted by anomalous transport where micro-turbulence is suspected to play a major role. Experimental documentation of this turbulence, its dependence on the plasma temperature, density, current will provide insights in the nature of this turbulence and the driving parameters. In this work advantage is taken of the ultra-fast sweep capabilities of the V and W band (50-110 GHz) reflectometers, developed by CEA, to record fast plasma turbulent events on ASDEX upgrade. The X-mode polarization will provide a rather large radial access to the plasma from the very edge to, under certain conditions, the center. The scope of the work is to exploit the specific strengths of the diagnostic in order to study the radial spectra of fluctuations, radial turbulence spreading and the fast dynamic profile evolution after confinement transitions or changes in the discharge control parameters. First experimental data obtained during the ASDEX upgrade campaign 2014 are presented.

  19. Experimental study of the radial structure of turbulence with an ultra-fast swept reflectometer in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Confinement of fusion plasmas is restricted by anomalous transport where micro-turbulence is suspected to play a major role. Experimental documentation of this turbulence, its dependence on the plasma temperature, density, current will provide insights in the nature of this turbulence and the driving parameters. In this work advantage is taken of the ultra-fast sweep capabilities of the V and W band (50-110 GHz) reflectometers, developed by CEA, to record fast plasma turbulent events on ASDEX Upgrade. The X-mode polarization provides a rather large radial access to the plasma from the very edge to, under certain conditions, the center. The scope of the work is to exploit the specific strengths of the diagnostic in order to study the radial spectra of fluctuations, radial turbulence spreading and the fast dynamic profile evolution after confinement transitions or changes in the discharge control parameters. The latest experimental data obtained during the ASDEX Upgrade campaign 2014 are presented.

  20. Summary discussion: An integrated advanced tokamak reactor

    International Nuclear Information System (INIS)

    The tokamak concept improvement workshop addressed a wide range of issues involved in the development of a more attractive tokamak. The agenda for the workshop progressed from a general discussion of the long-range energy context (with the objective being the identification of a set of criteria and ''figures of merit'' for measuring the attractiveness of a tokamak concept) to particular opportunities for the improvement of the tokamak concept. The discussions concluded with a compilation of research program elements leading to an improved tokamak concept

  1. STARFIRE: a commercial tokamak reactor

    International Nuclear Information System (INIS)

    The purpose of this document is to provide an interim status report on the STARFIRE project for the period of May to September 1979. The basic objective of the STARFIRE project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor

  2. LHCD experiments on tokamak CASTOR

    International Nuclear Information System (INIS)

    A short survey is given of the experimental activities at the small Prague tokamak CASTOR. They concern primarily the LH current drive using multijunction waveguide grills as launching antennae. During two last years the, efforts were focused on a study of the electrostatic and magnetic fluctuations under conditions of combined inductive/LHCD regimes and of the relation of the level of these fluctuations to the anomalous particles transport in tokamak CASTOR. Results of the study are discussed in some detail. (author). 24 figs., 51 refs

  3. The European Integrated Tokamak Modelling Effort: Achievements and First Physics Results

    International Nuclear Information System (INIS)

    Full text: The achievements and first physics results are presented of the European Integrated Tokamak Modelling Task Force (EFDA ITM-TF) effort, aiming at providing a standardized platform and an integrated modelling suite of validated numerical codes, for the simulation and prediction of a complete plasma discharge in any tokamak. The framework developed by the ITM-TF, based on a generic datastructure enclosing both simulated and experimental data, allowed for the development of sophisticated integrated simulations (workflows) for physics application. Those include the European Transport Solver (ETS), incorporating a sophisticated module for synergy effects between heating schemes, several equilibrium modules, pellets, impurities, neutrals, sawteeth and NTM modules, a variety of simple transport modules and neoclassical modules. The ETS workflows have been subject to an extensive verification and validation laying the foundations for the use of ETS for both predictive and interpretative transport simulations as well as scenario modeling on present devices and ITER. The equilibrium reconstruction and linear MHD stability simulation chain is being applied for production runs on several devices. In particular, an analysis of the edge MHD stability of ASDEX Upgrade type-I ELMy H-mode discharges and ITER hybrid scenario was performed, revealing the stabilizing effect of an increased Shafranov shift on edge modes. A successful benchmark among EC beam/ray-tracing codes (C3PO, GRAY, TORAY-FOM, TORBEAM, TRAVIS) has been performed in the ITM framework for an ITER case for different launching conditions from the Equatorial Launcher, showing good agreement of the computed absorbed power and driven current. Simulations performed within the ITM infrastructure with the turbulence code GEM for a JET hybrid discharge and the comparison of the simulated anomalous fluxes with TRANSP are presented, addressing in particular, the effect of the E x B shear on the thermal and particle

  4. First operations with the new Collective Thomson Scattering diagnostic on the Frascati Tokamak Upgrade device

    Science.gov (United States)

    Bin, W.; Bruschi, A.; D'Arcangelo, O.; Castaldo, C.; De Angeli, M.; Figini, L.; Galperti, C.; Garavaglia, S.; Granucci, G.; Grosso, G.; Korsholm, S. B.; Lontano, M.; Mellera, V.; Minelli, D.; Moro, A.; Nardone, A.; Nielsen, S. K.; Rasmussen, J.; Simonetto, A.; Stejner, M.; Tartari, U.

    2015-10-01

    Anomalous emissions were found over the last few years in spectra of Collective Thomson Scattering (CTS) diagnostics in tokamak devices such as TEXTOR, ASDEX and FTU, in addition to real CTS signals. The signal frequency, down-shifted with respect to the probing one, suggested a possible origin in Parametric Decay Instability (PDI) processes correlated with the presence of magnetic islands and occurring for pumping wave power levels well below the threshold predicted by conventional models. A threshold below or close to the Electron Cyclotron Resonance Heating (ECRH) power levels could limit, under certain circumstances, the use of the ECRH in fusion devices. An accurate characterization of the conditions for the occurrence of this phenomenon and of its consequences is thus of primary importance. Exploiting the front-steering configuration available with the real-time launcher, the implementation of a new CTS setup now allows studying these anomalous emission phenomena in FTU under conditions of density and wave injection geometry that are more similar to those envisaged for CTS in ITER. The upgrades of the diagnostic are presented as well as a few preliminary spectra detected with the new system during the very first operations in 2014. The present work has been carried out under an EUROfusion Enabling Research project. A shorter version of this contribution is due to be published in PoS at: 1st EPS conference on Plasma Diagnostics

  5. Possibilities for breakeven and ignition of D-3He fusion fuel in a near term tokamak

    International Nuclear Information System (INIS)

    The recent realization that the moon contains a large amount of the isotope 3He has rekindled interest in the D-3He fuel cycle. In this study we consider the feasibility of investigating D-3He reactor plasma conditions in a tokamak of the NET/INTOR class. We have found that, depending on the energy confinement scaling law, energy breakeven may be achieved without significant modification to the NET design. The best results are for the more optimistic ASDEX H-mode scaling law. Kaye-Goldston scaling with a modest improvement due to the H-mode is more pessimistic and makes achieving breakeven more difficult. Significant improvement in Q (ratio of the fusion power to the injected power), or the ignition margin, can be achieved by taking advantage of the much reduced neutron production of the D-3He fuel cycle. Removal of the tritium producing blanket and replacing the inboard neutron shield by a thinner shield optimized for the neutron spectrum in D-3He allows the plasma to be increased without changing the magnetic field at the toroidal field magnet. This allows the plasma to achieve higher beta and Q values up to about 3. The implications of D-3He operation for fast ion loss, neutron shielding, heat loads on the first wall and divertor, plasma refuelling, changes to the poloidal field coil system, and pumping of the helium from the vacuum chamber are considered in the report. (orig.)

  6. Soft-computing approach to plasma evolution tracking in tokamak reactors

    Science.gov (United States)

    Morabito, Francesco C.

    1997-10-01

    Qualitative information about the structure of a mapping can surely be of help in learning a mapping by a collection of input-output pairs. However, there are conditions in which time and some other constraints make guessing the only plausible means for interpreting data. In this paper, the problem of the plasma boundary reconstruction in 'Tokamak' nuclear fusion rectors is assessed. The problem is formulated as an inverse 'identification' problem and the mapping is derived by a properly generated database of simulated experiments. Real data coming from experiments are also available to validate both numerically generated data and extracted model. The identification problem is solved for two different databases by using neural networks and more conventional models. The introduction of techniques derived from soft computing is shown to improve the performance in various respects. Dynamic identification systems appear to be rather demanding also for such systems, for the need of rapidly interpreting real time data for discharge control. Soft computing approaches may yet yield some low cost ways to take decisions during plasma evolution. The approximate analysis of experimental data could also improve the knowledge on the particular problem allowing an evolution of the knowledge base. Experimental data related to ASDEX-Upgrade machine are presented in this work and preliminary processed. Soft computing techniques also allow to simply get ideas about two other interesting problems in plasma engineering, namely, the fault tolerance and the minimization of the number of sensors.

  7. First operations with the new Collective Thomson Scattering diagnostic on the Frascati Tokamak Upgrade device

    International Nuclear Information System (INIS)

    Anomalous emissions were found over the last few years in spectra of Collective Thomson Scattering (CTS) diagnostics in tokamak devices such as TEXTOR, ASDEX and FTU, in addition to real CTS signals. The signal frequency, down-shifted with respect to the probing one, suggested a possible origin in Parametric Decay Instability (PDI) processes correlated with the presence of magnetic islands and occurring for pumping wave power levels well below the threshold predicted by conventional models. A threshold below or close to the Electron Cyclotron Resonance Heating (ECRH) power levels could limit, under certain circumstances, the use of the ECRH in fusion devices. An accurate characterization of the conditions for the occurrence of this phenomenon and of its consequences is thus of primary importance. Exploiting the front-steering configuration available with the real-time launcher, the implementation of a new CTS setup now allows studying these anomalous emission phenomena in FTU under conditions of density and wave injection geometry that are more similar to those envisaged for CTS in ITER. The upgrades of the diagnostic are presented as well as a few preliminary spectra detected with the new system during the very first operations in 2014. The present work has been carried out under an EUROfusion Enabling Research project. A shorter version of this contribution is due to be published in PoS at: 1st EPS conference on Plasma Diagnostics

  8. Plasma geometry and current profile identification on ASDEX Upgrade using an integrated equilibrium generation and interpretation system

    International Nuclear Information System (INIS)

    The identification of ideal MHD equilibrium states at ASDEX Upgrade is the starting point for interpreting any diagnostic data dependent on knowledge of the flux surface geometry. The method of Function Parameterization (FP) starts with the Monte Carlo generation of a simulated equilibrium database, regression analysis of which yields simple functional representations of plasma geometry whose arguments are information-rich, uncorrelated linear combinations of simulated diagnostic signals. Once calculated, these FP expressions can be rapidly evaluated using experimental data. FP using magnetic data is in routine realtime use on ASDEX Upgrade for plasma position and shape control. An extension to FP using MSE data has recently been developed for realtime identification and control of the current profile on ASDEX Upgrade. Post-discharge interpretive equilibrium solutions are generated by the CLISTE code, which best fits a set of specified diagnostic data. CLISTE can include kinetic data and poloidal halo currents in the scrape-off layer as constraints on the equilibrium solution, a valuable feature which has been applied to ELM analysis. The code has recently been extended to interpret dB/dt data from magnetics and dγ/dt data from MSE to yield a best fit solution to the time derivative of the Grad-Shafranov equation -Δ*∂ψ/∂t = 2πμ0R ∂/∂t jφ. The ∂ψ/∂t solution is used to calculate the flux surface averaged profile which can be used to calculate current drive from auxiliary heating methods via the equation aux.heating = equil - σ - boot where boot is calculated from kinetic profiles and neoclassical theory and equil is an equilibrium output. This technique is being applied to analyse current profile modification by off-axis NBI on ASDEX Upgrade. (author)

  9. Joint research using small tokamaks

    Czech Academy of Sciences Publication Activity Database

    Gryaznevich, M.P.; Del Bosco, E.; Malaquias, A.; Mank, G.; Van Oost, G.; He, Yexi; Hegazy, H.; Hirose, A.; Hron, Martin; Kuteev, B.; Ludwig, G.O.; Nascimento, I.C.; Silva, C.; Vorobyev, G.M.

    2005-01-01

    Roč. 45, č. 10 (2005), S245-S254. ISSN 0029-5515. [Fusion Energy Conference contributions. Vilamoura, 1.11.2004-6.11.2004] Institutional research plan: CEZ:AV0Z20430508 Keywords : small tokamaks * thermonuclear fusion Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.418, year: 2005

  10. RF preionization in Tokamak thor

    International Nuclear Information System (INIS)

    During the study of the RF preionization in Tokamak Thor was observed that the starting of the plasma and its time behaviour were correlated with the presence of resonance conditions both at the electron cyclotron frequency Ωsub(deg) and at its sub-harmonics Ωsub(deg)/n. These results are supported by a simple qualitative calculation

  11. Integral torque balance in tokamaks

    International Nuclear Information System (INIS)

    The study is aimed at clarifying the balance between the sinks and sources in the problem of intrinsic plasma rotation in tokamaks reviewed recently by deGrassie (2009 Plasma Phys. Control. Fusion 51 124047). The integral torque on the toroidal plasma is calculated analytically using the most general magnetohydrodynamic (MHD) plasma model taking account of plasma anisotropy and viscosity. The contributions due to several mechanisms are separated and compared. It is shown that some of them, though, possibly, important in establishing the rotation velocity profile in the plasma, may give small input into the integral torque, but an important contribution can come from the magnetic field breaking the axial symmetry of the configuration. In tokamaks, this can be the error field, the toroidal field ripple or the magnetic perturbation created by the correction coils in the dedicated experiments. The estimates for the error-field-induced electromagnetic torque show that the amplitude of this torque is comparable to the typical values of torques introduced into the plasma by neutral beam injection. The obtained relations allow us to quantify the effect that can be produced by the existing correction coils in tokamaks on the plasma rotation, which can be used in experiments to study the origin and physics of intrinsic rotation in tokamaks. Several problems are proposed for theoretical studies and experimental tests.

  12. Edge plasma diagnostics in tokamaks

    Czech Academy of Sciences Publication Activity Database

    Stöckel, Jan; Brotánková, Jana; Hron, Martin; Adámek, Jiří; Ďuran, Ivan; Van Oost, G.; Peleman, P.; Gunn, J.; Devynck, P.; Martines, E.; Schrittwieser, R.; Kocan, M.

    Kudowa Zdrój : -, 2006, s. 910-935. [Sixth International Workshop and Summer School Towards Fusion Energy - Plasma Physics, Diagnostics, Spin-offs. Kudowa Zdrój (PL), 18.09.2006-22.09.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * diagnostics * heating Subject RIV: BL - Plasma and Gas Discharge Physics

  13. Tokamak experimental power reactor studies

    International Nuclear Information System (INIS)

    The principal results of a scoping and project definition study for the Tokamak Experimental Power Reactor are presented. Objectives are discussed; a preliminary conceptual design is described; detailed parametric, survey and sensitivity studies are presented; and research and development requirements are outlined. (U.S.)

  14. Assembly of Aditya upgrade tokamak

    International Nuclear Information System (INIS)

    The existing Aditya tokamak, a medium sized tokamak with limiter configuration is being upgraded to a tokamak with divertor configuration. At present the existing ADITYA tokamak has been dismantled up to bottom plinth on which the whole assembly of toroidal field (TF) coils and vacuum vessel rested. The major components of ADITYA machine includes 20 TF coils and its structural components, 9 Ohmic coils and its clamps, 4 BV coils and its clamps as well as their busbar connections, vacuum vessel and its supports and buckling cylinder, which are all being dismantled. The re-assembly of the ADITYA Upgrade tokamak started with installation and positioning of new buckling cylinder and central solenoid (TR1) coil. After that the inner sections of TF coils are placed following which in-situ winding, installation, positioning and support mounting of two pairs of new inner divertor coils have been carried out. After securing the TF coils with top I-beams the new torus shaped vacuum vessel with circular cross-section in 2 halves have been installed. The assembly of TF structural components such as top and bottom guiding wedges, driving wedges, top and bottom compression ring, inner and outer fish plates and top inverted triangle has been carried out in an appropriate sequence. The assembly of outer sections of TF coils along with the proper placements of top auxiliary TR and vertical field coils with proper alignment and positioning with the optical metrology instrument mainly completes the reassembly. Detailed re-assembly steps and challenges faced during re-assembly will be discussed in this paper. (author)

  15. Deformation measurement of internal components of ASDEX Upgrade using optical strain sensors

    Energy Technology Data Exchange (ETDEWEB)

    Vorpahl, C., E-mail: christian.vorpahl@ipp.mpg.de [Max Planck Institute for Plasma Physics, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany); Suttrop, W.; Ebner, M.; Streibl, B.; Zohm, H. [Max Planck Institute for Plasma Physics, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany)

    2013-10-15

    Highlights: ► A fibre-optic measurement for the deformation of in-vessel components has successfully been installed and commissioned at ASDEX Upgrade. ► This technology has thereby been qualified for in-vessel use at experimental nuclear fusion devices. ► The sensors were tested for their neutron tolerance and vacuum compatibility. ► Installation was done by copper–steel laser beam welding. ► The temporal and spatial resolutions of the system are sufficient to resolve oscillations due to internal coils and plasma disruptions. -- Abstract: A fibre-optic measurement system to analyse the deformation of in-vessel components has successfully been developed, installed and commissioned at ASDEX Upgrade (AUG). This technology has thereby been qualified for in-vessel use at experimental fusion devices. AUG is equipped with an internal conductor for passive plasma stabilisation called the Passive Stabilisation Loop (PSL), on which the recently installed 16 internal coils (B-coils) are directly mounted. The PSL structure is highly prone to vibrations, and the risk of resonant oscillations in response to B-coil induced forces necessitated the development of the present diagnostic. The diagnostic system consists of 34 fibre-optic strain sensors incorporated in two glass fibres. It is completely insensitive to electromagnetic disturbances. The fibres are customised to avoid inconvenient excess fibre length in the vacuum vessel. They were tested for their neutron tolerance and vacuum compatibility prior to installation. The actual sensors are embedded in stainless steel carriers that were attached to the PSL, which is made of copper, by laser welding. Appropriate welding parameters were determined in view of the metallurgical dissimilarity. The weld quality was approved by tensile tests and microscopic investigations. Accurate in-vessel positioning of the sensors was assured using a 3D measurement system and coordinates from CAD. The data acquisition allows a

  16. Experimental tests of confinement scale invariance on JET, DIIID, ASDEX Upgrade and CMOD

    International Nuclear Information System (INIS)

    An international collaboration between JET, DIIID, AUG and CMOD has resulted in four sets of Tokamak discharges which are approximately identical as regards a set of dimensionless plasma variables. The data demonstrates some measure of scale invariance of local and global confinement but a more accurate matching of scaled density, power etc. is required to make firmer conclusions. (author)

  17. Full Tokamak discharge simulation and kinetic plasma profile control for ITER

    International Nuclear Information System (INIS)

    Understanding non-linearly coupled physics between plasma transport and free-boundary equilibrium evolution is essential to operating future tokamak devices, such as ITER and DEMO, in the advanced tokamak operation regimes. To study the non-linearly coupled physics, we need a simulation tool which can self-consistently calculate all the main plasma physics, taking the operational constraints into account. As the main part of this thesis work, we have developed a full tokamak discharge simulator by combining a non-linear free-boundary plasma equilibrium evolution code, DINA-CH, and an advanced transport modelling code, CRONOS. This tokamak discharge simulator has been used to study the feasibility of ITER operation scenarios and several specific issues related to ITER operation. In parallel, DINA-CH has been used to study free-boundary physics questions, such as the magnetic triggering of edge localized modes (ELMs) and plasma dynamic response to disturbances. One of the very challenging tasks in ITER, the active control of kinetic plasma profiles, has also been studied. In the part devoted to free-boundary tokamak discharge simulations, we have studied dynamic responses of the free-boundary plasma equilibrium to either external voltage perturbations or internal plasma disturbances using DINA-CH. Firstly, the opposite plasma behaviour observed in the magnetic triggering of ELMs between TCV and ASDEX Upgrade has been investigated. Both plasmas experience similar local flux surface expansions near the upper G-coil set and passive stabilization loop (PSL) when the ELMs are triggered, due to the presence of the PSLs located inside the vacuum vessel of ASDEX Upgrade. Secondly, plasma dynamic responses to strong disturbances anticipated in ITER are examined to study the capability of the feedback control system in rejecting the disturbances. Specified uncontrolled ELMs were controllable with the feedback control systems. However, the specifications for fast H-L mode

  18. Transport of Dust Particles in Tokamak Devices

    Energy Technology Data Exchange (ETDEWEB)

    Pigarov, A Y; Smirnov, R D; Krasheninnikov, S I; Rognlien, T D; Rozenberg, M

    2006-06-06

    Recent advances in the dust transport modeling in tokamak devices are discussed. Topics include: (1) physical model for dust transport; (2) modeling results on dynamics of dust particles in plasma; (3) conditions necessary for particle growth in plasma; (4) dust spreading over the tokamak; (5) density profiles for dust particles and impurity atoms associated with dust ablation in tokamak plasma; and (6) roles of dust in material/tritium migration.

  19. Microwave Tokamak Experiment: Overview and status

    International Nuclear Information System (INIS)

    The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. 3 figs., 3 tabs

  20. Bootstrap Current in Spherical Tokamaks

    Institute of Scientific and Technical Information of China (English)

    王中天; 王龙

    2003-01-01

    Variational principle for the neoclassical theory has been developed by including amomentum restoring term in the electron-electron collisional operator, which gives an additionalfree parameter maximizing the heat production rate. All transport coefficients are obtained in-cluding the bootstrap current. The essential feature of the study is that the aspect ratio affects thefunction of the electron-electron collision operator through a geometrical factor. When the aspectratio approaches to unity, the fraction of circulating particles goes to zero and the contribution toparticle flux from the electron-electron collision vanishes. The resulting diffusion coefficient is inrough agreement with Hazeltine. When the aspect ratio approaches to infinity, the results are inagreement with Rosenbluth. The formalism gives the two extreme cases a connection. The theoryis particularly important for the calculation of bootstrap current in spherical tokamaks and thepresent tokamaks, in which the square root of the inverse aspect ratio, in general, is not small.

  1. Comprehensive numerical modelling of tokamaks

    International Nuclear Information System (INIS)

    We outline a plan for the development of a comprehensive numerical model of tokamaks. The model would consist of a suite of independent, communicating packages describing the various aspects of tokamak performance (core and edge transport coefficients and profiles, heating, fueling, magnetic configuration, etc.) as well as extensive diagnostics. These codes, which may run on different computers, would be flexibly linked by a user-friendly shell which would allow run-time specification of packages and generation of pre- and post-processing functions, including workstation-based visualization of output. One package in particular, the calculation of core transport coefficients via gyrokinetic particle simulation, will become practical on the scale required for comprehensive modelling only with the advent of teraFLOP computers. Incremental effort at LLNL would be focused on gyrokinetic simulation and development of the shell

  2. Frascati Tokamak transformer switching system

    International Nuclear Information System (INIS)

    Plasma ionization and heating, in the Frascati Tokamak, is obtained generating an emf along the plasma column, by switching the dc current flowing in the Tokamak transformer. 30 kA flowing in the 60 mH transformer inductance must be commutated into a resistance to generate 40 kV across the transformer itself. Studies and tests to solve this problem have been conducted, on different types of breakers, in cooperation between Tecnomasio Italiano Brown Boveri, Milan and Laboratori Gas Ionizzati, Frascati. Satisfactory results have finally been obtained using a DLF commercial air blast breaker in a chopper type circuit. A capacitor bank in parallel to the breaker is discharged immediately after the contacts separation and the arc in the switching element is extinguished at the first current zero. A saturable reactance in series with the breaker reduces the current decay rate to allow sufficient deionization time

  3. Burn Control Mechanisms in Tokamaks

    Science.gov (United States)

    Hill, Maxwell; Stacey, Weston

    2013-10-01

    Burn control and passive safety in accident scenarios will be an important design consideration in future tokamaks, especially those used as a neutron source for fusion-fission hybrid reactors, such as the Subcritical Advanced Burner Reactor (SABR) concept. At Georgia Tech, we are developing a new burning plasma dynamics code to investigate passive safety mechanisms that could prevent power excursions in tokamak reactors. This code solves the coupled set of balance equations governing burning plasmas in conjunction with a two-point SOL-divertor model. Predictions have been benchmarked against data from DIII-D. We are examining several potential negative feedback mechanisms to limit power excursions: i) ion-orbit loss, ii) thermal instabilities, iii) the degradation of alpha-particle confinement resulting from ripples in the toroidal field, iv) modifications to the radial current profile, v) ``divertor choking'' and vi) Type 1 ELMs.

  4. Equilibrium Reconstruction in EAST Tokamak

    Institute of Scientific and Technical Information of China (English)

    QIAN Jinping; WAN Baonian; L. L. LAO; SHEN Biao; S. A. SABBAGH; SUN Youwen; LIU Dongmei; XIAO Singjia; REN Qilong; GONG Xianzu; LI Jiangang

    2009-01-01

    Reconstruction of experimental axisymmetric equilibria is an important part of toka-mak data analysis. Fourier expansion is applied to reconstruct the vessel current distribution in EFIT code. Benchmarking and testing calculations are performed to evaluate and validate this algorithm. Two cases for circular and non-circular plasma discharges are presented. Fourier ex-pansion used to fit the eddy current is a robust method and the real time EFIT can be introduced to the plasma control system in the coming campaign.

  5. Shear Alfven waves in tokamaks

    International Nuclear Information System (INIS)

    Shear Alfven waves in an axisymmetric tokamak are examined within the framework of the linearized ideal MHD equations. Properties of the shear Alfven continuous spectrum are studied both analytically and numerically. Implications of these results in regards to low frequency rf heating of toroidally confined plasmas are discussed. The structure of the spatial singularities associated with these waves is determined. A reduced set of ideal MHD equations is derived to describe these waves in a very low beta plasma

  6. Machine safety issues with respect to the extension of ECRH systems at ASDEX Upgrade

    Directory of Open Access Journals (Sweden)

    Schuberta Martin

    2015-01-01

    Full Text Available The beam intensity of electron cyclotron resonance heating at ASDEX Upgrade has the potential to seriously damage in-vessel components, whenever not fully absorbed by the plasma. Operation is, therefore, interlocked with both plasma current and density above a given threshold. Microwave protection detectors installed in several ports on the low field side switch the heating system off, in case the stray radiation exceeds a given threshold. During regular inspections, however, damages were reported in the vicinity of the launchers and in particular around the tiles of the heat shield. On one hand, it was found that insulating material, which may not face the plasma, degraded due to millimetre wave absorption. The waves entered the free space behind the heat shield through gaps. On the other hand, local damage even of metallic components was observed on surfaces, which were directly exposed to the microwave beam. Polarisation errors, which led to a local shine through of significant beam power, were responsible. We note that this happened mainly on the high field side in a certain distance to the microwave protection detectors, which were not triggered by the events. In order to increase the level of protection, we identify three necessary measures: Firstly, polarisation control is to be automated such, that mode content and shine through can be monitored. Secondly, by installing additional detectors, the spatial coverage of stray radiation monitoring is enlarged. Thirdly, the heat shield tiles will be redesigned in order to increase the shielding against millimetre waves.

  7. Investigation of scrape-off layer and divertor heat transport in ASDEX Upgrade L-mode

    Science.gov (United States)

    Sieglin, B.; Eich, T.; Faitsch, M.; Herrmann, A.; Scarabosio, A.; the ASDEX Upgrade Team

    2016-05-01

    Power exhaust is one of the major challenges for the development of a fusion power plant. Predictions based upon a multimachine database give a scrape-off layer power fall-off length {λq}≤slant 1 mm for large fusion devices such as ITER. The power deposition profile on the target is broadened in the divertor by heat transport perpendicular to the magnetic field lines. This profile broadening is described by the power spreading S. Hence both {λq} and S need to be understood in order to estimate the expected divertor heat load for future fusion devices. For the investigation of S and {λq} L-Mode discharges with stable divertor conditions in hydrogen and deuterium were conducted in ASDEX Upgrade. A strong dependence of S on the divertor electron temperature and density is found which is the result of the competition between parallel electron heat conductivity and perpendicular diffusion in the divertor region. For high divertor temperatures it is found that the ion gyro radius at the divertor target needs to be considered. The dependence of the in/out asymmetry of the divertor power load on the electron density is investigated. The influence of the main ion species on the asymmetric behaviour is shown for hydrogen, deuterium and helium. A possible explanation for the observed asymmetry behaviour based on vertical drifts is proposed.

  8. Mechanical braking system for the pulsed power supply system of ASDEX Upgrade

    International Nuclear Information System (INIS)

    Highlights: ► Compact and innovative solution for dumping of large kinetic energy. ► Small mass of energy converter at the shaft due to circulating storage medium. ► Design of the active parts ensures flat torque/power characteristics. ► Also suitable for spending a great part of operating life in “Freewheeling” mode. -- Abstract: A few years ago, IPP reviewed the safety of the ASDEX Upgrade pulsed power supply system. Two critical sub-systems had been identified: The (electrical) braking system for the flywheel generators and the oil lubrication system for the shaft bearings. A simultaneous failure of these two systems may lead to severe damages and could have consequences for the safety of operating personnel. Therefore a second, independent braking possibility for every generator was stipulated. Especially the challenges adapting a dynamometer, originally designed for motor test benches, towards a plant safety system for generator EZ4 will be described in the paper. Further on, the paper will present the problems, implementing such a system into an existing installation, including the calculation of the required supporting structure, balancing of the extended shaft line and required water cooling and control. Finally it will report on the performance achieved during operation

  9. Density response to central electron heating: theoretical investigations and experimental observations in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Theory of ion temperature gradient (ITG) and trapped electron modes (TEMs) is applied to the study of particle transport in experimental conditions with central electron heating. It is shown that in the unstable domain of TEMs, the electron thermo diffusive flux is directed outwards. By means of such a flux, a mechanism is identified likely to account for density flattening with central electron heating. Theoretical predictions are compared with experimental observations in ASDEX Upgrade. A parameter domain (including L- and H-mode plasmas) is identified, in which flattening with central electron heating is observed in the experiments. In general, this domain turns out to be the same domain in which the dominant plasma instability is a TEM. On the contrary, the dominant instability is an ITG in plasmas whose density profile is not affected significantly by central electron heating. The flattening predicted by quasi-linear theory for low density L-mode plasmas is too small compared to the experimental observations. At very high density, even when the dominant instability is an ITG, electron heating can provide density flattening, via the coupling with the ion heat channel. In these conditions the anomalous diffusivity increases in response to the increased ion heat flux, while the large collisionality makes the anomalous pinch small and the Ware pinch important. (author)

  10. Effect of radial electric field and ripple on edge neutral beam ion distribution in ASDEX Upgrade

    Science.gov (United States)

    Hynönen, V.; Kurki-Suonio, T.; Suttrop, W.; Stäbler, A.; ASDEX Upgrade Team

    2008-03-01

    The neutral beam injected fast ion distribution at the ASDEX Upgrade edge region is studied focusing on the difference between co- and counter-injected neutral beams. The slowing-down distribution of beam ions is simulated using the orbit-following Monte Carlo code ASCOT. The edge fast ion density and its gradient are higher for counter-injection than for co-injection. Also the distribution in the velocity space is different: for co-injection, there exists a population of untrapped particles which for counter-injection is found only when the effect of a non-constant, experimentally obtained quiescent H-mode radial electric field is included in the simulation. Toroidal ripple removes ions having small particle pitch, thereby reducing the density and density gradient, whereas the radial electric field has the opposite effect. Including simultaneously the effects of both ripple and the radial electric field restores the distribution close to the ideal case where both of them are neglected. The radial electric field is found to squeeze the orbit of a counter-injected neutral beam ion but to widen the orbit of a co-injected ion, and to cause transitions in the orbit topologies which are reflected in the fast ion distribution.

  11. Overview on plasma operation with a full tungsten wall in ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Neu, R., E-mail: Rudolf.Neu@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, EURATOM-Association, D-85748 Garching (Germany); Kallenbach, A.; Balden, M.; Bobkov, V. [Max-Planck-Institut für Plasmaphysik, EURATOM-Association, D-85748 Garching (Germany); Coenen, J.W. [Institut für Energie- und Klimaforschung IV, Forschungszentrum Jülich, TEC, Association EURATOM-FZJ, D-52425 Jülich (Germany); Drube, R.; Dux, R.; Greuner, H.; Herrmann, A.; Hobirk, J. [Max-Planck-Institut für Plasmaphysik, EURATOM-Association, D-85748 Garching (Germany); Höhnle, H. [Institut für Plasmaforschung, Universität Stuttgart, Stuttgart (Germany); Krieger, K.; Kočan, M.; Lang, P.; Lunt, T.; Maier, H.; Mayer, M.; Müller, H.W.; Potzel, S.; Pütterich, T. [Max-Planck-Institut für Plasmaphysik, EURATOM-Association, D-85748 Garching (Germany); and others

    2013-07-15

    Operation with all tungsten plasma facing components has become routine in ASDEX Upgrade. The conditioning of the device is strongly simplified and short glow discharges are used only on a daily basis. The long term fuel retention was reduced by more than a factor of 5 as demonstrated in gas balance as well as in post mortem analyses. Injecting nitrogen for radiative cooling, discharges with additional heating power up to 23 MW have been achieved, providing good confinement (H98{sub y2} = 1), divertor power loads around 5 MW m{sup −2} and divertor temperatures below 10 eV. ELM mitigation by pellet ELM pacemaking or magnetic perturbation coils reduces the deposited energy during ELMs, but also keeps the W density at the pedestal low. As a recipe to keep the central W concentration sufficiently low, central (wave) heating is well established and low density H-Modes could be re-established with the newly available ECRH power of up to 4 MW. The ICRH induced W sources could be strongly reduced by applying boron coatings to the poloidal guard limiters.

  12. Comparison of local transport studies with the profile consistency concept for ASDEX pellet-refuelled discharges

    International Nuclear Information System (INIS)

    Strongly peaked electron density profiles have been obtained in ASDEX by different refuelling methods: pellet fuelling, NBI counter-injection and recently by reduced gas puff fuelling scenarios. These discharges show in common increased density limits, a canonical electron temperature profile independent of the density profile and an improvement of the particle and energy confinement. Whereas the changes in particle transport are not fully understood, local transport analyses point out that the improved energy transport can be explained by reduced ion conduction losses coming close to the neoclassical ones. The different results for the ion transport with flat and peaked density profiles are quantitatively consistent with that expected from ηj-driven modes. So all cases showing confinement improvement through density peaking correspond to ηj and ηe) E with ηe for flat density profiles and the extension of the linear dependence for peaked ones in OH discharges then fits with a continuing inverse density dependence of the electron thermal diffusivity χe is also in agreement with τE enhancement when going from D+ to H+ ions. With additional heating χe is largely responsible for the confinement degradation in the L-mode and again the improvement at the H-mode transition. Near the plasma boundary χe is higher than χi in all cases investigated. (author). 9 refs, 7 figs

  13. Mechanical braking system for the pulsed power supply system of ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Käsemann, C.-P., E-mail: c.p.kaesemann@ipp.mpg.de [Max Planck Institute for Plasma Physics, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany); Huart, M. [Max Planck Institute for Plasma Physics, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany); Michel Huart Personal Coaching and Consulting, Georgenschwaigstraße 23 RG, 80807 München (Germany); Stobbe, F.; Goldstein, I.; Sigalov, A. [Max Planck Institute for Plasma Physics, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany); Sachs, E. [Siemens AG, Industrial Automation Systems, Gleiwitzer Straße 555, 90475 Nürnberg (Germany); Perk, E. [Piper Test and Measurement Ltd., The Barn, Bilsington, Ashford, Kent TN25 7JT, England (United Kingdom)

    2013-10-15

    Highlights: ► Compact and innovative solution for dumping of large kinetic energy. ► Small mass of energy converter at the shaft due to circulating storage medium. ► Design of the active parts ensures flat torque/power characteristics. ► Also suitable for spending a great part of operating life in “Freewheeling” mode. -- Abstract: A few years ago, IPP reviewed the safety of the ASDEX Upgrade pulsed power supply system. Two critical sub-systems had been identified: The (electrical) braking system for the flywheel generators and the oil lubrication system for the shaft bearings. A simultaneous failure of these two systems may lead to severe damages and could have consequences for the safety of operating personnel. Therefore a second, independent braking possibility for every generator was stipulated. Especially the challenges adapting a dynamometer, originally designed for motor test benches, towards a plant safety system for generator EZ4 will be described in the paper. Further on, the paper will present the problems, implementing such a system into an existing installation, including the calculation of the required supporting structure, balancing of the extended shaft line and required water cooling and control. Finally it will report on the performance achieved during operation.

  14. Non-monotonic growth rates of sawtooth precursors evidenced with a new method on ASDEX Upgrade

    Science.gov (United States)

    Vezinet, D.; Igochine, V.; Weiland, M.; Yu, Q.; Gude, A.; Meshcheriakov, D.; Sertoli, M.; the Asdex Upgrade Team; the EUROfusion MST1 Team

    2016-08-01

    This paper describes a new method to derive, from soft x-ray (SXR) tomography, robust estimates of the core displacement, growth rate and frequency of a 1/1 sawtooth crash precursor. The method is valid for very peaked SXR profiles and is robust against both the inversion algorithm and the presence of tungsten in a rotating plasma. Three typical ASDEX Upgrade crashes are then analysed. In all cases a postcursor is observed, suggesting incomplete reconnection. Despite different dynamics, in all three cases the growth rate of the core displacement shows similar features. First, it is not constant, supporting the idea of non-linear growth. Second, it can be divided into clearly identified phases with quasi-constant growth rates, suggesting sudden change of growth regime rather than smooth transitions. Third, its evolution is non-monotonic, with phases of accelerated growth followed by damped phases. This damping is interpreted for two cases respectively as an effect of fast ions and of mode coupling, based on the result of a MHD simulation. The mode frequency is observed in all cases to be closely related to the plasma bulk rotation profile, with little or no visible effect of the electron diamagnetic drift frequency. The onset criterion could not be clearly identified and it is shown that the role of the pressure gradient is not as expected from a naive extrapolation of the linear stability theory.

  15. A new compact solid-state neutral particle analyser at ASDEX Upgrade: Setup and physics modeling

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, P. A.; Blank, H.; Geiger, B.; Mank, K.; Martinov, S.; Ryter, F.; Weiland, M.; Weller, A. [Max-Planck-Institut für Plasmaphysik, Garching (Germany)

    2015-07-15

    At ASDEX Upgrade (AUG), a new compact solid-state detector has been installed to measure the energy spectrum of fast neutrals based on the principle described by Shinohara et al. [Rev. Sci. Instrum. 75, 3640 (2004)]. The diagnostic relies on the usual charge exchange of supra-thermal fast-ions with neutrals in the plasma. Therefore, the measured energy spectra directly correspond to those of confined fast-ions with a pitch angle defined by the line of sight of the detector. Experiments in AUG showed the good signal to noise characteristics of the detector. It is energy calibrated and can measure energies of 40-200 keV with count rates of up to 140 kcps. The detector has an active view on one of the heating beams. The heating beam increases the neutral density locally; thereby, information about the central fast-ion velocity distribution is obtained. The measured fluxes are modeled with a newly developed module for the 3D Monte Carlo code F90FIDASIM [Geiger et al., Plasma Phys. Controlled Fusion 53, 65010 (2011)]. The modeling allows to distinguish between the active (beam) and passive contributions to the signal. Thereby, the birth profile of the measured fast neutrals can be reconstructed. This model reproduces the measured energy spectra with good accuracy when the passive contribution is taken into account.

  16. Critical issues identified by the ASDEX Upgrade edge and divertor modelling

    International Nuclear Information System (INIS)

    A detailed comparison between the ASDEX Upgrade (AUG) experimental data and results of the SOLPS 2D edge code simulations has recently been performed. High quality upstream profiles of electron density and ion and electron temperatures in the scrape-off layer (SOL) of AUG have been collected for two shots with different upstream collisionalities: a low density ELMy H-mode shot (low collisionality) and a medium density Ohmic shot (higher collisionality). A generally broad agreement, within a factor 2, considering basic parameters characterising the divertor, has been reached between simulations and experiment. In both Ohmic and H-mode shots, however, the tendency of SOLPS solutions to underestimate the divertor electron temperature and overestimate its density has been reliably established. Two main possible causes of the discrepancies have been considered: some deficiencies in the neutral modelling (e.g. missing atomic and molecular reactions in EIRENE, the Monte-Carlo neutral part of SOLPS), and the presence of a significant population of supra-thermal ions and electrons in the SOL and divertor plasma. The results of dedicated SOLPS runs where the sensitivity of the code solution to various assumptions of the neutral model and parallel heat transport of ions and electrons are described. A comparison between simulated and experimentally measured Mach numbers of the parallel ion flow in the SOL is presented, and conditions necessary for obtaining fast flows in the code are analysed. (author)

  17. Gyrokinetic studies of core turbulence features in ASDEX Upgrade: Can gyrokinetic simulations match the fluctuation measurements?

    Science.gov (United States)

    Banon Navarro, Alejandro

    2015-11-01

    Worldwide, gyrokinetic codes are used to predict the dominant micro-instabilities as well as the resulting anomalous transport in fusion experiments. A careful verification and validation of these codes is crucial to develop confidence in the model and improving the predictive capabilities of the numerical simulations. To date, the validation of gyrokinetic simulations versus experiments is mainly done at a macroscopic level, namely, by comparing turbulent heat fluxes. This is usually achieved by varying the profile gradients within the experimental error bars until a match with the experimental heat fluxes is obtained. However, since the turbulent fluxes are caused by plasma fluctuations on microscopic scales, it is also necessary to validate gyrokinetic codes on a microscopic level. We will describe a recent step in this direction by presenting simulation results with the gyrokinetic code GENE for an ASDEX Upgrade discharge. In particular, after flux-matched simulations are achieved, density fluctuations measured by means of Doppler reflectometry are compared with results of gyrokinetic simulations. We will also show that density and temperature fluctuation amplitudes and even the fluctuation spectra can be very sensitive to small changes in the profile gradients. This implies that a match of gyrokinetic simulations with experiment measurements for these quantities can be very difficult to achieve. However, it is observed that cross-phases between different quantities are robust to changes in this parameter, indicating that cross-phases could be a better observable for comparisons with experimental measurements.

  18. Overview of ASDEX Upgrade results—development of integrated operating scenarios for ITER

    Science.gov (United States)

    Günter, S.; Angioni, C.; Apostoliceanu, M.; Atanasiu, C.; Balden, M.; Becker, G.; Becker, W.; Behler, K.; Behringer, K.; Bergmann, A.; Bilato, R.; Bizyukov, I.; Bobkov, V.; Bolzonella, T.; Borba, D.; Borrass, K.; Brambilla, M.; Braun, F.; Buhler, A.; Carlson, A.; Chankin, A.; Chen, J.; Chen, Y.; Cirant, S.; Conway, G.; Coster, D.; Dannert, T.; Dimova, K.; Drube, R.; Dux, R.; Eich, T.; Engelhardt, K.; Fahrbach, H.-U.; Fantz, U.; Fattorini, L.; Foley, M.; Franzen, P.; Fuchs, J. C.; Gafert, J.; Gal, K.; Gantenbein, G.; García Muñoz, M.; Gehre, O.; Geier, A.; Giannone, L.; Gruber, O.; Haas, G.; Hartmann, D.; Heger, B.; Heinemann, B.; Herrmann, A.; Hobirk, J.; Hohenöcker, H.; Horton, L.; Huart, M.; Igochine, V.; Jacchia, A.; Jakobi, M.; Jenko, F.; Kallenbach, A.; Kálvin, S.; Kardaun, O.; Kaufmann, M.; Keller, A.; Kendl, A.; Kick, M.; Kim, J.-W.; Kirov, K.; Klose, S.; Kochergov, R.; Kocsis, G.; Kollotzek, H.; Konz, C.; Kraus, W.; Krieger, K.; Kurki-Suonio, T.; Kurzan, B.; Lackner, K.; Lang, P. T.; Lauber, P.; Laux, M.; Leuterer, F.; Likonen, J.; Lohs, A.; Lorenz, A.; Lorenzini, R.; Lyssoivan, A.; Maggi, C.; Maier, H.; Mank, K.; Manini, A.; Manso, M.-E.; Mantica, P.; Maraschek, M.; Martin, P.; Mast, K. F.; Mayer, M.; McCarthy, P.; Meyer, H.; Meisel, D.; Meister, H.; Menmuir, S.; Meo, F.; Merkel, P.; Merkel, R.; Merkl, D.; Mertens, V.; Monaco, F.; Mück, A.; Müller, H. W.; Münich, M.; Murmann, H.; Na, Y.-S.; Narayanan, R.; Neu, G.; Neu, R.; Neuhauser, J.; Nishijima, D.; Nishimura, Y.; Noterdaeme, J.-M.; Nunes, I.; Pacco-Düchs, M.; Pautasso, G.; Peeters, A. G.; Pereverzev, G.; Pinches, S.; Poli, E.; Posthumus-Wolfrum, E.; Pütterich, T.; Pugno, R.; Quigley, E.; Radivojevic, I.; Raupp, G.; Reich, M.; Riedl, R.; Ribeiro, T.; Rohde, V.; Roth, J.; Ryter, F.; Saarelma, S.; Sandmann, W.; Santos, J.; Schall, G.; Schilling, H.-B.; Schirmer, J.; Schneider, W.; Schramm, G.; Schweinzer, J.; Schweizer, S.; Scott, B.; Seidel, U.; Serra, F.; Sihler, C.; Silva, A.; Sips, A.; Speth, E.; Stäbler, A.; Steuer, K.-H.; Stober, J.; Streibl, B.; Strintzi, D.; Strumberger, E.; Suttrop, W.; Tardini, G.; Tichmann, C.; Treutterer, W.; Troppmann, M.; Tsalas, M.; Urano, H.; Varela, P.; Wagner, D.; Wesner, F.; Würsching, E.; Ye, M. Y.; Yoon, S.-W.; Yu, Q.; Zaniol, B.; Zasche, D.; Zehetbauer, T.; Zehrfeld, H.-P.; Zilker, M.; Zohm, H.

    2005-10-01

    Significant progress has been made on ASDEX Upgrade during the last two years in the basic understanding of transport, in the extension of the improved H-mode in parameter space and towards an integrated operating scenario and in the development of control methods for major performance limiting instabilities. The important features were the understanding of particle transport and the control of impurity accumulation based on it, the satisfactory operation with predominantly tungsten-clad walls, the improved H-mode operation over density ranges and for temperature ratios covering (non-simultaneously) the ITER requirements on ν*, n/nGW and Te/Ti, the ELM frequency control by pellet injection and the optimization of NTM suppression by DC-ECCD through variation of the launching angle. From these experiments an integrated scenario has emerged which extrapolates to a 50% improvement in n T τ or a 30% reduction of the required current when compared with the ITER base-line assumptions, with moderately peaked electron and controllable high-Z density profiles.

  19. Ten years of W programme in ASDEX Upgrade-challenges and conclusions

    International Nuclear Information System (INIS)

    Since 1999 ASDEX Upgrade increased its tungsten plasma-facing components (PFCs) and finally reached a full W coverage in 2007. Most of the initial goals of the investigations were successfully achieved. A highlight of the investigations was multiple start-ups and operation without any boronization demonstrating that performance and confinement similar to boronized operation with carbon PFCs can be reached in high power, high density discharges. This also allowed the investigation of the hydrogen retention without disturbing effects from the low-Z coating. A strong reduction of hydrogen retention was found in gas balance measurements as well as in post-mortem analyses. On the other hand, an almost complete suppression of low-Z divertor radiation was achieved after boronization, providing valuable information on the control requirements of radiative cooling by artificially introduced impurities. Among the challenges remains the strong increase of the W source and W concentration resulting from ICRH. At the same time it helped to identify the underlying physics and may lead to solutions superior to the presently used ones.

  20. Improved Collective Thomson Scattering measurements of fast ions at ASDEX Upgrade

    CERN Document Server

    Rasmussen, J; Stejner, M; Salewski, M; Jacobsen, A S; Korsholm, S B; Leipold, F; Meo, F; Michelsen, P K; Moseev, D; Schubert, M; Stober, J; Tardini, G; Wagner, D

    2013-01-01

    Understanding the behaviour of the confined fast ions is important in both current and future fusion experiments. These ions play a key role in heating the plasma and will be crucial for achieving conditions for burning plasma in next-step fusion devices. Microwave-based Collective Thomson Scattering (CTS) is well suited for reactor conditions and offers such an opportunity by providing measurements of the confined fast-ion distribution function resolved in space, time and 1D velocity space. We currently operate a CTS system at ASDEX Upgrade using a gyrotron which generates probing radiation at 105 GHz. A new setup using two independent receiver systems has enabled improved subtraction of the background signal, and hence the first accurate characterization of fast-ion properties. Here we review this new dual-receiver CTS setup and present results on fast-ion measurements based on the improved background characterization. These results have been obtained both with and without NBI heating, and with the measurem...

  1. Shear strength of the ASDEX Upgrade TF coil insulation: rupture, fatigue and creep behaviour

    International Nuclear Information System (INIS)

    The interlaminar shear strength of the insulation system for the 16 toroidal field (TF) coils of ASDEX Upgrade is investigated. The interlaminar shear properties of the glass-epoxy insulation are primarily determined by the resin system and its curing procedure. The pure resin was therefore tested first in tension. The results were taken into account for setting up the method of curing the TF coils. Shear tests of the complete glass-epoxy system were then conducted with tubular torque specimens providing a nearly homogeneous stress distribution. In particular, the influence of the amount of flexibilizer (5, 10, 15 parts of resin weight = PoW) on the rupture and fatigue strengths was assessed at a temperature T = 60 C, as also was the temperature dependence of the creep rate (40 C, 60 C, 80 C). The results obtained are not based on safe statistics. Nevertheless, they show clear trends. Finally, a visco-elastic model was set up to describe the creep behaviour of the insulation system. (author)

  2. Shear strength of the ASDEX upgrade TF coil insulation: Rupture, fatigue and creep behaviour

    International Nuclear Information System (INIS)

    This paper is concerned with the interlaminar shear strength of the insulation system for the 16 toroidal field (TF) coils of ASDEX Upgrade. The interlaminar shear properties of the glass-epoxy insulation are primarily determined by the resin system and its curing procedure. The pure resin was therefore tested first in tension. The results were taken into account for setting up the method of curing the TF coils. Shear tests of the complete glass-epoxy system were then conducted with tubular torque specimens providing a nearly homogeneous stress distribution. In particular, the influence of the amount of flexibilizer (5, 10, 15 parts of resin weight = PoW) on the rupture and fatigue strengths was assessed at a temperature T = 60C, as also was the temperature dependence of the creep rate (40 C, 60C, 80 C). The results obtained are not based on safe statistics. Nevertheless, they show clear trends. Finally, a visco-elastic model was set up to describe the creep behavior of the insulation system

  3. Profile evaluation techniques for O-mode broadband microwave reflectometry on ASDEX

    International Nuclear Information System (INIS)

    Density profiles from reflectometry can be obtained, in principle, with phase or time delay measurements. In the first case frequency-modulated continuous waves (FM-CW) are launched into the plasma, and in the second one different types of signals, namely pulses, are used. Whereas in the ionosphere density profiles are normally obtained with pulsed radar techniques, in fusion plasmas FM-CW reflectometry has been mostly used. In both techniques the localization of each reflecting layer cannot be deducted from single measurements as, for the same measured phase shift or time delay, the location depends on the density of the plasma that the waves have encountered in their propagating path. So, in order to determine the correct position of each layer all the layers with lower densities have to be probed. As microwaves are very sensitive to plasma modes and broadband turbulence the resulting phase or time delay perturbations may lead to the incorrect interpretation of the data, causing large errors in the evaluated profiles. Also, in some cases, it is not possible to probe the complete plasma and deviations may occur due to the missing information. The evaluation of the profiles must, therefore, include data analysis procedures that take into account both the effect of plasma fluctuations and the limitations of the diagnostic. Here we present the techniques developed to analyse the ASDEX data, and discuss their potentialities for the routine evaluation of the density profiles from broadband reflectometry. (author). 3 refs, 12 figs

  4. Improved time-frequency analysis of ASDEX Upgrade reflectometry data using the reassigned spectrogram technique

    International Nuclear Information System (INIS)

    The spectrogram is one of the best-known time-frequency distributions suitable to analyze signals whose energy varies both in time and frequency. In reflectometry, it has been used to obtain the frequency content of FM-CW signals for density profile inversion and also to study plasma density fluctuations from swept and fixed frequency data. Being implemented via the short-time Fourier transform, the spectrogram is limited in resolution, and for that reason several methods have been developed to overcome this problem. Among those, we focus on the reassigned spectrogram technique that is both easily automated and computationally efficient requiring only the calculation of two additional spectrograms. In each time-frequency window, the technique reallocates the spectrogram coordinates to the region that most contributes to the signal energy. The application to ASDEX Upgrade reflectometry data results in better energy concentration and improved localization of the spectral content of the reflected signals. When combined with the automatic (data driven) window length spectrogram, this technique provides improved profile accuracy, in particular, in regions where frequency content varies most rapidly such as the edge pedestal shoulder.

  5. Real-time diagnostic integration with the ASDEX upgrade control system

    International Nuclear Information System (INIS)

    Operating in advanced plasma scenarios has become one of the major goals on the way towards ITER and a future fusion reactor. This implies the reinforced control of physics quantities like pressure profiles or magnetic island structures, which have to be reconstructed in real-time for this purpose. To achieve this goal the collaboration scheme between diagnostic data processing and real-time control has to be fundamentally changed from strict separation to open bidirectional information exchange in real-time. ASDEX Upgrade is currently designing and building a distributed computer cluster to implement such an integrated diagnostic and control system. The main topics comprise modular framework design, low-latency data communication via networks, and built-in synchronisation methods. We show how the discharge control system's modular framework is extended to publish calculated quantities via the network transparent to algorithm designers. A communication layer allows the exchange of real-time information between computation nodes with automatised routing even over several networks. Finally, in order to consistently combine data from independent sources, synchronisation methods are developed. Thus, sophisticated feedback control with time scales of milliseconds become feasible.

  6. Magnetic confinement experiment. I: Tokamaks

    International Nuclear Information System (INIS)

    Reports were presented at this conference of important advances in all the key areas of experimental tokamak physics: Core Plasma Physics, Divertor and Edge Physics, Heating and Current Drive, and Tokamak Concept Optimization. In the area of Core Plasma Physics, the biggest news was certainly the production of 9.2 MW of fusion power in the Tokamak Fusion Test Reactor, and the observation of unexpectedly favorable performance in DT plasmas. There were also very important advances in the performance of ELM-free H- (and VH-) mode plasmas and in quasi-steady-state ELM'y operation in JT-60U, JET, and DIII-D. In all three devices ELM-free H-modes achieved nTτ's ∼ 2.5x greater than ELM'ing H-modes, but had not been sustained in quasi-steady-state. Important progress has been made on the understanding of the physical mechanism of the H-mode in DIII-D, and on the operating range in density for the H-mode in Compass and other devices

  7. Characterization and scaling of the tokamak edge transport barrier

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, Philip Adrian

    2012-04-24

    The high confinement regime (H-mode) in a tokamak plasma displays a remarkable edge region. On a small spatial scale of 1-2 cm the properties of the plasma change significantly. Certain parameters vary 1-2 orders of magnitude in this region, called the pedestal. Currently, there is no complete understanding of how the pedestal forms or how it is sustained. The goal of this thesis is to contribute to the theoretical understanding of the pedestal and provide scalings towards larger machines, like ITER and DEMO. A pedestal database was built with data from different tokamaks: ASDEX Upgrade, DIIID and JET. The pedestal was characterized with the same method for all three machines. This gives the maximum value, gradient and width of the pedestal in n{sub e}, T{sub e} and T{sub i}. These quantities were analysed along with quantities derived from them, such as the pressure or the confinement time. For this purpose two parameter sets were used: normalized parameters (pressure {beta}, time {nu}{sub *}, length {rho}{sub *}, shape f{sub q}) and machine parameters (size a, magnetic field B{sub t}, plasma current I{sub p}, heating P). All results are dependent on the choice of the coordinate system: normalized poloidal flux {Psi}{sub N} and real space r/a. The most significant result, which was obtained with both parameter sets, shows a different scaling of the pedestal width for the electron temperature and the electron density. The presented scalings predict that in ITER and DEMO the temperature pedestal will be appreciably wider than the density pedestal. The pedestal top scaling for the pressure reveals differences between the electron and the ion pressure. In extrapolations this results in values for T{sub e,ped} of 4 keV (ITER) and 10 keV (DEMO), but significantly lower values for the ion temperature. A two-term method was applied to use the pedestal pressure to determine the pedestal contribution to the global confinement time {tau}{sub E}. The dependencies in the

  8. Metrology measurements for Aditya tokamak upgradation

    International Nuclear Information System (INIS)

    After 25 years of Aditya tokamak (midsized, air-core, R0= 75 cm, a = 25 cm) operation achieving high temperature circular plasmas in limiter configuration, upgrading it to Aditya-U tokamak with divertor configuration has been planned and the upgradation is under progress. The upgradation process include dismantling of the existing Aditya tokamak to its base level and re-erect it by placing new subsystems like new vacuum vessel of circular cross-section, new buckling cylinder etc. Apposite alignment of subsystems, mainly all the magnetic coil systems in all grades and scales of tokamak is very crucial and essential, as misaligned magnetic coil system scan generate error magnetic fields, which can significantly impact the plasma formation and sustainment in a tokamak. With this motivation, position and alignment measurement of the existing magnetic coils and structural components of ADITYA tokamak is carried out for the very first time with the optical metrology instrument. Prior to carrying out measurement exercise, machine datum has been transferred to the reference on the wall of tokamak hall using five-point laser and the machine center has been transformed to the four wall of tokamak hall. All position measurements are done with respect to machine major axis in cylindrical geometry. Measurement includes existing radial (R) and elevation (Z) positions of all magnetic coils and various structural components within the accuracy of ± 1 mm. More than 5000 data points are recorded using optical metrology instrument. Again the elevation references are transferred to the primary network established and the angular references are transformed on the floor of the tokamak hall. These results will serve as ready reference for reassembly and alignment of Aditya - Upgrade tokamak. In this paper detailed position measurements of different subsystems of old Aditya tokamak and the relocation of them along with new ones using the optical metrology instruments will be presented

  9. Assessment of compatibility of ICRF antenna operation with full W wall in ASDEX Upgrade

    Czech Academy of Sciences Publication Activity Database

    Bobkov, Vl.V.; Braun, F.; Dux, R.; Herrmann, A.; Giannone, L.; Kallenbach, A.; Křivská, Alena; Müller, H.W.; Neu, R.; Noterdaeme, J.-M.; Pütterich, T.; Rohde, V.; Schweinzer, J.; Sips, A.; Zammuto, I.

    2010-01-01

    Roč. 50, č. 3 (2010), 035004-035004. ISSN 0029-5515. [Workshop on Stochasticity in Fusion Plasmas/4./. Julich, 02.03.2009-04.03.2009] Institutional support: RVO:61389021 Keywords : tokamak * ICRF Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.303, year: 2010 http://iopscience.iop.org/0029-5515/50/3/035004/pdf/0029-5515_50_3_035004.pdf

  10. Fast ion dynamics in ASDEX upgrade and TEXTOR measured by collective Thomson scattering

    Energy Technology Data Exchange (ETDEWEB)

    Moseev, D.

    2011-11-15

    Fast ions are an essential ingredient in burning nuclear fusion plasmas: they are responsible for heating the bulk plasma, carry a significant amount of plasma current and moreover interact with various magnetohydrodynamic (MHD) instabilities. The collective Thomson scattering (CTS) diagnostic is sensitive to the projection of fast ion velocity distribution function. This thesis is mainly devoted to investigations of fast ion physics in tokamak plasmas by means of CTS. (Author)

  11. Fast ion dynamics in ASDEX upgrade and TEXTOR measured by collective Thomson scattering

    International Nuclear Information System (INIS)

    Fast ions are an essential ingredient in burning nuclear fusion plasmas: they are responsible for heating the bulk plasma, carry a significant amount of plasma current and moreover interact with various magnetohydrodynamic (MHD) instabilities. The collective Thomson scattering (CTS) diagnostic is sensitive to the projection of fast ion velocity distribution function. This thesis is mainly devoted to investigations of fast ion physics in tokamak plasmas by means of CTS. (Author)

  12. Theory of high-beta tokamaks

    International Nuclear Information System (INIS)

    The theoretical researches on high beta tokamak are reviewed. The ballooning mode instability is thought to be the most serious problem for the high beta tokamaks, and the theoretical results on the ballooning mode instability are discussed in detail. The experimental results in high beta belt pinch devices are also discussed. (author)

  13. Tokamak plasma position dynamics and feedback control

    International Nuclear Information System (INIS)

    The perturbation equations of a tokamak plasma equilibrium position are developed. Solution of the approximated perturbation equations is carried out. A unique, simple, and useful plasma displacement dynamics transfer function of a tokamak is developed. The dominant time constants of the dynamics transfer function are determined in a symbolic form

  14. Economic evaluation of tokamak power plants

    International Nuclear Information System (INIS)

    This study reports the impact of plasma operating characteristics, engineering options, and technology on the capital cost trends of tokamak power plants. Tokamak power systems are compared to other advanced energy systems and found to be economically competitive. A three-phase strategy for demonstrating commercial feasibility of fusion power, based on a common-site multiple-unit concept, is presented

  15. The disruptive instability in Tokamak plasmas

    NARCIS (Netherlands)

    Salzedas, F.J.B.

    2001-01-01

    Studies performed in RTP (Rijnhuizen Tokamak Project) of the most violent and dangerous instability in tokamak plasmas, the major disruption, are presented. A particular class of disruptions is analyzed, namely the density limit disruption, which occur in high density plasmas. The radiative te

  16. Physics of compact ignition tokamak designs

    International Nuclear Information System (INIS)

    Models for predicting plasma performance in compact ignition experiments are constructed on the basis of theoretical and empirical constraints and data from tokamak experiments. Emphasis is placed on finding transport and confinement models which reproduce results of both ohmically and auxiliary heated tokamak data. Illustrations of the application of the models to compact ignition designs are given

  17. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    This report contains an overview of the Aries-I tokamak reactor study. The following topics are discussed on this tokamak: Systems studies; equilibrium, stability, and transport; summary and conclusions; current drive; impurity control system; tritium systems; magnet engineering; fusion-power-core engineering; power conversion; Aries-I safety design and analysis; design layout and maintenance; and start-up and operations

  18. Engineering Design of KSTAR tokamak main structure

    International Nuclear Information System (INIS)

    The main components of the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak including vacuum vessel, plasma facing components, cryostat, thermal shield and magnet supporting structure are in the final stage of engineering design. Hundai Heavy Industries (HHI) has been involved in the engineering design of these components. The current configuration and the final engineering design results for the KSTAR main structure are presented. (author)

  19. Summary report on tokamak confinement experiments

    International Nuclear Information System (INIS)

    There are currently five major US tokamaks being operated and one being constructed under the auspices of the Division of Toroidal Confinement Systems. The currently operating tokamaks include: Alcator C at the Massachusetts Institute of Technology, Doublet III at the General Atomic Company, the Impurity Studies Experiment (ISX-B) at the Oak Ridge National Laboratory, and the Princeton Large Torus (PLT) and the Poloidal Divertor Experiment (PDX) at the Princeton Plasma Physics Laboratory. The Tokamak Fusion Test Reactor (TFTR) is under construction at Princeton and should be completed by December 1982. There is one major tokamak being funded by the Division of Applied Plasma Physics. The Texas Experimental Tokamak (TEXT) is being operated as a user facility by the University of Texas. The TEXT facility includes a complete set of standard diagnostics and a data acquisition system available to all users

  20. Natural current profiles in a tokamak

    International Nuclear Information System (INIS)

    In this paper I show how one may arrive at a universal, or natural, family of Tokamak profiles using only accepted physical principles. These particular profiles are similar to ones proposed previously on the basis of ad hoc variational principles and the point of the present paper is to provide a justification for them. However in addition, the present work provides an interesting view of Tokamak fluctuations and leads to a new result -- a relationship between the inward particle pinch velocity, the diffusion coefficient and the current profile. The basic Tokamak model is described in this paper. Then an analogy is developed between Tokamak profiles and the equilibrium of a realisable dynamical system. Then the equations governing the natural Tokamak profiles are derived by applying standard statistical mechanics to this analog. The profiles themselves are calculated and some other results of the theory are described

  1. Measurements of the fast-ion distribution function at ASDEX upgrade by collective Thomson scattering (CTS) using active and passive views

    DEFF Research Database (Denmark)

    Nielsen, Stefan Kragh; Stejner Pedersen, Morten; Rasmussen, Jesper;

    2015-01-01

    Collective Thomson scattering (CTS) can provide measurements of the confined fast-ion distribution function resolved in space, time and 1D velocity space. On ASDEX Upgrade, the measured spectra include an additional signal which previously has hampered data interpretation. A new set-up using two...

  2. Radial transport in the far scrape-off layer of ASDEX upgrade during L-mode and ELMy H-mode

    DEFF Research Database (Denmark)

    Ionita, C.; Naulin, Volker; Mehlmann, F.;

    2013-01-01

    The radial turbulent particle flux and the Reynolds stress in the scrape-off layer (SOL) of ASDEX Upgrade were investigated for two limited L-mode (low confinement) and one ELMy H-mode (high confinement) discharge. A fast reciprocating probe was used with a probe head containing five Langmuir pro...

  3. Influence of gas injection location and magnetic perturbations on ICRF antenna performance in ASDEX Upgrade

    International Nuclear Information System (INIS)

    In ASDEX Upgrade H-modes with H98≈0.95, similar effect of the ICRF antenna loading improvement by local gas injection was observed as previously in L-modes. The antenna loading resistance Ra between and during ELMs can increase by more than 25% after a switch-over from a deuterium rate of 7.5⋅1021 D/s injected from a toroidally remote location to the same amount of deuterium injected close to an antenna. However, in contrast to L-mode, this effect is small in H-mode when the valve downstream w.r.t. parallel plasma flows is used. In L-mode, a non-linearity of Ra at PICRPa>30% with no effect of spectrum and phase of MPs on Ra found so far. In the case ELMs are fully mitigated, the antenna loading is higher and steadier. In the case ELMs are not fully mitigated, the value of Ra between ELMs is increased. Looking at the W source modification for the improved loading, the local gas injection is accompanied by decreased values of tungsten (W) influx ΓW from the limiters and its effective sputtering yield Yw, with the exception of the locations directly at the antenna gas valve. Application of MPs leads to increase of ΓW and Yw for some of the MP phases. With nitrogen seeding in the divertor, ICRF is routinely used to avoid impurity accumulation and that despite enhanced ΓW and YW at the antenna limiters

  4. Real-time diagnostics at ASDEX Upgrade-Architecture and operation

    International Nuclear Information System (INIS)

    Diagnostics at ASDEX Upgrade have available a very large number of highly developed measuring channels. The prospect of making this wealth of information usable for plasma optimisation led to the implementation of a number of diagnostics running data acquisition in real-time (RT). Ultimately, this development aims to achieve a network of intelligent diagnostics delivering analysed data for high-level plasma performance control such as profile shaping and NTM stabilisation. The new RT diagnostics consist of standard industrial 19 in. servers organised in clusters and running a standard UNIX multiprocessor RT-capable operating system (RT OS). Built-to-purpose computer interface cards deliver data (e.g. via serial links) from the data acquisition (DAQ) front-ends directly into the main memory of the DAQ servers. An RT data analysis task immediately following the running direct memory access (DMA) data transfers processes the data and delivers the results to follow-up systems in the control chain. Whereas the first systems were implemented in a simple just a bunch of computers (JBOC) configuration being operated as a number of single diagnostics, newer systems are integrated into diagnostic clusters using parallel computing techniques such as message passing interface (MPI). The paper describes the hardware (ADC front-ends, serial I/O, selection criteria and performance of the involved computer busses and systems) and software (DAQ, DA, RT OS, MPI) architecture of the assembled systems. Benchmark results for DAQ and MPI bandwidth and latencies as well as for the behaviour of the RT OS will be given

  5. Monitoring millimeter wave stray radiation during ECRH operation at ASDEX Upgrade

    Directory of Open Access Journals (Sweden)

    Wagner D.

    2012-09-01

    Full Text Available Due to imperfection of the single path absorption, ECRH at ASDEX Upgrade (AUG is always accompanied by stray radiation in the vacuum vessel. New ECRH scenarios with O2 and X3 heating schemes extend the operational space, but they have also the potential to increase the level of stray radiation. There are hazards for invessel components. Damage on electric cables has already been encountered. It is therefore necessary to monitor and control the ECRH with respect to the stray radiation level. At AUG a system of Sniffer antennas equipped with microwave detection diodes is installed. The system is part of the ECRH interlock circuit. We notice, however, that during plasma operation the variations of the Sniffer antenna signal are very large. In laboratory measurements we see variations of up to 20 dB in the directional sensitivity and we conclude that an interference pattern is formed inside the copper sphere of the antenna. When ECRH is in plasma operation at AUG, the plasma is acting as a phase and mode mixer for the millimeter waves and thus the interference pattern inside the sphere changes with the characteristic time of the plasma dynamics. In order to overcome the difficulty of a calibrated measurement of the average stray radiation level, we installed bolometer and pyroelectric detectors, which intrinsically average over interference structures due to their large active area. The bolometer provides a robust calibration but with moderate temporal resolution. The pyroelectric detector provides high sensitivity and a good temporal resolution, but it raises issues of possible signal drifts in long pulses.

  6. ECRH on ASDEX Upgrade - System Status, Feed-Back Control, Plasma Physics Results -

    Directory of Open Access Journals (Sweden)

    Flamm J.

    2012-09-01

    Full Text Available The ASDEX Upgrade (AUG ECRH system now delivers a total of 3.9 MW to the plasma at 140 GHz. Three new units are capable of 2-frequency operation and may heat the plasma alternatively with 2.1 MW at 105 GHz. The system is routinely used with X2, O2, and X3 schemes. For Bt = 3.2 T also an ITER-like O1-scheme can be run using 105 GHz. The new launchers are capable of fast poloidal movements necessary for real-time control of the location of power deposition. Here real-time control of NTMs is summarized, which requires a fast analysis of massive data streams (ECE and Mirnov correlation and extensive calculations (equilibria, ray-tracing. These were implemented at AUG using a modular concept of standardized real-time diagnostics. The new realtime capabilities have also been used during O2 heating to keep the first reflection of the non-absorbed beam fraction on the holographic reflector tile which ensures a well defined second pass of the beam through the central plasma. Sensors for the beam position are fast thermocouples at the edge of the reflector tile. The enhanced ECRH power was used for several physics studies related to the unique feature of pure electron heating without fueling and without momentum input. As an example the effect of the variation of the heating mix in moderately heated H-modes is demonstrated using the three available heating systems, i.e. ECRH, ICRH and NBI. Keeping the total input power constant, strong effects are seen on the rotation, but none on the pedestal parameters. Also global quantities as the stored energy are hardly modified. Still it is found that the central ion temperature drops as the ECRH fraction exceeds a certain threshold.

  7. Fast particle-driven ion cyclotron emission (ICE) in tokamak plasmas and the case for an ICE diagnostic in ITER

    CERN Document Server

    McClements, K G; Dendy, R O; Carbajal, L; Chapman, S C; Cook, J W S; Harvey, R W; Heidbrink, W W; Pinches, S D

    2014-01-01

    Fast particle-driven waves in the ion cyclotron frequency range (ion cyclotron emission or ICE) have provided a valuable diagnostic of confined and escaping fast ions in many tokamaks. This is a passive, non-invasive diagnostic that would be compatible with the high radiation environment of deuterium-tritium plasmas in ITER, and could provide important information on fusion {\\alpha}-particles and beam ions in that device. In JET, ICE from confined fusion products scaled linearly with fusion reaction rate over six orders of magnitude and provided evidence that {\\alpha}-particle confinement was close to classical. In TFTR, ICE was observed from super-Alfv\\'enic {\\alpha}-particles in the plasma edge. The intensity of beam-driven ICE in DIII-D is more strongly correlated with drops in neutron rate during fishbone excitation than signals from more direct beam ion loss diagnostics. In ASDEX Upgrade ICE is produced by both super-Alfv\\'enic DD fusion products and sub-Alfv\\'enic deuterium beam ions.

  8. Potential turbulence in tokamak plasmas

    International Nuclear Information System (INIS)

    Microscopic potential turbulence in tokamak plasmas are investigated by a multi-sample-volume heavy ion beam probe. The wavenumber/frequency spectra S(k,ω) of the plasmas potential fluctuation as well as density fluctuation are obtained for the first time. The instantaneous turbulence-driven particle flux, calculated from potential and density turbulence has oscillations of which amplitude is about 100 times larger than the steady-state outwards flux, showing sporadic behaviours. We also observed large-scale coherent potential oscillations with the frequency around 10-40 kHz. (author)

  9. The bootstrap current in tokamaks

    International Nuclear Information System (INIS)

    The properties of the Hirshman equation for the bootstrap in the tokamak and the difference between it and the simpler Hinton-Hazeltine equation are discussed. The Hirshman model, which takes into account finite-aspect-ratio effects, is used to calculate the bootstrap current in the plasma in a circular cross section with Te = Ti. Approximate upper and lower bounds on the bootstrap current are obtained. These restrict the range of variation of the current as the temperature and density profiles vary. 16 refs., 9 figs

  10. Breakdown in the pretext tokamak

    International Nuclear Information System (INIS)

    Data are presented on the application of ion cyclotron resonance RF power to preionization in tokamaks. We applied 0.3-3 kW at 12 MHz to hydrogen and obtained a visible discharge, but found no scaling of breakdown voltage with any parameter we were able to vary. A possible explanation for this, which implies that higher RF power would have been much more effective, is discussed. Finally, we present our investigation of the dV/dt dependence of breakdown voltage in PRETEXT, a phenomenon also seen in JFT-2. The breakdown is discussed in terms of the physics of Townsend discharges

  11. Cluster storage for COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Písačka, Jan; Hron, Martin; Janky, Filip; Pánek, Radomír

    2012-01-01

    Roč. 87, č. 12 (2012), s. 2238-2241. ISSN 0920-3796. [IAEA Technical Meeting on Control, Data Acquisition, and Remote Participation for Fusion Research/8./. San Francisco, 20.06.2011-24.06.2011] R&D Projects: GA ČR GAP205/11/2470; GA MŠk 7G10072; GA MŠk(CZ) LM2011021 Institutional research plan: CEZ:AV0Z20430508 Keywords : COMPASS * Tokamak * Codac * Cluster * GlusterFS * Storage Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.842, year: 2012 http://dx.doi.org/10.1016/j.fusengdes.2012.09.006

  12. Anomalous particle pinch in Tokamaks

    International Nuclear Information System (INIS)

    The diffusion coefficient in phase space usually varies with the particle energy. A consequence is the dependence of the fluid particle flux on the temperature gradient. If the diffusion coefficient in phase space decreases with the energy in the bulk of the thermal distribution function, the particle thermodiffusion coefficient which links the particle flux to the temperature gradient is negative. This is a possible explanation for the inward particle pinch that is observed in tokamaks. A quasilinear theory shows that such a thermodiffusion is generic for a tokamak electrostatic turbulence at low frequency. This effect adds to the particle flux associated with the radial gradient of magnetic field. This behavior is illustrated with a perturbed electric potential, for which the trajectories of charged particle guiding centers are calculated. The diffusion coefficient of particles is computed and compared to the quasilinear theory, which predicts a divergence at low velocity. It is shown that at low velocity, the actual diffusion coefficient increases, but remains lower than the quasilinear value. Nevertheless, this differential diffusion between cold and fast particles leads to an inward flux of particles. (author)

  13. Enhancement of confinement in tokamaks

    International Nuclear Information System (INIS)

    A plausible interpretation of the experimental evidence is that energy confinement in tokamaks is governed by two separate considerations: (1) the need for resistive MHD kink-stability, which limits the permissible range of current profiles - and therefore normally also the range of temperature profiles; and (2) the presence of strongly anomalous microscopic energy transport near the plasma edge, which calibrates the amplitude of the global temperature profile, thus determining the energy confinement time tau/sub E/. Correspondingly, there are two main paths towards the enhancement of tokamak confinement: (1) Configurational optimization, to increase the MHD-stable energy content of the plasma core, can evidently be pursued by varying the cross-sectional shape of the plasma and/or finding stable radial profiles with central q-values substantially below unity - but crossing from ''first'' to ''second'' stability within the peak-pressure region would have the greatest ultimate potential. (2) Suppression of edge turbulence, so as to improve the heat insulation in the outer plasma shell, can be pursued by various local stabilizing techniques, such as use of a poloidal divertor. The present confinement model and initial TFTR pellet-injection results suggest that the introduction of a super-high-density region within the plasma core should be particularly valuable for enhancing ntau/subE/. In D-T operation, a centrally peaked plasma pressure profile could possibly lend itself to alpha-particle-driven entry into the second-stability regime

  14. Cluster storage for COMPASS tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Pisacka, J., E-mail: pisacka@ipp.cas.cz [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Hron, M., E-mail: hron@ipp.cas.cz [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Janky, F., E-mail: jankyf@ipp.cas.cz [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Department of Surface and Plasma Science, Faculty of Mathematics and Physics, Charles University, V Holesovickach 2, 180 00 Praha 8 (Czech Republic); Panek, R., E-mail: panek@ipp.cas.cz [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer New data storage system needed for the COMPASS tokamak. Black-Right-Pointing-Pointer Distributed, fault-tolerant, parallel, scalable, non-proprietary. Black-Right-Pointing-Pointer GlusterFS selected for testing on a small test bed. Black-Right-Pointing-Pointer Aggregated reading throughput reached 300 MiB/s for 6 clients - very good result. - Abstract: The COMPASS tokamak is expected to produce several gigabytes of data per shot in near future. A new storage system is needed to accommodate and access all the data. It should be scalable, fault-tolerant, and parallel. It should not be based on proprietary solutions to maintain independence from hardware and software manufacturers and preferably it should be built on inexpensive commodity hardware. One of the promising distributed parallel fault-tolerant file systems, GlusterFS, was selected for testing. The aim of the work was to make initial tests of a particular small GlusterFS setup to confirm its aptitude for the COMPASS storage system. Aggregated reading throughput from multiple NFS clients was one of the most important figures that were benchmarked, it scaled well with the number of clients, starting just above 60 MiB/s for 1 client and going slightly over 300 MiB/s for 6 clients.

  15. Cluster storage for COMPASS tokamak

    International Nuclear Information System (INIS)

    Highlights: ► New data storage system needed for the COMPASS tokamak. ► Distributed, fault-tolerant, parallel, scalable, non-proprietary. ► GlusterFS selected for testing on a small test bed. ► Aggregated reading throughput reached 300 MiB/s for 6 clients – very good result. - Abstract: The COMPASS tokamak is expected to produce several gigabytes of data per shot in near future. A new storage system is needed to accommodate and access all the data. It should be scalable, fault-tolerant, and parallel. It should not be based on proprietary solutions to maintain independence from hardware and software manufacturers and preferably it should be built on inexpensive commodity hardware. One of the promising distributed parallel fault-tolerant file systems, GlusterFS, was selected for testing. The aim of the work was to make initial tests of a particular small GlusterFS setup to confirm its aptitude for the COMPASS storage system. Aggregated reading throughput from multiple NFS clients was one of the most important figures that were benchmarked, it scaled well with the number of clients, starting just above 60 MiB/s for 1 client and going slightly over 300 MiB/s for 6 clients.

  16. Predictive Modeling of Tokamak Configurations*

    Science.gov (United States)

    Casper, T. A.; Lodestro, L. L.; Pearlstein, L. D.; Bulmer, R. H.; Jong, R. A.; Kaiser, T. B.; Moller, J. M.

    2001-10-01

    The Corsica code provides comprehensive toroidal plasma simulation and design capabilities with current applications [1] to tokamak, reversed field pinch (RFP) and spheromak configurations. It calculates fixed and free boundary equilibria coupled to Ohm's law, sources, transport models and MHD stability modules. We are exploring operations scenarios for both the DIII-D and KSTAR tokamaks. We will present simulations of the effects of electron cyclotron heating (ECH) and current drive (ECCD) relevant to the Quiescent Double Barrier (QDB) regime on DIII-D exploring long pulse operation issues. KSTAR simulations using ECH/ECCD in negative central shear configurations explore evolution to steady state while shape evolution studies during current ramp up using a hyper-resistivity model investigate startup scenarios and limitations. Studies of high bootstrap fraction operation stimulated by recent ECH/ECCD experiments on DIIID will also be presented. [1] Pearlstein, L.D., et al, Predictive Modeling of Axisymmetric Toroidal Configurations, 28th EPS Conference on Controlled Fusion and Plasma Physics, Madeira, Portugal, June 18-22, 2001. * Work performed under the auspices of the U.S. Department of Energy by the University of California, Lawrence Livermore National Laboratory under contract No. W-7405-Eng-48.

  17. Tokamak Physics Experiment divertor design

    International Nuclear Information System (INIS)

    The Tokamak Physics Experiment (TPX) tokamak requires a symmetric up/down double-null divertor capable of operation with steady-state heat flux as high as 7.5 MW/m2. The divertor is designed to operate in the radiative mode and employs a deep slot configuration with gas puffing lines to enhance radiative divertor operation. Pumping is provided by cryopumps that pump through eight vertical ports in the floor and ceiling of the vessel. The plasma facing surface is made of carbon-carbon composite blocks (macroblocks) bonded to multiple parallel copper tubes oriented vertically. Water flowing at 6 m/s is used, with the critical heat flux (CHF) margin improved by the use of enhanced heat transfer surfaces. In order to extend the operating period where hands on maintenance is allowed and to also reduce dismantling and disposal costs, the TPX design emphasizes the use of low activation materials. The primary materials used in the divertor are titanium, copper, and carbon-carbon composite. The low activation material selection and the planned physics operation will allow personnel access into the vacuum vessel for the first 2 years of operation. The remote handling system requires that all plasma facing components (PFCs) are configured as modular components of restricted dimensions with special provisions for lifting, alignment, mounting, attachment, and connection of cooling lines, and instrumentation and diagnostics services

  18. Atomic physics in tokamak plasmas

    International Nuclear Information System (INIS)

    Tokamak discharges produce hydrogen-isotope plasmas in a quasi-steady state, with radial electron temperature, Tsub(e)(r), and density nsub(e)(r), distribution usually centrally peaked, with typical values Tsub(e)(0) approx.= 1 - 3 keV, nsub(e)(r) approx.= 1014 cm-3. Besides hydrogen, the plasma contains small quantities of carbon, oxygen, various construction or wall-conditioning materials such as Fe, Cr, Ni, Ti, Zr, Mo, and perhaps elements added for special diagnostic purposes, e.g., Si, Sc, Al, or noble gases. These elements are spatially fairly homogeneously distributed, with the different ionization states occurring near radial locations where Tsub(e)(r) approx.= Esub(i), the ionization potential. Thus, spectroscopic measurements of various plasma properties, such as ion temperatures, plasma motions or oscillations, radial transport rates, etc. are automatically endowed with spatial resolution. Furthermore the emitted spectra, even of heavier elements such as Fe or Ni, are fairly simple because only the ground levels are appreciably populated under the prevailing plasma conditions. Identification of near-ground transitions, including particularly magnetic dipole and intercombination transitions of ions with ionization potentials in the several keV range, and determination of their collisional and radiative transition probabilities will be required for development of appropriate diagnostics of tokamak-type plasma approaching the prospective fusion reactor conditions. (orig.)

  19. Soft X-Ray measurements and analysis on Tokamaks in view of real-time control

    International Nuclear Information System (INIS)

    This thesis focuses on measuring and interpreting the Soft X-Ray (SXR) radiation (approximately [1 keV; 15 keV]) in Tokamaks. As explained in Chapter 2, this radiation conveys information about the plasma density, temperature, magnetic equilibrium and impurity content. However, the measured data is spectrally and spatially-integrated and results from several physical phenomena affecting every ion species. Tore Supra's SXR diagnostics is based on semiconductor diodes presented in Chapter 3, along with a new gas detector successfully tested in laboratory and on Tore Supra. A new methodology for absolute spectral characterisation of photo detectors using a portable SXR tube is presented. Tomographic inversion algorithms, that grant access to reconstructions of the SXR emissivity field in a poloidal cross-section, are presented in Chapter 4. Improvements implemented on one particular algorithm are detailed with examples of application. A comparison between the position of the SXR emissivity maximum and the magnetic axis reconstructed by an equilibrium code is presented in Chapter 5. Chapter 6 presents an approach used to derive an impurity density from its SXR emissivity using the robustness of its SXR cooling factor with respect to impurity transport. The physics accounting for this robustness is studied and a first map of the domain of validity of this method is provided. Chapter 7 addresses poloidal asymmetries of the SXR emissivity field. Two types of asymmetries are presented as well as experiments conducted on ASDEX-U to verify their parametric dependences. A new type of SXR asymmetry, observed on Tore Supra is introduced. (author)

  20. Control of a burning tokamak plasma

    Energy Technology Data Exchange (ETDEWEB)

    Burmeister, R.E.; Mandrekas, J.; Stacey, W.M.

    1993-03-01

    This report is a review of the literature relevant to the control of the thermonuclear burn in a tokamak plasma. Some basic tokamak phenomena are reviewed, and then control by modulation of auxiliary heating and fueling is discussed. Other possible control methods such as magnetic ripple, plasma compression, and impurity injection as well as more recent proposed methods such as divertor biasing and L- to H-mode transition are also reviewed. The applications of modern control theory to the tokamak burn control problem are presented. The control results are summarized and areas of further research are identified.

  1. Fast IR diodes thermometer for tokamak

    International Nuclear Information System (INIS)

    A 30 channel fast IR pyrometry array has been constructed for tokamak, which has 0.5 μs time response, 10 mm diameter spatial resolution and 5 degree C temperature resolution. The temperature measuring range is from 250 degree C to 1200 degree C. The two dimensional temperature profiles of the first wall during both major and minor disruptions can be measured with an accuracy of about 1% measuring temperature, which is adequate for tokamak experiments. This gives a very useful tool for the disruption study, especially for the divertor physics and edge heat flux research on tokamak and other magnetic confinement devices

  2. Ball-Pen Probe Measurements in L-Mode and H-Mode on ASDEX Upgrade

    Czech Academy of Sciences Publication Activity Database

    Adámek, Jiří; Horáček, Jan; Müller, H. W.; Rohde, V.; Ionita, C.; Schrittwieser, R.; Mehlmann, F.; Kurzan, B.; Stöckel, Jan; Dejarnac, Renaud; Weinzettl, Vladimír; Seidl, Jakub; Peterka, M.

    2010-01-01

    Roč. 50, č. 9 (2010), s. 854-859. ISSN 0863-1042. [International Workshop on Electric Probes in Magnetized Plasmas/8th./. Innsbruck, 21.09.2009-24.09.2009] R&D Projects: GA AV ČR KJB100430901; GA ČR GA202/09/1467 Institutional research plan: CEZ:AV0Z20430508 Keywords : Tokamak * ball-pen probe * electron temperature * L-mode * H-mode * ELMs Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.006, year: 2010 http://onlinelibrary.wiley.com/doi/10.1002/ctpp.201010145/pdf

  3. Development of an Edge Transport Barrier at the H-Mode Transition of ASDEX

    Science.gov (United States)

    Wagner, F.; Fussmann, G.; Grave, T.; Keilhacker, M.; Kornherr, M.; Lackner, K.; McCormick, K.; Müller, E. R.; Stäbler, A.; Becker, G.; Bernhardi, K.; Ditte, U.; Eberhagen, A.; Gehre, O.; Gernhardt, J.; Gierke, G. V.; Glock, E.; Gruber, O.; Haas, G.; Hesse, M.; Janeschitz, G.; Karger, F.; Kissel, S.; Klüber, O.; Lisitano, G.; Mayer, H. M.; Meisel, D.; Mertens, V.; Murmann, H.; Poschenrieder, W.; Rapp, H.; Röhr, H.; Ryter, F.; Schneider, F.; Siller, G.; Smeulders, P.; Söldner, F.; Speth, E.; Steuer, K.-H.; Szymanski, Z.; Vollmer, O.

    1984-10-01

    The thermal wave of a minor disruption can initiate the H phase of a neutral-beam-heated divertor tokamak discharge. Its propagation is used to probe the plasma edge conditions at the H transition. The results show the existence of a transport barrier which forms at the plasma edge and impedes the flow of particles and energy across the plasma surface, giving rise to improved confinement properties. Location and extension of the barrier coincide with the edge zone of increased shear specific to the divertor configuration.

  4. Tokamak Plasmas : Mirnov coil data analysis for tokamak ADITYA

    Indian Academy of Sciences (India)

    D Raju; R Jha; P K Kaw; S K Mattoo; Y C Saxena; Aditya Team

    2000-11-01

    The spatial and temporal structures of magnetic signal in the tokamak ADITYA is analysed using recently developed singular value decomposition (SVD) technique. The analysis technique is first tested with simulated data and then applied to the ADITYA Mirnov coil data to determine the structure of current peturbation as the discharge progresses. It is observed that during the current rise phase, current perturbation undergoes transition from = 5 poloidal structure to = 4 and then to = 3. At the time of current termination, = 2 perturbation is observed. It is observed that the mode frequency remains nearly constant (≈10 kHz) when poloidal mode structure changes from = 4 to = 2. This may be either an indication of mode coupling or a consequences of changes in the plasma electron temperature and density scale length.

  5. Plasma equilibrium and instabilities in tokamaks

    International Nuclear Information System (INIS)

    A phenomenological introduction of some of the main theoretical and experimental features on equilibrium and instabilities in tokamaks is presented. In general only macroscopic effects are considered, being the plasma described as a fluid. (L.C.)

  6. Power and particle exhaust in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Stambaugh, R.D.

    1998-01-01

    The status of power and particle exhaust research in tokamaks is reviewed in the light of ITER requirements. There is a sound basis for ITER`s nominal design positions; important directions for further research are identified.

  7. Robust Sliding Mode Control for Tokamaks

    Directory of Open Access Journals (Sweden)

    I. Garrido

    2012-01-01

    Full Text Available Nuclear fusion has arisen as an alternative energy to avoid carbon dioxide emissions, being the tokamak a promising nuclear fusion reactor that uses a magnetic field to confine plasma in the shape of a torus. However, different kinds of magnetohydrodynamic instabilities may affect tokamak plasma equilibrium, causing severe reduction of particle confinement and leading to plasma disruptions. In this sense, numerous efforts and resources have been devoted to seeking solutions for the different plasma control problems so as to avoid energy confinement time decrements in these devices. In particular, since the growth rate of the vertical instability increases with the internal inductance, lowering the internal inductance is a fundamental issue to address for the elongated plasmas employed within the advanced tokamaks currently under development. In this sense, this paper introduces a lumped parameter numerical model of the tokamak in order to design a novel robust sliding mode controller for the internal inductance using the transformer primary coil as actuator.

  8. Tokamak research in the Soviet Union

    International Nuclear Information System (INIS)

    Important milestones on the way to the tokamak fusion reactor are recapitulated. Soviet tokamak research concentrated at the I.V. Kurchatov Institute in Moscow, the A.F. Ioffe Institute in Leningrad and the Physical-Technical Institute in Sukhumi successfully provides necessary scientific and technological data for reactor design. Achievments include, the successful operation of the first tokamak with superconducting windings (T-7) and the gyrotron set for microwave plasma heating in the T-10 tokamak. The following problems have intensively been studied: Various methods of additional plasma heating, heat and particle transport, and impurity control. The efficiency of electron-cyclotron resonance heating was demonstrated. In the Joule heating regime, both the heat conduction and diffusion rates are anomalously high, but the electron heat conduction rate decreases with increasing plasma density. Progress in impurity control makes it possible to obtain a plasma with effective charge approaching unity. (J.U.)

  9. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    This report discusses the following topics on the Aries-I Tokamak: Design description; systems studies and economics; reactor plasma physics; magnet engineering; fusion-power-ore engineering; and environmental and safety features

  10. Synchrotron radiation in inhomogeneous tokamak plasmas

    International Nuclear Information System (INIS)

    Synchrotron emission in a tokamak configuration with inhomogeneous plasma parameters is considered to investigate the effects of the temperature profile and vertical elongation on the radiation loss. Using the numerical solution of the transfer equation for ITER-like plasma parameters, several new results on the radiated energy in a Maxwellian plasma have been derived. In particular: (i) synchrotron loss is profile dependent, namely, at constant average thermal energy, the emitted radiation increases with the peak temperature, (ii) an analytical formula of the global loss in inhomogeneous tokamak plasmas with arbitrary vertical elongation is established, (iii) the maximum of the frequency emission spectrum is a linear function of the volume average temperature, (iiii) high frequency synchrotron radiation is entirely due to electrons with energy much greater than the thermal energy. The need for experimental investigations on synchrotron emission in present-day large tokamaks to determine the effect of reflections of the complex tokamak first wall is stressed

  11. Assessment of compatibility of ICRF antenna operation with full W wall in ASDEX Upgrade

    International Nuclear Information System (INIS)

    The compatibility of ICRF (ion cyclotron range of frequencies) antenna operation with high-Z plasma facing components is assessed in ASDEX Upgrade (AUG) with its tungsten (W) first wall. The mechanism of ICRF-related W sputtering was studied by various diagnostics including the local spectroscopic measurements of W sputtering yield YW on antenna limiters. Modification of one antenna with triangular shields, which cover the locations where long magnetic field lines pass only one out of two (0π)-phased antenna straps, did not influence the locally measured YW values markedly. In the experiments with antennas powered individually, poloidal profiles of YW on limiters of powered antennas show high YW close to the equatorial plane and at the very edge of the antenna top. The YW-profile on an unpowered antenna limiter peaks at the location projecting to the top of the powered antenna. An interpretation of the YW measurements is presented, assuming a direct link between the W sputtering and the sheath driving RF voltages deduced from parallel electric near-field (E||) calculations and this suggests a strong E|| at the antenna limiters. However, uncertainties are too large to describe the YW poloidal profiles. In order to reduce ICRF-related rise in W concentration CW, an operational approach and an approach based on calculations of parallel electric fields with new antenna designs are considered. In the operation, a noticeable reduction in YW and CW in the plasma during ICRF operation with W wall can be achieved by (a) increasing plasma-antenna clearance; (b) strong gas puffing; (c) decreasing the intrinsic light impurity content (mainly oxygen and carbon in AUG). In calculations, which take into account a realistic antenna geometry, the high E|| fields at the antenna limiters are reduced in several ways: (a) by extending the antenna box and the surrounding structures parallel to the magnetic field; (b) by increasing the average strap-box distance, e.g. by increasing the

  12. Confinement of 'Improved H-Modes' in the All-Tungsten ASDEX Upgrade

    International Nuclear Information System (INIS)

    Full text: 'Improved H-mode' discharges in ASDEX Upgrade (AUG) are characterized by enhanced confinement factors H98 > 1, βN =2 - 3.5 and a q-profile with almost zero shear in the core of the plasma at q(0) ∼ 1. One of the major goals of the AUG tungsten programme has been to demonstrate the compatibility of such high performance scenarios with an all-W wall. After the all-W AUG was boronised a clear reduction of the concentration of light impurities such as carbon and oxygen (C: 0.1-1%, O< 0.1%) was observed. The radiated power decreased, especially in the divertor, and the thermal load on the W-coated divertor tiles reached values above technical capabilities. Therefore, high performance discharges in the boronized AUG were only conducted with active cooling of the divertor plasma by enhancing the radiation with N seeding. As a positive surprise it turned out that N seeding does not only protect the divertor tiles, but also improves significantly the energy confinement. This is a reproducible effect which holds for all D fuelling rates under both freshly boronised and unboronised conditions. In contrast to earlier studies of improved confinement following impurity seeding, density peaking, which would be detrimental in an all-W device, can be excluded as a contributor. The main contribution is the increase in the plasma temperature both in the core and in the edge. Stability analyses of comparable discharges with and without N seeding using the GS2 and the GENE codes highlight the role of deuterium dilution in the reduction of the core ion heat transport due to the ITG mode, which is dominant under the experimental conditions. The reduced core heat transport, however, explains the experimentally observed total confinement improvement only to a certain extent. This paper will deal with the present status of AUG plasma operation of 'improved H-Mode' scenarios at optimized performance with boronized and unboronized tungsten walls. It will focus on confinement

  13. Edge plasma studies on the CASTOR tokamak

    Czech Academy of Sciences Publication Activity Database

    Hron, Martin; Peleman, P.; Spolaore, M.; Martines, E.; Hronová-Bilyková, Olena; Dejarnac, Renaud; Devynck, P.; Brotánková, Jana; Sentkerestiová, Jana; Ďuran, Ivan; Gunn, J.; Stöckel, Jan; Van Oost, G.; Adámek, Jiří; van de Peppel, L.; Štěpán, Michal

    Krakow : Euratom - IPPLM Association, 2006 - (Zagorski, R.), - [IEA Large Tokamak IA Workshop on Edge Transport in Fusion plasmas. Kraków (PL), 11.09.2006-13.09.2006] Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * plasma * scrape-off layer * turbulence * interchange instability Subject RIV: BL - Plasma and Gas Discharge Physics http://www.etfp2006.ifpilm.waw.pl/presentations.html

  14. The ETE spherical Tokamak project. IAEA report

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Gerson Otto; Del Bosco, E.; Berni, L.A.; Ferreira, J.G.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Barroso, J.J.; Castro, P.J.; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma]. E-mail: ludwig@plasma.inpe.br

    2002-07-01

    This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the historical development of the ETE (Spherical Tokamak Experiment) project, its research program, technical characteristics and operating conditions as of October, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  15. D-D tokamak reactor studies

    International Nuclear Information System (INIS)

    A tokamak D-D reactor design, utilizing the advantages of a deuterium-fueled reactor but with parameters not unnecessarily extended from existing D-T designs, is presented. Studies leading to the choice of a design and initial studies of the design are described. The studies are in the areas of plasma engineering, first-wall/blanket/shield design, magnet design, and tritium/fuel/vacuum requirements. Conclusions concerning D-D tokamak reactors are stated

  16. Plasma diagnostics using synchrotron radiation in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Fidone, I.; Giruzzi, G.; Granata, G.

    1995-09-01

    This report deal with the use of synchrotron radiation in tokamaks. The main advantage of this new method is that it enables to overcome several deficiencies, caused by cut-off, refraction, and harmonic overlap. It also makes it possible to enhance the informative contents of the familiar low harmonic scheme. The basic theory of the method is presented and illustrated by numerical applications, for plasma parameters of relevance in present and next step tokamaks. (TEC). 10 refs., 13 figs.

  17. Thermonuclear ignition in the next generation tokamaks

    International Nuclear Information System (INIS)

    The extrapolation of experimental rules describing energy confinement and magnetohydrodynamic - stability limits, in known tokamaks, allow to show that stable thermonuclear ignition equilibria should exist in this configuration, if the product aBtx of the dimensions by a magnetic-field power is large enough. Quantitative application of this result to several next-generation tokamak projects show that those kinds of equilibria could exist in such devices, which would also have enough additional heating power to promote an effective accessible ignition

  18. Epoxide insulation for Tokamak coils

    International Nuclear Information System (INIS)

    The construction and testing of 12-tonne toroidal-field electromagnets for the Joint European Torus by Brown Boveri and Cie (Mannheim) are described. The principle of Tokamak confinement of a plasma which acts as the secondary winding of a transformer is explained. The Cu conductors are sanded and coated with epoxide adhesive before being wrapped in 7mm thick woven glass fibre, dried by heating under vacuum, impregnated and encapsulated in 1.2 tonnes of Araldite, which is solidified under pressure of 4 atmospheres and hardened for ten hours at 1500C. The prototype withstood tests involving 25,000 flexure cycles at 1.1 MN and 2 Hz, 2,000 quarter-hour 10kA heating cycles between 840 and 200C, and exposure to 500 million rads. 32 such coils were constructed at the rate of one every three weeks. (M.B.D.)

  19. Tokamak plasma interaction with limiters

    International Nuclear Information System (INIS)

    The importance of plasma purity is first discussed in terms of the general requirements of controlled thermonuclear fusion. The tokamak approach to fusion and its inherent problem of plasma contamination are introduced. A main source of impurities is due to the bombardment of the limiter by energetic particles and thus the three main aspects of the plasma-limiter interaction are reviewed, boundary plasma conditions, fuelling/recycling and impurity production. The experiments, carried out on the DITE tokamak at Culham Laboratory, UK, investigated these three topics and the results are compared with predicted behaviour; new physical phenomena are presented in all three areas. Simple one-dimensional fluid equations are found to adequately describe the SOL plasma, except in regard to the pre-sheath electric field and ambipolarity; that is, the electric field adjacent to the limiter surface appears to be weak and the associated plasma flow can be non-ambipolar. Recycling of fuel particles from the limiter is observed to be near unity at all times. The break-up behaviour of recycled and gas puffed D2 molecules is dependent on the electron temperature, as expected. Impurity production at the limiter is chemical erosion of graphite being negligible. Deposition of limiter and wall-produced impurities is found on the limiter. The spatial distributions of impurities released from the limiter are observed and are in good agreement with a sputtered atom transport code. Finally, preliminary experiments on the transport of impurity ions along field lines away from the limiter have been performed and compared with simple analytic theory. The results suggest that the pre-sheath electric field in the SOL is much weaker than the simple fluid model would predict

  20. Global migration of {sup 13}C impurities in high-density L-mode plasmas in ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Hakola, A., E-mail: antti.hakola@vtt.fi [VTT, Association EURATOM-Tekes, P.O. Box 1000, 02044 VTT (Finland); Koivuranta, S.; Likonen, J. [VTT, Association EURATOM-Tekes, P.O. Box 1000, 02044 VTT (Finland); Groth, M.; Kurki-Suonio, T.; Lindholm, V.; Miettunen, J. [Aalto University, Association EURATOM-Tekes, P.O. Box 14100, 00076 AALTO (Finland); Krieger, K.; Mayer, M.; Müller, H.W.; Neu, R.; Rohde, V. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Boltzmannstr. 2, 85748 Garching (Germany); Petersson, P. [Royal Institute of Technology, Association EURATOM-VR, Teknikringen 31, 10044 Stockholm (Sweden)

    2013-07-15

    We have studied the migration of {sup 13}C in ASDEX Upgrade after a global impurity injection experiment in 2011. The main chamber was observed to be the largest deposition region for carbon: almost 35% of the injected atoms end up there. Moreover, gaps between wall tiles account for surface densities which are comparable to those on the plasma-facing surfaces. SOLPS modeling of the experiment produced a set of background plasmas and poloidal flow profiles for simulating the transport of {sup 13}C with ASCOT; a match with measured deposition, however, required using an imposed flow profile. ASCOT reproduced the observed localized deposition at the outer midplane but work is needed to explain the measured deposition at the inner side of the torus and at the top of the vessel.

  1. High resolution scanning transmission electron microscopy (HR STEM) analysis of re-deposited layer on ASDEX Upgrade tile

    Energy Technology Data Exchange (ETDEWEB)

    Rasinski, M., E-mail: mrasin@o2.pl [Warsaw University of Technology, Faculty of Material Science and Engineering, Woloska 141, 02-507 Warsaw (Poland); Fortuna-Zalesna, E. [Warsaw University of Technology, Faculty of Material Science and Engineering, Woloska 141, 02-507 Warsaw (Poland); Mayer, M.; Neu, R. [Max-Planck-Institut fuer Plasmaphysik, Euratom Association, Boltzmannstrasse 2, D-85748 Garching (Germany); Plocinski, T.; Lewandowska, M.; Kurzydlowski, K.J. [Warsaw University of Technology, Faculty of Material Science and Engineering, Woloska 141, 02-507 Warsaw (Poland)

    2011-10-15

    Erosion and re-deposition of plasma-facing components (PFCs) is one of the most important issues in fusion devices and as such it is an area of interest for many research groups. However, the structure and composition of re-deposited layers as well as the mechanism and condition of their formation are not yet fully described and understood. In the present study, the structure and the composition of co-deposited layers, which developed at the outer divertor strike point tiles in ASDEX Upgrade during the 2009 campaign were examined. High resolution scanning transmission electron microscopy (HRSTEM) combined with energy-dispersive X-ray spectroscopy (EDS) and electron energy loss spectroscopy (EELS) have been used to identify deposits composition and morphology. Tungsten foam like structure and co-deposits rich in tungsten, oxygen, carbon, boron and nitrogen were observed.

  2. High resolution scanning transmission electron microscopy (HR STEM) analysis of re-deposited layer on ASDEX Upgrade tile

    International Nuclear Information System (INIS)

    Erosion and re-deposition of plasma-facing components (PFCs) is one of the most important issues in fusion devices and as such it is an area of interest for many research groups. However, the structure and composition of re-deposited layers as well as the mechanism and condition of their formation are not yet fully described and understood. In the present study, the structure and the composition of co-deposited layers, which developed at the outer divertor strike point tiles in ASDEX Upgrade during the 2009 campaign were examined. High resolution scanning transmission electron microscopy (HRSTEM) combined with energy-dispersive X-ray spectroscopy (EDS) and electron energy loss spectroscopy (EELS) have been used to identify deposits composition and morphology. Tungsten foam like structure and co-deposits rich in tungsten, oxygen, carbon, boron and nitrogen were observed.

  3. Progress in the prediction of disruptions in ASDEX-Upgrade via neural and fuzzy-neural techniques

    International Nuclear Information System (INIS)

    The paper addresses the problem of predicting the onset of a disruption on the basis of some known precursors possibly announcing the event. The availability in real time of a large set of diagnostic signals allows us to collectively interpret the data in order to decide whether we are near a disruption or during a normal operation scenario. As a relevant experimental example, a database of disruptive discharges in ASDEX-Upgrade has been analysed in this work. Both Neural Networks (NN's) and Fuzzy Inference Systems (FIS) have been investigated as suitable tools to cope with the prediction problem. The experimental database has been exploited aiming to gain information about the mechanisms which drive the plasma column to a disruption. The proposed processor will operate by implementing a classification of the shot type, and outputting a real number that indicates the time left before the disruption will effectively take place (ttd). (author)

  4. Reflectometry measurements of the m=1 satellite mode in L- and H-mode plasmas in ASDEX

    International Nuclear Information System (INIS)

    In ASDEX, with strong NBI heating, often a large central m=1, n=1 mode is observed on the SXR emission. For PNBI ≥ 1 MW a mode rotating with the same frequency, the so-called 'm=1 satellite', is seen on the magnetic pick-up coils in the L and H-phases. Magnetic measurements in the divertor chamber suggest that the satellite mode might be located outside the separatrix, on open field lines reaching the divertor. Here we present results from localized microwave reflectometric measurements. The time evolution of the satellite mode frequency is studied for plasmas with different qa and the mode localization is estimated, confirming that it should be close to but outside the separatrix. The central toroidal rotation velocities of the plasma can be inferred from the measured frequencies of the satellite modes. (orig.)

  5. Simulations of gas puff effects on edge density and ICRF coupling in ASDEX upgrade using EMC3-Eirene

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, W., E-mail: wei.zhang@ipp.mpg.de [Applied Physics Department, University of Ghent, Ghent (Belgium); Max-Planck-Institut für Plasmaphysik, Garching (Germany); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Lunt, T.; Bobkov, V.; Coster, D.; Brida, D. [Max-Planck-Institut für Plasmaphysik, Garching (Germany); Noterdaeme, J.-M. [Applied Physics Department, University of Ghent, Ghent (Belgium); Max-Planck-Institut für Plasmaphysik, Garching (Germany); Jacquet, P. [CCFE, Culham Science Centre, Abingdon (United Kingdom); Feng, Y. [Max-Planck-Institut für Plasmaphysik, Greifswald (Germany)

    2015-12-10

    Simulations were carried out with the 3D plasma transport code EMC3-EIRENE, to study the deuterium gas (D{sub 2}) puff effects on edge density and the coupling of Ion Cyclotron Range of Frequency (ICRF) power in ASDEX Upgrade. Firstly we simulated an inter-ELM phase of an H-mode discharge with a moderate (1.2 × 10{sup 22} electrons/s) lower divertor gas puff. Then we changed the gas source positions to the mid-plane or top of machine while keeping other conditions the same. Cases with different mid-plane or top gas valves are investigated. Our simulations indicate that compared to lower divertor gas puffing, the mid-plane gas puff can enhance the local density in front of the antennas most effectively, while a rather global (toroidally uniform) but significantly smaller enhancement is found for top gas puffing. Our results show quantitative agreement with the experiments.

  6. Comparison of experiment and models of geodesic acoustic mode frequency and amplitude geometric scaling in ASDEX Upgrade

    Science.gov (United States)

    Simon, P.; Conway, G. D.; Stroth, U.; Biancalani, A.; Palermo, F.; the ASDEX Upgrade Team

    2016-04-01

    In a set of dedicated ASDEX Upgrade shape-scan experiments, the influence of plasma geometry on the frequency and amplitude behaviour of the geodesic acoustic mode (GAM), measured by Doppler reflectometry, is studied. In both limiter and divertor configurations, the plasma elongation was varied between circular and highly elongated states (1.1effects. The experimentally observed effect of decreasing {ω\\text{GAM}} with increasing κ is predicted by most models. Other geometric factors, such as inverse aspect ratio ε and Shafranov shift gradient {Δ\\prime} are also seen to be influential in determining a reliable lower {ω\\text{GAM}} boundary. The GAM amplitude is found to vary with boundary elongation {κ\\text{b}} and safety factor q. The collisional damping is compared to multiple models for the collisionless damping. Collisional damping appears to play a stronger role in the divertor configuration, while collisional and collisionless damping both may contribute to the GAM amplitude in the limiter configuration.

  7. Sawtooth control using electron cyclotron current drive in the presence of energetic particles in high performance ASDEX Upgrade plasmas

    CERN Document Server

    Chapman, I T; Maraschek, M; McCarthy, P J; Tardini, G

    2013-01-01

    Sawtooth control using steerable electron cyclotron current drive (ECCD) has been demonstrated in ASDEX Upgrade plasmas with a significant population of energetic ions in the plasma core and long uncontrolled sawtooth periods. The sawtooth period is found to be minimised when the ECCD resonance is swept to just inside the q = 1 surface. By utilising ECCD inside q = 1 for sawtooth control, it is possible to avoid the triggering of neoclassical tearing modes, even at significnatly higher pressure than anticipated in the ITER baseline scenario. Operation at 25% higher normalised pressure has been achieved when only modest ECCD power is used for sawtooth control compared to identical discharges without sawtooth control when neo-classical tearing modes are triggered by the sawteeth. Modelling suggests that the destabilisation arising from the change in the local magnetic shear caused by the ECCD is able to compete with the stabilising influence of the energetic particles inside the q = 1 surface.

  8. Simulations of gas puff effects on edge density and ICRF coupling in ASDEX upgrade using EMC3-Eirene

    International Nuclear Information System (INIS)

    Simulations were carried out with the 3D plasma transport code EMC3-EIRENE, to study the deuterium gas (D2) puff effects on edge density and the coupling of Ion Cyclotron Range of Frequency (ICRF) power in ASDEX Upgrade. Firstly we simulated an inter-ELM phase of an H-mode discharge with a moderate (1.2 × 1022 electrons/s) lower divertor gas puff. Then we changed the gas source positions to the mid-plane or top of machine while keeping other conditions the same. Cases with different mid-plane or top gas valves are investigated. Our simulations indicate that compared to lower divertor gas puffing, the mid-plane gas puff can enhance the local density in front of the antennas most effectively, while a rather global (toroidally uniform) but significantly smaller enhancement is found for top gas puffing. Our results show quantitative agreement with the experiments

  9. JT-60SA project for JA-EU broader approach satellite tokamak and national centralized tokamak

    International Nuclear Information System (INIS)

    JT-60 Super Advanced (JT-60SA) project is the joint project of ITER satellite tokamak by Japan and EU with Japanese Tokamak. The background, objects, device design, management of JT-60SA is stated. It consists of six chapters: the first chapter describes introduction, the second chapter states the objects of tokamak device complementing ITER, the third chapter contains research subjects and device performance such as plasma performance and demand for devices, operation scenario, control of MHD instability, and control of heat and particles, the forth chapter design of devices, the fifth chapter management and the sixth conclusion. In order to realize prototype reactor, improvement research of tokamak, development of reactor engineering technology, fusion reactor researches, tokamak theory and simulation, and social and environment safety research has to be advanced. (S.Y.)

  10. Vacuum system for HL-2A tokamak

    International Nuclear Information System (INIS)

    The vacuum system for HL-2A was built in 2003. The test results indicated that this system is feasible. It consists of three main parts: a pumping system, a pumping divertor and a glow discharge cleaning (GDC) system. For the pumping system, there are three main functions: (1) evacuating the vacuum vessel thus to produce an ultra high vacuum, (2) removal of impurities released during baking and (3) pumping during GDC. The pumping divertor controls the particles at the plasma edge and the GDC system provides a clean wall conditioning. During the first campaign of physical trial experiment on HL-2A, the ultimate pressure reached 4.6 x 10-6 Pa, and the total leakage and outgassing rate in 12 hours was 1.8 x 10-5 Pa·m3/s, which is close to that of ASDEX. (authors)

  11. Vacuum System for HL-2A Tokamak

    Institute of Scientific and Technical Information of China (English)

    曹曾; 崔成和; 刘德权; 蔡萧; 高霄燕

    2005-01-01

    The vacuum system for HL-2A was built in 2003. The test results indicated that this system is feasible. It consists of three main parts: a pumping system, a pumping divertor and a glow discharge cleaning (GDC) system. For the pumping system, there are three main functions:(1) evacuating the vacuum vessel thus to produce an ultra high vacuum, (2) removal of impurities released during baking and (3) pumping during GDC. The pumping divertor controls the particles at the plasma edge and the GDC system provides a clean wall conditioning. During the first campaign of physical trial experiment on HL-2A, the ultimate pressure reached 4.6×10-6 Pa, and of ASDEX.

  12. Automatic setting of machine control with physics operation parameters at ASDEX Upgrade

    International Nuclear Information System (INIS)

    Performing plasma fusion experiments in a tokamak relies on close cooperation between discharge control (DCS) and machine control (MCS) systems, where the DCS's task is generating and handling the plasma, and the MCS is responsible for providing and configuring the necessary actuators. With a reduced number of experiment cycles per day, in consequence of long pulses or ambitious discharge scenarios, optimal utilization of resources as well as avoiding mismatches between the MCS' and DCS' configurations becomes increasingly important. Setting up the MCS for discharges currently relies on informal communication and requires careful attention of operators and session leaders. To enforce correspondence between MCS and DCS parametrizations, an automatic mediator agent formalizes the pre-shot configuration procedure. Functionality of the mediator and protocols for embedding into the existing system are discussed

  13. Particle and energy balances in tokamak plasmas

    International Nuclear Information System (INIS)

    Computational and experimental studies on particle and energy balances in tokamak plasmas are described. Firstly, concerning the modeling of tokamak plasmas, the particle balance considering diffusion and recycling, and the energy balance considering transport and energy losses due to impurities are discussed. Production mechanisms of gaseous and metallic impurities, which play important role in tokamak plasmas, are also discussed from a viewpoint of plasma-wall interactions. Scaling laws of density, temperature and energy confinement time are shown on the basis of recent data. Secondarily, tokamak plasmas are simulated with the above model, and anomalous diffusion and electron thermal conduction are indicated. Characteristics of a future tokamak plasma are also simulated. Stationary impurity density distributions and related energy losses, such as bremsstrahlung, ionization and excitation, are calculated taking into account diffusion and ionization processes. Edge cooling by oxygen impurities is described quantitatively compared with experiments. Permissible impurity levels of carbon, oxygen and iron in future large tokamaks are estimated. Thirdly, experimental studies on surface cleaning methods of the first wall are described; discharge cleaning in JFT-2, baking effect on the outgassing rates of wall materials, surface treatment of high-temperature molybdenum by oxygen and hydrogen gases, and in-situ coating of molybdenum by a coaxial magnetron sputter method. Lastly, problems in future large tokamaks aiming at break-even or self-ignited plasma are discussed quantitatively, such as trapped particle instabilities, impurities and additional heating. It is predicted that new conceptions will be necessary to overcome the problems and attain the fusion goal. (auth.)

  14. The JT-60 tokamak machine

    International Nuclear Information System (INIS)

    JT-60 is a large tokamak experimental device under construction at JAERI with main device parameters of R=3.0m, a=0.95m, Bsub(t)=45kG, and Isub(p)=2.7Ma. Its basic aim is to produce and confine hydrogen plasmas of temperatures in a multi-keV range and of confinement times comparable to a second, and to study its plasma-physics properties as well as engineering problems associated with them. The JT-60 tokamak machine is mainly composed of a vacuum vessel, toroidal field (TF) coils, poloidal field (PF) coils, and support structures. The vacuum vessel is a high toroidal chamber with an egg-shaped crossection, consisting of sectorial rigid rings and parallel bellows made from Inconel 625. It is baked out at a maximum temperature up to 5000C. Several kinds of first walls made from molybdenum are bolt-jointed to the vacuum vessel for its protection. The vacuum vessel is almost completely finished with design and is deeply into manufacturing. The TF system consists of 18 unit coils located around a torus axis at regular intervals. The unit coil composed of two pancakes are wedge-shaped at the section close to a torus axis and encased in a high-manganese non-magnetic steel case. Fabrication of the TF coils will be finished in May 1981. The PF coils are composed of ohmic heating coils, vertical field coils, horizontal field coils, and quadrupole field coils located inside the TF coil bore and outside the vacuum vessel, and magnetic limiter coils placed in the vacuum vessel. Its mechanical and thermal design is almost completed are composed of the upper and lower support structures, support comuns of the vacuum vessel, and central column made from high-manganese non-magnetic steel. The structural analysis was completed including a seismic analysis and the fabrication is now in progress. The first plasma is expected to be produced in October 1984. (orig.)

  15. Plasma Physics Regimes in Tokamaks with Li Walls

    International Nuclear Information System (INIS)

    Low recycling regimes with a plasma limited by a lithium wall surface suggest enhanced stability and energy confinement, both necessary for tokamak reactors. These regimes could make ignition feasible in compact tokamaks. Ignited Spherical Tokamaks (IST), self-sufficient in the bootstrap current, are introduced as a necessary step for development of the physics and technology of power reactors

  16. Small tokamaks for fusion technology testing

    International Nuclear Information System (INIS)

    Small steady-state tokamaks for testing divertors and fusion nuclear technologies are considered. Based on present physics and technology data and explanation to reduce R0/a, H-D-fueled tokamaks with R0 ∼ 0.6--0.75 m, R0/a ∼ 1.8--2.5, and Bt0 ∼ 1.4--2.2 T can be driven with Ptot ∼ 4.5 MW to maintain Ip ∼ 0.5 MA and produce the ITER-level plasma edge and divertor conditions. Given an adequate steady-state divertor solution and Q∼1 operation based on fusion through the suprathermal component, D-T-fueled tokamaks with R0 ∼ 0.8 m, R0/a ∼ 2, and Bt0 ∼ 4 T can be driven with Ptot ∼ 15 MW to maintain Ip ∼ 4.6 MA and produce an peak neutron wall load WL ∼ 1 MW/m2. Such devices appear possible if the plasma properties at the power R0/a remain tokamak-like and, for the D-T case, can unshielded center core is feasible. The use of a single conductor as the inboard leg of the toroidal field coils for this purpose is discussed. The physics issues and the design features are identified for such tokamaks with a testing duty for factor goal of 10--20%

  17. Three novel tokamak plasma regimes in TFTR

    International Nuclear Information System (INIS)

    Aside from extending ''standard'' ohmic and neutral beam heating studies to advanced plasma parameters, TFTR has encountered a number of special plasma regimes that have the potential to shed new light on the physics of tokamak confinement and the optimal design of future D-T facilities: (1) High-powered, neutral beam heating at low plasma densities can maintain a highly reactive hot-ion population (with quasi-steady-state beam fueling and current drive) in a tokamak configuration of modest bulk-plasma confinement requirements. (2) Plasma displacement away from limiter contact lends itself to clarification of the role of edge-plasma recycling and radiation cooling within the overall pattern of tokamak heat flow. (3) Noncentral auxiliary heating (with a ''hollow'' power-deposition profile) should serve to raise the central tokamak plasma temperature without deterioration of central region confinement, thus facilitating the study of alpha-heating effects in TFTR. The experimental results of regime (3) support the theory that tokamak profile consistency is related to resistive kink stability and that the global energy confinement time is determined by transport properties of the plasma edge region

  18. Three novel tokamak plasma regimes in TFTR

    Energy Technology Data Exchange (ETDEWEB)

    Furth, H.P.

    1985-10-01

    Aside from extending ''standard'' ohmic and neutral beam heating studies to advanced plasma parameters, TFTR has encountered a number of special plasma regimes that have the potential to shed new light on the physics of tokamak confinement and the optimal design of future D-T facilities: (1) High-powered, neutral beam heating at low plasma densities can maintain a highly reactive hot-ion population (with quasi-steady-state beam fueling and current drive) in a tokamak configuration of modest bulk-plasma confinement requirements. (2) Plasma displacement away from limiter contact lends itself to clarification of the role of edge-plasma recycling and radiation cooling within the overall pattern of tokamak heat flow. (3) Noncentral auxiliary heating (with a ''hollow'' power-deposition profile) should serve to raise the central tokamak plasma temperature without deterioration of central region confinement, thus facilitating the study of alpha-heating effects in TFTR. The experimental results of regime (3) support the theory that tokamak profile consistency is related to resistive kink stability and that the global energy confinement time is determined by transport properties of the plasma edge region.

  19. Electron thermal transport in tokamak plasmas

    International Nuclear Information System (INIS)

    The process of fusion of small nuclei thereby releasing energy, as it occurs continuously in the sun, is essential for the existence of mankind. The same process applied in a controlled way on earth would provide a clean and an abundant energy source, and be the long term solution of the energy problem. Nuclear fusion requires an extremely hot (108 K) ionized gas, a plasma, that can only be maintained if it is kept insulated from any material wall. In the so called 'tokamak' this is achieved by using magnetic fields. The termal insulation, which is essential if one wants to keep the plasma at the high 'fusion' temperature, can be predicted using basic plasma therory. A comparison with experiments in tokamaks, however, showed that the electron enery losses are ten to hundred times larger than this theory predicts. This 'anomalous transport' of thermal energy implies that, to reach the condition for nuclear fusion, a fusion reactor must have very large dimensions. This may put the economic feasibility of fusion power in jeopardy. Therefore, in a worldwide collaboration, physicists study tokamak plasmas in an attempt to understand and control the energy losses. From a scientific point of view, the mechanisms driving anomalous transport are one of the challenges in fudamental plasma physics. In Nieuwegein, a tokamak experiment (the Rijnhuizen Tokamak Project, RTP) is dedicated to the study of anomalous transport, in an international collaboration with other laboratories. (orig./WL)

  20. Simulation of burning tokamak plasmas

    International Nuclear Information System (INIS)

    To simulate dynamical behaviour of tokamak fusion reactors, a zero-dimensional time-dependent particle and power balance code has been developed. The zero-dimensional plasma model is based on particle and power balance equations that have been integrated over the plasma volume using prescribed profiles for plasma parameters. Therefore, the zero-dimensional model describes the global dynamics of a fusion reactor. The zero-dimensional model has been applied to study reactor start-up, and plasma responses to changes in the plasma confinement, fuelling rate, and impurity concentration, as well as to study burn control via fuelling modulation. Predictions from the zero-dimensional code have been compared with experimental data and with transport calculations of a higher dimensionality. In all cases, a good agreement was found. The advantage of the zero-dimensional code, as compared to higher-dimensional transport codes, is the possibility to quickly scan the interdependencies between reactor parameters. (88 refs., 58 figs., 6 tabs.)

  1. THOR tokamak magnetic field system

    International Nuclear Information System (INIS)

    The THOR Machine is an iron cored Tokamak having a major radius of 0.52 m and a minor radius of 0.17 m giving an aspect ratio of 3:1. It has a low ripple toroidal field of 1 T and an iron core giving 0.24 Vs. The maximum plasma current is expected to be in the region of 80x103 A. The maximum toroidal field ripple on axis is of the order of 0.01% and 2.5% at the plasma edge. The equilibrium of the plasma is achieved by means of a D.C. vertical field and a 1 cm thick copper shell. The D.C. field is cancelled during the rise time of the plasma current by means of pulsed reverse vertical field windings placed between the copper shell and the vacuum vessel. The design of this field system represents a compromise between obtaining adequate field penetration through the relatively thin vacuum vessel and maintaining the mechanical strength necessary to withstand the transient magnetic forces. Energy for the toroidal field system is supplied by a 15 kV 600 kJ capacitor bank and for the ohmic heating and reverse vertical fields by 5 kV 25 kJ and 50 kJ banks respectively. The problems encountered in the design, development and manufacture of these field systems are discussed. (author)

  2. Stability analysis of tokamak plasmas

    International Nuclear Information System (INIS)

    In a tokamak plasma, the energy transport is mainly turbulent. In order to increase the fusion reactions rate, it is needed to improve the energy confinement. The present work is dedicated to the identification of the key parameters leading to plasmas with a better confined energy in order to guide the future experiments. For this purpose, a numerical code has been developed. It calculates the growth rates characterizing the instabilities onset. The stability analysis is completed by the evaluation of the shearing rate of the rotation due to the radial electric field. When this shearing rate is greater than the growth rate the ion turbulence is fully stabilised. The shearing rate and the growth rate are determined from the density, temperature and security factor profiles of a given plasma. Three types of plasmas have been analysed. In the Radiative Improved modes of TEXTOR, high charge number ions seeding lowers the growth rates. In Tore Supra-high density plasmas, a strong magnetic shear and/or a more efficient ion heating linked to a bifurcation of the toroidal rotation direction (which is not understood) trigger the improvement of the confinement. In other Tore Supra plasmas, locally steep electron pressure gradients have been obtained following magnetic shear reversal. This locally negative magnetic shear has a stabilizing effect. In these three families of plasmas, the growth rates decrease, the confinement improves, the density and temperature profiles are steeper. This steepening induces an increase of the rotation shearing rate, which then maintains the confinement high quality. (author)

  3. Microtearing modes in tokamak discharges

    Science.gov (United States)

    Rafiq, T.; Weiland, J.; Kritz, A. H.; Luo, L.; Pankin, A. Y.

    2016-06-01

    Microtearing modes (MTMs) have been identified as a source of significant electron thermal transport in tokamak discharges. In order to describe the evolution of these discharges, it is necessary to improve the prediction of electron thermal transport. This can be accomplished by utilizing a model for transport driven by MTMs in whole device predictive modeling codes. The objective of this paper is to develop the dispersion relation that governs the MTM driven transport. A unified fluid/kinetic approach is used in the development of a nonlinear dispersion relation for MTMs. The derivation includes the effects of electrostatic and magnetic fluctuations, arbitrary electron-ion collisionality, electron temperature and density gradients, magnetic curvature, and the effects associated with the parallel propagation vector. An iterative nonlinear approach is used to calculate the distribution function employed in obtaining the nonlinear parallel current and the nonlinear dispersion relation. The third order nonlinear effects in magnetic fluctuations are included, and the influence of third order effects on a multi-wave system is considered. An envelope equation for the nonlinear microtearing modes in the collision dominant limit is introduced in order to obtain the saturation level. In the limit that the mode amplitude does not vary along the field line, slab geometry, and strong collisionality, the fluid dispersion relation for nonlinear microtearing modes is found to agree with the kinetic dispersion relation.

  4. Tokamak x ray diagnostic instrumentation

    International Nuclear Information System (INIS)

    Three classes of x-ray diagnostic instruments enable measurement of a variety of tokamak physics parameters from different features of the x-ray emission spectrum. (1) The soft x-ray (1 to 50 keV) pulse-height-analysis (PHA) diagnostic measures impurity concentrations from characteristic line intensities and the continuum enhancement, and measures the electron temperature from the continuum slope. (2) The Bragg x-ray crystal spectrometer (XCS) measures the ion temperature and neutral-beam-induced toroidal rotation velocity from the Doppler broadening and wavelength shift, respectively, of spectral lines of medium-Z impurity ions. Impurity charge state distributions, precise wavelengths, and inner-shell excitation and recombination rates can also be studied. X rays are diffracted and focused by a bent crystal onto a position-sensitive detector. The spectral resolving power E/ΔE is greater than 104 and time resolution is 10 ms. (3) The x-ray imaging system (XIS) measures the spatial structure of rapid fluctuations (0.1 to 100 kHZ) providing information on MHD phenomena, impurity transport rates, toroidal rotation velocity, plasma position, and the electron temperature profile. It uses an array of silicon surface-barrier diodes which view different chords of the plasma through a common slot aperture and operate in current (as opposed to counting) mode. The effectiveness of shields to protect detectors from fusion-neutron radiation effects has been studied both theoretically and experimentally

  5. Simulation of runaway electrons in Tokamak disruptions

    International Nuclear Information System (INIS)

    Self-consistent modelling of the generation of runaway electrons and the evolution of the toroidal electric field during tokamak disruptions is presented. The process of runaway generation is analysed by combining a relativistic kinetic equation for the electrons with Maxwell's equations for the electric field. Such modelling allows for a quantitative assessment of the runaway generation during disruptions in present day tokamak experiments, and to extrapolate to future tokamaks like ITER. It is found that the current profile can change dramatically during a disruption, such that the post disruption current, carried mainly by the runaway electrons, is significantly more peaked than the current profile before the disruption. In fact, it is found that the central current density can increase in spite of a reduction in the total current. (authors)

  6. Activation analysis of the compact ignition tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Selcow, E.C.

    1986-01-01

    The US fusion program has completed the conceptual design of a compact tokamak device that achieves ignition. The high neutron wall loadings associated with this compact deuterium-tritium-burning device indicate that radiation-related issues may be significant considerations in the overall system design. Sufficient shielding will be requied for the radiation protection of both reactor components and occupational personnel. A close-in igloo shield has been designed around the periphery of the tokamak structure to permit personnel access into the test cell after shutdown and limit the total activation of the test cell components. This paper describes the conceptual design of the igloo shield system and discusses the major neutronic concerns related to the design of the Compact Ignition Tokamak.

  7. Activation analysis of the compact ignition tokamak

    International Nuclear Information System (INIS)

    The US fusion program has completed the conceptual design of a compact tokamak device that achieves ignition. The high neutron wall loadings associated with this compact deuterium-tritium-burning device indicate that radiation-related issues may be significant considerations in the overall system design. Sufficient shielding will be requied for the radiation protection of both reactor components and occupational personnel. A close-in igloo shield has been designed around the periphery of the tokamak structure to permit personnel access into the test cell after shutdown and limit the total activation of the test cell components. This paper describes the conceptual design of the igloo shield system and discusses the major neutronic concerns related to the design of the Compact Ignition Tokamak

  8. Effect of impurity radiation on tokamak equilibrium

    International Nuclear Information System (INIS)

    The energy loss from a tokamak plasma due to the radiation from impurities is of great importance in the overall energy balance. Taking the temperature dependence of this loss for two impurities characteristic of those present in existing tokamak plasmas, the condition for radial power balance is derived. For the impurities considered (oxygen and iron) it is found that the radiation losses are concentrated in a thin outer layer of the plasma and the equilibrium condition places an upper limit on the plasma paraticle number density in this region. This limiting density scales with mean current density in the same manner as is experimentally observed for the peak number density of tokamak plasmas. The stability of such equilibria is also discussed. (author)

  9. Mass spectrometry instrumentation in TN (Novillo Tokamak)

    International Nuclear Information System (INIS)

    The mass spectrophotometry in the residual gases analysis in high vacuum systems, in particular in the Novillo Tokamak (TN), where pressures are required to be of the order 10-7 Torr, is carried out through an instrumental support with infrastructure configured in parallel to the experimental planning in this device. In the Novillo as well as other Tokamaks, it is necessary to condition the vacuum chamber for improving the main discharge parameters. At the present time, in this Tokamak the conditioning quality is presented determined by means of a mass spectrophotometer. A general instrumental description is presented associated with the Novillo conditioning, as well as the spectras obtained before and after operation. (Author)

  10. Current drive by spheromak injection into a tokamak

    OpenAIRE

    Brown, M. R.; Bellan, P. M.

    1990-01-01

    We report the first observation of current drive by injection of a spheromak plasma into a tokamak (Caltech ENCORE small reasearch tokamak) due to the process of helicity injection. After an abrupt 30% increase, the tokamak current decays by a factor of 3 due to plasma cooling caused by the merging of the relatively cold spheromak with the tokamak. The tokamak density profile peaks sharply due to the injected spheromak plasma (n¯3 increases by a factor of 6) then becomes hollow, suggestive of...

  11. Periodic disruptions in the MT-1 tokamak

    International Nuclear Information System (INIS)

    Disruptive instabilities are common phenomena in toroidal devices, especially in tokamaks. Three types can be distinguished: internal, minor and major disruptions. Periodic minor disruptions in the MT-1 tokamak were measured systematically with values of the limiter safety factor between 4 and 10. The density limit as a function of plasma current and horizontal displacement was investigated. Precursor oscillations always appear before the instability with increasing amplitude but can be observed at the density limit with quasi-stationary amplitude. Phase correlation between precursor oscillations were measured with Mirnov coils and x-ray detectors, and they show good agreement with a simple magnetic island model. (R.P.) 11 refs.; 6 figs

  12. Can better modelling improve tokamak control?

    International Nuclear Information System (INIS)

    The control of present day tokamaks usually relies upon primitive modelling and TCV is used to illustrate this. A counter example is provided by the successful implementation of high order SISO controllers on COMPASS-D. Suitable models of tokamaks are required to exploit the potential of modern control techniques. A physics based MIMO model of TCV is presented and validated with experimental closed loop responses. A system identified open loop model is also presented. An enhanced controller based on these models is designed and the performance improvements discussed. (author) 5 figs., 9 refs

  13. Tokamak power systems studies, FY 1985

    Energy Technology Data Exchange (ETDEWEB)

    Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Smith, D.L.; Sze, D.K.

    1985-12-01

    The Tokamak Power System Studies (TPSS) at ANL in FY-1985 were devoted to exploring innovative design concepts which have the potential for making substantial improvements in the tokamak as a commercial power reactor. Major objectives of this work included improved reactor economics, improved environmental and safety features, and the exploration of a wide range of reactor plant outputs with emphasis on reduced plant sizes compared to STARFIRE. The activities concentrated on three areas: plasma engineering, impurity control, and blanket/first wall/shield technology. 205 refs., 125 figs., 107 tabs.

  14. First experiments with SST-1 tokamak

    International Nuclear Information System (INIS)

    Full text: SST-1, a steady state superconducting tokamak, is at advanced stage of erection at the Institute for Plasma Research. The objectives of SST-1 include studying the physics of the plasma processes in tokamak under steady state conditions and learning technologies related to the steady state operation of the tokamak. These studies are expected to contribute to the tokamak physics database for very long pulse operations. The SST-1 tokamak is a large aspect ratio tokamak, configured to run double null diverted plasmas with significant elongation and triangularity. The machine has a major radius of 1.1 m, minor radius of 0.20 m, a toroidal field of 3.0 T at plasma center and a plasma current of 220 kA. Hydrogen gas will be used and plasma discharge duration will be 1000 s. Superconducting (SC) magnets are deployed for both the toroidal and poloidal field coils in SST-1. An Ohmic transformer is provided for plasma breakdown and initial current ramp up. SST-1 deploys a fully welded ultra high vacuum vessel, made up of 16 vessel sectors having ports and 16 rings with D- shaped cross-section, which are welded in-situ during the SST-1 assembly. Liquid nitrogen cooled radiation shield are deployed between the vacuum vessel and SC magnets as well as Sc magnets and cryostat, to minimize the radiation losses at the Sc magnets. In SST-1 tokamak, the auxiliary current drive will be based on 1.0 MW of Lower Hybrid current drive (LHCD) at 3.7 GHz. Auxiliary heating systems include 1 MW of Ion Cyclotron Resonance Frequency system (ICRF) at 22 MHz to 91 MHz, 0.2 MW of Electron Cyclotron Resonance heating at 84 GHz and a Neutral Beam Injection (NBI) system with peak power of 0.8 MW (at 80 keV) with variable beam energy in range of 10-80 keV. The ICRF system would also be used for initial breakdown and wall conditioning experiments. The assembly of the SST-1 tokamak is nearing completion. The cool down of the Superconducting magnets is scheduled to start by middle of year 2004

  15. Electron cyclotron emission diagnostics on KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, S. H. [Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Daejeon 305-353 (Korea, Republic of); Lee, K. D.; Kwon, M. [National Fusion Research Institute, 113 Gwahangno, Daejeon 305-333 (Korea, Republic of); Kogi, Y. [Fukuoka Institute of Technology, Higashiku, Fukuoka 811-0295 (Japan); Kawahata, K.; Nagayama, Y. [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Mase, A. [KASTEC, Kyushu University, Kasuga, Fukuoka 816-8580 (Japan)

    2010-10-15

    A new electron cyclotron emission (ECE) diagnostics system was installed for the Second Korea Superconducting Tokamak Advanced Research (KSTAR) campaign. The new ECE system consists of an ECE collecting optics system, an overmode circular corrugated waveguide system, and 48 channel heterodyne radiometer with the frequency range of 110-162 GHz. During the 2 T operation of the KSTAR tokamak, the electron temperatures as well as its radial profiles at the high field side were measured and sawtooth phenomena were also observed. We also discuss the effect of a window on in situ calibration.

  16. Electron cyclotron emission diagnostics on KSTAR tokamak.

    Science.gov (United States)

    Jeong, S H; Lee, K D; Kogi, Y; Kawahata, K; Nagayama, Y; Mase, A; Kwon, M

    2010-10-01

    A new electron cyclotron emission (ECE) diagnostics system was installed for the Second Korea Superconducting Tokamak Advanced Research (KSTAR) campaign. The new ECE system consists of an ECE collecting optics system, an overmode circular corrugated waveguide system, and 48 channel heterodyne radiometer with the frequency range of 110-162 GHz. During the 2 T operation of the KSTAR tokamak, the electron temperatures as well as its radial profiles at the high field side were measured and sawtooth phenomena were also observed. We also discuss the effect of a window on in situ calibration. PMID:21033954

  17. A method for tokamak neutronics calculations

    International Nuclear Information System (INIS)

    This paper presents a new method for neutron transport calculation in tokamak fusion reactors. The computational procedure is based on the solution of the even-parity transport equation in a toroidal geometry. The angular neutron distribution is treated by even-parity spherical harmonic expansion, while the spatial dependence is approximated by using R-function finite elements that are defined for regions of arbitrary geometric shape. In order to test the method, calculation of a simplified tokamak model is carried out. The results are compared with the results from the literature and for the same order of accuracy a reduction of the number of spatial unknowns is shown. (author)

  18. Electronic system of TBR tokamak device

    International Nuclear Information System (INIS)

    The electronics developed as a part of the TBR project, which involves the construction of a small tokamak at the Physics Institute of the University of Sao Paulo, is described. On the basis of tokamak parameter values, the electronics for the toroidal field, ohmic/heating and vertical field systems is presented, including capacitors bank, switches, triggering circuits and power supplies. A controlled power oscilator used in discharge cleaning and pre-ionization is also described. The performance of the system as a function of the desired plasma parameters is discussed. (Author)

  19. Tokamak Engineering Technology Facility scoping study

    Energy Technology Data Exchange (ETDEWEB)

    Stacey, W.M. Jr.; Abdou, M.A.; Bolta, C.C.

    1976-03-01

    A scoping study for a Tokamak Engineering Technology Facility (TETF) is presented. The TETF is a tokamak with R = 3 m and I/sub p/ = 1.4 MA based on the counterstreaming-ion torus mode of operation. The primary purpose of TETF is to demonstrate fusion technologies for the Experimental Power Reactor (EPR), but it will also serve as an engineering and radiation test facility. TETF has several technological systems (e.g., superconducting toroidal-field coil, tritium fuel cycle, impurity control, first wall) that are prototypical of EPR.

  20. Radial electric fields for improved tokamak performance

    International Nuclear Information System (INIS)

    The influence of externally-imposed radial electric fields on the fusion energy output, energy multiplication, and alpha-particle ash build-up in a TFTR-sized, fusing tokamak plasma is explored. In an idealized tokamak plasma, an externally-imposed radial electric field leads to plasma rotation, but no charge current flows across the magnetic fields. However, a realistically-low neutral density profile generates a non-zero cross-field conductivity and the species dependence of this conductivity allows the electric field to selectively alter radial particle transport

  1. Tokamak Spectroscopy for X-Ray Astronomy

    Science.gov (United States)

    Fournier, Kevin B.; Finkenthal, M.; Pacella, D.; May, M. J.; Soukhanovskii, V.; Mattioli, M.; Leigheb, M.; Rice, J. E.

    2000-01-01

    This paper presents the measured x-ray and Extreme Ultraviolet (XUV) spectra of three astrophysically abundant elements (Fe, Ca and Ne) from three different tokamak plasmas. In every case, each spectrum touches on an issue of atomic physics that is important for simulation codes to be used in the analysis of high spectral resolution data from current and future x-ray telescopes. The utility of the tokamak as a laboratory test bed for astrophysical data is demonstrated. Simple models generated with the HULLAC suite of codes demonstrate how the atomic physics issues studied can affect the interpretation of astrophysical data.

  2. Multichannel submillimeter interferometer for tokamak density measurements

    International Nuclear Information System (INIS)

    A two-channel, submillimeter (SMM) laser, electron-density interferometer has been operated successfully on the ISX tokamak. The interferometer is the first phase of a diagnostic system to measure the tokamak plasma current density using the Faraday rotation of the polarization vector of SMM laser beams. Deuterated formic acid lasers (lambda = 0.381 mm) have produced cw power of 10 mW. The interferometer has performed successfully for line-averaged electron densities as high as 8 x 1013 cm-3

  3. Tokamak power systems studies, FY 1985

    International Nuclear Information System (INIS)

    The Tokamak Power System Studies (TPSS) at ANL in FY-1985 were devoted to exploring innovative design concepts which have the potential for making substantial improvements in the tokamak as a commercial power reactor. Major objectives of this work included improved reactor economics, improved environmental and safety features, and the exploration of a wide range of reactor plant outputs with emphasis on reduced plant sizes compared to STARFIRE. The activities concentrated on three areas: plasma engineering, impurity control, and blanket/first wall/shield technology. 205 refs., 125 figs., 107 tabs

  4. A need for non-tokamak approaches to magnetic fusion energy

    International Nuclear Information System (INIS)

    Focusing exclusively on conventional tokamak physics in the quest for commercial fusion power is premature, and the options for both advanced-tokamak and non-tokamak concepts need continued investigation. The basis for this claim is developed, and promising advanced-tokamak and non-tokamak options are suggested

  5. Theoretical description of heavy impurity transport and its application to the modelling of tungsten in JET and ASDEX upgrade

    Czech Academy of Sciences Publication Activity Database

    Casson, F.J.; Angioni, C.; Belli, E.A.; Bilato, R.; Mantica, P.; Odstrčil, T.; Pütterich, T.; Valisa, M.; Garzotti, L.; Giroud, C.; Hobirk, J.; Maggi, C.F.; Mlynář, Jan; Reinke, M.L.

    2015-01-01

    Roč. 57, č. 1 (2015), 014031-014031. ISSN 0741-3335 Institutional support: RVO:61389021 Keywords : tokamak * impurity * transport * neoclassical * validation * modelling * tungsten Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.186, year: 2014 http://iopscience.iop.org/article/10.1088/0741-3335/57/1/014031/ meta

  6. Study and optimization of magnetized ICRF discharges for tokamak wall conditioning and assessment of the applicability to ITER

    International Nuclear Information System (INIS)

    This work is devoted to the study and optimization of the Ion Cyclotron Wall Conditioning (ICWC) technique. ICWC, operated in presence of the toroidal magnetic field, makes use of four main tokamak systems: the ICRF antennas to initiate and sustain the conditioning discharge, the gas injection valves to provide the discharge gas, the machine pumps to remove the wall desorbed particles, and the poloidal magnetic field system to optimize the discharge homogeneity. Additionally neutral gas and plasma diagnostics are required to monitor the discharge and the conditioning efficiency. In chapter 2 a general overview on ICWC is given. Chapter 3 treats the ICRF discharge homogeneity and the confinement properties of the employed magnetic field. In the first part we will discuss experimental facts on plasma homogeneity, and how experimental optimization led to its improvement. In the second part of the chapter the confinement properties of a partially ionized plasma in a toroidal magnetic field configuration with additional small vertical component are discussed. Chapter 4 gives an overview of experimental results on the efficiency of ICWC, obtained on TORE SUPRA, TEXTOR, JET and ASDEX Upgrade. In chapter 5 a 0D kinetic description of hydrogen-helium RF plasmas is outlined. The model, describing the evolution of ICRF plasmas from discharge initiation to the (quasi) steady state plasma stage, is developed to obtain insight on ICRF plasma parameters, particle fluxes to the walls and the main collisional processes. Chapter 6 presents a minimum structure for a 0D reservoir model of the wall to investigate in deeper detail the ICWC plasma wall interaction during isotopic exchange experiments. The hypothesis used to build up the wall model is that the same model structure should be able to describe the wall behavior during normal plasmas and conditioning procedures. Chapter 7 extrapolates the results to the envisaged application of ICWC on ITER

  7. Availability of ITER and Fusion Power Plants considering the experience of the present tokamaks and ITER operational constraints

    International Nuclear Information System (INIS)

    ITER represents a very important step forward in nuclear fusion field from both physics and technology point of view towards Fusion Power Plants (FPPs). Its availability is vital to allow the accomplishment of the planned experimental programmes. On the other side ITER is a first kind of nuclear machine with respect to size and operational and safety issues. Present tokamaks are prototypical as well. Their operating experience has now reached several hundred device-years and gives useful indications for the devices themselves and for those under design or construction. Most of the results are expressed in terms of good plasma pulse number, operating time and delays, effective operation time versus planned operation time. Thanks to the learning process, the availability of present fusion machines, mainly affected by reliability (less from maintainability) has been maintained constant and in some case improved all over the years in spite of the ageing of systems and components and of more demanding and complex plasma scenarios. ITER (and FPPs) will face further challenges, like more significant steady state and transient loads, much higher neutronic flux and fluence, activation and dose rates, that will ask for a periodic replacement of important components with remote handling techniques as well as the control of safety operating limits and conditions (OLCs), like inventories of tritium and dust. The strict quality requirements on components and maintenance equipment for safety and investment protection could compensate partially the reduction of availability. The paper will present in a qualitative way the main factors affecting the availability of ITER, extrapolating also availability data and the lessons learnt from JET, JT-60, DIII-D, ASDEX-U, TORE SUPRA, etc. It will report also reflections about the areas where to concentrate more efforts in order to optimise the availability considering the main constraints present in the design and in the operation

  8. Spontaneous generation of rotation in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Parra Diaz, Felix [Oxford University

    2013-12-24

    Three different aspects of intrinsic rotation have been treated. i) A new, first principles model for intrinsic rotation [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has been implemented in the gyrokinetic code GS2. The results obtained with the code are consistent with several experimental observations, namely the rotation peaking observed after an L-H transition, the rotation reversal observed in Ohmic plasmas, and the change in rotation that follows Lower Hybrid wave injection. ii) The model in [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has several simplifying assumptions that seem to be satisfied in most tokamaks. To check the importance of these hypotheses, first principles equations that do not rely on these simplifying assumptions have been derived, and a version of these new equations has been implemented in GS2 as well. iii) A tokamak cross-section that drives large intrinsic rotation has been proposed for future large tokamaks. In large tokamaks, intrinsic rotation is expected to be very small unless some up-down asymmetry is introduced. The research conducted under this contract indicates that tilted ellipticity is the most efficient way to drive intrinsic rotation.

  9. INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

    International Nuclear Information System (INIS)

    OAK-B135 Advanced tokamaks (AT) are distinguished from conventional tokamaks by their high degree of shaping, achievement of profiles optimized for high confinement and stability characteristics, and active stabilization of MHD instabilities to attain high values of normalized beta and confinement. These high performance fusion devices thus require accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating, as well as simultaneous and well-coordinated MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Satisfying the simultaneous demands on control accuracy, reliability, and performance for all of these subsystems requires a high degree of integration in both design and operation of the plasma control system in an advanced tokamak. The present work describes the approach, benefits, and progress made in integrated plasma control with application examples drawn from the DIII-D tokamak. The approach includes construction of plasma and system response models, validation of models against operating experiments, design of integrated controllers which operate in concert with one another as well as with supervisory modules, simulation of control action against off-line and actual machine control platforms, and iteration of the design-test loop to optimize performance

  10. Microinstabilities in weak density gradient tokamak systems

    International Nuclear Information System (INIS)

    A prominent characteristic of auxiliary-heated tokamak discharges which exhibit improved (''H-mode type'') confinement properties is that their density profiles tend to be much flatter over most of the plasma radius. Depsite this favorable trend, it is emphasized here that, even in the limit of zero density gradient, low-frequency microinstabilities can persist due to the nonzero temperature gradient

  11. High βp bootstrap tokamak reactor

    International Nuclear Information System (INIS)

    Basic characteristics of a steady state tokamak fusion reactor is presented. The minimum required energy multiplication factor Q is found to be 20 to 30 for the feasibility of the fusion reactor. Such a high Q steady state tokamak operation is possible, within our present knowledge of the operational constraints and the current drive physics, when a large fraction of the plasma current is carried by the bootstrap current. Operation at high βp (≥2.0) and high qψ (=4-5) with relatively small εβp (3) and fusion output power (2.5 GW) and is consistent with the present knowledges of the plasma physics of the tokamak, namely the Troyon limit, the energy confinement scalings, the bootstrap current, the current drive efficiency (NB current drive with the total power of 70 MW and the beam energy of 1 MeV) with a favorable aspect on the formation of the cold and dense diverter plasma-condition. From the economical aspect of the tokamak fusion reactor, a more compact reactor is favorable. The use of the high field magnet with Bmax = 16T (for example Ti-doped Nb3Sn conductor) enables to reduce the total machine size to 50% of the above-described conventional design, namely Rp = 7m, Vp = 760m-3, PF = 2.8 GW. (author)

  12. Tokamak fusion test reactor. Final design report

    International Nuclear Information System (INIS)

    Detailed data are given for each of the following areas: (1) system requirements, (2) the tokamak system, (3) electrical power systems, (4) experimental area systems, (5) experimental complex, (6) neutral beam injection system, (7) diagnostic system, and (8) central instrumentation control and data acquisition system

  13. Advanced tokamak concepts and reactor designs

    NARCIS (Netherlands)

    Oomens, A. A. M.

    2000-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described, some examples

  14. Plasma-gun fueling for tokamak reactors

    International Nuclear Information System (INIS)

    In light of the uncertain extrapolation of gas puffing for reactor fueling and certain limitations to pellet injection, the snowplow plasma gun has been studied as a fueling device. Based on current understanding of gun and plasma behavior a design is proposed, and its performance is predicted in a tokamak reactor environment

  15. UCLA Tokamak Program Close Out Report.

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, Robert John [UCLA/retired

    2014-02-04

    The results of UCLA experimental fusion program are summarized. Starting with smaller devices like Microtor, Macrotor, CCT and ending the research on the large (5 m) Electric Tokamak. CCT was the most diagnosed device for H-mode like physics and the effects of rotation induced radial fields. ICRF heating was also studied but plasma heating of University Type Tokamaks did not produce useful results due to plasma edge disturbances of the antennae. The Electric Tokamak produced better confinement in the seconds range. However, it presented very good particle confinement due to an "electric particle pinch". This effect prevented us from reaching a quasi steady state. This particle accumulation effect was numerically explained by Shaing's enhanced neoclassical theory. The PI believes that ITER will have a good energy confinement time but deleteriously large particle confinement time and it will disrupt on particle pinching at nominal average densities. The US fusion research program did not study particle transport effects due to its undue focus on the physics of energy confinement time. Energy confinement time is not an issue for energy producing tokamaks. Controlling the ash flow will be very expensive.

  16. Toroidal Alfven wave stability in ignited tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, C.Z.; Fu, G.Y.; Van Dam, J.W.

    1989-01-01

    The effects of fusion-product alpha particles on the stability of global-type shear Alfven waves in an ignited tokamak plasma are investigated in toroidal geometry. Finite toroidicity can lead to stabilization of the global Alfven eigenmodes, but it induces a new global shear Alfven eigenmodes, which is strongly destabilized via transit resonance with alpha particles. 8 refs., 2 figs.

  17. Radioactivity evaluation for the KSTAR tokamak

    International Nuclear Information System (INIS)

    The deuterium-deuterium (D-D) reaction in the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak generates neutrons with a peak yield of 2.5 x 1016 s-1 through a pulse operation of 300 s. Since the structure material of the tokamak is irradiated with neutrons, this environment will restrict work around and inside the tokamak from a radiation protection physics point of view after shutdown. Identification of neutron-produced radionuclides and evaluation of absorbed dose in the structure material are needed to develop a guiding principle for radiation protection. The activation level was evaluated by MCNP4C2 and an inventory code, FISPACT. The absorbed dose in the working area decreased by 4.26 x 10-4 mrem h-1 in the inner vessel 1.5 d after shutdown. Furthermore, tritium strongly contributes to the contamination in the graphite tile. The amount of tritium produced by neutrons was 3.03 x 106 Bq kg-1 in the carbon graphite of a plasma-facing wall. (authors)

  18. Compact tokamak reactors. Part 1 (analytic results)

    International Nuclear Information System (INIS)

    We discuss the possible use of tokamaks for thermonuclear power plants, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First we review and summarize the existing literature. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamak power plant, by including the power required to drive the toroidal field, and considering two extremes of plasma current drive efficiency. The analytic results will be augmented by a numerical calculation which permits arbitrary plasma current drive efficiency; the results of which will be presented in Part II. Third, a scaling from any given reference reactor design to a copper toroidal field coil device is discussed. Throughout the paper the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculating electric power. We conclude that the latest published reactor studies, which show little advantage in using low aspect ratio unless remarkably high efficiency plasma current drive and low safety factor are combined, can be reproduced with the analytic model

  19. Analysis of sawtooth relaxation oscillations in tokamaks

    International Nuclear Information System (INIS)

    Sawtooth relaxation oscillations are analyzed using the Kadomtsev's disruption model and a thermal relaxation model. The sawtooth period is found to be very sensitive to the thermal conduction loss. Qualitative agreement between these calculations and the sawtooth period observed in several tokamaks is demonstrated

  20. Material erosion and migration in tokamaks

    International Nuclear Information System (INIS)

    The issue of first wall and divertor target lifetime represents one of the greatest challenges facing the successful demonstration of integrated tokamak burning plasma operation, even in the case of the planned next step device, ITER, which will run at a relatively low duty cycle in comparison to future fusion power plants. Material erosion by continuous or transient plasma ion and neutral impact, the subsequent transport of the released impurities through and by the plasma and their deposition and/or eventual re-erosion constitute the process of migration. Its importance is now recognized by a concerted research effort throughout the international tokamak community, comprising a wide variety of devices with differing plasma configurations, sizes and plasma-facing component material. No single device, however, operates with the first wall material mix currently envisaged for ITER, and all are far from the ITER energy throughput and divertor particle fluxes and fluences. This paper aims to review the basic components of material erosion and migration in tokamaks, illustrating each by way of examples from current research and attempting to place them in the context of the next step device. Plans for testing an ITER-like first wall material mix on the JET tokamak will also be briefly outlined