WorldWideScience

Sample records for asdex upgrade tokamak

  1. 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

  2. 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.)

  3. 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.

  4. 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

  5. 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.

  6. 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.

  7. 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

  8. 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.

  9. 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.

  10. 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...

  11. 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)

  12. 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

  13. 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.

  14. 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

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

    Science.gov (United States)

    Nowak, S.; Lazzaro, E.; Esposito, B.; Granucci, G.; Maraschek, M.; Sauter, O.; Zohm, H.; Brunetti, D.; ASDEX Upgrade Team

    2012-09-01

    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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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

  1. 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

  2. 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

  3. 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...

  4. 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)

  5. 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

  6. 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.

  7. 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

  8. 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.)

  9. 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.

  10. 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.

  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...... is neoclassical, explaining the strong inward pinch of high-Z impurities in between ELMs. In improved H-mode, the width of the temperature pedestal increases with heating power, consistent with a scaling. In the area of MHD instabilities, disruption mitigation experiments using massive Ne injection reach volume...... for NTMs, TAEs and also beta-induced Alfven eigenmodes (BAEs). Specific studies addressing the first ITER operational phase show that O1 ECRH at the HFS assists reliable low-voltage breakdown. During ramp-up, additional heating can be used to vary li to fit within the ITER range. Confinement and power...

  12. Growth of axisymmetric instabilities in ASDEX upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Sehmer, Till; Lackner, Karl; Strumberger, Erika; Fable, Emiliano; Kardaun, Otto [Max-Planck-Institut fuer Plasma-Physik, EURATOM Association Boltzmannstrasse 2, 85748 Garching (Germany); McCarthy, Patrick [University College Cork (Ireland)

    2014-07-01

    Modern poloidal divertor tokamaks, such as ASDEX upgrade (AUG), produce elongated plasmas, which are unstable against vertical displacement. The growth rate of this 2D instability in the presence of stabilizing passive conductors (PSL) with finite resistivity was calculated for 5416 AUG typical equilibria. For this, a general ideal MHD code package (NEMEC, CAS3DN, STARWALL) was used, which is able to take into account also the 3D structure of the PSL. The comparison of the resulting growth rates with the previously used rigid displacement model (movement only in z-direction, no skin effect for PSL considered, no induced surface currents) shows that the latter simplified model gives a consistently lower limit for typical AUG parameters (elongation, triangularity, current profile and axis position in radial direction). A statistical analysis of the results of the rigid displacement model shows the expected dependencies except for the triangularity, which has a stabilizing effect in this model. Based on results of our present, more general model, we conclude that a rigid displacement model gives an over-optimistic result regarding the effect of triangularity, in line with the experimental observation on AUG of an increasing discrepancy between previously predicted and observed growth rates for strongly triangular plasmas.

  13. 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.

  14. 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...

  15. Status of the new multi-frequency ECRH system for ASDEX Upgrade

    DEFF Research Database (Denmark)

    Wagner, D.; Grünwald, G.; Leuterer, F.;

    2008-01-01

    Currently, a new multi-frequency ECRH system is under construction at the ASDEX Upgrade tokamak experiment. This system employs, for the first time in a fusion device, multi-frequency gyrotrons, step-tunable in the range 105-140 GHz. The first two gyrotrons, working at 105 and 140 GHz, were insta...

  16. 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)

  17. 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.

  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. The enhanced ASDEX Upgrade pellet centrifuge launcher

    Science.gov (United States)

    Plöckl, B.; Lang, P. T.

    2013-10-01

    Pellets played an important role in the program of ASDEX Upgrade serving both for investigations on efficient particle fuelling and high density scenarios but also for pioneering work on Edge Localised Mode (ELM) pacing and mitigation. Initially designed for launching fuelling pellets from the magnetic low field side, the system was converted already some time ago to inject pellets from the magnetic high field side as much higher fuelling efficiency was found using this configuration. In operation for more than 20 years, the pellet launching system had to undergo a major revision and upgrading, in particular of its control system. Furthermore, the control system installed adjacent to the launcher had to be transferred to a more distant location enforcing a complete galvanic separation from torus potential and a fully remote control solution. Changing from a hybrid system consisting of PLC S5/S7 and some hard wired relay control to a state of the art PLC system allowed the introduction of several new operational options enabling more flexibility in the pellet experiments. This article describes the new system architecture of control hardware and software, the operating procedure, and the extended operational window. First successful applications for ELM pacing and triggering studies are presented as well as utilization for the development of high density scenarios.

  1. 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

  2. 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.

  3. 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 profiles...... during on-axis injection and outwards shifted profiles during off-axis injection. Due to this change of the fast-ion population, a clear modification of the plasma current profile is predicted but not observed by a motional Stark effect diagnostic. The fast-ion transport caused by MHD activity has been...

  4. A fast feedback controlled magnetic drive for the ASDEX Upgrade fast-ion loss detectors

    Science.gov (United States)

    Ayllon-Guerola, J.; Gonzalez-Martin, J.; Garcia-Munoz, M.; Rivero-Rodriguez, J.; Herrmann, A.; Vorbrugg, S.; Leitenstern, P.; Zoletnik, S.; Galdon, J.; Garcia Lopez, J.; Rodriguez-Ramos, M.; Sanchis-Sanchez, L.; Dominguez, A. D.; Kocan, M.; Gunn, J. P.; Garcia-Vallejo, D.; Dominguez, J.

    2016-11-01

    A magnetically driven fast-ion loss detector system for the ASDEX Upgrade tokamak has been designed and will be presented here. The device is feedback controlled to adapt the detector head position to the heat load and physics requirements. Dynamic simulations have been performed taking into account effects such as friction, coil self-induction, and eddy currents. A real time positioning control algorithm to maximize the detector operational window has been developed. This algorithm considers dynamical behavior and mechanical resistance as well as measured and predicted thermal loads. The mechanical design and real time predictive algorithm presented here may be used for other reciprocating systems.

  5. 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...

  6. 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.

  7. 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.

  8. Development of the Q  =  10 scenario for ITER on ASDEX Upgrade (AUG)

    Science.gov (United States)

    Schweinzer, J.; Beurskens, M.; Frassinetti, L.; Joffrin, E.; Bobkov, V.; Dux, R.; Fischer, R.; Fuchs, C.; Kallenbach, A.; Hopf, C.; Lang, P. T.; Mlynek, A.; Pütterich, T.; Ryter, F.; Stober, J.; Tardini, G.; Wolfrum, E.; Zohm, H.; the EUROfusion MST1 Team; the ASDEX Upgrade Team

    2016-10-01

    The development of the baseline H-mode scenario foreseen for ITER on the ASDEX Upgrade tokamak, i.e. discharges at q 95  =  3, relatively low β N ~ 1.8, high normalized density n/n GW ~ 0.85 and high triangularity δ  =  0.4, focused on the integration of elements foreseen for ITER and available on ASDEX Upgrade, such as ELM mitigation techniques and impurity seeding in combination with a metallic wall. Values for density and energy confinement simultaneously came close to the requirements of the ITER baseline scenario as long as β N stayed above 2. At lower heating power and thus lower β N normalized energy confinement H 98y2 ~ 0.85 is obtained. It has been found that stationary discharges are not easily achieved under these conditions due to the low natural ELM frequency occurring at the low q 95/high δ operational point. Up until now the ELM parameters were uncontrollable with the tools developed in other scenarios. Therefore studies on an alternative operational point at higher β N and q 95 have been conducted. In order to prepare for the ITER first non-activation operational phase, Helium operation has been investigated as well.

  9. 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)

  10. 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...

  11. 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.

  12. 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.

  13. 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

  14. A new beam emission polarimetry diagnostic for measuring the magnetic field line angle at the plasma edge of ASDEX Upgrade

    Science.gov (United States)

    Viezzer, E.; Dux, R.; Dunne, M. G.

    2016-11-01

    A new edge beam emission polarimetry diagnostic dedicated to the measurement of the magnetic field line angle has been installed on the ASDEX Upgrade tokamak. The new diagnostic relies on the motional Stark effect and is based on the simultaneous measurement of the polarization direction of the linearly polarized π (parallel to the electric field) and σ (perpendicular to the electric field) lines of the Balmer line Dα. The technical properties of the system are described. The calibration procedures are discussed and first measurements are presented.

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

    DEFF Research Database (Denmark)

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

    2015-01-01

    and wave number show that an increase of R/LTe introduced by off-axis electron cyclotron resonance heating (ECRH) mainly increases the large scale fluctuations. The radial variation of the fluctuation level is in agreement with simulations using the GENE code. Fast particles are shown to undergo classical...... slowing down in the absence of large scale magnetohydrodynamic (MHD) events and for low heating power, but show signs of anomalous radial redistribution at large heating power, consistent with a broadened off-axis neutral beam current drive current profile under these conditions. Neoclassical tearing mode...... 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...

  16. 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.

  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. 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)

  19. 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.)

  20. 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.

  1. 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.

  2. 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.

  3. 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 ...

  4. 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

  5. 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.

  6. 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.)

  7. 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.)

  8. 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.

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

    Science.gov (United States)

    Křivská, A.; Bobkov, V.; Colas, L.; Jacquot, J.; Milanesio, D.; Ochoukov, R.

    2015-12-01

    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.

  10. 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.

  11. 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

  12. 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;

    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 sce...

  13. 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

  14. 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...

  15. 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 di...

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

    DEFF Research Database (Denmark)

    Michelsen, S.; Korsholm, Søren Bang; Bindslev, H.;

    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...

  17. Gaseous electron multiplier-based soft x-ray plasma diagnostics development: Preliminary tests at ASDEX Upgrade

    Science.gov (United States)

    Chernyshova, M.; Malinowski, K.; Czarski, T.; Wojeński, A.; Vezinet, D.; Poźniak, K. T.; Kasprowicz, G.; Mazon, D.; Jardin, A.; Herrmann, A.; Kowalska-Strzeciwilk, E.; Krawczyk, R.; Kolasiński, P.; Zabołotny, W.; Zienkiewicz, P.

    2016-11-01

    A Gaseous Electron Multiplier (GEM)-based detector is being developed for soft X-ray diagnostics on tokamaks. Its main goal is to facilitate transport studies of impurities like tungsten. Such studies are very relevant to ITER, where the excessive accumulation of impurities in the plasma core should be avoided. This contribution provides details of the preliminary tests at ASDEX Upgrade (AUG) with a focus on the most important aspects for detector operation in harsh radiation environment. It was shown that both spatially and spectrally resolved data could be collected, in a reasonable agreement with other AUG diagnostics. Contributions to the GEM signal include also hard X-rays, gammas, and neutrons. First simulations of the effect of high-energy photons have helped understanding these contributions.

  18. 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.

  19. 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.)

  20. 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

  1. 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

  2. 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.)

  3. 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

  4. 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.

  5. 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....

  6. 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...... 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....

  7. 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)....

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

    Energy Technology Data Exchange (ETDEWEB)

    Treutterer, Wolfgang, E-mail: Wolfgang.Treutterer@ipp.mpg.de [Max-Planck Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany); Neu, Gregor; Raupp, Gerhard; Zasche, Dieter; Zehetbauer, Thomas [Max-Planck Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany); Cole, Richard; Lueddecke, Klaus [Unlimited Computer Systems, Iffeldorf (Germany)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer Control system architectures with data-driven workflows are efficient, flexible and maintainable. Black-Right-Pointing-Pointer Signal groups provide coherence of interrelated signals and increase the efficiency of process synchronisation. Black-Right-Pointing-Pointer Sample tags indicating sample quality form the fundament of a local event handling strategy. Black-Right-Pointing-Pointer 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

  9. 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

  10. 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.

  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. 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.

  13. 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.)

  14. 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;

    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...... of the diagnostic. It then describes the commissioning activities carried out to date. These activities include gyrotron studies, transmission line alignment, and beam pattern measurements in the vacuum vessel. Overlap experiments in near perpendicular and near parallel have confirmed the successful alignment...

  15. 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...

  16. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Mlynek, Alexander

    2010-07-20

    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 2{pi} 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

  18. 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

  19. 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.

  20. 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.

  1. ICRF antennas optimized for operation with a high-Z metallic wall in ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Zammuto, I. [Euratom Association, Max Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2, D-85748 Garching (Germany)], E-mail: i.zammuto@libero.it; Krivska, A. [Euratom Association, Institute of Plasma Physics ASCR, Prague (Czech Republic); Telecommunication Engineering Department, Czech Technical University of Prague, Prague (Czech Republic); Bobkov, V.; Braun, F.; Bilato, R. [Euratom Association, Max Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2, D-85748 Garching (Germany); Noterdaeme, J.-M. [Euratom Association, Max Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2, D-85748 Garching (Germany); EESA Department, University of Gent, Gent (Belgium)

    2009-06-15

    High-Z materials are considered the best candidates for plasma facing components (PFCs) in future fusion devices, while ion cyclotron range of frequencies (ICRF) heating is a method of choice because of its flexibility, cost effectiveness, and plug-to-power efficiency. But high-Z materials and ICRF have been a difficult combination. ASDEX Upgrade (AUG) is pioneering the use of a tungsten (W) wall and in the next years it will prove (or disprove) the appropriateness of this combination for the future. This paper describes the optimization procedure for and the main features of a new proposed ICRF antenna for AUG specifically designed to improve the compatibility of ICRF with a high-Z metallic wall. The optimization criteria for a new design are based on a reduction of the parasitic parallel electric field responsible of the impurity production. The evolutionary design of the 2-strap antenna is compared with a redesigned 4-strap antenna embedded in the wall. Moreover, options and constraints on how to integrate a wider antenna in the present configuration of ASDEX Upgrade and its integration on the stabilizing wall are discussed.

  2. 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.

  3. On the determination of the poloidal velocity and the shear layer in the SOL of ASDEX Upgrade

    DEFF Research Database (Denmark)

    Mehlmann, F.; Costea, S.; Naulin, Volker;

    A reciprocating probe with six pins was used for localized measurements in the scrape-off layer (SOL) of ASDEX Upgrade (AUG) up to the shear layer (SL) and a few mm inside it. The probe was used to determine the poloidal velocity with three different methods which are critically compared to each...

  4. Measurements and modeling of Alfven eigenmode induced fast ion transport and loss in DIII-D and ASDEX Upgrade

    NARCIS (Netherlands)

    VanZeeland, M. A.; Heidbrink, W. W.; Fisher, R. K.; Munoz, M. G.; Kramer, G. J.; Pace, D. C.; White, R. B.; Akaslompolo, S.; Austin, M. E.; Boom, J. E.; Classen, I.G.J.; da Graca, S.; Geiger, B.; Gorelenkova, M.; Gorelenkov, N. N.; Hyatt, A. W.; Luhmann, N.; Maraschek, M.; McKee, G. R.; Moyer, R. A.; Muscatello, C. M.; Nazikian, R.; Park, H.; Sharapov, S.; Suttrop, W.; Tardini, G.; Tobias, B. J.; Zhu, Y. B.

    2011-01-01

    Neutral beam injection into reversed magnetic shear DIII-D and ASDEX Upgrade plasmas produces a variety of Alfvenic activity including toroidicity-induced Alfven eigenmodes and reversed shear Alfven eigenmodes (RSAEs). These modes are studied during the discharge current ramp phase when incomplete c

  5. 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

  6. 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

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

    DEFF Research Database (Denmark)

    Bin, W.; Bruschi, A.; D'Arcangelo, O.;

    2015-01-01

    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...... 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...

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

    DEFF Research Database (Denmark)

    Bin, W.; Bruschi, A.; D'Arcangelo, O.;

    2015-01-01

    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......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...

  9. 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.;

    2011-01-01

    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......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....... These results show that ECCD can be used as a tool for control of sawteeth also in the presence of fast particles....

  10. 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...

  11. 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.

  12. 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.

  13. Gyrokinetic studies of core turbulence features in ASDEX Upgrade H-mode plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, A. Bañón, E-mail: banon@physics.ucla.edu; Told, D. [Max-Planck-Institut für Plasmaphysik, Boltzmannstrase 2, 85748 Garching (Germany); Department of Physics and Astronomy, University of California, Los Angeles, California 90095 (United States); Happel, T.; Görler, T.; Abiteboul, J.; Bustos, A.; Doerk, H. [Max-Planck-Institut für Plasmaphysik, Boltzmannstrase 2, 85748 Garching (Germany); Jenko, F. [Max-Planck-Institut für Plasmaphysik, Boltzmannstrase 2, 85748 Garching (Germany); Department of Physics and Astronomy, University of California, Los Angeles, California 90095 (United States); Max-Planck/Princeton Center for Plasma Physics (United States)

    2015-04-15

    Gyrokinetic validation studies are crucial for developing confidence in the model incorporated in numerical simulations and thus improving their predictive capabilities. As one step in this direction, we simulate an ASDEX Upgrade discharge with the GENE code, and analyze various fluctuating quantities and compare them to experimental measurements. The approach taken is the following. First, linear simulations are performed in order to determine the turbulence regime. Second, the heat fluxes in nonlinear simulations are matched to experimental fluxes by varying the logarithmic ion temperature gradient within the expected experimental error bars. Finally, the dependence of various quantities with respect to the ion temperature gradient is analyzed in detail. It is found that density and temperature fluctuations can vary significantly with small changes in this parameter, thus making comparisons with experiments very sensitive to uncertainties in the experimental profiles. However, cross-phases are more robust, indicating that they are better observables for comparisons between gyrokinetic simulations and experimental measurements.

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

    Energy Technology Data Exchange (ETDEWEB)

    Igochine, V; Bobkov, V; Guenter, S; Maraschek, M; Pereversev, G; Reich, M; Stober, J [MPI fuer Plasmaphysik, Euratom-Association, D-85748 Garching (Germany); Chapman, I T [EURATOM/CCFE Association, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Moseev, D, E-mail: valentin.igochine@ipp.mpg.de [EURATOM-Risoe DTU, Risoe National Laboratory for Sustainable Energy, Technical University of Denmark, DK-4000 Roskilde (Denmark)

    2011-02-15

    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. These results show that ECCD can be used as a tool for control of sawteeth also in the presence of fast particles. (brief communication)

  15. 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.

  16. 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)

  17. 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.

  18. 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.

  19. 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

  20. 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.

  1. Fast ion measurements by collective Thomson scattering in TEXTOR and ASDEX Upgrade and proposal for the ITER CTS system

    DEFF Research Database (Denmark)

    Korsholm, Søren Bang; Bindslev, Henrik; Furtula, Vedran;

    ) provides the possibility of revealing the velocity distribution of the confined fast ions along a given direction – resolved both in time and space. Recently, the ITER baseline design has been expanded to include the enabling of the front end of a fast ion CTS diagnostic system resolving dynamics...... perpendicular to the magnetic field. The feasibility study and conceptual design of this diagnostic was provided by the CTS group at Risø DTU. The development of the ITER CTS diagnostic builds on the experiences and expertise gained from the construction and current operation of the CTS diagnostic systems...... on TEXTOR and ASDEX Upgrade. This contribution will briefly introduce the technique of CTS, give an overview of the results of the current diagnostic systems at TEXTOR and ASDEX Upgrade, and present the chosen solution and the status of the design of the ITER CTS diagnostic system....

  2. 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.

  3. 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

  4. 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

  5. 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.

  6. 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

  7. 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...

  8. Electric probe measurements of the poloidal velocity in the scrape-off layer of ASDEX upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Mehlmann, F.; Costea, S.; Schrittwieser, R.; Lux, C.; Ionita, C. [Institute for Ion Physics and Applied Physics, University of Innsbruck, Association EURATOM/OeAW (Austria); Naulin, V.; Rasmussen, J.J.; Nielsen, A.H. [Association EURATOM-DTU, Dept. of Physics, Technical University of Denmark, Lyngby (Denmark); Mueller, H.W.; Carralero, D.; Rohde, V. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany); Vianello, N. [Consorzio RFX, Associazione Euratom-ENEA sulla Fusione, Padova (Italy); Collaboration: ASDEX Upgrade Team

    2014-04-15

    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 vθ 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{sub r} measured by two radially staggered probe pins, with vθ being due to the E{sub r} x B{sub φ}-drift (B{sub φ} 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 correlation were used to determine the poloidal velocity and its jump, yielding comparable results to the first method. Also the method of conditional averaging was applied to the latter signals. (copyright 2014 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

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

    Science.gov (United States)

    Schneider, P A; Blank, H; Geiger, B; Mank, K; Martinov, S; Ryter, F; Weiland, M; Weller, A

    2015-07-01

    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.

  10. 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.

  11. 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.

  12. 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.

  13. 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)

  14. 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.

  15. 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

  16. 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.)

  17. 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.

  18. 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

  19. 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

  20. 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.

  1. 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.)

  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. 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

  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. High power ECRH and ECCD in moderately collisional ASDEX Upgrade Hmodes and status of EC system upgrade

    Directory of Open Access Journals (Sweden)

    Stober J.

    2015-01-01

    Full Text Available This contribution deals with H-modes with significant heat exchange between electrons and ions, but which can still show large differences between electron and ion-temperatures especially inside half minor radius. These conditions are referred to as moderately collisional. A systematic study shows that an increasing fraction of electron heating increases the transport in the ion channel mainly due to the dependence of the ITG dominated ion transport on the ratio Te/Ti in agreement with modeling. The rotational shear in the plasmas under study was so small that it hardly influences ITG stability, such that variations of the rotation profile due to a change of the heating method were of minor importance. These findings connect to studies of advanced tokamak scenarios using ECCD as a tool to modify the q-profile. The electron heating connected to the ECCD tends to increase the transport in the ion channel quite in contrast to the goal to operate at reduced current but with increased confinement. The confinement only increases as the fraction of ion heating is increased by adding more NBI. An ITER case was modeled as well. Due to the larger value of νei ・ τE the ratio Te/Ti is only moderately reduced even with strong electron heating and the confinement reduction is small even for the hypothetic case of using only ECRH as additional heating. Finally the paper discusses the ongoing upgrade of the AUG ECRH-system.

  6. 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.

  7. On the role of the edge density profile for the L-H transition power threshold in ASDEX Upgrade

    Science.gov (United States)

    Shao, L. M.; Wolfrum, E.; Ryter, F.; Birkenmeier, G.; Laggner, F. M.; Viezzer, E.; Fischer, R.; Willensdorfer, M.; Kurzan, B.; Lunt, T.; the ASDEX Upgrade Team

    2016-02-01

    The L-H transition power threshold ({{P}\\text{L-\\text{H}}} ) in full tungsten (W) wall discharges is lower by 25% compared to those with graphite (C) mix tungsten walls in ASDEX Upgrade (Ryter et al 2013 Nucl. Fusion 53 113003). The lower power threshold in the full tungsten wall discharges has been found to correlate with higher edge density as well as steeper edge density gradient. An estimate of the minimum in the neoclassical radial electric field well inside the separatrix yields a constant value for all analyzed L-H transitions at fixed toroidal magnetic field ({{B}\\text{T}} ). The decrease of the threshold power is explained by the steeper edge density gradient in the discharges with full tungsten wall.

  8. 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.

  9. 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.)

  10. 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...

  11. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Giannone, L., E-mail: Louis.Giannone@ipp.mpg.d [Max-Planck-Institut fuer Plasmaphysik, EURATOM-IPP Association, D-85748 Garching (Germany); Cerna, M. [National Instruments, Austin, TX 78759-3504 (United States); Cole, R.; Fitzek, M. [Unlimited Computer Systems GmbH, 82393 Iffeldorf (Germany); Kallenbach, A. [Max-Planck-Institut fuer Plasmaphysik, EURATOM-IPP Association, D-85748 Garching (Germany); Lueddecke, K. [Unlimited Computer Systems GmbH, 82393 Iffeldorf (Germany); McCarthy, P.J. [Department of Physics, University College Cork, Association EURATOM-DCU, Cork (Ireland); Scarabosio, A.; Schneider, W.; Sips, A.C.C.; Treutterer, W. [Max-Planck-Institut fuer Plasmaphysik, EURATOM-IPP Association, D-85748 Garching (Germany); Vrancic, A.; Wenzel, L.; Yi, H. [National Instruments, Austin, TX 78759-3504 (United States); Behler, K.; Eich, T.; Eixenberger, H.; Fuchs, J.C.; Haas, G.; Lexa, G. [Max-Planck-Institut fuer Plasmaphysik, EURATOM-IPP Association, D-85748 Garching (Germany)

    2010-07-15

    The existing VxWorks real-time system for the position and shape control in ASDEX Upgrade has been extended to calculate magnetic flux surfaces in real-time using a multi-core PCI Express system running LabVIEW RT 8.6. real-time signal processing of bolometers and manometers is performed with the on-board FPGA to calculate the measured radiated power flux and particle flux respectively from the raw data. Radiation feedback experiments use halo current measurements from the outer divertor with real-time median filter pre-processing to remove the excursions produced by ELMs. Integration of these plasma diagnostics into the control system by the exchange of XML sheets for communicating the real-time variables to be produced and consumed is in operation. Reflective memory and UDP are employed by the LabVIEW RT plasma diagnostics to communicate with the control system and other plasma diagnostics in a multi-platform real-time network.

  13. Deployment and future prospects of high performance diagnostics featuring serial I/O (SIO) data acquisition (DAQ) at ASDEX Upgrade

    International Nuclear Information System (INIS)

    Highlights: ► The high sustained data rates transferring measured data from periphery into memory of computers. ► The achieved low latency in real-time interrupt handling under Solaris 10. ► The new prototype of an even more powerful 2nd generation SIO II device. ► The fusion of all blocks of board logic (serializer, FIFO, TDC, merge engine, PCIe controller) into one single FPGA simplifying the boards physical layout significantly. - Abstract: The SIO DAQ concept used at the ASDEX Upgrade fusion experiment features data acquisition from a modular front-end (a modular crate-and-interface-cards concept for analog and digital input and output) over standardized serial lines and via a serial input/output computer interface card (the SIO card) in real-time directly into the main memory of a host computer. Deployment of a series of diagnostics using SIO led to various solutions and configurations for the different requirements. Experience has been gained and lessons learned applying the SIO concept at its technical limits. Requirements for a further development of the SIO concept have been identified, and a performance improvement by a factor of 4–8 beyond its current limits seems achievable. An effort has been started to develop a SIO version 2 (SIO II) featuring upgraded serial links and a more powerful FPGA for merging and forwarding data streams to host computer memory. (Compatibility with the existing SIO (SIO I) front-end system has to be maintained.) This paper presents results achieved and experiences gained in the deployment of SIO I, the status of SIO II development (currently in the prototype phase), and projected enhancements and updates to existing implementations.

  14. 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)

  15. 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)

  16. Optical emission measurements of H 2 and D 2 molecules in the divertor region of ASDEX Upgrade

    Science.gov (United States)

    Fantz, U.; Behringer, K.; Gafert, J.; Coster, D.; ASDEX Upgrade Team

    A spectroscopic method has been developed for measuring molecular influxes and particle densities in fusion edge plasmas, which is based on the H 2 and D 2 Fulcher emission bands around 600 nm wavelength. A first application to the ASDEX Upgrade divertor plasma is described. The influx of hydrogen molecules was determined from the population of the upper Fulcher state using the theoretical number of ionization and dissociation events per Fulcher photon ( Seff + Deff)/XB Ful, as calculated by a collisional-radiative model. These results were compared with expectations on the basis of the atomic hydrogen fluxes and a typical molecule/atom ratio. Measurements and calculations agree in their time dependence, but the experimental values are somewhat lower, which may be within the error margin or of more significance. The Fulcher radiation was also compared directly to B2-EIRENE predictions, resulting in a higher discrepancy. In addition, the vibrational population of the ground state molecules was determined from that of the excited state using a method based on Franck-Condon factors. It can be characterized by a Tvib between 3000 and 9000 K, inversely correlated with electron temperature. This variation is predicted by the collisional-radiative code and even allows an estimate of Te. Vibrational excitation increases ionization and dissociation rate coefficients, as clearly demonstrated by the code calculations. It is therefore very likely that the observed discrepancy in molecular intensity is mainly caused by the omission of vibrational excitation in the present version of B2-EIRENE. The described flux measurements are expected to be accurate above Te=5 eV, but are more difficult at lower temperatures due to the strong Te dependence of ( Seff + Deff)/XB Ful in that region.

  17. Upgraded data service system for HT-7 Tokamak

    Institute of Scientific and Technical Information of China (English)

    QU Lian-Zheng; LUO Jia-Rong; WEI Pei-Jie; LI Gui-Ming; CHENG Ting; QI Na

    2005-01-01

    A data service system plays an indispensable role in HT-7 Tokamak experiment. Since the former system doesn't provide the function of timely data procession and analysis, and all client software are based on Windows, it can't fulfill virtual fusion laboratory for remote researchers. Therefore, a new system which is simplified by three kinds of data servers and one data analysis and visualization software tool has been developed. The data servers include a data acquisition server based on file system, an MDSplus server used as the central repository for analysis data, and a web server. Users who prefer the convenience of application that can be run in a Web Browser can easily access the experiment data without knowing X-Windows. In order to adjust instruments to control experiment the operators need to plot data duly as soon as they are gathered. To satisfy their requirement, an upgraded data analysis and visualization software GT-7 is developed. It not only makes 2D data visualization more efficient, but also it can be capable of processing, analyzing and displaying interactive 2D and 3D graph of raw, analyzed data by the format of ASCII, LZO and MDSplus.

  18. A new thermal He-beam diagnostic for electron density and temperature measurements in the scrape-off layer of ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Griener, Michael; Wolfrum, Elisabeth; Eich, Thomas; Herrmann, Albrecht; Rohde, Volker [Max Planck Institute for Plasma Physics, Garching (Germany); Schmitz, Oliver [Engineering Physics Department, University of Wisconsin-Madison (United States); Stroth, Ulrich [Max Planck Institute for Plasma Physics, Garching (Germany); Physik Department E28, Technische Universitaet Muenchen, Garching (Germany); Collaboration: the ASDEX Upgrade Team

    2015-05-01

    In a nuclear fusion device power is exhausted across the last closed flux surface into the so-called 'scrape-off layer', SOL. In order to study the transport dynamics to (a) the divertor via parallel heat flux and (b) to the wall via filaments, a diagnostic for the determination of n{sub e} and T{sub e} with high spatial and temporal resolution is required. Although the diagnostic capabilities of the ASDEX Upgrade edge plasma are excellent, there is a lack of spatially and temporally highly resolved electron temperature measurements in the SOL. Therefore a piezo valve will be installed in ASDEX Upgrade in April 2015. It allows fast chopping of a thermal He-beam which is part of the new diagnostic. In the first campaign, existing lines of sight of the CXRS diagnostic will be used to measure various He I transitions to confirm the collisional radiative model for He. The principle of the thermal He-diagnostic as well as calculations of the achievable spatial resolution of the initial set-up are presented.

  19. Convective and Diffusive Energetic Particle Losses Induced by Shear Alfven Waves in the ASDEX Upgrade Tokamak

    NARCIS (Netherlands)

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

    2010-01-01

    We present here the first phase-space characterization of convective and diffusive energetic particle losses induced by shear Alfven waves in a magnetically confined fusion plasma. While single toroidal Alfven eigenmodes (TAE) and Alfven cascades (AC) eject resonant fast ions in a convective process

  20. 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

    unstable by fast ions from ICRH as well as NBI origin. In ICRF heated plasmas, diffusive and convective fast-ion losses induced by AEs have been characterized in fast-ion phase space. While single RSAEs and TAEs eject resonant fast ions in a convective process directly proportional to the fluctuation...... amplitude, δB/B, the overlapping of multiple RSAE and TAE spatial structures and wave–particle resonances leads to a large diffusive loss, scaling as (δB/B)2. In beam heated discharges, coherent fast-ion losses have been observed primarily due to TAEs. Core localized, low amplitude NBI driven RSAEs have...

  1. Toroidal modelling of RMP response in ASDEX Upgrade: coil phase scan, q 95 dependence, and toroidal torques

    Science.gov (United States)

    Liu, Yueqiang; Ryan, D.; Kirk, A.; Li, Li; Suttrop, W.; Dunne, M.; Fischer, R.; Fuchs, J. C.; Kurzan, B.; Piovesan, P.; Willensdorfer, M.; the ASDEX Upgrade Team; the EUROfusion MST1 Team

    2016-05-01

    The plasma response to the vacuum resonant magnetic perturbation (RMP) fields, produced by the ELM control coils in ASDEX Upgrade experiments, is computationally modelled using the MARS-F/K codes (Liu et al 2000 Phys. Plasmas 7 3681, Liu et al 2008 Phys. Plasmas 15 112503). A systematic investigation is carried out, considering various plasma and coil configurations as in the ELM control experiments. The low q plasmas, with {{q}95}˜ 3.8 (q 95 is the safety factor q value at 95% of the equilibrium poloidal flux), responding to low n (n is the toroidal mode number) field perturbations from each single row of the ELM coils, generates a core kink amplification effect. Combining two rows, with different toroidal phasing, thus leads to either cancellation or reinforcement of the core kink response, which in turn determines the poloidal location of the peak plasma surface displacement. The core kink response is typically weak for the n  =  4 coil configuration at low q, and for the n  =  2 configuration but only at high q ({{q}95}˜ 5.5 ). A phase shift of around 60 degrees for low q plasmas, and around 90 degrees for high q plasmas, is found in the coil phasing, between the plasma response field and the vacuum RMP field, that maximizes the edge resonant field component. This leads to an optimal coil phasing of about 100 (-100) degrees for low (high) q plasmas, that maximizes both the edge resonant field component and the plasma surface displacement near the X-point of the separatrix. This optimal phasing closely corresponds to the best ELM mitigation observed in experiments. A strong parallel sound wave damping moderately reduces the core kink response but has minor effect on the edge peeling response. For low q plasmas, modelling shows that both the resonant electromagnetic torque and the neoclassical toroidal viscous (NTV) torque (due to the presence of 3D magnetic field perturbations) contribute to the toroidal flow damping, in particular near the

  2. 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.

  3. Note: Upgrade of electron cyclotron emission imaging system and preliminary results on HL-2A tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, M., E-mail: jiangm@swip.ac.cn; Shi, Z. B.; Zhong, W. L.; Chen, W.; Liu, Z. T.; Ding, X. T.; Yang, Q. W.; Zhang, B. Y.; Shi, P. W.; Liu, Y.; Fu, B. Z.; Xu, Y. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Domier, C. W.; Luhmann, N. C. [Department of Electrical and Computer Engineering, University of California, Davis, California 95616 (United States); Yang, Z. C. [School of Physics and Chemistry, Xihua University, Chengdu 610039 (China)

    2015-07-15

    The electron cyclotron emission imaging system on the HL-2A tokamak has been upgraded to 24 (poloidally) × 16 (radially) channels based on the previous 24 × 8 array. The measurement region can be flexibly shifted due to the independence of the two local oscillator sources, and the field of view can be adjusted easily by changing the position of the zoom lenses. The temporal resolution is about 2.5 μs and the achievable spatial resolution is 1 cm. After laboratory calibration, it was installed on HL-2A tokamak in 2014, and the local 2D mode structures of MHD activities were obtained for the first time.

  4. 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.

  5. 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.

  6. 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)

  7. 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)

  8. 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.)

  9. Upgrade of the COMPASS tokamak real-time control system

    Energy Technology Data Exchange (ETDEWEB)

    Janky, F., E-mail: filip.janky.work@gmail.com [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 18200 Prague (Czech Republic); Charles University in Prague, Faculty of Mathematics and Physics, V Holesovickach 2, 18000 Prague (Czech Republic); Havlicek, J. [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 18200 Prague (Czech Republic); Charles University in Prague, Faculty of Mathematics and Physics, V Holesovickach 2, 18000 Prague (Czech Republic); Batista, A.J.N. [Associação EURATOM/IST, Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade Técnica de Lisboa, 1049-001 Lisboa (Portugal); Kudlacek, O.; Seidl, J. [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 18200 Prague (Czech Republic); Neto, A.C. [Associação EURATOM/IST, Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade Técnica de Lisboa, 1049-001 Lisboa (Portugal); Pipek, J.; Hron, M. [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 18200 Prague (Czech Republic); Mikulin, O. [Institute of Plasma Physics, AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 18200 Prague (Czech Republic); Czech Technical University, Faculty of Nuclear Sciences and Physical Engineering, V Holesovickach 2, 18000 Prague (Czech Republic); and others

    2014-03-15

    Highlights: • An upgrade of the COMPASS real-time system has been made to generally improve the plasma performance. • Stability of discharges in SNT configuration has been increased. • Plasma flat-top phase length has been extended. • Central solenoid protection has been developed. • Plasma position estimation has been improved. - Abstract: The COMPASS plasma control system is based on the MARTe real-time framework. Thanks to MARTe modularity and flexibility new algorithms have been developed for plasma diagnostic (plasma position calculation), control (shaping field control), and protection systems (central solenoid protection). Moreover, the MARTe framework itself was modified to broaden the communication capabilities via Aurora. This paper presents the recent upgrades and improvements made to the COMPASS real-time plasma control system, focusing on the issues related to precision of the real-time calculations, and discussing the improvements in terms of discharge parameters and stability. In particular, the new real-time system has given the possibility to analyze and to minimize the transport delays of each control loop.

  10. Localized Scrape-Off Layer density modifications by Ion Cyclotron near fields in JET and ASDEX-Upgrade L-mode plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Colas, L., E-mail: laurent.colas@cea.fr [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Jacquet, Ph. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Van Eester, D. [LPP-ERM-KMS, TEC partner, Brussels (Belgium); Bobkov, V. [Max-Planck-Institut für Plasmaphysik, Garching (Germany); Brix, M.; Meneses, L. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Tamain, P. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Marsen, S. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Silva, C. [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa (Portugal); Carralero, D.; Kočan, M.; Müller, H.-W. [Max-Planck-Institut für Plasmaphysik, Garching (Germany); Crombé, K.; Křivska, A. [LPP-ERM-KMS, TEC partner, Brussels (Belgium); Goniche, M. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Lerche, E. [LPP-ERM-KMS, TEC partner, Brussels (Belgium); Rimini, F.G. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)

    2015-08-15

    Combining Lithium beam emission spectroscopy and edge reflectometry, localized Scrape-Off Layer (SOL) density modifications by Ion Cyclotron Range of Frequencies (ICRF) near fields were characterized in JET L-mode plasmas. When using the ICRF wave launchers connected magnetically to the Li-beam chord, the density decreased more steeply 2–3 cm outside the last closed flux surface (mapped onto the outer mid-plane) and its value at the outer limiter radial position was half the ohmic value. The depletion depends on the ICRF power and on the phasing between adjacent radiating straps. Convection due to ponderomotive effects and/or E × B{sub 0} drifts is suspected: during ICRF-heated H-mode discharges in 2013, DC potentials up to 70 V were measured locally in the outer SOL by a floating reciprocating probe, located toroidally several metres from the active antennas. These observations are compared with probe measurements on ASDEX-Upgrade. Their implications for wave coupling, heat loads and impurity production are discussed.

  11. Modelling of the ICRF induced E  ×  B convection in the scrape-off-layer of ASDEX Upgrade

    Science.gov (United States)

    Zhang, W.; Feng, Y.; Noterdaeme, J.-M.; Bobkov, V.; Colas, L.; Coster, D.; Lunt, T.; Bilato, R.; Jacquot, J.; Ochoukov, R.; Van Eester, D.; Křivská, A.; Jacquet, P.; Guimarais, L.; the ASDEX Upgrade Team

    2016-09-01

    In magnetic controlled fusion devices, plasma heating with radio-frequency (RF) waves in the ion cyclotron (IC) range of frequency relies on the electric field of the fast wave to heat the plasma. However, the slow wave can be generated parasitically. The electric field of the slow wave can induce large biased plasma potential (DC potential) through sheath rectification. The rapid variation of the rectified potential across the equilibrium magnetic field can cause significant convective transport (E  ×  B drifts) in the scrape-off layer (SOL). In order to understand this phenomenon and reproduce the experiments, 3D realistic simulations are carried out with the 3D edge plasma fluid and kinetic neutral code EMC3-Eirene in ASDEX Upgrade. For this, we have added the prescribed drift terms to the EMC3 equations and verified the 3D code results against the analytical ones in cylindrical geometry. The edge plasma potential derived from the experiments is used to calculate the drift velocities, which are then treated as input fields in the code to obtain the final density distributions. Our simulation results are in good agreement with the experiments.

  12. Characterization of edge profiles and fluctuations in discharges with type-II and nitrogen-mitigated edge localized modes in ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Wolfrum, E; Bernert, M; Burckhart, A; Classen, I G J; Conway, G D; Eich, T; Fischer, R; Gude, A; Herrmann, A; Maraschek, M; McDermott, R; Puetterich, T; Wieland, B [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany); Boom, J E [FOM Institute for Plasmaphysics, Rijnhuizen, Association EURATOM-FOM, Nieuwegein (Netherlands); Luhmann, N C Jr [University of California at Davis, Davis, CA95616 (United States); Park, H K [POSTECH, Pohang, Gyeongbuk, 790-784 (Korea, Republic of); Vicente, J [Associacao EURATOM/IST, Instituto de Plasmas e Fosao Nuclear - Laboratorio Associado, IST, Lisbon (Portugal); Willensdorfer, M, E-mail: e.wolfrum@ipp.mpg.de [Institute of Applied Physics, Vienna University of Technology, Association EURATOM-OEAW, Vienna (Austria)

    2011-08-15

    Edge localized modes (ELMs) with high frequency and low power loss (type-II ELMs) occur in high triangularity, near double null configurations in ASDEX Upgrade with full tungsten plasma facing components. The transition from type-I to type-II ELMs is shown to occur above a collisionality threshold. For the first time the characteristic MHD fluctuations around 40 kHz have been localized. The fluctuations are observed in a wide region extending from the pedestal inward to normalized poloidal radius {rho}{sub pol} = 0.7. Their amplitudes on the low-field side of the plasma exhibit maxima above and below the mid-plane. The fluctuations move in the electron drift direction and lead to a reduced edge electron temperature gradient. The reduction in the edge pressure gradient is connected with these MHD fluctuations, which affect the electron temperature but not the electron density profiles. A comparison with nitrogen-mitigated type-I ELMs in the same plasma shape shows that core profiles are also affected. The electron temperature profile is self-similar for type-I and nitrogen-mitigated type-I ELMs but is not self-similar in the case of type-II ELMs.

  13. Upgrade of plasma density feedback control system in HT-7 tokamak

    Institute of Scientific and Technical Information of China (English)

    ZHAO Da-Zheng; LUO Jia-Rong; LI Gang; JI Zhen-Shan; WANG Feng

    2004-01-01

    The HT-7 is a superconducting tokamak in China used to make researches on the controlled nuclear fusion as a national project for the fusion research. The plasma density feedback control subsystem is the one of the subsystems of the distributed control system in HT-7 tokamak (HT7DCS). The main function of the subsystem is to control the plasma density on real-time. For this reason, the real-time capability and good stability are the most significant factors, which will influence the control results. Since the former plasma density feedback control system (FPDFCS) based on Windows operation system could not fulfill such requirements well, a new subsystem has to be developed. The paper describes the upgrade of the plasma density feedback control system (UPDFCS), based on the dual operation system (Windows and Linux), in detail.

  14. 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...

  15. Upgrade of the diagnostic neutral beam injector for the TCV tokamak

    International Nuclear Information System (INIS)

    A diagnostic neutral beam injector (DNBI) [CRPP report LRP 710/01, CRPP-EPFL, 2001; EPS Conf. Contr. Fusion Plasma Phys., 25A (2001) 365] has been installed on tokamak a configuration variable (TCV) [Plasma Phys. Control Fusion, 36 (1994) B277; Plasma Phys. Control Fusion, 43 (2001) A161; Plasma Phys. Control Fusion, to be published] for the purpose of providing local measurements of plasma ion temperature, velocity and impurity density by Charge eXchange recombination spectroscopy (CXRS) [EPS Conf. Contr. Fusion Plasma Phys., 25A (2001) 365]. The system recently underwent a technical upgrade, which allowed to increase the full neutral beam current density by a factor of two (from 0.5 to 1 A at 52 keV injection energy) and to extend the operational range of the diagnostic. This was achieved by means of a new, larger ion source, with an increased extraction area and corresponding enhancements of the power supplies

  16. New plasma configurations in the TCV tokamak, TCV upgrades. Two steps towards the realisation of fusion

    International Nuclear Information System (INIS)

    In view of realising the full potential of fusion as an abundant energy source, some challenges must still be solved. They are identified and will be addressed by the implementation of the EU fusion roadmap. The TCV tokamak, with its high plasma shaping capability and the flexibility of its heating and current drive systems, is strongly contributing to this effort, as one of a small number of devices selected by the EU community for the 2014-2020 period. One of the primary challenges lies in the heat exhaust from tokamak plasmas. Indeed, the currently foreseen operational regime of ITER implies heat flows impinging onto the facing materials that are not compatible with a fully operating fusion reactor. TCV has developed alternative plasma configurations, termed 'snowflakes', that strongly reduce the heat flow towards the vessel walls, via an increase in the number of deposition surface areas, as shown in Fig. 1. Measurement of particle fluxes, together with IR camera imaging, show a clear reduction of the heat flow onto the walls. The TCV tokamak is going through major upgrades of its heating systems to expand its operational domain towards burning plasma regimes. The installation of a 1MW neutral beam injector will allow the achievement of high temperature plasmas with equal ion and electron temperatures. An additional 2MW of electron cyclotron resonance heating power will be installed to increase the plasma pressure near the range in which ITER will operate. This will also improve access to and control of high confinement regimes. Varying the power ratio between the two heating systems will furthermore lead to improved understanding of the different plasma turbulence regimes that develop in plasmas with different electron to ion temperature ratios. Acknowledgement: This work was partly supported by the Swiss National Science Foundation. (author)

  17. 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

  18. Overview of ASDEX Upgrade results

    DEFF Research Database (Denmark)

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

    2011-01-01

    the oscillatory geodesic acoustic modes, turbulent fluctuations and the mean equilibrium E × B flow in the edge negative Er well region just inside the separatrix. Improved pedestal diagnostics revealed also a refined picture of the pedestal transport in the fully developed H-mode type-I ELM cycle. Impurity ion...

  19. Overview of ASDEX Upgrade results

    DEFF Research Database (Denmark)

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

    2013-01-01

    resonance heating (ECRH) power, by installing 2 × 8 internal magnetic perturbation coils, and by improving the ion cyclotron range of frequency compatibility with the tungsten wall. With the perturbation coils, reliable suppression of large type-I edge localized modes (ELMs) could be demonstrated in a wide...... operational window, which opens up above a critical plasma pedestal density. The pellet fuelling efficiency was observed to increase which gives access to H-mode discharges with peaked density profiles at line densities clearly exceeding the empirical Greenwald limit. Owing to the increased ECRH power of 4 MW...... simultaneously available heating power of 23 MW, high performance discharges have been carried out where feed-back controlled radiative cooling in the core and the divertor allowed the divertor peak power loads to be maintained below 5 MW m−2. Under attached divertor conditions, a multi-device scaling expression...

  20. FT Tokamak Upgrade (FTU) vacuum vessel section cleaning by glow discharge in hydrogen

    Energy Technology Data Exchange (ETDEWEB)

    Ciotti, M.; Apicella, M.L.; Verdini, L.; Ferro, C.

    1991-09-01

    The possibility of applying glow discharge in hydrogen for the cleaning of the FTU (Frascati Tokamak Upgrade) vacuum chamber was analyzed on a 1:1 scale toroidal section by using the same operating conditions as foreseen for the machine. The discharge was maintained for six hours in the chamber with the wall temperature kept at 150 degrees C. The partial pressures at the end of the cleaning run were compared with those obtained by using only thermal outgassing at the same temperature. A reduction of about a factor of two in the H/sub 2/0 and C0/sub 2/ partial pressures was observed, related to a better cleanness of the surface. It was found that the high temperature during the glow discharge cleaning not only increases the efficiency of the discharge, but it is an efficient tool to remove impurities from the hidden regions, defined by the thermal shields that cover all the vacuum vessel walls not directly exposed to the glow discharge.

  1. Measurement of the Radiative Cooling Coefficient of Krypton Gas in the Frascati Tokamak Upgrade

    Science.gov (United States)

    Fournier, K. B.; Goldstein, W. H.; Pacella, D.; Mazzitelli, G.; Gabellieri, L.; Leigheb, M.; de Angelis, R.; May, M. J.; Regan, S. P.; Stutman, D.; Soukhanovskii, V.; Finkenthal, M.; Moos, H. W.

    1997-11-01

    For future fusion reactors, a careful balance must be achieved between the cooling of the outer plasma via impurity radiation and the deleterious effects of inevitable core penetration by impurity ions. We extract the krypton impurity radial profile and the radiative cooling rate for krypton gas in the Frascati Tokamak Upgrade (FTU). The measured bolometric, soft x-ray and visible bremmstrhalung signals are Abel inverted and then incorporated in an analytic model. Using the known (calculated) ionization state distribution, the radial power loss profile for krypton is derived. Anamolous transport is assumed to have a negligible affect on the total krypton radiation profile; this assumption is confirmed using the derived krypton radiation rate in a plasma transport modeling code. The level of intrinsic impurities (Mo, Cr, Mn and Fe) in the plasma during the krypton puffing is monitored with a VUV SPRED spectrometer. Models for krypton emissivity from the literature are compared to our measured results. These initial results are part of a multiwavelength impurity spectroscopy campaign that will measure transport profiles and basic atomic data in the FTU. Work carried out under the auspices of the U.S. DoE, Contract No. W-7405-ENG-48.

  2. Planned upgrade to the coaxial plasma source facility for high heat flux plasma flows relevant to tokamak disruption simulations

    International Nuclear Information System (INIS)

    Plasma disruptions in tokamaks remain serious obstacles to the demonstration of economical fusion power. In disruption simulation experiments, some important effects have not been taken into account. Present disruption simulation experimental data do not include effects of the high magnetic fields expected near the PFCs in a tokamak major disruption. In addition, temporal and spatial scales are much too short in present simulation devices to be of direct relevance to tokamak disruptions. To address some of these inadequacies, an experimental program is planned at North Carolina State University employing an upgrade to the Coaxial Plasma Source (CPS-1) magnetized coaxial plasma gun facility. The advantages of the CPS-1 plasma source over present disruption simulation devices include the ability to irradiate large material samples at extremely high areal energy densities, and the ability to perform these material studies in the presence of a high magnetic field. Other tokamak disruption relevant features of CPS-1U include a high ion temperature, high electron temperature, and long pulse length

  3. 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

  4. Upgrade of Langmuir probe diagnostic in ITER-like tungsten mono-block divertor on experimental advanced superconducting tokamak

    Science.gov (United States)

    Xu, J. C.; Wang, L.; Xu, G. S.; Luo, G. N.; Yao, D. M.; Li, Q.; Cao, L.; Chen, L.; Zhang, W.; Liu, S. C.; Wang, H. Q.; Jia, M. N.; Feng, W.; Deng, G. Z.; Hu, L. Q.; Wan, B. N.; Li, J.; Sun, Y. W.; Guo, H. Y.

    2016-08-01

    In order to withstand rapid increase in particle and power impact onto the divertor and demonstrate the feasibility of the ITER design under long pulse operation, the upper divertor of the EAST tokamak has been upgraded to actively water-cooled, ITER-like tungsten mono-block structure since the 2014 campaign, which is the first attempt for ITER on the tokamak devices. Therefore, a new divertor Langmuir probe diagnostic system (DivLP) was designed and successfully upgraded on the tungsten divertor to obtain the plasma parameters in the divertor region such as electron temperature, electron density, particle and heat fluxes. More specifically, two identical triple probe arrays have been installed at two ports of different toroidal positions (112.5-deg separated toroidally), which can provide fundamental data to study the toroidal asymmetry of divertor power deposition and related 3-dimension (3D) physics, as induced by resonant magnetic perturbations, lower hybrid wave, and so on. The shape of graphite tip and fixed structure of the probe are designed according to the structure of the upper tungsten divertor. The ceramic support, small graphite tip, and proper connector installed make it possible to be successfully installed in the very narrow interval between the cassette body and tungsten mono-block, i.e., 13.5 mm. It was demonstrated during the 2014 and 2015 commissioning campaigns that the newly upgraded divertor Langmuir probe diagnostic system is successful. Representative experimental data are given and discussed for the DivLP measurements, then proving its availability and reliability.

  5. Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Meglicki, Z

    1995-09-19

    We describe in detail the implementation of a weighted differences code, which is used to simulate a tokamak using the Maschke-Perrin solution as an initial condition. The document covers the mainlines of the program and the most important problem-specific functions used in the initialization, static tests, and dynamic evolution of the system. The mathematics of the Maschke-Perrin solution is discussed in parallel with its realisation within the code. The results of static and dynamic tests are presented in sections discussing their implementation.The code can also be obtained by ftp -anonymous from cisr.anu.edu.au Directory /pub/papers/meglicki/src/tokamak. This code is copyrighted. (author). 13 refs.

  6. 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.)

  7. 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.)

  8. UPGRADES

    CERN Multimedia

    J. Spalding and D. Contardo

    2012-01-01

      The CMS Upgrade Programme consists of four classes of projects: (a) Detector and Systems upgrades which are ongoing and largely (though not entirely) target LS1. (b) Full system upgrades for three projects that are preparing TDRs: Pixels, HCAL and L1 Trigger. The projects target completion by LS2. (c) Infrastructure consolidation and upgrades to improve operational robustness and to support the above projects. (d) Phase 2 replacement of the Tracker and major upgrades of the Trigger and Forward Detectors. For (a) and (c), detailed costing exists and is being integrated into a common reporting system. The schedule milestones for each project will be linked into the overall schedule planning for LS1. For the three TDR projects, the designs have progressed significantly since the Technical Proposal in 2010. Updated detailed cost estimates and schedules will be prepared with the TDRs to form the basis for tracking the projects through completion. To plan the upgrades and the supporting simulati...

  9. 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.

  10. UPGRADES

    CERN Multimedia

    J. Butler and J. Nash

    2011-01-01

    Recent progress on the CMS upgrades was summarised, in a workshop held at Fermilab between 7th and 10th November, attended by more than 150 people, many of whom came from Europe and Asia. Important goals of the workshop were to begin to formulate a schedule for the upgrades and to determine project interdependencies. Input was received from all the upgrade working groups and will be combined into a first-pass schedule over the next several weeks. In addition, technical progress on each of the major subtasks was presented and plans for the near-term future were established. Slides from the more than 100 talks are located at: https://indico.cern.ch/conferenceDisplay.py?confId=153564 In the opening plenary session, Frank Zimmermann, of the CERN Beams Department, gave his view of the LHC luminosity evolution. The luminosity will increase faster than we assumed in designing the upgrades. CMS will need to re-evaluate the current upgrade plans and revise them if necessary. CMS Upgrade Physics coordinator...

  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. Applications of ECH on the DIII-D tokamak and projections for future ECH upgrades

    Directory of Open Access Journals (Sweden)

    Solomon W.M.

    2012-09-01

    Full Text Available Electron Cyclotron Heating and Current Drive plays an important role in the DIII-D program. In high performance discharges EC power contributes greatly to MHD stability, and this is particularly important for discharges with low rotational torque applied, as will be the case for ITER. Off-axis EC current drive also plays a key role in the actualization of steady-state scenarios by supporting the desired current profile. In order to carry out these applications at higher beta and higher field, an upgrade of the EC power to 15 MW is needed, and the best gyrotron frequency for the DIII-D program is 117.5 GHz.

  13. UPGRADES

    CERN Document Server

    D. Contardo and J. Spalding

    2013-01-01

    There is very good progress in the execution of the LS1 projects and in launching construction of the Phase 1 upgrades. We focus here on two main achievements since the last CMS Week. The approval of the third Phase 1 TDR The preparation of the L1 Trigger Upgrade Technical Design Report has been a major effort of the collaboration at the beginning of this year, especially to develop supporting Trigger menu and physics performance studies. These studies have demonstrated the efficiency of the upgraded system to ensure low lepton and jet trigger thresholds, leading to a significant increase of the acceptance for the Higgs measurements, in the associated production mode and in the ττ decays, as well as for the stop searches involving multiple jets in the final state. The TDR was submitted to the LHCC in May and approved at the June committee meeting. It is now a public document, completing the series of the three TDRs describing the Phase 1 upgrades, with the new Pixel system and the HCAL rea...

  14. 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-...

  15. Electron cyclotron emission radiometer upgrade on the J-TEXT tokamak

    Science.gov (United States)

    Yang, Z. J.; Pan, X. M.; Ma, X. D.; Ruan, B. W.; Zhou, R. B.; Zhang, C.

    2016-11-01

    To meet experimental requirements, the J-TEXT electron cyclotron emission (ECE) diagnostic is being upgraded. The front end antenna and transmission line have been modified and a new 8-channel W-band detecting unit has been developed. The improved ECE system will extend the frequency range from 94.5-124.5 GHz to 80.5-124.5 GHz. This will enable the system to cover the most plasma in the radius direction for BT = 1.8-2.2 T, and it even can cover almost the whole plasma range ρ = - 0.8-0.9 (minus means the high field side) at BT = 1.8 T. A new auxiliary channel bank with 8 narrow band, tunable yttrium iron garnet filters is planned to add to the ECE system. Due to observations along a major radius, perpendicular to BT, and relatively low electron temperature, Doppler and relativistic broadening are minimal and thus high spatial resolution measurements can be made at variable locations with these tunable channels.

  16. B2-Eirene modelling of ASDEX Upgrade

    Science.gov (United States)

    Coster, D. P.; Schneider, R.; Neuhauser, J.; Bosch, H.-S.; Wunderlich, R.; Fuchs, C.; Mast, F.; Kallenbach, A.; Dux, R.; Becker, G.; Braams, B. J.; Reiter, D.; ASDEX Upgrade Team

    1997-02-01

    The extension of the computational region of the coupled fluid plasma, Monte-Carlo neutrals code, B2-Eirene, to the plasma center is discussed. The simulation of completely detached H-mode plasma is presented, as is the modelling of He and Ne compression.

  17. 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.

  18. UPGRADES

    CERN Multimedia

    D. Contardo and J. Spalding

    2012-01-01

      Good progress is being made on the projects that will be installed during LS1. CSC chamber production for ME4/2 is progressing at a rate of four chambers per month, with 25 built so far, and the new electronics for ME1/1 is undergoing a pre-production integration testing. For the RPC chambers, gap production is underway with first deliveries to the chamber assembly sites at CERN and Ghent. The third site at Mumbai will begin production next month. For the PMT replacement in the forward hadron calorimeters (HF), the 1728 PMTs are all characterised and ready to be installed. Testing of the electronics boards is going well. Preparations to replace the HPDs in the outer calorimeter (HO) with SiPMs are also on-track. All components are at CERN and burn-in of the new front-end electronics is proceeding. There are three major upgrade projects targeting the period from LS1 through LS2: a new pixel detector, upgraded photo-detectors and electronics for HCAL, and development of a new L1 Trigger. The new ...

  19. UPGRADES

    CERN Multimedia

    D. Contardo and J. Spalding

    2013-01-01

    The three post-LS1 Phase 1 Upgrade projects (the L1-Trigger, Pixel Tracker, and HCAL) are all making excellent progress and are transitioning from the prototype to the execution phase. Meanwhile plans are developing for Phase 2, a major Upgrade programme targeting the third long shutdown, LS3. News on Phase 1 is included under the respective projects; we only provide a brief summary here. Phase 1 The plan for the L1 Trigger relies on the installation during the present shutdown of optical splitting for the Trigger input signals. This will allow the new Trigger system to be brought online and fully commissioned during beam operation in 2015, while CMS relies on the existing legacy Trigger for physics. Once fully commissioned the experiment can switch over to the new Trigger, which will provide greatly improved performance at high event pile-up, by 2016. System tests of the splitter system, and of the new architecture of the calorimeter trigger were very successful, and the work in LS1 is on-track. Prototype ...

  20. UPGRADES

    CERN Multimedia

    Didier Contardo

    2012-01-01

      The CMS Upgrade Programme is making good progress on the LS1 and Phase 1 projects, in the planning for Phase 2. The construction of the ME4/2 muon chambers to be installed during LS1 has started and the two first CSC production chambers have been fully qualified. The three muon groups have recently established a set of milestones towards the completion of their project, that will be integrated in the detailed planning and scheduling for the shutdown work established by Technical Coordination. The project to replace the photo-detectors in the HF and HO calorimeters is also well advanced and at the validation stage. The operation of an HF slice with new multi-anode PMTs and back-end electronics has already been demonstrated in 2012. For the Phase 1 data-taking, as discussed in the Chamonix workshop, it is likely that the LHC performance will exceed the nominal luminosity and pile-up before the second shutdown, still scheduled in 2018. The collaboration is therefore pursuing a strategy to upgrade ...

  1. 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.)

  2. UPGRADES

    CERN Multimedia

    D. Contardo and J. Spalding

    2013-01-01

      LS1 and Phase 1 The detector projects targeting LS1 are progressing well, and a fully integrated schedule developed by Technical Coordination includes installation milestones and a detailed work-plan. The first chambers of the RPC system were produced and are being qualified. Production will ramp up this year to a rate of 20 chambers per month. 32 chambers of the CSC system have been fabricated for the ME4/2 CSC stations, and production proceeds at a rate of 4 per month. The new ME1/1 Front-End Board is in production and the off-detector electronics integration tests are ongoing. The new Theta Trigger Boards for the DT readout production is started and the relocation of the Sector Collector boards with new Optical Links as been successfully tested. All the components for the upgrade of the Forward Hadron Calorimeter PMTs have been received at CERN and assemblies are being qualified. The situation is similar for the Hadron Outer Calorimeter new SiPMs and readout modules. Three projects are plan...

  3. 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.

  4. 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.)

  5. Two steps towards the realization of fusion: New plasma configurations in the TCV tokamak and its ongoing upgrades

    OpenAIRE

    Martin Yves; Duval Basil P.; Karpushov Alexander N.; Labit Benoit; Reimerdes Holger

    2014-01-01

    To realise the potential of fusion as an abundant energy source, several challenges remain. The TCV tokamak, featuring high shaping capability and a flexible heating system, is strongly contributing to solving these challenges. A fundamental challenge remains in controlling heat exhaust from the plasma. ITER's currently foreseen operational regime implies heat flows to the plasma facing materials that are not compatible with a commercial fusion reactor. TCV has demonstrated alternative plasma...

  6. 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.)

  7. 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...

  8. 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

  9. 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.)

  10. 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.

  11. 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.)

  12. 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.)

  13. 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

  14. 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.

  15. 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.

  16. 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.)

  17. 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.

  18. 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.

  19. 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.

  20. 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

  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. 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

  3. 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

  4. 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

  5. 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.

  6. MAST Upgrade - Construction Status

    CERN Document Server

    Milnes, Joe; Dhalla, Fahim; Fishpool, Geoff; Hill, John; Katramados, Ioannis; Martin, Richard; Naylor, Graham; O'Gorman, Tom; Scannell, Rory

    2015-01-01

    The Mega Amp Spherical Tokamak (MAST) is the centre piece of the UK fusion research programme. In 2010, a MAST Upgrade programme was initiated with three primary objectives, to contribute to: 1) Testing reactor concepts (in particular exhaust solutions via a flexible divertor allowing Super-X and other extended leg configurations); 2) Adding to the knowledge base for ITER (by addressing important plasma physics questions and developing predictive models to help optimise ITER performance of ITER) and 3) Exploring the feasibility of using a spherical tokamak as the basis for a fusion Component Test Facility. With the project mid-way through its construction phase, progress will be reported on a number of the critical subsystems. This will include manufacture and assembly of the coils, armour and support structures that make up the new divertors, construction of the new set coils that make up the centre column, installation of the new power supplies for powering the divertor coils and enhanced TF coil set, progr...

  7. ISTTOK control system upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, Ivo S., E-mail: ivoc@ipfn.ist.utl.pt; Duarte, Paulo; Fernandes, Horácio; Valcárcel, Daniel F.; Carvalho, Pedro J.; Silva, Carlos; Duarte, André S.; Neto, André; Sousa, Jorge; Batista, António J.N.; Carvalho, Bernardo B.

    2013-10-15

    Highlights: •ISTTOK fast controller. •All real-time diagnostic and actuators were integrated in the control platform. •100 μs control cycle under the MARTe framework. •The ISTTOK control system upgrade provides reliable operation with an improved operational space. -- Abstract: The ISTTOK tokamak (Ip = 4 kA, BT = 0.5 T, R = 0.46 m, a = 0.085 m) is one of the few tokamaks with regular alternate plasma current (AC) discharges scientific programme. In order to improve the discharge stability and to increase the number of AC discharge cycles a novel control system was developed. The controller acquires data from 50 analog-to-digital converter (ADC) channels of real-time diagnostics and measurements: tomography, Mirnov coils, interferometer, electric probes, sine and cosine probes, bolometer, current delivered by the power supplies, loop voltage and plasma current. The system has a control cycle of 100 μs during which it reads all the diagnostics connected to the advanced telecommunications computing architecture (ATCA) digitizers and sends the control reference to ISTTOK actuators. The controller algorithms are executed on an Intel{sup ®} Q8200 chip with 4 cores running at 2.33 GHz and connected to the I/O interfaces through an ATCA based environment. The real-time control system was programmed in C++ on top of the Multi-threaded Application Real-Time executor (MARTe). To extend the duration of the AC discharges and the plasma stability a new magnetising field power supply was commissioned and the horizontal and vertical field power supplies were also upgraded. The new system also features a user-friendly interface based on HyperText Markup Language (HTML) and Javascript to configure the controller parameters. This paper presents the ISTTOK control system and the consequent update of real-time diagnostics and actuators.

  8. The Pegasus-Upgrade Experiment

    Science.gov (United States)

    Fonck, R. J.; Bongard, M. W.; Barr, J. L.; Frerichs, H. G.; Lewicki, B. T.; Reusch, J. A.; Schmitz, O.; Winz, G. R.

    2015-11-01

    Tokamak operation at near-unity aspect ratio provides access to advanced tokamak physics at modest parameters. High plasma current is accessible at very low toroidal field. This offers H-mode performance at Te levels that allow use of electrostatic and magnetic probe arrays through the edge pedestal region into the plasma core. An upgrade to the Pegasus ST is planned to exploit these features and pursue unique studies in three areas: local measurements of pedestal and ELM dynamics at Alfvenic timescales; direct measurement of the local plasma response to application of 3D magnetic perturbations with high spectral flexibility; and extension of Local Helicity Injection for nonsolenoidal startup to NSTX-U-relevant confinement and stability regimes. Significant but relatively low-cost upgrades to the facility are proposed: a new centerstack with larger solenoid and 2x the number of toroidal field conductors; a new TF power supply and conversion of the 200 MVA OH power supply to a cascaded multilevel inverter configuration; and installation of an extensive 3D-magnetic perturbation coil system for ELM mitigation and suppression studies. The upgraded facility will provide 0.3 MA plasmas with pulse lengths of 50-100 msec flattop, aspect ratio <1.25, and toroidal field up to 0.4 T. These research activities will be integrated into related efforts on DIII-D and NSTX-U. Work supported by US DOE grant DE-FG02-96ER54375.

  9. 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

  10. 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

  11. Performance and development of the DIII-D tokamak core

    International Nuclear Information System (INIS)

    The DIII-D tokamak is an upgrade of the Doublet III configuration which has operated since early 1986. This paper presents recent advances in performance using the upper divertor, fabrication development for vanadium components, operation of the helium leak checking in a high deuterium background, and restoration of the damaged Ohmic heating solenoid

  12. ECE Imaging Bandwidth Upgrade for TEXTOR

    Science.gov (United States)

    Domier, C. W.; Zhang, P.; Luhmann, N. C., Jr.; Park, H. K.; van de Pol, M. J.; Spakman, G. W.; Jaspers, R.; Donne, A. J. H.

    2007-11-01

    The 128 channel 2-D Electron Cyclotron Emission (ECE) Imaging system collects time-resolved 16x8 images of electron temperature profiles and fluctuations on the TEXTOR tokamak. This instrument was upgraded in February 2007 with new wideband ECE electronics which increased the instantaneous frequency coverage by >50% to 6.4 GHz with a corresponding increase in horizontal plasma coverage. Frequency extenders have been developed to combine modules together to double the instantaneous coverage to 12.8 GHz. Technical details regarding both the electronics upgrade and the frequency extenders as well as the preliminary physics results will be presented. Implementation of a similar but new ECEI instrument on the DIII-D tokamak will be extensively discussed.

  13. ASDEX upgrade results - publications and conference contributions. Period 10/93 to 7/94

    International Nuclear Information System (INIS)

    The report contains the papers of IPP members contributed to the conferences as follows: 1) 4th H-Mode Workshop, NAKA-JAERI, Japan Nov. 1993; 2) 11th Topical Meeting on the Technology of Fusion Energy, New Orleans, June 1994; 3) 11th International Conference on Plasma Surface Interactions in Controlled Fusion Devices, Mito-shi, Japan, May 23-27, 1994; 4) 21th EPS Conference on Controlled Fusion and Plasma Physics, Montpellier, June 27 to July 1, 1994. (HP)

  14. On the asymmetries of ELM divertor power deposition in JET and ASDEX Upgrade

    DEFF Research Database (Denmark)

    Eich, T.; Kallenbach, A.; Fundamenski, W.;

    2009-01-01

    An analytical expression was derived for describing the divertor target power during ELMs based on the model discussed in [W. Fundamenski, R.A. Pitts, Plasma Phys. Control. Fus. 48 (2006) 109] where the power load arises from a Maxwellian distribution of particles released into the SOL region......-streaming-particle (FSP) approach predicts a dependence of the ELM in/out energy balance of the pedestal Mach number as well as an inversion of the in/out balance by a change of the field line helicity as observed experimentally. From the FSP approach the value for EτIR (see text) is predicted to be 18–25% in good...

  15. High frequency magnetic fluctuations correlated with the inter-ELM pedestal evolution in ASDEX Upgrade

    Science.gov (United States)

    Laggner, F. M.; Wolfrum, E.; Cavedon, M.; Mink, F.; Viezzer, E.; Dunne, M. G.; Manz, P.; Doerk, H.; Birkenmeier, G.; Fischer, R.; Fietz, S.; Maraschek, M.; Willensdorfer, M.; Aumayr, F.; the EUROfusion MST1 Team; the ASDEX Upgrade Team

    2016-06-01

    In order to understand the mechanisms that determine the structure of the high confinement mode (H-mode) pedestal, the evolution of the plasma edge electron density and temperature profiles between edge localised modes (ELMs) is investigated. The onset of radial magnetic fluctuations with frequencies above 200 kHz is found to correlate with the stagnation of the electron temperature pedestal gradient. During the presence of these magnetic fluctuations the gradients of the edge electron density and temperature are clamped and stable against the ELM onset. The detected magnetic fluctuation frequency is analysed for a variety of plasma discharges with different electron pressure pedestals. It is shown that the magnetic fluctuation frequency scales with the neoclassically estimated \\text{E} × \\text{B} velocity at the plasma edge. This points to a location of the underlying instability in the gradient region. Furthermore, the magnetic signature of these fluctuations indicates a global mode structure with toroidal mode numbers of approximately 10. The fluctuations are also observed on the high field side with significant amplitude, indicating a mode structure that is symmetric on the low field side and high field side. The associated fluctuations in the current on the high field side might be attributed to either a strong peeling part or the presence of non-adiabatic electron response.

  16. Fast Ion Dynamics in ASDEX Upgrade and TEXTOR Measured by Collective Thomson Scattering

    DEFF Research Database (Denmark)

    Moseev, Dmitry

    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...

  17. Trends of W behavior in ICRF assisted discharges in ASDEX Upgrade

    Science.gov (United States)

    Czarnecka, A.; Bobkov, V.; Dux, R.; Pütterich, T.; Sertoli, M.

    2015-08-01

    During Ion Cyclotron Range of Frequencies (ICRF) operation with metallic plasma facing components, plasma wall interaction processes are modified, which needs to be understood in order to reduce the ICRF-related rise in impurity concentration and maximize ICRF coupling. This contribution is focused on gathering and analyzing a database on W-behavior during ICRF spanning a multi-parameter space. The gas puff dependence is investigated in selected pulses that feature the injection of deuterium (D2) main gas, with impurity gas nitrogen (N2), argon (Ar) and krypton (Kr). CW was found to decrease with the total gas injection rate, which was also correlated with the changes of electron density in the location of the CW measurements. Although the addition of N2, Ar and Kr impacted on the increase in sputtering, to achieve a similar CW level with mixed puffing, one needs to inject considerably more gas. Moreover, CW is shown to change after implementation of boron coatings on the ICRF antenna limiters.

  18. 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;

    2014-01-01

    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...

  19. ASDEX upgrade results. Publications and conference contributions. Period 6/92 to 9/93

    International Nuclear Information System (INIS)

    The report contains the papers of IPP members contributed to the conferences as follows: 1. 29th EPS Conference on Controlled Fusion and Plasma Physics, Innsbruck, June 19-July 3, 1992; 2. 14th International Conference on Plasma Physics and Controlled Nuclear Fusion Research, Wuerzburg, Sept. 30-Oct.7, 1992; 3. 10th APS Conference on Radio Frequency Power to Plasmas, Boston, 1993, 4. 20th EPS Conference on Controlled Fusion and Plasma Physics, Lisbon, July 26-30, 1993; 4. 18th International Conference on Infrared and Millimeter Waves, University of Essex, Sept. 6-10, 1993. (HP)

  20. 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

    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...

  1. Collective Thomson scattering measurements of fast-ion transport due to sawtooth crashes in ASDEX Upgrade

    DEFF Research Database (Denmark)

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

    2016-01-01

    Sawtooth instabilities can modify heating and current-drive profiles and potentially increase fast-ion losses. Understanding how sawteeth redistribute fast ions as a function of sawtooth parameters and of fast-ion energy and pitch is hence a subject of particular interest for future fusion devices...

  2. 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..;

    2014-01-01

    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...... correlation were used to determine the poloidal velocity and its jump, yielding comparable results to the first method. Also the method of conditional averaging was applied to the latter signals. ((c) 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)....

  3. 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

  4. 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)

  5. 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

  6. Subsidiary Upgrading?

    OpenAIRE

    Dörrenbächer, Christoph; Gammelgård, Jens

    2004-01-01

    Abstract This study reports the results of interviews with 65 managers in 11 German headquarters and in their 13 Hungarian subsidiaries. We focused on the role of the subsidiary with regard to market, product and value-adding mandates. Further, we investigated whether the Hungarian subsidiaries had experienced an upgrade of their role during the first 10 years of transition. The host country economy was supportive to role development, but inadequate subsidiary capabilities a...

  7. Time dependent neutral gas transport in tokamak edge plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Reiter, D. [Institut fuer Plasmaphysik, Forschungszentrum Juelich GmbH, Association EURATOM-KFA Postfach 1913, D-52425, Juelich (Germany); May, C. [Institut fuer Plasmaphysik, Forschungszentrum Juelich GmbH, Association EURATOM-KFA Postfach 1913, D-52425, Juelich (Germany); Coster, D. [Max-Planck-Institut fuer Plasmaphysik, EURATOM-Association, D-85740, Garching (Germany); Schneider, R. [Max-Planck-Institut fuer Plasmaphysik, EURATOM-Association, D-85740, Garching (Germany)

    1995-04-01

    The effects of neutral particles on the edge plasma conditions play a key role in divertor and limiter physics. In computational models they are usually treated in steady state approximation (instantaneous relaxation). However, the characteristic transport time scale is comparable to the ion acustic time scale. Thus neutral atoms relax to their steady state distributions much slower than electron temperature profiles along the fieldlines are established. A computational assessment of divertor or limiter dynamics requires ultimately an extension to time dependent algorithms. The numerical procedure in the EIRENE Monte Carlo code is presented. A first numerical study of ELM`s in the ASDEX-Upgrade divertor plasma has been carried out and the results are briefly discussed. ((orig.)).

  8. 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

  9. Study of anomalous inward drift in tokamaks by transport analysis and simulations

    Science.gov (United States)

    Becker, G.

    2004-09-01

    The origin of the anomalous inward drift is explored by transport analysis of Ohmic, L- and H-mode discharges in ASDEX Upgrade using a special version of the 1.5-D BALDUR transport code. It is shown that the anomalous particle pinch significantly affects the density profile, in contrast to the Ware pinch. The measured density profiles can be modelled by the anomalous inward drift velocity \\smash{v_in=C_{v}2xD/({\\rho_{w}x^{2}_{s}})} with Cv equal to 0.2 for H-mode and 1.1 for Ohmic plasmas and by a strongly rising vin/D near the edge. Here, x = rgr/rgrw and xs = rgrs/rgrw with effective radii rgr, rgrw and rgrs of flux surface, wall contour and separatrix contour, respectively. At low densities, beam fuelling alone yields peaked density profiles. With increasing density the beam fuelling is shifted to the edge which causes the observed density flattening. Evaluation of measured electron density and temperature profiles in deuterium and hydrogen discharges with various heating schemes yields \\smash{C_{v} \\propto L_{T_{e}}^{-2}} with L_{T_{e}} the electron temperature gradient length. A semi-empirical scaling \\smash{v_in=C_{t}({\\rho_{s}/L_{T_{e}}})^{2}D/({\\rho_{w}x})} is set up and validated against ASDEX Upgrade, DIII-D, JET and ASDEX discharges. It is shown to work in the core and edge regions of Ohmic, L- and H-mode plasmas. The ratio vin/D is independent of density, plasma current, toroidal magnetic field, hydrogenic atomic mass number, collisionality and Zeff. The anomalous inward flux is driven by the square of the electron temperature gradient. Simulations of ITER using the new vin scaling predict peaked density profiles for gas puffed scenarios because of central heating due to alpha particles.

  10. 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

  11. 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)

  12. SST and ADITYA tokamak research in India

    International Nuclear Information System (INIS)

    Steady state operation of tokamaks plays an important role in high temperature magnetically confined plasma research. Steady state Superconducting Tokamak (SST) programme in India deals with the development of various technologies in this direction. SST-1 machine has been engineered and is being fabricated at the Institute for Plasma Research. The objectives of the machine are to study physics of plasma processes under steady state condition and develop the technologies related to steady state operation. Various sub-systems are being prototyped and developed. SST-1 is a large aspect ratio machine with a major radius of 1.1 m and a plasma minor radius of 0.2 m with elongation of 1.7 to 1.9 and triangularity of 0.5 to 0.7. It has been designed for 1000 sec operation at 3 T toroidal magnetic eld. Neutral beam Injection and Radio frequency heating systems are being developed to heat the plasma. Lower hybrid Current Drive system would sustain 200 kA of plasma current during 1000 sec operation. ADITYA tokamak has been upgraded with new diagnostics and RF heating systems. Thomson Scattering and ECE diagnostics have been operated. 200 kW Ion Cyclotron Resonance Heating (ICRH) and 200 kW Electron Cyclotron Resonance Heating (ECRH) systems have been successfully commissioned. RF assisted initial breakdown experiments have been initiated with these systems. (author)

  13. 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

  14. TCS Upgrade

    Science.gov (United States)

    Grossnickle, J. A.; Miller, K. E.

    2004-11-01

    The original TCS experiment has demonstrated the robust ability to form and sustain FRCs in steady-state using Rotating Magnetic Fields (RMF). Radiation levels, which are due in large part to Oxygen, are seen to increase dramatically after the initial formation phase ( ˜0.5 msec), causing a severe drop in the plasma temperature. Since the RMF magnitude and frequency determine the plasma density, as the temperature is limited, so is the FRC's external field and energy confinement time. In order to improve temperatures and flux levels, TCS is being extensively upgraded. All o-ring sealed flanges will be replaced with wire sealed flanges, and heating blankets installed to bake the system to 200 C. Internal flux rings, shielded with Tantalum, will be installed to shield the quartz and stainless steel vacuum wall from the plasma. Unique aspects of this design are related to the interface between the quartz section needed to allow penetration of the RMF from the external antennas and the adjacent stainless steel vacuum chambers. Wall conditioning will include glow discharge, Ti gettering, siliconization, and/or boronization. The total system will be described.

  15. 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)

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. Stabilization of Neoclassical Tearing Modes in Tokamaks by Electron Cyclotron Current Drive

    Science.gov (United States)

    La Haye, R. J.

    2009-04-01

    Resistive neoclassical tearing modes (NTMs) are anticipated to be the principal limit on stability and performance in ITER as the resulting islands break up the magnetic surfaces confining the plasma. Drag from island-induced eddy currents in the resistive wall can slow plasma rotation, produce locking to the wall, and cause loss of the high-confinement H-mode and disruption. NTMs are destabilized by helical perturbations to the pressure-gradient-driven "bootstrap" current. NTMs can be stabilized by applying co-electron-cyclotron current drive (ECCD) at the island rational surface. Such stabilization and/or preemption is successful in ASDEX Upgrade, DIII-D, and JT-60U, if the peak off-axis current density is comparable to the local bootstrap current density and well-aligned. ASDEX Upgrade has used a feed-forward sweep of the toroidal field to get ECCD alignment on the island. JT-60U has used feed-forward sweeps of the launching mirror for the same purpose, followed up by real-time adjustment of the mirror using the electron cyclotron emission (ECE) diagnostic to locate the island rational surface. In DIII-D, ECCD alignment techniques include applying "search and suppress" real-time control to find and lock onto optimum alignment (adjusting the field or shifting the plasma major radius in equivalent small steps). Most experimental work to date uses narrow, cw ECCD; the relatively wide ECCD in ITER may be less effective if it is also cw: the stabilization effect of replacing the "missing" bootstrap current on the island O-point could be nearly cancelled by the destabilization effect on the island X-point if the ECCD is very broad. Modulating the ECCD so that it is absorbed only on the m/n = 3/2 rotating island O-point is proving successful in recovering ECCD effectiveness in ASDEX Upgrade when the ECCD is configured for wider deposition. The ECCD in ITER is relatively broad, with current deposition full width half maximum almost twice the marginal island width. This

  5. 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

  6. 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

  7. 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)

  8. 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.)

  9. 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

  10. [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

  11. Divertor coil power supply in Aditya Tokamak for improved plasma operation

    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. This moderate field Tokamak is capable of producing 250 kA of plasma current with 300 ms duration. Two new sets of diverter coils will be added to the system with an objective of producing double null plasmas in Aditya Upgrade Tokamak. Diverter coils, made up of continuously transposed conductor, are low voltage high current carrying poloidal field coils. One set of inner divertor coil has radius of 460 mm containing 6 turns and the other set of 1075 mm radius coil with 1 turn makes the outer divertor coils. The simulated plasma double null equilibrium demands 150 kAT of NI for the inner divertor coils and 10 - 20 kAT of NI for outer divertor coils. To energize the divertor coils with required power, a pulsed DC power supply of 3 MW (100V, 30 kA) has been designed. The designed pulsed DC power supply will be a 3-phase, 12-pulse rectifier based convertor power supply having a duty cycle of 300 ms on-time and 15 minutes off-time. The current rise time in the divertor coils will be ∼ 0.6 MA/sec. Detailed design of the divertor power supply with active controls for real time control of the plasma shape will be discussed in this paper. (author)

  12. NSTX-U Control System Upgrades

    Energy Technology Data Exchange (ETDEWEB)

    Erickson, K.G., E-mail: kerickso@pppl.gov; Gates, D.A.; Gerhardt, S.P.; Lawson, J.E.; Mozulay, R.; Sichta, P.; Tchilinguirian, G.J.

    2014-06-15

    The National Spherical Tokamak Experiment (NSTX) is undergoing a wealth of upgrades (NSTX-U). These upgrades, especially including an elongated pulse length, require broad changes to the control system that has served NSTX well. A new fiber serial Front Panel Data Port input and output (I/O) stream will supersede the aging copper parallel version. Driver support for the new I/O and cyber security concerns require updating the operating system from Redhat Enterprise Linux (RHEL) v4 to RedHawk (based on RHEL) v6. While the basic control system continues to use the General Atomics Plasma Control System (GA PCS), the effort to forward port the entire software package to run under 64-bit Linux instead of 32-bit Linux included PCS modifications subsequently shared with GA and other PCS users. Software updates focused on three key areas: (1) code modernization through coding standards (C99/C11), (2) code portability and maintainability through use of the GA PCS code generator, and (3) support of 64-bit platforms. Central to the control system upgrade is the use of a complete real time (RT) Linux platform provided by Concurrent Computer Corporation, consisting of a computer (iHawk), an operating system and drivers (RedHawk), and RT tools (NightStar). Strong vendor support coupled with an extensive RT toolset influenced this decision. The new real-time Linux platform, I/O, and software engineering will foster enhanced capability and performance for NSTX-U plasma control.

  13. FTU bolometer electronic system upgrade

    International Nuclear Information System (INIS)

    Highlights: ► Design and realization of a new bolometer electronic system. ► Many improvements over the actual commercial system. ► Architecture based on digital electronic hardware with minimal analog front end. ► Auto off-set correction, real time visualization features and small system size. ► Test results for the electronic system. -- Abstract: The FTU (Frascati Tokamak Upgrade) requires a bolometer diagnostic in order to measure the total plasma radiation. The current diagnostic architecture is based on a full analog multichannel AC bolometer system, which uses a carrier frequency amplifier with a synchronous demodulation. Taking into account the technological upgrades in the field of electronics, it was decided to realize an upgrade for the bolometric electronic system by using a hybrid analog/digital implementation. The new system developed at the ENEA Frascati laboratories has many improvements, and mainly a massive system volume reduction, a good measurement linearity and a simplified use. The new hardware system consists of two subsystems: the Bolometer Digital Control and the Bolometer Analog System. The Bolometer Digital Control can control 16 bolometer bridges through the Bolometer Analog System. The Bolometer Digital Control, based on the FPGA architecture, is connected via Ethernet with a PC; therefore, it can receive commands settings from the PC and send the stream of bolometric measurements in real time to the PC. In order to solve the cross-talk between the bridges and the cables, each of the four bridges in the bolometer head receives a different synthesized excitation frequency. Since the system is fully controlled by a PC GUI (Graphic User Interface), it is very user friendly. Moreover, some useful features have been developed, such as: auto off-set correction, bridge amplitude regulation, software gain setting, real time visualization, frequency excitation selection and noise spectrum analyzer embedded function. In this paper, the

  14. FTU bolometer electronic system upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Pollastrone, Fabio, E-mail: fabio.pollastrone@enea.it [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Neri, Carlo; Florean, Marco; Ciccone, Giovanni [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy)

    2013-10-15

    Highlights: ► Design and realization of a new bolometer electronic system. ► Many improvements over the actual commercial system. ► Architecture based on digital electronic hardware with minimal analog front end. ► Auto off-set correction, real time visualization features and small system size. ► Test results for the electronic system. -- Abstract: The FTU (Frascati Tokamak Upgrade) requires a bolometer diagnostic in order to measure the total plasma radiation. The current diagnostic architecture is based on a full analog multichannel AC bolometer system, which uses a carrier frequency amplifier with a synchronous demodulation. Taking into account the technological upgrades in the field of electronics, it was decided to realize an upgrade for the bolometric electronic system by using a hybrid analog/digital implementation. The new system developed at the ENEA Frascati laboratories has many improvements, and mainly a massive system volume reduction, a good measurement linearity and a simplified use. The new hardware system consists of two subsystems: the Bolometer Digital Control and the Bolometer Analog System. The Bolometer Digital Control can control 16 bolometer bridges through the Bolometer Analog System. The Bolometer Digital Control, based on the FPGA architecture, is connected via Ethernet with a PC; therefore, it can receive commands settings from the PC and send the stream of bolometric measurements in real time to the PC. In order to solve the cross-talk between the bridges and the cables, each of the four bridges in the bolometer head receives a different synthesized excitation frequency. Since the system is fully controlled by a PC GUI (Graphic User Interface), it is very user friendly. Moreover, some useful features have been developed, such as: auto off-set correction, bridge amplitude regulation, software gain setting, real time visualization, frequency excitation selection and noise spectrum analyzer embedded function. In this paper, the

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

    NARCIS (Netherlands)

    Box, F. M. A.; Howard, J.; VandeKolk, E.; Meijer, F. G.

    1997-01-01

    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. S

  16. The D0 Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Abachi, S.; D0 Collaboration

    1995-07-01

    In this paper we describe the approved DO Upgrade detector, and its physics capabilities. The DO Upgrade is under construction and will run during the next Fermilab collider running period in early 1999 (Run II). The upgrade is designed to work at the higher luminosities and shorter bunch spacings expected during this run. The major elements of t he upgrade are: a new tracking system with a silicon tracker, scintillating fiber tracker, a 2T solenoid, and a central preshower detector; new calorimeter electronics; new muon trigger and tracking detectors with new muon system electronics; a forward preshower detector; new trigger electronics and DAQ improvements to handle the higher rates.

  17. 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

  18. Magnetic confinement experiment -- 1: Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Goldston, R.J.

    1994-12-31

    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.

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

    DEFF Research Database (Denmark)

    Geiger, B.; Weiland, M.; Mlynek, A.;

    2015-01-01

    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...

  20. Numerical and experimental study of the redistribution of energetic and impurity ions by sawteeth in ASDEX Upgrade

    DEFF Research Database (Denmark)

    Jaulmes, F.; Geiger, B.; Odstrčil, T.;

    2016-01-01

    with tungsten impurity that include the centrifugal force are achieved and recover the soft x-ray measurements. Based on this full-reconnection description of the sawtooth, a simple tool dedicated to estimate the duration of the reconnection is introduced. This work then studies the redistribution of fast ions...

  1. Effect of resonant magnetic perturbations on low collisionality discharges in MAST and a comparison with ASDEX Upgrade

    CERN Document Server

    Kirk, A; Liu, Yueqiang; Chapman, I T; Cahyna, P; Eich, T; Fuchs, C; Ham, C; Harrison, J R; Jakubowski, M W; Pamela, S; Peterka, M; Ryan, D; Saarelma, S; Scannell, R; Thornton, A J; Valovic, M; Sieglin, B; Orte, L Barrera; Willensdorfer, M; Kurzan, B; Fischer, R; Upgrade, ASDEX

    2014-01-01

    Sustained ELM mitigation has been achieved on MAST and AUG using RMPs with a range of toroidal mode numbers over a wide region of low to medium collisionality discharges. The ELM energy loss and peak heat loads at the divertor targets have been reduced. The ELM mitigation phase is typically associated with a drop in plasma density and overall stored energy. In one particular scenario on MAST, by carefully adjusting the fuelling it has been possible to counteract the drop in density and to produce plasmas with mitigated ELMs, reduced peak divertor heat flux and with minimal degradation in pedestal height and confined energy. While the applied resonant magnetic perturbation field can be a good indicator for the onset of ELM mitigation on MAST and AUG there are some cases where this is not the case and which clearly emphasise the need to take into account the plasma response to the applied perturbations. The plasma response calculations show that the increase in ELM frequency is correlated with the size of the e...

  2. 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

  3. Upgrading Uncompetitive Products Economically

    DEFF Research Database (Denmark)

    Lu, Hua; Jensen, C.S.

    2012-01-01

    uncompetitive products to become competitive, but wants to take into account the upgrading cost. We study the top-k product upgrading problem. Given a set P of competitor products, a set T of products that are candidates for upgrade, and an upgrading cost function f that applies to T , the problem is to return...... the k products in T that can be upgraded to not be dominated by any products in P at the lowest cost. This problem is non-trivial due to not only the large data set sizes, but also to the many possibilities for upgrading a product. We identify and provide solutions for the different options...... for upgrading an uncompetitive product, and combine the solutions into a single solution. We also propose a spatial join-based solution that assumes P and T are indexed by an R-tree. Given a set of products in the same R-tree node, we derive three lower bounds on their upgrading costs. These bounds are employed...

  4. 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

  5. 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.

  6. 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.

  7. Edge turbulence in tokamaks

    Science.gov (United States)

    Nedospasov, A. V.

    1992-12-01

    Edge turbulence is of decisive importance for the distribution of particle and energy fluxes to the walls of tokamaks. Despite the availability of extensive experimental data on the turbulence properties, its nature still remains a subject for discussion. This paper contains a review of the most recent theoretical and experimental studies in the field, including mainly the studies to which Wootton (A.J. Wooton, J. Nucl. Mater. 176 & 177 (1990) 77) referred to most in his review at PSI-9 and those published later. The available theoretical models of edge turbulence with volume dissipation due to collisions fail to fully interpret the entire combination of experimental facts. In the scrape-off layer of a tokamak the dissipation prevails due to the flow of current through potential shifts near the surface of limiters of divertor plates. The different origins of turbulence at the edge and in the core plasma due to such dissipation are discussed in this paper. Recent data on the electron temperature fluctuations enabled one to evaluate the electric probe measurements of turbulent flows of particles and heat critically. The latest data on the suppression of turbulence in the case of L-H transitions are given. In doing so, the possibility of exciting current instabilities in biasing experiments (rather than only to the suppression of existing turbulence) is given some attention. Possible objectives of further studies are also discussed.

  8. Dust Measurements in Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Rudakov, D; Yu, J; Boedo, J; Hollmann, E; Krasheninnikov, S; Moyer, R; Muller, S; Yu, A; Rosenberg, M; Smirnov, R; West, W; Boivin, R; Bray, B; Brooks, N; Hyatt, A; Wong, C; Fenstermacher, M; Groth, M; Lasnier, C; McLean, A; Stangeby, P; Ratynskaia, S; Roquemore, A; Skinner, C; Solomon, W M

    2008-04-23

    Dust production and accumulation impose safety and operational concerns for ITER. Diagnostics to monitor dust levels in the plasma as well as in-vessel dust inventory are currently being tested in a few tokamaks. Dust accumulation in ITER is likely to occur in hidden areas, e.g. between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering resolves size of particles between 0.16-1.6 {micro}m in diameter; the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast-framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in 2D with a single camera or 3D using multiple cameras, but determination of particle size is problematic. In order to calibrate diagnostics and benchmark dust dynamics modeling, pre-characterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase of carbon atomic, C2 dimer, and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey 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. 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)

  11. 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.)

  12. 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

  13. Trajectory for Industrial Upgrade

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    The Ministry of Industry and Information Technology(MIIT) ordered the closure of outdated production lines in 18 industries as part of the country’s plan to upgrade its industrial structure and move up the

  14. The LHCb Muon Upgrade

    CERN Multimedia

    Cardini, A

    2013-01-01

    The LHCb collaboration is currently working on the upgrade of the experiment to allow, after 2018, an efficient data collection while running at an instantaneous luminosity of 2x10$^{33}$/cm$^{-2}$s$^{-1}$. The upgrade will allow 40 MHz detector readout, and events will be selected by means of a very flexible software-based trigger. The muon system will be upgraded in two phases. In the first phase, the off-detector readout electronics will be redesigned to allow complete event readout at 40 MHz. Also, part of the channel logical-ORs, used to reduce the total readout channel count, will be removed to reduce dead-time in critical regions. In a second phase, higher-granularity detectors will replace the ones installed in highly irradiated regions, to guarantee efficient muon system performances in the upgrade data taking conditions.

  15. LEP is upgraded

    CERN Multimedia

    1995-01-01

    A superconducting radio-frequency cavity is installed on the Large Electron-Positron (LEP) collider. This upgrade, known as LEP-2, allowed the accelerator to reach new, higher energies and so investigate new areas of physics.

  16. Trajectory for Industrial Upgrade

    Institute of Scientific and Technical Information of China (English)

    LIU YUNYUN

    2010-01-01

    @@ The Ministry of Industry and Information Technology (MIIT) ordered the closure of outdated production lines in 18 industries as part of the country's plan to upgrade its industrial structure and move up the value chain.

  17. 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...

  18. 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

  19. 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)

  20. Spherical tokamak development in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma; Barbosa, L.F.W. [Universidade do Vale do Paraiba (UNIVAP), Sao Jose dos Campos, SP (Brazil). Faculdade de Engenharia, Arquitetura e Urbanismo; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Div. de Mecanica Espacial e Controle; The high-power microwave sources group

    2003-12-01

    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 steps in the development of the ETE (Experimento Tokamak Esferico) project, its research program, technical characteristics and operating conditions as of December, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)

  1. 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.)

  2. 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

  3. 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

  4. 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.)

  5. Optics upgrade for switchyard

    Energy Technology Data Exchange (ETDEWEB)

    Kobilarcik, Thomas R.; /Fermilab

    2005-08-01

    An upgrade of the Switchyard optics is proposed. This upgrade extends the P3 (old Main Ring) lattice through enclosure C. The septa for the 3-way Meson Area split is moved from enclosure F1 to enclosure M01. The functionality of the Meson Target Train is preserved. Finally, for the purpose of demonstrating that the resulting split can be transported, a straw-man lattice is proposed for enclosure M02 and beyond.

  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. 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

  8. LHC Upgrade Scenarios

    CERN Document Server

    Zimmermann, F

    2007-01-01

    The EU CARE-HHH and US-LARP studies for an LHC luminosity upgrade aim at increasing the peak luminosity by a factor of 10, to 1035 cm-2s-1. The luminosity can be raised by rebuilding the interaction regions (IRs) in combination with a consistent change of beam parameters. In addition to advanced low-beta quadrupoles, the upgraded IRs may accommodate other new elements such as slim s.c. dipoles or quadrupoles embedded deep inside the detectors, global low-angle crab cavities, and wire compensators of long-range beam-beam effects. Important constraints on the upgrade path are the maximum acceptable number of detector pile-up events, favoring many closely spaced bunches, and the heat load on the cold-magnet beam screens, pointing towards fewer and more intense bunches. In order to translate the increased peak luminosity into a correspondingly higher integrated luminosity, the upgrade of the LHC ring should be complemented by an upgrade of the injector complex. I will present preferred upgrade scenarios for the L...

  9. Upgrades to the profile and Doppler reflectometer systems on EAST

    Science.gov (United States)

    Hu, Jian Qiang; Liu, A. Di; Doyle, Edward J.; Wang, Guiding; Li, Hong; Zhou, Chu; Zhang, Xiao Hui; Wang, Ming Yuan; Zhang, Jin; Yu, Chang Xuan

    2015-11-01

    The USTC reflectometer systems on the EAST Tokamak have been upgraded, including new Q- and V-band monostatic FMCW profile reflectometer systems with dynamic calibration, efficient transition lines with quasi-optical lenses and corrugated waveguides, dual polarization operation. The profile system is integrated with an 8-channel Doppler backscattering (DBS) system in a new flexible microwave front-end, and a second DBS system is at a separate toroidal location. The new systems cater for variable scenarios and allow for poloidal and toroidal turbulence correlations. We present the designs for these upgraded systems, system calibrations and measurements of the beam profile in laboratory, as well as the primary experimental results from EAST operation. Work supported by the Natural Science Foundation of China 11475173, National Magnetic Confinement Fusion Energy Development Program of China 2013GB106002 and 2014GB109002, US DOE Grants DE-SC0010424 and DE-SC0010469, and China Scholarship Council 3026.

  10. 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

  11. 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

  12. 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.

  13. 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.

  14. Cyclotron Institute Upgrade Project

    Energy Technology Data Exchange (ETDEWEB)

    Clark, Henry [Texas A& M University; Yennello, Sherry [Texas A& M University; Tribble, Robert [Texas A& M University

    2014-08-26

    The Cyclotron Institute at Texas A&M University has upgraded its accelerator facilities to extend research capabilities with both stable and radioactive beams. The upgrade is divided into three major tasks: (1) re-commission the K-150 (88”) cyclotron, couple it to existing beam lines to provide intense stable beams into the K-500 experimental areas and use it as a driver to produce radioactive beams; (2) develop light ion and heavy ion guides for stopping radioactive ions created with the K-150 beams; and (3) transport 1+ ions from the ion guides into a charge-breeding electron-cyclotron-resonance ion source (CB-ECR) to produce highly-charged radioactive ions for acceleration in the K-500 cyclotron. When completed, the upgraded facility will provide high-quality re-accelerated secondary beams in a unique energy range in the world.

  15. The LHCb VELO upgrade

    CERN Document Server

    Collins, P; Poikela, T; Crossley, M; Kucharczyk, M; Whitehead, M; Dumps, R; Mountain, R; Artuso, M; Rodrigues, E; Tlustos, L; Papadelis, A; Buytaert, J; Blusk, S; Parkes, C; Xing, Z; Eklund, L; Coco, V; Michel, T; Campbell, M; Bowcock, T J V; Wang, J C; Akiba, K; Gligorov, V; Huse, T; Llin, L F; Gandelman, M; Plackett, R; Esperante, D; Maneuski, D; Bayer, F; Llopart, X; Alexander, M; Gallas, A; Nichols, M; van Beuzekom, M G; John, M

    2011-01-01

    The LHCb experiment at the LHC plans to massively increase its data taking capabilities by running at a higher luminosity with a fully upgraded detector around 2016. This scheme is independent of (but compatible with) the plans for the SLHC upgrades. The silicon detector will be upgraded to provide a 40 MHz readout and to be able to cope with the increased radiation environment. This paper describes the options currently under consideration. A highlight of the R\\&D so far undertaken is a beam test during summer 2009 using the Timepix chip to track charged particles. Preliminary results are presented, including a measurement of the resolution achieved by the 55 mu m pitch pixel array of better than 9.5 mu m for perpendicular tracks and 55 mu m for angled tracks. (C) 2010 Elsevier B.V. All rights reserved.

  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. MAST-Upgrade Divertor Facility and Assessing Performance of Long-Legged Divertors

    OpenAIRE

    Fishpool, G.; Canik, J.; Cunningham, G.; Harrison, J.; Katramados, I.; Kirk, A.; Kovari, M.; H. Meyer; Scannell, R.; Team, the MAST-Upgrade

    2013-01-01

    A potentially important feature in a divertor design for a high-power tokamak is an extended and expanded divertor leg. The upgrade to MAST will allow a wide range of such divertor leg geometries to be produced, and hence will allow the roles of greatly increased connection length and flux expansion to be experimentally tested. This will include testing the potential of the Super-X configuration [1]. The design process for the upgrade has required analysis of producing and controlling the mag...

  18. MAST-Upgrade Divertor Facility and Assessing Performance of Long-Legged Divertors

    CERN Document Server

    Fishpool, G; Cunningham, G; Harrison, J; Katramados, I; Kirk, A; Kovari, M; Meyer, H; Scannell, R

    2013-01-01

    A potentially important feature in a divertor design for a high-power tokamak is an extended and expanded divertor leg. The upgrade to MAST will allow a wide range of such divertor leg geometries to be produced, and hence will allow the roles of greatly increased connection length and flux expansion to be experimentally tested. This will include testing the potential of the Super-X configuration [1]. The design process for the upgrade has required analysis of producing and controlling the magnetic configurations, and has included consideration of the roles that divertor closure and increasing magnetic connection length will play.

  19. The FNAL injector upgrade

    CERN Document Server

    Tan, C Y; Duel, K L; Lackey, J R; Pellico, W A

    2012-01-01

    The present FNAL H- injector has been operational since the 1970s and consists of two magnetron H- sources and two 750 keV Cockcroft-Walton Accelerators. In the upgrade, both slit-type magnetron sources will be replaced with circular aperture sources, and the Cockcroft-Waltons with a 200 MHz RFQ (radio frequency quadrupole). Operational experience at BNL (Brookhaven National Laboratory) has shown that the upgraded source and RFQ will be more reliable, improve beam quality and require less manpower than the present system.

  20. The LHCb VELO upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho Akiba, Kazuyoshi, E-mail: Kazuyoshi.Akiba@cern.ch

    2013-12-11

    The LHCb experiment plans to have a fully upgraded detector and data acquisition system in order to take data with instantaneous luminosities up to 5 times greater than currently. For this reason the first tracking and vertexing detector, the VELO, will be completely redesigned to be able to cope with the much larger occupancies and data acquisition rates. Two main design alternatives, micro-strips or pixel detectors, are under consideration to build the upgraded detector. This paper describes the options presently under consideration, as well as a few highlights of the main aspects of the current R and D. Preliminary results using a pixel telescope are also presented.

  1. ATLAS/CMS Upgrades

    CERN Document Server

    AUTHOR|(INSPIRE)INSPIRE-00370685; The ATLAS collaboration

    2016-01-01

    Precision studies of the Standard Model (SM) and the searches of the physics beyond the SM are ongoing at the ATLAS and CMS experiments at the Large Hadron Collider (LHC). A luminosity upgrade of LHC is planned, which provides a significant challenge for the experiments. In this report, the plans of the ATLAS and CMS upgrades are introduced. Physics prospects for selected topics, including Higgs coupling measurements, Bs,d -> mumu decays, and top quark decays through flavor changing neutral current, are also shown.

  2. 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.)

  3. 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.)

  4. 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

  5. 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.

  6. 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

  7. 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

  8. Upgrading Undergraduate Biology Education

    Science.gov (United States)

    Musante, Susan

    2011-01-01

    On many campuses throughout the country, undergraduate biology education is in serious need of an upgrade. During the past few decades, the body of biological knowledge has grown exponentially, and as a research endeavor, the practice of biology has evolved. Education research has also made great strides, revealing many new insights into how…

  9. OMEGA Upgrade preliminary design

    International Nuclear Information System (INIS)

    The OMEGA laser system at the Laboratory for Laser Energetics of the University of Rochester is the only major facility in the United States capable of conducting fully diagnosed, direct-drive, spherical implosion experiments. As such, it serves as the national Laser Users Facility, benefiting scientists throughout the country. The University's participation in the National Inertial Confinement Fusion (ICF) program underwent review by a group of experts under the auspices of the National Academy of Sciences (the Happer Committee) in 1985. The Happer Committee recommended that the OMEGA laser be upgraded in energy to 30 kJ. To this end, Congress appropriated $4,000,000 for the preliminary design of the OMEGA Upgrade, spread across FY88 and FY89. This document describes the preliminary design of the OMEGA Upgrade. The proposed enhancements to the existing OMEGA facility will result in a 30-kHJ, 351-nm, 60-beam direct-drive system, with a versatile pulse-shaping facility and a 1%--2% uniformity of target drive. The Upgrade will allow scientists to explore the ignition-scaling regime, and to study target behavior that is hydrodynamically equivalent to that of targets appropriate for a laboratory microfusion facility (LMF). In addition, it will be possible to perform critical interaction experiments with large-scale-length uniformly irradiated plasmas

  10. Upgrade of telephone exchange

    CERN Multimedia

    2006-01-01

    As part of the upgrade of telephone services, work will be carried out on the CERN switching centre between Monday 23 October 8.00 p.m. and Tuesday 24 October 2.00 a.m. Telephone services may be disrupted and possibly even interrupted during this operation. We apologise in advance for any inconvenience this may cause. CERN TELECOM Service

  11. Lightweight incremental application upgrade

    NARCIS (Netherlands)

    Storm, T. van der

    2006-01-01

    I present a lightweight approach to incremental application upgrade in the context of component-based software development. The approach can be used to efficiently implement an automated update feature in a platform and programming language agnostic way. A formal release model is presented which ens

  12. 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.)

  13. DIII-D UPGRADE PROJECT FINAL REPORT FOR THE PERIOD OCTOBER 1, 1993 THROUGH MAY 31, 2003

    International Nuclear Information System (INIS)

    OAK-B135 Under DOE Contracts DE-AC03-89ER51114 and DE-AC03-99ER54463 to General Atomics (GA), three ''capital project'' upgrade projects were accomplished on DIII-D from FY93 to FY03 at a total GA cost of $27.2M. These projects included the Fast Wave Current Drive (FWCD) Upgrade ($8.2M), the Radiative Divertor Upgrade ($7.2M) and the Electron Cyclotron Heating (ECH) Upgrade ($11.8M). The ECH and FWCD upgrades provided DIII-D rf and microwave power for electron heating, driving plasma current, controlling the plasma current profile, controlling tearing mode instabilities, and modulated transport studies.The divertor provided adequate density and impurity control for high triangularity single null plasmas in the Advanced Tokamak (AT) Program and information for International Thermonuclear Experimental Reactor (ITER) divertor design. These upgrades provide the power and density control required to initiate the active control of advanced tokamak discharges, which is the key element in the DIII-D program

  14. The LHCb Upgrade

    CERN Document Server

    Alessio, Federico

    2013-01-01

    The LHCb experiment is designed to perform high-precision measurements of CP violation and search for New Physics using the enormous flux involving beauty and charm quarks produced at the LHC. The operation and the results obtained from the data collected in 2010 and 2011 demonstrate that the detector is robust and functioning very well. However, the limit of 1 fb^-1 of data per nominal year cannot be overcome without improving the detector. We therefore plan for an upgraded spectrometer by 2018 with a 40 MHz readout and a much more flexible software-based triggering system that will increase the data rate as well as the efficiency specially in the hadronic channels. Here we present the LHCb detector upgrade plans, based on the Letter of Intent and Framework Technical Design Report.

  15. A Neutral Beam Injector Upgrade for NSTX

    Energy Technology Data Exchange (ETDEWEB)

    T. Stevenson; B McCormack; G.D. Loesser; M. Kalish; S. Ramakrishnan; L. Grisham; J. Edwards; M. Cropper; G. Rossi; A. von Halle; M. Williams

    2002-01-18

    The National Spherical Torus Experiment (NSTX) capability with a Neutral Beam Injector (NBI) capable of 80 kiloelectronvolt (keV), 5 Megawatt (MW), 5 second operation. This 5.95 million dollar upgrade reused a previous generation injector and equipment for technical, cost, and schedule reasons to obtain these specifications while retaining a legacy capability of 120 keV neutral particle beam delivery for shorter pulse lengths for possible future NSTX experiments. Concerns with NBI injection included power deposition in the plasma, aiming angles from the fixed NBI fan array, density profiles and beam shine through, orbit losses of beam particles, and protection of the vacuum vessel wall against beam impingement. The upgrade made use of the beamline and cryo panels from the Neutral Beam Test Stand facility, existing power supplies and controls, beamline components and equipment not contaminated by tritium during DT [deuterium-tritium] experiments, and a liquid Helium refrigerator plant to power and cryogenically pump a beamline and three ion sources. All of the Tokamak Fusion Test Reactor (TFTR) ion sources had been contaminated with tritium, so a refurbishment effort was undertaken on selected TFTR sources to rid the three sources destined for the NSTX NBI of as much tritium as possible. An interconnecting duct was fabricated using some spare and some new components to attach the beamline to the NSTX vacuum vessel. Internal vacuum vessel armor using carbon tiles was added to protect the stainless steel vacuum vessel from beam impingement in the absence of plasma and interlock failure. To date, the NBI has operated to 80 keV and 5 MW and has injected requested power levels into NSTX plasmas with good initial results, including high beta and strong heating characteristics at full rated plasma current.

  16. CMS upgrades for SLHC

    CERN Document Server

    Palla, Fabrizio

    2006-01-01

    I will discuss the impact of the LHC luminosity upgrade on CMS detector. While most of the CMS can possibly cope with the increased luminosity, the Tracker must undergo a major redesign in technology both in terms of detector substrates as well as in the data transfer links. I will show the impact on CMS of reduced bunch length and machine elements close to the interaction point.

  17. The VELO Upgrade

    CERN Document Server

    Jans, E

    2015-01-01

    A significant upgrade of the LHCb detector is scheduled to be installed in 2018-2019. Afterwards all sub-detectors will be read out at the LHC bunch crossing frequency of 40 MHz and the trigger will be fully implemented in software. The silicon strip vertex detector will be replaced by a hybrid pixel detector. In these proceedings the following items are discussed: frontend ASIC, data rates, data transmission, cooling, radiation hard sensors, module design and simulated performance.

  18. ATLAS/CMS Upgrades

    CERN Document Server

    Horii, Yasuyuki; The ATLAS collaboration

    2016-01-01

    Precise Higgs measurements and new physics searches are planned at LHC (HL-LHC) with integrated luminosity of 300 fb^{-1} (3000 fb^{-1}). An increased peak luminosity provides a significant challenge for the experiments. In this presentation, the plans for the ATLAS and CMS upgrades are introduced. Physics prospects for some topics related with ‘flavour’, e.g Higgs couplings, B_{s, d}->mumu, and FCNC top decays, are also shown.

  19. Optimizing pyrolysis gasoline upgrading

    Energy Technology Data Exchange (ETDEWEB)

    Coupard, V.; Cosyns, J.; Debuisschert, Q.; Travers, Ph. [Axens (France). Kinetics and Catalysis Div.

    2002-06-01

    Stringent environmental regulations for European Gasoline will mean decrease in Pygas in Gasoline pool. Pygas upgrading routes have been developed to produce added value products such as dicyclopentadiene, cyclopentane, improved olefin cracking stocks and desulfurized aromatic streams. Examples will be presented with Economics. New generation Nickel/Palladium catalysts in the 1{sup st} stage Pygas hydrogenation units will be discussed related to increasing capacity and service life. (orig.)

  20. The LHCb VELO upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Rodríguez Pérez, Pablo, E-mail: pablo.rodriguez@usc.es

    2013-12-01

    LHCb is a forward spectrometer experiment dedicated to the study of new physics in the decays of beauty and charm hadrons produced in proton collisions at the Large Hadron Collider (LHC) at CERN. The VErtex LOcator (VELO) is the microstrip silicon detector surrounding the interaction point, providing tracking and vertexing measurements. The upgrade of the LHCb experiment, planned for 2018, will increase the luminosity up to 2×10{sup 33} cm{sup −2} s{sup −1} and will perform the readout as a trigger-less system with an event rate of 40 MHz. Extremely non-uniform radiation doses will reach up to 5×10{sup 15} 1 MeV n{sub eq}/cm{sup 2} in the innermost regions of the VELO sensors, and the output data bandwidth will be increased by a factor of 40. An upgraded detector is under development based in a pixel sensor of the Timepix/Medipix family, with 55×55μm{sup 2} pixels. In addition a microstrip solution with finer pitch, higher granularity and thinner than the current detector is being developed in parallel. The current status of the VELO upgrade program will be described together with recent testbeam results.

  1. The LHCb VELO upgrade

    Science.gov (United States)

    Dosil Suárez, Álvaro

    2016-07-01

    The upgrade of the LHCb experiment, planned for 2019, will transform the experiment to a trigger-less system reading out the full detector at 40 MHz event rate. All data reduction algorithms will be executed in a high-level software farm. The upgraded detector will run at luminosities of 2×1033 cm-2 s-1 and probe physics beyond the Standard Model in the heavy flavour sector with unprecedented precision. The Vertex Locator (VELO) is the silicon vertex detector surrounding the interaction region. The current detector will be replaced with a hybrid pixel system equipped with electronics capable of reading out at 40 MHz. The detector comprises silicon pixel sensors with 55×55 μm2 pitch, read out by the VeloPix ASIC, based on the TimePix/MediPix family. The hottest region will have pixel hit rates of 900 Mhits/s yielding a total data rate more than 3 Tbit/s for the upgraded VELO. The detector modules are located in a separate vacuum, separated from the beam vacuum by a thin custom made foil. The detector halves are retracted when the beams are injected and closed at stable beams, positioning the first sensitive pixel at 5.1 mm from the beams. The material budget will be minimised by the use of evaporative CO2 coolant circulating in microchannels within 400 μm thick silicon substrates.

  2. LHCb VELO Upgrade

    CERN Document Server

    Hennessy, Karol

    2016-01-01

    The upgrade of the LHCb experiment, scheduled for LHC Run-III, scheduled to start in 2021, will transform the experiment to a trigger-less system reading out the full detector at 40 MHz event rate. All data reduction algorithms will be executed in a high-level software farm enabling the detector to run at luminosities of $2\\times10^{33} \\mathrm{cm}^{-2}\\mathrm{s}^{-1}$. The Vertex Locator (VELO) is the silicon vertex detector surrounding the interaction region. The current detector will be replaced with a hybrid pixel system equipped with electronics capable of reading out at 40 MHz. The upgraded VELO will provide fast pattern recognition and track reconstruction to the software trigger. The silicon pixel sensors have $55\\times55 \\mu m^{2}$ pitch, and are read out by the VeloPix ASIC, from the Timepix/Medipix family. The hottest region will have pixel hit rates of 900 Mhits/s yielding a total data rate of more than 3 Tbit/s for the upgraded VELO. The detector modules are located in a separate vacuum, separate...

  3. Engineering design for the HL-2M tokamak components

    International Nuclear Information System (INIS)

    Highlights: • The HL-2M is upgraded from HL-2A tokamak which is being constructed in SWIP. • The plasma major radius is 1.78 m, the minor radius is 0.65 m, and the plasma current will reach up to 3 MA with the elongation 1.6–1.8 and tri-angularity more than 0.5. • The new machine has been designed with demountable toroidal field coils and the vacuum vessel is an all-welded torus. • The coils, vacuum vessel and support structure are under fabricated in factories, and the construction of device will be finished in the end of next year. -- Abstract: The HL-2A tokamak will be modified into HL-2M. The Bt at the plasma center (major radius R = 1.78 m) is 2.2 T, the minor radius is 0.65 m. The plasma current IP of HL-2M will reach up to 2.5 MA, the elongation and triangularity is more than 1.8 and more than 0.5, respectively. The vacuum vessel torus consists of 20 sectors with “D” shaped cross-section and double wall structure. 20 toroidal field coil bundles comprise 140 turns which are designed with demountable joints, the poloidal field coils system consists of 25 coils. The engineering design and calculation for field coil system, vacuum vessel, support structure, etc. are finished, many key issues for manufacture process have been discussed with industry and the fabrication of main components of HL-2M tokamak will be carried out in factories

  4. 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.

  5. Big Dee upgrade of the Doublet III diagnostic data acquisition computer system

    Energy Technology Data Exchange (ETDEWEB)

    McHarg, B.B. Jr.

    1983-12-01

    The Big Dee upgrade of the Doublet III tokamak facility will begin operation in 1986 with an initial quantity of data expected to be 10 megabytes per shot and eventually attaining 20 to 25 megabytes per shot. This is in comparison to the 4 to 5 megabytes of data currently acquired. To handle this greater quantity of data and to serve physics needs for significantly improved between-shot processing of data will require a substantial upgrade of the existing data acquisition system. The key points of the philosophy that have been adopted for the upgraded system to handle the greater quantity of data are (1) preserve existing hardware; (2) preserve existing software; (3) configure the system in a modular fashion; and (4) distribute the data acquisition over multiple computers. The existing system using ModComp CLASSIC 16 bit minicomputers is capable of handling 5 megabytes of data per shot.

  6. Upgrades for the ECH System on DIII-D

    Science.gov (United States)

    Lohr, J.; Cengher, M.; Doane, J. L.; Gorelov, Y. A.; Moeller, C. P.; Ponce, D.; Noraky, S.; Penaflor, B. G.; Kolemen, E.

    2012-10-01

    The gyrotron system for electron cyclotron heating on the DIII-D tokamak is being upgraded with the addition of higher efficiency gyrotrons having collector potential depression. Two new gyrotrons, operating at the present frequency of 110 GHz and generating 1.2 MW per unit, have been manufactured and are being installed and tested. The subsequent group of gyrotrons have been designed to generate 1.5 MW for 10 s pulses at 117.5 GHz. The first of these tubes is presently being manufactured at Communications and Power Industries. By the end of 2013, the system will comprise eight high power gyrotrons and, pending the successful performance of the 1.5 MW tube, an upgrade to a 15 MW system will begin. High voltage power supplies, transmission lines, launchers and associated control and data acquisition systems are included in the upgrades as is enhanced ability to steer the rf beams under a variety of pre-programed and reactive scenarios.

  7. 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.

  8. Options for an ignited tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Sheffield, J.

    1984-02-01

    It is expected that the next phase of the fusion program will involve a tokamak with the goals of providing an ignited plasma for pulses of hundreds of seconds. A simple model is described in this memorandum which establishes the physics conditions for such a self-sustaining plasma, for given ion and electron thermal diffusivities, in terms of R/a, b/a, I, B/q, epsilon ..beta../sub p/, anti T/sub i/, and anti T/sub e//anti T/sub i/. The model is used to produce plots showing the wide range of tokamaks that may ignite or have a given ignition margin. The constraints that limit this range are discussed.

  9. A compact Tokamak transmutation reactor

    Institute of Scientific and Technical Information of China (English)

    QiuLi-Jian; XiaoBing-Jia

    1997-01-01

    The low aspect ration tokamak is proposed for the driver of a transmutation reactor.The main parameters of the reactor core,neutronic analysis of the blanket are given>the neutron wall loading can be lowered from the magnitude order of 1 MW/m2 to 0.5MW/m2 which is much easier to reach in the near future,and the transmutation efficiency (fission/absorption ratio)is raised further.The blanket power density is about 200MW/m3 which is not difficult to deal with.The key components such as diverter and center conductor post are also designed and compared with conventional TOkamak,Finally,by comparison with the other drivers such as FBR,PWR and accelerator,it can be anticipated that the low aspect ratio transmutation reactor would be one way of fusion energy applications in the near future.

  10. 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.

  11. LHCb Upgrade: Upstream Tracker

    CERN Document Server

    AUTHOR|(CDS)2082337

    2015-01-01

    The upgraded LHCb detector will run at an instantaneous luminosity of 2 X 10$^{33}$ cm$^{-2}$ s$^{-1}$, five times higher than in the current configuration, and will have a full 40 MHz readout. In order to cope with these higher instantaneous rates, the tracking detector upstream of the LHCb dipole magnet, called Tracker Turicensis (TT) [1], will be replaced by the Upstream Tracker (UT) [2]. The conceptual design of the UT and the current status of the R&D are presented here.

  12. Upgrading of the tandem

    International Nuclear Information System (INIS)

    The program of the tandem-linac accelerator system is summarized under the following headings: operating experience for the tandem, operation of the superconducting linac, upgrading of the tandem (ion sources, vacuum systems, terminal box, stripping foils, beam bunching), installation of the booster, planned accelerator system improvements, experimental facilities development at the super conducting-linac booster (new beam line, layout and installation of the 00 beam line in the new experiment area, beam optics calculations, 65-in. scattering chamber, split-pole spectrograph, sum/multiplicity detector, nuclear target making and development), and university use of the tandem accelerator

  13. 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

  14. Magnetic confinement experiment. I: Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Goldston, R.J.

    1995-08-01

    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 nT{tau}`s {approximately} 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.

  15. 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

  16. First Results from the Lithium Tokamak eXperiment

    International Nuclear Information System (INIS)

    Full text: The Lithium Tokamak eXperiment (LTX) is a newly commissioned, modest-scale spherical tokamak with R =0.4 m, a =0.26 m, and elongation of 1.5. Design targets are a toroidal field of 3.2 kG, plasma current up to 400 kA, and a discharge duration of order 100 msec. LTX is the first tokamak designed to investigate modifications to equilibrium and transport when global recycling is reduced to 10 - 20 %. To reduce recycling, LTX is fitted with a 1 cm thick heated (300 deg. C) copper shell, conformal to the last closed flux surface, over 85% of the plasma surface area. The plasma-facing surface of the shell will be evaporatively coated with a thin (< 100 micron) layer of molten lithium, retained by surface tension. The shell is replaceable, and a second version has been constructed, which was plasma-sprayed with 100 - 200 microns of molybdenum to form a high-Z substrate for subsequent coating with lithium. After the installation of the second shell (in 2011), a high temperature (500 - 600 deg. C) operating phase for LTX is planned. LTX is the first tokamak designed to operate with a full hot high-Z wall, near the projected operating temperature for reactor PFCs. The engineering design and construction of the hot high-Z shell, as well as the vessel and diagnostics to tolerate both lithium and 500 deg. C internal components, will be discussed. LTX will employ short-pulse fueling with a new hydrogen molecular cluster injector, to transiently eliminate edge gas (between puffs). This fueling system will be briefly discussed. Diagnostics include single-pulse multipoint Thomson scattering, Lyman alpha arrays, microwave interferometers, spectrometers, and an edge Langmuir probe. LTX is now progressing through the shakedown phase, and first operation with a liquid lithium film wall is scheduled for Spring 2010. Later in 2010, a new diagnostic (Digital Holography) for core density variations will be tested on LTX. In 2011 a 5 A, 20 kV, 1 second hydrogen neutral beam

  17. The Fermilab Linac Upgrade

    International Nuclear Information System (INIS)

    The Fermilab Linac Upgrade is planned to increase the energy of the H- linac from 200 to 400 MeV. This is intended to reduce the incoherent space-charge tuneshift at injection into the 8 GeV Booster which can limit either the brightness or the total intensity of the beam. The Linac Upgrade will be achieved by replacing the last four 201.25 MHz drift-tube tanks which accelerate the beam from 116 to 200 MeV, with seven 805 MHz side-coupled cavity modules operating at an average axial field of abut 7.5 MV/m. This will allow acceleration to 400 MeV in the existing Linac enclosure. Each accelerator module will be driven with a klystron-based rf power supply. A prototype rf modulator has been built and tested at Fermilab, and a prototype 12 MW klystron is being fabricated by Litton Electron Devices. Fabrication of production accelerator modules is in progress. 8 figs., 4 tabs

  18. The Fermilab Linac Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Noble, R.J.

    1991-02-01

    The Fermilab Linac Upgrade is planned to increase the energy of the H- linac from 200 to 400 MeV. This is intended to reduce the incoherent space-charge tuneshift at injection into the 8 GeV Booster which can limit either the brightness or the total intensity of the beam. The Linac Upgrade will be achieved by replacing the last four 201.25 MHz drift-tube tanks which accelerate the beam from 116 to 200 MeV, with seven 805 MHz side-coupled cavity modules operating at an average axial field of abut 7.5 MV/m. This will allow acceleration to 400 MeV in the existing Linac enclosure. Each accelerator module will be driven with a klystron-based rf power supply. A prototype rf modulator has been built and tested at Fermilab, and a prototype 12 MW klystron is being fabricated by Litton Electron Devices. Fabrication of production accelerator modules is in progress. 8 figs., 4 tabs.

  19. LHCb VELO Upgrade

    CERN Multimedia

    van Beuzekom, Martin; Ketel, Tjeerd; Gershon, Timothy; Parkes, Christopher; Reid, Matthew

    2011-01-01

    The VErtex LOcator (VELO) is a vital piece of apparatus for allowing precision measurements in hadronic physics. It provides not only superb impact parameter resolutions but also excellent momentum resolution, both important discriminating tools for precision high energy physics. This poster focuses on the R&D going into the future LHCb VELO detector. At present there are two proposed options for the upgrade; pixel chips or strip detectors. The LHCb upgrade is designed with higher luminosities and increased yields in mind. In order to get more out of the LHCb detector changes to the front end electronics will have to be made. At present, the first level hardware trigger is sets a limiting factor on the maximum efficiency for hadronic channels. As the VELO is positioned so close the proton-proton interaction region, whatever the choice of sensor, we will require efficient cooling and some proposed solutions are outlined. The LHCb TimePix telescope has had a very successful years running, with various devic...

  20. ATLAS Detector Upgrade Prospects

    CERN Document Server

    Dobre, Monica; The ATLAS collaboration

    2016-01-01

    After the successful operation at the center-of-mass energies of 7 and 8 TeV in 2010 - 2012, the LHC is ramped up and successfully took data at the center-of-mass energies of 13 TeV in 2015. Meanwhile, plans are actively advancing for a series of upgrades of the accelerator, culminating roughly ten years from now in the high-luminosity LHC (HL-LHC) project, delivering of the order of five times the LHC nominal instantaneous luminosity along with luminosity leveling. The ultimate goal is to extend the dataset from about few hundred fb−1 expected for LHC running to 3000 fb−1 by around 2035 for ATLAS and CMS. The challenge of coping with the HL-LHC instantaneous and integrated luminosity, along with the associated radiation levels, requires further major changes to the ATLAS detector. The designs are developing rapidly for a new all-silicon tracker, significant upgrades of the calorimeter and muon systems, as well as improved triggers and data acquisition. ATLAS is also examining potential benefits of extens...

  1. ATLAS Strip Upgrade

    CERN Document Server

    Bernabeu, J; The ATLAS collaboration

    2012-01-01

    A phased upgrade of the Large Hadron Collider (LHC) at CERN is planned. The last upgrade phase (HL-LHC) is currently foreseen in 2022-2023. It aims to increase the integrated luminosity to about ten times the original LHC design luminosity. To cope with the harsh conditions in terms of particle rates and radiation dose expected during HL-LHC operation, the ATLAS collaboration is developing technologies for a complete tracker replacement. This new detector will need to provide extreme radiation hardness and a high granularity, within the tight constraints imposed by the existing detectors and their services. An all-silicon high granularity tracking detector is proposed. An international R&D collaboration is working on the strip layers for this new tracker. A number of large area prototype planar detectors produced on p-type wafers have been designed and fabricated for use at HL-LHC. These prototype detectors and miniature test detectors have been irradiated to a set of fluences matched to HL-LHC expectatio...

  2. ATLAS Strip Upgrade

    CERN Document Server

    Bernabeu, J; The ATLAS collaboration

    2012-01-01

    A phased upgrade of the Large Hadron Collider (LHC) at CERN is planned. The last upgrade phase (HL-LHC) is currently foreseen in 2022-2023. It aims to increase the integrated luminosity to about ten times the original LHC design luminosity. To cope with the harsh conditions in terms of particle rates and radiation dose expected during HL-LHC operation, the ATLAS collaboration is developing technologies for a complete tracker replacement. This new detector will need to provide extreme radiation hardness and a high granularity, within the tight constraints imposed by the existing detectors and their services. An all-silicon high-granularity tracking detector is proposed. An international R&D collaboration is working on the strip layers for this new tracker. A number of large area prototype planar detectors produced on p-type wafers have been designed and fabricated for use at HL-LHC. These prototype detectors and miniature test detectors have been irradiated to a set of fluences matched to HL-LHC expectatio...

  3. 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

  4. 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

  5. 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

  6. 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

  7. The role of limiter in Egyptor Tokamak

    CERN Document Server

    Ei-Sisi, A B

    2002-01-01

    In Egyptor Tokamak, the limiter is used for separation of the plasma from the vessel. In this work an overview of limiter types, and construction of limiter in Egyptor Tokamak is discussed. Also simulation results of the radial electron density distribution in case of limiter are presented. The results of the simulation are in agreement with the experimental and analytical results.

  8. Simulations of edge and scrape off layer turbulence in mega ampere spherical tokamak plasmas

    DEFF Research Database (Denmark)

    Militello, F; Fundamenski, W; Naulin, Volker;

    2012-01-01

    The L-mode interchange turbulence in the edge and scrape-off-layer (SOL) of the tight aspect ratio tokamak MAST is investigated numerically. The dynamics of the boundary plasma are studied using the 2D drift-fluid code ESEL, which has previously shown good agreement with large aspect ratio machines...... of the edge/SOL density and temperature. In addition, we also discuss how the system changes when the length of the divertor leg is modified. This allows one to better understand the regime of operation of the Super-X divertor which will be implemented on MAST-Upgrade. The results obtained qualitatively agree...

  9. Enhanced Energy Confinement and Performance in a Low-Recycling Tokamak

    International Nuclear Information System (INIS)

    Extensive lithium wall coatings and liquid lithium plasma-limiting surfaces reduce recycling, with dramatic improvements in Ohmic plasma discharges in the Current Drive Experiment-Upgrade. Global energy confinement times increase by up to 6 times. These results exceed confinement scalings such as ITER98P(y,1) by 2-3 times, and represent the largest increase in energy confinement ever observed for an Ohmic tokamak plasma. Measurements of Dα emission indicate that global recycling coefficients decrease to approximately 0.3, the lowest documented for a magnetically confined hydrogen plasma

  10. Performance of V-4Cr-4Ti material exposed to the DIII-D tokamak environment

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Smith, D.L.; Chung, H.M. [Argonne National Lab., IL (United States); Johnson, W.R.; Smith, J.P. [General Atomics, San Diego, CA (United States); Wampler, W.R. [Sandia National Labs. (United States)

    1998-03-01

    A series of tests is being conducted in the DIII-D tokamak to determine the effects of environmental exposure on a V-4Cr-4Ti vanadium alloy. These tests are part of the effort to build and install a water-cooled vanadium V alloy structure in the upgrade of the DIII-D radiative divertor. Data from the test series indicate that the performance of the V-4Cr-4Ti alloy would not be significantly affected by environmental exposure. Interstitial absorption by the material appears to be limited to the surface, and neither the tensile nor the impact properties of the material appear to be affected by the exposure.

  11. Upgrade of JT-60 pellet injector for higher velocity

    International Nuclear Information System (INIS)

    Pellet injection experiments have been performed to improve the plasma performance by the JT-60 tokamak from June, 1988. From the results of the experiments, it was found that the plasma confinement time increased up to 40% with pellet injection (velocity over 1.5 km/s), in which was obtained with 10 MW neutral beam injection highly peaked electron density profile. The experimental results suggested that improvement of the plasma confinement time depends on the penetration depth of the pellet into the plasma column, especially into 'q2 to 100 kg/cm2 and from 80degC to 200degC respectively. The upgraded pellet injector can inject, independently, four pellets, two of which are 3.0 mm in diameter x 3.0 mm in length and the other two of which are 4.0 mm in diameter x 4.0 mm in length. (author)

  12. The STAR Tracking Upgrade

    CERN Document Server

    Simon, Frank

    2007-01-01

    The STAR experiment at the Relativistic Heavy Ion Collider RHIC studies the new state of matter produced in relativistic heavy ion collisions and the spin structure of the nucleon in collisions of polarized protons. In order to improve the capabilities for heavy flavor measurements and the reconstruction of charged vector bosons an upgrade of the tracking system both in the central and the forward region is pursued. The challenging environments of high track multiplicity in heavy ion collisions and of high luminosity in polarized proton collisions require the use of new technologies. The proposed inner tracking system, optimized for heavy flavor identification, is using active pixel sensors close to the collision point and silicon strip technology further outward. Charge sign determination for electrons and positrons from the decay of W bosons will be provide by 6 large-area triple GEM disks currently under development. A prototype of the active pixel detectors has been tested in the STAR experiment, and an e...

  13. 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

  14. The RHIC polarized source upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Zelenski, A.; Atoian, G.; Davydenko, V.; Ivanov, A.; Kolmogorov, A.; Ritter, J.; Steski, D.; Zubets, V.

    2010-09-27

    The RHIC polarized H{sup -} ion source is being upgraded to higher intensity (5-10 mA) and polarization for use in the RHIC polarization physics program at enhanced luminosity RHIC operation. The higher beam peak intensity will allow reduction of the transverse beam emittance at injection to AGS to reduce polarization losses in AGS. There is also a planned RHIC luminosity upgrade by using the electron beam lens to compensate the beam-beam interaction at collision points. This upgrade is also essential for future BNL plans for a high-luminosity electron - proton (ion) Collider eRHIC.

  15. 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.)

  16. 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.

  17. Altair performance and upgrades

    Science.gov (United States)

    Lai, Olivier; Véran, Jean-Pierre; Herriot, Glen; White, John; Ball, Jesse; Trujillo, Chad

    2014-07-01

    Altair is the facility single conjugate AO system for Gemini North. Although it has been in operation for more than 10 years (and upgraded to LGS in 2007), Altair's performance is degraded by three main issues: vibrations of the telescope and instrument support structure, spatial aliasing on centroid offsets from the M2 support structure print-through on the optical surface and static non-common path aberrations. Monte-Carlo simulations can reproduce the behavior of Altair when including these three effects and they are roughly of the same order of magnitude. Solutions or mitigations are being investigated to overcome these nefarious effects and restore Altair's performance to its nominal level. A simplex algorithm as well as a phase diversity approach are being investigated to measure and correct for static aberrations. A high accuracy phase map of the M2 print-through has been obtained and is being used to calibrate and/or filter centroids affected by aliasing. A new real time computer is under consideration, to be able to handle more advanced controllers, especially notch filters to combat vibrations. In this paper we will report on the various simulations and on-sky results of this rejuvenation of one of Gemini's workhorse instruments.

  18. ATLAS Future Upgrade

    CERN Document Server

    Vankov, Peter; The ATLAS collaboration

    2016-01-01

    After the successful operation at the center-of-mass energies of 7 and 8 TeV in 2010 - 2012, the LHC is ramped up and successfully took data at the center-of-mass energies of 13 TeV in 2015. Meanwhile, plans are actively advancing for a series of upgrades of the accelerator, culminating roughly ten years from now in the high-luminosity LHC (HL-LHC) project, delivering of the order of five times the LHC nominal instantaneous luminosity along with luminosity leveling. The ultimate goal is to extend the dataset from about few hundred fb−1 expected for LHC running to 3000 fb−1 by around 2035 for ATLAS and CMS. In parallel, the experiments need to be keep lockstep with the accelerator to accommodate running beyond the nominal luminosity this decade. Along with maintenance and consolidation of the detector in the past few years, ATLAS has added inner b-layer to its tracking system. The challenge of coping with the HL-LHC instantaneous and integrated luminosity, along with the associated radiation levels, requir...

  19. Superhilac upgrade project

    Energy Technology Data Exchange (ETDEWEB)

    Feinberg, B.; Brown, I.G.

    1985-05-01

    This project will increase the uranium output of the Bevalac heavy-ion facility from the currently available 10/sup 7/ to 5 x 10/sup 7/ ions/pulse, allowing accurate Lamb shift measurements to be made in U/sup 90 +/ and U/sup 91 +/ with important applications to the testing of quantum electrodynamics and the development of an x-ray laser. The injected beam intensity will be increased to make better use of the 10emA output space-charge limit of the Wideroe linac. Components will include a new high current MEtal Vapor Vacuum Arc (MEVVA) ion source along with an improved high current, high voltage Cockcroft-Walton power supply to handle the increased beam current. The Low Energy Beam Transport (LEBT) line will be upgraded with additional focusing to manage the increased space-charge forces and with an improved vacuum to reduce charge exchange losses. Finally, the phase matching between the 23MHz Wideroe linac and the 70MHz Alvarez linac will be improved by the addition of the appropriate buncher cavities. Physics design is underway and detailed engineering is scheduled to begin in October 1985, with installation slated for the 1986 summer shutdown.

  20. 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)

  1. IPP annual report 1981

    International Nuclear Information System (INIS)

    In part A of this annual report the tokamak and stellarator projects at the IPP are reported: ASDEX, ASDEX upgrade, JET collaboration, NET collaboration, Wendelstein VII-7, Wendelstein VII-AS, Wendelstein VII-X and stellarator reactor system studies. In part B the departments and research groups give a brief, but detailed report of the results in the field of research and development. In part C a review is presented of the IPP organisation. Part D includes the papers and conference reports published in 1981. Finally a brief description of the IPP projects at German universities is presented. (GG)

  2. Recent results from the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Petersen, P.I.

    1998-02-01

    The DIII-D national fusion research program focuses on establishing the scientific basis for optimization of the tokamak approach to fusion energy production. The symbiotic development of research, theory, and hardware continues to fuel the success of the DIII-D program. During the last year, a radiative divertor and a second cryopump were installed in the DIII-D vacuum vessel, an array of central and boundary diagnostics were added, and more sophisticated computer models were developed. These new tools have led to substantial progress in the understanding of the plasma. The authors now have a better understanding of the divertor as a means to manage the heat, particle, and impurity transport pumping of the plasma edge using the in situ divertor cryopumps effectively controls the plasma density. The evolution of diagnostics that probe the interior of the plasma, particularly the motional Stark effect diagnostic, has led to a better understanding of the core of the plasma. This understanding, together with tools to control the profiles, including electron cyclotron waves, pellet injection, and neutral beam injection, has allowed them to progress in making plasma configurations that give rise to both low energy transport and improved stability. Most significant here is the use of transport barriers to improve ion confinement to neoclassical values. Commissioning of the first high power (890 kW) 110 GHz gyrotron validates an important tool for managing the plasma current profile, key to maintaining the transport barriers. An upgraded plasma control system, ``isoflux control,`` which exploits real time MHD equilibrium calculations to determine magnetic flux at specified locations within the tokamak vessel and provides the means for precisely controlling the plasma shape and, in conjunction with other heating and fueling systems, internal profiles.

  3. 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.

  4. 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.

  5. Analysis of NSTX Upgrade OH Magnet and Center Stack

    Energy Technology Data Exchange (ETDEWEB)

    A. Zolfaghari, P. Titus, J. Chrzanowski, A. Salehzadeh, F. Dahlgren

    2010-11-30

    The new ohmic heating (OH) coil and center stack for the National Spherical Torus Experiment (NSTX) upgrade are required to meet cooling and structural requirements for operation at the enhanced 1 Tesla toroidal field and 2 MA plasma current. The OH coil is designed to be cooled in the time between discharges by water flowing in the center of the coil conductor. We performed resistive heating and thermal hydraulic analyses to optimize coolant channel size to keep the coil temperature below 100 C and meet the required 20 minute cooling time. Coupled electromagnetic, thermal and structural FEA analyses were performed to determine if the OH coil meets the requirements of the structural design criteria. Structural response of the OH coil to its self-field and the field from other coils was analyzed. A model was developed to analyze the thermal and electromagnetic interaction of centerstack components such as the OH coil, TF inner legs and the Bellville washer preload mechanism. Torsional loads from the TF interaction with the OH and poloidal fields are transferred through the TF flag extensions via a torque transfer coupling to the rest of the tokamak structure. A 3D FEA analysis was performed to qualify this design. The results of these analyses, which will be presented in this paper, have led to the design of OH coil and centerstack components that meet the requirements of the NSTX-upgrade structural design criteria.

  6. 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.)

  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. 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

  9. 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)

  10. 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

  11. Simulating Plasma Turbulence in Tokamaks

    CERN Document Server

    Kepner, J V; Decyk, V; Kepner, Jeremy; Parker, Scott; Decyk, Viktor

    1997-01-01

    A challenging and fundamental research problem is the better understanding and control of the turbulent transport of heat in present-day tokamak fusion experiments. Recent developments in numerical methods along with enormous gains in computing power have made large-scale simulations an important tool for improving our understanding of this phenomena. Simulating this highly non-linear behavior requires solving for the perturbations of the phase space distribution function in five dimensions. We use a particle-in-cell approach to solve the equations. The code has been parallelized for a variety of architectures (C90, CM-5, T3D) using a 1-D domain decomposition along the toroidal axis, for which the number of particles in each cell remains approximately constant. The quasi-uniform distribution of particles, which minimizes load imbalance, coupled with the relatively small movement of particles across cells, which minimizes communications, makes this problem ideally suited to massively parallel architectures. We...

  12. 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

  13. Upgrade and validation on plasma of the Tore Supra CW LHCD generator

    International Nuclear Information System (INIS)

    A one year-long major upgrade of the 3.7 GHz Lower Hybrid Current Drive (LHCD) generator for the Tore Supra (TS) tokamak has been performed. It consisted in installing a first series of eight Thales Electron Devices (TED) 700 kW CW klystrons, new CW components and auxiliaries, and in modifying the transmitter control and protection software. Modifications and calibration of the sensors and the RF subsystems were completed as well. Finally, the RF power available in the generator has been increased by 35% and the pulse duration could reach 1000 s. A complete validation and optimization of the klystrons have been performed in 2010 on matched load before the generator could enter into operation. The eight klystrons connected with the Full Active Multijunction (FAM) antenna delivered 3.5 MW/50s in December 2010. The upgrade of the generator and the steps to validate the modifications are described.

  14. 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

    , the measured spectra agree quantitatively with the synthetic spectra in periods with and without NBI heating. For the discharges investigated, the central velocity distribution of neutral beam ions can be described by classical slowing down. These results will have a major impact on ITER physics exploration......, since CTS is presently the only diagnostic to measure the confined alpha particles produced by the fusion reactions....

  15. The upgraded MAGIC Cherenkov telescopes

    Energy Technology Data Exchange (ETDEWEB)

    Tescaro, D., E-mail: dtescaro@iac.es [Instituto de Astrofísica de Canarias (IAC), E-38205 La Laguna, Tenerife (Spain); Universidad de La Laguna (ULL), Dept. Astrofísica, E-38206 La Laguna, Tenerife (Spain)

    2014-12-01

    The MAGIC Cherenkov telescopes underwent a major upgrade in 2011 and 2012. A new 1039-pixel camera and a larger area digital trigger system were installed in MAGIC-I, making it essentially identical to the newer MAGIC-II telescope. The readout systems of both telescopes were also upgraded, with fully programmable receiver boards and DRS4-chip-based digitization systems. The upgrade eased the operation and maintenance of the telescopes and also improved significantly their performance. The system has now an integral sensitivity as good as 0.6% of the Crab Nebula flux (for E>400GeV), with an effective analysis threshold at 70 GeV. This allows MAGIC to secure one of the leading roles among the current major ground-based Imaging Atmospheric Cherenkov telescopes for the next 5–10 years. - Highlights: • In 2011 and 2012 the MAGIC telescopes underwent a two-stage major upgrade. • The new camera of MAGIC-I allows us to exploit a 1.4 larger trigger area. • The novel DRS4-based readout systems allow a cost-effective ultra-fast digitization. • The upgrade greatly improved the maintainability of the system. • MAGIC has now an optimal integral sensitivity of 0.6% of the Crab Nebula flux.

  16. Project scenarios for bitumen upgrading

    International Nuclear Information System (INIS)

    The established reserves of Alberta's heavy oil resources are 178 billion barrels, and potential recoverable reserves are 315 billion barrels. The challenge of production includes the logistics of recovery, upgrading and transportation to market. Utilization of the bitumen is not simple because bitumen is too viscous to transport by pipeline. In addition, it is not processable by most existing refineries unless it can be upgraded through dilution. This paper examined different factors regarding the economic viability of various upgrading methods of a wide range of bitumen feedstocks. The study also examined the sensitivity of refinery demand to the prices of these feedstocks, along with the competitiveness among bitumen-based feedstock and conventional crudes. Western Canada, Ontario and the PADD II district in the United States are the 3 major markets for western Canadian bitumen based feedstock, the demand for which depends on refinery configurations and asphalt demand. This paper described the following 4 generic scenarios that describe Alberta bitumen upgrading projects: (1) adjacent to open pit mines, (2) adjacent to steam assisted gravity drainage (SAGD) facilities, (3) remotely located from resource production at an existing refinery, and (4) pipeline bitumen. It was noted that producers should determine the best way to upgrade the bitumen to ensure there is an economic market for the product, but they should also be aware not to over process the bitumen so as not to leave existing refinery facilities under-utilized. 2 refs., 1 tab., 3 figs

  17. 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)

  18. 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.

  19. LHCb DAQ network upgrade tests

    CERN Document Server

    Pisani, Flavio

    2013-01-01

    My project concerned the evaluation of new technologies for the DAQ network upgrade of LHCb. The first part consisted in developing and Open Flow-based Clos network. This new technology is very interesting and powerful but, as shown by the results, it still needs further improvements. The second part consisted in testing and benchmarking 40GbE network equipment: Mellanox MT27500, Chelsio T580 and Huawei Cloud Engine 12804. An event-building simulation is currently been performed in order to check the feasibility of the DAQ network upgrade in LS2. The first results are promising.

  20. 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.)

  1. Deuterium-tritium experiments on the Tokamak Fusion Test reactor

    International Nuclear Information System (INIS)

    The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to ∼9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the PFUS ∼6 MW level. Instability in the TAE mode frequency range has been observed at PFUS > 7 MW and its effect on performance in under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fully explored

  2. Deuterium-tritium experiments on the Tokamak Fusion Test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hosea, J.; Adler, J.H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D. [and others

    1994-09-01

    The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to {approx}9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the PFUS {approx}6 MW level. Instability in the TAE mode frequency range has been observed at PFUS > 7 MW and its effect on performance in under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fully explored.

  3. Characteristics of Plasma Turbulence in the Mega Amp Spherical Tokamak

    CERN Document Server

    Ghim, Young-chul

    2013-01-01

    Turbulence is a major factor limiting the achievement of better tokamak performance as it enhances the transport of particles, momentum and heat which hinders the foremost objective of tokamaks. Hence, understanding and possibly being able to control turbulence in tokamaks is of paramount importance, not to mention our intellectual curiosity of it.

  4. Electron thermal transport in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Konings, J.A.

    1994-11-30

    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 (10{sup 8} 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).

  5. 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)

  6. 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.)

  7. 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.

  8. Monte Carlo simulation for the calibration of neutron source strength measurement of JT-60 upgrade

    International Nuclear Information System (INIS)

    The calibration of the relation between the neutron source strength in the whole plasma and the output of neutron monitor is important to evaluate the fusion gain in tokamaks with DD or DT operation. JT-60 will be modified to be tokamak of deuterium plasma with Ip≤7MA and V≤110 m3. The source strength of JT-60 Upgrade will be measured with 235U and 238U fission chambers. Detection efficiencies for source neutron are calculated by the Monte Carlo code MCNP with 3-dimensional modelling of JT-60 Upgrade and with the poloidally distributed neutron source. More than 90% of fission chamber's counts are contributed by source of -85deg235U and 238U detectors, respectively. Detection efficiencies are sensitive to major radius of the detector position, but not so sensitive to vertical and toroidal shift of the detector positions. And total uncertainties combined detector position errors are ±13% and ±9% for 235U and 238U detectors, respectively. The modelling errors of the detection efficiencies are so large for the 238U detector that more precise modelling including the port boxes is needed. (author)

  9. Upgrading of the West Area

    CERN Multimedia

    1983-01-01

    The rejigged main hall (EHW1) in the West Area: on background, below the crane, is the brown yoke of the Omega magnet which had been resited. The upgrading was completed by the time in July when 400 GeV protons arrived. See Annual Report 1983 p. 107.

  10. Preparing the ALICE DAQ upgrade

    Science.gov (United States)

    Carena, F.; Carena, W.; Chapeland, S.; Chibante Barroso, V.; Costa, F.; Dénes, E.; Divià, R.; Fuchs, U.; Grigore, A.; Kiss, T.; Rauch, W.; Rubin, G.; Simonetti, G.; Soós, C.; Telesca, A.; Vande Vyvre, P.; Von Haller, B.

    2012-12-01

    In November 2009, after 15 years of design and installation, the ALICE experiment started to detect and record the first collisions produced by the LHC. It has been collecting hundreds of millions of events ever since with both proton and heavy ion collisions. The future scientific programme of ALICE has been refined following the first year of data taking. The physics targeted beyond 2018 will be the study of rare signals. Several detectors will be upgraded, modified, or replaced to prepare ALICE for future physics challenges. An upgrade of the triggering and readout systems is also required to accommodate the needs of the upgraded ALICE and to better select the data of the rare physics channels. The ALICE upgrade will have major implications in the detector electronics and controls, data acquisition, event triggering and offline computing and storage systems. Moreover, the experience accumulated during more than two years of operation has also lead to new requirements for the control software. We will review all these new needs and the current R&D activities to address them. Several papers of the same conference present in more details some elements of the ALICE online system.

  11. ATLAS Detector : Performance and Upgrades

    CERN Document Server

    Oliveira Damazio, Denis; The ATLAS collaboration

    2016-01-01

    Describe the ATLAS detector and summarize most relevant and recent information about the detector performance in 2016 with LHC colliding bunches at sqrt(s)=13 TeV with luminosity above the nominal value. Describe the different upgrade phases previewed for the detector and main activities already ongoing.

  12. First escaping fast ion mesurements in ITER-like geometry using an activation probe

    OpenAIRE

    BONHEURE Georges; Hult, Mikael; Fenyvesi, A.; ÄKÄSLOMPOLO S.; Carralero, D.; DEGERING Detlef; DE-VISMES OTT A.; Garcia-Munoz, M; Gmeiner, B.; Herrmann, A.; Laubenstein, Matthias; LUTTER GUILLAUME; Mlynar, J.; Mueller, H. W.; ROHDE V.

    2015-01-01

    More research is needed to develop suitable diagnostics for measuring alpha particle confinement in ITER and techniques relevant for fusion reactor conditions need further development. Based on nuclear reactions, the activation probe is a novel concept first tested in JET. It may offer a more robust solution for performing alpha particle measurements in ITER. This paper describes the first escaping fast ion measurements performed at ASDEX Upgrade (AUG) tokamak using an activation probe. A det...

  13. Max-Planck-Institut fuer Plasmaphysik. Annual report 1993

    International Nuclear Information System (INIS)

    In 1993 the first particle injector of the ASDEX Upgrade divertor tokamak was put into operation up to 6 MW of heating power. The main diagnostics were put into operation as also was a newly developed pellet centrifuge. At the Wendelstein 7-AS stellarator experiment the cooperation with Russian and German research institutes an electron cyclotron resonance heating was successful. The works of the Berlin Division of IPP and the coordination of research efforts with Kernforschungszentrum Karlsruhe are reported. (DG)

  14. Studies of ELM heat load, SOL flow and carbon erosion from existing tokamak experiments, and projections for ITER

    International Nuclear Information System (INIS)

    Three important physics issues for the ITER divertor design and operation are summarized based on the experimental and numerical work from multi-machine database (JET, JT-60U, ASDEX Upgrade, DIII-D, Alcator C-Mod and TEXTOR). (i) The energy load associated with Type-I ELMs is of great concern for the lifetime of the ITER divertor target. In order t o understand the physics base of the scaling models[1], the ELM heat and particle transport from the edge pedestal to the divertor is investigated. Convective transport during ELMs plays an important role in heat transport to the divertor. (ii) Determination of the SOL flow pattern and the driving mechanism has progressed experimentally and numerically. Influences of the drift effects on the SOL and divertor plasma transport were discussed. (iii) Carbon erosion and redeposition are of great importance in particular for tritium retention via codeposition. Characteristics of chemical yield at two different deposited carbon surfaces, i.e. erosion- and redeposition-dominated areas, have been studied. Progress in the understanding of the chemical erosion is reviewed. (author)

  15. 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)

  16. 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)

  17. The Spherical Tokamak MEDUSA for Mexico

    Science.gov (United States)

    Ribeiro, C.; Salvador, M.; Gonzalez, J.; Munoz, O.; Tapia, A.; Arredondo, V.; Chavez, R.; Nieto, A.; Gonzalez, J.; Garza, A.; Estrada, I.; Jasso, E.; Acosta, C.; Briones, C.; Cavazos, G.; Martinez, J.; Morones, J.; Almaguer, J.; Fonck, R.

    2011-10-01

    The former spherical tokamak MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R Mexican Fusion Network. Strong liaison within national and international plasma physics communities is expected. New activities on plasma & engineering modeling are expected to be developed in parallel by using the existing facilities such as a multi-platform computer (Silicon Graphics Altix XE250, 128G RAM, 3.7TB HD, 2.7GHz, quad-core processor), ancillary graph system (NVIDIA Quadro FE 2000/1GB GDDR-5 PCI X16 128, 3.2GHz), and COMSOL Multiphysics-Solid Works programs.

  18. 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.

  19. 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.

  20. 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.

  1. 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)

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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)

  7. Take control of upgrading to Leopard

    CERN Document Server

    Kissell, Joe

    2009-01-01

    Start on the right foot with Mac OS X 10.5 Leopard! Little is more exciting and unnerving than a major operating system upgrade for your Mac, but thousands of people have upgraded to Panther and Tiger calmly and successfully with the advice in Joe Kissell's previous hit Take Control of Upgrading... titles. Joe's expert guidance, developed over innumerable test installations, walks you through the six steps necessary before upgrading, which of Leopard's three installation options is right for you, how to perform the actual upgrade, and post-installation checking and cleanup.

  8. MIPP Plastic Ball electronics upgrade

    International Nuclear Information System (INIS)

    An upgrade electronics design for Plastic Ball detector is described. The Plastic Ball detector was a part of several experiments in the past and its back portion (proposed to be used in MIPP) consists of 340 photomultipliers equipped with a sandwich scintillator. The scintillator sandwich has fast and slow signal component with decay times 10 ns and 1 (micro)s respectively. The upgraded MIPP experiment will collect up to 12,000 events during each 4 second spill and read them out in ∼50 seconds between spills. The MIPP data acquisition system will employ deadtime-less concept successfully implemented in Muon Electronics of Dzero experiment at Fermilab. An 8-channel prototype design of the Plastic Ball Front End (PBFE) implementing these requirements is discussed. Details of the schematic design, simulation and prototype test results are discussed.

  9. MIPP Plastic Ball electronics upgrade

    International Nuclear Information System (INIS)

    An upgrade electronics design for Plastic Ball detector is described. The Plastic Ball detector was a part of several experiments in the past and its back portion (proposed to be used in Main Injector Particle Production (MIPP)) consists of 340 photomultipliers equipped with a sandwich scintillator. The scintillator sandwich has fast and slow signal component with decay times 10 ns and 1 μs, respectively. The upgraded MIPP experiment will collect up to 12,000 events during each 4 s spill and read them out in ∼50 s between spills. The MIPP data acquisition system will employ deadtime-less concept successfully implemented in Muon Electronics of Dzero experiment at Fermilab An 8-channel prototype design of the Plastic Ball Front-End (PBFE) implementing these requirements is discussed. Details of the schematic design, simulation and prototype test results are discussed.

  10. B physics with upgraded detector

    CERN Document Server

    Palla, Fabrizio

    2016-01-01

    The CMS potential for B-Physics with the Upgraded Phase-I and Phase-II detectors will be discussed, with the $\\mathrm{B}_{s}^{0}\\to\\mu^{+}\\mu^{-}$ and $\\mathrm{B}^{0}\\to\\mu^{+}\\mu^{-}$ benchmark channels, for the runs of the LHC at $\\sqrt{s}=14$~TeV up to an integrated luminosity of 3000~$\\mathrm{fb}^{-1}$. With the upgraded CMS detector it will be possible to efficiently trigger and reconstruct both processes, with reduced statistical and systematic uncertainties leading to high precision measurements of the branching fractions of the $\\mathrm{B}^{0}\\to\\mu^{+}\\mu^{-}$ and $\\mathrm{B}_{s}^{0}\\to\\mu^{+}\\mu^{-}$ decays. This will allow in turn stringent tests of the Standard Model.

  11. Tevatron Beam Position Monitor Upgrade

    CERN Document Server

    Wolbers, Stephen; Barker, B; Bledsoe, S; Boes, T; Bowden, Mark; Cancelo, Gugstavo I; Dürling, G; Forster, B; Haynes, B; Hendricks, B; Kasza, T; Kutschke, Robert K; Mahlum, R; Martens, Michael A; Mengel, M; Olsen, M; Pavlicek, V; Pham, T; Piccoli, Luciano; Steimel, Jim; Treptow, K; Votava, Margaret; Webber, Robert C; West, B; Zhang, D

    2005-01-01

    The Tevatron Beam Position Monitor (BPM) readout electronics and software have been upgraded to improve measurement precision, functionality and reliability. The original system, designed and built in the early 1980s, became inadequate for current and future operations of the Tevatron. The upgraded system consists of 960 channels of new electronics to process analog signals from 240 BPMs, new front-end software, new online and controls software, and modified applications to take advantage of the improved measurements and support the new functionality. The new system reads signals from both ends of the existing directional stripline pickups to provide simultaneous proton and antiproton position measurements. Measurements using the new system are presented that demonstrate its improved resolution and overall performance.

  12. Upgrade of the DIII-D vacuum vessel protection system

    International Nuclear Information System (INIS)

    An upgrade of the General Atomics DIII-D tokamak armor protection system has been completed. The upgrade consisted of armoring the outer wall and the divertor gas baffle with monolithic graphite tiles and cleaning the existing floor, ceiling, and inner wall tiles to remove any deposited impurity layer from the tile surfaces. The new tiles replace the graphite tiles used as local armor for neutral beam shine through, three graphite poloidal back-up limiter bands, and miscellaneous Inconel protection tiles. The total number of tiles increased from 1636 to 3200 and corresponding vessel coverage from 40% to 90%. A new, graphite armored, toroidally continuous, gas baffle between the outer wall and the biased divertor ring was installed in order to accommodate the cryocondensation pump that was installed in parallel with the outer wall tiles. To eliminate a source of copper in the plasma, GRAFOIL gaskets replaced the copper felt metal gaskets previously used as a compliant heat transfer interface between the inertially cooled tiles and the vessel wall. GRAFOIL, an exfoliated, flexible graphite material from Union Carbide, Inc., was used between each tile and the vessel wall and also between each tile and its hold-down hardware. Testing was performed to determine the mechanical compliance, thermal conductance, and vacuum characteristics of the GRAFOIL material. To further decrease the quantity of high Z materials exposed to the plasma, the 1636 existing graphite tiles were identified, removed, and grit blasted to eliminate a thin layer of deposited metals which included nickel, chromium, and molybdenum. Prior to any processing, a selected set of tiles was tested for radioactivity, including tritium contamination. The tiles were grit blasted in a negative-pressure blasting cabinet using 37 μm boron carbide powder as the blast media and dry nitrogen as the propellant

  13. Conclusion. Upgrading the marketing function

    OpenAIRE

    Lambin, Jean-Jacques

    2013-01-01

    More countries, more customer segments, more distribution channels, more direct and substitute competitors, new market actors, new communication technologies, new virtual market places, fast changing technologies, new media controlled by consumers, new civil society stakeholders, … The growing complexity of the global market has numerous causes. To cope with this increased complexity, the firm has to redefine the way the marketing function operates by broadening and upgrading its role, by cha...

  14. CMS upgrade and future plans

    OpenAIRE

    Hoepfner Kerstin

    2015-01-01

    CMS plans for operation at the LHC phase-II unprecedented in terms of luminosity thus resulting in serious consequences for detector performance. To achieve the goal to maintain the present excellent performance of the CMS detector, several upgrades are necessary. To handle the high phase-II data rates, the readout and trigger systems are redesigned using recent technology developments. The high particle rates will accelerate detector aging and require replacement of the tracker and forward c...

  15. RHIC and its upgrade programmes.

    Energy Technology Data Exchange (ETDEWEB)

    Roser,T.

    2008-06-23

    As the first hadron accelerator and collider consisting of two independent superconducting rings RHIC has operated with a wide range of beam energies and particle species. After a brief review of the achieved performance the presentation will give an overview of the plans, challenges and status of machine upgrades, that range from a new heavy ion pre-injector and beam cooling at 100 GeV to a high luminosity electron-ion collider.

  16. AliPDU Package Upgrade

    CERN Document Server

    "Martin, Michael

    2015-01-01

    "AliPDU Package" is a set of script, panels, and datapoints designed in WinCC to manage and monitor PDU's. PDU is an essential component in the data center, in order to make data center working properly through the monitoring of power distribution and environmental condition of the data center. In this project "AliPDU Package" is upgraded so it can be used to monitor environmental condition of data center using PDU's and external environmental sensor connected to PDU.

  17. AliPDU Package Upgrade

    CERN Document Server

    Martin, Michael

    2015-01-01

    AliPDU Package is a set of script, panels, and datapoints designed in WinCC to manage and monitor PDU's. PDU is an essential component in the data center, in order to make data center working properly through the monitoring of power distribution and environmental condition of the data center. In this project "AliPDU Package" is upgraded so it can be used to monitor environmental condition of data center using PDU's and external environmental sensor connected to PDU.

  18. 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.

  19. 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

  20. INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

    Energy Technology Data Exchange (ETDEWEB)

    HUMPHREYS,D.A; FERRON,J.R; JOHNSON,R.D; LEUER,J.A; PENAFLOR,B.G; WALKER,M.L; WELANDER,A.S; KHAYRUTDINOV,R.R; DOKOUKA,V; EDGELL,D.H; FRANSSON,C.M

    2003-10-01

    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.

  1. 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.

  2. 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.

  3. 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

  4. 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

  5. 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)

  6. 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

  7. Tokamak Transport Studies Using Perturbation Analysis

    NARCIS (Netherlands)

    Cardozo, N. J. L.; Dehaas, J. C. M.; Hogeweij, G. M. D.; Orourke, J.; Sips, A.C.C.; Tubbing, B. J. D.

    1990-01-01

    Studies of the transport properties of tokamak plasmas using perturbation analysis are discussed. The focus is on experiments with not too large perturbations, such as sawtooth induced heat and density pulse propagation, power modulation and oscillatory gas-puff experiments. The approximations made

  8. RIPS upgrade and physics programs

    Science.gov (United States)

    Ueno, Hideki; Yoshimi, Akihiro; Asahi, Koichiro

    2009-10-01

    The upgrade of RIPS has been proposed in the phase-II program of RIKEN RI Beam Factory (RIBF) project. In this upgrade, the former fragment separator RIPS will be equipped with a new beam line that delivers beams of 115A-MeV heavy io ns extracted from the IRC cyclotron by skipping the final acceleration of SRC. This beam energy is high enough to produce radioactive isotope beams (RIBs) via the projectile-fragmentation reaction. Thus, compared with RIBs produced in the present AVF-RRC acceleration scheme, their production yield are drastically increased by this upgrade, especially in the mass region heavier than Kr. Remarkably, RIPS further enhances research opportunities on spin-related subjects such as nuclear structure studies through electromagnetic nuclear moments: it has been revealed that RIBs produced at this energy can be spin-oriented independently of their atomic and chemical properties. Also, the research subjects include not only nuclear moments but also material science by means, e.g., of the β-NMR, γ-PAD, γ-PAC, laser, and in-beam M"ossbauer methods, because RIBs of this energy allow for a scheme to implant them into sample materials with limited thickness and thus stopped-RI type experiments will be conveniently carried out.

  9. Scenarios for the LHC Upgrade

    CERN Document Server

    Scandale, Walter

    2008-01-01

    The projected lifetime of the LHC low-beta quadrupoles, the evolution of the statistical error halving time, and the physics potential all call for an LHC luminosity upgrade by the middle of the coming decade. In the framework of the CARE-HHH network three principal scenarios have been developed for increasing the LHC peak luminosity by more than a factor of 10, to values above 1035 cm−2s−1. All scenarios imply a rebuilding of the high-luminosity interaction regions (IRs) in combination with a consistent change of beam parameters. However, their respective features, bunch structures, IR layouts, merits and challenges, and luminosity variation with β∗ differ substantially. In all scenarios luminosity leveling during a store would be advantageous for the physics experiments. An injector upgrade must complement the upgrade measures in the LHC proper in order to provide the beam intensity and brightness needed as well as to reduce the LHC turnaround time for higher integrated luminosity.

  10. Modelling of wall and SOL processes and contamination of ITER plasma after impurity injection with the tokamak code TOKES

    International Nuclear Information System (INIS)

    In the future tokamak ITER the damage to the wall after the disruptions can be mitigated using preventive massive gas injection (MGI) of noble gases into confined plasma during the thermal quench. The gas gets ionized in the plasma, and then the ions dump into the scrape-off layer (SOL) and impact on the target. The contamination of core plasma results in fast loss of plasma energy by radiation. The radiation distributes rather homogeneously over the wall. However, enhanced radiation load in e.g. vicinity of gas jet entry is an issue for ITER design that can be addressed numerically. For the modelling the tokamak code TOKES is applied, after upgrading it with toroidally symmetric 2D plasma model. This allowed detailed radiation fluxes and the expansion of noble ions both across and along the magnetic surfaces. In the work one- and two-dimensional (2D) MGI models are evaluated. 2D model is preliminary compared with the tokamak DIII-D. Substantial discrepancies were explained, and then predictive simulations for ITER performed, with the conclusion that after the radiation flush in front of jet entry the wall temperature can exceed the beryllium melting point.

  11. Fishbone activity in experimental advanced superconducting tokamak neutral beam injection plasma

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Liqing; Zhang, Jizong; Chen, Kaiyun, E-mail: Kychen@ipp.cas.cn, E-mail: lqhu@ipp.cas.cn; Hu, Liqun, E-mail: Kychen@ipp.cas.cn, E-mail: lqhu@ipp.cas.cn; Li, Erzhong; Lin, Shiyao; Shi, Tonghui; Duan, Yanmin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Zhu, Yubao [Department of Physics and Astronomy, University of California, Irvine, California 92697-4575 (United States)

    2015-12-15

    Repetitive fishbones near the trapped ion procession frequency were observed for the first time in the neutral beam injection high confinement plasmas in Experimental Advanced Superconducting Tokamak (EAST) tokamak, and diagnosed using a solid-state neutral particle analyzer based on a compact silicon photodiode together with an upgraded high spatial-temporal-resolution multi-arrays soft X-ray (SX) system. This 1/1 typical internal kink mode propagates in the ion-diamagnetism direction with a rotation speed faster than the bulk plasma in the plasma frame. From the SX measurements, this mode frequency is typical of chirping down and the energetic particle effect related to the twisting mode structure. This ion fishbone was found able to trigger a multiple core sawtooth crashes with edge-2/1 sideband modes, as well as to lead to a transition from fishbone to long lived saturated kink mode to fishbone. Furthermore, using SX tomography, a correlation between mode amplitude and mode frequency was found. Finally, a phenomenological prey–predator model was found to reproduce the fishbone nonlinear process well.

  12. Calculated radiative power losses from mid- and high-Z impurities in Tokamak plasmas

    Science.gov (United States)

    Fournier, Kevin B.; May, M. J.; Pacella, D.; Gregory, B. C.; Rice, J. E.; Terry, J. L.; Finkenthal, M.; Goldstein, W. H.

    1998-09-01

    This paper summarizes recent calculations of the radiative cooling coefficient for molybdenum (Z=42), krypton (Z=36) and argon (Z=18). The radiative processes considered are collisional-radiative line emission, dielectronic recombination line emission, and radiative recombination and bremsstrahlung continuum emission. Collisional-radiative line emission dominates the power loss channels for a given impurity at all but the highest plasma electron temperatures. The atomic data for the line emission are computed ab initio with the HULLAC atomic physics suite of codes. Relativistic, ab initio atomic physics data are used to compute ionization and recombination rate coefficients; the resulting charge state distribution and recombination rates are used to estimate the radiative power from recombination processes. The calculations in the present work are benchmarked against absolute measurements of ion brightness profiles in the Frascati Tokamak Upgrade plasma. Integrated measurements from tokamak plasmas such as bolometry are then simulated. The atomic physics data used to predict the emissivity of individual ions is validated; the calculated cooling coefficients agree well with bolometric measurements.

  13. Calculated radiative power losses from mid- and high-Z impurities in Tokamak plasmas

    International Nuclear Information System (INIS)

    This paper summarizes recent calculations of the radiative cooling coefficient for molybdenum (Z=42), krypton (Z=36) and argon (Z=18). The radiative processes considered are collisional-radiative line emission, dielectronic recombination line emission, and radiative recombination and bremsstrahlung continuum emission. Collisional-radiative line emission dominates the power loss channels for a given impurity at all but the highest plasma electron temperatures. The atomic data for the line emission are computed ab initio with the HULLAC atomic physics suite of codes. Relativistic, ab initio atomic physics data are used to compute ionization and recombination rate coefficients; the resulting charge state distribution and recombination rates are used to estimate the radiative power from recombination processes. The calculations in the present work are benchmarked against absolute measurements of ion brightness profiles in the Frascati Tokamak Upgrade plasma. Integrated measurements from tokamak plasmas such as bolometry are then simulated. The atomic physics data used to predict the emissivity of individual ions is validated; the calculated cooling coefficients agree well with bolometric measurements

  14. Topology of toroidal helical fields in non-circular cross-sectional tokamaks

    Institute of Scientific and Technical Information of China (English)

    Zha Xue-Jun; Zhu Si-Zheng; Yu Qing-Quan; Wang Yan

    2005-01-01

    The ordinary differential magnetic field line equations are solved numerically; the tokamak magnetic structure is studied on Hefei Tokamak-7 Upgrade (HT-7U) when the equilibrium field with a monotonic q-profile is perturbed by a helical magnetic field. We find that a single mode (m, n) helical perturbation can cause the formation of islands on rational surfaces with q = m/n and q = (m ± 1,±2, ±3,...)/n due to the toroidicity and plasma shape (i.e.elongation and triangularity), while there are many undestroyed magnetic surfaces called Kolmogorov-Arnold-Moser (KAM) barriers on irrational surfaces. The islands on the same rational surface do not have the same size. When the ratio between the perturbing magnetic field (B)r(r) and the toroidal magnetic field amplitude Bφ0 is large enough, the magnetic island chains on different rational surfaces will overlap and chaotic orbits appear in the overlapping area, and the magnetic field becomes stochastic. It is remarkable that the stochastic layer appears first in the plasma edge region.

  15. Experiments with Liquid Metal Walls: Status of the Lithium Tokamak Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Kaita, Robert; Boyle, Dennis; Gray, Timothy; Granstedt, Erik; Hammett, Gregory; Jacobson, Craig M; Jones, Andrew; Kozub, Thomas; Kugel, Henry; Leblanc, Benoit; Logan, Nicholas; Lucia, Matthew; Lundberg, Daniel; Majeski, Richard; Mansfield, Dennis; Menard, Jonathan; Spaleta, Jeffrey; Strickler, Trevor

    2010-02-16

    Liquid metal walls have been proposed to address the first wall challenge for fusion reactors. The Lithium Tokamak Experiment (LTX) at the Princeton Plasma Physics Laboratory (PPPL) is the first magnetic confinement device to have liquid metal plasma-facing components (PFC's) that encloses virtually the entire plasma. In the Current Drive Experiment-Upgrade (CDX-U), a predecessor to LTX at PPPL, the highest improvement in energy confinement ever observed in Ohmically-heated tokamak plasmas was achieved with a toroidal liquid lithium limiter. The LTX extends this liquid lithium PFC by using a conducting conformal shell that almost completely surrounds the plasma. By heating the shell, a lithium coating on the plasma-facing side can be kept liquefied. A consequence of the low-recycling conditions from liquid lithium walls is the need for efficient plasma fueling. For this purpose, a molecular cluster injector is being developed. Future plans include the installation of a neutral beam for core plasma fueling, and also ion temperature measurements using charge-exchange recombination spectroscopy. Low edge recycling is also predicted to reduce temperature gradients that drive drift wave turbulence. Gyrokinetic simulations are in progress to calculate fluctuation levels and transport for LTX plasmas, and new fluctuation diagnostics are under development to test these predictions. __________________________________________________

  16. Experiments with liquid metal walls: Status of the lithium tokamak experiment

    Energy Technology Data Exchange (ETDEWEB)

    Kaita, Robert, E-mail: kaita@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Berzak, Laura; Boyle, Dennis; Gray, Timothy; Granstedt, Erik; Hammett, Gregory; Jacobson, Craig M.; Jones, Andrew; Kozub, Thomas; Kugel, Henry; Leblanc, Benoit; Logan, Nicholas; Lucia, Matthew; Lundberg, Daniel; Majeski, Richard; Mansfield, Dennis; Menard, Jonathan; Spaleta, Jeffrey; Strickler, Trevor; Timberlake, John [Princeton Plasma Physics Laboratory, Princeton, NJ (United States)

    2010-11-15

    Abstarct: Liquid metal walls have been proposed to address the first wall challenge for fusion reactors. The lithium tokamak experiment (LTX) at the Princeton Plasma Physics Laboratory (PPPL) is the first magnetic confinement device to have liquid metal plasma-facing components (PFC's) that encloses virtually the entire plasma. In the current drive experiment-upgrade (CDX-U), a predecessor to LTX at PPPL, the highest improvement in energy confinement ever observed in ohmically heated tokamak plasmas was achieved with a toroidal liquid lithium limiter. The LTX extends this liquid lithium PFC by using a conducting conformal shell that almost completely surrounds the plasma. By heating the shell, a lithium coating on the plasma-facing side can be kept liquefied. A consequence of the low-recycling conditions from liquid lithium walls is the need for efficient plasma fueling. For this purpose, a molecular cluster injector is being developed. Future plans include the installation of a neutral beam for core plasma fueling, and also ion temperature measurements using charge-exchange recombination spectroscopy (CHERS). Low edge recycling is also predicted to reduce temperature gradients that drive drift wave turbulence. Gyrokinetic simulations are in progress to calculate fluctuation levels and transport for LTX plasmas, and new fluctuation diagnostics are under development to test these predictions.

  17. Fishbone activity in experimental advanced superconducting tokamak neutral beam injection plasma

    International Nuclear Information System (INIS)

    Repetitive fishbones near the trapped ion procession frequency were observed for the first time in the neutral beam injection high confinement plasmas in Experimental Advanced Superconducting Tokamak (EAST) tokamak, and diagnosed using a solid-state neutral particle analyzer based on a compact silicon photodiode together with an upgraded high spatial-temporal-resolution multi-arrays soft X-ray (SX) system. This 1/1 typical internal kink mode propagates in the ion-diamagnetism direction with a rotation speed faster than the bulk plasma in the plasma frame. From the SX measurements, this mode frequency is typical of chirping down and the energetic particle effect related to the twisting mode structure. This ion fishbone was found able to trigger a multiple core sawtooth crashes with edge-2/1 sideband modes, as well as to lead to a transition from fishbone to long lived saturated kink mode to fishbone. Furthermore, using SX tomography, a correlation between mode amplitude and mode frequency was found. Finally, a phenomenological prey–predator model was found to reproduce the fishbone nonlinear process well

  18. A survey of electron Bernstein wave heating and current drive potential for spherical tokamaks

    CERN Document Server

    Urban, Jakub; Peysson, Yves; Preinhaelter, Josef; Shevchenko, Vladimir; Taylor, Gary; Vahala, Linda; Vahala, George

    2011-01-01

    The electron Bernstein wave (EBW) is typically the only wave in the electron cyclotron (EC) range that can be applied in spherical tokamaks for heating and current drive (H&CD). Spherical tokamaks (STs), which feature relatively high neutron flux and good economy, operate generally in high-beta regimes, in which the usual EC O- and X- modes are cut-off. In this case, EBWs seem to be the only option that can provide features similar to the EC waves---controllable localized H&CD that can be utilized for core plasma heating as well as for accurate plasma stabilization. The EBW is a quasi-electrostatic wave that can be excited by mode conversion from a suitably launched O- or X-mode; its propagation further inside the plasma is strongly influenced by the plasma parameters. These rather awkward properties make its application somewhat more difficult. In this paper we perform an extensive numerical study of EBW H&CD performance in four typical ST plasmas (NSTX L- and H-mode, MAST Upgrade, NHTX). Coupled...

  19. Pellet penetration in ASDEX: a comparison of results computed by means of the ORNL ablation model with measured data

    International Nuclear Information System (INIS)

    The neutral gas plasma shielding (NGPS) ablation model recently proposed by Houlberg et al. has been extensively tested on pellet penetration depths measured in JET. The best fit among calculated and measured penetration depths was obtained by assuming a shielding cloud radius 1 mm larger than the local pellet radius: Rcl = rp + 1 mm, yielding maximum shielding at the end of the pellet liftime (rp = 0). Recently, a model was developed that describes the time evolution of particle clouds in plasmas. With the help of this model, the ionization radius, i.e. the radius of the shielding cloud, can be calculated as a function of the local ablation rate. The results of these calculations show that the shielding cloud radius is proportional to the number of particles locally deposited. The cloud expansion code can be combined with the ORNL ablation model with an Rcl feedback option between the two models. Calculations are performed here for a number of randomly selected pellet-fuelled ASDEX shots. The pellet penetration is calculated for the measured Te(r) and ne(r) profiles by means of the ORNL ablation code with and without Rcl feedback active. (author) 3 figs., 2 tabs

  20. Upgrade to the Tritium Remote Control and Monitoring System for TFTR D and D

    International Nuclear Information System (INIS)

    Since 1988, the Tritium Remote Control and Monitoring System (TRECAMS) has performed crucial functions in support of D-T [deuterium-tritium] operations of the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory (PPPL). Although plasma operations on TFTR were completed in 1997, the need for TRECAMS continued. During this period TRECAMS supported the TFTR tritium systems, the TFTR's Shutdown and Safing phase, and the TFTR Decontamination and Decommissioning (D and D) project. The most critical function of the TRECAMS in the post-TFTR era has been to provide a real-time indication of the airborne tritium levels in the tritium areas and the (HVAC) stacks. TRECAMS is a critical tool in conducting safe TFTR D and D tritium-line breaks and other tritium-related work activities. Beginning in 1998, the failure rate of the system's hardware sharply increased. Furthermore, the specialized knowledge required to maintain the original software and hardware was diminishing. It soon became apparent that a failure of the TRECAMS could significantly impact the TFTR D and D project's cost and schedule. To preclude this, the TRECAMS hardware and software was upgraded in the year 2000 to use modern components. This paper will describe that successful upgrade, including a review of the engineering processes and our operating experiences with the upgraded system

  1. Overview of the physics and engineering design of NSTX upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Menard, J. E. [Princeton Plasma Physics Laboratory (PPPL); Gerhardt, S. P. [Princeton Plasma Physics Laboratory (PPPL); Bell, M. [Princeton Plasma Physics Laboratory (PPPL); Bialek, J [Columbia University; Brooks, A. [Princeton Plasma Physics Laboratory (PPPL); Canik, John [ORNL; Chrzanowski, J. [Princeton Plasma Physics Laboratory (PPPL); Denault, M [Princeton Plasma Physics Laboratory (PPPL)

    2012-01-01

    The spherical tokamak (ST) is a leading candidate for a Fusion Nuclear Science Facility (FNSF) due to its compact size and modular configuration. The National Spherical Torus eXperiment (NSTX) is a MA-class ST facility in the US actively developing the physics basis for an ST-based FNSF. In plasma transport research, ST experiments exhibit a strong (nearly inverse) scaling of normalized confinement with collisionality, and if this trend holds at low collisionality, high fusion neutron fluences could be achievable in very compact ST devices. A major motivation for the NSTX Upgrade (NSTX-U) is to span the next factor of 3-6 reduction in collisionality. To achieve this collisionality reduction with equilibrated profiles, NSTX-U will double the toroidal field, plasma current, and NBI heating power and increase the pulse length from 1-1.5 s to 5-8 s. In the area of stability and advanced scenarios, plasmas with higher aspect ratio and elongation, high beta(N), and broad current profiles approaching those of an ST-based FNSF have been produced in NSTX using active control of the plasma beta and advanced resistive wall mode control. High non-inductive current fractions of 70% have been sustained for many current diffusion times, and the more tangential injection of the 2nd NBI of the Upgrade is projected to increase the NBI current drive by up to a factor of 2 and support 100% non-inductive operation. More tangential NBI injection is also projected to provide non-solenoidal current ramp-up as needed for an ST-based FNSF. In boundary physics, NSTX measures an inverse relationship between the scrape-off layer heat-flux width and plasma current that could unfavourably impact next-step devices. Recently, NSTX has successfully demonstrated substantial heat-flux reduction using a snowflake divertor configuration, and this type of divertor is incorporated in the NSTX-U design. The physics and engineering design supporting NSTX Upgrade is described.

  2. Tritium pellet injector for the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    The tritium pellet injector (TPI) for the Tokamak Fusion Test Reactor (TFTR) will provide a tritium pellet fueling capability with pellet speeds in the 1- to 3-km/s range for the TFTR deuterium-tritium (D-T) plasma phase. An existing deuterium pellet injector (DPI) was modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed for frozen pellets ranging in size from 3 to 4 mm in diameter in arbitrarily programmable firing sequences at tritium pellet speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller (PLC). The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were also made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed and the TPI was tested at ORNL with deuterium pellets. Results of the testing program at ORNL are described. The TPI has been installed and operated on TFTR in support of the CY-92 deuterium plasma run period. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and tritium gloveboxes and integrated into TFTR tritium processing systems to provide full tritium pellet capability

  3. Emission system upgrades for older vehicles

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, R.R.; Finkenbiner, K.; Sommerville, R.J.

    1996-09-01

    Thirteen 1975--1980 model year vehicles were equipped with a set of components to upgrade their emission control systems. Each vehicle was tested before maintenance (as-received), after tune-up and correction of original equipment emission system defects (baseline), and after installation of the emission upgrade system (upgrade). Average emissions of non-methane hydrocarbons (NHMC), carbon monoxide (CO), and nitrogen oxides (NOx) with the emission upgrade system installed were reduced more than 60% from the baseline immediately after upgrade. Six of the vehicles accumulated 48,000 kilometers with the upgrade system. After 48,000 kilometers, average emissions of NMHC and NOx were still reduced approximately 50% compared to the baseline and average emissions of CO were reduced approximately 20%.

  4. Characterization of the Tokamak Novillo in cleaning regime; Caracterizacion del Tokamak Novillo en regimen de limpieza

    Energy Technology Data Exchange (ETDEWEB)

    Lopez C, R.; Melendez L, L.; Valencia A, R.; Chavez A, E.; Colunga S, S.; Gaytan G, E

    1992-02-15

    In this work the obtained results of the investigation about the experimental characterization of those low energy pulsed discharges of the Tokamak Novillo are reported. With this it is possible to fix the one operation point but appropriate of the Tokamak to condition the chamber in the smallest possible time for the cleaning discharges regime before beginning the main discharge. The characterization of the cleaning discharges in those Tokamaks is an unique process and characteristic of each device, since the good points of operation are consequence of those particularities of the design of the machine. In the case of the Tokamak Novillo, besides characterizing it a contribution is made to the cleaning discharges regime which consists on the one product of the current peak to peak of plasma by the duration of the discharge Ip{sub t} like reference parameter for the optimization of the operation of the device in the cleaning discharge regime. The maximum value of the parameter I{sub (p)}t, under different work conditions, allowed to find the good operation point to condition the discharges chamber of the Tokamak Novillo in short time and to arrive to a regime in which is not necessary the preionization for the obtaining of the cleaning discharges. (Author)

  5. First experiments on the TO-2 tokamak with a divertor

    International Nuclear Information System (INIS)

    Long stable discharges have been obtained in a recetrack tokamak with toroidal divertors in low plasma density regime. Divertors sharply limit plasma filament cross section, plasma density decreasing by an order at 1 cm length near the separatrix. 8 mm thick well formed flux of plasma appears at the divertor plate. Divertor power efficiency at different modes of operation is 50- 70 %. As compared to the TO-1 nondivertor tokamak some plasma filament hot zone expansion is recorded in the TO-2 tokamak

  6. Banana orbits in elliptic tokamaks with hole currents

    Science.gov (United States)

    Martin, P.; Castro, E.; Puerta, J.

    2015-03-01

    Ware Pinch is a consequence of breaking of up-down symmetry due to the inductive electric field. This symmetry breaking happens, though up-down symmetry for magnetic surface is assumed. In previous work Ware Pinch and banana orbits were studied for tokamak magnetic surface with ellipticity and triangularity, but up-down symmetry. Hole currents appear in large tokamaks and their influence in Ware Pinch and banana orbits are now considered here for tokamaks magnetic surfaces with ellipticity and triangularity.

  7. Systems studies of high-field tokamak ignition experiments

    International Nuclear Information System (INIS)

    A study of the interaction between the physics of ignition and the engineering constraints in the design of compact, high-field tokamak ignition demonstration devices is presented. The studies investigate the effects the various electron and ion thermal diffusivities, which result from the many tokamak scaling laws, have on the design parameters of an ignition device and show the feasibility of building and igniting a compact tokamak (R<1m). The relevant machine technology is discussed

  8. Steady State Advanced Tokamak (SSAT): The mission and the machine

    International Nuclear Information System (INIS)

    Extending the tokamak concept to the steady state regime and pursuing advances in tokamak physics are important and complementary steps for the magnetic fusion energy program. The required transition away from inductive current drive will provide exciting opportunities for advances in tokamak physics, as well as important impetus to drive advances in fusion technology. Recognizing this, the Fusion Policy Advisory Committee and the US National Energy Strategy identified the development of steady state tokamak physics and technology, and improvements in the tokamak concept, as vital elements in the magnetic fusion energy development plan. Both called for the construction of a steady state tokamak facility to address these plan elements. Advances in physics that produce better confinement and higher pressure limits are required for a similar unit size reactor. Regimes with largely self-driven plasma current are required to permit a steady-state tokamak reactor with acceptable recirculating power. Reliable techniques of disruption control will be needed to achieve the availability goals of an economic reactor. Thus the central role of this new tokamak facility is to point the way to a more attractive demonstration reactor (DEMO) than the present data base would support. To meet the challenges, we propose a new ''Steady State Advanced Tokamak'' (SSAT) facility that would develop and demonstrate optimized steady state tokamak operating mode. While other tokamaks in the world program employ superconducting toroidal field coils, SSAT would be the first major tokamak to operate with a fully superconducting coil set in the elongated, divertor geometry planned for ITER and DEMO

  9. CMS upgrade and future plans

    Directory of Open Access Journals (Sweden)

    Hoepfner Kerstin

    2015-01-01

    Full Text Available CMS plans for operation at the LHC phase-II unprecedented in terms of luminosity thus resulting in serious consequences for detector performance. To achieve the goal to maintain the present excellent performance of the CMS detector, several upgrades are necessary. To handle the high phase-II data rates, the readout and trigger systems are redesigned using recent technology developments. The high particle rates will accelerate detector aging and require replacement of the tracker and forward calorimeters. In addition, the muon system will be extended.

  10. The Upgraded D0 Detector

    CERN Document Server

    Abazov, V M; Abolins, M; Acharya, B S; Adams, D L; Adams, M; Adams, T; Agelou, M; Agram, J L; Ahmed, S N; Ahn, S H; Ahsan, M; Alexeev, G D; Alkhazov, G; Alton, A; Alverson, G; Alves, G A; Anastasoaie, M; Andeen, T; Anderson, J T; Anderson, S; Andrieu, B; Angstadt, R; Anosov, V; Arnoud, Y; Arov, M; Askew, A; Åsman, B; Assis-Jesus, A C S; Atramentov, O; Autermann, C; Avila, C; Babukhadia, L; Bacon, Trevor C; Badaud, F; Baden, A; Baffioni, S; Bagby, L; Baldin, B; Balm, P W; Banerjee, P; Banerjee, S; Barberis, E; Bardon, O; Barg, W; Bargassa, P; Baringer, P; Barnes, C; Barreto, J; Bartlett, J F; Bassler, U; Bhattacharjee, M; Baturitsky, M A; Bauer, D; Bean, A; Baumbaugh, B; Beauceron, S; Begalli, M; Beaudette, F; Begel, M; Bellavance, A; Beri, S B; Bernardi, G; Bernhard, R; Bertram, I; Besançon, M; Besson, A; Beuselinck, R; Beutel, D; Bezzubov, V A; Bhat, P C; Bhatnagar, V; Binder, M; Biscarat, C; Bishoff, A; Black, K M; Blackler, I; Blazey, G; Blekman, F; Blessing, S; Bloch, D; Blumenschein, U; Bockenthein, E; Bodyagin, V; Böhnlein, A; Boeriu, O; Bolton, T A; Bonamy, P; Bonifas, D; Borcherding, F; Borissov, G; Bos, K; Bose, T; Boswell, C; Bowden, M; Brandt, A; Briskin, G; Brock, R; Brooijmans, G; Bross, A; Buchanan, N J; Buchholz, D; Bühler, M; Büscher, V; Burdin, S; Burke, S; Burnett, T H; Busato, E; Buszello, C P; Butler, D; Butler, J M; Cammin, J; Caron, S; Bystrický, J; Canal, L; Canelli, F; Carvalho, W; Casey, B C K; Casey, D; Cason, N M; Castilla-Valdez, H; Chakrabarti, S; Chakraborty, D; Chan, K M; Chandra, A; Chapin, D; Charles, F; Cheu, E; Chevalier, L; Chi, E; Chiche, R; Cho, D K; Choate, R; Choi, S; Choudhary, B; Chopra, S; Christenson, J H; Christiansen, T; Christofek, L; Churin, I; Cisko, G; Claes, D; Clark, A R; Clement, B; Clément, C; Coadou, Y; Colling, D J; Coney, L; Connolly, B; Cooke, M; Cooper, W E; Coppage, D; Corcoran, M; Coss, J; Cothenet, A; Cousinou, M C; Cox, B; Crepe-Renaudin, S; Cristetiu, M; Cummings, M A C; Cutts, D; Da Motta, H; Das, M; Davies, B; Davies, G; Davis, G A; Davis, W; De, K; de Jong, P; De Jong, S J; De La Cruz-Burelo, E; de La Taille, C; De Oliveira Martins, C; Dean, S; Degenhardt, J D; Déliot, F; Delsart, P A; Del Signore, K; De Maat, R; Demarteau, M; Demina, R; Demine, P; Denisov, D; Denisov, S P; Desai, S; Diehl, H T; Diesburg, M; Doets, M; Doidge, M; Dong, H; Doulas, S; Dudko, L V; Duflot, L; Dugad, S R; Duperrin, A; Dvornikov, O; Dyer, J; Dyshkant, A; Eads, M; Edmunds, D; Edwards, T; Ellison, J; Elmsheuser, J; Eltzroth, J T; Elvira, V D; Eno, S; Ermolov, P; Eroshin, O V; Estrada, J; Evans, D; Evans, H; Evdokimov, A; Evdokimov, V N; Fagan, J; Fast, J; Fatakia, S N; Fein, D; Feligioni, L; Ferapontov, A V; Ferbel, T; Ferreira, M J; Fiedler, F; Filthaut, F; Fisher, W; Fisk, H E; Fleck, I; Fitzpatrick, T; Flattum, E; Fleuret, F; Flores, R; Foglesong, J; Fortner, M; Fox, H; Franklin, C; Freeman, W; Fu, S; Fuess, S; Gadfort, T; Galea, C F; Gallas, E; Galyaev, E; Gao, M; García, C; García-Bellido, A; Gardner, J; Gavrilov, V; Gay, A; Gay, P; Gelé, D; Gelhaus, R; Genser, K; Gerber, C E; Gershtein, Yu; Gillberg, D; Geurkov, G; Ginther, G; Gobbi, B; Goldmann, K; Golling, T; Gollub, N; Golovtsov, V L; Gómez, B; Gómez, G; Gómez, R; Goodwin, R W; Gornushkin, Y; Gounder, K; Goussiou, A; Graham, D; Graham, G; Grannis, P D; Gray, K; Greder, S; Green, D R; Green, J; Green, J A; Greenlee, H; Greenwood, Z D; Gregores, E M; Grinstein, S; Gris, P; Grivaz, J F; Groer, L; Grünendahl, S; Grünewald, M W; Gu, W; Guglielmo, J; Sen-Gupta, A; Gurzhev, S N; Gutíerrez, G; Gutíerrez, P; Haas, A; Hadley, N J; Haggard, E; Haggerty, H; Hagopian, S; Hall, I; Hall, R E; Han, C; Han, L; Hance, R; Hanagaki, K; Hanlet, P; Hansen, S; Harder, K; Harel, A; Harrington, R; Hauptman, J M; Hauser, R; Hays, C; Hays, J; Hazen, E; Hebbeker, T; Hebert, C; Hedin, D; Heinmiller, J M; Heinson, A P; Heintz, U; Hensel, C; Hesketh, G; Hildreth, M D; Hirosky, R; Hobbs, J D; Hoeneisen, B; Hohlfeld, M; Hong, S J; Hooper, R; Hou, S; Houben, P; Hu, Y; Huang, J; Huang, Y; Hynek, V; Huffman, D; Iashvili, I; Illingworth, R; Ito, A S; Jabeen, S; Jacquier, Y; Jaffré, M; Jain, S; Jain, V; Jakobs, K; Jayanti, R; Jenkins, A; Jesik, R; Jiang, Y; Johns, K; Johnson, M; Johnson, P; Jonckheere, A; Jonsson, P; Jöstlein, H; Jouravlev, N I; Juárez, M; Juste, A; Kaan, A P; Kado, M; Käfer, D; Kahl, W; Kahn, S; Kajfasz, E; Kalinin, A M; Kalk, J; Kalmani, S D; Karmanov, D; Kasper, J; Katsanos, I; Kau, D; Kaur, R; Ke, Z; Kehoe, R; Kermiche, S; Kesisoglou, S; Khanov, A; Kharchilava, A I; Kharzheev, Yu M; Kim, H; Kim, K H; Kim, T J; Kirsch, N; Klima, B; Klute, M; Kohli, J M; Konrath, J P; Komissarov, E V; Kopal, M; Korablev, V M; Kostritskii, A V; Kotcher, J; Kothari, B; Kotwal, A V; Koubarovsky, A; Kozelov, A V; Kozminski, J; Kryemadhi, A; Kuznetsov, O; Krane, J; Kravchuk, N; Krempetz, K; Krider, J; Krishnaswamy, M R

    2005-01-01

    The D0 experiment enjoyed a very successful data-collection run at the Fermilab Tevatron collider between 1992 and 1996. Since then, the detector has been upgraded to take advantage of improvements to the Tevatron and to enhance its physics capabilities. We describe the new elements of the detector, including the silicon microstrip tracker, central fiber tracker, solenoidal magnet, preshower detectors, forward muon detector, and forward proton detector. The uranium/liquid-argon calorimeters and central muon detector, remaining from Run I, are discussed briefly. We also present the associated electronics, triggering, and data acquisition systems, along with the design and implementation of software specific to D0.

  11. VISIR upgrade overview and status

    Science.gov (United States)

    Kerber, Florian; Käufl, Hans-Ulrich; Baksai, Pedro; Di Lieto, Nicola; Dobrzycka, Danuta; Duhoux, Philippe; Finger, Gert; Heikamp, Stephanie; Ives, Derek; Jakob, Gerd; Lundin, Lars; Mawet, Dimitri; Mehrgan, Leander; Momany, Yazan; Moreau, Vincent; Pantin, Eric; Riquelme, Miguel; Sandrock, Stefan; Siebenmorgen, Ralf; Smette, Alain; Taylor, Julian; van den Ancker, Mario; Valdes, Guillermo; Venema, Lars; Weilenmann, Ueli

    2014-07-01

    We present an overview of the VISIR upgrade project. VISIR is the mid-infrared imager and spectrograph at ESO's VLT. The project team is comprised of ESO staff and members of the original VISIR consortium: CEA Saclay and ASTRON. The project plan is based on input from the ESO user community with the goal of enhancing the scientific performance and efficiency of VISIR by a combination of measures: installation of improved hardware, optimization of instrument operations and software support. The cornerstone of the upgrade is the 1k by 1k Si:As AQUARIUS detector array (Raytheon) which has been carefully characterized in ESO's IR detector test facility (modified TIMMI 2 instrument). A prism spectroscopic mode will cover the N-band in a single observation. New scientific capabilities for high resolution and high-contrast imaging will be offered by sub-aperture mask (SAM) and phase-mask coronagraphic (4QPM/AGPM) modes. In order to make optimal use of favourable atmospheric conditions a water vapour monitor has been deployed on Paranal, allowing for real-time decisions and the introduction of a user-defined constraint on water vapour. During the commissioning in 2012 it was found that the on-sky sensitivity of the AQUARIUS detector was significantly below expectations and that VISIR was not ready to go back to science operations. Extensive testing of the detector arrays in the laboratory and on-sky enabled us to diagnose the cause for the shortcoming of the detector as excess low frequency noise (ELFN). It is inherent to the design chosen for this detector and can't be remedied by changing the detector set-up. Since this is a form of correlated noise its impact can be limited by modulating the scene recorded by the detector. We have studied several mitigation options and found that faster chopping using the secondary mirror (M2) of the VLT offers the most promising way forward. Faster M2 chopping has been tested and is scheduled for implementation before the end of 2014

  12. Voltage Upgrading of Overhead Lines

    OpenAIRE

    Olsen, Anders Tuhus

    2010-01-01

    Statnett wants to increase the transmission capacity in their 300 kV overhead lines by upgrading the operating voltage to 420 kV. To make this possible some modifications must be done. Insulator strings have to be elongated by two to four insulators and the air clearances must be checked. EN standards provide guidelines for how to calculate the air clearances adequately to provide required safety margins.It turns out that the formulas given by the standards provide greater safety margin than ...

  13. PF-AR upgrading project

    CERN Document Server

    Kasuga, T

    2002-01-01

    The upgrading project of the dedicated pulse X-ray source PF-AR has been completed by the end of the 2001 fiscal year. Machine commissioning exclusively using the injector linac was successfully accomplished in the beginning of January 2002. After fine tuning of the machine and cleaning of the vacuum system with the beams from the middle of January to the middle of March, routine operation for users has begun in April. The historical details, commissioning and results of the project are reported. (author)

  14. The Upgraded D0 detector

    Energy Technology Data Exchange (ETDEWEB)

    Abazov, V.M.; Abbott, B.; Abolins, M.; Acharya, B.S.; Adams, D.L.; Adams, M.; Adams, T.; Agelou, M.; Agram, J.-L.; Ahmed, S.N.; Ahn, S.H.; Ahsan, M.; Alexeev, G.D.; Alkhazov, G.; Alton, A.; Alverson, G.; Alves, G.A.; Anastasoaie, M.; Andeen, T.; Anderson, J.T.; Anderson, S.; /Buenos Aires U. /Rio de Janeiro, CBPF /Sao Paulo, IFT /Alberta U.

    2005-07-01

    The D0 experiment enjoyed a very successful data-collection run at the Fermilab Tevatron collider between 1992 and 1996. Since then, the detector has been upgraded to take advantage of improvements to the Tevatron and to enhance its physics capabilities. We describe the new elements of the detector, including the silicon microstrip tracker, central fiber tracker, solenoidal magnet, preshower detectors, forward muon detector, and forward proton detector. The uranium/liquid-argon calorimeters and central muon detector, remaining from Run I, are discussed briefly. We also present the associated electronics, triggering, and data acquisition systems, along with the design and implementation of software specific to D0.

  15. Upgrading Status Of Bandung Triga 2000 Reactor

    International Nuclear Information System (INIS)

    Upgrading Status Of Bandung TRIGA 2000 Reactor. Upgrading of TRIGA Mark II Reactor from 1000 k W to 2000 k W has been done. On June 24, 2000 it has been inaugurated by the Vice President, Madame Megawati Soekarnoputri. The solution of the problems faced in the upgrading should be described here since some experiences got during the process probably are very useful, especially the methods in finishing the project

  16. Pipeline transportation of emerging partially upgraded bitumen

    International Nuclear Information System (INIS)

    The recoverable reserves of Canada's vast oil deposits is estimated to be 335 billion barrels (bbl), most of which are in the Alberta oil sands. Canada was the largest import supplier of crude oil to the United States in 2001, followed by Saudi Arabia. By 2011, the production of oil sands is expected to increase to 50 per cent of Canada's oil, and conventional oil production will decline as more production will be provided by synthetic light oil and bitumen. This paper lists the announced oil sands projects. If all are to proceed, production would reach 3,445,000 bbl per day by 2011. The three main challenges regarding the transportation and marketing of this new production were described. The first is to expand the physical capacity of existing pipelines. The second is the supply of low viscosity diluent (such as natural gas condensate or synthetic diluent) to reduce the viscosity and density of the bitumen as it passes through the pipelines. The current pipeline specifications and procedures to transport partially upgraded products are presented. The final challenge is the projected refinery market constraint to process the bitumen and synthetic light oil into consumer fuel products. These challenges can be addressed by modifying refineries and increasing Canadian access in Petroleum Administration Defense District (PADD) II and IV. The technology for partial upgrading of bitumen to produce pipeline specification oil, reduce diluent requirements and add sales value, is currently under development. The number of existing refineries to potentially accept partially upgraded product is listed. The partially upgraded bitumen will be in demand for additional upgrading to end user products, and new opportunities will be presented as additional pipeline capacity is made available to transport crude to U.S. markets and overseas. The paper describes the following emerging partial upgrading methods: the OrCrude upgrading process, rapid thermal processing, CPJ process for

  17. Upgrading existing evaporators to reduce energy consumption

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    This manual is intended to assist the evaporator engineer who will be performing the technical and economic analyses to determine the most suitable evaporator upgrading technique for his particular plant. Information is included on potentials for upgrading evaporators; correctable operating factors; heat recovery and other improvements in energy use with minor capital investments; upgrading through major capital investments; guidelines for formulating an upgrading program; and new technologies encompassing advanced designs, use of solar and low-grade heat sources, and heat transfer enhancement. A 36 item bibliography is included. (LCL)

  18. Numerical studies of edge localized instabilities in tokamaks

    International Nuclear Information System (INIS)

    A new computational tool, edge localized instabilities in tokamaks equilibria (ELITE), has been developed to help our understanding of short wavelength instabilities close to the edge of tokamak plasmas. Such instabilities may be responsible for the edge localized modes observed in high confinement H-mode regimes, which are a serious concern for next step tokamaks because of the high transient power loads which they can impose on divertor target plates. ELITE uses physical insight gained from analytic studies of peeling and ballooning modes to provide an efficient way of calculating the edge ideal magnetohydrodynamic stability properties of tokamaks. This paper describes the theoretical formalism which forms the basis for the code

  19. KTM Tokamak operation scenarios software infrastructure

    Energy Technology Data Exchange (ETDEWEB)

    Pavlov, V.; Baystrukov, K.; Golobkov, YU.; Ovchinnikov, A.; Meaentsev, A.; Merkulov, S.; Lee, A. [National Research Tomsk Polytechnic University, Tomsk (Russian Federation); Tazhibayeva, I.; Shapovalov, G. [National Nuclear Center (NNC), Kurchatov (Kazakhstan)

    2014-10-15

    One of the largest problems for tokamak devices such as Kazakhstan Tokamak for Material Testing (KTM) is the operation scenarios' development and execution. Operation scenarios may be varied often, so a convenient hardware and software solution is required for scenario management and execution. Dozens of diagnostic and control subsystems with numerous configuration settings may be used in an experiment, so it is required to automate the subsystem configuration process to coordinate changes of the related settings and to prevent errors. Most of the diagnostic and control subsystems software at KTM was unified using an extra software layer, describing the hardware abstraction interface. The experiment sequence was described using a command language. The whole infrastructure was brought together by a universal communication protocol supporting various media, including Ethernet and serial links. The operation sequence execution infrastructure was used at KTM to carry out plasma experiments.

  20. Magnetic sensor for steady state tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Neyatani, Yuzuru; Mori, Katsuharu; Oguri, Shigeru; Kikuchi, Mitsuru [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1996-06-01

    A new type of magnetic sensor has been developed for the measurement of steady state magnetic fields without DC-drift such as integration circuit. The electromagnetic force induced to the current which leads to the sensor was used for the measurement. For the high frequency component which exceeds higher than the vibration frequency of sensor, pick-up coil was used through the high pass filter. From the results using tokamak discharges, this sensor can measure the magnetic field in the tokamak discharge. During {approx}2 hours measurement, no DC drift was observed. The sensor can respond {approx}10ms of fast change of magnetic field during disruptions. We confirm the extension of measured range to control the current which leads to the sensor. (author).

  1. Boundary Plasma Turbulence Simulations for Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Xu, X; Umansky, M; Dudson, B; Snyder, P

    2008-05-15

    The boundary plasma turbulence code BOUT models tokamak boundary-plasma turbulence in a realistic divertor geometry using modified Braginskii equations for plasma vorticity, density (ni), electron and ion temperature (T{sub e}; T{sub i}) and parallel momenta. The BOUT code solves for the plasma fluid equations in a three dimensional (3D) toroidal segment (or a toroidal wedge), including the region somewhat inside the separatrix and extending into the scrape-off layer; the private flux region is also included. In this paper, a description is given of the sophisticated physical models, innovative numerical algorithms, and modern software design used to simulate edge-plasmas in magnetic fusion energy devices. The BOUT code's unique capabilities and functionality are exemplified via simulations of the impact of plasma density on tokamak edge turbulence and blob dynamics.

  2. The Spherical Tokamak MEDUSA for Costa Rica

    Science.gov (United States)

    Ribeiro, Celso; Vargas, Ivan; Guadamuz, Saul; Mora, Jaime; Ansejo, Jose; Zamora, Esteban; Herrera, Julio; Chaves, Esteban; Romero, Carlos

    2012-10-01

    The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R>=3.6, under design[2]) and also the ongoing activities in low temperature plasmas. Courses in plasma physics at undergraduate and post-graduate joint programme levels are regularly conducted. The scientific programme is intend to clarify several issues in relevant physics for conventional and mainly STs, including transport, heating and current drive via Alfv'en wave, and natural divertor STs with ergodic magnetic limiter[3,4]. [1] G.D.Garstka, PhD thesis, University of Wisconsin at Madison, 1997 [2] L.Barillas et al., Proc. 19^th Int. Conf. Nucl. Eng., Japan, 2011 [3] C.Ribeiro et al., IEEJ Trans. Electrical and Electronic Eng., 2012(accepted) [4] C.Ribeiro et al., Proc. 39^th EPS Conf. Contr. Fusion and Plasma Phys., Sweden, 2012

  3. Module description of TOKAMAK equilibrium code MEUDAS

    International Nuclear Information System (INIS)

    The analysis of an axisymmetric MHD equilibrium serves as a foundation of TOKAMAK researches, such as a design of devices and theoretical research, the analysis of experiment result. For this reason, also in JAERI, an efficient MHD analysis code has been developed from start of TOKAMAK research. The free boundary equilibrium code ''MEUDAS'' which uses both the DCR method (Double-Cyclic-Reduction Method) and a Green's function can specify the pressure and the current distribution arbitrarily, and has been applied to the analysis of a broad physical subject as a code having rapidity and high precision. Also the MHD convergence calculation technique in ''MEUDAS'' has been built into various newly developed codes. This report explains in detail each module in ''MEUDAS'' for performing convergence calculation in solving the MHD equilibrium. (author)

  4. Rapidly Moving Divertor Plates In A Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S. Zweben

    2011-05-16

    It may be possible to replace conventional actively cooled tokamak divertor plates with a set of rapidly moving, passively cooled divertor plates on rails. These plates would absorb the plasma heat flux with their thermal inertia for ~10-30 sec, and would then be removed from the vessel for processing. When outside the tokamak, these plates could be cooled, cleaned, recoated, inspected, and then returned to the vessel in an automated loop. This scheme could provide nearoptimal divertor surfaces at all times, and avoid the need to stop machine operation for repair of damaged or eroded plates. We describe various possible divertor plate designs and access geometries, and discuss an initial design for a movable and removable divertor module for NSTX-U.

  5. Module description of TOKAMAK equilibrium code MEUDAS

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Masaei; Hayashi, Nobuhiko; Matsumoto, Taro; Ozeki, Takahisa [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2002-01-01

    The analysis of an axisymmetric MHD equilibrium serves as a foundation of TOKAMAK researches, such as a design of devices and theoretical research, the analysis of experiment result. For this reason, also in JAERI, an efficient MHD analysis code has been developed from start of TOKAMAK research. The free boundary equilibrium code ''MEUDAS'' which uses both the DCR method (Double-Cyclic-Reduction Method) and a Green's function can specify the pressure and the current distribution arbitrarily, and has been applied to the analysis of a broad physical subject as a code having rapidity and high precision. Also the MHD convergence calculation technique in ''MEUDAS'' has been built into various newly developed codes. This report explains in detail each module in ''MEUDAS'' for performing convergence calculation in solving the MHD equilibrium. (author)

  6. Rapidly Moving Divertor Plates In A Tokamak

    International Nuclear Information System (INIS)

    It may be possible to replace conventional actively cooled tokamak divertor plates with a set of rapidly moving, passively cooled divertor plates on rails. These plates would absorb the plasma heat flux with their thermal inertia for ∼10-30 sec, and would then be removed from the vessel for processing. When outside the tokamak, these plates could be cooled, cleaned, recoated, inspected, and then returned to the vessel in an automated loop. This scheme could provide nearoptimal divertor surfaces at all times, and avoid the need to stop machine operation for repair of damaged or eroded plates. We describe various possible divertor plate designs and access geometries, and discuss an initial design for a movable and removable divertor module for NSTX-U.

  7. Scattering of diffracting beams of electron cyclotron waves by random density fluctuations in inhomogeneous plasmas

    Science.gov (United States)

    Weber, Hannes; Maj, Omar; Poli, Emanuele

    2015-03-01

    The physics and first results of the new WKBeam code for electron cyclotron beams in tokamak plasmas are presented. This code is developed on the basis of a kinetic radiative transfer model which is general enough to account for the effects of diffraction and density fluctuations on the beam. Our preliminary numerical results show a significant broadening of the power deposition profile in ITER due to scattering from random density fluctuations at the plasma edge, while such scattering effects are found to be negligible in medium-size tokamaks like ASDEX upgrade.

  8. ALICE Upgrades: Plans and Potentials

    CERN Document Server

    Tieulent, Raphael

    2015-01-01

    The ALICE collaboration consolidated and completed the installation of current detectors during LS1 with the aim to accumulate 1 nb$^{-1}$ of Pb-Pb collisions during Run 2 corresponding to about 10 times the Run 1 integrated luminosity. In parallel, the ALICE experiment has a rich detector upgrade programme scheduled during the second LHC long shutdown (LS2, 2018-2019) in order to fully exploit the LHC Runs 3 and 4. The main objectives of this programme are: improving the tracking precision and enabling the read-out of all Pb-Pb interactions at a rate of up to 50 kHz, with the goal to record an integrated luminosity of 10 nb$^{-1}$ after LS2 in minimum-bias trigger mode. This sample would represent an increase by a factor of one hundred with respect to the minimum-bias sample expected during Run 2. The implementation of this upgrade programme, foreseen in LS2, includes: a new low-material Inner Tracking System at central rapidity with a forward rapidity extension to add vertexing capabilities to the current M...

  9. MPTS Operation and Recent Upgrade

    Science.gov (United States)

    Leblanc, B. P.; Diallo, A.; Labik, G.; Stevens, D. R.

    2011-10-01

    NSTX's Multi-Point Thomson Scattering (MPTS) diagnostic has supported plasma operation for over ten years, during which time a phased implementation has been pursued. The measurements span the horizontal midplane covering around 90 percent of the full-bore confined plasma and the scrape-off layer (SOL). While beginning with one 30-Hz Nd:YAG laser and 10 radial positions, MPTS has operated with a second laser - combined frequency of 60 Hz - and 30 radial positions during the past six years. A recent upgrade brings the total number of radial positions to 42. While most of the 12 new channels are set to improve spatial resolution in the pedestal and internal transport barrier (ITB) regions, a limited number of extra channels have been added to the inner edge and the SOL. Many of the new channels resulted from the splitting of existing fiber bundles, an option that had been left open in MPTS's original design. The 42-channel configuration is planned to begin operation during the 2011 NSTX experimental run. Experimental results will be presented. Future plans for the upcoming NSTX center-stack upgrade will be discussed. U.S. Dept. of Energy Contract No. DE-AC02-09CH11466.

  10. ALICE upgrades its powerful eyes

    CERN Multimedia

    Yuri Kharlov, ALICE Collaboration

    2013-01-01

    The ALICE Photon Spectrometer (PHOS) is a high-resolution photon detector that measures the photons coming out of the extremely hot plasma created in the lead-lead collisions at the LHC. Taking advantage of the long accelerator shut-down, the ALICE teams are now repairing and upgrading the existing modules and getting ready to install the brand-new module in time for the next run. The upgraded PHOS detector will be faster and more stable with wider acceptance and improved photon identification.   PHOS crystal matrix during repair. The key feature and the main complexity of the ALICE PHOS detector is that it operates at a temperature of -25°C, which makes it the second-coldest equipment element at the LHC after the cryogenic superconducting magnets. Since 2009 when it was installed, the PHOS detector, with its cold and warm volumes, has been immersed in airtight boxes to avoid condensation in the cold volumes. The 10,752 lead tungstate crystals of the PHOS were completely insulated fr...

  11. Upgrading of Boundary Dam spillway

    Energy Technology Data Exchange (ETDEWEB)

    McPhail, Gordon; MacMillan, Dave; Smith, Bert [KGS Group, Winnipeg, (Canada); Lacelle, Justin [SaskPower, Regina, (Canada)

    2010-07-01

    An initial dam safety review was performed in 2005 and identified a number of concerns; the most critical were insufficient spillway capacity and deficiencies in the condition of the existing spillways. This paper described the challenges faced by the upgrading operation on the 50 year old Boundary Dam spillway started in 2008. SaskPower retained the KGS Group to increase the design spillway capacity to 1200 m3/s and remedy observed defects. The construction project involved maintaining the reservoir at full supply level while the 20m long spillway chute and stilling basin below were completely replaced. The difficulties came from the need to complete each year's construction such that the spillway could potentially pass spring flood flows. This paper showed that the upgrade measures selected for implementation were developed through close dialogue between the owner and the designer, with valuable input provided by a panel of external experts as well as from contractors participating in the design process.

  12. Analysis Efforts Supporting NSTX Upgrades

    International Nuclear Information System (INIS)

    The National Spherical Torus Experiment (NSTX) is a low aspect ratio, spherical torus (ST) configuration device which is located at Princeton Plasma Physics Laboratory (PPPL) This device is presently being updated to enhance its physics by doubling the TF field to 1 Tesla and increasing the plasma current to 2 Mega-amperes. The upgrades include a replacement of the centerstack and addition of a second neutral beam. The upgrade analyses have two missions. The first is to support design of new components, principally the centerstack, the second is to qualify existing NSTX components for higher loads, which will increase by a factor of four. Cost efficiency was a design goal for new equipment qualification, and reanalysis of the existing components. Showing that older components can sustain the increased loads has been a challenging effort in which designs had to be developed that would limit loading on weaker components, and would minimize the extent of modifications needed. Two areas representing this effort have been chosen to describe in more details: analysis of the current distribution in the new TF inner legs, and, second, analysis of the out-of-plane support of the existing TF outer legs.

  13. EU Integrated Tokamak Modelling (ITM) Task Force

    Institute of Scientific and Technical Information of China (English)

    A Becoulet

    2007-01-01

    @@ At the end of 2003, the European Fusion Development Agreement (EFDA) structure set-up a long-term European task force (TF) in charge of "co-ordinating the development of a coherent set of validated simulation tools for the purpose of benchmarking on existing tokamak experiments, with the ultimate aim of providing a comprehensive simulation package for ITER plasmas" [http://www.efda-taskforce-itm.org/].

  14. Smaller coil systems for tokamak reactors

    International Nuclear Information System (INIS)

    Ripple reduction by ferro-magnetic iron shielding is used to reduce the size of the toroidal field coils down to 7.8 by 10.4 m bore for a commercial tokamak reactor design with plasma parameters similar to STARFIRE. For maximum effectiveness, it is found that the blocks of ferromagnetic iron shielding should have triangular cross section and should be placed as close to the plasma as possible

  15. Resistive interchange instability in reversed shear tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Furukawa, Masaru; Nakamura, Yuji; Wakatani, Masahiro [Graduate School of Energy Science, Kyoto University, Uji, Kyoto (Japan)

    1999-04-01

    Resistive interchange modes become unstable due to the magnetic shear reversal in tokamaks. In the present paper, the parameter dependences, such as q (safety factor) profile and the magnetic surface shape are clarified for improving the stability, using the local stability criterion. It is shown that a significant reduction of the beta limit is obtained for the JT-60U reversed shear configuration with internal transport barrier, since the local pressure gradient increases. (author)

  16. Tore Supra. Basic design Tokamak system

    International Nuclear Information System (INIS)

    This document describes the basic design for the main components of the Tokamak system of Tora Supra. As such, it focuses on the engineering problems, and refers to last year report on Tora Supra (EUR-CEA-1021) for objectives and experimental programme of the apparatus on one hand, and for qualifying tests of the main technical solutions on the other hand. Superconducting toroidal field coil system, vacuum vessels and radiation shields, poloidal field system and cryogenic system are described

  17. Self-Organized Stationary States of Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Jardin, S. C. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Ferraro, N. [General Atomics, San Diego, CA (United States); Krebs, I. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Max-Plank-Institut fur Plasmaphysik, Garching, Germany

    2015-11-01

    We demonstrate that in a 3D resistive magnetohydrodynamic simulation, for some parameters it is possible to form a stationary state in a tokamak where a saturated interchange mode in the center of the discharge drives a near helical flow pattern that acts to nonlinearly sustain the configuration by adjusting the central loop voltage through a dynamo action. This could explain the physical mechanism for maintaining stationary nonsawtoothing "hybrid" discharges, often referred to as "flux pumping."

  18. Self-Organized Stationary States of Tokamaks.

    Science.gov (United States)

    Jardin, S C; Ferraro, N; Krebs, I

    2015-11-20

    We demonstrate that in a 3D resistive magnetohydrodynamic simulation, for some parameters it is possible to form a stationary state in a tokamak where a saturated interchange mode in the center of the discharge drives a near helical flow pattern that acts to nonlinearly sustain the configuration by adjusting the central loop voltage through a dynamo action. This could explain the physical mechanism for maintaining stationary nonsawtoothing "hybrid" discharges, often referred to as "flux pumping."

  19. Microtearing modes and anomalous transport in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Drake, J.F.; Gladd, N.T.; Liu, C.S.; Chang, C.L.

    1980-04-14

    Microtearing (high-m) modes driven by the electron temperature gradient are found to be unstable for present tokamak parameters. A self-consistent calculation of the nonlinear saturation of this instability yields magnetic fluctuations vertical-barBvertical-bar/B approx. = rho/sub e//L/sub T/. The associated crossfield electron thermal conductivity is shown to be inversely proportional to density, consistent with Alcator scaling, and comparable in magnitude with that inferred from experiments.

  20. Neoclassical transport in high [beta] tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Cowley, S.C.

    1992-12-01

    Neoclassical, transport in high [beta] large aspect ratio tokamaks is calculated. The variational method introduced by Rosenbluth, et al., is used to calculate the full Onsager matrix in the banana regime. These results are part of a continuing study of the high [beta] large aspect ratio equilibria introduced in Cowley, et al. All the neoclassical coefficients are reduced from their nominal low [beta] values by a factor ([var epsilon]/q[sup 2][beta])[sup [1/2