WorldWideScience

Sample records for frascati tokamak upgrade

  1. Analysis of images from videocameras in the Frascati Tokamak Upgrade tokamak

    International Nuclear Information System (INIS)

    De Angelis, R.; Migliori, S.; Borioni, S.; Bracco, G.; Pierattini, S.; Perozziello, A.

    2004-01-01

    The plasma edge interaction in FTU tokamak is monitored by wide angle videocameras. Data are acquired as movies or single frames at a rate of 50 frames/s. The images show interesting features of the plasma such as the presence of Marfes or runaways and give useful information on the status of large parts of the vacuum vessel and toroidal limiter. Due to the large number of data available visual inspection of the movies is often insufficient to correlate the images to the experimental findings. This article illustrates a number of applications developed in order to correlate the images with plasma signals and to search the image database for specific features relevant to the discharge

  2. 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

  3. 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 i...

  4. JUST: Joint Upgraded Spherical Tokamak

    International Nuclear Information System (INIS)

    Azizov, E.A.; Dvorkin, N.Ya.; Filatov, O.G.

    1997-01-01

    The main goals, ideas and the programme of JUST, spherical tokamak (ST) for the plasma burn investigation, are presented. The place and prospects of JUST in thermonuclear investigations are discussed. (author)

  5. The magnet system of the Tokamak T-15 upgrade

    International Nuclear Information System (INIS)

    Khvostenko, P.P.; Azizov, E.A.; Alfimov, D.E.; Belyakov, V.A.; Bondarchuk, E.N.; Chudnovsky, A.N.; Dokuka, V.N.; Kavin, A.A.; Khayrutdinov, R.R.; Khokhlov, M.V.; Kitaev, B.A.; Krasnov, S.V.; Maximova, I.I.; Labusov, A.N.; Lukash, V.E.; Mineev, A.B.; Muratov, V.P.

    2015-01-01

    Highlights: • T-15U project is the initial technical base for creating fusion neutron sources. • Magnet system of T-15U will confine the hot plasma in the divertor configuration. • Toroidal magnetic field at the plasma axis is 2 T. • T-15U should begin operations in 2016. - Abstract: Presently, the Tokamak T-15 is being upgraded. The magnet system of the Tokamak T-15 upgrade will obtain and confine the hot plasma in the divertor configuration. Plasma parameters are a major radius of 1.48 m, a minor radius of 0.67 m, an elongation of 1.7–1.9 and a triangularity of 0.3–0.4. The magnet system includes the toroidal winding and the poloidal magnet system. The poloidal magnet system generates the divertor with single null and double null magnetic configurations. The power supply system provides the necessary current scenarios in the windings of the magnet system. All elements of the magnet system will be manufactured by the end of 2015. The Tokamak T-15 upgrade should begin operations in 2016.

  6. The magnet system of the Tokamak T-15 upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Khvostenko, P.P., E-mail: ppkhvost@rambler.ru [National Research Centre ‘Kurchatov Institute’, Institute of Tokamak Physics, Kurchatov sq. 1, 123182 Moscow (Russian Federation); Azizov, E.A.; Alfimov, D.E. [National Research Centre ‘Kurchatov Institute’, Institute of Tokamak Physics, Kurchatov sq. 1, 123182 Moscow (Russian Federation); Belyakov, V.A.; Bondarchuk, E.N. [Joint Stock Company “D.V. Efremov Institute of Electrophysical Apparatus”, Metallostroy, 196641 St. Petersburg (Russian Federation); Chudnovsky, A.N.; Dokuka, V.N. [National Research Centre ‘Kurchatov Institute’, Institute of Tokamak Physics, Kurchatov sq. 1, 123182 Moscow (Russian Federation); Kavin, A.A. [Joint Stock Company “D.V. Efremov Institute of Electrophysical Apparatus”, Metallostroy, 196641 St. Petersburg (Russian Federation); Khayrutdinov, R.R. [National Research Centre ‘Kurchatov Institute’, Institute of Tokamak Physics, Kurchatov sq. 1, 123182 Moscow (Russian Federation); Khokhlov, M.V.; Kitaev, B.A.; Krasnov, S.V.; Maximova, I.I.; Labusov, A.N. [Joint Stock Company “D.V. Efremov Institute of Electrophysical Apparatus”, Metallostroy, 196641 St. Petersburg (Russian Federation); Lukash, V.E. [National Research Centre ‘Kurchatov Institute’, Institute of Tokamak Physics, Kurchatov sq. 1, 123182 Moscow (Russian Federation); Mineev, A.B.; Muratov, V.P. [Joint Stock Company “D.V. Efremov Institute of Electrophysical Apparatus”, Metallostroy, 196641 St. Petersburg (Russian Federation); and others

    2015-10-15

    Highlights: • T-15U project is the initial technical base for creating fusion neutron sources. • Magnet system of T-15U will confine the hot plasma in the divertor configuration. • Toroidal magnetic field at the plasma axis is 2 T. • T-15U should begin operations in 2016. - Abstract: Presently, the Tokamak T-15 is being upgraded. The magnet system of the Tokamak T-15 upgrade will obtain and confine the hot plasma in the divertor configuration. Plasma parameters are a major radius of 1.48 m, a minor radius of 0.67 m, an elongation of 1.7–1.9 and a triangularity of 0.3–0.4. The magnet system includes the toroidal winding and the poloidal magnet system. The poloidal magnet system generates the divertor with single null and double null magnetic configurations. The power supply system provides the necessary current scenarios in the windings of the magnet system. All elements of the magnet system will be manufactured by the end of 2015. The Tokamak T-15 upgrade should begin operations in 2016.

  7. Divertor radiation in the ASDEX upgrade tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Sehmer, Till; Bernert, Matthias; Koll, Juergen; Meister, Hans; Wischmeier, Marco; Fantz, Ursel [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2, 85748 Garching (Germany); Reimold, Felix [Forschungszentrum Juelich GmbH, Institut fuer Energie- und Klimaforschung - Plasmaphysik, 52425 Juelich (Germany); Collaboration: The ASDEX Upgrade Team

    2016-07-01

    To reduce in ITER the expected power flux density onto the divertor target, the plasma-wall interaction in the divertor needs to be strongly reduced. The fundamental path to achieve this is using radiation from seeded impurities, whereas the localization of this radiation (e.g. inside/outside confined region), which could have an impact onto the power balance, is a key challenge. The absolute radiated power distribution can be measured by foil bolometers. To study at the ASDEX Upgrade tungsten divertor the localization and quantification of radiation, the respective line of sight density of the bolometers has been improved by two additional cameras. The divertor radiation enhanced by nitrogen (N{sub 2}) seeding has been investigated, using variations of (1) the external heating power or (2) the N{sub 2} seeding rate. While in both cases the inner divertor stays fully detached, measurements indicate that the region of dominant radiation moves from the inner divertor through the X-Point into the confined region. In the outer divertor however, the measurements indicate either an immediate upwards shift or a continuous movement of the radiation away from the target, depending on experimental conditions.

  8. A cross-tokamak neural network disruption predictor for the JET and ASDEX Upgrade tokamaks

    Science.gov (United States)

    Windsor, C. G.; Pautasso, G.; Tichmann, C.; Buttery, R. J.; Hender, T. C.; EFDA Contributors, JET; ASDEX Upgrade Team

    2005-05-01

    First results are reported on the prediction of disruptions in one tokamak, based on neural networks trained on another tokamak. The studies use data from the JET and ASDEX Upgrade devices, with a neural network trained on just seven normalized plasma parameters. In this way, a simple single layer perceptron network trained solely on JET correctly anticipated 67% of disruptions on ASDEX Upgrade in advance of 0.01 s before the disruption. The converse test led to a 69% success rate in advance of 0.04 s before the disruption in JET. Only one overall time scaling parameter is allowed between the devices, which can be introduced from theoretical arguments. Disruption prediction performance based on such networks trained and tested on the same device shows even higher success rates (JET, 86%; ASDEX Upgrade, 90%), despite the small number of inputs used and simplicity of the network. It is found that while performance for networks trained and tested on the same device can be improved with more complex networks and many adjustable weights, for cross-machine testing the best approach is a simple single layer perceptron. This offers the basis of a potentially useful technique for large future devices such as ITER, which with further development might help to reduce disruption frequency and minimize the need for a large disruption campaign to train disruption avoidance systems.

  9. A cross-tokamak neural network disruption predictor for the JET and ASDEX Upgrade tokamaks

    International Nuclear Information System (INIS)

    Windsor, C.G.; Buttery, R.J.; Hender, T.C.; Pautasso, G.; Tichmann, C.

    2005-01-01

    First results are reported on the prediction of disruptions in one tokamak, based on neural networks trained on another tokamak. The studies use data from the JET and ASDEX Upgrade devices, with a neural network trained on just seven normalized plasma parameters. In this way, a simple single layer perceptron network trained solely on JET correctly anticipated 67% of disruptions on ASDEX Upgrade in advance of 0.01 s before the disruption. The converse test led to a 69% success rate in advance of 0.04 s before the disruption in JET. Only one overall time scaling parameter is allowed between the devices, which can be introduced from theoretical arguments. Disruption prediction performance based on such networks trained and tested on the same device shows even higher success rates (JET, 86%; ASDEX Upgrade, 90%), despite the small number of inputs used and simplicity of the network. It is found that while performance for networks trained and tested on the same device can be improved with more complex networks and many adjustable weights, for cross-machine testing the best approach is a simple single layer perceptron. This offers the basis of a potentially useful technique for large future devices such as ITER, which with further development might help to reduce disruption frequency and minimize the need for a large disruption campaign to train disruption avoidance systems

  10. Impurity seeding in ASDEX upgrade tokamak modeled by COREDIV code

    Energy Technology Data Exchange (ETDEWEB)

    Galazka, K.; Ivanova-Stanik, I.; Czarnecka, A.; Zagoerski, R. [Institute of Plasma Physics and Laser Microfusion, Warsaw (Poland); Bernert, M.; Kallenbach, A. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Collaboration: ASDEX Upgrade Team

    2016-08-15

    The self-consistent COREDIV code is used to simulate discharges in a tokamak plasma, especially the influence of impurities during nitrogen and argon seeding on the key plasma parameters. The calculations are performed with and without taking into account the W prompt redeposition in the divertor area and are compared to the experimental results acquired on ASDEX Upgrade tokamak (shots 29254 and 29257). For both impurities the modeling shows a better agreement with the experiment in the case without prompt redeposition. It is attributed to higher average tungsten concentration, which on the other hand seriously exceeds the experimental value. By turning the prompt redeposition process on, the W concentration is lowered, what, in turn, results in underestimation of the radiative power losses. By analyzing the influence of the transport coefficients on the radiative power loss and average W concentration it is concluded that the way to compromise the opposing tendencies is to include the edge-localized mode flushing mechanism into the code, which dominates the experimental particle and energy balance. Also performing the calculations with both anomalous and neoclassical diffusion transport mechanisms included is suggested. (copyright 2016 The Authors. Contributions to Plasma Physics published by Wiley-VCH Verlag GmbH and Co. KGaA Weinheim. This)

  11. Overview of the TCV tokamak program: scientific progress and facility upgrades

    DEFF Research Database (Denmark)

    Coda, S.; Ahn, J.; Albanese, R.

    2017-01-01

    The TCV tokamak is augmenting its unique historical capabilities (strong shaping, strong electron heating) with ion heating, additional electron heating compatible with high densities, and variable divertor geometry, in a multifaceted upgrade program designed to broaden its operational range with...

  12. Overview of the TCV tokamak program : scientific progress and facility upgrades

    NARCIS (Netherlands)

    Coda, S.; Ahn, J.; Albanese, R.; Alberti, S.; Alessi, E.; Allan, S.; Anand, H.; Anastassiou, G.; Andrèbe, Y.; Angioni, C.; Ariola, M.; Bernert, M.; Beurskens, M.N.A.; Bin, W.; Blanchard, P.; Blanken, T.C.; Boedo, J.A.; Bolzonella, T.; Bouquey, F.; Braunmüller, F.H.; Bufferand, H.; Buratti, P.; Calabró, G.; Camenen, Y.; Carnevale, D.; Carpanese, F.; Causa, F.; Cesario, R.; Chapman, I.T.; Chellai, O.; Choi, D.; Cianfarani, C.; Ciraolo, G.; Citrin, J.; Costea, S.; Crisanti, F.; Cruz, N.; Czarnecka, A.; Decker, J.; De Masi, G.; De Tommasi, G.; Douai, D.; Dunne, M.; Duval, B.P.; Eich, T.; Elmore, S.; Esposito, B.; Faitsch, M.; Fasoli, A.; Fedorczak, N.; Felici, F.; Février, O.; Ficker, O.; Fietz, S.; Fontana, M.; Frassinetti, L.; Furno, I.; Galeani, S.; Gallo, A.; Galperti, C.; Garavaglia, S.; Garrido, I.; Geiger, B.; Giovannozzi, E.; Gobbin, M.; Goodman, T.P.; Gorini, G.; Gospodarczyk, M.; Granucci, G.; Graves, J.P.; Guirlet, R.; Hakola, A.; Ham, C.; Harrison, J.; Hawke, J.; Hennequin, P.; Hnat, B.; Hogeweij, D.; Hogge, J.- P.; Honoré, C.; Hopf, C.; Horáček, J.; Huang, Z.; Igochine, V.; Innocente, P.; Ionita-Schrittwieser, C.; Isliker, H.; Jacquier, R.; Jardin, A.; Kamleitner, J.; Karpushov, A.; Keeling, D.L.; Kirneva, N.; Kong, M.; Koubiti, M.; Kovacic, J.; Krämer-Flecken, A.; Krawczyk, N.; Kudlacek, O.; Labit, B.; Lazzaro, E.; Le, H.B.; Lipschultz, B.; Llobet, X.; Lomanowski, B.; Loschiavo, V.P.; Lunt, T.; Maget, P.; Maljaars, E.; Malygin, A.; Maraschek, M.; Marini, C.; Martin, P.; Martin, Y.; Mastrostefano, S.; Maurizio, R.; Mavridis, M.; Mazon, D.; McAdams, R.; McDermott, R.; Merle, A.; Meyer, H.; Militello, F.; Miron, I.G.; Molina Cabrera, P.A.; Moret, J.M.; Moro, A.; Moulton, D.; Naulin, V.; Nespoli, F.; Nielsen, A.H.; Nocente, M.; Nouailletas, R.; Nowak, S.; Odstrčil, T.; Papp, G.; Papřok, R.; Pau, A.; Pautasso, G.; Pericoli Ridolfini, V.; Piovesan, P.; Piron, C.; Pisokas, T.; Porte, L.; Preynas, M.; Ramogida, G.; Rapson, C.; Rasmussen, J.J.; Reich, M.; Reimerdes, H.; Reux, C.; Ricci, P.; Rittich, D.; Riva, F.; Robinson, T.; Saarelma, S.; Saint-Laurent, F.; Sauter, O.; Scannell, R.; Schlatter, Ch.; Schneider, B.; Schneider, P.; Schrittwieser, R.; Sciortino, F.; Sertoli, M.; Sheikh, U.; Sieglin, B.; Silva, M.; Sinha, J.; Sozzi, C.; Spolaore, M.; Stange, T.; Stoltzfus-Dueck, T.; Tamain, P.; Teplukhina, A.; Testa, D.; Theiler, C.; Thornton, A.; Tophøj, L.; Tran, M.Q.; Tsironis, C.; Tsui, C.; Uccello, A.; Vartanian, S.; Verdoolaege, G.; Verhaegh, K.H.A.; Vermare, L.; Vianello, N.; Vijvers, W.A.J.; Vlahos, L.; Vu, N.M.T.; Walkden, N.; Wauters, T.; Weisen, H.; Wischmeier, M.; Zestanakis, P.; Zuin, M.

    2017-01-01

    The TCV tokamak is augmenting its unique historical capabilities (strong shaping, strong electron heating) with ion heating, additional electron heating compatible with high densities, and variable divertor geometry, in a multifaceted upgrade program designed to broaden its operational range without

  13. Upgraded data service system for HT-7 tokamak

    International Nuclear Information System (INIS)

    Qu Lianzheng; Luo Jiarong; Wei Peijie; Li Guiming; 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. (authors)

  14. Upgraded ECE radiometer on the Tore Supra Tokamak

    International Nuclear Information System (INIS)

    Segui, J.L.; Molina, D.; Goniche, M.; Maget, P.; Udintsev, V.S.; Kraemer-Flecken, A.

    2004-01-01

    An upgraded 32-channel heterodyne radiometer, 1 GHz spaced, is used on the Tore-Supra tokamak to measure the electron cyclotron emission (ECE) in the frequency range 78-110 GHz for the ordinary mode (O1) and 94-126.5 GHz for the extraordinary mode (X2). From now radial resolution is essentially limited by ECE relativistic effects related to electron temperature and density, not by the channels frequency spacing. For example, this leads to precise electron temperature mapping during magneto hydrodynamic activities (MHD). In the equatorial plane, we use a dual polarisation Gaussian optics lens antenna. It has low spreading and a perpendicular line-of-sight that gives ECE measurements very low refraction and Doppler effects. Assuming that the plasma is a black body and there is no overlap between ECE harmonics, one can deduce the electron temperature profile by using the first harmonic ordinary mode (O1) or the second harmonic extraordinary mode (X2). The principle radio frequency emitter (RF) has its frequencies down shifted into intermediary frequencies (IF) that span from 2 to 18 GHz in the single side band mode (SSB). It is amplified by low noise IF amplifiers before forming channels. A separate O/X mode RF front-end allows the use of an IF electronic mode selector. This gives the potentiality of simultaneous O/X mode measurements in the 94-110 GHz. RF and IF filters reject the gyrotron frequency (118 GHz) in order to perform electron temperature measurements during electron cyclotron resonance heated plasmas. A precise absolute spectral calibration is performed outside the tokamak vacuum vessel by using a 600 deg C black body hot source, a double coherent digital signal averaging (trigger, turn and clock) on the waveform generated by a mechanical chopper, and a simulated tokamak window. The use of differential electronics and strong electromagnetic shielding improves also the calibration precision. The fast and slow data acquisition systems are free of aliasing

  15. Upgraded ECE radiometer on the Tore Supra Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Segui, J.L.; Molina, D.; Goniche, M.; Maget, P.; Udintsev, V.S. [Association Euratom-CEA, Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Antar, G.Y. [Center for Energy Research, UCSD, La Jolla CA (United States); Kraemer-Flecken, A. [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Plasmaphysik

    2004-07-01

    An upgraded 32-channel heterodyne radiometer, 1 GHz spaced, is used on the Tore-Supra tokamak to measure the electron cyclotron emission (ECE) in the frequency range 78-110 GHz for the ordinary mode (O1) and 94-126.5 GHz for the extraordinary mode (X2). From now radial resolution is essentially limited by ECE relativistic effects related to electron temperature and density, not by the channels frequency spacing. For example, this leads to precise electron temperature mapping during magneto hydrodynamic activities (MHD). In the equatorial plane, we use a dual polarisation Gaussian optics lens antenna. It has low spreading and a perpendicular line-of-sight that gives ECE measurements very low refraction and Doppler effects. Assuming that the plasma is a black body and there is no overlap between ECE harmonics, one can deduce the electron temperature profile by using the first harmonic ordinary mode (O1) or the second harmonic extraordinary mode (X2). The principle radio frequency emitter (RF) has its frequencies down shifted into intermediary frequencies (IF) that span from 2 to 18 GHz in the single side band mode (SSB). It is amplified by low noise IF amplifiers before forming channels. A separate O/X mode RF front-end allows the use of an IF electronic mode selector. This gives the potentiality of simultaneous O/X mode measurements in the 94-110 GHz. RF and IF filters reject the gyrotron frequency (118 GHz) in order to perform electron temperature measurements during electron cyclotron resonance heated plasmas. A precise absolute spectral calibration is performed outside the tokamak vacuum vessel by using a 600 deg C black body hot source, a double coherent digital signal averaging (trigger, turn and clock) on the waveform generated by a mechanical chopper, and a simulated tokamak window. The use of differential electronics and strong electromagnetic shielding improves also the calibration precision. The fast and slow data acquisition systems are free of aliasing

  16. Advanced tokamak research with integrated modeling in JT-60 Upgrade

    International Nuclear Information System (INIS)

    Hayashi, N.

    2010-01-01

    Researches on advanced tokamak (AT) have progressed with integrated modeling in JT-60 Upgrade [N. Oyama et al., Nucl. Fusion 49, 104007 (2009)]. Based on JT-60U experimental analyses and first principle simulations, new models were developed and integrated into core, rotation, edge/pedestal, and scrape-off-layer (SOL)/divertor codes. The integrated models clarified complex and autonomous features in AT. An integrated core model was implemented to take account of an anomalous radial transport of alpha particles caused by Alfven eigenmodes. It showed the reduction in the fusion gain by the anomalous radial transport and further escape of alpha particles. Integrated rotation model showed mechanisms of rotation driven by the magnetic-field-ripple loss of fast ions and the charge separation due to fast-ion drift. An inward pinch model of high-Z impurity due to the atomic process was developed and indicated that the pinch velocity increases with the toroidal rotation. Integrated edge/pedestal model clarified causes of collisionality dependence of energy loss due to the edge localized mode and the enhancement of energy loss by steepening a core pressure gradient just inside the pedestal top. An ideal magnetohydrodynamics stability code was developed to take account of toroidal rotation and clarified a destabilizing effect of rotation on the pedestal. Integrated SOL/divertor model clarified a mechanism of X-point multifaceted asymmetric radiation from edge. A model of the SOL flow driven by core particle orbits which partially enter the SOL was developed by introducing the ion-orbit-induced flow to fluid equations.

  17. Real-time horizontal position control for Aditya-upgrade tokamak

    International Nuclear Information System (INIS)

    Kumar, Rohit; Ghosh, Joydeep; Tanna, Rakesh L.

    2015-01-01

    Position of plasma column is required to be controlled in real time for improved operation of any tokamak. A PID based system for real-time horizontal plasma position control has been designed for Aditya Upgrade tokamak. Modelling of transfer functions of actuators, plasma and diagnostic system are carried out for ADITYA-U tokamak. The PID controller is optimized using MATLAB-SIMULINK for horizontal position control. Further feed-forward loop is implemented where disturbance due to density variation is suppressed, which results in improved performance as compared to conventional PID operation. In this paper the detailed design of the whole system for real time control of plasma horizontal position in Aditya Upgrade tokamak is presented. (author)

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

    NARCIS (Netherlands)

    M. García-Muñoz,; 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

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

    NARCIS (Netherlands)

    García-Munoz, M.; Hicks, N.; Voornveld, van 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 Alfv´en 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 Alfv´en eigenmodes (TAEs) and

  20. Limiter and divertor systems - conceptual and mechanical design for Aditya Tokamak upgrade

    International Nuclear Information System (INIS)

    Patel, Kaushal; Rathod, Kulav; Jadeja, Kumarpalsinh A.

    2015-01-01

    Existing Aditya tokamak with limiter configuration is being upgraded into a machine to have both the limiter and divertor configurations. Necessary modifications have been carried out to accommodate divertor coils by replacing the old vacuum vessel with a new circular section vacuum vessel. The upgraded Aditya tokamak will have different set of limiters and divertors, such as Safety limiter, Toroidal Inner limiter, outer limiter of smaller toroidal extent, Upper and lower divertor plates. The limiter and divertor locations inside the Aditya tokamak upgrade are decided based on the numerical simulation of the plasma equilibrium profiles. Initially graphite will be used as plasma facing material (PFM) in all the limiter and divertor plates. The dimensions of the limiter and divertor tiles are decided based on their installation inside the vacuum vessel as well as on the total plasma heat loads (∼ 1 MW) falling on them. Depending upon the heat loads; the thickness of graphite tiles for limiter and divertor plates is estimated. Shaped graphite tiles will be fixed on specially designed support structures made out of SS-304L inside the torus shaped vacuum vessel. In this paper mechanical structural design of limiter and divertor of Aditya Upgrade Tokamak is presented. (author)

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

    DEFF Research Database (Denmark)

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

    2011-01-01

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

  2. q=1 advanced tokamak experiments in JET and comparison with ASDEX Upgrade

    International Nuclear Information System (INIS)

    Joffrin, E.; Wolf, R.; Alper, B.

    2002-01-01

    The ASDEX Upgrade advanced tokamak scenario with central q close to 1 has been reproduced on JET. For almost identical q profiles, the comparative analysis does show similar features like the fishbone activity and the current profile evolution. In JET, transport analyses indicates that an internal transport barrier (ITB) has been produced. Gradient length criterions based on the ion temperature gradient turbulence stabilization are used to characterize the ITBs in both devices. The trigger of ITBs is associated with rational surfaces in both devices although the underlying physics for this triggering seems different. This experiment has the prospect to get closer to identity experiments between the two tokamaks. (author)

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

    Czech Academy of Sciences Publication Activity Database

    Janky, Filip; Havlíček, Josef; Batista, A.J.N.; Kudláček, Ondřej; Seidl, Jakub; Neto, A.C.; Pipek, Jan; Hron, Martin; Mikulín, Ondřej; Duarte, A.S.; Carvalho, B.B.; Stöckel, Jan; Pánek, Radomír

    2014-01-01

    Roč. 89, č. 3 (2014), s. 186-194 ISSN 0920-3796 R&D Projects: GA ČR GAP205/11/2470; GA MŠk 7G10072; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : Real-time * Feedback control * Real-time framework * MARTe * COMPASS tokamak Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.152, year: 2014 http://www.sciencedirect.com/science/article/pii/S0920379613007564

  4. TIBER II: an upgraded tokamak igntion/burn experimental reactor

    International Nuclear Information System (INIS)

    Henning, C.D.; Logan, B.G.; Perkins, L.J.

    1986-01-01

    We are disIgning a minimum-size Tokamak ignition/Burn Reactor (TIBER II). This design incorporates physics requirements, neutron wall loading and fluence parameters that will make it compatible with a nuclear testing mission. Reactor relevant physics will be tested by using current drive and steady-state operation. Although the design accommodates several current drive options, including neutral beams, the base case uses a combination of lower hybrid and electron-cyclotron radio frequency power. Minimum neutron shielding, compact structures, high magnet-current densities, and remotely maintainable vacuum seals, all contribute to the compact size

  5. 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.

  6. DIII-D tokamak control and neutral beam computer system upgrades

    International Nuclear Information System (INIS)

    Penaflor, B.G.; McHarg, B.B.; Piglowski, D.A.; Pham, D.; Phillips, J.C.

    2004-01-01

    This paper covers recent computer system upgrades made to the DIII-D tokamak control and neutral beam computer systems. The systems responsible for monitoring and controlling the DIII-D tokamak and injecting neutral beam power have recently come online with new computing hardware and software. The new hardware and software have provided a number of significant improvements over the previous Modcomp AEG VME and accessware based systems. These improvements include the incorporation of faster, less expensive, and more readily available computing hardware which have provided performance increases of up to a factor 20 over the prior systems. A more modern graphical user interface with advanced plotting capabilities has improved feedback to users on the operating status of the tokamak and neutral beam systems. The elimination of aging and non supportable hardware and software has increased overall maintainability. The distinguishing characteristics of the new system include: (1) a PC based computer platform running the Redhat version of the Linux operating system; (2) a custom PCI CAMAC software driver developed by general atomics for the kinetic systems 2115 serial highway card; and (3) a custom developed supervisory control and data acquisition (SCADA) software package based on Kylix, an inexpensive interactive development environment (IDE) tool from borland corporation. This paper provides specific details of the upgraded computer systems

  7. Using Real Time Workshop for rapid and reliable control implementation in the Frascati Tokamak Upgrade Feedback Control System running under RTAI-GNU/Linux

    International Nuclear Information System (INIS)

    Centioli, C.; Iannone, F.; Ledauphin, M.; Panella, M.; Pangione, L.; Podda, S.; Vitale, V.; Zaccarian, L.

    2005-01-01

    The Feedback Control System running at FTU has been recently ported from a commercial platform (based on LynxOS) to an open-source GNU/Linux-based RTAI-LXRT platform, thereby, obtaining significant performance and cost improvements. Based on the new open-source platform, it is now possible to experiment novel control strategies aimed at improving the robustness and accuracy of the feedback control. Nevertheless, the implementation of control ideas still requires a great deal of coding of the control algorithms that, if carried out manually, may be prone to coding errors, therefore time consuming both in the development phase and in the subsequent validation tests consisting of dedicated experiments carried out on FTU. In this paper, we report on recent developments based on Mathworks' Simulink and Real Time Workshop (RTW) packages to obtain a user-friendly environment where the real time code implementing novel control algorithms can be easily generated, tested and validated. Thanks to this new tool, the control designer only needs to specify the block diagram of the control task (namely, a high level and functional description of the new algorithm under consideration) and the corresponding real time code generation and testing is completely automated without any need of dedicated experiments. In the paper, the necessary work carried out to adapt the Real Time Workshop to our RTAI-LXRT context will be illustrated. A necessary re-organization of the previous real time software, aimed at incorporating the code coming from the adapted RTW, will also be discussed. Moreover, we will report on a performance comparison between the code obtained using the automated RTW-based procedure and the hand-written C code, appropriately optimised; at the moment, a preliminary performance comparison consisting of dummy algorithms has shown that the code automatically generated from RTW is faster (about 30% up) than the manually written one. This preliminary result combined with the fact that the use of RTW eliminates the coding workload leads to the conclusion that this approach to control implementation grants significant advantages. According to the FTU experiment program, the actual deployment of the new tool is scheduled for the next experimental campaign

  8. Multifractality in edge localized modes in Japan Atomic Energy Research Institute Tokamak-60 Upgrade

    International Nuclear Information System (INIS)

    Bak, P.E.; Asakura, N.; Miura, Y.; Nakano, T.; Yoshino, R.

    2001-01-01

    The temporal losses of confinement during edge localized modes in the Japan Atomic Energy Research Institute Tokamak-60 Upgrade (JT-60U) show multifractal scaling and the spectra are generally smooth, but in some cases there are signs of discontinuous derivatives. Dynamics of the Sugama-Horton model, interpreted as edge localized modes, also display multifractal scaling. The spectra display singularities in the derivative, which can be interpreted as a phase transition. It is argued that the multifractal spectra of edge localized modes can be used to discriminate between different experimental discharges and validate edge localized mode models

  9. A Neutral Beam for the Lithium Tokamak eXperiment Upgrade (LTX-U)

    Science.gov (United States)

    Merino, Enrique; Majeski, Richard; Kaita, Robert; Kozub, Thomas; Boyle, Dennis; Schmitt, John; Smirnov, Artem

    2015-11-01

    Neutral beam injection into tokamaks is a proven method of plasma heating and fueling. In LTX, high confinement discharges have been achieved with low-recycling lithium walls. To further improve plasma performance, a neutral beam (NB) will be installed as part of an upgrade to LTX (LTX-U). The NB will provide core plasma fueling with up to 700 kW of injected power. Requirements for accommodating the NB include the addition of injection and beam-dump ports onto the vessel and enhancement of the vacuum vessel pumping capability. Because the NB can also serve as a source of neutrals for charge-exchange recombination spectroscopy, ``active'' spectroscopic diagnostics will also be developed. An overview of these plans and other improvements for upgrading LTX to LTX-U will be presented. Supported by US DOE contracts DE-AC02-09CH11466 and DE-AC52-07NA27344.

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

    International Nuclear Information System (INIS)

    Nakamura, Y.; Pautasso, G.; Gruber, O.; Jardin, S.C.

    2002-01-01

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

  11. Measurement of turbulent electron temperature fluctuations on the ASDEX Upgrade tokamak using correlated electron cyclotron emission

    Energy Technology Data Exchange (ETDEWEB)

    Freethy, S. J., E-mail: simon.freethy@ipp.mpg.de [Max Planck Institute for Plasma Physics, 85748 Garching (Germany); Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Conway, G. D.; Happel, T.; Köhn, A. [Max Planck Institute for Plasma Physics, 85748 Garching (Germany); Classen, I.; Vanovac, B. [FOM Institute DIFFER, 5612 AJ Eindhoven (Netherlands); Creely, A. J.; White, A. E. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States)

    2016-11-15

    Turbulent temperature fluctuations are measured on the ASDEX Upgrade tokamak using pairs of closely spaced, narrow-band heterodyne radiometer channels and a standard correlation technique. The pre-detection spacing and bandwidth of the radiometer channel pairs is chosen such that they are physically separated less than a turbulent correlation length, but do not overlap. The radiometer has 4 fixed filter frequency channels and two tunable filter channels for added flexibility in the measurement position. Relative temperature fluctuation amplitudes are observed in a helium plasma to be δT/T = (0.76 ± 0.02)%, (0.67 ± 0.02)%, and (0.59 ± 0.03)% at normalised toroidal flux radius of ρ{sub tor} = 0.82, 0.75, and 0.68, respectively.

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

    International Nuclear Information System (INIS)

    1982-03-01

    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.)

  13. Stability analysis for the Big Dee upgrade of the Doublet III tokamak

    International Nuclear Information System (INIS)

    Helton, F.J.; Luxon, J.L.

    1987-01-01

    Ideal magnetohydrodynamic stability analysis has been carried out for configurations expected in the Big Dee tokamak, an upgrade of the Doublet III tokamak into a non-circular cross-section device which began operation early in 1986. The results of this analysis support theoretical predictions as follows: Since the maximum value of beta stable to ballooning and Mercier modes, which we denote β c , increases with inverse aspect ratio, elongation and triangularity, the Big Dee is particularly suited to obtain high values of β c and there exist high β c Big Dee equilibria for large variations in all relevant plasma parameters. The beta limits for the Big Dee are consistent with established theory as summarized in present scaling laws. High beta Big Dee equilibria are continuously accessible when approached through changes in all relevant input parameters and are structurally stable with respect to variations of input plasma parameters. Big Dee beta limits have a smooth dependence on plasma parameters such as β p and elongation. These calculations indicate that in the actual running of the device the Big Dee high beta equilibria should be smoothly accessible. Theory predicts that the limiting plasma parameters, such as beta, total plasma current and plasma pressure, which can be obtained within the operating limits of the Big Dee are reactor relevant. Thus the Big Dee should be able to use its favourable ideal MHD scaling and controlled plasma shaping to attain reactor relevant parameters in a moderate sized device. (author)

  14. Diagnostics upgrade and capability available for physics study on EAST tokamak

    International Nuclear Information System (INIS)

    Hu Liqun

    2013-01-01

    As a consequence of employment of many new techniques and upgrade of EAST superconducting tokamak to enhance divertor plasma performance, significant achievement has been realized in 2012, including 400s long pulse plasma, stationary 35s H-mode and 3.45s H-mode with only ion cyclotron resonant heating (ICRH) etc. To approach steady-state (SS) operation of high-performance plasmas and address key physics on fusion reactor-relevent subjects, recently, capability of the plasma heating and current drive of EAST machine are doubled with total auxiliary injection power over 20 MW, including new methodology of neutral beam injection (NBI) and electron cyclotron resonant heating (ECRH). Most diagnostics have been upgraded to be more compact and integrated due to limited port window and space available, and new advanced neutral-beam related diagnostics have been developed as well, to provide profile of all key parameters for study and understanding critical issues specific to SS high performance plasma. (author)

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

    International Nuclear Information System (INIS)

    Dunne, Michael G.

    2014-01-01

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Dunne, Michael G.

    2014-02-15

    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.

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

    International Nuclear Information System (INIS)

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

    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 cascades (ACs) have allowed to identify both loss mechanisms. While single ACs and TAEs eject resonant fast-ions in a convective process, the overlapping of AC and TAE spatial structures leads to a large fast-ion diffusion and loss. The threshold for diffusive fast-ion losses depends on the ion energy (gyroradius). Diffusive fast-ion losses with gyroradius ∼70 mm have been observed with a single TAE for local radial displacements of the magnetic field lines larger than ∼2 mm. Multiple frequency chirping ACs cause an enhancement of the diffusive losses. The ACs and TAEs radial structures have been reconstructed by means of cross-correlation techniques between the fast-ion loss detector and the electron cyclotron emission radiometer.

  18. Impact of lithium on the plasma performance in the all-metal-wall tokamak ASDEX upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Lang, P.T.; Moreno Quicios, R.; Arredondo Parra, R.; Ploeckl, B.; McDermott, R.; Neu, R.; Wolfrum, E. [MPI fuer Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Maingi, R.; Mansfield, D.K.; Diallo, A. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Collaboration: ASDEX Upgrade Team

    2016-07-01

    Several tokamaks reported improvement in key plasma parameters concurrent with the presence of lithium in the plasma. At ASDEX Upgrade explorative experiments have been performed to find out if such effects can be observed when operating with an all-metal-wall. A gas gun launcher was developed capable to inject pellets containing about 1.6 x 10{sup 20} Li atoms at 2 Hz. The speed of about 600 m/s is sufficient to achieve core penetration and to create a homogeneous Li concentration of up to 10 %. With a typical sustainment time on the order of 100 ms, only transient Li presence without any pile up was achieved. Deposition of Li on plasma facing components, which remained for several discharges after injection, was observed. This short lived wall conditioning showed beneficial effects during plasma start-up. However, the accompanying surface contamination negatively impacted some diagnostics. The Li impact on the confinement was investigated in a dedicated plasma scenario with a proven sensitivity to nitrogen and helium. In phases with N seeding enhancing the confinement by about 30 %, Li injection resulted in a very modest, transient loss of confinement (about 5 %). No Li impact was found for pure Deuterium plasmas.

  19. Optimized tokamak power exhaust with double radiative feedback in ASDEX Upgrade

    Science.gov (United States)

    Kallenbach, A.; Bernert, M.; Eich, T.; Fuchs, J. C.; Giannone, L.; Herrmann, A.; Schweinzer, J.; Treutterer, W.; the ASDEX Upgrade Team

    2012-12-01

    A double radiative feedback technique has been developed on the ASDEX Upgrade tokamak for optimization of power exhaust with a standard vertical target divertor. The main chamber radiation is measured in real time by a subset of three foil bolometer channels and controlled by argon injection in the outer midplane. The target heat flux is in addition controlled by nitrogen injection in the divertor private flux region using either a thermoelectric sensor or the scaled divertor radiation obtained by a bolometer channel in the outer divertor. No negative interference of the two radiation controllers has been observed so far. The combination of main chamber and divertor radiative cooling extends the operational space of a standard divertor configuration towards high values of P/R. Pheat/R = 14 MW m-1 has been achieved so far with nitrogen seeding alone as well as with combined N + Ar injection, with the time-averaged divertor peak heat flux below 5 MW m-2. Good plasma performance can be maintained under these conditions, namely H98(y,2) = 1 and βN = 3.

  20. 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.

  1. Numerical simulations of fast ion loss measurements induced by magnetic islands in the ASDEX Upgrade tokamak

    International Nuclear Information System (INIS)

    Gobbin, M.; Marrelli, L.; Martin, P.; Fahrbach, H.U.; Garcia-Munoz, M.; Guenter, S.; White, R.B.

    2009-01-01

    A test particle approach, implemented with the Hamiltonian code ORBIT, is used to simulate measurements of fast ion losses induced by magnetic islands in the ASDEX Upgrade tokamak. In particular, the numerical simulations reproduce the toroidal localization of losses and the lost ions pitch angle and energy distribution experimentally measured with the fast ion losses detector (FILD) in the presence of a neoclassical tearing mode (NTM). The simulated NTM induced losses occurring on time scales longer than 100 μs are composed of mainly trapped or barely passing particles, consistently with the slow decay of the experimental signal from one FILD channel after the beam switch-off. The numerical simulations have been performed by taking into account the D-shaped plasma geometry, the collision mechanisms, the losses due to ripple effects and the rotation of the mode. The radial profile of the magnetic perturbation is adjusted in order to match ECE measurements. While statistical properties of FILD measurements are rather well reproduced, the simulated total amount of losses is found to be significantly affected by edge details of the magnetic perturbation as it determines the loss mechanism.

  2. Confinement and transport properties during current ramps in the ASDEX Upgrade tokamak

    Science.gov (United States)

    Fable, E.; Angioni, C.; Hobirk, J.; Pereverzev, G.; Fietz, S.; Hein, T.; ASDEX Upgrade Team

    2011-04-01

    A detailed analysis of experimental data from the ASDEX Upgrade tokamak is carried out to shed light on the properties of confinement and transport in the current ramp-up and ramp-down phases of the plasma discharge. The experimental database is used to identify the relevant ranges of parameters explored during the ramp-up and the ramp-down. The energy confinement time observed in the two ramps displays interesting evolution, in many cases attaining different values at the same current level between ramp-up and ramp-down. The possible reasons for this behaviour are investigated. Interpretative transport simulations are used as a tool to clarify the interplay between different parameters, which are coupled in a non-linear way. In addition, a theory-based transport model is used to understand the behaviour of confinement as observed in the experiment, evidencing the role of both turbulent and neoclassical transport. Linear gyrokinetic calculations are performed to identify the relevant turbulence regime, showing that a broad range of frequencies, in the trapped electron modes (TEMs) and in the ion temperature gradient modes (ITGs) regimes, is explored during both the ramp-up and ramp-down. In the same framework, a quasi-linear model is applied to calculate the value of the local logarithmic density gradient and compare it with the experimental value. Finally, first non-linear simulations of heat transport during the current ramps are presented.

  3. Study of EAST LH antennas coupling at ENEA-Frascati

    Energy Technology Data Exchange (ETDEWEB)

    Panaccione, L.; Mirizzi, F. [Consorzio CREATE, Via Claudio 21, 80125, Napoli (Italy); Ceccuzzi, S.; Cesario, R.; Tuccillo, A. A., E-mail: angelo.tuccillo@enea.it [ENEA for EUROfusion, via E. Fermi 45, 00044 Frascati (RM) (Italy); Ding, B. J.; Li, M.; Liu, F.; Liu, L.; Shan, J. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2015-12-10

    The two Lower Hybrid (LH) launchers of the EAST tokamak have been analysed using some tools available at ENEA-Frascati research centre. The antennas, working at 2.45 and 4.6 GHz, have been assessed in terms of reflection coefficient and launched power spectrum for several plasma loads differing in the electron density profile. Fitting an experimental profile we derived a set of parameterised realistic density profiles to compute the coupling performances of different spectra, launched by considering different phasing between antenna modules. The sensitivity to the tilt of the magnetic field with respect to the equatorial plane as well as to an additional progressive phasing at the mouth due to the toroidal curvature has been studied too. The most suitable operational conditions for the minimization of reflected power and side lobes in the n{sub ||} spectra are identified.

  4. Tokamak

    International Nuclear Information System (INIS)

    Wesson, John.

    1996-01-01

    This book is the first compiled collection about tokamak. At first chapter tokamak is represented from fusion point of view and also the necessary conditions for producing power. The following chapters are represent plasma physics, the specifications of tokamak, plasma heating procedures and problems related to it, equilibrium, confinement, magnetohydrodynamic stability, instabilities, plasma material interaction, plasma measurement and experiments regarding to tokamak; an addendum is also given at the end of the book

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

    International Nuclear Information System (INIS)

    Saarelma, S.

    2000-01-01

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

  6. Overview of the TCV tokamak program: scientific progress and facility upgrades

    Science.gov (United States)

    Coda, S.; Ahn, J.; Albanese, R.; Alberti, S.; Alessi, E.; Allan, S.; Anand, H.; Anastassiou, G.; Andrèbe, Y.; Angioni, C.; Ariola, M.; Bernert, M.; Beurskens, M.; Bin, W.; Blanchard, P.; Blanken, T. C.; Boedo, J. A.; Bolzonella, T.; Bouquey, F.; Braunmüller, F. H.; Bufferand, H.; Buratti, P.; Calabró, G.; Camenen, Y.; Carnevale, D.; Carpanese, F.; Causa, F.; Cesario, R.; Chapman, I. T.; Chellai, O.; Choi, D.; Cianfarani, C.; Ciraolo, G.; Citrin, J.; Costea, S.; Crisanti, F.; Cruz, N.; Czarnecka, A.; Decker, J.; De Masi, G.; De Tommasi, G.; Douai, D.; Dunne, M.; Duval, B. P.; Eich, T.; Elmore, S.; Esposito, B.; Faitsch, M.; Fasoli, A.; Fedorczak, N.; Felici, F.; Février, O.; Ficker, O.; Fietz, S.; Fontana, M.; Frassinetti, L.; Furno, I.; Galeani, S.; Gallo, A.; Galperti, C.; Garavaglia, S.; Garrido, I.; Geiger, B.; Giovannozzi, E.; Gobbin, M.; Goodman, T. P.; Gorini, G.; Gospodarczyk, M.; Granucci, G.; Graves, J. P.; Guirlet, R.; Hakola, A.; Ham, C.; Harrison, J.; Hawke, J.; Hennequin, P.; Hnat, B.; Hogeweij, D.; Hogge, J.-Ph.; Honoré, C.; Hopf, C.; Horáček, J.; Huang, Z.; Igochine, V.; Innocente, P.; Ionita Schrittwieser, C.; Isliker, H.; Jacquier, R.; Jardin, A.; Kamleitner, J.; Karpushov, A.; Keeling, D. L.; Kirneva, N.; Kong, M.; Koubiti, M.; Kovacic, J.; Krämer-Flecken, A.; Krawczyk, N.; Kudlacek, O.; Labit, B.; Lazzaro, E.; Le, H. B.; Lipschultz, B.; Llobet, X.; Lomanowski, B.; Loschiavo, V. P.; Lunt, T.; Maget, P.; Maljaars, E.; Malygin, A.; Maraschek, M.; Marini, C.; Martin, P.; Martin, Y.; Mastrostefano, S.; Maurizio, R.; Mavridis, M.; Mazon, D.; McAdams, R.; McDermott, R.; Merle, A.; Meyer, H.; Militello, F.; Miron, I. G.; Molina Cabrera, P. A.; Moret, J.-M.; Moro, A.; Moulton, D.; Naulin, V.; Nespoli, F.; Nielsen, A. H.; Nocente, M.; Nouailletas, R.; Nowak, S.; Odstrčil, T.; Papp, G.; Papřok, R.; Pau, A.; Pautasso, G.; Pericoli Ridolfini, V.; Piovesan, P.; Piron, C.; Pisokas, T.; Porte, L.; Preynas, M.; Ramogida, G.; Rapson, C.; Rasmussen, J. Juul; Reich, M.; Reimerdes, H.; Reux, C.; Ricci, P.; Rittich, D.; Riva, F.; Robinson, T.; Saarelma, S.; Saint-Laurent, F.; Sauter, O.; Scannell, R.; Schlatter, Ch.; Schneider, B.; Schneider, P.; Schrittwieser, R.; Sciortino, F.; Sertoli, M.; Sheikh, U.; Sieglin, B.; Silva, M.; Sinha, J.; Sozzi, C.; Spolaore, M.; Stange, T.; Stoltzfus-Dueck, T.; Tamain, P.; Teplukhina, A.; Testa, D.; Theiler, C.; Thornton, A.; Tophøj, L.; Tran, M. Q.; Tsironis, C.; Tsui, C.; Uccello, A.; Vartanian, S.; Verdoolaege, G.; Verhaegh, K.; Vermare, L.; Vianello, N.; Vijvers, W. A. J.; Vlahos, L.; Vu, N. M. T.; Walkden, N.; Wauters, T.; Weisen, H.; Wischmeier, M.; Zestanakis, P.; Zuin, M.; the EUROfusion MST1 Team

    2017-10-01

    The TCV tokamak is augmenting its unique historical capabilities (strong shaping, strong electron heating) with ion heating, additional electron heating compatible with high densities, and variable divertor geometry, in a multifaceted upgrade program designed to broaden its operational range without sacrificing its fundamental flexibility. The TCV program is rooted in a three-pronged approach aimed at ITER support, explorations towards DEMO, and fundamental research. A 1 MW, tangential neutral beam injector (NBI) was recently installed and promptly extended the TCV parameter range, with record ion temperatures and toroidal rotation velocities and measurable neutral-beam current drive. ITER-relevant scenario development has received particular attention, with strategies aimed at maximizing performance through optimized discharge trajectories to avoid MHD instabilities, such as peeling-ballooning and neoclassical tearing modes. Experiments on exhaust physics have focused particularly on detachment, a necessary step to a DEMO reactor, in a comprehensive set of conventional and advanced divertor concepts. The specific theoretical prediction of an enhanced radiation region between the two X-points in the low-field-side snowflake-minus configuration was experimentally confirmed. Fundamental investigations of the power decay length in the scrape-off layer (SOL) are progressing rapidly, again in widely varying configurations and in both D and He plasmas; in particular, the double decay length in L-mode limited plasmas was found to be replaced by a single length at high SOL resistivity. Experiments on disruption mitigation by massive gas injection and electron-cyclotron resonance heating (ECRH) have begun in earnest, in parallel with studies of runaway electron generation and control, in both stable and disruptive conditions; a quiescent runaway beam carrying the entire electrical current appears to develop in some cases. Developments in plasma control have benefited from

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-01

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

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

    International Nuclear Information System (INIS)

    Caress, R.W.; Mayo, R.M.; Carter, T.A.

    1995-01-01

    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

  9. Validation of neutral point on JT-60U, Alcator C-Mod and ASDEX-Upgrade tokamaks

    International Nuclear Information System (INIS)

    Nakamura, Yukiharu; Yoshino, Ryuji; Pautasso, Gabriella; Gruber, Otto; Jardin, Stephen

    2002-01-01

    Validation studies of a neutrally balanced vertical plasma position, so-called ''neutral point'', have been carried out by computational simulations and experiments under trilateral Japan-US-EU collaborations. It was clarified that the neutral point, where VDEs (Vertical Displacement Events) are hardly occurred, does exit in the Alcator C-Mod and ASDEX-Upgrade tokamaks as well as the JT-60U, consistent with the simulations. Meanwhile, precise details of the VDE behavior exhibit their own characters according to the individual of the tokamaks such as an up-down asymmetry of plasma shape. Sensitivity of the neutral point to the plasma shape and current profile was also addressed in detail. (author)

  10. Design of test kits for the RF characterization of the PAM antenna of LHCD system for Aditya-upgrade Tokamak

    International Nuclear Information System (INIS)

    Jain, Yogesh M.; Sharma, P.K.; Parmar, P.R.; Ambulkar, K.K.

    2017-01-01

    The Lower Hybrid Current Drive (LHCD) system of the ADITYA-Upgrade tokamak will employ a Passive Active Multijunction (PAM) antenna to launch 250 kW of RF power at 3.7 GHz to drive plasma current non inductively in the tokamak. To evaluate the RF performance of the designed PAM antenna, it is characterized with the help of VNA measurements. The performance of the PAM antenna is mainly decided by the integrated performance of the entire antenna (with a differential phase shift of 270° and equal power distribution between each of the output waveguides) and the performance of mode converter, which transforms input TE 10 mode to TE 30 mode (with a mode purity of 98.5% at the output). This poster thus reports the design and analysis of these testing kits. Also, the test results of PAM antenna obtained by using these test kits would also be presented and discussed in this poster

  11. Plasma radiation dynamics with the upgraded Absolute Extreme Ultraviolet tomographical system in the Tokamak à Configuration Variable

    Energy Technology Data Exchange (ETDEWEB)

    Tal, B.; Nagy, D.; Veres, G. [Institute for Particle and Nuclear Physics, Wigner Research Centre for Physics, Hungarian Academy of Sciences, Association EURATOM, P. O. Box 49, H-1525 Budapest (Hungary); Labit, B.; Chavan, R.; Duval, B. [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, Association EURATOM-Confédération Suisse, EPFL SB CRPP, Station 13, CH-1015 Lausanne (Switzerland)

    2013-12-15

    We introduce an upgraded version of a tomographical system which is built up from Absolute Extreme Ultraviolet-type (AXUV) detectors and has been installed on the Tokamak à Configuration Variable (TCV). The system is suitable for the investigation of fast radiative processes usually observed in magnetically confined high-temperature plasmas. The upgrade consists in the detector protection by movable shutters, some modifications to correct original design errors and the improvement in the data evaluation techniques. The short-term sensitivity degradation of the detectors, which is caused by the plasma radiation itself, has been monitored and found to be severe. The results provided by the system are consistent with the measurements obtained with the usual plasma radiation diagnostics installed on TCV. Additionally, the coupling between core plasma radiation and plasma-wall interaction is revealed. This was impossible with other available diagnostics on TCV.

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

    International Nuclear Information System (INIS)

    Bernert, Matthias

    2013-01-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Maingi, R. [Oak Ridge Associated Universities, TN (United States); Terreault, B. [Inst. National de la Recherche Scientifique, Varennes, Quebec (Canada); Haas, G. [Max Planck Inst. fuer Plasmaphysik, Garching (Germany)] [and others

    1996-06-01

    The authors present a comparison of the wall deuterium retention and plasma fueling requirements of three diverted tokamaks, DIII-D, TdeV, and ASDEX-Upgrade, with different fractions of graphite coverage of stainless steel or Inconel outer walls and different heating modes. Data from particle balance experiments on each tokamak demonstrate well-defined differences in wall retention of deuterium gas, even though all three tokamaks have complete graphite coverage of divertor components and all three are routinely boronized. This paper compares the evolution of the change in wall loading and net fueling efficiency for gas during dedicated experiments without Helium Glow Discharge Cleaning on the DIII-D and TdeV tokamaks. On the DIII-D tokamak, it was demonstrated that the wall loading could be increased by > 1,250 Torr-1 (equivalent to 150 {times} plasma particle content) plasma inventories resulting in an increase in fueling efficiency from 0.08 to 0.25, whereas the wall loading on the TdeV tokamak could only be increased by < 35 Torr-{ell} (equivalent to 50{times} plasma particle content) plasma inventories at a maximum fueling efficiency {approximately} 1. Data from the ASDEX-Upgrade tokamak suggests qualitative behavior of wall retention and fueling efficiency similar to DIII-D.

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

    Energy Technology Data Exchange (ETDEWEB)

    Nowak, S.; Lazzaro, E.; Granucci, G. [Associazione Euratom-CNR sulla Fusione, IFP-CNR, Via R. Cozzi 53, 20125 Milano (Italy); Esposito, B. [Associazione Euratom-CNR sulla Fusione, CR Frascati, C.P. 65, 00044 Frascati (Italy); Maraschek, M.; Zohm, H. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Boltzmannstr.2, 85748 Garching bei Munchen (Germany); Sauter, O.; Brunetti, D. [CRPP, Association Euratom-Confederation Suisse, EPFL, 1015 Lausanne (Switzerland); Collaboration: ASDEX Upgrade Team

    2012-09-15

    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. 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...

  17. The gravitational wave detection experiment in Frascati

    International Nuclear Information System (INIS)

    Maischberger, K.

    1974-01-01

    The Frascati detector is a copy of the Weber (USA) instrument, with a signal/noise ratio improved by a factor 2.5. Computerized data analyses allows a study of the amplitude and phase of the signal. It is observed that vibrations of energy 0.25kT were never detected more than once a day, which leads to the logical conclusion that Weber's claim can only be taken into consideration of his equipment is capable of detecting vibrations below 0.1kT

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

    Energy Technology Data Exchange (ETDEWEB)

    Fietz, Sina Marie Ariane

    2013-12-16

    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

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

    International Nuclear Information System (INIS)

    Fietz, Sina Marie Ariane

    2013-01-01

    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

  20. Development and application of poloidal correlation reflectometry to study turbulent structures in the ASDEX Upgrade tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Prisiazhniuk, Dmitrii

    2017-06-05

    One of the key question of high temperature plasma confinement in a magnetic field is how plasma turbulence influences the radial transport of particles and energy. A better understanding of transport processes caused by turbulence would allow to improve the plasma confinement in fusion devices. To this end a deeper understanding of the mechanisms controlling the development, saturation and stabilization of turbulence is needed. From the experimental point of view a main challenge in these investigations is the measurement of plasma parameters on both small temporal (μs) and spatial (mm) scales. In this thesis a new microwave heterodyne poloidal correlation reflectometry diagnostic has been developed and installed at the ASDEX Upgrade tokamak to investigate the cross-correlation of turbulent density fluctuations. This diagnostic yields information on fundamental turbulence parameters such as the perpendicular propagation velocity v {sub perpendicular} {sub to}, the perpendicular correlation length l {sub perpendicular} {sub to} (characteristic size of the turbulent eddies) and the decorrelation time τ{sub d} (characteristic life time of the turbulent eddies) over a wide range of plasma densities. The inclination of the turbulent eddies α in the poloidal-toroidal plane spanned by the magnetic flux surfaces of a tokamak, being a measure of the magnetic field pitch angle, can also be obtained. The turbulence investigations were performed in low confinement mode (L-mode) plasmas for a range of plasma parameters. All measurements were interpreted taking into account the transfer function of reflectometry in the Born approximation. The results are compared with theoretical predictions and simulations. In the first part of this thesis the inclination and the propagation of turbulent structures are investigated. It is shown that eddies are nearly aligned to the magnetic field line and, therefore, the magnetic field pitch angle can be measured with a precision of about 1

  1. A real-time data acquisition and elaboration system for instabilities control in the FTU tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Alessi, E., E-mail: alessi@ifp.cnr.it [Associazione EURATOM-ENEA, IFP-CNR, Milano (Italy); Boncagni, L. [Associazione EURATOM-ENEA, C.R. Frascati (Italy); Galperti, C.; Marchetto, C.; Nowak, S.; Sozzi, C. [Associazione EURATOM-ENEA, IFP-CNR, Milano (Italy); Apruzzese, G. [Associazione EURATOM-ENEA, C.R. Frascati (Italy); Bin, W. [Associazione EURATOM-ENEA, IFP-CNR, Milano (Italy); Belli, F.; Botrugno, A. [Associazione EURATOM-ENEA, C.R. Frascati (Italy); Bruschi, A.; Cirant, S. [Associazione EURATOM-ENEA, IFP-CNR, Milano (Italy); D' Antona, G.; Davoudi, M. [Politecnico di Milano, Dipartimento di Elettrotecnica (Italy); Figini, L. [Associazione EURATOM-ENEA, IFP-CNR, Milano (Italy); Ferrero, R. [Politecnico di Milano, Dipartimento di Elettrotecnica (Italy); Gabellieri, L. [Associazione EURATOM-ENEA, C.R. Frascati (Italy); Garavaglia, S.; Granucci, G. [Associazione EURATOM-ENEA, IFP-CNR, Milano (Italy); Grosso, A. [Associazione EURATOM-ENEA, C.R. Frascati (Italy); and others

    2013-08-21

    A real-time data acquisition and elaboration system is being implemented to control the new ECH launcher recently installed at FTU (Frascati Tokamak Upgrade). The system is aimed at controlling different kinds of magnetohydrodynamic instabilities, in particular the deleterious 3/2 and 2/1 (neoclassical) tearing modes, (N)TM, and the saw teeth period in order to prevent the seeding of NTMs. The complete system is presented here together with preliminary offline and real-time tests. © 2001 Elsevier Science. All rights reserved.

  2. Impurity screening in high density plasmas in tokamaks with a limiter configuration

    International Nuclear Information System (INIS)

    Ferro, C.; Zanino, R.

    1992-01-01

    Impurity screening in high density plasmas in tokamaks with a limiter configuration is investigated by means of a simple semi-analytical model. An iterative scheme is devised, in order to determine self-consistently the values of scrape-off layer thickness, edge electron density and temperature, and main plasma contamination parameter Z eff , as a function of given average electron density and temperature in the main plasma and given input power. The model is applied to the poloidal limiter case of the Frascati Tokamak Upgrade, and results are compared with experimental data. A reasonable agreement between the trends is found, emphasizing the importance of a high edge plasma density for obtaining a clean main plasma in limiter tokamaks. (orig.)

  3. Overview of the TCV tokamak program: scientific progress and facility upgrades.

    Czech Academy of Sciences Publication Activity Database

    Coda, S.; Ficker, Ondřej; Horáček, Jan; Papřok, Richard

    2017-01-01

    Roč. 57, October (2017), č. článku 102011. ISSN 0029-5515 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : TCV * tokamak * overview Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 3.307, year: 2016

  4. 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...

  5. Upgrade of the TCV tokamak, first phase: Neutral beam heating system

    Energy Technology Data Exchange (ETDEWEB)

    Karpushov, Alexander N., E-mail: alexander.karpushov@epfl.ch [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Alberti, Stefano; Chavan, René [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Davydenko, Vladimir I. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Duval, Basil P. [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Ivanov, Alexander A. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Fasel, Damien; Fasoli, Ambrogio [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Gorbovsky, Aleksander I. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Goodman, Timothy [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Kolmogorov, Vyacheslav V. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); Martin, Yves; Sauter, Olivier [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, CH-1015 Lausanne (Switzerland); Sorokin, Aleksey V. [Budker Institute of Nuclear Physics SB RAS, 630090 Novosibirsk (Russian Federation); and others

    2015-10-15

    Highlights: • Widening the parameter range of reactor relevant regimes on the TCV tokamak. • Installation of 1 MW, 30 keV neutral beam, direct ion heating, access to T{sub i}/T{sub e} ≥ 1. • ASTRA simulation of plasma response to NB and EC heating in different regimes. • Specific low divergency neutral beam injector with tunable beam power and energy. - Abstract: Experiments on TCV are designed to complement the work at large integrated tokamak facilities (such as JET) to provide a stepwise approach to extrapolation to ITER and DEMO in areas where medium-size tokamaks can often exploit their experimental capabilities and flexibility. Improving the understanding and control requirements of burning plasmas is a major scientific challenge, requiring access to plasma regimes and configurations with high normalized plasma pressure and a wide range of ion to electron temperature ratios, including T{sub e}/T{sub i} ∼ 1. These conditions will be explored by adding a 1 MW neutral heating beam to TCV's auxiliary for direct ion heating (2015) and increasing the ECH power injected in X-mode at the third harmonic (2 MW in 2015–2016). The manufacturing of the neutral beam injector was launched in 2014.

  6. 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...

  7. Upgrade of the COMPASS tokamak microwave reflectometry system with I/Q modulation and detection.

    Czech Academy of Sciences Publication Activity Database

    Zajac, Jaromír; Bogár, Ondrej; Varavin, Mykyta; Žáček, František; Hron, Martin; Pánek, Radomír; Nanobashvili, S.; Silva, A.

    2017-01-01

    Roč. 123, November (2017), s. 911-914 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] R&D Projects: GA ČR(CZ) GA14-35260S Institutional support: RVO:61389021 Keywords : Microwave reflectometry * Heterodyne detection * I/Q modulator * COMPASS tokamak Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379617303101

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

    International Nuclear Information System (INIS)

    Anastassiadis, A.; Tsingas, A.C.; Tsois, N.N.; Zoumbos, G.A.

    1985-05-01

    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)

  9. Velocity space resolved absolute measurement of fast ion losses induced by a tearing mode in the ASDEX Upgrade tokamak

    Science.gov (United States)

    Galdon-Quiroga, J.; Garcia-Munoz, M.; Sanchis-Sanchez, L.; Mantsinen, M.; Fietz, S.; Igochine, V.; Maraschek, M.; Rodriguez-Ramos, M.; Sieglin, B.; Snicker, A.; Tardini, G.; Vezinet, D.; Weiland, M.; Eriksson, L. G.; The ASDEX Upgrade Team; The EUROfusion MST1 Team

    2018-03-01

    Absolute flux of fast ion losses induced by tearing modes have been measured by means of fast ion loss detectors (FILD) for the first time in RF heated plasmas in the ASDEX Upgrade tokamak. Up to 30 MW m-2 of fast ion losses are measured by FILD at 5 cm from the separatrix, consistent with infra-red camera measurements, with energies in the range of 250-500 keV and pitch angles corresponding to large trapped orbits. A resonant interaction between the fast ions in the high energy tail of the ICRF distribution and a m/n  =  5/4 tearing mode leads to enhanced fast ion losses. Around 9.3 +/- 0.7 % of the fast ion losses are found to be coherent with the mode and scale linearly with its amplitude, indicating the convective nature of the transport mechanism. Simulations have been carried out to estimate the contribution of the prompt losses. A good agreement is found between the simulated and the measured velocity space of the losses. The velocity space resonances that may be responsible for the enhanced fast ion losses are identified.

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

    Science.gov (United States)

    García-Muñoz, M; Fahrbach, H-U; Zohm, H

    2009-05-01

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

  11. 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 ...

  12. 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 ...

  13. 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 ...

  14. High spatial resolution upgrade of the electron cyclotron emission radiometer for the DIII-D tokamak.

    Science.gov (United States)

    Truong, D D; Austin, M E

    2014-11-01

    The 40-channel DIII-D electron cyclotron emission (ECE) radiometer provides measurements of Te(r,t) at the tokamak midplane from optically thick, second harmonic X-mode emission over a frequency range of 83-130 GHz. The frequency spacing of the radiometer's channels results in a spatial resolution of ∼1-3 cm, depending on local magnetic field and electron temperature. A new high resolution subsystem has been added to the DIII-D ECE radiometer to make sub-centimeter (0.6-0.8 cm) resolution Te measurements. The high resolution subsystem branches off from the regular channels' IF bands and consists of a microwave switch to toggle between IF bands, a switched filter bank for frequency selectivity, an adjustable local oscillator and mixer for further frequency down-conversion, and a set of eight microwave filters in the 2-4 GHz range. Higher spatial resolution is achieved through the use of a narrower (200 MHz) filter bandwidth and closer spacing between the filters' center frequencies (250 MHz). This configuration allows for full coverage of the 83-130 GHz frequency range in 2 GHz bands. Depending on the local magnetic field, this translates into a "zoomed-in" analysis of a ∼2-4 cm radial region. Expected uses of these channels include mapping the spatial dependence of Alfven eigenmodes, geodesic acoustic modes, and externally applied magnetic perturbations. Initial Te measurements, which demonstrate that the desired resolution is achieved, are presented.

  15. High spatial resolution upgrade of the electron cyclotron emission radiometer for the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Truong, D. D., E-mail: dtruong@wisc.edu [Department of Engineering Physics, University of Wisconsin-Madison, Madison, Wisconsin 53706 (United States); Austin, M. E. [Institute for Fusion Studies, University of Texas, Austin, Texas, 78712 (United States)

    2014-11-15

    The 40-channel DIII-D electron cyclotron emission (ECE) radiometer provides measurements of T{sub e}(r,t) at the tokamak midplane from optically thick, second harmonic X-mode emission over a frequency range of 83–130 GHz. The frequency spacing of the radiometer's channels results in a spatial resolution of ∼1–3 cm, depending on local magnetic field and electron temperature. A new high resolution subsystem has been added to the DIII-D ECE radiometer to make sub-centimeter (0.6–0.8 cm) resolution T{sub e} measurements. The high resolution subsystem branches off from the regular channels’ IF bands and consists of a microwave switch to toggle between IF bands, a switched filter bank for frequency selectivity, an adjustable local oscillator and mixer for further frequency down-conversion, and a set of eight microwave filters in the 2–4 GHz range. Higher spatial resolution is achieved through the use of a narrower (200 MHz) filter bandwidth and closer spacing between the filters’ center frequencies (250 MHz). This configuration allows for full coverage of the 83–130 GHz frequency range in 2 GHz bands. Depending on the local magnetic field, this translates into a “zoomed-in” analysis of a ∼2–4 cm radial region. Expected uses of these channels include mapping the spatial dependence of Alfven eigenmodes, geodesic acoustic modes, and externally applied magnetic perturbations. Initial T{sub e} measurements, which demonstrate that the desired resolution is achieved, are presented.

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

    International Nuclear Information System (INIS)

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

    2010-01-01

    The migration of carbon in low-density, low-confinement plasmas of ASDEX Upgrade was studied by injecting 13 C 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 13 C 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, 13 C was deposited mainly on the high-field side tiles, while barely detectable amounts of 13 C 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 13 C experiments in ASDEX Upgrade. In the case of W-coated tiles, the 13 C measurements were complemented by determining the total amount of deposited carbon ( 12 C) on the tiles, which also shows strong deposition at the inner parts of the lower divertor. The estimated deposition of 13 C on W at the divertor areas was less than 1.5% of the injected amount of 13 C atoms. The 13 C 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 13 C 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.

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

    Science.gov (United States)

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

    2016-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Adamek, J., E-mail: adamek@ipp.cas.cz; Horacek, J.; Bilkova, P.; Böhm, P.; Aftanas, M.; Vondracek, P.; Stöckel, J.; Panek, R. [Institute of Plasma Physics, Prague (Czech Republic); Müller, H. W. [Max-Planck-Institute for Plasma Physics, Garching near Munich (Germany); Institute of Materials Chemistry & Research, University of Vienna, Vienna (Austria); Silva, C.; Fernandes, H.; Figueiredo, H. [Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, 1049-001 Lisboa (Portugal); Schrittwieser, R.; Ionita, C.; Mehlmann, F.; Costea, S. [Institute for Ion Physics and Applied Physics, University of Innsbruck, Innsbruck (Austria); Kurzan, B. [Max-Planck-Institute for Plasma Physics, Garching near Munich (Germany)

    2016-04-15

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  20. 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...

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

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Z. J.; Pan, X. M., E-mail: panxiaoming@hust.edu.cn; Ma, X. D.; Ruan, B. W.; Zhou, R. B.; Zhang, C. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan, Hubei 430074 (China)

    2016-11-15

    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 B{sub T} = 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 B{sub T} = 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 B{sub T}, 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.

  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. 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

  4. Fast-ion transport studies using FIDA spectroscopy at the ASDEX Upgrade tokamak

    International Nuclear Information System (INIS)

    Geiger, Benedikt

    2013-01-01

    A good confinement of fast-ions, i.e. ions with energies above the thermal energy, is essential for the success of fusion devices as it determines, amongst others, the plasma performance and the heating and current drive efficiencies. In case of a turbulent or magneto-hydrodynamic (MHD) active background plasma, various mechanisms have to be considered in order to estimate the spatial distribution of the fast-ions: the slowing down and radial diffusion by Coulomb collisions on electrons and ions, the effect of potential fluctuations and the effect of perturbations of the magnetic field structure. These can lead to a broadening of the fast-ion distribution function which is not yet completely understood. At the fusion experiment ASDEX Upgrade, the fast-ions are generated by heating sources such as neutral beam injection (NBI). Their transport properties can be studied by a fast-ion D-alpha (FIDA) spectroscopy diagnostic which has been built in the framework of this thesis. Through charge exchange reactions with neutrals, fast-ions can receive a bound electron and emit Balmer alpha line radiation. This so-called FIDA radiation can be measured with large Doppler shifts and is localized along the NBI path where a high density of neutrals is present. The FIDA diagnostic uses radially distributed lines of sight that intersect, in the horizontal and in the vertical plane, the path of a 2.5 MW NBI heating source. Thereby different parts of the fast-ion phase space above 25 keV can be analyzed. To interpret the FIDA radiation quantitatively, a forward modelling code has been implemented, tested and further developed. The code calculates, based on theoretical fast-ion distribution functions, synthetic FIDA spectra that can be compared to the measurement. In MHD-quiescent plasmas, the possible effect of turbulence on the fast-ion transport has been investigated with the FIDA diagnostic. The measurements obtained under different experimental conditions, such as during on- and

  5. 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

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

    International Nuclear Information System (INIS)

    Garcia-Munoz, M; Rodriguez-Ramos, M; Äkäslompolo, S; De Marne, P; Dunne, M G; Dux, R; Fietz, S; Fuchs, C; Geiger, B; Herrmann, A; Hoelzl, M; Kurzan, B; McDermott, R M; Strumberger, E; Evans, T E; Ferraro, N M; Pace, D C; Lazanyi, N; Nocente, M; Shinohara, K

    2013-01-01

    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 q 95 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

  7. Comparison of particle confinement in the high confinement mode plasmas with the edge localized mode of the Japan Atomic Energy Research Institute Tokamak-60 Upgrade and the DIII-D tokamak

    International Nuclear Information System (INIS)

    Takenaga, H.; Mahdavi, M.A.; Baker, D.R.

    2001-01-01

    Particle confinement was compared for the high confinement mode plasmas with the edge localized mode in the Japan Atomic Energy Research Institute Tokamak-60 Upgrade (JT-60U) [S. Ishida, JT-60 Team, Nucl. Fusion 39, 1211 (1999)] and the DIII-D tokamak [J. L. Luxon et al., Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159] considering separate confinement times for particles supplied by neutral beam injection (NBI) (center fueling) and by recycling and gas-puffing (edge fueling). Similar dependence on the NBI power was obtained in JT-60U and DIII-D. The particle confinement time for center fueling in DIII-D was smaller by a factor of 4 in the low density discharges and by a factor of 1.8 in the high density discharges than JT-60U scaling, respectively, suggesting the stronger dependence on the density in DIII-D. The particle confinement time for edge fueling in DIII-D was comparable with JT-60U scaling in the low density discharges. However, it decreased to a much smaller value in the high density discharges

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

    Czech Academy of Sciences Publication Activity Database

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

    2016-01-01

    Roč. 87, č. 4 (2016), č. článku 043510. ISSN 0034-6748 R&D Projects: GA ČR(CZ) GA15-10723S; GA ČR(CZ) GAP205/12/2327; GA ČR(CZ) GA14-35260S; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : ball- pen probe (BPP) * ASDEX Upgrade * Langmuir probe (LP) * ISTTOK (Instituto Superior Tecnico TOKamak) * COMPASS (COMPact ASSembly), Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: 2.11 Other engineering and technologies Impact factor: 1.515, year: 2016 http://scitation.aip.org/content/aip/journal/rsi/87/4/10.1063/1.4945797

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

    Czech Academy of Sciences Publication Activity Database

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

    2016-01-01

    Roč. 87, č. 4 (2016), č. článku 043510. ISSN 0034-6748 R&D Projects: GA ČR(CZ) GA15-10723S; GA ČR(CZ) GAP205/12/2327; GA ČR(CZ) GA14-35260S; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : ball-pen probe (BPP) * ASDEX Upgrade * Langmuir probe (LP) * ISTTOK (Instituto Superior Tecnico TOKamak) * COMPASS (COMPact ASSembly), Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: 2.11 Other engineering and technologies Impact factor: 1.515, year: 2016 http://scitation.aip.org/content/aip/journal/rsi/87/4/10.1063/1.4945797

  10. The effect of off-axis neutral beam injection on sawtooth stability in ASDEX Upgrade and Mega-Ampere Spherical Tokamak

    International Nuclear Information System (INIS)

    Chapman, I. T.; de Bock, M. F.; Pinches, S. D.; Turnyanskiy, M. R.; Igochine, V. G.; Maraschek, M.; Tardini, G.

    2009-01-01

    Sawtooth behavior has been investigated in plasmas heated with off-axis neutral beam injection in ASDEX Upgrade [A. Herrmann and O. Gruber, Fusion Sci. Technol. 44, 569 (2003)] and the Mega-Ampere Spherical Tokamak (MAST) [A. Sykes et al., Nucl. Fusion 41, 1423 (2001)]. Provided that the fast ions are well confined, the sawtooth period is found to decrease as the neutral beam is injected further off-axis. Drift kinetic modeling of such discharges qualitatively shows that the passing fast ions born outside the q=1 rational surface can destabilize the n=1 internal kink mode, thought to be related to the sawtooth instability. This effect can be enhanced by optimizing the deposition of the off-axis beam energetic particle population with respect to the mode location.

  11. 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.

  12. Phase-space resolved measurement of 2nd harmonic ion cyclotron heating using FIDA tomography at the ASDEX Upgrade tokamak

    DEFF Research Database (Denmark)

    Weiland, M.; Bilato, R.; Geiger, B.

    2017-01-01

    Recent upgrades to the FIDA (fast-ion D-alpha) diagnostic at ASDEX Upgrade allow to reconstruct the fast-ion phase space at several radial positions with decent energy and pitch resolution. These new diagnostic capabilities are applied to study the physics of 2nd harmonic ion cyclotron heating, w....... Furthermore, comparisons to other fast-ion diagnostics (neutron yield and neutral particle analyzers) are discussed....

  13. Deuterium retention and surface modification of tungsten macrobrush samples exposed in FTU Tokamak

    Science.gov (United States)

    Maddaluno, G.; Giacomi, G.; Rufoloni, A.; Verdini, L.

    2007-06-01

    The effect of discrete structures such as macrobrush or castellated surfaces on power handling and deuterium retention of plasma facing components is to be assessed since such geometrical configurations are needed for increasing the lifetime of the armour to heat-sink joint. Four small macrobrush W and W + 1%La2O3 samples have been exposed in the Frascati Tokamak Upgrade (FTU) scrape-off layer up to the last closed flux surface by means of the Sample Introduction System. FTU is an all metal machine with no carbon source inside vacuum vessel; it exhibits ITER relevant energy and particle fluxes on the plasma facing components. Here, results on morphological surface changes (SEM), chemical composition (EDX) and deuterium retention (TDS) are reported.

  14. INFN-Laboratori Nazionali di Frascati School 2011

    CERN Document Server

    Black objects in supergravity

    2013-01-01

    This is the sixth volume in a series of books on the general topics of supersymmetry, supergravity, black holes and the attractor mechanism. The present volume is based upon lectures held in May 2011 at the INFN-Laboratori Nazionali di Frascati School on Black Objects in Supergravity (BOSS2011), directed by Stefano Bellucci, with the participation of prestigious lecturers, including G. Lopes Cardoso, W. Chemissany, T. Ortin, J. Perz, O. Vaughan, D. Turton, L. Lusanna and S. Ferrara. All lectures were at a pedagogical, introductory level, a feature which is reflected in the specific "flavor" of this volume, which also benefited greatly from extensive discussions and related reworking of the various contributions.

  15. The activity on linear accelerators at the ENEA Frascati center

    International Nuclear Information System (INIS)

    Picardi, L.; Messina, G.; Ronsivalle, C.; Vignati, A.

    1992-01-01

    In the last ten years four small linear accelerators have been built at the ENEA Frascati Center, used for a 100 MeV racetrack microtron and for electron beam processing tests of materials. ENEA is also involved in infrared free electron laser (FEL) research. After good results from a 2 mm wavelength FEL driven by a 2.5 MeV linac, design has been started on a linac for a far infrared FEL facility devoted to an experiment in muonic hydrogen spectroscopy. 3 figs., 3 tabs., ref

  16. Texas Experimental Tokamak

    International Nuclear Information System (INIS)

    Wootton, A.J.

    1990-04-01

    This paper discusses the following work on the text tokamak: data systems; particle confinement; impurity transport; plasma rotation; runaway electrons; electron cyclotron heating; FIR system; transient transport; internal turbulence; edge turbulence; ion temperature; EML experiments; impurity pellet experiments; MHD experiments and analysis; TEXT Upgrade; and Upgrade diagnostics

  17. Studies of tungsten erosion at the inner and outer main chamber wall of the ASDEX Upgrade tokamak

    Science.gov (United States)

    Tabasso, A.; Maier, H.; Roth, J.; Krieger, K.; ASDEX Upgrade Team

    2001-03-01

    A critical issue for the choice of main chamber first wall materials in future fusion devices such as ITER is the erosion rate due to bombardment by charge-exchange (CX) neutrals. Due to the relatively small flux density of impacting particles, respective measurements are only possible using long term samples (LTS) exposed for a full experimental campaign. In ASDEX Upgrade, CX erosion has been studied extensively for tungsten on the inner heat shield by placing four W coated tiles at different poloidal positions in one toroidal sector. During the same campaign, several LTS were placed at different poloidal and toroidal positions of the outer wall. 13C and Cu coated graphite probes were also used in order to test and compare W low and medium Z alternatives. The erosion results from the probes are compared with the calculated erosion [W. Eckstein, C. Garcia-Rosales, J. Roth, W. Ottenberger IPP Report, IPP 9/82]; [H. Verbeek, J. Stober, D. Coster, W. Eckstein, R. Schneider Nucl. Fus. 38 (1998) 12] and a figure of merit (F. of M.) between several materials is proposed which also takes into account the plasma isotope effect in CX erosion.

  18. Studies of tungsten erosion at the inner and outer main chamber wall of the ASDEX Upgrade tokamak

    International Nuclear Information System (INIS)

    Tabasso, A.; Maier, H.; Roth, J.; Krieger, K.

    2001-01-01

    A critical issue for the choice of main chamber first wall materials in future fusion devices such as ITER is the erosion rate due to bombardment by charge-exchange (CX) neutrals. Due to the relatively small flux density of impacting particles, respective measurements are only possible using long term samples (LTS) exposed for a full experimental campaign. In ASDEX Upgrade, CX erosion has been studied extensively for tungsten on the inner heat shield by placing four W coated tiles at different poloidal positions in one toroidal sector. During the same campaign, several LTS were placed at different poloidal and toroidal positions of the outer wall. 13 C and Cu coated graphite probes were also used in order to test and compare W low and medium Z alternatives. The erosion results from the probes are compared with the calculated erosion [W. Eckstein, C. Garcia-Rosales, J. Roth, W. Ottenberger IPP Report, IPP 9/82]; [H. Verbeek, J. Stober, D. Coster, W. Eckstein, R. Schneider Nucl. Fus. 38 (1998) 12] and a figure of merit (F. of M.) between several materials is proposed which also takes into account the plasma isotope effect in CX erosion

  19. Tritium contamination and monitoring at Frascati Neutron Generator

    Energy Technology Data Exchange (ETDEWEB)

    Lucci, F.; Sandri, S.; Ianni, A. [ENEA, Frascati (Italy). Dipartimento Ambiente; Vasselli, R. [ANPA, Roma (Italy); Pillon, M.; Bettinali, L. [ENEA, Frascati (Italy). Dipartimento Energia

    1994-11-01

    The Frascati Neutron Generator (FGN) is a specialised 300 keV, 3 mA direct electrostatic deuteron accelerator which produces about 5-10{sup 1}1 14 MeV neutrons per second by D-T reactions on a tritium-titanium fixed target. This paper concerns the tritium contamination control and monitoring aspects after some months of testing and a preliminary period of operation of the plant. The tritium monitoring system is composed of both on-line and off-line devices to control the tritium concentration in the atmosphere measured from different parts of the plant: vacuum exhaust clean up (VECU) system, stack, etc. The on-line devices are three flux monitors, that sample continuosly the air from up to eight different points in the plant. The passive sampling system is designed to select the chemical form of tritium and to collect respectively HTO and HT in two different cartridges filled with an appropriate drying material. The response of the on-line tritium monitor system are exposed and discussed: some measurements performed with atmosphere dehumidifying apparatus of this system are described and the relevant results are analysed.

  20. Tritium contamination and monitoring at Frascati Neutron Generator

    International Nuclear Information System (INIS)

    Lucci, F.; Sandri, S.; Ianni, A.; Pillon, M.; Bettinali, L.

    1994-11-01

    The Frascati Neutron Generator (FGN) is a specialised 300 keV, 3 mA direct electrostatic deuteron accelerator which produces about 5-10 1 1 14 MeV neutrons per second by D-T reactions on a tritium-titanium fixed target. This paper concerns the tritium contamination control and monitoring aspects after some months of testing and a preliminary period of operation of the plant. The tritium monitoring system is composed of both on-line and off-line devices to control the tritium concentration in the atmosphere measured from different parts of the plant: vacuum exhaust clean up (VECU) system, stack, etc. The on-line devices are three flux monitors, that sample continuosly the air from up to eight different points in the plant. The passive sampling system is designed to select the chemical form of tritium and to collect respectively HTO and HT in two different cartridges filled with an appropriate drying material. The response of the on-line tritium monitor system are exposed and discussed: some measurements performed with atmosphere dehumidifying apparatus of this system are described and the relevant results are analysed

  1. Radiation levels at the boundaries of the Frascati center and natural background

    International Nuclear Information System (INIS)

    Lucci, F.; Pelliccioni, M.

    1975-01-01

    The results of dosimetric controls carried out with thermoluminescent dosimeters along the boundaries of the Frascati Center since 1969 are presented. The average yearly exposure observed during the period 1969-1974 is 206mR/y, while no significant tendency is evident. In particular neither effects of accelerator shutdowns nor seasonal ones are noticed. The measurements at the boundary are also compared with measurements of natural background carried out near Frascati and in Rome, Thus we can estimate that the machine contribution to the total dose along the boundaries is equal to a small fraction of either the natural background or the general population dose limit [fr

  2. Texas Experimental Tokamak

    International Nuclear Information System (INIS)

    Wootton, A.J.

    1993-04-01

    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

  3. On the evaluation of the Munich-Frascati Werber-type experiment

    International Nuclear Information System (INIS)

    Kafka, Peter

    1974-01-01

    The theory of optimal evaluation is sketched and applied to the case of short pulses of gravitational radiation. The concepts of one-detector and multi-detector sensitivity are clarified. Comparison of the equivalent circuits of various detectors and tests with artificial pulses show that the pulses reported by Weber should be detectable in coincidence between Munich and Frascati. First negative results are presented

  4. The basics of spherical tokamaks and progress in European research

    International Nuclear Information System (INIS)

    Gusev, V K; Alladio, F; Morris, A W

    2003-01-01

    When the aspect ratio of a tokamak (A = R/a) decreases significantly, there is a transformation of the well studied tokamak toroidal magnetic configuration into the spherical tokamak (ST) configuration. This configuration has high natural plasma elongation and triangularity and other unique equilibrium and stability properties of ST configuration, which are discussed in this paper. European research into ST physics is well advanced in spite of the young age of this branch of fusion science. An overview of selected experimental and theoretical results obtained at Ioffe, Culham and Frascati is given with the emphasis on their complementarity and links to the main stream of tokamak research, such as ITER. An outline of the basic ST advantages and the potential of ST research for new insights into magnetic confinement is also given. More detailed descriptions of recent advances in ST theory and experiment may be found in the invited papers by Akers and Ono in the proceedings of this conference

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

    Czech Academy of Sciences Publication Activity Database

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

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Adamek, J., E-mail: adamek@ipp.cas.c [Institute of Plasma Physics, Association EURATOM/IPP.CR, Prague, Za Slovankou 3, 182 00, Prague 8 (Czech Republic); Rohde, V.; Mueller, H.W.; Herrmann, A. [Institute of Plasma Physics, Association EURATOM/IPP, Garching (Germany); Ionita, C.; Schrittwieser, R.; Mehlmann, F. [Institute for Ion Physics and Applied Physics, University of Innsbruck, Association EURATOM/OAW (Austria); Stoeckel, J.; Horacek, J.; Brotankova, J. [Institute of Plasma Physics, Association EURATOM/IPP.CR, Prague, Za Slovankou 3, 182 00, Prague 8 (Czech Republic)

    2009-06-15

    Experimental investigations of the plasma potential and electric field were performed for ELMy H-mode plasmas in the vicinity of the limiter shadow of ASDEX Upgrade. A fast reciprocating probe with a probe head containing four ball-pen probes (BPPs) [J. Adamek et al., Czech. J. Phys. 54 (2004) C95 - C99.] was used on the midplane manipulator. Different gradients of the plasma potential were observed during ELMs and in between them. The temporal resolution of the direct plasma potential measurements with BPP was determined by using power-spectra analysis.

  7. Amorphous silicon crystallization by laser. Report of the experiments at Frascati (Project Foto); Cristallizzazione di silicio amorfo via laser. Rapporto degli esperimenti a frascati (Progetto Foto)

    Energy Technology Data Exchange (ETDEWEB)

    Bollanti, S; Di Lazzaro, P; Murra, D [ENEA, Centro Ricerche Frascati, Frascati, RM (Italy). Div. Fisica Applicata; Imparato, A; Privato, C [ENEA, Centro Ricerche Portici, Naples (Italy). Div. Fonti Rinnovabili; Carluccio, R; Fortunato, G; Mariucci, L; Pecora, A [CNR Istituto di Elettronica dello Stato Solido, Rome (Italy)

    2000-07-01

    The final goal of the Project FOTO is the construction of a laboratory in a clean room for the production of active matrix which can be used to obtain Active Matrix Liquid Crystal Displays (AMLCD). The AMLCD are based on Thin Film Transistors (TFT), which can be obtained by poly-silicon (poly-Si) thin films, achieved, e.g., by irradiating films of amorphous silicon (a-Si) by ultraviolet laser radiation. In this report, are presented the results of the a-Si irradiation by using the laser-facility Hercules (excimer XeCl, l=0,308 mm) done at the ENEA Frascati Centre. The transformation of a-Si into poly-Si is commented upon the variation of the space-time characteristics of the laser pulses, of the irradiation conditions and of the characteristics of the irradiated a-Si films. [Italian] Il macro-obiettivo del Progetto FOTO e' la realizzazione di un laboratorio in camera pulita per lo sviluppo di processi atti a fabbricare matrici attive utilizzabili per ottenere schermi piatti a cristalli liquidi (AMLCD, Active Matrix Liquid Crystal Display). Uno dei primi passi del processo consiste nel creare transistori a film sottile (TFT, Thin Film Transistor). A tal fine, e' necessario ottenere strati sottili di Silicio policristallino irragiando films di silicio amorfo con luce laser ultravioletta. In questo rapporto, sono presentati i risultati degli irraggiamenti di film sottili di silicio amorfo tramite la laser-facility Hercules (eccimero XeCl, l=0,308 mm) effettuati presso il C.R. ENEA di Frascati. La trasformazione di silicio amorfo in silicio policristallino cosi' ottenuta e' commentata al variare delle caratteristiche spazio-temporali dell'impulso laser, delle condizioni di irraggiamento e delle caratteristiche del film di silicio amorfo irraggiato.

  8. Amorphous silicon crystallization by laser. Report of the experiments at Frascati (Project Foto); Cristallizzazione di silicio amorfo via laser. Rapporto degli esperimenti a frascati (Progetto Foto)

    Energy Technology Data Exchange (ETDEWEB)

    Bollanti, S.; Di Lazzaro, P.; Murra, D. [ENEA, Centro Ricerche Frascati, Frascati, RM (Italy). Div. Fisica Applicata; Imparato, A.; Privato, C. [ENEA, Centro Ricerche Portici, Naples (Italy). Div. Fonti Rinnovabili; Carluccio, R.; Fortunato, G.; Mariucci, L.; Pecora, A. [CNR Istituto di Elettronica dello Stato Solido, Rome (Italy)

    2000-07-01

    The final goal of the Project FOTO is the construction of a laboratory in a clean room for the production of active matrix which can be used to obtain Active Matrix Liquid Crystal Displays (AMLCD). The AMLCD are based on Thin Film Transistors (TFT), which can be obtained by poly-silicon (poly-Si) thin films, achieved, e.g., by irradiating films of amorphous silicon (a-Si) by ultraviolet laser radiation. In this report, are presented the results of the a-Si irradiation by using the laser-facility Hercules (excimer XeCl, l=0,308 mm) done at the ENEA Frascati Centre. The transformation of a-Si into poly-Si is commented upon the variation of the space-time characteristics of the laser pulses, of the irradiation conditions and of the characteristics of the irradiated a-Si films. [Italian] Il macro-obiettivo del Progetto FOTO e' la realizzazione di un laboratorio in camera pulita per lo sviluppo di processi atti a fabbricare matrici attive utilizzabili per ottenere schermi piatti a cristalli liquidi (AMLCD, Active Matrix Liquid Crystal Display). Uno dei primi passi del processo consiste nel creare transistori a film sottile (TFT, Thin Film Transistor). A tal fine, e' necessario ottenere strati sottili di Silicio policristallino irragiando films di silicio amorfo con luce laser ultravioletta. In questo rapporto, sono presentati i risultati degli irraggiamenti di film sottili di silicio amorfo tramite la laser-facility Hercules (eccimero XeCl, l=0,308 mm) effettuati presso il C.R. ENEA di Frascati. La trasformazione di silicio amorfo in silicio policristallino cosi' ottenuta e' commentata al variare delle caratteristiche spazio-temporali dell'impulso laser, delle condizioni di irraggiamento e delle caratteristiche del film di silicio amorfo irraggiato.

  9. Overview of ASDEX Upgrade results

    NARCIS (Netherlands)

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

    2013-01-01

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

  10. Overview of ASDEX Upgrade results

    DEFF Research Database (Denmark)

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

    2013-01-01

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

  11. INFN-Laboratori Nazionali di Frascati School on the Attractor Mechanism 2009

    CERN Document Server

    4th School on Attractor Mechanism : Supersymmetric Gravity and Black Holes

    2013-01-01

    This book is based upon lectures presented in the summer of 2009 at the INFN-Laboratori Nazionali di Frascati School on Attractor Mechanism, directed by Stefano Bellucci. The symposium included such prestigious lecturers as S. Ferrara,  G. Dall'Agata, J.F. Morales, J. Simón and M. Trigiante. All lectures were given at a pedagogical, introductory level, which is reflected in the specific "flavor" of this volume. The book also benefits from extensive discussions about, and the related reworking of, the various contributions. It is the fifth volume in a series of books on the general topics of supersymmetry, supergravity, black holes and the attractor mechanism.

  12. The on-line data acquisition system for the MHD facility of Frascati

    International Nuclear Information System (INIS)

    Di Bartolomeo, M.; Papalia, B.; Gay, P.; Panaccione, L.

    1975-01-01

    An on-line data acquisition system for the MHD facility of the Laboratorio Conversione Diretta at Frascati is described. After a brief description of the MHD facility and of the measurement requirements, the criteria a,d the configuration of the minicomputer-based data acquisition system chosen are presented. Then the general philosophy and the flow-charts of the software implemented are shown, with particular emphasis to the real-time requirements of the measurement system. At last it is illustrated an off-line program, running on a large computer, that elaborates the output data of the data acquisition system

  13. Centralized multiprocessor control system for the frascati storage rings DAΦNE

    International Nuclear Information System (INIS)

    Di Pirro, G.; Milardi, C.; Serio, M.

    1992-01-01

    We describe the status of the DANTE (DAΦne New Tools Environment) control system for the new DAΦNE Φ-factory under construction at the Frascati National Laboratories. The system is based on a centralized communication architecture for simplicity and reliability. A central processor unit coordinates all communications between the consoles and the lower level distributed processing power, and continuously updates a central memory that contains the whole machine status. We have developed a system of VME Fiber Optic interfaces allowing very fast point to point communication between distant processors. Macintosh II personal computers are used as consoles. The lower levels are all built using the VME standard. (author)

  14. Quantitative AMS depth profiling of the hydrogen isotopes collected in graphite divertor and wall tiles of the tokamak ASDEX-Upgrade

    International Nuclear Information System (INIS)

    Sun, G.Y.; Friedrich, M.; Groetzschel, R.; Buerger, W.; Behrisch, R.; Garcia-Rosales, C.

    1997-01-01

    The accelerator mass spectrometry (AMS) facility at the 3 MV Tandetron in Rossendorf has been applied for quantitative depth profiling of deuterium and tritium in samples cut from graphite protection tiles at the vessel walls of the fusion experiment ASDEX-Upgrade at the Max-Planck-Institut fuer Plasmaphysik in Garching. The tritium originates from D(d,p)T fusion reactions in the plasma and it is implanted in the vessel walls together with deuterium atoms and ions from the plasma. The T concentrations in the surface layers down to the analyzing depth of about 25 μm are in the range of 10 11 to 5 x 10 15 T-atoms/cm 3 corresponding to a tritium retention of 3 x 10 10 to 3.5 x 10 12 T-atoms/cm 2 . The much higher deuterium concentrations in the samples were simultaneously measured by calibrated conventional SIMS. In the surface layers down to the analyzing depth of about 25 μm the deuterium concentrations are between 3 x 10 18 and 8 x 10 21 atoms/cm 3 , corresponding to a deuterium retention of 2.5 x 10 16 to 2.5 x 10 18 atoms/cm 2 The estimated total amount of tritium in the vessel walls is of the same order of magnitude as the total number of neutrons produced in D(d,n) 3 He reactions. (orig.)

  15. The stainless steel bulk shielding benchmark experiment at the Frascati Neutron Generator (FNG)

    International Nuclear Information System (INIS)

    Batistoni, P.; Angelone, M.; Martone, M.; Petrizzi, L.; Pillon, M.; Rado, V.; Santamarina, A.; Abidi, I.; Gastaldi, G.; Joyer, P.; Marquette, J.P.; Martini, M.

    1994-01-01

    In the framework of the European Technology Program for NET/ITER, ENEA (Ente Nazionale per le Nuove Tecnologie, l'Energia e l'Ambiente), Frascati and CEA (Commissariat a l'Energie Atomique), Cadarache, are collaborating on a bulk shielding benchmark experiment using the 14 MeV Frascati Neutron Generator (FNG). The aim of the experiment is to obtain accurate experimental data for improving the nuclear database and methods used in the shielding designs, through a rigorous analysis of the results. The experiment consists of the irradiation of a stainless steel block by 14 MeV neutrons. The neutron flux and spectra at different depths, up to 65 cm inside the block, are measured by fission chambers and activation foils characterized by different energy response ranges. The γ-ray dose measurements are performed with ionization chambers and thermo-luminescent dosimeters (TLD). The first results are presented, as well as the comparison with calculations using the cross section library EFF (European Fusion File). ((orig.))

  16. The Benchmark experiment on stainless steel bulk shielding at the Frascati neutron generator

    International Nuclear Information System (INIS)

    Batistoni, P.; Angelone, M.; Martone, M.; Pillon, M.; Rado, V.

    1994-11-01

    In the framework of the European Technology Program for NET/ITER, ENEA (Italian Agency for New Technologies, Energy and Environment) - Frascati and CEA (Commissariat a L'Energie Atomique) - Cadarache collaborated on a Bulk Shield Benchmark Experiment using the 14-MeV Frascati Neutron Generator (FNG). The aim of the experiment was to obtain accurate experimental data for improving the nuclear database and methods used in shielding designs, through a rigorous analysis of the results. The experiment consisted of the irradiation of a stainless steel block by 14-MeV neutrons. The neutron reaction rates at different depths inside the block were measured by fission chambers and activation foils characterized by different energy response ranges. The experimental results have been compared with numerical results calculated using both S N and Monte Carlo transport codes and as transport cross section library the European Fusion File (EFF). In particular, the present report describes the experimental and numerical activity, including neutron measurements and Monte Carlo calculations, carried out by the ENEA Italian Agency for New Technologies, Energy and Environment) team

  17. Measuring main-ion temperatures in ASDEX upgrade using scattering of ECRH radiation

    DEFF Research Database (Denmark)

    Pedersen, Morten Stejner; Nielsen, Stefan Kragh; Jacobsen, Asger Schou

    2016-01-01

    We demonstrate that collective Thomson scattering of millimeter wave electron cyclotron resonance heating radiation can be used for measurements of the main-ion temperature in the ASDEX Upgrade tokamak.......We demonstrate that collective Thomson scattering of millimeter wave electron cyclotron resonance heating radiation can be used for measurements of the main-ion temperature in the ASDEX Upgrade tokamak....

  18. The poloidal OHM's law and a profile constraint in tokamaks

    International Nuclear Information System (INIS)

    Segre, S.E.; Zanza, V.

    1991-01-01

    Explicit use of the poloidal Ohm's law, together with the radial plasma velocity (obtained from the distribution of plasma sources), leads to a very general constraint on the possible radial profiles of plasma density and temperature. The constraint does not require any ad hoc assumption; it can place severe restrictions on the allowed profiles and is independent of energy and particle transport; also, it may be the underlying principle of profile consistency. The constraint is discussed in the framework of neoclassical theory, using results from the Frascati tokamak. (author). 23 refs, 7 figs

  19. Continuous tokamaks

    International Nuclear Information System (INIS)

    Peng, Y.K.M.

    1978-04-01

    A tokamak configuration is proposed that permits the rapid replacement of a plasma discharge in a ''burn'' chamber by another one in a time scale much shorter than the elementary thermal time constant of the chamber first wall. With respect to the chamber, the effective duty cycle factor can thus be made arbitrarily close to unity minimizing the cyclic thermal stress in the first wall. At least one plasma discharge always exists in the new tokamak configuration, hence, a continuous tokamak. By incorporating adiabatic toroidal compression, configurations of continuous tokamak compressors are introduced. To operate continuous tokamaks, it is necessary to introduce the concept of mixed poloidal field coils, which spatially groups all the poloidal field coils into three sets, all contributing simultaneously to inducing the plasma current and maintaining the proper plasma shape and position. Preliminary numerical calculations of axisymmetric MHD equilibria in continuous tokamaks indicate the feasibility of their continued plasma operation. Advanced concepts of continuous tokamaks to reduce the topological complexity and to allow the burn plasma aspect ratio to decrease for increased beta are then suggested

  20. Tokamak experiments

    International Nuclear Information System (INIS)

    Robinson, D.C.

    1987-01-01

    With the advent of the new large tokamaks JET, JT-60 and TFTR important advances in magnetic confinement have been made. These include the exploitation of radio frequency and neutral beam heating on a much larger scale than previously, the demonstration of regimes of improved confinement and the demonstration of current drive at the Megamp level. A number of small and medium sized tokamaks have also come into operation recently such as WT-3 in Japan with an emphasis on radio frequency current drive and HL-1 a medium sized tokamak in China. Each of these new tokamaks is addressing specific problems which remain for the future development of the system. Of these particular problems: β, density and q limits remain important issues for the future development of the tokamak. β limits are being addressed on the DIII-D device in the USA. The anomalous confinement that the tokamak displays is being explored in detail on the TEXT device in the USA. Two other problems are impurity control and current drive. There is significant emphasis on divertor configurations at the present time with their enhanced confinement in the so called H mode. Due to improved discharge cleaning techniques and the ability to repetitively refuel using pellets, purer plasmas can be obtained even without divertors. Current drive remains a crucial issue for quasi of near steady state operation of the tokamak in the future and many current drive schemes are being investigated. (author) [pt

  1. Copper benchmark experiment at the Frascati Neutron Generator for nuclear data validation

    Energy Technology Data Exchange (ETDEWEB)

    Angelone, M., E-mail: maurizio.angelone@enea.it; Flammini, D.; Loreti, S.; Moro, F.; Pillon, M.; Villari, R.

    2016-11-01

    Highlights: • A benchmark experiment was performed using pure copper with 14 MeV neutrons. • The experiment was performed at the Frascati Neutron Generator (FNG). • Activation foils, thermoluminescent dosimeters and scintillators were used to measure reactions rates (RR), nuclear heating and neutron spectra. • The paper presents the RR measurements and the post analysis using MCNP5 and JEFF-3.1.1, JEFF-3.2 and FENDL-3.1 libraries. • C/Es are presented showing the need for deep revision of Cu cross sections. - Abstract: A neutronics benchmark experiment on a pure Copper block (dimensions 60 × 70 × 60 cm{sup 3}), aimed at testing and validating the recent nuclear data libraries for fusion applications, was performed at the 14-MeV Frascati Neutron Generator (FNG) as part of a F4E specific grant (F4E-FPA-395-01) assigned to the European Consortium on Nuclear Data and Experimental Techniques. The relevant neutronics quantities (e.g., reaction rates, neutron flux spectra, doses, etc.) were measured using different experimental techniques and the results were compared to the calculated quantities using fusion relevant nuclear data libraries. This paper focuses on the analyses carried-out by ENEA through the activation foils techniques. {sup 197}Au(n,γ){sup 198}Au, {sup 186}W(n,γ){sup 187}W, {sup 115}In(n,n′){sup 115}In, {sup 58}Ni(n,p){sup 58}Co, {sup 27}Al(n,α){sup 24}Na, {sup 93}Nb(n,2n){sup 92}Nb{sup m} activation reactions were used. The foils were placed at eight different positions along the Cu block and irradiated with 14 MeV neutrons. Activation measurements were performed by means of High Purity Germanium (HPGe) detector. Detailed simulation of the experiment was carried-out using MCNP5 Monte Carlo code and the European JEFF-3.1.1 and 3.2 nuclear cross-sections data files for neutron transport and IRDFF-v1.05 library for the reaction rates in activation foils. The calculated reaction rates (C) were compared to the experimental quantities (E) and

  2. Conceptual design of the COMPASS upgrade tokamak.

    Czech Academy of Sciences Publication Activity Database

    Pánek, Radomír; Markovič, Tomáš; Cahyna, Pavel; Dejarnac, Renaud; Havlíček, Josef; Horáček, Jan; Hron, Martin; Imríšek, Martin; Junek, Pavel; Komm, Michael; Šesták, David; Urban, Jakub; Varju, Jozef; Weinzettl, Vladimír; Adámek, Jiří; Bílková, Petra; Böhm, Petr; Dimitrova, Miglena; Háček, Pavel; Kovařík, Karel; Krbec, Jaroslav; Mlynář, Jan; Podolník, Aleš; Seidl, Jakub; Stöckel, Jan; Tomeš, Matěj; Zajac, Jaromír; Mitošinková, Klára; Peterka, Matěj; Vondráček, Petr

    2017-01-01

    Roč. 123, October (2017), s. 11-16 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] R&D Projects: GA MŠk(CZ) LM2015045 Institutional support: RVO:61389021 Keywords : COMPASS-U * Closed divertor * High magnetic field * Snowflake divertor Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379617302053

  3. Tokamak Physics Experiment (TPX) design

    International Nuclear Information System (INIS)

    Schmidt, J.A.

    1995-01-01

    TPX is a national project involving a large number of US fusion laboratories, universities, and industries. The element of the TPX requirements that is a primary driver for the hardware design is the fact that TPX tokamak hardware is being designed to accommodate steady state operation if the external systems are upgraded from the 1,000 second initial operation. TPX not only incorporates new physics, but also pioneers new technologies to be used in ITER and other future reactors. TPX will be the first tokamak with fully superconducting magnetic field coils using advanced conductors, will have internal nuclear shielding, will use robotics for machine maintenance, and will remove the continuous, concentrated heat flow from the plasma with new dispersal techniques and with special materials that are actively cooled. The Conceptual Design for TPX was completed during Fiscal Year 1993. The Preliminary Design formally began at the beginning of Fiscal Year 1994. Industrial contracts have been awarded for the design, with options for fabrication, of the primary tokamak hardware. A large fraction of the design and R and D effort during FY94 was focused on the tokamak and in turn on the tokamak magnets. The reason for this emphasis is because the magnets require a large design and R and D effort, and are critical to the project schedule. The magnet development is focused on conductor development, quench protection, and manufacturing R and D. The Preliminary Design Review for the Magnets is planned for fall, 1995

  4. TCV divertor upgrade for alternative magnetic configurations

    Directory of Open Access Journals (Sweden)

    H. Reimerdes

    2017-08-01

    Full Text Available The Swiss Plasma Center (SPC is planning a divertor upgrade for the TCV tokamak. The upgrade aims at extending the research of conventional and alternative divertor configurations to operational scenarios and divertor regimes of greater relevance for a fusion reactor. The main elements of the upgrade are the installation of an in-vessel structure to form a divertor chamber of variable closure and enhanced diagnostic capabilities, an increase of the pumping capability of the divertor chamber and the addition of new divertor poloidal field coils. The project follows a staged approach and is carried out in parallel with an upgrade of the TCV heating system. First calculations using the EMC3-Eirene code indicate that realistic baffles together with the planned heating upgrade will allow for a significantly higher compression of neutral particles in the divertor, which is a prerequisite to test the power dissipation potential of various divertor configurations.

  5. 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.

  6. ISTTOK control system upgrade

    International Nuclear Information System (INIS)

    Carvalho, Ivo S.; 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.

    2013-01-01

    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 ® 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

  7. Tokamak COMPASS

    Czech Academy of Sciences Publication Activity Database

    Řípa, Milan; Křenek, Petr

    2011-01-01

    Roč. 17, č. 1 (2011), s. 32-34 ISSN 1210-4612 Institutional research plan: CEZ:AV0Z20430508 Keywords : fusion * tokamak * Compass * Golem * Institute of Plasma Physics AVCR v.v * NBI * diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics

  8. MAST Upgrade – Construction Status

    Energy Technology Data Exchange (ETDEWEB)

    Milnes, Joe, E-mail: Joe.Milnes@ccfe.ac.uk; Ayed, Nizar Ben; Dhalla, Fahim; Fishpool, Geoff; Hill, John; Katramados, Ioannis; Martin, Richard; Naylor, Graham; O’Gorman, Tom; Scannell, Rory

    2015-10-15

    Highlights: • Outlines unique capability of MAST-U, including divertor and diagnostic capability. • Describes progress made in the manufacture and assembly of key MAST-U components. • Highlights the design challenges that have been overcome. • Lists the key lessons learned thus far in the project. - Abstract: 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, progress in delivering the upgraded diagnostic capability, the modification and upgrading of the NBI heating systems and the complete overhaul of the machine control infrastructure, including a new control room with full remote participation facilities.

  9. Tokamaks - Third Edition

    International Nuclear Information System (INIS)

    Rogister, A L

    2004-01-01

    an introduction to diagnostics for tokamaks. The complexity of fusion plasmas is attested to by the discovery of new phenomena and new operational regimes as machine size and power increased and the diagnostic tools improved over the forty years of research on magnetic confinement. The history of those discoveries in the devices which have been built worldwide after the results obtained on the first tokamaks at the Kurchatov Institute had been confirmed is outlined in chapters 11-12. Particular emphasis is naturally given to the results from the larger tokamaks: ASDEX Upgrade, DIII-D, TFTR, JT-60/JT-60U and JET. Chapter 13 is devoted to the International Tokamak Experimental Reactor and prospects beyond ITER. Examples of operational regimes and of often unexpected phenomena are the linear and saturated ohmic confinement modes, confinement degradation when auxiliary heating is applied, the high energy confinement mode, the formation of internal transport barriers in weak or negative central shear discharges, sawtooth relaxations, disruptions, multifaceted asymmetric radiation from the edge, edge localised modes, etc. The relevant observations are described very thoroughly with the support of numerous selected figures and their physical interpretation, a major topic of the book, is carefully discussed on the basis of simplified but convincing mathematical models. With respect to the previous edition (1997), a few additions have been introduced; those concern plasma rotation (section 3.13), internal transport barriers (4.14), the role of radial electric field shear (4.19), turbulence simulations (4.21), impurity transport (4.22) and neoclassical drive of tearing modes (7.3). It is my personal feeling that some of those additions should have been somewhat more elaborated. A few pages have finally been added concerning the TCV, START, MAST, NSTX and ASDEX Upgrade tokamaks. With this book, John Wesson offers the fusion community a very precious and thorough survey of

  10. Recent Upgrades and Extensions of the ASDEX Upgrade ECRH System

    Science.gov (United States)

    Wagner, Dietmar; Stober, Jörg; Leuterer, Fritz; Monaco, Francesco; Münich, Max; Schmid-Lorch, Dominik; Schütz, Harald; Zohm, Hartmut; Thumm, Manfred; Scherer, Theo; Meier, Andreas; Gantenbein, Gerd; Flamm, Jens; Kasparek, Walter; Höhnle, Hendrik; Lechte, Carsten; Litvak, Alexander G.; Denisov, Gregory G.; Chirkov, Alexey; Popov, Leonid G.; Nichiporenko, Vadim O.; Myasnikov, Vadim E.; Tai, Evgeny M.; Solyanova, Elena A.; Malygin, Sergey A.

    2011-03-01

    The multi-frequency Electron Cyclotron Heating (ECRH) system at the ASDEX Upgrade tokamak employs depressed collector gyrotrons, step-tunable in the range 105-140 GHz. The system is equipped with a fast steerable launcher allowing for remote steering of the ECRH RF beam during the plasma discharge. The gyrotrons and the mirrors are fully integrated in the discharge control system. The polarization can be controlled in a feed-forward mode. 3 Sniffer probes for millimeter wave stray radiation detection have been installed.

  11. Contributions to the 14th Symposium on fusion technology

    International Nuclear Information System (INIS)

    Andreani, R.

    1987-01-01

    The ENEA contributions to the 14. Symposium on fusion technology is represented by 15 papers. They are dealing mainly with the FTU (Frascati Tokamak Upgrade), a device under construction, through which high densities and confinement times will be obtained

  12. PSI: Upgrading

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    The accelerator complex at the Paul Scherrer Institute in Villigen near Zurich (PSI - formed in 1988 by combining the Federal Institute for Reactor Research and the Swiss Institute for Nuclear Research) is in the throes of a major and lengthy upgrade

  13. PSI: Upgrading

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1992-09-15

    The accelerator complex at the Paul Scherrer Institute in Villigen near Zurich (PSI - formed in 1988 by combining the Federal Institute for Reactor Research and the Swiss Institute for Nuclear Research) is in the throes of a major and lengthy upgrade.

  14. [Fusion research/tokamak]. Final report, 1 May 1988 - 30 April 1994

    International Nuclear Information System (INIS)

    1994-01-01

    The objectives of the Fusion Research Center Program are: (1) to advance /the transport studies of tokamaks, including the development and maintenance of the Magnetic Fusion Energy Database, and (2) to provide theoretical interpretation, modeling and equilibrium and stability studies for the text-upgrade tokamak. Work is described on five basic categories: (1) magnetic fusion energy database; (2) computational support and numerical modeling; (3) support for TEXT-upgrade and diagnostics; (4) transport studies; and (5) Alfven waves

  15. The physics of an ignited tokamak

    International Nuclear Information System (INIS)

    Troyon, F.

    1990-10-01

    There appears to be a consensus that time has come to embark on the design and construction of the next generation of tokamaks which is at the origin of the ITER initiative. Different proposals have been made based on different appreciation as to the size of the step which can be taken, related to considerations of cost, risk and duration of construction. A class of devices which may be considered the last the very high-field, high density ALCATOR-Frascati line of tokamaks have been proposed for some years specifically for this purpose. Today there remain three such projects: Ignitor, Ignitex and CIT. The technology chosen limits the pulse length to a few seconds. These devices have evolved through the years becoming larger and much more expensive than originally anticipated, increasing the pressure to do more than just a simple demonstration of ignition. There is another class of more ambitious devices which aim at creating long burning plasmas in conditions as close as possible to those of a tokamak reactor in order to address all the plasma physics problems associated with long burn. Three such projects, NET, the european next step after JET, ITER and JIT are good examples of this approach. The ideal would be to design a device with sufficient margin to study burning plasmas over a wide range of parameters. The object of this didactic presentation is to describe the common physics basis of all these projects, compare their expected performance using present knowledge and list the physics problems associated with a burning plasma experiment. The comparison is not meant to be a judgement since the important parameter is the cost/benefit ratio which is a matter of appreciation at this stage. 6 refs., 3 figs., 1 tab

  16. The Discharge Design of HL-2M with the Tokamak Simulation Code (TSC)

    International Nuclear Information System (INIS)

    Yudong Pan; Jardin, S.C.; Kes, C.

    2007-01-01

    We present results on the discharge design of the HL-2M tokamak, which is to be an upgrade to the existing HL-2A tokamak. We present simulation results for complete 5-sec. discharges, both double null and lower single null, for both ohmic and auxiliary heated discharges. We also discuss the vertical stability properties of the device

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

    International Nuclear Information System (INIS)

    Kendl, A.; Scott, B.D.

    2004-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Kendl, A. [Innsbruck Univ., Institut fuer Theoretische Physik, Association EURATOM (Austria); Scott, B.D. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Garching bei Muenchen (Germany)

    2004-07-01

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

  19. Varennes Tokamak

    International Nuclear Information System (INIS)

    Cumyn, P.B.

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

  20. Tokamak physics

    International Nuclear Information System (INIS)

    Haines, M.G.

    1984-01-01

    The physical conditions required for breakeven in thermonuclear fusion are derived, and the early conceptual ideas of magnetic confinement and subsequent development are followed, leading to present-day large scale tokamak experiments. Confinement and diffusion are developed in terms of particle orbits, whilst magnetohydrodynamic stability is discussed from energy considerations. From these ideas are derived the scaling laws that determine the physical size and parameters of this fusion configuration. It becomes clear that additional heating is required. However there are currently several major gaps in our understanding of experiments; the causes of anomalous electron energy loss and the major current disruption, the absence of the 'bootstrap' current and what physics determines the maximum plasma pressure consistent with stability. The understanding of these phenomena is a major challenge to plasma physicists. (author)

  1. INFN-Laboratori Nazionali di Frascati School 2013: Breaking of Supersymmetry and Ultraviolet Divergences in Extended Supergravity

    CERN Document Server

    BUDS 2013

    2014-01-01

    This is the seventh volume in a series on the general topics of supersymmetry, supergravity, black objects (including black holes), and the attractor mechanism. The present volume is based on lectures held in March 2013 at the INFN-Laboratori Nazionali di Frascati during the Breaking of supersymmetry and Ultraviolet Divergences in extended Supergravity Workshop (BUDS 2013), organized by Stefano Bellucci, with the participation of prestigious speakers including P. Aschieri, E. Bergshoeff, M. Cederwall, T. Dennen, P. Di Vecchia, S. Ferrara, R. Kallosh, A. Karlsson, M. Koehn, B. Ovrut, A. Van Proeyen, G. Ruppeiner. Special attention is devoted to discussing topics related to the cancellation of ultraviolet divergences in extended supergravity and Born-Infeld-like actions. All talks were followed by extensive discussions and subsequent reworking of the various contributions, a feature which is reflected in the unique "flavor" of this volume.

  2. Coal upgrading

    Energy Technology Data Exchange (ETDEWEB)

    Nunes, S. [IEA Clean Coal Centre, London (United Kingdom)

    2009-10-15

    This report examines current technologies and those likely to be used to produce cleaner coal and coal products, principally for use in power generation and metallurgical applications. Consideration is also given to coal production in the leading coal producing countries, both with developed and developing industries. A range of technologies are considered. These include the coal-based liquid fuel called coal water mixture (CWM) that may compete with diesel, the production of ultra-clean coal (UCC) and coal liquefaction which competes with oil and its products. Technologies for upgrading coal are considered, especially for low rank coals (LRC), since these have the potential to fill the gap generated by the increasing demand for coal that cannot be met by higher quality coals. Potential advantages and downsides of coal upgrading are outlined. Taking into account the environmental benefits of reduced pollution achieved through cleaner coal and reduced transport costs, as well as other positive aspects such as a predictable product leading to better boiler design, the advantages appear to be significant. The drying of low rank coals improves the energy productively released during combustion and may also be used as an adjunct or as part of other coal processing procedures. Coal washing technologies vary in different countries and the implications of this are outlined. Dry separation technologies, such as dry jigging and electrostatic separation, are also described. The demonstration of new technologies is key to their further development and demonstrations of various clean coal technologies are considered. A number of approaches to briquetting and pelletising are available and their use varies from country to country. Finally, developments in upgrading low rank coals are described in the leading coal producing countries. This is an area that is developing rapidly and in which there are significant corporate and state players. 81 refs., 32 figs., 3 tabs.

  3. Performance of the first ASDEX Upgrade neutral beam injector

    International Nuclear Information System (INIS)

    Staebler, A.; Vollmer, O.; Feist, J.H.; Speth, E.; Heinemann, B.; Melkus, W.; Obermayer, S.; Riedl, R.; Schaerich, W.; Wittenbecher, K.

    1995-01-01

    Plasmas of the ASDEX Upgrade tokamak have been heated with H 0 beams of up to 7 MW and D 0 beams of up to 10 MW. Beam modulation allows to inject at any power level between zero and full power. Measurements characterizing the NBI system performance, the power accountability, and the operational experience obtained so far are discussed. (orig.)

  4. PPPL tokamak program

    International Nuclear Information System (INIS)

    Furth, H.P.

    1984-10-01

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

  5. Status of tokamak research

    International Nuclear Information System (INIS)

    Rawls, J.M.

    1979-10-01

    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

  6. Tokamak Systems Code

    International Nuclear Information System (INIS)

    Reid, R.L.; Barrett, R.J.; Brown, T.G.

    1985-03-01

    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

  7. Diagnosing transient plasma status: from solar atmosphere to tokamak divertor

    International Nuclear Information System (INIS)

    Giunta, A.S.; Henderson, S.; O'Mullane, M.; Summers, H.P.; Harrison, J.; Doyle, J.G.

    2016-01-01

    This work strongly exploits the interdisciplinary links between astrophysical (such as the solar upper atmosphere) and laboratory plasmas (such as tokamak devices) by sharing the development of a common modelling for time-dependent ionisation. This is applied to the interpretation of solar flare data observed by the UVSP (Ultraviolet Spectrometer and Polarimeter), on-board the Solar Maximum Mission and the IRIS (Interface Region Imaging Spectrograph), and also to data from B2-SOLPS (Scrape Off Layer Plasma Simulations) for MAST (Mega Ampère Spherical Tokamak) Super-X divertor upgrade. The derived atomic data, calculated in the framework of the ADAS (Atomic Data and Analysis Structure) project, allow equivalent prediction in non-stationary transport regimes and transients of both the solar atmosphere and tokamak divertors, except that the tokamak evolution is about one thousand times faster.

  8. Tokamak devices: towards controlled fusion

    International Nuclear Information System (INIS)

    Trocheris, M.

    1975-01-01

    The Tokamak family is from Soviet Union. These devices were exclusively studied at the Kurchatov Institute in Moscow for more than ten years. The first occidental Tokamak started in 1970 at Princeton. The TFR (Tokamak Fontenay-aux-Roses) was built to be superior to the Russian T4. Tokamak future is now represented by the JET (Joint European Tokamak) [fr

  9. Disruption studies on ASDEX upgrade

    International Nuclear Information System (INIS)

    Pautasso, G.; Egorov, S.; Finken, K.H.

    2003-01-01

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

  10. Disruption studies in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Pautasso, G.

    2002-01-01

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

  11. The physics of tokamak start-up

    International Nuclear Information System (INIS)

    Mueller, D.

    2013-01-01

    Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases, inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. International Thermonuclear Experimental Reactor, the National Spherical Torus Experiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in the solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection, and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current

  12. Njoy modules used at Enea, Frascati to produce an Ace format neutron cross section library from Eff-1 for the Monte Carlo Mcnp

    International Nuclear Information System (INIS)

    Petrizzi, L.

    1989-01-01

    A note is presented about the experience had in using the NJOY 87.1 module to produce an ACE format library for MCNP from the European Fusion File EFF-1. The IBM 3090 computer, MVS system at ENEA, Bologna was used. The library, called MCNP. EFF1 is at the moment available at Frascati. Few words are said about the met processing problems and the more general topics related to our activity

  13. Tokamak engineering mechanics

    International Nuclear Information System (INIS)

    Song, Yuntao; Wu, Weiyue; Du, Shijun

    2014-01-01

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

  14. Studies of runaway electrons via Cherenkov effect in tokamaks

    Science.gov (United States)

    Zebrowski, J.; Jakubowski, L.; Rabinski, M.; Sadowski, M. J.; Jakubowski, M. J.; Kwiatkowski, R.; Malinowski, K.; Mirowski, R.; Mlynar, J.; Ficker, O.; Weinzettl, V.; Causa, F.; COMPASS; FTU Teams

    2018-01-01

    The paper concerns measurements of runaway electrons (REs) which are generated during discharges in tokamaks. The control of REs is an important task in experimental studies within the ITER-physics program. The NCBJ team proposed to study REs by means of Cherenkov-type detectors several years ago. The Cherenkov radiation, induced by REs in appropriate radiators, makes it possible to identify fast electron beams and to determine their spatial- and temporal-characteristics. The results of recent experimental studies of REs, performed in two tokamaks - COMPASS in Prague and FTU in Frascati, are summarized and discussed in this paper. Examples of the electron-induced signals, as recorded at different experimental conditions and scenarios, are presented. Measurements performed with a three-channel Cherenkov-probe in COMPASS showed that the first fast electron peaks can be observed already during the current ramp-up phase. A strong dependence of RE-signals on the radial position of the Cherenkov probe was observed. The most distinct electron peaks were recorded during the plasma disruption. The Cherenkov signals confirmed the appearance of post-disruptive RE beams in circular-plasma discharges with massive Ar-puffing. During experiments at FTU a clear correlation between the Cherenkov detector signals and the rotation of magnetic islands was identified.

  15. 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

  16. Advanced Tokamak Stability Theory

    Science.gov (United States)

    Zheng, Linjin

    2015-03-01

    The intention of this book is to introduce advanced tokamak stability theory. We start with the derivation of the Grad-Shafranov equation and the construction of various toroidal flux coordinates. An analytical tokamak equilibrium theory is presented to demonstrate the Shafranov shift and how the toroidal hoop force can be balanced by the application of a vertical magnetic field in tokamaks. In addition to advanced theories, this book also discusses the intuitive physics pictures for various experimentally observed phenomena.

  17. Tokamak confinement scaling laws

    International Nuclear Information System (INIS)

    Connor, J.

    1998-01-01

    The scaling of energy confinement with engineering parameters, such as plasma current and major radius, is important for establishing the size of an ignited fusion device. Tokamaks exhibit a variety of modes of operation with different confinement properties. At present there is no adequate first principles theory to predict tokamak energy confinement and the empirical scaling method is the preferred approach to designing next step tokamaks. This paper reviews a number of robust theoretical concepts, such as dimensional analysis and stability boundaries, which provide a framework for characterising and understanding tokamak confinement and, therefore, generate more confidence in using empirical laws for extrapolation to future devices. (author)

  18. Tokamak concept innovations

    International Nuclear Information System (INIS)

    1986-04-01

    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

  19. Plea for stellarator funding raps tokamaks

    International Nuclear Information System (INIS)

    Blake, M.

    1992-01-01

    The funding crunch in magnetic confinement fusion development has moved the editor of a largely technical publication to speak out on a policy issue. James A. Rome, who edits Stellarator News from the Fusion Energy Division at Oak Ridge National Laboratory, wrote an editorial that appeared on the front page of the May 1992 issue. It was titled open-quotes The US Stellarator Program: A Time for Renewal,close quotes and while it focused chiefly on that subject (and lamented the lack of funding for the operation of the existing ATF stellarator at Oak Ridge), it also cited some of the problems inherent in the mainline MCF approach--the tokamak--and stated that if the money can be found for further tokamak design upgrades, it should also be found for stellarators. Rome wrote, open-quotes There is growing recognition in the US, and elsewhere, that the conventional tokamak does not extrapolate to a commercially competitive energy source except with very high field coils ( 1000 MWe).close quotes He pointed up open-quotes the difficulty of simultaneously satisfying conflicting tokamak requirements for efficient current drive, high bootstrap-current fraction, complete avoidance of disruptions, adequate beta limits, and edge-plasma properties compatible with improved (H-mode) confinement and acceptable erosion of divertor plates.close quotes He then called for support for the stellarator as open-quotes the only concept that has performance comparable to that achieved in tokamaks without the plasma-current-related limitations listed above.close quotes

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

    OpenAIRE

    Boyle, Dennis; Gray, Timothy; Granstedt, Erik; Kozub, Thomas; Berzak, Laura; Hammett, Gregory; Kugel, Henry; Leblanc, Benoit; Logan, Nicholas; Jacobson, Craig M.; Lucia, Matthew; Jones, Andrew; Lundberg, Daniel; Timberlake, John; Majeski, Richard

    2010-01-01

    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 wa...

  1. 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 obtain...... for an extension of the ESEL model from 2D to 3D to fully resolve the parallel dynamics, and the coupling from the plasma to the sheath....

  2. Tokamak control simulator

    International Nuclear Information System (INIS)

    Edelbaum, T.N.; Serben, S.; Var, R.E.

    1976-01-01

    A computer model of a tokamak experimental power reactor and its control system is being constructed. This simulator will allow the exploration of various open loop and closed loop strategies for reactor control. This paper provides a brief description of the simulator and some of the potential control problems associated with this class of tokamaks

  3. Fabrication of the vacuum vessel for JT-60 machine upgrade

    International Nuclear Information System (INIS)

    Uchikawa, T.; Takanabe, K.; Tsujimura, S.; Ue, K.; Oka, K.; Kuri, S.; Ioki, K.; Namiki, K.; Suzuki, Y.; Horliike, H.; Ninomiya, H.; Yamamoto, M.; Neyatani, Y.; Ando, T.; Matsukawa, M.

    1992-01-01

    The JT-60 tokamak was upgraded to double the plasma current to 6 MA. In the JT-60 machine upgrade (JT-60U), the vacuum vessel and poloidal field (PF) coils were renewed. The new vacuum vessel features a three-dimensionally curved, thin double-skin torus with multi-arc D-shaped cross section. The double-skin structure is strengthened with square pipes placed in between the outer and inner skins. Fabrication and site installation of the vessel was smoothly completed in February, 1991. This paper describes an overview of the JT-60U vacuum vessel construction

  4. Physical design of JT-60 Super Upgrade

    International Nuclear Information System (INIS)

    Nagashima, K.; Kikuchi, M.; Kurita, G.; Ozeki, T.; Aoyagi, T.; Ushigusa, K.; Neyatani, Y.; Kubo, T.; Mori, K.; Nakagawa, S.; Kuriyama, M.; Nagami, M.

    1997-01-01

    The JT-60 Super Upgrade (JT-60SU) is an upgraded tokamak device of JT-60U for developing the steady-state reactor and advanced tokamak operation in the International Thermonuclear Experimental Reactor. The device is planned to utilize the JT-60 facilities fully and to minimize the needed modification. The major radius is 4.8 m and the maximum plasma current is 10 MA. Neutral beam injection with 750 keV beam energy is the primary heating method. The machine is capable of steady-state operation with high density up to 8.8 x 10 19 m -3 at 5 MA plasma current. The high operating density, over the Greenwald et al. limit, is critically important in order to achieve high bootstrap current fraction. Ballooning mode and low n ideal magnetohydrodynamic (MHD) mode including the bootstrap current were analyzed for steady-state operation. The current profile must be optimized to obtain a normalized beta up to 3. The plasma configuration with high triangularity was adopted in order to get good MHD stability and high energy confinement. A compact divertor was designed in order to get the large plasma space. (orig.)

  5. Tokamaks. 2. ed.

    International Nuclear Information System (INIS)

    Wesson, John; Campbell, D.J.; Connor, J.W.

    1997-01-01

    It is interesting to recall the state of tokamak research when the first edition of this book was written. My judgement of the level of real understanding at that time is indicated by the virtual absence of comparisons of experiment with theory in that edition. The need then was for a 'handbook' which collected in a single volume the concepts and models which form the basis of everyday tokamak research. The experimental and theoretical endeavours of the subsequent decade have left almost all of this intact, but have brought a massive development of the subject. Firstly, there are now several areas where the experimental behaviour is described in terms of accepted theory. This is particularly true of currents parallel to the magnetic field, and of the stability limitations on the plasma pressure. Next there has been the research on large tokamaks, hardly started at the writing of the first edition. Now our thinking is largely based on the results from these tokamaks and this work has led to the long awaited achievement of significant amounts of fusion power. Finally, the success of tokamak research has brought us face to face with the problems involved in designing and building a tokamak reactor. The present edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes an account of the advances outlined above. (Author)

  6. Plasma diagnostics for the DIII-D divertor upgrade (abstract)

    International Nuclear Information System (INIS)

    Hill, D.N.; Futch, A.; Buchenauer, D.; Doerner, R.; Lehmer, R.; Schmitz, L.; Klepper, C.C.; Menon, M.; Leikind, B.; Lippmann, S.; Mahdavi, M.A.; Schaffer, M.; Smith, J.; Salmonson, J.; Watkins, J.

    1990-01-01

    The DIII-D tokamak is being upgraded to allow for divertor biasing, baffling, and pumping experiments. This paper gives an overview of the new diagnostics added to DIII-D as part of this advanced divertor program. They include tile current monitors, fast reciprocating Langmuir probes, a fixed probe array in the divertor, fast neutral pressure gauges, and H α measurements with TV cameras and fiber optics coupled to a high-resolution spectrometer

  7. Diagnostics Neutral Beam Injector at the TCV Tokamak

    International Nuclear Information System (INIS)

    Mlynar, J.; Shukaev, A.N.; Bosshard, P.; Duval, B.P.; Ivanov, A.A.; Kollegov, M.; Kolmogorov, V.V.; Llobet, X.; Pitts, R.A.; Weisen, H.

    2001-10-01

    Within this report we summarize the technical and experimental effort made on diagnostics neutral beam injector (DNBI) which was installed at tokamak TCV last year. Basic components of DNBI are reviewed, its remote control is presented in more detail. Profile and attenuation studies are referred to. First experimental results obtained with DNBI, which led to a decision to upgrade the machine, are discussed in the last section. (author)

  8. Tokamak reactor studies

    International Nuclear Information System (INIS)

    Baker, C.C.

    1981-01-01

    This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

  9. Tokamak ARC damage

    International Nuclear Information System (INIS)

    Murray, J.G.; Gorker, G.E.

    1985-01-01

    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

  10. Survey of Tokamak experiments

    International Nuclear Information System (INIS)

    Bickerton, R.J.

    1977-01-01

    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)

  11. Tokamak ARC damage

    Energy Technology Data Exchange (ETDEWEB)

    Murray, J.G.; Gorker, G.E.

    1985-01-01

    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.

  12. Lidar fluoro sensor measurements of pine trees affected by the Blastophagus destruens Wollaston in the Frascati ENEA centre; Monitoraggio mediante lidar fluorosensore di piante di pino domestico attaccate dal coleottero Scolitide Blastofago Distruttore (Blastophagus Destruens W.) al centro ENEA di Frascati

    Energy Technology Data Exchange (ETDEWEB)

    Barbini, R.; Colao, F.; Fantoni, R.; Palucci, A.; Ribezzo, S. [ENEA, Centro Ricerche Frascati, Rome (Italy). Dip. Innovazione; Cinti, S. [Osservatorio per le Malattie delle Piante, Regione Lazio, Rome (Italy); Paparatti, B. [Tuscia Univ., Viterbo (Italy). Dip. Protezione delle Piante, Sez. Entomologia

    1996-10-01

    The lidar fluoro sensor available at ENEA-Frascati has been used for a series of laboratory measurements to monitor the health of pine trees (Pinus Pinea L.) cut branches. The present campaign has been carried out in the Frascati Research Center to investigate pine plants affected by a parasite (Blastophagus destruens Wollaston), and to follow the effects of the phyto-chemical substances used in the therapy. The laboratory tests have been carried on during the first week of therapy and in the following months. A probe laser has been used to measure the chlorophyll LIF (laser induced fluorescence) signals at 690 nm, before and after switching on either a saturating actinic lamp or an intense laser pulse. The detected induction curves supplied us with information on the plant photosynthetic activity and on its time adaptation to the light exposure.

  13. Tokamak simulation code manual

    International Nuclear Information System (INIS)

    Chung, Moon Kyoo; Oh, Byung Hoon; Hong, Bong Keun; Lee, Kwang Won

    1995-01-01

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

  14. Major upgrades of the high frequency B-dot probe diagnostic suite on ASDEX Upgrade

    Directory of Open Access Journals (Sweden)

    Ochoukov Roman

    2017-01-01

    Full Text Available The high frequency B-dot (HFB probe diagnostic on the ASDEX Upgrade tokamak has undergone a considerable upgrade during the 2016 opening of the torus. The probe coverage is now greatly expanded toroidally, as well as radially with the addition of probes on the high field side and the removable manipulator head. A new 2-channel fast digitizer now allows to examine and record radio frequency (RF wave emissions emanating from the plasma in the ion cyclotron range of frequencies (ICRF. Possible studies that can be achieved now include: a study of core ICRF power absorption efficiency; a study of ion cyclotron emissions from the plasma generated by energetic ions; and study of ICRF wave/plasma turbulence interactions in the scrape-off layer region.

  15. ANITA-2000 activation code package - updating of the decay data libraries and validation on the experimental data of the 14 MeV Frascati Neutron Generator

    Directory of Open Access Journals (Sweden)

    Frisoni Manuela

    2016-01-01

    Full Text Available ANITA-2000 is a code package for the activation characterization of materials exposed to neutron irradiation released by ENEA to OECD-NEADB and ORNL-RSICC. The main component of the package is the activation code ANITA-4M that computes the radioactive inventory of a material exposed to neutron irradiation. The code requires the decay data library (file fl1 containing the quantities describing the decay properties of the unstable nuclides and the library (file fl2 containing the gamma ray spectra emitted by the radioactive nuclei. The fl1 and fl2 files of the ANITA-2000 code package, originally based on the evaluated nuclear data library FENDL/D-2.0, were recently updated on the basis of the JEFF-3.1.1 Radioactive Decay Data Library. This paper presents the results of the validation of the new fl1 decay data library through the comparison of the ANITA-4M calculated values with the measured electron and photon decay heats and activities of fusion material samples irradiated at the 14 MeV Frascati Neutron Generator (FNG of the NEA-Frascati Research Centre. Twelve material samples were considered, namely: Mo, Cu, Hf, Mg, Ni, Cd, Sn, Re, Ti, W, Ag and Al. The ratios between calculated and experimental values (C/E are shown and discussed in this paper.

  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. ETA-II accelerator upgrades

    International Nuclear Information System (INIS)

    Nilson, D.G.; Deadrick, F.J.; Hibbs, S.M.; Sampayan, S.E.; Petersen, D.E.

    1991-09-01

    We discuss recent improvements to the ETA-II linear induction electron accelerator. The accelerator's cells have been carefully reconditioned to raise the maximum accelerating gap voltage from approximately 100 kV to 125 kV. Insulators of Rexolite plastic in a new ''zero-gap'' arrangement replaced the alumina originals after several alternative materials were investigated. A new multi-cable current feed system will be used to eliminate pulse reflection interactions encountered in earlier experiments. Improved alignment fixtures have been installed to help minimize beam perturbation due to poorly aligned intercell magnets between 10-cell groups. A stretched wire alignment technique (SWAT) has been utilized to enhance overall magnetic alignment, and to characterize irreducible alignment errors. These changes are in conjunction with an expansion of the accelerator from a 20-cell to a 60-cell configuration. When completed, the upgraded accelerator is expected to deliver 2.5 kA of electron beam current at 7.5 MeV in bursts of up to fifty 70-ns pulses at a 5-kHz repetition rate. A 5.5-meter-long wiggler will convert the energy into 3-GW microwave pulses at 140 GHz for plasma heating experiments in the Microwave Tokamak Experiment (MTX)

  18. Regime identification in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Giannone, L; Sips, A C C; Kardaun, O; Spreitler, F; Suttrop, W

    2004-01-01

    The ability to recognize the transition from the L-mode to the H-mode or from the H-mode to the improved H-mode reliably from a conveniently small number of measurements in real time is of increasing importance for machine control. Discriminant analysis has been applied to regime identification of plasma discharges in the ASDEX Upgrade tokamak. An observation consists of a set of plasma parameters averaged over a time slice in a discharge. The data set consists of all observations over different discharges and time slices. Discriminant analysis yields coefficients allowing the classification of a new observation. The results of a frequentist and a formal Bayesian approach to discriminant analysis are compared. With five plasma variables, a failure rate of 1.3% for predicting the L-mode and the H-mode confinement regime was achieved. With five plasma variables, a failure rate of 5.3% for predicting the H-mode and the improved H-mode confinement regime was achieved. The coefficients derived by discriminant analysis have been applied subsequently to discharges to illustrate the operation of regime identification in a real time control system

  19. NSTX-U Control System Upgrades

    International Nuclear Information System (INIS)

    Erickson, K.G.; Gates, D.A.; Gerhardt, S.P.; Lawson, J.E.; Mozulay, R.; Sichta, P.; Tchilinguirian, G.J.

    2014-01-01

    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

  20. 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.

  1. Plasma shut-down with fast impurity puff on ASDEX Upgrade

    International Nuclear Information System (INIS)

    Pautasso, G.; Fuchs, C.J.; Gruber, O.; Maggi, C.F.; Maraschek, M.; Puetterich, T.; Rohde, V.; Wittmann, C.; Wolfrum, E.; Cierpka, P.; Beck, M.

    2007-01-01

    The massive injection of impurity gas into a plasma has been proved to reduce forces and localized thermal loads caused by disruptions in tokamaks. This mitigation system is routinely used on ASDEX Upgrade to shut down plasmas with a locked mode. The plasma response to impurity injection and the mechanism of reduction of the mechanical forces is discussed in the paper

  2. Overview of physics results from MAST towards ITER/DEMO and the MAST Upgrade

    DEFF Research Database (Denmark)

    Meyer, H.; Abel, I.G.; Akers, R.J.

    2013-01-01

    New diagnostic, modelling and plant capability on the Mega Ampère Spherical Tokamak (MAST) have delivered important results in key areas for ITER/DEMO and the upcoming MAST Upgrade, a step towards future ST devices on the path to fusion currently under procurement. Micro-stability analysis...

  3. 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...

  4. Overview of progress in European medium sized tokamaks towards an integrated plasma-edge/wall solution

    NARCIS (Netherlands)

    Meyer, H.; Eich, T.; Beurskens, M.N.A.; Coda, S.; Hakola, A.; Martin, P.; Adamek, J.; Agostini, M.; Aguiam, D.; Ahn, J.; Aho-Mantila, L.; Akers, R.; Albanese, R.; Aledda, R.; Alessi, E.; Allan, S.; Alves, D.; Ambrosino, R.; Amicucci, L.; Anand, H.; Anastassiou, G.; Andrèbe, Y.; Angioni, C.; Apruzzese, G.; Ariola, M.; Arnichand, H.; Arter, W.; Baciero, A.; Barnes, M.; Barrera, L.; Behn, R.; Bencze, A.; Bernardo, J.; Bernert, M.; Bettini, P.; Bilková, P.; Bin, W.; Birkenmeier, G.; Bizarro, J. P.S.; Blanchard, P.; Blanken, T.; Bluteau, M.; Bobkov, V.; Bogar, O.; Böhm, P.; Bolzonella, T.; Boncagni, L.; Botrugno, A.; Bottereau, C.; Bouquey, F.; Bourdelle, C.; Brémond, S.; Brezinsek, S.; Brida, D.; Brochard, F.; Buchanan, J.; Bufferand, H.; Buratti, P.; Cahyna, P.; Calabrò, G.; Camenen, Y.; Caniello, R.; Cannas, B.; Canton, A.; Cardinali, A.; Carnevale, D.; Carr, M.; Carralero, D.; Carvalho, P.; Casali, L.; Castaldo, C.; Castejón, F.; Castro, R.; Causa, F.; Cavazzana, R.; Cavedon, M.; Cecconello, M.; Ceccuzzi, S.; Cesario, R.; Challis, C.D.; Chapman, I.T.; Chapman, S.; Chernyshova, M.; Choi, D.; Cianfarani, C.; Ciraolo, G.; Citrin, J.; Clairet, F.; Classen, I.; Coelho, R.; Coenen, J. W.; Colas, L.; Conway, G.; Corre, Y.; Costea, S.; Crisanti, F.; Cruz, N.; Cseh, G.; Czarnecka, A.; D'Arcangelo, O.; De Angeli, M.; De Masi, G.; De Temmerman, G.; De Tommasi, G.; Decker, J.; Delogu, R. S.; Dendy, R.; Denner, P.; Di Troia, C.; Dimitrova, M.; D'Inca, R.; Dorić, V.; Douai, D.; Drenik, A.; Dudson, B.; Dunai, D.; Dunne, M.; Duval, B. P.; Easy, L.; Elmore, S.; Erdös, B.; Esposito, B.; Fable, E.; Faitsch, M.; Fanni, A.; Fedorczak, N.; Felici, F.; Ferreira, J.; Février, O.; Ficker, O.; Fietz, S.; Figini, L.; Figueiredo, A.; Fil, A.; Fishpool, G.; Fitzgerald, M.; Fontana, M.; Ford, O.; Frassinetti, L.; Fridström, R.; Frigione, D.; Fuchert, G.; Fuchs, C.; Furno Palumbo, M.; Futatani, S.; Gabellieri, L.; Gałazka, K.; Galdon-Quiroga, J.; Galeani, S.; Gallart, D.; Gallo, A.; Galperti, C.; Gao, Y.; Garavaglia, S.; Garcia, J.; Garcia-Carrasco, A.; Garcia-Lopez, J.; Garcia-Munoz, M.; Gardarein, J. L.; Garzotti, L.; Gaspar, J.; Gauthier, E.; Geelen, P.; Geiger, B.; Ghendrih, P.; Ghezzi, F.; Giacomelli, L.; Giannone, L.; Giovannozzi, E.; Giroud, C.; Gleason González, C.; Gobbin, M.; Goodman, T. P.; Gorini, G.; Gospodarczyk, M.; Granucci, G.; Gruber, M.; Gude, A.; Guimarais, L.; Guirlet, R.; Gunn, J.; Hacek, P.; Hacquin, S.; Hall, S.; Ham, C.; Happel, T.; Harrison, J.; Harting, D.; Hauer, V.; Havlickova, E.; Hellsten, T.; Helou, W.; Henderson, S.; Hennequin, P.; Heyn, M.; Hnat, B.; Hölzl, M.; Hogeweij, D.; Honoré, C.; Hopf, C.; Horáček, J.; Hornung, G.; Horváth, L.; Huang, Z.; Huber, A.; Igitkhanov, J.; Igochine, V.; Imrisek, M.; Innocente, P.; Ionita-Schrittwieser, C.; Isliker, H.; Ivanova-Stanik, I.; Jacobsen, A. S.; Jacquet, P.; Jakubowski, M.; Jardin, A.; Jaulmes, F.; Jenko, F.; Jensen, T.; Jeppe Miki Busk, O.; Jessen, M.; Joffrin, E.; Jones, O.; Jonsson, T.; Kallenbach, A.; Kallinikos, N.; Kálvin, S.; Kappatou, A.; Karhunen, J.; Karpushov, A.; Kasilov, S.; Kasprowicz, G.; Kendl, A.; Kernbichler, W.; Kim, D.; Kirk, A.; Kjer, S.; Klimek, I.; Kocsis, G.; Kogut, D.; Komm, M.; Korsholm, S. B.; Koslowski, H. R.; Koubiti, M.; Kovacic, J.; Kovarik, K.; Krawczyk, N.; Krbec, J.; Krieger, K.; Krivska, A.; Kube, R.; Kudlacek, O.; Kurki-Suonio, T.; Labit, B.; Laggner, F. M.; Laguardia, L.; Lahtinen, A.; Lalousis, P.; Lang, P.; Lauber, P.; Lazányi, N.; Lazaros, A.; Le, H.B.; Lebschy, A.; Leddy, J.; Lefévre, L.; Lehnen, M.; Leipold, F.; Lessig, A.; Leyland, M.; Li, L.; Liang, Y.; Lipschultz, B.; Liu, Y.Q.; Loarer, T.; Loarte, A.; Loewenhoff, T.; Lomanowski, B.; Loschiavo, V. P.; Lunt, T.; Lupelli, I.; Lux, H.; Lyssoivan, A.; Madsen, J.; Maget, P.; Maggi, C.; Maggiora, R.; Magnussen, M. L.; Mailloux, J.; Maljaars, B.; Malygin, A.; Mantica, P.; Mantsinen, M.; Maraschek, M.; Marchand, B.; Marconato, N.; Marini, C.; Marinucci, M.; Markovic, T.; Marocco, D.; Marrelli, L.; Martin, Y.; Martin Solis, J. R.; Martitsch, A.; Mastrostefano, S.; Mattei, M.; Matthews, G.; Mavridis, M.; Mayoral, M. L.; Mazon, D.; McCarthy, P.; McAdams, R.; McArdle, G.; McCarthy, P.; McClements, K.; McDermott, R.; McMillan, B.; Meisl, G.; Merle, A.; Meyer, O.; Milanesio, D.; Militello, F.; Miron, I. G.; Mitosinkova, K.; Mlynar, J.; Mlynek, A.; Molina, D.; Molina, P.; Monakhov, I.; Morales, J.; Moreau, D.; Morel, P.; Moret, J. M.; Moro, A.; Moulton, D.; Müller, H. W.; Nabais, F.; Nardon, E.; Naulin, V.; Nemes-Czopf, A.; Nespoli, F.; Neu, R.; Nielsen, A. H.; Nielsen, S. K.; Nikolaeva, V.; Nimb, S.; Nocente, M.; Nouailletas, R.; Nowak, S.; Oberkofler, M.; Oberparleiter, M.; Ochoukov, R.; Odstrčil, T.; Olsen, J.; Omotani, J.; O'Mullane, M. G.; Orain, F.; Osterman, N.; Paccagnella, R.; Pamela, S.; Pangione, L.; Panjan, M.; Papp, G.; Papřok, R.; Parail, V.; Parra, F. I.; Pau, A.; Pautasso, G.; Pehkonen, S. P.; Pereira, A.; Perelli Cippo, E.; Pericoli Ridolfini, V.; Peterka, M.; Petersson, P.; Petrzilka, V.; Piovesan, P.; Piron, C.; Pironti, A.; Pisano, F.; Pisokas, T.; Pitts, R.; Ploumistakis, I.; Plyusnin, V.; Pokol, G.; Poljak, D.; Pölöskei, P.; Popovic, Z.; Pór, G.; Porte, L.; Potzel, S.; Predebon, I.; Preynas, M.; Primc, G.; Pucella, G.; Puiatti, M. E.; Pütterich, T.; Rack, M.; Ramogida, G.; Rapson, C.; Rasmussen, J. Juul; Rasmussen, J.; Rattá, G. A.; Ratynskaia, S.; Ravera, G.; Réfy, D.; Reich, M.; Reimerdes, H.; Reimold, F.; Reinke, M.; Reiser, D.; Resnik, M.; Reux, C.; Ripamonti, D.; Rittich, D.; Riva, G.; Rodriguez-Ramos, M.; Rohde, V.; Rosato, J.; Ryter, F.; Saarelma, S.; Sabot, R.; Saint-Laurent, F.; Salewski, M.; Salmi, A.; Samaddar, D.; Sanchis-Sanchez, L.; Santos, J.; Sauter, O.; Scannell, R.; Scheffer, M.; Schneider, M.; Schneider, B.; Schneider, P.; Schneller, M.; Schrittwieser, R.; Schubert, M.; Schweinzer, J.; Seidl, J.; Sertoli, M.; Šesnić, S.; Shabbir, A.; Shalpegin, A.; Shanahan, B.; Sharapov, S.; Sheikh, U.; Sias, G.; Sieglin, B.; Silva, C.; Silva, A.; Silva Fuglister, M.; Simpson, J.; Snicker, A.; Sommariva, C.; Sozzi, C.; Spagnolo, S.; Spizzo, G.; Spolaore, M.; Stange, T.; Stejner Pedersen, M.; Stepanov, I.; Stober, J.; Strand, P.; Šušnjara, A.; Suttrop, W.; Szepesi, T.; Tál, B.; Tala, T.; Tamain, P.; Tardini, G.; Tardocchi, M.; Teplukhina, A.; Terranova, D.; Testa, D.; Theiler, C.; Thornton, A.; Tolias, P.; Tophj, L.; Treutterer, W.; Trevisan, G. L.; Tripsky, M.; Tsironis, C.; Tsui, C.; Tudisco, O.; Uccello, A.; Urban, J.; Valisa, M.; Vallejos, P.; Valovic, M.; Van Den Brand, H.; Vanovac, B.; Varoutis, S.; Vartanian, S.; Vega, J.; Verdoolaege, G.; Verhaegh, K.; Vermare, L.; Vianello, N.; Vicente, J.; Viezzer, E.; Vignitchouk, L.; Vijvers, W.A.J.; Villone, F.; Viola, B.; Vlahos, L.; Voitsekhovitch, I.; Vondráček, P.; Vu, N. M.T.; Wagner, D.; Walkden, N.; Wang, N.; Wauters, T.; Weiland, M.; Weinzettl, V.; Westerhof, E.; Wiesenberger, M.; Willensdorfer, M.; Wischmeier, M.; Wodniak, I.; Wolfrum, E.; Yadykin, D.; Zagórski, R.; Zammuto, I.; Zanca, P.; Zaplotnik, R.; Zestanakis, P.; Zhang, W.; Zoletnik, S.; Zuin, M.

    2017-01-01

    Integrating the plasma core performance with an edge and scrape-off layer (SOL) that leads to tolerable heat and particle loads on the wall is a major challenge. The new European medium size tokamak task force (EU-MST) coordinates research on ASDEX Upgrade (AUG), MAST and TCV. This multi-machine

  5. Overview of progress in European medium sized tokamaks towards an integrated plasma-edge/wall solution

    DEFF Research Database (Denmark)

    Meyer, H.; Eich, T.; Beurskens, M.

    2017-01-01

    Integrating the plasma core performance with an edge and scrape-off layer (SOL) that leads to tolerable heat and particle loads on the wall is a major challenge. The new European medium size tokamak task force (EU-MST) coordinates research on ASDEX Upgrade (AUG), MAST and TCV. This multi-machine ...

  6. Temperature and impurity transport studies of heated tokamak plasmas by means of a collisional-radiative model of x-ray emission from Mo30+ to Mo39+

    International Nuclear Information System (INIS)

    Pacella, D.; Fournier, K. B.; Zerbini, M.; Finkenthal, M.; Mattioli, M.; May, M. J.; Goldstein, W. H.

    2000-01-01

    This work presents and interprets, by means of detailed atomic calculations, observations of L-shell (n=3→n=2) transitions in highly ionized molybdenum, the main intrinsic heavy impurity in the Frascati tokamak upgrade plasmas. These hot plasmas were obtained by additional electron cyclotron resonance heating (ECRH), at the frequency of 140 Ghz, during the current ramp-up phase of the discharge. Injecting 400 kW on axis and 800 kW slightly off axis, the peak central electron temperature reached 8.0 and 7.0 keV, respectively, for a time much longer than the ionization equilibrium time of the molybdenum ions. X-ray emissions from rarely observed high charge states, Mo 30+ to Mo 39+ , have been studied with moderate spectral resolution (λ/Δλ∼150) and a time resolution of 5 ms. A sophisticated collisional-radiative model for the study of molybdenum ions in plasmas with electron temperature in the range 4-20 keV is presented. The sensitivity of the x-ray emission to the temperature and to impurity transport processes is discussed. This model has been then used to investigate two different plasma scenarios. In the first regime the ECRH heating occurs on axis during the current ramp up phase, when the magnetic shear is evolving from negative to zero up to the half radius. The spectrum is well reproduced with the molybdenum ions in coronal equilibrium and with a central impurity peaking. In the second regime, at the beginning of the current flat top when magnetic shear is monotonic and sawtoothing activity is appearing, the lowest charge states (Mo 33+ to Mo 30+ ), populated off axis, are affected by anomalous transport and the total molybdenum profile is found to be flat up to the half radius. We conclude with the presentation of ''synthetic spectra'' computed for even higher temperature plasmas that are expected in future experiments with higher ECRH power input. (c) 2000 The American Physical Society

  7. Joint research using small tokamaks

    International Nuclear Information System (INIS)

    Gryaznevich, M.P.; Del Bosco, E.; Malaquias, A.; Mank, G.; Oost, G. van

    2005-01-01

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

  8. Joint research using small tokamaks

    International Nuclear Information System (INIS)

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

    2005-01-01

    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

  9. Upgrading uncompetitive products economically

    DEFF Research Database (Denmark)

    Lu, Hua; Jensen, Christian Søndergaard

    2012-01-01

    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...

  10. Maintenance procedure upgrade programs

    International Nuclear Information System (INIS)

    Campbell, J.J.; Zimmerman, C.M.

    1988-01-01

    This paper describes a systematic approach to upgrading nuclear power plant maintenance procedures. The approach consists of four phases: diagnosis, program planning, program implementation, and program evaluation. Each phase is explained as a series of steps to ensure that all factors in a procedure upgrade program are considered

  11. NSLS control system upgrade

    International Nuclear Information System (INIS)

    Smith, J.D.; Ramamoorthy, Susila; Tang, Y.N.

    1994-01-01

    The NSLS consists of two storage rings, a booster and a linac. A major upgrade of the control system (installed in 1978) was undertaken and has been completed. The computer architecture is being changed from a three level star-network to a two level distributed system. The microprocessor subsystem, host computer and workstations, communication link and the main software components are being upgraded or replaced. Since the NSLS rings operate twenty four hours a day a year with minimum maintenance time, the key requirement during the upgrade phase is a non-disruptive transition with minimum downtime. Concurrent with the upgrade, some immediate improvements were required. This paper describes the various components of the upgraded system and outlines the future plans. ((orig.))

  12. NSLS control system upgrade

    International Nuclear Information System (INIS)

    Smith, J.D.; Ramamoorthy, S.; Tang, Yong N.

    1995-01-01

    The NSLS consists of two storage rings, a booster and a linac. A major upgrade of the control system (installed in 1978) was undertaken and has been completed. The computer architecture is being changed from a three level star-network to a two level distributed system. The microprocessor subsystem, host computer and workstations, communication link and the main software components are being upgraded or replaced. Since the NSLS rings operate twenty four hours a day a year with minimum maintenance time, the key requirement during the upgrade phase is a non-disruptive transition with minimum downtime. Concurrent with the upgrade, some immediate improvements were required. This paper describes the various components of the upgraded system and outlines the future plans

  13. Advanced commercial tokamak study

    International Nuclear Information System (INIS)

    Thomson, S.L.; Dabiri, A.E.; Keeton, D.C.; Brown, T.G.; Bussell, G.T.

    1985-12-01

    Advanced commercial tokamak studies were performed by the Fusion Engineering Design Center (FEDC) as a participant in the Tokamak Power Systems Studies (TPSS) project coordinated by the Office of Fusion Energy. The FEDC studies addressed the issues of tokamak reactor cost, size, and complexity. A scoping study model was developed to determine the effect of beta on tokamak economics, and it was found that a competitive cost of electricity could be achieved at a beta of 10 to 15%. The implications of operating at a beta of up to 25% were also addressed. It was found that the economics of fusion, like those of fission, improve as unit size increases. However, small units were found to be competitive as elements of a multiplex plant, provided that unit cost and maintenance time reductions are realized for the small units. The modular tokamak configuration combined several new approaches to develop a less complex and lower cost reactor. The modular design combines the toroidal field coil with the reactor structure, locates the primary vacuum boundary at the reactor cell wall, and uses a vertical assembly and maintenance approach. 12 refs., 19 figs

  14. To upgrade or not to upgrade

    DEFF Research Database (Denmark)

    Rasmussen, Torben Valdbjørn; Rose, Jørgen

    2017-01-01

    This paper identifies the key indicators that owners need to take into account in order to choose the most affordable extent of upgrading of a typical post-1945 building. The exterior look of the building is not to be changed. Indicators include measures, risk assessment related to the changes in...

  15. Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

    International Nuclear Information System (INIS)

    Riedel, K.S.

    1989-03-01

    This is a compendium of three separate articles on the statistical analysis of tokamak transport. The first article is an expository introduction to advanced statistics and scaling laws. The second analyzes two important problems of tokamak data---colinearity and tokamak to tokamak variation in detail. The third article generalizes the Swamy random coefficient model to the case of degenerate matrices. Three papers have been processed separately

  16. Microwave Tokamak Experiment

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

    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

  17. Magnetic ''islandography'' in tokamaks

    International Nuclear Information System (INIS)

    Callen, J.D.; Waddell, B.V.; Hicks, H.R.

    1978-09-01

    Tearing modes are shown to be responsible for most of the experimentally observed macroscopic behavior of tokamak discharges. The effects of these collective magnetic perturbations on magnetic topology and plasma transport in tokamaks are shown to provide plausible explanations for: internal disruptions (m/n = 1); Mirnov oscillations (m/n = 2,3...); and major disruptions (coupling of 2/1-3/2 modes). The nonlinear evolution of the tearing modes is followed with fully three-dimensional computer codes. The effects on plasma confinement of the magnetic islands or stochastic field lines induced by the macroscopic tearing modes are discussed and compared with experiment. Finally, microscopic magnetic perturbations are shown to provide a natural model for the microscopic anomalous transport processes in tokamaks

  18. Accelerator technology in tokamaks

    International Nuclear Information System (INIS)

    Kustom, R.L.

    1977-01-01

    This article presents the similarities in the technology required for high energy accelerators and tokamak fusion devices. The tokamak devices and R and D programs described in the text represent only a fraction of the total fusion program. The technological barriers to producing successful, economical tokamak fusion power plants are as many as the plasma physics problems to be overcome. With the present emphasis on energy problems in this country and elsewhere, it is very likely that fusion technology related R and D programs will vigorously continue; and since high energy accelerator technology has so much in common with fusion technology, more scientists from the accelerator community are likely to be attracted to fusion problems

  19. Upgrading during difficult times

    International Nuclear Information System (INIS)

    Tiefenbach, K.

    1993-01-01

    The Co-op Upgrader is part of an integrated refinery and upgrader complex in Regina, Saskatchewan. The upgrader processes 50,000 bbl/d heavy sour crude oil, mostly Fosterton and dilute Lloydminster crude, via hydrogen addition and carbon rejection, desulfurization, demetallization, and denitrification to yield a synthetic crude blend. The synthetic crude is refined to produce gasoline and diesel fuel. Byproducts from the integrated operation include 100,000 tonnes/y of petroleum coke, 65,000 tonnes/y of sulfur, propane, butane, fuel oil, and metals (Ni and V) in the form of spent catalysts. Recent operational and economic challenges faced by the upgrader are reviewed. Technical challenges include operating the upgrader's high-temperature high-pressure heavy oil hydrotreating unit and distillate hydrogenation unit, removal and replacement of the desulfurization catalyst, waste management, and producing coke of sufficient quality. Economic challenges include the shrinking differential between light and heavy oil, higher prices for natural gas (the main raw material for hydrogen production for upgrading), seasonal changes in product specifications, and lower prices for sulfur and metal byproducts. The upgrader is also affected by interest rates since borrowing costs are the single largest expenditure after crude oil purchases. 4 figs

  20. ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    Steiner, D.; Embrechts, M.

    1990-07-01

    This is a status report on technical progress relative to the tasks identified for the fifth year of Grant No. FG02-85-ER52118. The ARIES tokamak reactor study is a multi-institutional effort to develop several visions of the tokamak as an attractive fusion reactor with enhanced economic, safety, and environmental features. The ARIES study is being coordinated by UCLA and involves a number of institutions, including RPI. The RPI group has been pursuing the following areas of research in the context of the ARIES-I design effort: MHD equilibrium and stability analyses; plasma-edge modeling and blanket materials issues. Progress in these areas is summarized herein

  1. Internal disruption in tokamaks

    International Nuclear Information System (INIS)

    Kuvshinov, B.N.; Savrukhin, P.V.

    1990-01-01

    A review of results of experimental and theoretical investigations of internal disruption in tokamaks is given. Specific features of various types of saw-tooth oscillations are described and their classification is performed. Theoretical models of the process of development of internal disruption instability are discussed. Effect of internal disruption on parameters of plasma, confined in tokamak, is considered. Scalings of period and amplitude of saw-tooth oscillations, as well as version radius are presented. Different methods for stabilizing instability of internal disruption are described

  2. Overview of Tokamak Results

    International Nuclear Information System (INIS)

    Unterberg, Bernhard; Samm, Ulrich

    2004-01-01

    An overview is given of recent results obtained in tokamak devices. We introduce basic confinement scenarios as L-mode, H-mode and plasmas with an internal transport barrier and discuss methods for profile control. Important findings in DT-experiments at JET as α-particle heating are described. Methods for power exhaust like plasma regimes with a radiating mantle and radiative divertor scenarios are discussed. The overall impact of plasma edge conditions on the general plasma performance in tokamaks is illustrated by describing the impact of wall conditions on confinement and the edge operational diagram of H-mode plasmas

  3. Internal disruptions in tokamaks

    International Nuclear Information System (INIS)

    Kuvshinov, B.N.; Savrukhin, P.V.

    1990-01-01

    Experimental and theoretical studies of the phenomenon of internal disruptions in tokamaks are reviewed. A classification scheme is introduced and the features of different types of sawtooth oscillations are described. A theoretical model for the development of the internal disruption instability is discussed. The effect of internal disruptions on the parameters of plasma confined in tokamaks is discussed. Scaling laws for the period and amplitude of sawtooth oscillations, as well as for the inversion radius, are presented. Different methods of stabilizing the internal disruption instability are described

  4. The energy principle applied to diverted tokamak configurations

    International Nuclear Information System (INIS)

    Atanasiu, C. V.; Guenter, S.; Lackner, K.; Moraru, A.; Zakharov, L. E.; Subbotin, A. A.

    2008-01-01

    Writing the expression of the potential energy in terms of the perturbation of the flux function, and performing an Euler minimisation, one obtains a system of ordinary differential equations in that perturbation. For a diverted configuration, the usual vanishing boundary conditions for the perturbed flux function at the magnetic axis and at infinity can no longer be used. In place of the vanishing boundary conditions at infinity, an approach to fix 'natural' boundary conditions for the system of differential equations for the perturbed flux function, just at the plasma boundary has been developed. As an example of application of our approaches, a particular equilibrium configuration of the ASDEX Upgrade tokamak has been considered and a detailed investigation of the dependence of the tearing stability parameter Δ' on plasma shape is given for a realistic tokamak equilibrium. The results shown are at least in qualitative agreement with experimental observations on ASDEX Upgrade and JET of a stabilizing influence of triangularity. The knowledge of Δ' for realistic tokamak plasmas is especially important for understanding of the plasma stability against NTMs. (authors)

  5. 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.

  6. Bi-Provincial Upgrader

    International Nuclear Information System (INIS)

    1997-01-01

    Husky Oil's Bi-Provincial Upgrader is located in the rural municipality of Wilton, east of Lloydminster, Saskatchewan. It is jointly owned by Husky Oil and the Saskatchewan government. The upgrader is designed to produce 7.300 m 3 /cd of synthetic crude from 8.440 m 3 /cd of diluted Lloyd blend and distillates from the existing Lloydminster refinery. The designed coke and sulphur production is 415 t/day and 240 t/day respectively. Chronology of the Bi-Provincial Upgrader project was presented, along with details of the heavy oil feedstock properties for Cold Lake, Lloydminster and Lloyd Blend. Upgrader production history since start-up, synthetic crude oil customers, and the evolution of prices for Edmonton light vs. Husky LLB at Hardisty were also reviewed. 3 tabs., 9 figs

  7. 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.

  8. High beta tokamaks

    International Nuclear Information System (INIS)

    Dory, R.A.; Berger, D.P.; Charlton, L.A.; Hogan, J.T.; Munro, J.K.; Nelson, D.B.; Peng, Y.K.M.; Sigmar, D.J.; Strickler, D.J.

    1978-01-01

    MHD equilibrium, stability, and transport calculations are made to study the accessibility and behavior of ''high beta'' tokamak plasmas in the range β approximately 5 to 15 percent. For next generation devices, beta values of at least 8 percent appear to be accessible and stable if there is a conducting surface nearby

  9. Sawtooth phenomena in tokamaks

    International Nuclear Information System (INIS)

    Kuvshinov, B.N.; Savrukhin, P.V.

    1989-01-01

    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

  10. Reconnection in tokamaks

    International Nuclear Information System (INIS)

    Pare, V.K.

    1983-01-01

    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

  11. Research using small tokamaks

    International Nuclear Information System (INIS)

    1993-01-01

    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

  12. Research using small tokamaks

    International Nuclear Information System (INIS)

    1991-01-01

    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

  13. Compact tokamak reactors

    International Nuclear Information System (INIS)

    Wootton, A.J.; Wiley, J.C.; Edmonds, P.H.; Ross, D.W.

    1997-01-01

    The possible use of tokamaks for thermonuclear power plants is discussed, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First, the existing literature is reviewed and summarized. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamaks power plant, by including the power required to drive the toroidal field and by considering two extremes of plasma current drive efficiency. Third, the analytic results are augmented by a numerical calculation that permits arbitrary plasma current drive efficiency and different confinement scaling relationships. Throughout, the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculation of electric power. The latest published reactor studies show little advantage in using low aspect ratios to obtain a more compact device (and a low cost of electricity) unless either remarkably high efficiency plasma current drive and low safety factor are combined, or unless confinement (the H factor), the permissible elongation and the permissible neutron wall loading increase as the aspect ratio is reduced. These results are reproduced with the analytic model. (author). 22 refs, 3 figs

  14. Overview of the Tokamak de Varennes program

    International Nuclear Information System (INIS)

    Pacher, H.D.

    1986-01-01

    The Tokamak de Varennes will be the major Canadian experiment in the magnetic fusion domain. It has a toroidal field of 1.5 tesla, major radius of 0.85 m, a minor radius of 0.25 m, and will study long pulses, up to 30 seconds duration. Initially, a series of successive plasma pulses, each of the order of seconds, will yield a duty factor of over 50 percent. During this phase, the major emphasis will be on the study of impurity generation, transport, and control, plasma-wall interactions and material properties. The program will include studies of fast current rampdown and the resultant current profile modifications. The development of advanced diagnostics will also be undertaken. To attain a higher duty factor with continuous plasma operation, noninductive current drive by radio=frequency will be added as an early upgrade. This will introduce current drive investigations such as transformer recharge and profile relaxation, and enhance the wall and materials study program. In this context, the Tokamak de Varennes will concentrate on the study of impurity exhaust and retention as well as net erosion of the limiter and neutralization plate materials

  15. Tokamaks (Second Edition)

    Energy Technology Data Exchange (ETDEWEB)

    Stott, Peter [JET, UK (United Kingdom)

    1998-10-01

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

  16. Tokamaks (Second Edition)

    International Nuclear Information System (INIS)

    Stott, Peter

    1998-01-01

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

  17. Present status of Tokamak research

    International Nuclear Information System (INIS)

    Basu, Jayanta

    1991-01-01

    The scenario of thermonuclear fusion research is presented, and the tokamak which is the most promising candidate as a fusion reactor is introduced. A brief survey is given of the most noteworthy tokamaks in the global context, and fusion programmes relating to Next Step devices are outlined. Supplementary heating of tokamak plasma by different methods is briefly reviewed; the latest achievements in heating to fusion temperatures are also reported. The progress towards the high value of the fusion product necessary for ignition is described. The improvement in plasma confinement brought about especially by the H-mode, is discussed. The latest situation in pushing up Β for increasing the efficiency of a tokamak is elucidated. Mention is made of the different types of wall treatment of the tokamak vessel for impurity control, which has led to a significant improvement in tokamak performance. Different methods of current drive for steady state tokamak operation are reviewed, and the issue of current drive efficiency is addressed. A short resume is given of the various diagnostic methods which are employed on a routine basis in the major tokamak centres. A few diagnostics recently developed or proposed in the context of the advanced tokamaks as well as the Next Step devices are indicated. The important role of the interplay between theory, experiment and simulation is noted, and the areas of investigation requiring concerted effort for further progress in tokamak research are identified. (author). 17 refs

  18. Zeff from spectroscopic bremsstrahlung measurements at ASDEX Upgrade and JET

    International Nuclear Information System (INIS)

    Meister, H.; Fischer, R.; Horton, L.D.; Maggi, C.F.; Nishijima, D.; Giroud, C.; Zastrow, K.-D.; Zaniol, B.

    2004-01-01

    The effective ionic charge Z eff is a means to assess the impurity content of a fusion plasma. It can be derived from measurements of bremsstrahlung intensity. These have been extended at ASDEX Upgrade by the usage of the sight lines for the charge exchange recombination diagnostic. Together with a previously installed sight line array, it is now possible to routinely determine the bremsstrahlung intensity over the whole minor radius purely from spectroscopic measurements. In a tokamak where the plasma facing components are made up of various materials, this is necessary to check if measurements are contaminated by line radiation. The bremsstrahlung background of the respective spectra is determined using Bayesian probability theory, giving consistent and improved error statistics. Using the information for electron temperature and density profiles, the Z eff profile is determined by an integrated method. The same approach to assess the Z eff profile has been demonstrated to be successful also at the JET tokamak

  19. Large Aspect Ratio Tokamak Study

    International Nuclear Information System (INIS)

    Reid, R.L.; Holmes, J.A.; Houlberg, W.A.; Peng, Y.K.M.; Strickler, D.J.; Brown, T.G.; Wiseman, G.W.

    1980-06-01

    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

  20. The ALICE TPC Upgrad

    Science.gov (United States)

    Castro, Andrew; Alice-Usa Collaboration; Alice-Tpc Collaboration

    2017-09-01

    The Time Projection Chamber (TPC) currently used for ALICE (A Large Ion Collider Experiment at CERN) is a gaseous tracking detector used to study both proton-proton and heavy-ion collisions at the Large Hadron Collider (LHC) In order to accommodate the higher luminosit collisions planned for the LHC Run-3 starting in 2021, the ALICE-TPC will undergo a major upgrade during the next LHC shut down. The TPC is limited to a read out of 1000 Hz in minimum bias events due to the intrinsic dead time associated with back ion flow in the multi wire proportional chambers (MWPC) in the TPC. The TPC upgrade will handle the increase in event readout to 50 kHz for heavy ion minimum bias triggered events expected with the Run-3 luminosity by switching the MWPCs to a stack of four Gaseous Electron Multiplier (GEM) foils. The GEM layers will combine different hole pitches to reduce the dead time while maintaining the current spatial and energy resolution of the existing TPC. Undertaking the upgrade of the TPC represents a massive endeavor in terms of design, production, construction, quality assurance, and installation, thus the upgrade is coordinated over a number of institutes worldwide. The talk will go over the physics motivation for the upgrade, the ALICE-USA contribution to the construction of Inner Read Out Chambers IROCs, and QA from the first chambers built in the U.S

  1. Tokamak reactor startup power

    International Nuclear Information System (INIS)

    Weldon, D.M.; Murray, J.G.

    1983-01-01

    Tokamak startup with ohmic heating (OH)-induced voltages requires rather large voltages and power supplies. On present machines, with no radiofrequency (rf)-assist provisions, hundreds of volts have been specified for their designs. With the addition of electron cyclotron resonant heating (ECRH) assist, the design requirements have been lowered. To obtain information on the cost and complexity associated with this ECRH-assisted, OH-pulsed startup voltage for ignition-type machines, a trade-off study was completed. The Fusion Engineering Device (FED) configuration was selected as a model because information was available on the structure. The data obtained are applicable to all tokamaks of this general size and complexity, such as the Engineering Test Reactor

  2. Tokamak fusion reactor exhaust

    International Nuclear Information System (INIS)

    Harrison, M.F.A.; Harbour, P.J.; Hotston, E.S.

    1981-08-01

    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.)

  3. Theory of tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    White, R B [Princeton Univ., NJ (USA). Plasma Physics Lab.

    1989-01-01

    The book covers the consequences of ideal and resistive magnetohydrodynamics, these theories being responsible for most of what is well understood regarding the physics of tokamak discharges. The focus is on the description of equilibria, the linear and nonlinear theory of large scale modes, and single particle guiding center motion, including simple neoclassical effects. modern methods of general magnetic coordinates are used, and the student is introduced to the onset of chaos in Hamiltonian systems in the discussion of destruction of magnetic surfaces. Much of the book is devoted to the description of the limitations placed on tokamak operating parameters given by ideal and resistive modes, and current ideas about how to extend and optimize these parameters. (author). refs.; figs.

  4. Axisymmetric tokamak scapeoff transport

    International Nuclear Information System (INIS)

    Singer, C.E.; Langer, W.D.

    1982-08-01

    We present the first self-consistent estimate of the magnitude of each term in a fluid treatment of plasma transport for a plasma lying in regions of open field lines in an axisymmetric tokamak. The fluid consists of a pure hydrogen plasma with sources which arise from its interaction with neutral hydrogen atoms. The analysis and results are limited to the high collisionality regime, which is optimal for a gaseous neutralizer divertor, or to a cold plasma mantle in a tokamak reactor. In this regime, both classical and neoclassical transport processes are important, and loss of particles and energy by diamagnetic flow are also significant. The prospect of extending the analysis to the lower collisionality regimes encountered in many existing experiments is discussed

  5. Density limits in Tokamaks

    International Nuclear Information System (INIS)

    Tendler, M.

    1984-06-01

    The energy loss from a tokamak plasma due to neutral hydrogen radiation and recycling is of great importance for the energy balance at the periphery. It is shown that the requirement for thermal equilibrium implies a constraint on the maximum attainable edge density. The relation to other density limits is discussed. The average plasma density is shown to be a strong function of the refuelling deposition profile. (author)

  6. Tokamak pump limiters

    International Nuclear Information System (INIS)

    Conn, R.W.

    1984-05-01

    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

  7. Modular tokamak magnetic system

    International Nuclear Information System (INIS)

    Yang, T.F.

    1988-01-01

    This patent describes a tokamak reactor including a vacuum vessel, toroidal confining magnetic field coils disposed concentrically around the minor radius of the vacuum vessel, and poloidal confining magnetic field coils, an ohmic heating coil system comprising at least one magnetic coil disposed concentrically around a toroidal field coil, wherein the magnetic coil is wound around the toroidal field coil such that the ohmic heating coil enclosed the toroidal field coil

  8. Tokamak pump limiters

    International Nuclear Information System (INIS)

    Conn, R.W.; California Univ., Los Angeles

    1984-01-01

    Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6 MW of auxiliary neutral beam heating. Experiments have also been performed 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 scrape-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 regime 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. (orig.)

  9. The ASDEX Upgrade ICRF system: Operational experience and developments

    International Nuclear Information System (INIS)

    Faugel, H.; Angene, P.; Becker, W.; Braun, F.; Bobkov, Vl.V.; Eckert, B.; Fischer, F.; Hartmann, D.A.; Heilmaier, G.; Kneidl, J.; Noterdaeme, J.-M.; Siegl, G.; Wuersching, E.

    2005-01-01

    The ICRF system at the ASDEX Upgrade tokamak is in operation since May 1992. Following some modifications of which the major one was the installation of 3 dB couplers it has become a reliable additional heating system. The maximum power coupled into the plasma has been raised up to 7.2 MW (90% of the installed RF power) for short pulses and up to 6.2 MW for pulses several second long with energy of up to 29 MJ. A power of 5 MW is delivered on a regular basis to replace two NBI sources

  10. ATLAS Upgrade Programme

    CERN Document Server

    Hillier, S J; The ATLAS collaboration

    2012-01-01

    With the already outstanding LHC luminosity performance, and planned LHC upgrades in the upcoming shutdowns, it is expected that within a short time-scale, the general purpose LHC experiments will have to cope with luminosities beyond their original design. In order to maintain detector performance and sensitivity to expected and new physics processes, ATLAS has defined a continuous upgrade programme which foresees staged enhancements during the next 10 years of operation, and then more widespread changes before the transition to the highest luminosities after 2022. This talk will describe several components of the ATLAS upgrade, focusing in particular on the Inner Detector and Trigger. The Inner Detector faces two challenges in the higher luminosity environment: high particle multiplicities and increased radiation dose. These will be addressed in the short term by a new layer of Pixel detectors, and in the long term by a complete replacement. The Trigger faces an increasingly difficult task of distinguishing...

  11. Upgradation of Apsara reactor

    International Nuclear Information System (INIS)

    Mammen, S.; Mukherjee, P.; Bhatnagar, A.; Sasidharan, K.; Raina, V.K.

    2009-01-01

    Apsara is a 1 MW swimming pool type research reactor using high enriched uranium as fuel with light water as coolant and moderator. The reactor is in operation for more than five decades and has been extensively used for basic research, radioisotope production, neutron radiography, detector testing, shielding experiments etc. In view of its long service period, it is planned to carry out refurbishment of the reactor to extend its useful life. During refurbishment, it is also planned to upgrade the reactor to a 2 MW reactor to improve its utilization and to upgrade the structure, system and components in line with the current safety standards. This paper gives a brief account of the design features and safety aspects of the upgraded Apsara reactor. (author)

  12. TPX tokamak construction management

    International Nuclear Information System (INIS)

    Knutson, D.; Kungl, D.; Seidel, P.; Halfast, C.

    1995-01-01

    A construction management contract normally involves the acquisition of a construction management firm to assist in the design, planning, budget conformance, and coordination of the construction effort. In addition the construction management firm acts as an agent in the awarding of lower tier contracts. The TPX Tokamak Construction Management (TCM) approach differs in that the construction management firm is also directly responsible for the assembly and installation of the tokamak including the design and fabrication of all tooling required for assembly. The Systems Integration Support (SIS) contractor is responsible for the architect-engineering design of ancillary systems, such as heating and cooling, buildings, modifications and site improvements, and a variety of electrical requirements, including switchyards and >4kV power distribution. The TCM will be responsible for the procurement of materials and the installation of the ancillary systems, which can either be performed directly by the TCM or subcontracted to a lower tier subcontractor. Assurance that the TPX tokamak is properly assembled and ready for operation when turned over to the operations team is the primary focus of the construction management effort. To accomplish this a disciplined constructability program will be instituted. The constructability effort will involve the effective and timely integration of construction expertise into the planning, component design, and field operations. Although individual component design groups will provide liaison during the machine assembly operations, the construction management team is responsible for assembly

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

    NARCIS (Netherlands)

    Sommer, F.; Günter, S.; Kallenbach, A.; Maraschek, M.; Boom, J.E.; Fischer, R.; Hicks, N.; Luhmann, N.C.; Park, H.K.; Reiter, B.; Wenninger, R.; Wolfrum, E.

    2011-01-01

    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

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

    NARCIS (Netherlands)

    Geiger, B.; M. García-Muñoz,; 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

  15. The UKIRT Upgrades Programme

    Science.gov (United States)

    Adamson, Andy; Davies, John; Robson, Ian

    Tim Hawarden presented this paper to the 30th anniversary workshop, just a month before his untimely death. The editors have done their best to convert his talk into this paper, and gratefully acknowledge the assistance of Nick Rees (a member of the Upgrades team, now at Diamond Light Source). Tim's discussion concerned the UKIRT Upgrades Project, which ran through the 1990s and transformed the telescope and made it truly competitive on the world stage for operation into the twenty-first century. The reference list at the end of the paper is comprehensive; some of these are referred to in the paper itself and some are included for completeness only.

  16. AGS intensity upgrades

    International Nuclear Information System (INIS)

    Roser, T.

    1995-01-01

    After the successful completion of the AGS Booster and several upgrades of the AGS, a new intensity record of 6.3 x 10 13 protons per pulse accelerated to 24 GeV was achieved. The high intensity slow-extracted beam program at the AGS typically serves about five production targets and about eight experiments including three rare Kaon decay experiments. Further intensity upgrades are being discussed that could increase the average delivered beam intensity by up to a factor of four

  17. OMEGA upgrade staging options

    International Nuclear Information System (INIS)

    Kelly, J.H.; Shoup, M.J.; Smith, D.L.

    1989-01-01

    The authors discuss how they are designing an upgrade to its 24-beam OMEGA laser system, OMEGA is a frequency tripled, all-rod system capable of producing 2 kJ at 0.8 ns on target. Important direct-drive-target-ignition physics could be investigated with an upgraded system capable of producing a shaped pulse consisting of a long (5ns) low-intensity, foot, smoothly transitioning into a short (0.5 ns), intense, compression pulse. The total pulse energy is 30 kJ, which, from target-irradiation uniformity considerations, must be distributed over 60 beams

  18. Upgrades and Real Time Ntm Control Application of the Ece Radiometer on Asdex Upgrade

    Science.gov (United States)

    Hicks, N. K.; Suttrop, W.; Behler, K.; Giannone, L.; Manini, A.; Maraschek, M.; Raupp, G.; Reich, M.; Sips, A. C. C.; Stober, J.; Treutterer, W.; ASDEX Upgrade Team; Cirant, S.

    2009-04-01

    The 60-channel electron cyclotron emission (ECE) radiometer diagnostic on the ASDEX Upgrade tokamak is presently being upgraded to include a 1 MHz sampling rate data acquisition system. This expanded capability allows electron temperature measurements up to 500 kHz (anti-aliasing filter cut-off) with spatial resolution ~1 cm, and will thus provide measurement of plasma phenomena on the MHD timescale, such as neoclassical tearing modes (NTMs). The upgraded and existing systems may be run in parallel for comparison, and some of the first plasma measurements using the two systems together are presented. A particular planned application of the upgraded radiometer is integration into a real-time NTM stabilization loop using targeted deposition of electron cyclotron resonance heating (ECRH). For this loop, it is necessary to determine the locations of the NTM and ECRH deposition using ECE measurements. As the magnetic island of the NTM repeatedly rotates through the ECE line of sight, electron temperature fluctuations at the NTM frequency are observed. The magnetic perturbation caused by the NTM is independently measured using Mirnov coils, and a correlation profile between these magnetic measurements and the ECE data is constructed. The phase difference between ECE oscillations on opposite sides of the island manifests as a zero-crossing of the correlation profile, which determines the NTM location in ECE channel space. To determine the location of ECRH power deposition, the power from a given gyrotron may be modulated at a particular frequency. Correlation analysis of this modulated signal and the ECE data identifies a particular ECE channel associated with the deposition of that gyrotron. Real time equilibrium reconstruction allows the ECE channels to be translated into flux surface and spatial coordinates for use in the feedback loop.

  19. Upgrade of the CMS hardron calorimeter for an upgraded LHC

    OpenAIRE

    Anderson, Jake

    2012-01-01

    The CMS barrel and endcap hadron calorimeters (Hcal) upgrading the current photo-sensors are hybrid photodiodes (HPDs) to meet the demands of the upgraded luminosity of the LHC. A key aspect of the Hcal upgrade is to add longitudinal segmentation to improve background rejection, energy resolution, and electron isolation at L1 trigger. The increased segmentation can be achieved by replacing the HPD's with multi-pixel Geiger-mode avalanche photodiodes. The upgraded electron...

  20. Conceptual design of a neutron camera for MAST Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Weiszflog, M., E-mail: matthias.weiszflog@physics.uu.se; Sangaroon, S.; Cecconello, M.; Conroy, S.; Ericsson, G.; Klimek, I. [Department of Physics and Astronomy, Uppsala University, EURATOM-VR Association, Uppsala (Sweden); Keeling, D.; Martin, R. [CCFE, Culham Science Centre, Abingdon (United Kingdom); Turnyanskiy, M. [ITER Physics Department, EFDA CSU Garching, Boltzmannstrae 2, D-85748 Garching (Germany)

    2014-11-15

    This paper presents two different conceptual designs of neutron cameras for Mega Ampere Spherical Tokamak (MAST) Upgrade. The first one consists of two horizontal cameras, one equatorial and one vertically down-shifted by 65 cm. The second design, viewing the plasma in a poloidal section, also consists of two cameras, one radial and the other one with a diagonal view. Design parameters for the different cameras were selected on the basis of neutron transport calculations and on a set of target measurement requirements taking into account the predicted neutron emissivities in the different MAST Upgrade operating scenarios. Based on a comparison of the cameras’ profile resolving power, the horizontal cameras are suggested as the best option.

  1. LIDAR Thomson scattering for advanced tokamaks. Final report

    International Nuclear Information System (INIS)

    Molvik, A.W.; Lerche, R.A.; Nilson, D.G.

    1996-01-01

    The LIDAR Thomson Scattering for Advanced Tokamaks project made a valuable contribution by combining LLNL expertise from the MFE Program: tokamak design and diagnostics, and the ICF Program and Physics Dept.: short-pulse lasers and fast streak cameras. This multidisciplinary group evaluated issues involved in achieving a factor of 20 higher high spatial resolution (to as small as 2-3 mm) from the present state of the art in LIDAR Thomson scattering, and developed conceptual designs to apply LIDAR Thomson scattering to three tokamaks: Upgraded divertor measurements in the existing DIII-D tokamak; Both core and divertor LIDAR Thomson scattering in the proposed (now cancelled) TPX; and core, edge, and divertor LIDAR Thomson scattering on the presently planned International Tokamak Experimental Reactor, ITER. Other issues were evaluated in addition to the time response required for a few millimeter spatial resolution. These include the optimum wavelength, 100 Hz operation of the laser and detectors, minimizing stray light - always the Achilles heel of Thomson scattering, and time dispersion in optics that could prevent good spatial resolution. Innovative features of our work included: custom short pulsed laser concepts to meet specific requirements, use of a prism spectrometer to maintain a constant optical path length for high temporal and spatial resolution, the concept of a laser focus outside the plasma to ionize gas and form an external fiducial to use in locating the plasma edge as well as to spread the laser energy over a large enough area of the inner wall to avoid laser ablation of wall material, an improved concept for cleaning windows between shots by means of laser ablation, and the identification of a new physics issue - nonlinear effects near a laser focus which could perturb the plasma density and temperature that are to be measured

  2. Correlation electron cyclotron emission diagnostic and improved calculation of turbulent temperature fluctuation levels on ASDEX Upgrade

    Science.gov (United States)

    Creely, A. J.; Freethy, S. J.; Burke, W. M.; Conway, G. D.; Leccacorvi, R.; Parkin, W. C.; Terry, D. R.; White, A. E.

    2018-05-01

    A newly upgraded correlation electron cyclotron emission (CECE) diagnostic has been installed on the ASDEX Upgrade tokamak and has begun to perform experimental measurements of electron temperature fluctuations. CECE diagnostics measure small amplitude electron temperature fluctuations by correlating closely spaced heterodyne radiometer channels. This upgrade expanded the system from six channels to thirty, allowing simultaneous measurement of fluctuation level radial profiles without repeat discharges, as well as opening up the possibility of measuring radial turbulent correlation lengths. Newly refined statistical techniques have been developed in order to accurately analyze the fluctuation data collected from the CECE system. This paper presents the hardware upgrades for this system and the analysis techniques used to interpret the raw data, as well as measurements of fluctuation spectra and fluctuation level radial profiles.

  3. OMEGA Upgrade preliminary design

    International Nuclear Information System (INIS)

    Craxton, R.S.

    1989-10-01

    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

  4. Treat upgrade fuel fabrication

    International Nuclear Information System (INIS)

    Davidson, K.V.; Schell, D.H.

    1979-01-01

    An extrusion and thermal treatment process was developed to produce graphite fuel rods containing a dispersion of enriched UO 2 . These rods will be used in an upgraded version of the Transient Reactor Test Facility (TREAT). The improved fuel provides a higher graphite matrix density, better fuel dispersion and higher thermal capabilities than the existing fuel

  5. ATLAS Pixel Detector Upgrade

    CERN Document Server

    Flick, T; The ATLAS collaboration

    2009-01-01

    The first upgrade for higher luminosity at LHC for the ATLAS pixel detector is the insertion of a forth layer, the IBL. The talk gives an overview about what the IBL is and how it will be set up, as well as to give a status of the research and develoment work.

  6. 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…

  7. 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

  8. Status of the tokamak program

    Science.gov (United States)

    Sheffield, J.

    1981-08-01

    For a specific configuration of magnetic field and plasma to be economically attractive as a commercial source of energy, it must contain a high-pressure plasma in a stable fashion while thermally isolating the plasma from the walls of the containment vessel. The tokamak magnetic configuration is presently the most successful in terms of reaching the considered goals. Tokamaks were developed in the USSR in a program initiated in the mid-1950s. By the early 1970s tokamaks were operating not only in the USSR but also in the U.S., Australia, Europe, and Japan. The advanced state of the tokamak program is indicated by the fact that it is used as a testbed for generic fusion development - for auxiliary heating, diagnostics, materials - as well as for specific tokamak advancement. This has occurred because it is the most economic source of a large, reproducible, hot, dense plasma. The basic tokamak is considered along with tokamak improvements, impurity control, additional heating, particle and power balance in a tokamak, aspects of microscopic transport, and macroscopic stability.

  9. Magnetic confinement experiment -- 1: Tokamaks

    International Nuclear Information System (INIS)

    Goldston, R.J.

    1994-01-01

    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

  10. The density limit in Tokamaks

    International Nuclear Information System (INIS)

    Alladio, F.

    1985-01-01

    A short summary of the present status of experimental observations, theoretical ideas and understanding of the density limit in tokamaks is presented. It is the result of the discussion that was held on this topic at the 4th European Tokamak Workshop in Copenhagen (December 4th to 6th, 1985). 610 refs

  11. SRS control system upgrade requirements

    International Nuclear Information System (INIS)

    Hill, L.F.

    1998-01-01

    This document defines requirements for an upgrade of the Sodium Removal System (SRS) control system. The upgrade is being performed to solve a number of maintainability and operability issues. The upgraded system will provide the same functions, controls and interlocks as the present system, and in addition provide enhanced functionality in areas discussed in this document

  12. Performing instrumentation and controls upgrades

    International Nuclear Information System (INIS)

    Kessler, F. M.; Connell, T. J.; Ryan, M. P.

    1992-01-01

    I and C upgrades are comprised of a varying range of content, complexity, expansiveness, and criticality. There are common threads in all upgrades which can be simplified by the development of a long term I and C upgrade plan. The development of a such a plan can establish effective ground rules for upgrades, large and small. It can be the basis from which to begin an upgrade evaluation and the standard which is used to compare the degree of compliance of any upgrade regarding the plan or to define the differences from the plan and an individual upgrade. Primary motivation for I and C upgrades are obsolescence and unavailability of spare parts. Numerous other areas of consideration are also involved in an upgrade. Today's technology results in most upgrades largely or totally utilizing digital equipment. The use of digital equipment is fairly new in many I and C applications and requires an elaborate evaluation from functional, qualification, operational, and licensing perspectives as well as others. A well defined upgrade plan developed as a basis for I and C upgrades is a significant start to ensuring an effective upgrade process. Properly developed and implemented, the plan will support I and C upgrade efforts to ensure that the intricacies associated with such tasks eliminate the existing problems which require the upgrade to be performed. The upgrade plan also results in ensuring the maximum benefit from all perspectives of the plant enhancements being carried out and considered for future implementation. Instrumentation and controls aging and replacement are issues of growing importance due to the potential for significant impact on plant operation and efficiency. Obsolescence and unavailability of spare parts are major drivers towards evaluating the cost benefits of upgrading current equipment. In addition to these two primary factors, the advantages of utilizing digital equipment have also become of prime importance when evaluating instrumentation and

  13. Operational present status and reliability analysis of the upgraded EAST cryogenic system

    Science.gov (United States)

    Zhou, Z. W.; Y Zhang, Q.; Lu, X. F.; Hu, L. B.; Zhu, P.

    2017-12-01

    Since the first commissioning in 2005, the cryogenic system for EAST (Experimental Advanced Superconducting Tokamak) has been cooled down and warmed up for thirteen experimental campaigns. In order to promote the refrigeration efficiencies and reliability, the EAST cryogenic system was upgraded gradually with new helium screw compressors and new dynamic gas bearing helium turbine expanders with eddy current brake to improve the original poor mechanical and operational performance from 2012 to 2015. Then the totally upgraded cryogenic system was put into operation in the eleventh cool-down experiment, and has been operated for the latest several experimental campaigns. The upgraded system has successfully coped with various normal operational modes during cool-down and 4.5 K steady-state operation under pulsed heat load from the tokamak as well as the abnormal fault modes including turbines protection stop. In this paper, the upgraded EAST cryogenic system including its functional analysis and new cryogenic control networks will be presented in detail. Also, its operational present status in the latest cool-down experiments will be presented and the system reliability will be analyzed, which shows a high reliability and low fault rate after upgrade. In the end, some future necessary work to meet the higher reliability requirement for future uninterrupted long-term experimental operation will also be proposed.

  14. The tokamak as a neutron source

    International Nuclear Information System (INIS)

    Hendel, H.W.; Jassby, D.L.

    1989-11-01

    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

  15. X-ray crystal spectrometer upgrade for ITER-like wall experiments at JET

    Energy Technology Data Exchange (ETDEWEB)

    Shumack, A. E., E-mail: amy.shumack@ccfe.ac.uk [JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); FOM Institute DIFFER, P.O. Box 1207, NL-3430 BE Nieuwegein (Netherlands); Rzadkiewicz, J. [National Centre for Nuclear Research, Andrzeja Sołtana 7, 05-400 Otwock (Poland); Chernyshova, M.; Czarski, T.; Karpinski, L. [Institute of Plasma Physics and Laser Microfusion, Hery 23, 01-497 Warsaw (Poland); Jakubowska, K. [Institute of Plasma Physics and Laser Microfusion, Hery 23, 01-497 Warsaw (Poland); Université Bordeaux, CNRS, CEA, CELIA, UMR 5107, F-33405 Talence (France); Scholz, M. [Institute of Nuclear Physics PAN, ul. Radzikowskiego 152, 31-342 Kraków (Poland); Byszuk, A.; Cieszewski, R.; Kasprowicz, G.; Pozniak, K.; Wojenski, A.; Zabolotny, W. [Institute of Electronic Systems, Warsaw University of Technology, 00-665 Warsaw (Poland); Dominik, W. [Faculty of Physics, Institute of Experimental Physics, Warsaw University, 00-681 Warsaw (Poland); Conway, N. J.; Dalley, S.; Tyrrell, S.; Zastrow, K.-D. [JET-EFDA, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Figueiredo, J. [EFDA-CSU, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Associação EURATOM/IST, Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Av Rovisco Pais, 1049-001 Lisbon (Portugal); and others

    2014-11-15

    The high resolution X-Ray crystal spectrometer at the JET tokamak has been upgraded with the main goal of measuring the tungsten impurity concentration. This is important for understanding impurity accumulation in the plasma after installation of the JET ITER-like wall (main chamber: Be, divertor: W). This contribution provides details of the upgraded spectrometer with a focus on the aspects important for spectral analysis and plasma parameter calculation. In particular, we describe the determination of the spectrometer sensitivity: important for impurity concentration determination.

  16. Tokamak instrumentation and controls

    International Nuclear Information System (INIS)

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

    1979-02-01

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

  17. Demonstration tokamak power plant

    International Nuclear Information System (INIS)

    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

  18. Maximum entropy tokamak configurations

    International Nuclear Information System (INIS)

    Minardi, E.

    1989-01-01

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

  19. Texas Experimental Tokamak: A plasma research facility. Technical progress report, November 1, 1993--October 31, 1994

    Energy Technology Data Exchange (ETDEWEB)

    Wootton, A.J.

    1994-07-01

    The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics in order to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks and in particular to understand the role of turbulence. So that they can continue to study the physics that is most relevant to the fusion program, TEXT completed a significant device upgrade this year. The new capabilities of the device and new and innovative diagnostics were exploited in all main program areas including: (1) configuration studies; (2) electron cyclotron heating physics; (3) improved confinement modes; (4) edge physics/impurity studies; (5) central turbulence and transport; and (6) transient transport. Details of the progress in each of the research areas are described.

  20. Texas Experimental Tokamak: A plasma research facility. Technical progress report, November 1, 1993--October 31, 1994

    International Nuclear Information System (INIS)

    Wootton, A.J.

    1994-07-01

    The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics in order to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks and in particular to understand the role of turbulence. So that they can continue to study the physics that is most relevant to the fusion program, TEXT completed a significant device upgrade this year. The new capabilities of the device and new and innovative diagnostics were exploited in all main program areas including: (1) configuration studies; (2) electron cyclotron heating physics; (3) improved confinement modes; (4) edge physics/impurity studies; (5) central turbulence and transport; and (6) transient transport. Details of the progress in each of the research areas are described

  1. The radiation asymmetry in MGI rapid shutdown on J-TEXT tokamak

    Science.gov (United States)

    Tong, Ruihai; Chen, Zhongyong; Huang, Duwei; Cheng, Zhifeng; Zhang, Xiaolong; Zhuang, Ge; J-TEXT Team

    2017-10-01

    Disruptions, the sudden termination of tokamak fusion plasmas by instabilities, have the potential to cause severe material wall damage to large tokamaks like ITER. The mitigation of disruption damage is an essential part of any fusion reactor system. Massive gas injection (MGI) rapid shutdown is a technique in which large amounts of noble gas are injected into the plasma in order to safely radiate the plasma energy evenly over the entire plasma-facing first wall. However, the radiated energy during the thermal quench (TQ) in massive gas injection (MGI) induced disruptions is found toroidal asymmetric, and the degrees of asymmetry correlate with the gas penetration and MGI induced magnetohydrodynamics (MHD) activities. A toroidal and poloidal array of ultraviolet photodiodes (AXUV) has been developed to investigate the radiation asymmetry on J-TEXT tokamak. Together with the upgraded mirnov probe arrays, the relation between MGI triggered MHD activities with radiation asymmetry is studied.

  2. Contribution of ASDEX Upgrade to disruption studies for ITER

    International Nuclear Information System (INIS)

    Pautasso, G.; Reiter, B.; Giannone, L.; Gruber, O.; Herrmann, A.; Kardaun, O.; Maraschek, M.; Mlynek, A.; Schneider, W.; Zhang, Y.; Khayrutdinov, K.K.; Lukash, V.E.; Nakamura, Y.; Sias, G.; Sugihara, M.

    2011-01-01

    This paper describes the most recent contributions of ASDEX Upgrade to ITER in the field of disruption studies. (1) The ITER specifications for the halo current magnitude are based on data collected from several tokamaks and summarized in the plot of the toroidal peaking factor versus the maximum halo current fraction. Even if the maximum halo current in ASDEX Upgrade reaches 50% of the plasma current, the duration of this maximum lasts a fraction of a ms. (2) Long-lasting asymmetries of the halo current are rare and do not give rise to a large asymmetric component of the mechanical forces on the machine. Differently from JET, these asymmetries are neither locked nor exhibit a stationary harmonic structure. (3) Recent work on disruption prediction has concentrated on the search for a simple function of the most relevant plasma parameters, which is able to discriminate between the safe and pre-disruption phases of a discharge. For this purpose, the disruptions of the last four years have been classified into groups and then discriminant analysis is used to select the most significant variables and to derive the discriminant function. (4) The attainment of the critical density for the collisional suppression of the runaway electrons seems to be technically and physically possible on our medium size tokamak. The CO 2 interferometer and the AXUV diagnostic provide information on the highly 3D impurity transport process during the whole plasma quench.

  3. CMS pixel upgrade project

    CERN Document Server

    Kaestli, Hans-Christian

    2010-01-01

    The LHC machine at CERN finished its first year of pp collisions at a center of mass energy of 7~TeV. While the commissioning to exploit its full potential is still ongoing, there are plans to upgrade its components to reach instantaneous luminosities beyond the initial design value after 2016. A corresponding upgrade of the innermost part of the CMS detector, the pixel detector, is needed. A full replacement of the pixel detector is planned in 2016. It will not only address limitations of the present system at higher data rates, but will aggressively lower the amount of material inside the fiducial tracking volume which will lead to better tracking and b-tagging performance. This article gives an overview of the project and illuminates the motivations and expected improvements in the detector performance.

  4. CMS pixel upgrade project

    CERN Document Server

    INSPIRE-00575876

    2011-01-01

    The LHC machine at CERN finished its first year of pp collisions at a center of mass energy of 7 TeV. While the commissioning to exploit its full potential is still ongoing, there are plans to upgrade its components to reach instantaneous luminosities beyond the initial design value after 2016. A corresponding upgrade of the innermost part of the CMS detector, the pixel detector, is needed. A full replacement of the pixel detector is planned in 2016. It will not only address limitations of the present system at higher data rates, but will aggressively lower the amount of material inside the fiducial tracking volume which will lead to better tracking and b-tagging performance. This article gives an overview of the project and illuminates the motivations and expected improvements in the detector performance.

  5. Observations on the W-transport in the core plasma of JET and ASDEX Upgrade

    Czech Academy of Sciences Publication Activity Database

    Pütterich, T.; Dux, R.; Neu, R.; Bernert, M.; Beurskens, M.N.A.; Bobkov, V.; Brezinsek, S.; Challis, C.; Coenen, J.W.; Coffey, I.; Czarnecka, A.; Giroud, C.; Jacquet, P.; Joffrin, E.; Kallenbach, A.; Lehnen, M.; Lerche, E.; De La Luna, E.; Marsen, S.; Matthews, G.; Mayoral, M.-L.; McDermott, R.M.; Meigs, A.; Mlynář, Jan; Sertoli, M.; van Rooij, G.

    2013-01-01

    Roč. 55, č. 12 (2013), s. 124036-124036 ISSN 0741-3335. [European Physical Society Conference on Plasma Physics/40./. Espoo, 01.07.2013-05.07.2013] Institutional support: RVO:61389021 Keywords : tokamak * impurity transport * core plasma * fusion * tungsten * ASDEX Upgrade Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.386, year: 2013 http://iopscience.iop.org/0741-3335/55/12/124036/pdf/0741-3335_55_12_124036.pdf

  6. The ASDEX upgrade toroidal field magnet and poloidal divertor field coil system adapted to reactor requirements

    International Nuclear Information System (INIS)

    Koeppendoerfer, W.; Blaumoser, M.; Ennen, K.; Gruber, J.; Gruber, O.; Jandl, O.; Kaufmann, M.; Kollotzek, H.; Kotzlowski, H.; Lackner, E.; Lackner, K.; Larcher, T. von; Noterdaeme, J.M.; Pillsticker, M.; Poehlchen, R.; Preis, H.; Schneider, H.; Seidel, U.; Sombach, B.; Speth, E.; Streibl, B.; Vernickel, H.; Werner, F.; Wesner, F.; Wieczorek, A.

    1986-01-01

    ASDEX Upgrade is a tokamak experiment with external poloidal field coils that is now under construction at IPP Garching. It can produce elongated single-null (SN), double-null (DN) , and limiter (L) configurations. The SN is the reference configuration with asymmetric load distributions in the poloidal field (PF) system and the toroidal field (TF) magnet. Plasma control and stabilization require a rigid passive conductor close to the plasma. The design principles of the coils and support structure are described. (orig.)

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

    International Nuclear Information System (INIS)

    Meisl, Gerd Korbinian

    2015-01-01

    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.

  8. Topology of tokamak orbits

    International Nuclear Information System (INIS)

    Rome, J.A.; Peng, Y.K.M.

    1978-09-01

    Guiding center orbits in noncircular axisymmetric tokamak plasmas are studied in the constants of motion (COM) space of (v, zeta, psi/sub m/). Here, v is the particle speed, zeta is the pitch angle with respect to the parallel equilibrium current, J/sub parallels/, and psi/sub m/ is the maximum value of the poloidal flux function (increasing from the magnetic axis) along the guiding center orbit. Two D-shaped equilibria in a flux-conserving tokamak having β's of 1.3% and 7.7% are used as examples. In this space, each confined orbit corresponds to one and only one point and different types of orbits (e.g., circulating, trapped, stagnation and pinch orbits) are represented by separate regions or surfaces in the space. It is also shown that the existence of an absolute minimum B in the higher β (7.7%) equilibrium results in a dramatically different orbit topology from that of the lower β case. The differences indicate the confinement of additional high energy (v → c, within the guiding center approximation) trapped, co- and countercirculating particles whose orbit psi/sub m/ falls within the absolute B well

  9. ITER tokamak device

    International Nuclear Information System (INIS)

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

    1991-01-01

    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; and 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, fuelling, 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., (i) magnet systems (toroidal and poloidal field coils and cryogenic systems), (ii) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (iii) first wall, (iv) divertor plate (design and materials, performance and lifetime, a.o.), (v) blanket/shield system, (vi) maintenance equipment, (vii) current drive and heating, (viii) fuel cycle system, and (ix) diagnostics. 11 refs, figs and tabs

  10. Dust Measurements in Tokamaks

    International Nuclear Information System (INIS)

    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

    2008-01-01

    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

  11. Axisymmetric control in tokamaks

    International Nuclear Information System (INIS)

    Humphreys, D.A.

    1991-02-01

    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 κ sep = 1.85 can be stabilized robustly with the present control architecture and conductor/sensor configuration

  12. The CDF upgrade

    International Nuclear Information System (INIS)

    Newman-Holmes, C.

    1995-01-01

    The Collider Detector at Fermilab (CDF) has been used to study proton-antiproton collisions at the Fermilab Tevatron since 1985. Over the years, the detector has evolved steadily to increase its physics capability and to keep pace with improvements to the Tevatron. Fermilab is currently building a new Main Injector accelerator which will lead to even larger luminosity values. This paper describes upgrades to CDF that will allow one to exploit the higher luminosity of the Main Injector

  13. LHCb VELO upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Hennessy, Karol

    2017-02-11

    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 2×10{sup 33} cm{sup −2} s{sup −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 55×55 μm{sup 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, separated from the beam vacuum by a thin custom made foil. The foil will be manufactured through milling and possibly thinned further by chemical etching. The material budget will be minimised by the use of evaporative CO{sub 2} coolant circulating in microchannels within 400 μm thick silicon substrates. The current status of the VELO upgrade is described and latest results from operation of irradiated sensor assemblies are presented.

  14. Upgrade to the control system of the reflectometry diagnostic of ASDEX upgrade

    International Nuclear Information System (INIS)

    Graca, S.; Santos, J.; Manso, M.E.

    2004-01-01

    The broadband frequency modulation-continuous wave microwave/millimeter wave reflectometer of ASDEX upgrade tokamak (Institut fuer Plasma Physik (IPP), Garching, Germany) developed by Centro de Fusao Nuclear (Lisboa, Portugal) with the collaboration of IPP, is a complex system with 13 channels (O and X modes) and two types of operation modes (swept and fixed frequency). The control system that ensures remote operation of the diagnostic incorporates VME and CAMAC bus based acquisition/timing systems. Microprocessor input/output boards are used to control and monitor the microwave circuitry and associated electronic devices. The implementation of the control system is based on an object-oriented client/server model: a centralized server manages the hardware and receives input from remote clients. Communication is handled through transmission control protocol/internet protocol sockets. Here we describe recent upgrades of the control system aiming to: (i) accommodate new channels; (ii) adapt to the heterogeneity of computing platforms and operating systems; and (iii) overcome remote access restrictions. Platform and operating system independence was achieved by redesigning the graphical user interface in JAVA. As secure shell is the standard remote access protocol adopted in major fusion laboratories, secure shell tunneling was implemented to allow remote operation of the diagnostic through the existing firewalls

  15. Upgrade to the control system of the reflectometry diagnostic of ASDEX upgrade

    Science.gov (United States)

    Graça, S.; Santos, J.; Manso, M. E.

    2004-10-01

    The broadband frequency modulation-continuous wave microwave/millimeter wave reflectometer of ASDEX upgrade tokamak (Institut für Plasma Physik (IPP), Garching, Germany) developed by Centro de Fusão Nuclear (Lisboa, Portugal) with the collaboration of IPP, is a complex system with 13 channels (O and X modes) and two types of operation modes (swept and fixed frequency). The control system that ensures remote operation of the diagnostic incorporates VME and CAMAC bus based acquisition/timing systems. Microprocessor input/output boards are used to control and monitor the microwave circuitry and associated electronic devices. The implementation of the control system is based on an object-oriented client/server model: a centralized server manages the hardware and receives input from remote clients. Communication is handled through transmission control protocol/internet protocol sockets. Here we describe recent upgrades of the control system aiming to: (i) accommodate new channels; (ii) adapt to the heterogeneity of computing platforms and operating systems; and (iii) overcome remote access restrictions. Platform and operating system independence was achieved by redesigning the graphical user interface in JAVA. As secure shell is the standard remote access protocol adopted in major fusion laboratories, secure shell tunneling was implemented to allow remote operation of the diagnostic through the existing firewalls.

  16. The LHCb VELO upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Dosil Suárez, Álvaro, E-mail: alvaro.dosil@usc.es

    2016-07-11

    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 2×10{sup 33} cm{sup −2} s{sup −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 μm{sup 2} 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 CO{sub 2} coolant circulating in microchannels within 400 μm thick silicon substrates.

  17. The D0 upgrade

    International Nuclear Information System (INIS)

    Tuts, P.M.

    1992-10-01

    The original D0 detector was proposed in 1983, with a focus on high P T physics using precision measurements of e's, μ's, jets, and missing E T . This detector, as of the summer of 1992, has started data taking at the Fermilab Collider. However, by 1995/6 the luminosity will reach 10 31 cm -2 sec -1 , and the minimum bunch spacing will drop to 396ns from the present 3.5μs (by the Main Injector era, luminosities will approach 10 32 cm -2 sec -1 and minimum bunch spacings may reach 132ns). These changes in the accelerator conditions force us to upgrade or replace a number of detector subsystems in order to meet these new demands. In addition, the upgrade offers us the opportunity to expand the physics horizons to include not only the all important high P T physics menu, but also the low P T physics that has become increasingly important. In this paper we describe the D0 detector upgrade

  18. ATLAS Upgrade Plans

    CERN Document Server

    Hopkins, W; The ATLAS collaboration

    2014-01-01

    After the successful LHC operation at the center-of-mass energies of 7 and 8 TeV in 2010-2012, 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 final goal is to extend the dataset from about few hundred fb−1 expected for LHC running to 3000/fb 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. Current planning in ATLAS envisions significant upgrades to the detector during the consolidation of the LHC to reach full LHC energy and further upgrades. 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...

  19. The LHCb VELO upgrade

    International Nuclear Information System (INIS)

    Rodríguez Pérez, Pablo

    2013-01-01

    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 33 cm −2 s −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 15 1 MeV n eq /cm 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 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

  20. The LHCb VELO Upgrade

    CERN Document Server

    de Capua, Stefano

    2016-01-01

    The upgrade of the LHCb experiment, scheduled for LHC Run-3, will transform the experiment to a triggerless 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 2×1033 cm−2 s −1 . The Vertex Locator (VELO) is the silicon vertex detector surrounding the interaction region. The current strip detector will be replaced with a hybrid pixel system equipped with electronics capable of reading out at 40 MHz. The upgraded VELO will allow for fast pattern recognition and track reconstruction in the software trigger. The silicon pixel sensors have 55×55 µm2 pitch, and are read out by the VeloPix ASIC. The VeloPix builds on the currently available Timepix3, modified to deliver a radiation hard design capable of an order of magnitude increase in output rate. The hottest regions will have pixel hit rates of 900 Mhits/s, yielding a total data rate more than 3 Tbit/s for the upgraded VELO...

  1. 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...

  2. Vacuum system and cleaning techniques in the FTU machines

    International Nuclear Information System (INIS)

    Alessandrini, C.; Apicella, M.L.; Ferro, C.

    1988-01-01

    FTU (Frascati Tokamak Upgrade) is a high magnetic field (8T) tokamak under construction at the Frascati Energy Research Center (ENEA). Its vacuum systems has been already manifactured and is presently being assembled. It consist of an all metallic fully welded vessel, pumped by six turbomolecular pumps. The vacuum system has been dimensioned to allow a base pressure lower than 2.6 x 10 -6 Pa. The paper reports the design philosophy of the vacuum system. The results of the cleaning techniques performed on a 1:1 scale toroidal sector of FTU are also presented and discussed

  3. Modelling dust transport in tokamaks

    International Nuclear Information System (INIS)

    Martin, J.D.; Martin, J.D.; Bacharis, M.; Coppins, M.; Counsell, G.F.; Allen, J.E.; Counsell, G.F.

    2008-01-01

    The DTOKS code, which models dust transport through tokamak plasmas, is described. The floating potential and charge of a dust grain in a plasma and the fluxes of energy to and from it are calculated. From this model, the temperature of the dust grain can be estimated. A plasma background is supplied by a standard tokamak edge modelling code (B2SOLPS5.0), and dust transport through MAST (the Mega-Amp Spherical Tokamak) and ITER plasmas is presented. We conclude that micron-radius tungsten dust can reach the separatrix in ITER. (authors)

  4. Tokamak building-design considerations for a large tokamak device

    International Nuclear Information System (INIS)

    Barrett, R.J.; Thomson, S.L.

    1981-01-01

    Design and construction of a satisfactory tokamak building to support FED appears feasible. Further, a pressure vessel building does not appear necessary to meet the plant safety requirements. Some of the building functions will require safety class systems to assure reliable and safe operation. A rectangular tokamak building has been selected for FED preconceptual design which will be part of the confinement system relying on ventilation and other design features to reduce the consequences and probability of radioactivity release

  5. Research using small tokamaks

    International Nuclear Information System (INIS)

    1991-05-01

    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

  6. The DIII-D cryogenic system upgrade

    International Nuclear Information System (INIS)

    Schaubel, K.M.; Laughon, G.J.; Campbell, G.L.; Langhorn, A.R.; Stevens, N.C.; Tupper, M.L.

    1993-10-01

    The original DIII-D cryogenic system was commissioned in 1981 and was used to cool the cryopanel arrays for three hydrogen neutral beam injectors. Since then, new demands for liquid helium have arisen including: a fourth neutral beam injector, ten superconducting magnets for the electron cyclotron heating gyrotrons, and more recently, the advanced diverter cryopump which resides inside the tokamak vacuum vessel. The original cryosystem could not meet these demands. Consequently, the cryosystem was upgraded in several phases to increase capacity, improve reliability, and reduce maintenance. The majority of the original system has been replaced with superior equipment. The capacity now exists to support present as well as future demands for liquid helium at DIII-D including a hydrogen pellet injector, which is being constructed by Oak Ridge National Laboratory. Upgrades to the cryosystem include: a recently commissioned 150 ell/hr helium liquefier, two 55 g/sec helium screw compressors, a fully automated 20-valve cryogen distribution box, a high efficiency helium wet expander, and the conversion of equipment from manual or pneumatic to programmable logic controller (PLC) control. The distribution box was designed and constructed for compactness due to limited space availability. Overall system efficiency was significantly improved by replacing the existing neutral beam reliquefier Joule-Thomson valve with a reciprocating wet expander. The implementation of a PLC-based automatic control system has resulted in increased efficiency and reliability. This paper will describe the cryosystem design with emphasis on newly added equipment. In addition, performance and operational experience will be discussed

  7. A Neutral Beam Injector Upgrade for NSTX

    International Nuclear Information System (INIS)

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

    2002-01-01

    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

  8. The DIII-D cryogenic system upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Schaubel, K.M.; Laughon, G.J.; Campbell, G.L.; Langhorn, A.R.; Stevens, N.C.; Tupper, M.L.

    1993-10-01

    The original DIII-D cryogenic system was commissioned in 1981 and was used to cool the cryopanel arrays for three hydrogen neutral beam injectors. Since then, new demands for liquid helium have arisen including: a fourth neutral beam injector, ten superconducting magnets for the electron cyclotron heating gyrotrons, and more recently, the advanced diverter cryopump which resides inside the tokamak vacuum vessel. The original cryosystem could not meet these demands. Consequently, the cryosystem was upgraded in several phases to increase capacity, improve reliability, and reduce maintenance. The majority of the original system has been replaced with superior equipment. The capacity now exists to support present as well as future demands for liquid helium at DIII-D including a hydrogen pellet injector, which is being constructed by Oak Ridge National Laboratory. Upgrades to the cryosystem include: a recently commissioned 150 {ell}/hr helium liquefier, two 55 g/sec helium screw compressors, a fully automated 20-valve cryogen distribution box, a high efficiency helium wet expander, and the conversion of equipment from manual or pneumatic to programmable logic controller (PLC) control. The distribution box was designed and constructed for compactness due to limited space availability. Overall system efficiency was significantly improved by replacing the existing neutral beam reliquefier Joule-Thomson valve with a reciprocating wet expander. The implementation of a PLC-based automatic control system has resulted in increased efficiency and reliability. This paper will describe the cryosystem design with emphasis on newly added equipment. In addition, performance and operational experience will be discussed.

  9. Progress in diagnostics of the COMPASS tokamak

    Science.gov (United States)

    Weinzettl, V.; Adamek, J.; Berta, M.; Bilkova, P.; Bogar, O.; Bohm, P.; Cavalier, J.; Dejarnac, R.; Dimitrova, M.; Ficker, O.; Fridrich, D.; Grover, O.; Hacek, P.; Havlicek, J.; Havranek, A.; Horacek, J.; Hron, M.; Imrisek, M.; Komm, M.; Kovarik, K.; Krbec, J.; Markovic, T.; Matveeva, E.; Mitosinkova, K.; Mlynar, J.; Naydenkova, D.; Panek, R.; Paprok, R.; Peterka, M.; Podolnik, A.; Seidl, J.; Sos, M.; Stockel, J.; Tomes, M.; Varavin, M.; Varju, J.; Vlainic, M.; Vondracek, P.; Zajac, J.; Zacek, F.; Stano, M.; Anda, G.; Dunai, D.; Krizsanoczi, T.; Refy, D.; Zoletnik, S.; Silva, A.; Gomes, R.; Pereira, T.; Popov, Tsv.; Sarychev, D.; Ermak, G. P.; Zebrowski, J.; Jakubowski, M.; Rabinski, M.; Malinowski, K.; Nanobashvili, S.; Spolaore, M.; Vianello, N.; Gauthier, E.; Gunn, J. P.; Devitre, A.

    2017-12-01

    The COMPASS tokamak at IPP Prague is a small-size device with an ITER-relevant plasma geometry and operating in both the Ohmic as well as neutral beam assisted H-modes since 2012. A basic set of diagnostics installed at the beginning of the COMPASS operation has been gradually broadened in type of diagnostics, extended in number of detectors and collected channels and improved by an increased data acquisition speed. In recent years, a significant progress in diagnostic development has been motivated by the improved COMPASS plasma performance and broadening of its scientific programme (L-H transition and pedestal scaling studies, magnetic perturbations, runaway electron control and mitigation, plasma-surface interaction and corresponding heat fluxes, Alfvenic and edge localized mode observations, disruptions, etc.). In this contribution, we describe major upgrades of a broad spectrum of the COMPASS diagnostics and discuss their potential for physical studies. In particular, scrape-off layer plasma diagnostics will be represented by a new concept for microsecond electron temperature and heat flux measurements - we introduce a new set of divertor Langmuir and ball-pen probe arrays, newly constructed probe heads for reciprocating manipulators as well as several types of standalone probes. Among optical tools, an upgraded high-resolution edge Thomson scattering diagnostic for pedestal studies and a set of new visible light and infrared (plasma-surface interaction investigations) cameras will be described. Particle and beam diagnostics will be covered by a neutral particle analyzer, diagnostics on a lithium beam, Cherenkov detectors (for a direct detection of runaway electrons) and neutron detectors. We also present new modifications of the microwave reflectometer for fast edge density profile measurements.

  10. Natural current profiles in tokamaks

    International Nuclear Information System (INIS)

    Biskamp, D.

    1986-01-01

    It is proposed that a certain class of equilibrium, which follow from an elementary variational principle, are the natural current profiles in tokamaks, to which actual discharge profiles tend to relax. (orig.)

  11. Alcator C-Mod Tokamak

    Data.gov (United States)

    Federal Laboratory Consortium — Alcator C-Mod at the Massachusetts Institute of Technology is operated as a DOE national user facility. Alcator C-Mod is a unique, compact tokamak facility that uses...

  12. Preliminary Design of Alborz Tokamak

    Science.gov (United States)

    Mardani, M.; Amrollahi, R.; Saramad, S.

    2012-04-01

    The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. The most important part of the tokamak design is the design of TF coils. In this paper a refined design of the TF coil system for the Alborz tokamak is presented. This design is based on cooper cable conductor with 5 cm width and 6 mm thickness. The TF coil system is consist of 16 rectangular shape coils, that makes the magnetic field of 0.7 T at the plasma center. The stored energy in total is 160 kJ, and the power supply used in this system is a capacitor bank with capacity of C = 1.32 mF and V max = 14 kV.

  13. New directions in tokamak reactors

    International Nuclear Information System (INIS)

    Baker, C.C.

    1985-01-01

    New directions for tokamak research are briefly mentioned. Some of the areas for new considerations are the following: reactor size, beta ratio, current drivers, blankets, impurity control, and modular designs

  14. The Tokamak IST-TOK

    International Nuclear Information System (INIS)

    Varandas, C.A.F.; Cabral, J.A.C.; Manso, M.E.

    1991-01-01

    A small tokamak is under construction at the Portuguese Technical Superior Institute. The main objective is to create a home based laboratory in which an independent scientific program might be developed. (L.C.J.A.). 14 refs, 6 figs

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

    International Nuclear Information System (INIS)

    STAMBAUGH, RD

    2003-01-01

    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

  16. Numerical Tokamak Project code comparison

    International Nuclear Information System (INIS)

    Waltz, R.E.; Cohen, B.I.; Beer, M.A.

    1994-01-01

    The Numerical Tokamak Project undertook a code comparison using a set of TFTR tokamak parameters. Local radial annulus codes of both gyrokinetic and gyrofluid types were compared for both slab and toroidal case limits assuming ion temperature gradient mode turbulence in a pure plasma with adiabatic electrons. The heat diffusivities were found to be in good internal agreement within ± 50% of the group average over five codes

  17. Spherical tokamak development in Brazil

    International Nuclear Information System (INIS)

    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.; Barbosa, L.F.W.; Patire Junior, H.; The high-power microwave sources group

    2003-01-01

    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)

  18. Confinement and diffusion in tokamaks

    International Nuclear Information System (INIS)

    McWilliams, R.

    1988-01-01

    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*cT e /eB(δn i /n i ) rms which is also derived by a simple theory, the cross-field diffusion time, tp=a 2 /D, is calculated and compared to experimental results from 51 tokamak for standard Ohmic operation

  19. Enhancement of confinement in tokamaks

    International Nuclear Information System (INIS)

    Furth, H.P.

    1986-01-01

    The analysis begins by identifying a hypothetical model of tokamak confinement that is designed to take into account the conflict between Tsub(e)(r)-profile shapes arising from microscopic transport and J(r)-profile shapes required for gross stability. On the basis of this model, a number of hypothetical lines of advance are developed. Some TFTR experiments that may point the way to a particularly attractive type of tokamak reactor regime are discussed. (author)

  20. 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)

  1. 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)

  2. Spherical tokamak development in Brazil

    International Nuclear Information System (INIS)

    Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes

    2003-01-01

    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)

  3. High Beta Tokamak research

    International Nuclear Information System (INIS)

    Navratil, G.A.; Mauel, M.E.; Ivers, T.H.; Sankar, M.K.V.; Eisner, E.; Gates, D.; Garofalo, A.; Kombargi, R.; Maurer, D.; Nadle, D.; Xiao, Q.

    1993-01-01

    During the past 6 months, experiments have been conducted with the HBT-EP tokamak in order to (1) test and evaluate diagnostic systems, (2) establish basic machine operation, (3) document MHD behavior as a function of global discharge parameters, (4) investigate conditions leading to passive stabilization of MHD instabilities, and (5) quantify the external saddle coil current required for DC mode locking. In addition, the development and installation of new hardware systems has occurred. A prototype saddle coil was installed and tested. A five-position (n,m) = (1,2) external helical saddle coil was attached for mode-locking experiments. And, fabrication of the 32-channel UV tomography and the multipass Thomson scattering diagnostics have begun in preparation for installation later this year

  4. Anomalous transport in tokamaks

    International Nuclear Information System (INIS)

    Wootton, A.J.

    1989-01-01

    A review is presented of what is known about anomalous transport in tokamaks. It is generally thought that this anomalous transport is the result of fluctuations in various plasma parameters. In the plasma edge detailed measurements of the quantities required to directly determine the fluctuation driven fluxes are available. The total flux of particles is well explained by the measured electrostatic fluctuation driven flux. However, a satisfactory model to explain the origin of the fluctuations has not been identified. The processes responsible for determining the edge energy flux are less clear, but electrostatic convection plays an important part. In the confinement region experimental observations are presently restricted to measurements of density and potential fluctuations and their correlations. The characteristics of the measured fluctuations are discussed and compared with the predictions of various models. Comparisons between measured particle, electron heat and ion heat fluxes, and those fluxes predicted to result from the measured fluctuations, are made. Magnetic fluctuations is discussed

  5. Tokamak hybrid study

    International Nuclear Information System (INIS)

    Tenney, F.H.

    1976-09-01

    A report on one year of study of a tokamak hybrid reactor is presented. The plasma is maintained by both D and T beams. To obtain long burn times a poloidal field divertor is required. Both the single null and the double null style of divertor are considered. The blanket consists of a neutron multiplier region containing natural uranium followed by burner regions of molten salt (flibe) loaded with PuF 3 to enhance the energy multiplication. Economic analysis has been applied only recently to a variety of reactor sizes and plasma conditions. Early indications suggest that the most attractive hybrids will have large plasmas of major radius in excess of 8 meters

  6. Tokamak hybrid study

    International Nuclear Information System (INIS)

    Tenney, F.H.

    1976-01-01

    A report on one year of study of a tokamak hybrid reactor is given. The plasma is maintained by both D and T beams. To obtain long burn times a poloidal field divertor is required. Both the single null and the double null style of divertor are considered. The blanket consists of a neutron multiplier region containing natural uranium followed by burner regions of molten salt (flibe) loaded with PuF 3 to enhance the energy multiplication. Economic analysis has been applied only recently to a variety of reactor sizes and plasma conditions. Early indications suggest that the most attractive hybrids will have large plasmas of major radius in excess of 8 meters

  7. The Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Schmidt, J.

    1987-01-01

    The author discusses his lab's plan for completing the Compact Ignition Tokamak (CIT) conceptual design during calendar year 1987. Around July 1 they froze the subsystem envelopes on the device to continue with the conceptual design. They did this by formalizing a general requirements document. They have been developing the management plan and submitted a version to the DOE July 10. He describes a group of management activities. They released the vacuum vessel Request For Proposals (RFP) on August 5. An RFP to do a major part of the system engineering on the device is being developed. They intend to assemble the device outside of the test cell, then move it into the the test cell, install it there, and bring to the test cell many of the auxiliary facilities from TFTR, for example, power supplies

  8. Plasma turbulence in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Caldas, Ibere L.; Heller, M.V.A.P.; Brasilio, Z.A. [Sao Paulo Univ., SP, RJ (Brazil). Inst. de Fisica

    1997-12-31

    Full text. In this work we summarize the results from experiments on electrostatic and magnetic fluctuations in tokamak plasmas. Spectral analyses show that these fluctuations are turbulent, having a broad spectrum of wavectors and a broad spectrum of frequencies at each wavector. The electrostatic turbulence induces unexpected anomalous particle transport that deteriorates the plasma confinement. The relationship of these fluctuations to the current state of plasma theory is still unclear. Furthermore, we describe also attempts to control this plasma turbulence with external magnetic perturbations that create chaotic magnetic configurations. Accordingly, the magnetic field lines may become chaotic and then induce a Lagrangian diffusion. Moreover, to discuss nonlinear coupling and intermittency, we present results obtained by using numerical techniques as bi spectral and wavelet analyses. (author)

  9. Disruptions in Tokamaks

    International Nuclear Information System (INIS)

    Bondeson, A.

    1987-01-01

    This paper discusses major and minor disruptions in Tokamaks. A number of models and numerical simulations of disruptions based on resistive MHD are reviewed. A discussion is given of how disruptive current profiles are correlated with the experimentally known operational limits in density and current. It is argued that the q a =2 limit is connected with stabilization of the m=2/n=1 tearing mode for a approx.< 2.7 by resistive walls and mode rotation. Experimental and theoretical observations indicate that major disruptions usually occur in at least two phases, first a 'predisruption', or loss of confinement in the region 1 < q < 2, leaving the q approx.= 1 region almost unaffected, followed by a final disruption of the central part, interpreted here as a toroidal n = 1 external kink mode. (author)

  10. Upgrading of TREAT experimental capabilities

    International Nuclear Information System (INIS)

    Dickerman, C.E.; Rose, D.; Bhattacharyya, S.K.

    1982-01-01

    The TREAT facility at the Argonne National Laboratory site in the Idaho National Engineering Laboratory is being upgraded to provide capabilities for fast-reactor-safety transient experiments not possible at any other experimental facility. Principal TREAT Upgrade (TU) goal is provision for 37-pin size experiments on energetics of core-disruptive accidents (CDA) in fast breeder reactor cores with moderate sodium void coefficients. this goal requires a significant enhancement of the capabilities of the TREAT facility, specifically including reactor control, hardened neutron spectrum incident on the test sample, and enlarged building. The upgraded facility will retain the capability for small-size experiments of the types currently being performed in TREAT. Reactor building and crane upgrading have been completed. TU schedules call for the components of the upgraded reactor system to be finished in 1984, including upgraded TREAT fuel and control system, and expanded coverage by the hodoscope fuel-motion diagnostics system

  11. The ARIES tokamak reactor study

    International Nuclear Information System (INIS)

    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 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

  12. FTU cooled liquid lithium upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Iafrati, M., E-mail: matteo.iafrati@enea.it [Associazione Euratom-ENEA sulla Fusione, C. R. Frascati, C. P. 65-00044 Frascati, Rome (Italy); Apicella, M.L.; Boncagni, L. [Associazione Euratom-ENEA sulla Fusione, C. R. Frascati, C. P. 65-00044 Frascati, Rome (Italy); Lyublinski, I. [JSC “RED STAR”, Moscow (Russian Federation); Mazzitelli, G. [Associazione Euratom-ENEA sulla Fusione, C. R. Frascati, C. P. 65-00044 Frascati, Rome (Italy); Vertkov, A. [JSC “RED STAR”, Moscow (Russian Federation)

    2017-04-15

    In the framework of the liquid lithium limiter experiment in Frascati a new auxiliary system was developed in order to provide a better control of the energy fluid vector. The cooled liquid lithium system (CLL) was installed for the first time at the end of 2013, it uses overheated water to heat the lithium and to extract, at the same time, the heat from the metal surface when it gets wet by the plasma. A first version of the system, developed and presented in previous papers, has been modified to optimize the heat flux measurement on the liquid lithium surface. The changes include a new power supply logic for the heating system, new sensors and new read-out electronics compatible with the implementation of a real time control system. The prototype was updated with the aim of achieving a low cost and versatile control system.

  13. The Bevalac Upgrade Project

    International Nuclear Information System (INIS)

    Alonso, J.R.; Dwinell, R.D.; Feinberg, B.

    1987-03-01

    This paper describes a proposed upgrade of the Bevalac accelerator complex in which the present Bevatron is replaced with a modern, strong-focusing 17 T-m synchrotron. This new ring is designed to accelerate all ions throughout the periodic table with intensities 100 to 1000 times higher than the present Bevatron. It will also provide a substantially improved beam spill structure and will reduce operating costs. A fast extraction capability can be used to inject a future heavy ion storage ring. Pulse-to-pulse switching of energy and ion species is an important goal. The existing injectors, shielding, experimental facilities and utilities of the present Bevalac will remain substantially intact

  14. Upgrade Software and Computing

    CERN Document Server

    The LHCb Collaboration, CERN

    2018-01-01

    This document reports the Research and Development activities that are carried out in the software and computing domains in view of the upgrade of the LHCb experiment. The implementation of a full software trigger implies major changes in the core software framework, in the event data model, and in the reconstruction algorithms. The increase of the data volumes for both real and simulated datasets requires a corresponding scaling of the distributed computing infrastructure. An implementation plan in both domains is presented, together with a risk assessment analysis.

  15. Upgrading Enterprise Search

    Energy Technology Data Exchange (ETDEWEB)

    McDunn, R

    2005-04-28

    This presentation will describe the process we went through this past year to upgrade our enterprise search tool from a very old version of Inktomi to the latest version of Verity Ultraseek. We started with requirements gathering and then compared requirements against several available products to determine which product to choose. After purchasing the product, we worked through several defined phases of implementation and customization, with initial rollout late January 2004. Finally, we will show you where we are today and describe future search plans.

  16. Upgrade, rebuild or replace?

    International Nuclear Information System (INIS)

    Forbes, C.A.

    1990-01-01

    Ageing reactor simulators present some tough decisions for utility managers. Although most utilities have chosen the cheaper, upgrading solution as the best compromise between costs and outage length, some US utilities have found that for them, replacement represents the best option. Simulators may be less than ten years old, but they have limited instructor systems, older low fidelity models that cannot reproduce important training scenarios, and out of date, difficult to maintain computers that do not permit much expansion of the models anyway. Perhaps worse than this is the possibility that the simulator may no longer be a faithful reproduction of the referenced plant, or have poor (or non-existent) documentation. (author)

  17. Repairing and Upgrading Your PC

    CERN Document Server

    Thompson, Robert

    2009-01-01

    Repairing and Upgrading Your PC delivers start-to-finish instructions, simple enough for even the most inexperienced PC owner, for troubleshooting, repairing, and upgrading your computer. Written by hardware experts Robert Bruce Thompson and Barbara Fritchman Thompson, this book covers it all: how to troubleshoot a troublesome PC, how to identify which components make sense for an upgrade, and how to tear it all down and put it back together. This book shows how to repair and upgrade all of your PC's essential components.

  18. Immersive virtual walk-through development for tokamak using active head mounted display

    International Nuclear Information System (INIS)

    Dutta, Pramit

    2015-01-01

    A fully immersive virtual walk-through of the SST-1 tokamak has been developed. The virtual walkthrough renders the virtual model of SST-1 tokamak through a active stereoscopic head mounted display to visualize the virtual environment. All locations inside and outside of the reactor can be accessed and reviewed. Such a virtual walkthrough provides a 1:1 scale visualization of all components of the tokamak. To achieve such a virtual model, the graphical details of the tokamak CAD model are enhanced. Such enhancements are provided to improve lighting conditions at various locations, texturing of components to have a realistic visual effect and 360° rendering for ease of access. The graphical enhancements also include the redefinition of the facets to optimize the surface triangles to remove lags in display during visual rendering. Two separate algorithms are developed to interact with the virtual model. A fly-by algorithm, developed using C#, uses inputs from a commercial joystick to navigate within the virtual environment. The second algorithm uses the IR and gyroscopic tracking system of the head mounted display to render view as per the current pose of the user within the virtual environment and the direction of view. Such a virtual walk-thorough can be used extensively for design review and integration, review of new components, operator training for remote handling, operations, upgrades of tokamak, etc. (author)

  19. Upgrade of ICRF heating system on EAST

    International Nuclear Information System (INIS)

    Chen Gen; Zhao Yanpin; Mao Yuzhou

    2013-01-01

    ICRF (Ion Cyclotron Range of Frequency) heating is an essential heating and current drive tool on EAST (Experimental Advanced Superconducting Tokamak). The high-power steady-state transmitters were designed as a part of research and development of ICRF heating system which aimed at output power of 1.5 MW for 1000 s in a frequency range of 25 to 70 MHz. There are 3 stage power amplifiers for each transmitter. Tube TH525A and TH535 were chosen for drive power amplifier (DPA) and final power amplifier (FPA), respectively. The power supply system of DPA and FPA were upgraded by using reliable PSM high voltage sources, whose response time is less than 5 μs. The ICRF system, which consists of 8 transmitters, will give out more than 10 MW total output power in the future. Four of them have been already fabricated, and another four are under construction. Three liquid stub tuners are used for impedance matching between antennas and transmitters, which can be only tuned shot to shot. There are two fast wave heating antennas which are assembled at I port and B port on EAST. Several projects are in progress including fast response impedance matching, distributed data acquisition and control system and so on for EAST ICRF heating system. (author)

  20. Stochastic sawtooth reconnection in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Igochine, V.; Dumbrajs, O.; Zohm, H.; Flaws, A.

    2007-01-01

    In this paper we investigate non-complete sawtooth reconnection in the ASDEX Upgrade tokamak. Such reconnection phenomena are associated with internal m/n = 1/1 kink mode which does not vanish after the crash phase (as would be the case for complete reconnection). It is shown that this sawtooth cannot be fully described by pure m/n = 1/1 mode and that higher harmonics play an important role during the sawtooth crash phase. We employ the Hamiltonian formalism and reconstructed perturbations to model incomplete sawtooth reconnection. It is demonstrated that stochastization appears due to the excitation of low-order resonances which are present in the corresponding q-profiles inside the q = 1 surface which reflects the key role of the q 0 value. Depending on this value two completely different situations are possible for one and the same mode perturbations: (i) the resonant surfaces are present in the q-profile leading to stochasticity and sawtooth crash (q 0 ∼ 0.7 ± 0.1); (ii) the resonant surfaces are not present, which means no stochasticity in the system and no crash event (q 0 ∼ 0.9 ± 0.05). Accordingly the central safety factor value is always less than unity in the case of a non-complete sawtooth reconnection. Our investigations show that the stochastic model agrees well with the experimental observations and can be proposed as a promising candidate for an explanation of the sawtooth reconnection

  1. Alfven wave experiments on the TORTUS tokamak

    International Nuclear Information System (INIS)

    Ballico, M.J.; Bowden, M.; Brand, G.F.; Brennan, M.H.; Cross, R.C.; Fekete, P.; James, B.W.

    1989-01-01

    Results are presented on the first observations of the Discrete Alfven Wave (DAW) and the first measurements of laser scattering off the kinetic Alfven wave in the TORTUS tokamak. TORTUS is a relatively small device, with major radius R=0.44m, minor radius 0.1m and has previously been operated routinely with B Φ =0.7T, I p =20 kA and n e ∼ 1x10 19 m -3 . Under these conditions, and over a wide frequency range (1-14 MHz), there has been no evidence of the DAW modes observed on TCA. Recently, a minor upgrade of TORTUS has permitted routine operation at B Φ =1.0 T, I p =39 kA, q(a)∼5 and n e ∼1-4 x 10 19 m -3 . At the operating frequency, 3.2 MHz, chosen for this study, DAW modes are observed clearly at both low and high densities. The appearance of DAW modes appears to be due to a steeper current profile at the higher plasma currents now generated in TORTUS. The general behaviour of DAW modes is in fact quite sensitive to the density and current profiles, indicating that DAW modes should provide a useful current profile diagnostic. (author) 6 refs., 2 figs

  2. DOE FES FY2017 Joint Research Target Fourth Quarter Milestone Report for theNational Spherical Torus Experiment Upgrade.

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V. A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-09-13

    A successful high-performance plasma operation with a radiative divertor has been demonstrated on many tokamak devices, however, significant uncertainty remains in accurately modeling detachment thresholds, and in how detachment depends on divertor geometry. Whereas it was originally planned to perform dedicated divertor experiments on the National Spherical Tokamak Upgrade to address critical detachment and divertor geometry questions for this milestone, the experiments were deferred due to technical difficulties. Instead, existing NSTX divertor data was summarized and re-analyzed where applicable, and additional simulations were performed.

  3. 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...

  4. ATLAS detector upgrade prospects

    CERN Document Server

    AUTHOR|(INSPIRE)INSPIRE-00184940; The ATLAS collaboration

    2017-01-01

    After the successful operation at the centre-of-mass energies of 7 and 8 TeV in 2010-2012, the LHC is ramped up and successfully took data at the centre-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 levelling. 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 ...

  5. LHCb VELO Upgrade

    CERN Document Server

    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...

  6. ATLAS Detector Upgrade Prospects

    International Nuclear Information System (INIS)

    Dobre, M

    2017-01-01

    After the successful operation at the centre-of-mass energies of 7 and 8 TeV in 2010-2012, the LHC was ramped up and successfully took data at the centre-of-mass energies of 13 TeV in 2015 and 2016. 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, which will deliver of the order of five times the LHC nominal instantaneous luminosity along with luminosity levelling. The ultimate goal is to extend the dataset from about few hundred fb −1 expected for LHC running by the end of 2018 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 extensions to larger pseudorapidity, particularly in tracking and muon systems. This report summarizes various improvements to the ATLAS detector required to cope with the anticipated evolution of the LHC luminosity during this decade and the next. A brief overview is also given on physics prospects with a pp centre-of-mass energy of 14 TeV. (paper)

  7. Micro-NRA and micro-3HIXE with He-3 microbeam on samples exposed in ASDEX Upgrade and Pilot-PSI machines

    NARCIS (Netherlands)

    Kelemen, M.; Zaloznik, A.; Vavpetic, P.; Pecovnik, M.; Pelicon, P.; Hakola, A.; Lahtinen, A.; Karhunen, J.; Piip, K.; van der Meiden, H. J.; Paris, P.; Laan, M.; Krieger, K.; Oberkofler, M.; Markelj, S.; ASDEX Upgrade team,

    2017-01-01

    Micro nuclear reaction analysis (micro-NRA) exploiting the nuclear reaction D(He-3,p)He-4 was used for post-mortem analyses of special marker samples, exposed to deuterium plasma inside ASDEX Upgrade (AUG) tokamak and to the deuterium plasma jet in the Pilot-PSI linear plasma gun. Lateral

  8. Bibliography of fusion product physics in tokamaks

    International Nuclear Information System (INIS)

    Hively, L.M.; Sigmar, D.J.

    1989-09-01

    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

  9. A compact, smart Langmuir Probe control module for MAST-Upgrade

    Science.gov (United States)

    Lovell, J.; Stephen, R.; Bray, S.; Naylor, G.; Elmore, S.; Willett, H.; Peterka, M.; Dimitrova, M.; Havranek, A.; Hron, M.; Sharples, R.

    2017-11-01

    A new control module for the MAST-Upgrade Langmuir Probe system has been developed. It is based on a Xilinx Zynq FPGA, which allows for excellent configurability and ease of retrieving data. The module is capable of arbitrary bias voltage waveform generation, and digitises current and voltage readings from 16 probes. The probes are biased and measured one at a time in a time multiplexed fashion, with the multiplexing sequence completely configurable. In addition, simultaneous digitisation of the floating potential of all unbiased probes is possible. A suite of these modules, each coupled with a high voltage amplifier, enables biasing and digitisation of 640 Langmuir Probes in the MAST-Upgrade Super-X divertor. The system has been successfully tested on the York Linear Plasma Device and on the COMPASS tokamak. It will be installed on MAST-Upgrade ready for operations in 2018.

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

    International Nuclear Information System (INIS)

    Mcharg, B.B.

    1983-01-01

    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 20-25 megabytes per shot. This is in comparison to the 4-5 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

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

    International Nuclear Information System (INIS)

    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

  12. Tokamak engineering test reactor

    International Nuclear Information System (INIS)

    Conn, R.W.; Jassby, D.L.

    1975-07-01

    The design criteria for a tokamak engineering test reactor can be met by operating in the two-component mode with reacting ion beams, together with a new blanket-shield design based on internal neutron spectrum shaping. A conceptual reactor design achieving a neutron wall loading of about 1 MW/m 2 is presented. The tokamak has a major radius of 3.05 m, the plasma cross-section is noncircular with a 2:1 elongation, and the plasma radius in the midplane is 55 cm. The total wall area is 149 m 2 . The plasma conditions are T/sub e/ approximately T/sub i/ approximately 5 keV, and ntau approximately 8 x 10 12 cm -3 s. The plasma temperature is maintained by injection of 177 MW of 200-keV neutral deuterium beams; the resulting deuterons undergo fusion reactions with the triton-target ions. The D-shaped toroidal field coils are extended out to large major radius (7.0 m), so that the blanket-shield test modules on the outer portion of the torus can be easily removed. The TF coils are superconducting, using a cryogenically stable TiNb design that permits a field at the coil of 80 kG and an axial field of 38 kG. The blanket-shield design for the inner portion of the torus nearest the machine center line utilizes a neutron spectral shifter so that the first structural wall behind the spectral shifter zone can withstand radiation damage for the reactor lifetime. The energy attenuation in this inner blanket is 8 x 10 -6 . If necessary, a tritium breeding ratio of 0.8 can be achieved using liquid lithium cooling in the []outer blanket only. The overall power consumption of the reactor is about 340 MW(e). A neutron wall loading greater than 1 MW/m 2 can be achieved by increasing the maximum magnetic field or the plasma elongation. (auth)

  13. START: the creation of a spherical tokamak

    International Nuclear Information System (INIS)

    Sykes, Alan

    1992-01-01

    The START (Small Tight Aspect Ratio Tokamak) plasma fusion experiment is now operational at AEA Fusion's Culham Laboratory. It is the world's first experiment to explore an extreme limit of the tokamak - the Spherical Tokamak - which theoretical studies predict may have substantial advantages in the search for economic fusion power. The Head of the START project, describes the concept, some of the initial experimental results and the possibility of developing a spherical tokamak power reactor. (author)

  14. Slum Upgrading and Health Equity.

    Science.gov (United States)

    Corburn, Jason; Sverdlik, Alice

    2017-03-24

    Informal settlement upgrading is widely recognized for enhancing shelter and promoting economic development, yet its potential to improve health equity is usually overlooked. Almost one in seven people on the planet are expected to reside in urban informal settlements, or slums, by 2030. Slum upgrading is the process of delivering place-based environmental and social improvements to the urban poor, including land tenure, housing, infrastructure, employment, health services and political and social inclusion. The processes and products of slum upgrading can address multiple environmental determinants of health. This paper reviewed urban slum upgrading evaluations from cities across Asia, Africa and Latin America and found that few captured the multiple health benefits of upgrading. With the Sustainable Development Goals (SDGs) focused on improving well-being for billions of city-dwellers, slum upgrading should be viewed as a key strategy to promote health, equitable development and reduce climate change vulnerabilities. We conclude with suggestions for how slum upgrading might more explicitly capture its health benefits, such as through the use of health impact assessment (HIA) and adopting an urban health in all policies (HiAP) framework. Urban slum upgrading must be more explicitly designed, implemented and evaluated to capture its multiple global environmental health benefits.

  15. Upgrade trigger: Biannual performance update

    CERN Document Server

    Aaij, Roel; Couturier, Ben; Esen, Sevda; De Cian, Michel; De Vries, Jacco Andreas; Dziurda, Agnieszka; Fitzpatrick, Conor; Fontana, Marianna; Grillo, Lucia; Hasse, Christoph; Jones, Christopher Rob; Le Gac, Renaud; Matev, Rosen; Neufeld, Niko; Nikodem, Thomas; Polci, Francesco; Del Buono, Luigi; Quagliani, Renato; Schwemmer, Rainer; Seyfert, Paul; Stahl, Sascha; Szumlak, Tomasz; Vesterinen, Mika Anton; Wanczyk, Joanna; Williams, Mark Richard James; Yin, Hang; Zacharjasz, Emilia Anna

    2017-01-01

    This document presents the performance of the LHCb Upgrade trigger reconstruction sequence, incorporating changes to the underlying reconstruction algorithms and detector description since the Trigger and Online Upgrade TDR. An updated extrapolation is presented using the most recent example of an Event Filter Farm node.

  16. Major upgrade of the articulated inspection arm control system to fulfill daily operation requirements

    Energy Technology Data Exchange (ETDEWEB)

    Pastor, P., E-mail: patrick.pastor@cea.fr [CEA, IRFM, Institut de Recherche sur la Fusion par confinement Magnétique (France); Villedieu, E.; Allegretti, L.; Vincent, B.; Barbuti, A.; Bruno, V.; Coquillat, P.; Dechelle, C.; Gargiulo, L.; Le, R.; Malard, P.; Martinez, A.; Nouailletas, R. [CEA, IRFM, Institut de Recherche sur la Fusion par confinement Magnétique (France); Yuntao, Song; Yong, Cheng; Chen, Liu; Hansheng, Feng; Shanshuang, Shi [ASIPP, Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui (China)

    2015-10-15

    Highlights: • We propose an overview of the work which has been done to upgrade the control system of the AIA robot (articulated inspection arm) to fulfill daily operation requirements for tokamak inspection. • The control system is based on the use of new position sensors, new electronics design and new supervisor software. • Final tests are ongoing in the EAST scale 1 tokamak mock-up. Routine operation of the robot at EAST will start in the beginning of 2015. - Abstract: An articulated inspection arm (AIA) has been developed by CEA for visual inspection between pulses inside the Tore Supra tokamak vacuum vessel without breaking temperature and vacuum conditions. The eight meters length robot is composed of a shuttle and six articulated segments with a video camera at its end. A demonstration prototype has been achieved in 2008 at Tore Supra (Gargiulo, 2007; Houry, 2008; Perrot, 2003). A project to upgrade the AIA into a fully operational robot has been undertaken by IRFM and ASIPP in an Associated Laboratory. It will be in operation first in the EAST machine and afterwards in Tore Supra in its WEST (W/Tungsten Environment in Steady-state Tokamak) configuration where it is of paramount importance to survey possible degradation of W component surface. The control system of the robot has been extensively upgraded. The effort has been focused on three areas: (1) improvement of the arm position accuracy, (2) increase of the operational robustness, (3) use of a powerful graphical user interface including simulation of trajectories and robot deployment capabilities in a 3D viewer environment. The aim of this paper is to detail the architecture of the AIA control system.

  17. Upgrading the ATLAS control system

    International Nuclear Information System (INIS)

    Munson, F.H.; Ferraretto, M.

    1993-01-01

    Heavy-ion accelerators are tools used in the research of nuclear and atomic physics. The ATLAS facility at the Argonne National Laboratory is one such tool. The ATLAS control system serves as the primary operator interface to the accelerator. A project to upgrade the control system is presently in progress. Since this is an upgrade project and not a new installation, it was imperative that the development work proceed without interference to normal operations. An additional criteria for the development work was that the writing of additional ''in-house'' software should be kept to a minimum. This paper briefly describes the control system being upgraded, and explains some of the reasons for the decision to upgrade the control system. Design considerations and goals for the new system are described, and the present status of the upgrade is discussed

  18. LHC luminosity upgrade detector challenges

    CERN Multimedia

    CERN. Geneva; de Roeck, Albert; Bortoletto, Daniela; Wigmans, Richard; Riegler, Werner; Smith, Wesley H

    2006-01-01

    LHC luminosity upgrade: detector challenges The upgrade of the LHC machine towards higher luminosity (1035 cm -2s-1) has been studied over the last few years. These studies have investigated scenarios to achieve the increase in peak luminosity by an order of magnitude, as well as the physics potential of such an upgrade and the impact of a machine upgrade on the LHC DETECTORS. This series of lectures will cover the following topics: • Physics motivation and machine scenarios for an order of magnitude increase in the LHC peak luminosity (lecture 1) • Detector challenges including overview of ideas for R&D programs by the LHC experiments: tracking and calorimetry, other new detector developments (lectures 2-4) • Electronics, trigger and data acquisition challenges (lecture 5) Note: the much more ambitious LHC energy upgrade will not be covered

  19. Moving Divertor Plates in a Tokamak

    International Nuclear Information System (INIS)

    Zweben, S.J.; Zhang, H.

    2009-01-01

    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

  20. Fusion potential for spherical and compact tokamaks

    International Nuclear Information System (INIS)

    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 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

  1. 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.

  2. 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.

  3. The Fermilab ACNET upgrade

    International Nuclear Information System (INIS)

    Briegel, C.; Johnson, G.; Winterowd, L.

    1990-01-01

    The Fermilab Accelerator Controls Network (ACNET) upgrade consists of a new physical medium (IEEE 802.5 token ring), additions to the calling sequence and added processor support. ACNET is the accelerator control backbone network for all data communication. A proprietary network was replaced by an IEEE standard enabling an open network with excellent characteristics for the control system. The calling sequence was enhanced for the added capabilities of the token-ring interface such as 'gather-read' and 'scatter-write'. In addition to prior support of DEC PDP11s under RS11M and VAXs under VMS, the ACNET calling sequence was implemented in the language C for the IBM PC with MS-DOS and Motorola 680x0 with MTOS using VME bus. Additional support is in progress for Intel 80x86 with MTOS using Multibus II. (orig.)

  4. Resistive instabilities in tokamaks

    International Nuclear Information System (INIS)

    Rutherford, P.H.

    1985-10-01

    Low-m tearing modes constitute the dominant instability problem in present-day tokamaks. In this lecture, the stability criteria for representative current profiles with q(0)-values slightly less than unit are reviewed; ''sawtooth'' reconnection to q(0)-values just at, or slightly exceeding, unity is generally destabilizing to the m = 2, n = 1 and m = 3, n = 2 modes, and severely limits the range of stable profile shapes. Feedback stabilization of m greater than or equal to 2 modes by rf heating or current drive, applied locally at the magnetic islands, appears feasible; feedback by island current drive is much more efficient, in terms of the radio-frequency power required, then feedback by island heating. Feedback stabilization of the m = 1 mode - although yielding particularly beneficial effects for resistive-tearing and high-beta stability by allowing q(0)-values substantially below unity - is more problematical, unless the m = 1 ideal-MHD mode can be made positively stable by strong triangular shaping of the central flux surfaces. Feedback techniques require a detectable, rotating MHD-like signal; the slowing of mode rotation - or the excitation of non-rotating modes - by an imperfectly conducting wall is also discussed

  5. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

    1978-01-01

    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

  6. Classical tokamak transport theory

    International Nuclear Information System (INIS)

    Nocentini, Aldo

    1982-01-01

    A qualitative treatment of the classical transport theory of a magnetically confined, toroidal, axisymmetric, two-species plasma is presented. The 'weakly collisional' ('banana' and 'plateau') and 'collision dominated' ('Pfirsch-Schlueter' and 'highly collisional') regimes, as well as the Ware effect are discussed. The method used to evaluate the diffusion coffieicnts of particles and heat in the weakly collisional regime is based on stochastic argument, that requires an analysis of the characteristic collision frequencies and lengths for particles moving in a tokamak-like magnetic field. The same method is used to evaluate the Ware effect. In the collision dominated regime on the other hand, the particle and heat fluxes across the magnetic field lines are dominated by macroscopic effects so that, although it is possible to present them as diffusion (in fact, the fluxes turn out to be proportional to the density and temperature gradients), a macroscopic treatment is more appropriate. Hence, fluid equations are used to inveatigate the collision dominated regime, to which particular attention is devoted, having been shown relatively recently that it is more complicated than the usual Pfirsch-Schlueter regime. The whole analysis presented here is qualitative, aiming to point out the relevant physical mechanisms involved in the various regimes more than to develop a rigorous mathematical derivation of the diffusion coefficients, for which appropriate references are given. (author)

  7. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Bertoncini, P.J.

    1976-01-01

    A conceptual design has been developed for a tokamak Experimental Power Reactor to operate at net electrical power conditions with a plant capacity factor of 50 percent for 10 yr. The EPR operates in a pulsed mode at a frequency of approximately 1/min, with approximately 75 percent duty cycle, is capable of producing approximately 72 MWe and requires 42 MWe. The annual tritium consumption is 16 kg. The EPR vacuum chamber is 6.25 m in major radius and 2.4 m in minor radius, is constructed of 2 cm thick stainless steel, and has 2 cm thick detachable, beryllium-coated coolant panels mounted on the interior. A 0.28 m stainless steel blanket and a shield ranging from 0.6 to 1.0 m surround the vacuum vessel. The coolant is H 2 O. Sixteen niobium-titanium superconducting toroidal field coils provide a field of 10 T at the coil and 4.47 T at the plasma. Superconducting ohmic heating and equilibrium field coils provide 135 V-s to drive the plasma current. Plasma heating is accomplished by 12 neutral beam injectors which provide 60 MW. The energy transfer and storage system consists of a central superconducting storage ring, a homopolar energy storage unit, and a variety of inductor-convertors

  8. Design of multi-range tomographic system for transport studies in tokamak plasmas

    Czech Academy of Sciences Publication Activity Database

    Weinzettl, Vladimír; Naydenkova, Diana; Šesták, David; Vlček, Jiří; Mlynář, Jan; Melich, Radek; Jareš, Daniel; Malot, J.; Sarychev, D.; Igochine, V.

    2010-01-01

    Roč. 623, č. 2 (2010), s. 806-808 ISSN 0168-9002. [International Conference on Frontiers in Diagnostic Technologies/1st./. Frascati, 25.11.2009-27.11.2009] R&D Projects: GA ČR GA202/09/1467 Institutional research plan: CEZ:AV0Z20430508 Keywords : SXR * Bolometry * Visible light * Tokamak * COMPASS Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.142, year: 2010 http://www.sciencedirect.com/science?_ob=ArticleURL&_udi=B6TJM-4YTFBF2-C&_user=6542793&_coverDate=11%2F11%2F2010&_rdoc=1&_fmt=high&_orig=search&_origin=search&_sort=d&_docanchor=&view=c&_searchStrId=1546173561&_rerunOrigin= google &_acct=C000070123&_version=1&_urlVersion=0&_userid=6542793&md5=1642ebdad07f4009a3a7b81bb4d11c8f&searchtype=a

  9. The role of the spherical tokamak in clarifying tokamak physics

    International Nuclear Information System (INIS)

    Morris, A.W.; Akers, R.J.; Connor, J.W.; Counsell, G.F.; Gryaznevich, M.P.; Hender, T.C.; Maddison, G.P.; Martin, T.J.; McClements, K.G.; Roach, C.M.; Robinson, D.C.; Sykes, A.; Valovic, M.; Wilson, H.R.; Fonck, R.J.; Gusev, V.; Kaye, S.M.; Majeski, R.; Peng, Y.-K.M.; Medvedev, S.; Sharapov, S.; Walsh, M.J.

    1999-01-01

    The spherical tokamak (ST) provides a unique environment in which to perform complementary and exacting tests of the tokamak physics required for a burning plasma experiment of any aspect ratio, while also having the potential for long-term fusion applications in its own right. New experiments are coming on-line in the UK (MAST), USA (NSTX, Pegasus), Russia (Globus-M), Brazil (ETE) and elsewhere, and the status of these devices will be reported, along with newly-analysed data from START. Those physics issues where the ST provides an opportunity to remove degeneracy in the databases or clarify one's understanding will be emphasized. (author)

  10. Advanced tokamak burning plasma experiment

    International Nuclear Information System (INIS)

    Porkolab, M.; Bonoli, P.T.; Ramos, J.; Schultz, J.; Nevins, W.N.

    2001-01-01

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

  11. Large aspect ratio tokamak study

    International Nuclear Information System (INIS)

    Reid, R.L.; Holmes, J.A.; Houlberg, W.A.; Peng, Y.K.M.; Strickler, D.J.; Brown, T.G.; Sardella, C.; Wiseman, G.W.

    1979-01-01

    The Large Aspect Ratio Tokamak Study (LARTS) investigated the potential for producing a viable long burn tokamak reactor through enhanced volt-second capability of the ohmic heating transformer by employing high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were accessed in the context of extended burn operation. Plasma startup and burn parameters were addressed using a one-dimensional transport code. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the field in the ohmic heating coil and the wave shape of the ohmic heating discharge. A high aspect ratio reference reactor was chosen and configured

  12. Plasma boundary phenomena in tokamaks

    International Nuclear Information System (INIS)

    Stangeby, P.C.

    1989-06-01

    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 (n e and T e ) 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.)

  13. Computational studies of tokamak plasmas

    International Nuclear Information System (INIS)

    Takizuka, Tomonori; Tsunematsu, Toshihide; Tokuda, Shinji

    1981-02-01

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

  14. Summary discussion: An integrated advanced tokamak reactor

    International Nuclear Information System (INIS)

    Sauthoff, N.R.

    1994-01-01

    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

  15. Calculation of the electromagnetic forces on the ASDEX upgrade vacuum vessel on disruption of the plasma current

    International Nuclear Information System (INIS)

    Preis, H.

    1986-01-01

    This study investigates the magnetic field diffusion through the vacuum vessel of the ASDEX Upgrade tokamak that occurs on sudden disruption of the plasma current. Eddy currents are thereby produced in the vessel wall. Their time behaviour and distribution are determined. Furthermore, the vessel is permeated by various magnetic fields which, together with the eddy currents, exert magnetic forces in the vessel wall. These are also calculated. These numerical analyses are performed for two of the modes of operation envisaged for ASDEX Upgrade: the so-called limiter and single-null magnetic field configurations. (orig.)

  16. STARFIRE: a commercial tokamak reactor

    International Nuclear Information System (INIS)

    1979-12-01

    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

  17. LHCD experiments on tokamak CASTOR

    International Nuclear Information System (INIS)

    Zacek, F.; Badalec, J.; Jakubka, J.; Kryska, L.; Preinhaelter, J.; Stoeckel, J.; Valovic, M.; Nanobashvili, S.; Weixelbaum, L.; Wenzel, U.; Spineanu, F.; Vlad, M.

    1990-10-01

    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

  18. ALT-II armor tile design for upgraded TEXTOR operation

    International Nuclear Information System (INIS)

    Newberry, B.L.; McGrath, R.T.; Watson, R.D.; Kohlhaas, W.; Finken, K.H.

    1994-01-01

    The upgrade of the TEXTOR tokamak at KFA Juelich was recently completed. This upgrade extended the TEXTOR pulse length from 5 seconds to 10 seconds. The auxiliary heating was increased to a total of 8.0 MW through a combination of neutral beam injection and radio frequency heating. Originally, the inertially cooled armor tiles of the full toroidal belt Advanced Limiter Test -- II (ALT-II) were designed for a 5-second operation with total heating of 6.0 MW. The upgrade of TEXTOR will increase the energy deposited per pulse onto the ALT-II by about 300%. Consequently, the graphite armor tiles for the ALT-II had to be redesigned to avoid excessively high graphite armor surface temperatures that would lead to unacceptable contamination of the plasma. This redesign took the form of two major changes in the ALT-II armor tile geometry. The first design change was an increase of the armor tile thermal mass, primarily by increasing the radial thickness of each tile from 17 mm to 20 mm. This increase in the radial tile dimension reduces the overall pumping efficiency of the ALT-II pump limiter by about 30%. The reduction in exhaust efficiency is unfortunate, but could be avoided only by active cooling of the ALT-II armor tiles. The active cooling option was too complicated and expensive to be considered at this time. The second design change involved redefining the plasma facing surface of each armor tile in order to fully utilize the entire surface area. The incident charged particle heat flux was distributed uniformly over the armor tile surfaces by carefully matching the radial, poloidal and toroidal curvature of each tile to the plasma flow in the TEXTOR boundary layer. This geometry redefinition complicates the manufacturing of the armor tiles, but results in significant thermal performance gains. In addition to these geometry upgrades, several material options were analyzed and evaluated

  19. ALT-II armor tile design for upgraded TEXTOR operation

    International Nuclear Information System (INIS)

    Newberry, B.L.; McGrath, R.T.; Watson, R.D.

    1994-01-01

    The upgrade of the TEXTOR tokamak at KFA Julich will be completed in the spring of 1994. The upgrade will extend the TEXTOR pulse length from 5 seconds to 10 seconds. The auxiliary heating systems are also scheduled to be upgraded so that eventually a total of 8.0 MW auxiliary heating will be available through a combination of neutral beam injection and radio frequency heating. Originally, the inertially cooled armor tiles on the full toroidal belt Advanced Limiter Test - II (ALT-II) were designed for 5-second operation with a total heating power of 6.0 MW. The upgrade of TEXTOR will increase the energy deposited per pulse onto ALT-II by more than 300%. Consequently, the graphite armor tiles for ALT-II had to be redesigned in order to increase their thermal inertia and, thereby, avoid excessively high graphite armor surface temperatures that would lead to unacceptable contamination of the plasma. The armor tile thermal inertia had been increase primarily by expanding the radial thickness of the tiles from 17 mm to 20 mm. This increase in radial tile dimension will reduce the overall pumping efficiency of the ALT-II pump limiter by about 30%. The final armor tile design was a compromise between increasing the power handling capability and reducing the particle exhaust efficiency of ALT-II. The reduction in exhaust efficiency is unfortunate, but could only be avoided by active cooling of the ALT-II armor tiles. The active cooling option was too complicated and expensive to be considered at this time

  20. The tokamak hybrid reactor

    International Nuclear Information System (INIS)

    Kelly, J.L.; Rose, R.P.

    1981-01-01

    At a time when the potential benefits of various energy options are being seriously evaluated in many countries through-out the world, it is both timely and important to evaluate the practical application of fusion reactors for their economical production of nuclear fissile fuels from fertile fuels. The fusion hybrid reactor represents a concept that could assure the availability of adequate fuel supplies for a proven nuclear technology and have the potential of being an electrical energy source as opposed to an energy consumer as are the present fuel enrichment processes. Westinghouse Fusion Power Systems Department, under Contract No. EG-77-C-02-4544 with the Department of Energy, Office of Fusion Energy, has developed a preliminary conceptual design for an early twenty-first century fusion hybrid reactor called the commercial Tokamak Hybrid Reactor (CTHR). This design was developed as a first generation commercial plant producing fissile fuel to support a significant number of client Light Water Reactor (LWR) Plants. To the depth this study has been performed, no insurmountable technical problems have been identified. The study has provided a basis for reasonable cost estimates of the hybrid plants as well as the hybrid/LWR system busbar electricity costs. This energy system can be optimized to have a net cost of busbar electricity that is equivalent to the conventional LWR plant, yet is not dependent on uranium ore prices or standard enrichment costs, since the fusion hybrid can be fueled by numerous fertile fuel resources. A nearer-term concept is also defined using a beam driven fusion driver in lieu of the longer term ignited operating mode. (orig.)

  1. Tokamak Plasmas : Mirnov coil data analysis for tokamak ADITYA

    Indian Academy of Sciences (India)

    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 ...

  2. H-1 Upgrades (4BW/4BN) (H-1 Upgrades)

    Science.gov (United States)

    2015-12-01

    automatic blade fold of the new composite rotor blades, new performance matched transmissions, a new four-bladed tail rotor and drive system, upgraded...Upgrades December 2015 SAR March 18, 2016 10:59:17 UNCLASSIFIED 4 Col Steven Girard PMA-276 USMC Light/Attack Helicopter Program Executive Officer...attack helicopter is to provide rotary wing close air support, anti-armor, armed escort, armed/visual reconnaissance and fire support coordination

  3. Low frequency sawtooth precursor activity in ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Papp, G; Pokol, G I; Por, G; Magyarkuti, A; Lazanyi, N; Horvath, L [Department of Nuclear Techniques, Budapest University of Technology and Economics, Association EURATOM, Pf 91, H-1521 Budapest (Hungary); Igochine, V; Maraschek, M, E-mail: papp@reak.bme.h [Max-Planck-Institut fuer Plasmaphysik, Association EURATOM, D-85748 Garching (Germany)

    2011-06-15

    This paper describes the precursor activity observed in the ASDEX Upgrade tokamak before sawtooth crashes in various neutral beam heated plasmas, utilizing the soft x-ray diagnostic. In addition to the well-known (m, n) = (1,1) internal kink mode and its harmonics, a lower frequency mode is studied in detail. Power modulation of this mode is found to correlate with the power modulation of the (1, 1) kink mode in the quasistationary intervals indicating possible nonlinear interaction. Throughout the studied sawtooth crashes, the power of the lower frequency mode rose by several orders of magnitude just before the crash. In addition to its temporal behaviour, its spatial structure was estimated and the most likely value was found to be (1, 1). A possible role of this mode in the mechanism of the sawtooth crash is discussed.

  4. Motional stark effect upgrades on DIII-D

    International Nuclear Information System (INIS)

    Rice, B.W.; Nilson, D.G.; Wroblewski, D.

    1994-04-01

    The measurement and control of the plasma current density profile (or q profile) is critical to the advanced tokamak program on DIII-D. A complete understanding of the stability and transport properties of advanced operating regimes requires detail poloidal field measurements over the entire plasma radius from the core to the edge. In support of this effort, the authors have recently completed an upgrade of the existing MSE diagnostic, increasing the number of channels from 8 to 16. A new viewing geometry has been added to the outer edge of the plasma which improves the radial resolution in this region from 10 cm to < 4 cm. This view requires the use of a reflector that has been designed to minimize polarization amplitude and phase effects. Vacuum-compatible polarizers have also been added to the instrument for in-situ calibration. Future use of the MSE diagnostic for feedback control of the q profile will also be discussed

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-15

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

  6. ICRF induced edge plasma convection in ASDEX upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Wei [Max Planck Institute for Plasma Physics, Garching/Greifswald (Germany); University of Ghent, Ghent (Belgium); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Feng, Yuehe; Lunt, Tilmann; Jacquot, Jonathan; Coster, David; Bilato, Roberto; Bobkov, Volodymyr; Ochoukov, Roman [Max Planck Institute for Plasma Physics, Garching/Greifswald (Germany); Noterdaeme, Jean-Marie [Max Planck Institute for Plasma Physics, Garching/Greifswald (Germany); University of Ghent, Ghent (Belgium); Colas, Laurent [CEA, IRFM, Saint-Paul-Lez-Durance (France); Collaboration: ASDEX Upgrade Team

    2016-07-01

    Ion Cyclotron Range of Frequency (ICRF) heating is one of the main auxiliary plasma heating methods in tokamaks. It relies on the fast wave to heat the plasma. However the slow wave can also be generated parasitically. The parallel electric field of the slow wave can induce large biased plasma potential through sheath rectification. The rapid variation of this rectified potential across the magnetic field can cause significant E x B convection in the Scrape-Off Layer (SOL). The ICRF induced convection can affect the SOL density, influence the ICRF power coupling and enhance the strength of plasma-wall interactions. To explore these physics, we not only show the experimental evidences in ASDEX Upgrade, but also present the associated simulation results with the 3D edge plasma fluid code EMC3-Eirene. Further simulations via combination of EMC3-Eirene and a sheath code SSWICH in an iterative and quasi self-consistent way can give good predictions for future experiments.

  7. Development of upgraded pellet injector for JT-60

    International Nuclear Information System (INIS)

    Onozuka, M.; Shimomura, T.; Tanaka, N.; Iwamoto, S.; Hashiri, N.; Oda, Y.; Minami, M.; Hiratsuka, H.; Kawasaki, K.; Takatsu, H.; Shimizu, M.

    1989-01-01

    The pneumatic 4-shot pellet injector had been in use for JT-60 (JAERI Tokamak-60) contributing to plasma studies in 1988. It could propel the pellets up to 1.6 km/sec at 50 bar propellant gas. In 1989, the new gun assembly has been reinstalled in the upgraded system to provide higher performance and reliability. The supply pressure of the propellant gas is to be raised to 100 bar to obtain higher pellet velocity up to 2.3 km/sec. The device is now in use for JT-60, and is expected to contribute to further plasma studies. In this paper the outline of features and performance of the device is presented. 5 refs., 9 figs

  8. Low frequency sawtooth precursor activity in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Papp, G; Pokol, G I; Por, G; Magyarkuti, A; Lazanyi, N; Horvath, L; Igochine, V; Maraschek, M

    2011-01-01

    This paper describes the precursor activity observed in the ASDEX Upgrade tokamak before sawtooth crashes in various neutral beam heated plasmas, utilizing the soft x-ray diagnostic. In addition to the well-known (m, n) = (1,1) internal kink mode and its harmonics, a lower frequency mode is studied in detail. Power modulation of this mode is found to correlate with the power modulation of the (1, 1) kink mode in the quasistationary intervals indicating possible nonlinear interaction. Throughout the studied sawtooth crashes, the power of the lower frequency mode rose by several orders of magnitude just before the crash. In addition to its temporal behaviour, its spatial structure was estimated and the most likely value was found to be (1, 1). A possible role of this mode in the mechanism of the sawtooth crash is discussed.

  9. On the avalanche generation of runaway electrons during tokamak disruptions

    International Nuclear Information System (INIS)

    Martín-Solís, J. R.; Loarte, A.; Lehnen, M.

    2015-01-01

    A simple zero dimensional model for a tokamak disruption is developed to evaluate the avalanche multiplication of a runaway primary seed during the current quench phase of a fast disruptive event. Analytical expressions for the plateau runaway current, the energy of the runaway beam, and the runaway energy distribution function are obtained allowing the identification of the parameters dominating the formation of the runaway current during disruptions. The effect of the electromagnetic coupling to the vessel and the penetration of the external magnetic energy during the disruption current quench as well as of the collisional dissipation of the runaway current at high densities are investigated. Current profile shape effects during the formation of the runaway beam are also addressed by means of an upgraded one-dimensional model

  10. The Atlas upgrade project

    International Nuclear Information System (INIS)

    Bollinger, L.M.

    1988-01-01

    ATLAS is a heavy-ion accelerator system consisting of a 9-MV tandem electrostatic injector coupled to a superconducting linac. A project now well advanced will upgrade the capabilities of ATLAS immensely by replacing the tandem and its negative-ion source with a positive-ion injector that consists of an electron-cyclotron resonance (ECR) ion source and a 12-MV superconducting injector linac of novel design. This project will increase the beam intensity 100 fold and will extend the projectile-mass range up to uranium. Phase 1 of the work, which is nearing completion in late 1988, will provide an injector comprising the ECR source and its 350-kV voltage platform, beam analysis and bunching systems, beam lines, and a prototype 3-MV linac. The ECR source and its voltage platform are operational, development of the new class of low-frequency interdigital superconducting resonators required for the injector linac has been completed, and assembly of the whole system is in progress. Test runs and then routine use of the Phase 1 injector systems are planned for early 1989, and the final 12-MV injector linac will be commissioned in 1990. 12 refs., 6 figs

  11. Superhilac upgrade project

    International Nuclear Information System (INIS)

    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 7 to 5 x 10 7 ions/pulse, allowing accurate Lamb shift measurements to be made in U 90+ and U 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

  12. 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.

  13. Alberta propylene upgrading prospects

    International Nuclear Information System (INIS)

    2000-03-01

    A very significant byproduct recovery and purification scheme is at present being prepared by TransCanada Midstream (TCMS). Alberta Economic Development commissioned an independent study to identify propylene supply options while proceeding with the evaluation of various propylene derivatives with regard to their fit with the Alberta context. Identification of chemical companies with derivative interests was also accomplished. By 2005, it is estimated that 280 kilo-tonnes of propylene will be available on an annual basis from byproduct sources. Those sources are oil sands upgraders, ethylene plants and refineries. The ranges of impurities and supply costs vary between the different sources. An option being considered involves pipeline and rail receipt with a major central treating and distillation facility for the production of polymer grade (PG) propylene with propane and other smaller byproducts. Special consideration was given to three chemicals in this study, namely: polypropylene (PP), acrylonitrile (ACN), and acrylic acid (AA). Above average growth rates were identified for these chemicals: demand is growing at 6 to 7 per cent a year for both PP and ACN, while demand for AA grows at 8 per cent annually. Two other possibilities were identified, propylene oxide (PO) and phenol. The study led to the conclusion that low capital and operating costs and shipping costs to the Pacific Rim represent advantages to the development of propylene derivatives in the future in Alberta. 4 refs., 87 tabs., 7 figs

  14. Mongolia - Vocational Education - Equipment Upgrades

    Data.gov (United States)

    Millennium Challenge Corporation — Evaluation design The impact evaluation sought to identify the causal impact of exposure to equipment upgrades on subsequent outcomes. Insofar as we were not able to...

  15. Mining Upgrades to Reduce Pollution

    Science.gov (United States)

    Settlement with Southern Coal Corporation and 26 affiliates requires the companies to comprehensively upgrade their coal mining and processing operations to prevent polluted wastewater from threatening rivers and streams and communities across Appalachia.

  16. Energy losses on tokamak startup

    International Nuclear Information System (INIS)

    Murray, J.G.; Rothe, K.E.; Bronner, G.

    1983-01-01

    During the startup of a tokamak reactor using poloidal field (PF) coils to induce plasma currents, the conducting structures carry induced currents. The associated energy losses in the circuits must be provided by the startup coils and the PF system. This paper provides quantitative and comparitive values for the energies required as a function of the thickness or resistivity of the torus shells

  17. Prospects for Tokamak Fusion Reactors

    International Nuclear Information System (INIS)

    Sheffield, J.; Galambos, J.

    1995-01-01

    This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant

  18. Integral torque balance in tokamaks

    International Nuclear Information System (INIS)

    Pustovitov, V.D.

    2011-01-01

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

  19. ECRH Studies on Tokamak Plasmas.

    Science.gov (United States)

    1980-10-10

    r.I*cru.Dtrtibution uUnliited 300 Unicorn Pork Drive Woburn, Massachusetts 04801 ECRH STUDIES ON TOKAMAK PLASMAS JAYCOR Project No. 6183 Final Report...up techniques now in use or being suggested, include growing the plasma from a small minor radius or applying a negative voltage spike immediately

  20. Tokamak and RFP ignition requirements

    International Nuclear Information System (INIS)

    Werley, K.A.

    1991-01-01

    A plasma model is applied to calculate numerically transport- confinement (nτ E ) requirements and steady-state operation tokamak. The CIT tokamak and RFP ignition conditions are examined. Physics differences between RFP and tokamaks, and their consequences for a DT ignition machine, are discussed. The ignition RFP, compared to a tokamak, has many physics advantages, including ohmic heating to ignition (no need for auxiliary heating systems), higher beta, low ignition current, less sensitivity of ignition requirements to impurity effects, no hard disruptions (associated with beta or density limits), and successful operation with high radiation fractions (f RAD ∼ 0.95). These physics advantages, coupled with important engineering advantages associated with lower external magnetic fields, larger aspect ratios, and smaller plasma cross sections translate into significant cost reductions for both ignition and power reactor. The primary drawback of the RFP is the uncertainty that the present confinement scaling will extrapolate to reactor regimes. The 4-MA ZTH was expected to extend the nτ E transport scaling data three order of magnitude above ZT-40M results, and if the present scaling held, to achieve a DT-equivalent scientific energy breakeven, Q=1. A basecase RFP ignition point is identified with a plasma current of 8.1 MA and no auxiliary heating. 16 refs., 4 figs., 1 tab

  1. Tokamak impurity-control techniques

    International Nuclear Information System (INIS)

    Schmidt, J.A.

    1980-01-01

    A brief review is given of the impurity-control functions in tokamaks, their relative merits and disadvantages and some prominent edge-interaction-control techniques, and there is a discussion of a new proposal, the particle scraper, and its potential advantages. (author)

  2. An enhanced tokamak startup model

    Science.gov (United States)

    Goswami, Rajiv; Artaud, Jean-François

    2017-01-01

    The startup of tokamaks has been examined in the past in varying degree of detail. This phase typically involves the burnthrough of impurities and the subsequent rampup of plasma current. A zero-dimensional (0D) model is most widely used where the time evolution of volume averaged quantities determines the detailed balance between the input and loss of particle and power. But, being a 0D setup, these studies do not take into consideration the co-evolution of plasma size and shape, and instead assume an unchanging minor and major radius. However, it is known that the plasma position and its minor radius can change appreciably as the plasma evolves in time to fill in the entire available volume. In this paper, an enhanced model for the tokamak startup is introduced, which for the first time takes into account the evolution of plasma geometry during this brief but highly dynamic period by including realistic one-dimensional (1D) effects within the broad 0D framework. In addition the effect of runaway electrons (REs) has also been incorporated. The paper demonstrates that the inclusion of plasma cross section evolution in conjunction with REs plays an important role in the formation and development of tokamak startup. The model is benchmarked against experimental results from ADITYA tokamak.

  3. Multimegawatt neutral beams for tokamaks

    International Nuclear Information System (INIS)

    Kunkel, W.B.

    1979-03-01

    Most of the large magnetic confinement experiments today and in the near future use high-power neutral-beam injectors to heat the plasma. This review briefly describes this remarkable technique and summarizes recent results as well as near term expectations. Progress has been so encouraging that it seems probable that tokamaks will achieve scientific breakeven before 1990

  4. Joint research using small tokamaks

    Czech Academy of Sciences Publication Activity Database

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

    2005-01-01

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

  5. Application of AXUV diode detectors at ASDEX Upgrade

    Science.gov (United States)

    Bernert, M.; Eich, T.; Burckhart, A.; Fuchs, J. C.; Giannone, L.; Kallenbach, A.; McDermott, R. M.; Sieglin, B.

    2014-03-01

    In the ASDEX Upgrade tokamak, a radiation measurement for a wide spectral range, based on semiconductor detectors, with 256 lines of sight and a time resolution of 5μs was recently installed. In combination with the foil based bolometry, it is now possible to estimate the absolutely calibrated radiated power of the plasma on fast timescales. This work introduces this diagnostic based on AXUV (Absolute eXtended UltraViolet) n-on-p diodes made by International Radiation Detectors, Inc. The measurement and the degradation of the diodes in a tokamak environment is shown. Even though the AXUV diodes are developed to have a constant sensitivity for all photon energies (1 eV-8 keV), degradation leads to a photon energy dependence of the sensitivity. The foil bolometry, which is restricted to a time resolution of less than 1 kHz, offers a basis for a time dependent calibration of the diodes. The measurements of the quasi-calibrated diodes are compared with the foil bolometry and found to be accurate on the kHz time scale. Therefore, it is assumed, that the corrected values are also valid for the highest time resolution (200 kHz). With this improved diagnostic setup, the radiation induced by edge localized modes is analyzed on fast timescales.

  6. Application of AXUV diode detectors at ASDEX Upgrade

    International Nuclear Information System (INIS)

    Bernert, M.; Eich, T.; Burckhart, A.; Fuchs, J. C.; Giannone, L.; Kallenbach, A.; McDermott, R. M.; Sieglin, B.

    2014-01-01

    In the ASDEX Upgrade tokamak, a radiation measurement for a wide spectral range, based on semiconductor detectors, with 256 lines of sight and a time resolution of 5μs was recently installed. In combination with the foil based bolometry, it is now possible to estimate the absolutely calibrated radiated power of the plasma on fast timescales. This work introduces this diagnostic based on AXUV (Absolute eXtended UltraViolet) n-on-p diodes made by International Radiation Detectors, Inc. The measurement and the degradation of the diodes in a tokamak environment is shown. Even though the AXUV diodes are developed to have a constant sensitivity for all photon energies (1 eV-8 keV), degradation leads to a photon energy dependence of the sensitivity. The foil bolometry, which is restricted to a time resolution of less than 1 kHz, offers a basis for a time dependent calibration of the diodes. The measurements of the quasi-calibrated diodes are compared with the foil bolometry and found to be accurate on the kHz time scale. Therefore, it is assumed, that the corrected values are also valid for the highest time resolution (200 kHz). With this improved diagnostic setup, the radiation induced by edge localized modes is analyzed on fast timescales

  7. MHD Effects of a Ferritic Wall on Tokamak Plasmas

    Science.gov (United States)

    Hughes, Paul E.

    It has been recognized for some time that the very high fluence of fast (14.1MeV) neutrons produced by deuterium-tritium fusion will represent a major materials challenge for the development of next-generation fusion energy projects such as a fusion component test facility and demonstration fusion power reactor. The best-understood and most promising solutions presently available are a family of low-activation steels originally developed for use in fission reactors, but the ferromagnetic properties of these steels represent a danger to plasma confinement through enhancement of magnetohydrodynamic instabilities and increased susceptibility to error fields. At present, experimental research into the effects of ferromagnetic materials on MHD stability in toroidal geometry has been confined to demonstrating that it is still possible to operate an advanced tokamak in the presence of ferromagnetic components. In order to better quantify the effects of ferromagnetic materials on tokamak plasma stability, a new ferritic wall has been installated in the High Beta Tokamak---Extended Pulse (HBT-EP) device. The development, assembly, installation, and testing of this wall as a modular upgrade is described, and the effect of the wall on machine performance is characterized. Comparative studies of plasma dynamics with the ferritic wall close-fitting against similar plasmas with the ferritic wall retracted demonstrate substantial effects on plasma stability. Resonant magnetic perturbations (RMPs) are applied, demonstrating a 50% increase in n = 1 plasma response amplitude when the ferritic wall is near the plasma. Susceptibility of plasmas to disruption events increases by a factor of 2 or more with the ferritic wall inserted, as disruptions are observed earlier with greater frequency. Growth rates of external kink instabilities are observed to be twice as large in the presence of a close-fitting ferritic wall. Initial studies are made of the influence of mode rotation frequency

  8. Silicon Tracking Upgrade at CDF

    International Nuclear Information System (INIS)

    Kruse, M.C.

    1998-04-01

    The Collider Detector at Fermilab (CDF) is scheduled to begin recording data from Run II of the Fermilab Tevatron in early 2000. The silicon tracking upgrade constitutes both the upgrade to the CDF silicon vertex detector (SVX II) and the new Intermediate Silicon Layers (ISL) located at radii just beyond the SVX II. Here we review the design and prototyping of all aspects of these detectors including mechanical design, data acquisition, and a trigger based on silicon tracking

  9. WNP-2 core model upgrade

    International Nuclear Information System (INIS)

    Golightly, C.E.; Ravindranath, T.K.; Belblidia, L.A.; O'Farrell, D.; Andersen, P.S.

    2006-01-01

    The paper describes the core model upgrade of the WNP-2 training simulator and the reasons for the upgrade. The core model as well as the interface with the rest of the simulator are briefly described . The paper also describes the procedure that will be used by WNP-2 to update the simulator core data after future core reloads. Results from the fully integrated simulator are presented. (author)

  10. Impact of magnetic perturbation fields on tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Fietz, Sina; Maraschek, Marc; Suttrop, Wolfgang; Zohm, Hartmut [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Classen, Ivo [FOM-Institute DIFFER, Nieuwegein (Netherlands); Collaboration: the ASDEX Upgrade Team

    2015-05-01

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

  11. Investigation of magnetic modes in the ASDEX tokamak

    International Nuclear Information System (INIS)

    Zohm, H.

    1990-10-01

    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)

  12. Bayesian fuzzy logic-based estimation of electron cyclotron heating (ECH) power deposition in MHD control systems

    Energy Technology Data Exchange (ETDEWEB)

    Davoudi, Mehdi, E-mail: mehdi.davoudi@polimi.it [Department of Electrical and Computer Engineering, Buein Zahra Technical University, Buein Zahra, Qazvin (Iran, Islamic Republic of); Davoudi, Mohsen, E-mail: davoudi@eng.ikiu.ac.ir [Department of Electrical Engineering, Imam Khomeini International University, Qazvin, 34148-96818 (Iran, Islamic Republic of)

    2017-06-15

    Highlights: • A couple of algorithms to diagnose if Electron Cyclotron Heating (ECH) power is deposited properly on the expected deposition minor radius are proposed. • The algorithms are based on Bayesian theory and Fuzzy logic. • The algorithms are tested on the off-line experimental data acquired from Frascati Tokamak Upgrade (FTU), Frascati, Italy. • Uncertainties and evidences derived from the combination of online information formed by the measured diagnostic data and the prior information are also estimated. - Abstract: In the thermonuclear fusion systems, the new plasma control systems use some measured on-line information acquired from different sensors and prior information obtained by predictive plasma models in order to stabilize magnetic hydro dynamics (MHD) activity in a tokamak. Suppression of plasma instabilities is a key issue to improve the confinement time of controlled thermonuclear fusion with tokamaks. This paper proposes a couple of algorithms based on Bayesian theory and Fuzzy logic to diagnose if Electron Cyclotron Heating (ECH) power is deposited properly on the expected deposition minor radius (r{sub DEP}). Both algorithms also estimate uncertainties and evidences derived from the combination of the online information formed by the measured diagnostic data and the prior information. The algorithms have been employed on a set of off-line ECE channels data which have been acquired from the experimental shot number 21364 at Frascati Tokamak Upgrade (FTU), Frascati, Italy.

  13. Setting priorities for safeguards upgrades

    International Nuclear Information System (INIS)

    Al-Ayat, R.A.; Judd, B.R.; Patenaude, C.J.; Sicherman, A.

    1987-01-01

    This paper describes an analytic approach and a computer program for setting priorities among safeguards upgrades. The approach provides safeguards decision makers with a systematic method for allocating their limited upgrade resources. The priorities are set based on the upgrades cost and their contribution to safeguards effectiveness. Safeguards effectiveness is measured by the probability of defeat for a spectrum of potential insider and outsider adversaries. The computer program, MI$ER, can be used alone or as a companion to ET and SAVI, programs designed to evaluate safeguards effectiveness against insider and outsider threats, respectively. Setting the priority required judgments about the relative importance (threat likelihoods and consequences) of insider and outsider threats. Although these judgments are inherently subjective, MI$ER can analyze the sensitivity of the upgrade priorities to these weights and determine whether or not they are critical to the priority ranking. MI$ER produces tabular and graphical results for comparing benefits and identifying the most cost-effective upgrades for a given expenditure. This framework provides decision makers with an explicit and consistent analysis to support their upgrades decisions and to allocate the safeguards resources in a cost-effective manner

  14. Development and Operational Experiences of the JT-60U Tokamak and Power Supplies

    International Nuclear Information System (INIS)

    Hosogane, N.; Ninomiya, H.; Matsukawa, M.; Ando, T.; Neyatani, Y.; Horiike, H.; Sakurai, S.; Masaki, K.; Yamamoto, M.; Kodama, K.; Sasajima, T.; Terakado, T.; Ohmori, S.; Ohmori, Y.; Okano, J.

    2002-01-01

    The design of the JT-60U tokamak, the configuration of the coil power supplies, and the operational experiences gained to date are reviewed. JT-60U is a large tokamak upgraded from the original JT-60 in order to obtain high plasma current, large plasma volume, and highly elongated divertor configurations. All components inside the toroidal magnetic field coils, such as vacuum vessel, poloidal magnetic field coils, divertor, etc., were modified. Various technologies and ideas were introduced to develop these components; for example, a multi-arc double skin wall structure for the vacuum vessel and a functional poloidal magnetic field coil system with taps for obtaining various plasma configurations. Furthermore, boron-carbide coated carbon fiber composite (CFC) tiles were used as divertor tiles to reduce erosion of carbon-base tiles. Later, a semiclosed divertor with pumps, for which cryo-panels originally used for NBI units were converted, was installed in the replacement of the open divertor. These development and operational results provide data for future tokamaks. Major failures experienced in the long operational period of JT-60U, such as water leakage from the toroidal magnetic field coil, fracture of carbon tiles, and breakdown of a filter capacitor, are described. As a maintenance issue for tokamaks using deuterium fueling gas, a method for reducing radiation exposure of in-vessel workers is described

  15. Improvements in disruption prediction at ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Aledda, R., E-mail: raffaele.aledda@diee.unica.it; Cannas, B., E-mail: cannas@diee.unica.it; Fanni, A., E-mail: fanni@diee.unica.it; Pau, A., E-mail: alessandro.pau@diee.unica.it; Sias, G., E-mail: giuliana.sias@diee.unica.it

    2015-10-15

    Highlights: • A disruption prediction system for AUG, based on a logistic model, is designed. • The length of the disruptive phase is set for each disruption in the training set. • The model is tested on dataset different from that used during the training phase. • The generalization capability and the aging of the model have been tested. • The predictor performance is compared with the locked mode detector. - Abstract: In large-scale tokamaks disruptions have the potential to create serious damage to the facility. Hence disruptions must be avoided, but, when a disruption is unavoidable, minimizing its severity is mandatory. A reliable detection of a disruptive event is required to trigger proper mitigation actions. To this purpose machine learning methods have been widely studied to design disruption prediction systems at ASDEX Upgrade. The training phase of the proposed approaches is based on the availability of disrupted and non-disrupted discharges. In literature disruptive configurations were assumed appearing into the last 45 ms of each disruption. Even if the achieved results in terms of correct predictions were good, it has to be highlighted that the choice of such a fixed temporal window might have limited the prediction performance. In fact, it generates confusing information in cases of disruptions with disruptive phase different from 45 ms. The assessment of a specific disruptive phase for each disruptive discharge represents a relevant issue in understanding the disruptive events. In this paper, the Mahalanobis distance is applied to define a specific disruptive phase for each disruption, and a logistic regressor has been trained as disruption predictor. The results show that enhancements on the achieved performance on disruption prediction are possible by defining a specific disruptive phase for each disruption.

  16. Improvements in disruption prediction at ASDEX Upgrade

    International Nuclear Information System (INIS)

    Aledda, R.; Cannas, B.; Fanni, A.; Pau, A.; Sias, G.

    2015-01-01

    Highlights: • A disruption prediction system for AUG, based on a logistic model, is designed. • The length of the disruptive phase is set for each disruption in the training set. • The model is tested on dataset different from that used during the training phase. • The generalization capability and the aging of the model have been tested. • The predictor performance is compared with the locked mode detector. - Abstract: In large-scale tokamaks disruptions have the potential to create serious damage to the facility. Hence disruptions must be avoided, but, when a disruption is unavoidable, minimizing its severity is mandatory. A reliable detection of a disruptive event is required to trigger proper mitigation actions. To this purpose machine learning methods have been widely studied to design disruption prediction systems at ASDEX Upgrade. The training phase of the proposed approaches is based on the availability of disrupted and non-disrupted discharges. In literature disruptive configurations were assumed appearing into the last 45 ms of each disruption. Even if the achieved results in terms of correct predictions were good, it has to be highlighted that the choice of such a fixed temporal window might have limited the prediction performance. In fact, it generates confusing information in cases of disruptions with disruptive phase different from 45 ms. The assessment of a specific disruptive phase for each disruptive discharge represents a relevant issue in understanding the disruptive events. In this paper, the Mahalanobis distance is applied to define a specific disruptive phase for each disruption, and a logistic regressor has been trained as disruption predictor. The results show that enhancements on the achieved performance on disruption prediction are possible by defining a specific disruptive phase for each disruption.

  17. Upgrading inflatable door seals

    International Nuclear Information System (INIS)

    Sykes, T.M.; Metcalfe, R.; Welch, L.A.; Josefowich, J.M.

    1997-01-01

    Inflatable door seals are used for airlocks in CANDU stations. They have been a significant source of unreliability and maintenance cost. A program is underway to improve their performance and reliability, backed by environmental qualification testing. Only commercial products and suppliers existed in 1993. For historical reasons, these 'existing products' did not use the most durable material then available. In hindsight, neither had they been adapted nor optimized to combat conditions often experienced in the plants-sagging doors, damaged sealing surfaces, and many thousands of openings and closings per year. Initial attempts to involve the two existing suppliers in efforts to upgrade these seals were unsuccessful. Another suitable supplier had therefore to be found, and a 'new,' COG-owned seal developed; this was completed in 1997. This paper summarizes its testing, along with that of the two existing products. Resistance to aging has been improved significantly. Testing has shown that an accident can be safely withstood after 10 years of service or 40,000 openings-closings, whichever comes first. AECL's Fluid Sealing Technology Unit (FSTU) has invested in the special moulds, test fixtures and other necessary tooling and documentation required to begin commercial manufacture of this new quality product. Accordingly, as with FSTU's other nuclear products such as pump seals, the long-term supply of door seals to CANDU plants is now protected from many external uncertainties-e.g., commercial products being discontinued, materials being changed, companies going out of business. Manufacturing to AECL's detailed specifications is being subcontracted to the new supplier. FSTU is performing the quality surveillance, inspection, testing, and customer service activities concomitant with direct responsibility for supply to the plants. (author)

  18. Upgradation in SCADA and PLC of existing LN_2 control system for SST-1

    International Nuclear Information System (INIS)

    Panchal, Pradip; Mahesuria, Gaurang; Panchal, Rohit; Patel, Rakesh; Sonara, Dashrath; Pitroda, Dipen; Nimavat, Hiren; Tanna, Vipul; Pradhan, Subrata

    2016-01-01

    Highlights: • The control system of LN_2 Management System of SST-1 is designed on PLC and SCADA. • The implementation and results of up-gradation in PLC and SCADA are reported. • The up-gradation in PLC and SCADA has improved the reliability & availability of SST-1 LN_2 system. - Abstract: Helium Refrigerator/Liquefier system of Steady State Superconducting Tokamak (SST-1) incorporates Liquid Nitrogen (LN_2) pre-cooling system. LN_2 is used for 80 K thermal shields of SST-1, current feeder system and integrated flow distribution and control system. The LN_2 management system is distributed system and requires automatic control. Initially LN_2 control system had Citect based Supervisory Control and Data Acquisition (SCADA) and Koyo make Programmable Logic Controller (PLC). With the passage of time and due to unavailability of their hardware, it is being obsoleted. So, the requirements of new PLC and SCADA systems have been envisaged to make uninterruptable operation of SST-1 cryogenic system. Therefore, Wonderware SCADA and Schneider Electric make PLC is programmed to replace Citect SCADA and Koyo PLC. New control features have been added in upgraded control system for better management of LN_2 system. This upgradation of SCADA and PLC is completed, tested successfully and in operation. Operational performance highlights of the new upgraded system are presented in this paper.

  19. Upgradation in SCADA and PLC of existing LN{sub 2} control system for SST-1

    Energy Technology Data Exchange (ETDEWEB)

    Panchal, Pradip, E-mail: pradip@ipr.res.in; Mahesuria, Gaurang; Panchal, Rohit; Patel, Rakesh; Sonara, Dashrath; Pitroda, Dipen; Nimavat, Hiren; Tanna, Vipul; Pradhan, Subrata

    2016-11-15

    Highlights: • The control system of LN{sub 2} Management System of SST-1 is designed on PLC and SCADA. • The implementation and results of up-gradation in PLC and SCADA are reported. • The up-gradation in PLC and SCADA has improved the reliability & availability of SST-1 LN{sub 2} system. - Abstract: Helium Refrigerator/Liquefier system of Steady State Superconducting Tokamak (SST-1) incorporates Liquid Nitrogen (LN{sub 2}) pre-cooling system. LN{sub 2} is used for 80 K thermal shields of SST-1, current feeder system and integrated flow distribution and control system. The LN{sub 2} management system is distributed system and requires automatic control. Initially LN{sub 2} control system had Citect based Supervisory Control and Data Acquisition (SCADA) and Koyo make Programmable Logic Controller (PLC). With the passage of time and due to unavailability of their hardware, it is being obsoleted. So, the requirements of new PLC and SCADA systems have been envisaged to make uninterruptable operation of SST-1 cryogenic system. Therefore, Wonderware SCADA and Schneider Electric make PLC is programmed to replace Citect SCADA and Koyo PLC. New control features have been added in upgraded control system for better management of LN{sub 2} system. This upgradation of SCADA and PLC is completed, tested successfully and in operation. Operational performance highlights of the new upgraded system are presented in this paper.

  20. Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

    International Nuclear Information System (INIS)

    Riedel, K.S.

    1989-01-01

    This paper is an expository introduction to advanced statistics and scaling laws and their application to tokamak devices. Topics of discussion are as follows: implicit assumptions in the standard analysis; advanced regression techniques; specialized tools in statistics and their applications in fusion physics; and improved datasets for transport studies

  1. High power electron cyclotron heating in ISX and ORMAK Upgrade at ORNL

    International Nuclear Information System (INIS)

    England, A.C.; Eldridge, O.C.; Marcus, F.B.; Sprott, J.C.; Namkung, W.; Wilgen, J.B.

    1976-05-01

    A phased program of plasma heating at the electron cyclotron frequency is proposed for the Oak Ridge tokamaks ISX and ORMAK Upgrade. The past history of the program of electron cyclotron heating (ECH) at ORNL on mirrors and in the ELMO Bumpy Torus has been successful. Future technological developments in the production of high power high frequency microwave tubes look promising at this time. The physics of wave propagation and particle heating are fairly well understood and indicate the viability of this technique. Studies on breakdown and on runaway electron reduction will provide useful information for larger machines. Recent experiments in the USSR on small tokamaks have shown that ECH is a viable heating technique. Providing that the microwave tubes become available, the engineering considerations suggest that the technique is practical and workable, based on present day technology

  2. Upgrade and benchmarking of the NIFS physics-engineering-cost code

    International Nuclear Information System (INIS)

    Dolan, T.J.; Yamazaki, K.

    2004-07-01

    The NIFS Physics-Engineering-Cost (PEC) code for helical and tokamak fusion reactors is upgraded by adding data from three blanket-shield designs, a new cost section based on the ARIES cost schedule, more recent unit costs, and improved algorithms for various computations. The PEC code is also benchmarked by modeling the ARIES-AT (advanced technology) tokamak and the ARIES-SPPS (stellarator power plant system). The PEC code succeeds in predicting many of the pertinent plasma parameters and reactor component masses within about 10%. There are cost differences greater than 10% for some fusion power core components, which may be attributed to differences of unit costs used by the codes. The COEs estimated by the PEC code differ from the COEs of the ARIES-AT and ARIES-SPPS studies by 5%. (author)

  3. Track-mounted remote handling system for the Tokamak Fusion Engineering Test

    International Nuclear Information System (INIS)

    Kelly, V.P.; Berger, J.D.; Daubert, R.L.; Yount, J.A.

    1982-01-01

    Concepts for remote handling machines (IVM) designed to transverse the interior of toroidal vessels with guidance and support from track systems have been developed for the proposed Tokamak Fusion Engineering Test (TFET). TFET has been proposed as an upgrade for the Tokamak Fusion Test Reactor (TFTR), currently nearing completion. The track-mounted IVMs were conceived to perform in-vessel remote maintenance for TFET, including removal and replacement of pump limiter blades and protective tiles as well as other maintenance-related tasks such as vessel wall inspection leak testing and interior cleanup. The conceptual IVMs consist of three manipulator arms mounted on a common frame member: a single power manipulator arm with high load carrying capacity and two lower-capacity servomanipulator arms. Descriptions of the IVM concepts, in-vessel track systems, and ex-vessel handling systems are presented

  4. Characterization of axisymmetric disruption dynamics toward VDE avoidance in tokamaks

    International Nuclear Information System (INIS)

    Nakamura, Y.

    2002-01-01

    Disruption experiments on Alcator C-Mod and ASDEX-Upgrade tokamaks and axisymmetric MHD simulations using the TSC have explicated the underlying mechanisms of Vertical Displacement Events (VDEs) and a diversity of disruption dynamics. First, the neutral point, which is known as an initial vertical plasma position advantageous to VDE avoidance, is shown to be fairly insensitive to plasma shape and current profile parameters, while the VDE rate significantly depends on those parameters. Secondly, it is clarified that a rapid flattening of the plasma current profile frequently seen at the thermal quench drags a single null-diverted, up-down asymmetric plasma vertically toward divertor, whereas the dragging effect is absent in up-down symmetric limiter discharges. As a consequence, the occurrence of downward-going VDEs predominates over the upward-going ones in bottom-diverted discharges, being consistent with experiments in ASDEX-Upgrade. Together with the attractive force that arises from passive shell currents induced by the current quench and vanishes at the neutral point, the dragging effect explains many details of the VDE dynamics over the whole period of disruptive termination. (author)

  5. Simulation of enhanced tokamak performance on DIII-D using fast wave current drive

    International Nuclear Information System (INIS)

    Grassie, J.S. de; Lin-Liu, Y.R.; Petty, C.C.; Pinsker, R.I.; Chan, V.S.; Prater, R.; John, H. St.; Baity, F.W.; Goulding, R.H.; Hoffman, D.H.

    1993-01-01

    The fast magnetosonic wave is now recognized to be a leading candidate for noninductive current drive for the tokamak reactor due to the ability of the wave to penetrate to the hot dense core region. Fast wave current drive (FWCD) experiments on DIII-D have realized up to 120 kA of rf current drive, with up to 40% of the plasma current driven noninductively. The success of these experiments at 60 MHz with a 2 MW transmitter source capability has led to a major upgrade of the FWCD system. Two additional transmitters, 30 to 120 MHz, with a 2 MW source capability each, will be added together with two new four-strap antennas in early 1994. Another major thrust of the DIII-D program is to develop advanced tokamak modes of operation, simultaneously demonstrating improvements in confinement and stability in quasi-steady-state operation. In some of the initial advanced tokamak experiments on DIII-D with neutral beam heated (NBI) discharges it has been demonstrated that energy confinement time can be improved by rapidly elongating the plasma to force the current density profile to be more centrally peaked. However, this high-l i phase of the discharge with the commensurate improvement in confinement is transient as the current density profile relaxes. By applying FWCD to the core of such a κ-ramped discharge it may be possible to sustain the high internal inductance and elevated confinement. Using computational tools validated on the initial DIII-D FWCD experiments we find that such a high-l i advanced tokamak discharge should be capable of sustainment at the 1 MA level with the upgraded capability of the FWCD system. (author) 16 refs., 3 figs., 1 tab

  6. Microwave Tokamak Experiment: Overview and status

    International Nuclear Information System (INIS)

    1990-05-01

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

  7. Combined confinement system applied to tokamaks

    International Nuclear Information System (INIS)

    Ohkawa, Tihiro

    1986-01-01

    From particle orbit point of view, a tokamak is a combined confinement configuration where a closed toroidal volume is surrounded by an open confinement system like a magnetic mirror. By eliminating a cold halo plasma, the energy loss from the plasma becomes convective. The H-mode in diverted tokamaks is an example. Because of the favorable scaling of the energy confinement time with temperature, the performance of the tokamak may be significantly improved by taking advantage of this effect. (author)

  8. Presheath profiles in simulated tokamak edge plasmas

    International Nuclear Information System (INIS)

    LaBombard, B.; Conn, R.W.; Hirooka, Y.; Lehmer, R.; Leung, W.K.; Nygren, R.E.; Ra, Y.; Tynan, G.

    1988-04-01

    The PISCES plasma surface interaction facility at UCLA generates plasmas with characteristics similar to those found in the edge plasmas of tokamaks. Steady state magnetized plasmas produced by this device are used to study plasma-wall interaction phenomena which are relevant to tokamak devices. We report here progress on some detailed investigations of the presheath region that extends from a wall surface into these /open quotes/simulated tokamak/close quotes/ edge plasma discharges along magnetic field lines

  9. Project scenarios for bitumen upgrading

    International Nuclear Information System (INIS)

    Koppel, P.E.; Mazurek, W.L.; Harji, A.

    2002-01-01

    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

  10. Processing options for bitumen upgrading

    International Nuclear Information System (INIS)

    Harji, A.N.; Koppel, P.E.; Mazurek, W.L.; Meysami, P.

    2003-01-01

    It is estimated that 178 billion barrels of oil can be recovered from Alberta's vast heavy oil reserves. The challenge lies in the logistics of recovering, upgrading and transporting the oil to market. The Canadian Energy Research Institute conducted a recent study to determine market potential by 2007 for diluted bitumen and synthetic crude oil produced from upgraded bitumen. The viability for a wide range of bitumen feedstocks was assessed along with the sensitivity of refinery demand to their prices. The 3 major markets for western Canadian bitumen include PADD 2 in the United States, western Canada, and Ontario. Bitumen is too viscous to transport by pipeline and cannot be processed by most of the existing refineries. Therefore, in order to develop a mass market for the product, bitumen must undergo the energy intensive upgrading process at existing refineries. The factors impacting which method of upgrading is most suitable were discussed with particular attention to the impact that Canada's ratification of the Kyoto Protocol may have on Alberta's bitumen resource in terms of costs of complying with greenhouse gas reduction initiatives. The authors emphasized that it is crucial to customize an upgrading project to meet site and market specific factors. 8 refs., 3 tabs., 3 figs

  11. Numerical Simulation of Neoclassical Currents, Parallel Viscosity, and Radial Current Balance in Tokamak Plasmas

    International Nuclear Information System (INIS)

    Kiviniemi, T.

    2001-01-01

    One of the principal problems en route to a fusion reactor is that of insufficient plasma confinement, which has lead to both theoretical and experimental research into transport processes in the parameter range relevant for fusion energy production. The neoclassical theory of tokamak transport is well-established unlike the theory of turbulence driven anomalous transport in which extensive progress has been made during last few years. So far, anomalous transport has been dominant in experiments, but transport may be reduced to the neoclassical level in advanced tokamak scenarios. This thesis reports a numerical study of neoclassical fluxes, parallel viscosity, and neoclassical radial current balance in tokamaks. Neoclassical parallel viscosity and particle fluxes are simulated over a wide range of collisionalities, using the fully kinetic five-dimensional neoclassical orbit-following Monte Carlo code ASCOT. The qualitative behavior of parallel viscosity derived in earlier analytic models is shown to be incorrect for high poloidal Mach numbers. This is because the poloidal dependence of density was neglected. However, in high Mach number regime, it is the convection and compression terms, rather than the parallel viscosity term, that are shown to dominate the momentum balance. For fluxes, a reasonable agreement between numerical and analytical results is found in the collisional parameter regime. Neoclassical particle fluxes are additionally studied in the banana regime using the three-dimensional Fokker-Planck code DEPORA, which solves the drift-kinetic equation with finite differencing. Limitations of the small inverse aspect ratio approximation adopted in the analytic theory are addressed. Assuming that the anomalous transport is ambipolar, the radial electric field and its shear at the tokamak plasma edge can be solved from the neoclassical radial current balance. This is performed both for JET and ASDEX Upgrade tokamaks using the ASCOT code. It is shown that

  12. Improvement of the tokamak concept

    Energy Technology Data Exchange (ETDEWEB)

    Laurent, L

    1994-12-31

    Improvement of the tokamak concept is highly desirable to reduce the size and capital cost of a device able to ignite to increase the plasma pressure, i.e. the power density to reduce the cost of electricity, and to increase the fraction of bootstrap current to render the tokamak compatible with continuous operation. The most important results obtained in this field are summarized, and the options are shown which are still open and explored by the various experiments. Various effects of the plasma shaping are discussed, plasma configurations with both high {beta}{sub N} and H{sub G} are explored, and the issues of stable steady state and of the plasma edge are briefly discussed. (R.P.). 65 refs., 2 tabs.

  13. Advanced commercial Tokamak optimization studies

    International Nuclear Information System (INIS)

    Whitley, R.H.; Berwald, D.H.; Gordon, J.D.

    1985-01-01

    Our recent studies have concentrated on developing optimal high beta (bean-shaped plasma) commercial tokamak configurations using TRW's Tokamak Reactor Systems Code (TRSC) with special emphasis on lower net electric power reactors that are more easily deployable. A wide range of issues were investigated in the search for the most economic configuration: fusion power, reactor size, wall load, magnet type, inboard blanket and shield thickness, plasma aspect ratio, and operational β value. The costs and configurations of both steady-state and pulsed reactors were also investigated. Optimal small and large reactor concepts were developed and compared by studying the cost of electricity from single units and from multiplexed units. Multiplexed units appear to have advantages because they share some plant equipment and have lower initial capital investment as compared to larger single units

  14. Flux driven turbulence in tokamaks

    International Nuclear Information System (INIS)

    Garbet, X.; Ghendrih, P.; Ottaviani, M.; Sarazin, Y.; Beyer, P.; Benkadda, S.; Waltz, R.E.

    1999-01-01

    This work deals with tokamak plasma turbulence in the case where fluxes are fixed and profiles are allowed to fluctuate. These systems are intermittent. In particular, radially propagating fronts, are usually observed over a broad range of time and spatial scales. The existence of these fronts provide a way to understand the fast transport events sometimes observed in tokamaks. It is also shown that the confinement scaling law can still be of the gyroBohm type in spite of these large scale transport events. Some departure from the gyroBohm prediction is observed at low flux, i.e. when the gradients are close to the instability threshold. Finally, it is found that the diffusivity is not the same for a turbulence calculated at fixed flux than at fixed temperature gradient, with the same time averaged profile. (author)

  15. Options for an ignited tokamak

    International Nuclear Information System (INIS)

    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 β/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

  16. Plasma diagnostics on large tokamaks

    International Nuclear Information System (INIS)

    Orlinskij, D.V.; Magyar, G.

    1988-01-01

    The main tasks of the large tokamaks which are under construction (T-15 and Tore Supra) and of those which have already been built (TFTR, JET, JT-60 and DIII-D) together with their design features which are relevant to plasma diagnostics are briefly discussed. The structural features and principal characteristics of the diagnostic systems being developed or already being used on these devices are also examined. The different diagnostic methods are described according to the physical quantities to be measured: electric and magnetic diagnostics, measurements of electron density, electron temperature, the ion components of the plasma, radiation loss measurements, spectroscopy of impurities, edge diagnostics and study of plasma stability. The main parameters of the various diagnostic systems used on the six large tokamaks are summarized in tables. (author). 351 refs, 44 figs, 22 tabs

  17. Starfire: a commercial tokamak reactor

    International Nuclear Information System (INIS)

    Baker, C.C.; Abdou, M.A.; DeFreece, D.A.; Trachsel, C.A.; Graumann, D.; Kokoszenski, J.

    1979-01-01

    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. The STARFIRE Project was initiated in May 1979, with the goal of completing the design study by October 1980. The purpose of this paper is to present an overview of the major parameters and design features that have been tentatively selected for STARFIRE

  18. Comprehensive numerical modelling of tokamaks

    International Nuclear Information System (INIS)

    Cohen, R.H.; Cohen, B.I.; Dubois, P.F.

    1991-01-01

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

  19. Overview of ASDEX Upgrade results

    Czech Academy of Sciences Publication Activity Database

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

    2013-01-01

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

  20. Magnetic island formation in tokamaks

    International Nuclear Information System (INIS)

    Yoshikawa, S.

    1989-04-01

    The size of a magnetic island created by a perturbing helical field in a tokamak is estimated. A helical equilibrium of a current- carrying plasma is found in a helical coordinate and the helically flowing current in the cylinder that borders the plasma is calculated. From that solution, it is concluded that the helical perturbation of /approximately/10/sup /minus/4/ of the total plasma current is sufficient to cause an island width of approximately 5% of the plasma radius. 6 refs

  1. Equilibrium Reconstruction in EAST Tokamak

    International Nuclear Information System (INIS)

    Qian Jinping; Wan Baonian; Shen Biao; Sun Youwen; Liu Dongmei; Xiao Bingjia; Ren Qilong; Gong Xianzu; Li Jiangang; Lao, L. L.; Sabbagh, S. A.

    2009-01-01

    Reconstruction of experimental axisymmetric equilibria is an important part of tokamak 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 expansion 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. (magnetically confined plasma)

  2. Relaxed states of tokamak plasmas

    International Nuclear Information System (INIS)

    Kucinski, M.Y.; Okano, V.

    1993-01-01

    The relaxed states of tokamak plasmas are studied. It is assumed that the plasma relaxes to a quasi-steady state which is characterized by a minimum entropy production rate, compatible with a number of prescribed conditions and pressure balance. A poloidal current arises naturally due to the anisotropic resistivity. The minimum entropy production theory is applied, assuming the pressure equilibrium as fundamental constraint on the final state. (L.C.J.A.)

  3. Runaway electrons during tokamak startup

    International Nuclear Information System (INIS)

    Sharma, A.S.; Jayakumar, R.

    1988-01-01

    Runaway electrons significantly affect the plasma and impurity evolution during tokamak startup. During its rise, a runaway pulse stores magnetic flux inductively; this is then released during the decay phase of the runaway pulse. This process affects plasma formation, current initiation and current buildup. Because of their relativistic velocities the runaway electrons have higher ionization and excitation rates than the plasma electrons. This leads to a significant modification of the impurity behaviour and consequently the plasma evolution. (author). 20 refs, 8 figs

  4. Minimum scaling laws in tokamaks

    International Nuclear Information System (INIS)

    Zhang, Y.Z.; Mahajan, S.M.

    1986-10-01

    Scaling laws governing anomalous electron transport in tokamaks with ohmic and/or auxiliary heating are derived using renormalized Vlasov-Ampere equations for low frequency electromagnetic microturbulence. It is also shown that for pure auxiliary heating (or when auxiliary heating power far exceeds the ohmic power), the energy confinement time scales as tau/sub E/ ∼ P/sub inj//sup -1/3/, where P/sub inj/ is the injected power

  5. Gyrosheath near the tokamak edge

    International Nuclear Information System (INIS)

    Hazeltine, R.D.; Xiao, H.; Valanju, P.M.

    1993-03-01

    A new model for the structure of the radial electric field profile in the edge during the H-mode is proposed. Charge separation caused by the difference between electron and ion gyromotion, or more importantly in a tokamak, the banana motion (halo effect) can self-consistently produce an electric dipole moment that causes the sheared radial electric field. The calculated results based on the model are consistent with D-III D and TEXTOR experimental results

  6. Tokamak plasma boundary layer model

    International Nuclear Information System (INIS)

    Volkov, T.F.; Kirillov, V.D.

    1983-01-01

    A model has been developed for the limiter layer and for the boundary region of the plasma column in a tokamak to facilitate analytic calculations of the thickness of the limiter layers, the profiles and boundary values of the temperature and the density under various conditions, and the difference between the electron and ion temperatures. This model can also be used to analyze the recycling of neutrals, the energy and particle losses to the wall and the limiter, and other characteristics

  7. Shear Alfven waves in tokamaks

    International Nuclear Information System (INIS)

    Kieras, C.E.

    1982-12-01

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

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

    Czech Academy of Sciences Publication Activity Database

    Horáček, Jan; Adámek, Jiří; Müller, H. W.; Seidl, J.; Nielsen, A.H.; Rohde, V.; Mehlmann, F.; Ionita, C.; Havlíčková, Eva

    2010-01-01

    Roč. 50, č. 10 (2010), s. 105001-105001 ISSN 0029-5515 R&D Projects: GA AV ČR KJB100430901; GA ČR GA202/09/1467; GA MŠk 7G09042 Institutional research plan: CEZ:AV0Z20430508 Keywords : Ball- pen probe * Langmuir probe * tokamak * plasma * SOL * turbulence * blob * interchange instability * ASDEX Upgrade Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.303, year: 2010 http://iopscience.iop.org/0029-5515/50/10/105001

  9. 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...

  10. 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...... potential fluctuations measured by a Langmuir probe are dominated by plasma electron temperature rather than potential. Spatial and temporal scales are found consistent with expectations based on interchange-driven turbulence. Conditionally averaged signals found for both potential and density are also...

  11. Intermittent transport across the scrape-off layer: latest results from ASDEX Upgrade

    Czech Academy of Sciences Publication Activity Database

    Kočan, M.; Müller, H.W.; Nold, B.; Lunt, T.; Adámek, Jiří; Allan, S.Y.; Bernert, M.; Conway, G.D.; de Marné, P.; Eich, T.; Elmore, S.; Gennrich, F.P.; Herrmann, A.; Horáček, Jan; Huang, Z.; Kallenbach, A.; Komm, Michael; Maraschek, M.; Mehlmann, F.; Müller, S.; Ribeiro, T.T.; Rohde, V.; Schrittwieser, R.; Scott, B.; Stroth, U.; Suttrop, W.; Wolfrum, E.

    2013-01-01

    Roč. 53, č. 7 (2013), 073047-073047 ISSN 0029-5515 R&D Projects: GA MŠk(CZ) LG11018; GA ČR(CZ) GAP205/12/2327; GA ČR GA202/09/1467 Institutional support: RVO:61389021 Keywords : ASDEX Upgrade scrape-off layer * plasma * tokamak * edge-localized mode (ELM) Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.243, year: 2013 http://iopscience.iop.org/0029-5515/53/7/073047/pdf/0029-5515_53_7_073047.pdf

  12. H-mode confinement properties close to the power threshold in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Ryter, F; Fuchs, J; Schneider, W; Sips, A; Staebler, A; Stober, J

    2008-01-01

    Confinement properties close to the H-mode power threshold are studied in the ASDEX Upgrade tokamak. The results show that good confinement can be obtained close to the threshold with Type-I ELMs. The existence of Type-I ELMs does not necessarily require the heating power to be higher than the H-Mode power threshold, but it requires collisionality to be low enough. At higher collisionality Type-III ELMs replace the Type-I ELMs and confinement time is reduced by about 20%

  13. Physical protection upgrades in Ukraine

    International Nuclear Information System (INIS)

    Djakov, A.

    1998-01-01

    The U.S. DOE is providing nuclear material safeguards assistance in both material control and accountability and in physical protection to several facilities in Ukraine. This paper summarizes the types of physical protection upgrades that have been or are presently being implemented at these facilities. These facilities include the Kiev Institute for Nuclear Research, Kharkov Institute of Physics and Technology, Sevastopol Institute of Nuclear Energy and Industry, and the South Ukraine Nuclear Power Plant. Typical upgrades include: hardening of storage areas; improvements in access control, intrusion detection, and CCTV assessment; central alarm station improvements; and implementation of new voice communication systems. Methods used to implement these upgrades and problems encountered are discussed. Training issues are also discussed

  14. The D0 detector upgrade

    International Nuclear Information System (INIS)

    Bross, A.D.

    1995-02-01

    The Fermilab collider program is undergoing a major upgrade of both the accelerator complex and the two detectors. Operation of the Tevatron at luminosities upwards of ten time that currently provided will occur in early 1999 after the commissioning of the new Fermilab Main Injector. The D0 upgrade program has been established to deliver a detector that will meet the challenges of this environment. A new magnetic tracker consisting of a superconducting solenoid, a silicon vertex detector, a scintillating fiber central tracker, and a central preshower detector will replace the current central tracking and transition radiation chambers. We present the design and performance capabilities of these new systems and describe results from physics simulations that demonstrate the physics reach of the upgraded detector

  15. Environmental upgrading of a landfill

    International Nuclear Information System (INIS)

    Agostinetto, V.; Vendrame, G.

    1999-01-01

    This article refers to an experimental study concerning the vegetative upgrading of a closed-down landfill (once used for industrial waste disposal). The aim was to check the possibility of reconstructing or aiding the natural growth of a vegetation in keeping with the surrounding area, in a tried environment such as that of landfills. The original idea contained in the approved project - which meant to generically upgrade the territory by planting species belonging to the grassy layer, shrubs and trees - has, with time, undergone some changes. On the basis of both the knowledge acquired during management and of a more accurate analysis of the territory, the experiment was preferred to aim at finding out which were the species, both continental and Mediterranean, able to gradually adjust to the surrounding landscape, leaving to natural selection the task to decide which species were more suitable to the upgrading of closed-down landfills, and which planting technique was more effective [it

  16. Hydrogen assisted biological biogas upgrading

    DEFF Research Database (Denmark)

    Bassani, Ilaria

    Wind and biomass are promoted worldwide as sustainable forms of energy. Anaerobic digestion of biomass produces biogas with ∼50−70% CH4 and 30−50% CO2. However, biogas with >90% CH4 content has higher heating value, can be injected into the natural gas grid or used as alternative to natural gas...... as vehicle fuel. Methods currently available for biogas upgrading mainly consists of physicochemical CO2 removal, requiring the use of chemical substances and energy input and, thus, increasing process costs. This PhD project proposes an alternative to existing biogas upgrading technologies, where H2......, produced by water electrolysis, using excess of electricity from wind mills, is coupled with the CO2 contained in the biogas to convert them to CH4. This process is defined as biological biogas upgrading and is carried out by hydrogenotrophic methanogenic archaea that couples CO2 with H2 to produce...

  17. The Role of an Electric Field in the Formation of a Detached Regime in Tokamak Plasma

    Science.gov (United States)

    Senichenkov, I.; Kaveeva, E.; Rozhansky, V.; Sytova, E.; Veselova, I.; Voskoboynikov, S.; Coster, D.

    2018-03-01

    Modeling of the transition to the detachment of ASDEX Upgrade tokamak plasma with increasing density is performed using the SOLPS-ITER numerical code with a self-consistent account of drifts and currents. Their role in plasma redistribution both in the confinement region and in the scrape-off layer (SOL) is investigated. The mechanism of high field side high-density formation in the SOL in the course of detachment is suggested. In the full detachment regime, when the cold plasma region expands above the X-point and reaches closed magnetic-flux surfaces, plasma perturbation in a confined region may lead to a change in the confinement regime.

  18. Overview of progress in European medium sized tokamaks towards an integrated plasma-edge/wall solution

    OpenAIRE

    Meyer, H.; Eich, T.; Beurskens, M.N.A.; Coda, S.; Hakola, A.; Martin, P.; Adamek, J.; Agostini, M.; Aguiam, D.; Ahn, J.; Aho-Mantila, L.; Akers, R.; Albanese, R.; Aledda, R.; Alessi, E.

    2017-01-01

    Integrating the plasma core performance with an edge and scrape-off layer (SOL) that leads to tolerable heat and particle loads on the wall is a major challenge. The new European medium size tokamak task force (EU-MST) coordinates research on ASDEX Upgrade (AUG), MAST and TCV. This multi-machine approach within EU-MST, covering a wide parameter range, is instrumental to progress in the field, as ITER and DEMO core/pedestal and SOL parameters are not achievable simultaneously in present day de...

  19. Discharge cleaning for a tokamak

    International Nuclear Information System (INIS)

    Ishii, Shigeyuki

    1983-01-01

    Various methods of discharge cleaning for tokamaks are described. The material of the first walls of tokamaks is usually stainless steel, inconel, titanium and so on. Hydrogen is exclusively used as the discharge gas. Glow discharge cleaning (GDC), Taylor discharge cleaning (TDC), and electron cyclotron resonance discharge cleaning (ECR-DC) are discussed in this paper. The cleaning by GDC is made by moving a movable anode to the center of a tokamak vassel. Taylor found the good cleaning effect of induced discharge by high pressure and low power discharge. This is called TDC. When the frequency of high frequency discharge in a magnetic field is equal to that of the electron cyclotron resonance, the break down potential is lowered if the pressure is sufficiently low. The ECR-CD is made by using this effect. In TDC and ECR-DC, the electron temperature, which has a close relation to the production rate of H 0 , can be controlled by the pressure. In GDC, the operating pressure was improved by the radio frequency glow (RG) method. However, there is still the danger of arcing. In case of GDC and ECR-DC, the position of plasma can be controlled, but not in case of TDC. The TDC is accepted by most of takamak devices in the world. (Kato, T.)

  20. Magnetic confinement experiment. I: Tokamaks

    International Nuclear Information System (INIS)

    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 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

  1. Status of TMX upgrade diagnostics construction

    International Nuclear Information System (INIS)

    Hornady, R.S.; Davis, J.C.; Simonen, T.C.

    1981-01-01

    This report describes the status of the initial TMX Upgrade diagnostics and the state of development of additional diagnostics being prepared for later TMX Upgrade experiments. The initial diagnostic instrument set has been described in the TMX Upgrade Proposal. This set is required to get TMX Upgrade operational and to evaluate its initial performance. Additional diagnostic instruments are needed to then carry out the more detailed experiments outlined by the TMX Upgrade program milestones. The relation of these new measurements to the physics program is described in The TMX Upgrade Program Plan

  2. Upgrade of reactor operation technology

    International Nuclear Information System (INIS)

    Itoh, Hideaki; Suzuki, Toshiaki; O-kawa, Toshikatsu

    2003-01-01

    To improve operational reliability and availability, the operation technology for a fast reactor was developed in the ''JOYO''. This report describes the upgrading of the simulator, plant operation management tools and fuel handling system for the MK-III core operation. The simulator was modified to the MK-III version to verify operation manuals, and to train operators in MK-III operation. The plant operation management tool was replaced on the operation experience to increase the reliability and efficiency of plant management works relating to plant operation and maintenance. To shorten the refueling period, the fuel handling system was upgraded to full automatic remote control. (author)

  3. Upgrading the BEPC control system

    International Nuclear Information System (INIS)

    Yang Liping; Wang Lizheng; Liu Shiyao

    1992-01-01

    The BEPC control system has been put into operation and operated normally since the end of 1987. Three years's experience shows this system can satisfy basically the operation requirements, also exhibits some disadvantages araised from the original centralized system architecture based on the VAX-VCC-CAMAC, such as slow response, bottle neck of VCC, less CPU power for control etc.. This paper describes the method and procedure for upgrading the BEPC control system which will be based on DEC net and DEC-WS, and thus intend to upgrade the control system architecture from the centralized to the distributed and improve the integral system performance. (author)

  4. Argonne's atlas control system upgrade

    International Nuclear Information System (INIS)

    Munson, F.; Quock, D.; Chapin, B.; Figueroa, J.

    1999-01-01

    The ATLAS facility (Argonne Tandem-Linac Accelerator System) is located at the Argonne National Laboratory. The facility is a tool used in nuclear and atomic physics research, which focuses primarily on heavy-ion physics. The accelerator as well as its control system are evolutionary in nature, and consequently, continue to advance. In 1998 the most recent project to upgrade the ATLAS control system was completed. This paper briefly reviews the upgrade, and summarizes the configuration and features of the resulting control system

  5. Reflections on a digital upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Tadjalli, M.

    2013-07-01

    Upon receiving US NRC's approval in 2010, the first RPS/ESPS digital upgrade using TELEPERM® XS technology was successfully installed in Spring of 2011 at the first Unit of a three-unit station, followed by the 2nd Unit installation in spring of 2012. Both Units' systems have been operating flawlessly since installation. After about two years of operation, a reflection on digital upgrades and lessons learned, from a vendor perspective, provides valuable insight for the commercial nuclear power industry.

  6. 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.

  7. Experiences with tungsten coatings in high heat flux tests and under plasma load in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Herrmann, A; Greuner, H; Fuchs, J C; Marne, P de; Neu, R

    2009-01-01

    ASDEX Upgrade was operated with about 6400 s plasma discharge during the scientific program in 2007/2008 exploring tungsten as a first wall material in tokamaks. In the first phase, the heating power was restricted to 10 MW. It was increased to 15 MW in the second phase. During this operational period, a delamination of the 200 μm W-VPS coating happened at 2 out of 128 tiles of the outer divertor and an unscheduled opening was required. In the third phase, ASDEX Upgrade was operated with partly predamaged tiles and up to 15 MW heating power. The target load was actively controlled by N 2 -seeding. This paper presents the screening test of target tiles in the high heat flux test facility GLADIS, experiences with operation and detected damages of the outer divertor as well as the heat load to the outer divertor and the reasons for the toroidal asymmetry of the divertor load.

  8. Initial results of the high resolution edge Thomson scattering upgrade at DIII-D.

    Science.gov (United States)

    Eldon, D; Bray, B D; Deterly, T M; Liu, C; Watkins, M; Groebner, R J; Leonard, A W; Osborne, T H; Snyder, P B; Boivin, R L; Tynan, G R

    2012-10-01

    Validation of models of pedestal structure is an important part of predicting pedestal height and performance in future tokamaks. The Thomson scattering diagnostic at DIII-D has been upgraded in support of validating these models. Spatial and temporal resolution, as well as signal to noise ratio, have all been specifically enhanced in the pedestal region. This region is now diagnosed by 20 view-chords with a spacing of 6 mm and a scattering length of just under 5 mm sampled at a nominal rate of 250 Hz. When mapped to the outboard midplane, this corresponds to ~3 mm spacing. These measurements are being used to test critical gradient models, in which pedestal gradients increase in time until a threshold is reached. This paper will describe the specifications of the upgrade and present initial results of the system.

  9. Design of a real-time two-color interferometer for MAST Upgrade

    International Nuclear Information System (INIS)

    O’Gorman, T.; Naylor, G.; Scannell, R.; Cunningham, G.; Martin, R.; Croft, D.; Brunner, K. J.

    2014-01-01

    A single chord two-color CO 2 /HeNe (10.6/0.633 μm) heterodyne laser interferometer has been designed to measure the line integral electron density along the mid-plane of the MAST Upgrade tokamak, with a typical error of 1 × 10 18 m −3 (∼2° phase error) at 4 MHz temporal resolution. To ensure this diagnostic system can be restored from any failures without stopping MAST Upgrade operations, it has been located outside of the machine area. The final design and initial testing of this system, including details of the optics, vibration isolation, and a novel phase detection scheme are discussed in this paper

  10. Old PCs: Upgrade or Abandon?

    Science.gov (United States)

    Perez, Ernest

    1997-01-01

    Examines the practical realities of upgrading Intel personal computers in libraries, considering budgets and technical personnel availability. Highlights include adding RAM; putting in faster processor chips, including clock multipliers; new hard disks; CD-ROM speed; motherboards and interface cards; cost limits and economic factors; and…

  11. Overview of ASDEX Upgrade results

    DEFF Research Database (Denmark)

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

    2011-01-01

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

  12. Overview of ASDEX Upgrade results

    NARCIS (Netherlands)

    Kallenbach, A.; Adamek, J.; Aho-Mantila, L.; Akaslompolo, S.; Angioni, C.; Atanasiu, C. V.; Balden, M.; Behler, K.; Belonohy, E.; Bergmann, A.; Bernert, M.; Bilato, R.; Bobkov, V.; Boom, J.; Bottino, A.; Braun, F.; Brudgam, M.; Buhler, A.; Burckhart, A.; Chankin, A.; Classen, I.G.J.; Conway, G. D.; Coster, D. P.; de Marne, P.; D' Inca, R.; Drube, R.; Dux, R.; Eich, T.; Endstrasser, N.; Engelhardt, K.; Esposito, B.; Fable, E.; Fahrbach, H. U.; Fattorini, L.; Fischer, R.; Flaws, A.; Funfgelder, H.; Fuchs, J. C.; Gal, K.; Munoz, M. G.; Geiger, B.; Adamov, M. G.; Giannone, L.; Giroud, C.; Gorler, T.; da Graca, S.; Greuner, H.; Gruber, O.; Gude, A.; Gunter, S.; Haas, G.; Hakola, A. H.; Hangan, D.; Happel, T.; Hauff, T.; Heinemann, B.; Herrmann, A.; Hicks, N.; Hobirk, J.; Hohnle, H.; Holzl, M.; Hopf, C.; Horton, L.; Huart, M.; Igochine, V.; Ionita, C.; Janzer, A.; Jenko, F.; Kasemann, C. P.; Kalvin, S.; Kardaun, O.; Kaufmann, M.; Kirk, A.; Klingshirn, H. J.; Kocan, M.; Kocsis, G.; Kollotzek, H.; Konz, C.; Koslowski, R.; Krieger, K.; Kurki-Suonio, T.; Kurzan, B.; Lackner, K.; Lang, P. T.; Lauber, P.; Laux, M.; Leipold, F.; Leuterer, F.; Lohs, A.; N C Luhmann Jr.,; Lunt, T.; Lyssoivan, A.; Maier, H.; Maggi, C.; Mank, K.; Manso, M. E.; Maraschek, M.; Martin, P.; Mayer, M.; McCarthy, P. J.; McDermott, R.; Meister, H.; Menchero, L.; Meo, F.; Merkel, P.; Merkel, R.; Mertens, V.; Merz, F.; Mlynek, A.; Monaco, F.; Muller, H. W.; Munich, M.; Murmann, H.; Neu, G.; Neu, R.; Nold, B.; Noterdaeme, J. M.; Park, H. K.; Pautasso, G.; Pereverzev, G.; Podoba, Y.; Pompon, F.; Poli, E.; Polochiy, K.; Potzel, S.; Prechtl, M.; Puschel, M. J.; Putterich, T.; Rathgeber, S. K.; Raupp, G.; Reich, M.; Reiter, B.; Ribeiro, T.; Riedl, R.; Rohde, V.; Roth, J.; Rott, M.; Ryter, F.; Sandmann, W.; Santos, J.; Sassenberg, K.; Sauter, P.; Scarabosio, A.; Schall, G.; Schmid, K.; Schneider, P. A.; Schneider, W.; Schramm, G.; Schrittwieser, R.; Schweinzer, J.; Scott, B.; Sempf, M.; Serra, F.; Sertoli, M.; Siccinio, M.; Sigalov, A.; Silva, A.; Sips, A.C.C.; Sommer, F.; Stabler, A.; Stober, J.; Streibl, B.; Strumberger, E.; Sugiyama, K.; Suttrop, W.; Szepesi, T.; Tardini, G.; Tichmann, C.; Told, D.; Treutterer, W.; Urso, L.; Varela, P.; Vincente, J.; Vianello, N.; Vierle, T.; Viezzer, E.; Vorpahl, C.; Wagner, D.; Weller, A.; Wenninger, R.; Wieland, B.; Wigger, C.; Willensdorfer, M.; Wischmeier, M.; Wolfrum, E.; Wursching, E.; Yadikin, D.; Yu, Q.; Zammuto, I.; Zasche, D.; Zehetbauer, T.; Zhang, Y.; Zilker, M.; Zohm, H.

    2011-01-01

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

  13. Overview of ASDEX Upgrade results

    DEFF Research Database (Denmark)

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

    2009-01-01

    ASDEX Upgrade was operated with a fully W-covered wall in 2007 and 2008. Stationary H-modes at the ITER target values and improved H-modes with H up to 1.2 were run without any boronization. The boundary conditions set by the full W wall (high enough ELM frequency, high enough central heating and...

  14. Upgrading of the Budapest reactor

    International Nuclear Information System (INIS)

    Rosta, L.

    1986-10-01

    The increasing importance of neutron sources, high demand for irradiation and experimental facilities as well as improved safety requirements in the 'eighties, necessitate not only a technical modernization of the Reactor but an overall upgrade including instrumentation. Such a reconstruction was decided by the Hungarian governement in 1983

  15. 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.

  16. LHC challenges and upgrade options

    Energy Technology Data Exchange (ETDEWEB)

    Bruning, O [CERN AB/ABP, Y03600, 1211 Geneva 23 (Switzerland)], E-mail: Oliver.Bruning@cern.ch

    2008-05-15

    The presentation summarizes the key parameters of the LHC collider. Following a discussion of the main challenges for reaching the nominal machine performance the presentation identifies options for increasing the operation tolerances and the potential performance reach of the LHC by means of future hardware upgrades of the LHC and its injector complex.

  17. LHC challenges and upgrade options

    International Nuclear Information System (INIS)

    Bruning, O

    2008-01-01

    The presentation summarizes the key parameters of the LHC collider. Following a discussion of the main challenges for reaching the nominal machine performance the presentation identifies options for increasing the operation tolerances and the potential performance reach of the LHC by means of future hardware upgrades of the LHC and its injector complex

  18. Five-Megajoule Homopolar Upgrade

    Science.gov (United States)

    1981-06-01

    near the bearing lands to monitor temperature rise during motoring. AISI 4140 steel sleeves were shrunk onto the beryllium copper shaft of the new...MJ HPG in this upgrade. Rotor and Shaft A 0.75-m diameter, 0.28-m thick AISI 4340 air- craft quality steel rotor is shrunk on to a 0. 14-m diameter

  19. Overview of ASDEX Upgrade results

    DEFF Research Database (Denmark)

    Kallenbach, A.; Aguiam, D.; Aho-Mantila, L.

    2017-01-01

    The ASDEX Upgrade (AUG) programme is directed towards physics input to critical elements of the ITER design and the preparation of ITER operation, as well as addressing physics issues for a future DEMO design. Since 2015, AUG is equipped with a new pair of 3-strap ICRF antennas, which were design...

  20. Magnetic confinement by Tokamak: physical aspects

    International Nuclear Information System (INIS)

    Tachon, J.

    1980-01-01

    After describing the Tokamak configuration concept, the author provides an analysis of the principal physical aspects of this type of installation and concludes by estimating that the Tokamak concept is a 'plausible candidate' as a means of producing controlled thermonuclear fusion [fr

  1. Economic evaluation of tokamak power plants

    International Nuclear Information System (INIS)

    Reid, R.L.; Steiner, D.

    1977-01-01

    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

  2. Simulation of a major tokamak disruption

    International Nuclear Information System (INIS)

    White, R.B.; Monticello, D.A.; Rosenbluth, M.N.

    1977-08-01

    It is known that the internal tokamak disruption leads to a current profile which is flattened inside the surface where the safety factor equals unity. It is shown that such a profile can lead to m = 2 magnetic islands which grow to fill a substantial part of the tokamak cross section in a time consistent with the observations of the major disruption

  3. Diagnostics for the Rijnhuizen Tokamak Project

    NARCIS (Netherlands)

    Donne, A. J. H.

    1994-01-01

    The research program of the Rijnhuizen Tokamak Project is concentrated on the study of plasma transport processes. The RTP tokamak is therefore equipped with an extensive set of multichannel diagnostics, including a 19-channel FIR interferometer, a 20-channel heterodyne ECE system, an 80-channel

  4. MAST: a Mega Amp Spherical Tokamak

    International Nuclear Information System (INIS)

    Darke, A.C.; Harbar, J.R.; Hay, J.H.; Hicks, J.B.; Hill, J.W.; McKenzie, J.S.; Morris, A.W.; Nightingale, M.P.S.; Todd, T.N.; Voss, G.M.; Watkins, J.R.

    1995-01-01

    The highly successful tight aspect ratio tokamak research pioneered on the START machine at Culham, together with the attractive possibilities of the concept, suggest a larger device should be considered. The design of a Mega Amp Spherical Tokamak is described, operating at much higher currents and over longer pulses than START and compatible with strong additional heating. (orig.)

  5. Mercier criterion for high-β tokamaks

    International Nuclear Information System (INIS)

    Galvao, R.M.O.

    1984-01-01

    An expression, for the application of the Mercier criterion to numerical studies of diffuse high-β tokamaks (β approximatelly Σ,q approximatelly 1), which contains only leading order contributions in the high-β tokamak approximation is derived. (L.C.) [pt

  6. The ARIES-I tokamak reactor study

    International Nuclear Information System (INIS)

    1991-01-01

    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

  7. Engineering Design of KSTAR tokamak main structure

    International Nuclear Information System (INIS)

    Im, K.H.; Cho, S.; Her, N.I.

    2001-01-01

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

  8. Summary report on tokamak confinement experiments

    International Nuclear Information System (INIS)

    1982-03-01

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

  9. Cluster storage for COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

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

    2012-01-01

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

  10. Surface tearing modes in tokamaks

    International Nuclear Information System (INIS)

    Takizuka, Tomonori; Kurita, Gen-ichi; Azumi, Masafumi; Takeda, Tatsuoki

    1985-10-01

    Surface tearing modes in tokamaks are studied numerically and analytically. The eigenvalue problem is solved to obtain the growth rate and the mode structure. We investigate in detail dependences of the growth rate of the m/n = 2/1 resistive MHD modes on the safety factor at the plasma surface, current profile, wall position, and resistivity. The surface tearing mode moves the plasma surface even when the wall is close to the surface. The stability diagram for these modes is presented. (author)

  11. Major disruption process in tokamak

    International Nuclear Information System (INIS)

    Kurita, Gen-ichi; Azumi, Masafumi; Tuda, Takashi; Takizuka, Tomonori; Tsunematsu, Toshihide; Tokuda, Shinji; Itoh, Kimitaka; Takeda, Tatsuoki

    1981-11-01

    The major disruption in a cylindrical tokamak is investigated by using the multi-helicity code, and the destabilization of the 3/2 mode by the mode coupling with the 2/1 mode is confirmed. The evolution of the magnetic field topology caused by the major disruption is studied in detail. The effect of the internal disruption on the 2/1 magnetic island width is also studied. The 2/1 magnetic island is not enhanced by the flattening of the q-profile due to the internal disruption. (author)

  12. Initial results in SST-1 after up-gradation

    Science.gov (United States)

    Pradhan, S.; Khan, Z.; Tanna, V. L.; Prasad, U.; Paravastu, Y.; Raval, D. C.; Masand, H.; Kumar, Aveg; Dhongde, J. R.; Jana, S.; Kakati, B.; Patel, K. B.; Bhandarkar, M. K.; Shukla, B. K.; Ghosh, D.; Patel, H. S.; Parekh, T. J.; Mansuri, I. A.; Dhanani, K. R.; Varadharajulu, A.; Khristi, Y. S.; Biswas, P.; Gupta, C. N.; George, S.; Semwal, P.; Sharma, D. K.; Gulati, H. K.; Mahajan, K.; Praghi, B. R.; Banaudha, M.; Makwana, A. R.; Chudasma, H. H.; Kumar, M.; Manchanda, R.; Joisa, Y. S.; Asudani, K.; Pandya, S. N.; Pathak, S. K.; Banerjee, S.; Patel, P. J.; Santra, P.; Pathan, F. S.; Chauhan, P. K.; Khan, M. S.; Thankey, P. L.; Prakash, A.; Panchal, P. N.; Panchal, R. N.; Patel, R. J.; Mahsuria, G. I.; Sonara, D. P.; Patel, K. M.; Jayaswal, S. P.; Sharma, M.; Patel, J. C.; Varmora, P.; Srikanth, G. L. N.; Christian, D. R.; Garg, A.; Bairagi, N.; Babu, G. R.; Panchal, A. G.; Vora, M. M.; Singh, A. K.; Sharma, R.; Nimavat, H. D.; Shah, P. R.; Purwar, G.; Raval, T. Y.; Sharma, A. L.; Ojha, A.; Kumar, S.; Ramaiya, N. K.; Siju, V.; Gopalakrishna, M. V.; Kumar, A.; Sharma, P. K.; Atrey, P. K.; Kulkarni, SV; Ambulkar, K. K.; Parmar, P. R.; Thakur, A. L.; Raval, J. V.; Purohit, S.; Mishra, P. K.; Adhiya, A. N.; Nagora, U. C.; Thomas, J.; Chaudhari, V. K.; Patel, K. G.; Dalakoti, S.; Virani, C. G.; Gupta, S.; Kumar, Ajay; Chaudhari, B.; Kaur, R.; Srinivasan, R.; Raju, D.; Kanabar, D. H.; Jha, R.; Das, A.; Bora, D.

    2017-04-01

    SST-1 Tokamak has recently completed the 1st phase of up-gradation with successful installation and integration of all its First Wall components. The First Wall of SST-1 comprises of ∼ 3800 high heat flux compatible graphite tiles being assembled and installed on 132 CuCrZr heat sink back plates engraved with ∼ 4 km of leak tight baking and cooling channels in five major sub groups equipped with ∼ 400 sensors and weighing ∼ 6000 kg in total in thirteen isolated galvanic and six isolated hydraulic circuits. The phase-1 up-gradation spectrum also includes addition of Supersonic Molecular Beam Injection (SMBI) both on the in-board and out-board side, installation of fast reciprocating probes, adding some edge plasma probe diagnostics in the SOL region, installation and integration of segmented and up-down symmetric radial coils aiding/controlling plasma rotations, introduction of plasma position feedback and density controls etc. Post phase-I up-gradation spanning from Nov 2014 till June 2016, initial plasma experiments in up-graded SST-1 have begun since Aug 2016 after a brief engineering validation period in SST-1. The first experiments in SST-1 have revealed interesting aspects on the ‘eddy currents in the First Wall support structures’ influencing the ‘magnetic Null evolution dynamics’ and the subsequent plasma start-up characteristics after the ECH pre-ionization, the influence of the first walls on the ‘field errors’ and the resulting locked modes observed, the magnetic index influencing the evolution of the equilibrium of the plasma column, low density supra-thermal electron induced discharges and normal ohmic discharges etc. Presently; repeatable ohmic discharges regimes in SST-1 having plasma currents in excess of 65 KA (qa ∼ 3.8, BT = 1.5 T) with a current ramp rates ∼ 1.2 MA/s over a duration of ∼ 300 ms with line averaged densities ∼ 0.8 × 1019 and temperatures ∼ 200 eV with copious MHD signatures have been experimentally

  13. Recent results from the DIII-D tokamak

    International Nuclear Information System (INIS)

    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

  14. Nuclear fusion research at Tokamak Energy Ltd

    International Nuclear Information System (INIS)

    Windridge, Melanie J.; Gryaznevich, Mikhail; Kingham, David

    2017-01-01

    Tokamak Energy's approach is close to the mainstream of nuclear fusion, and chooses a spherical tokamak, which is an economically developed form of Tokamak reactor design, as research subjects together with a high-temperature superconducting magnet. In the theoretical prediction, it is said that spherical tokamak can make tokamak reactor's scale compact compared with ITER or DEMO. The dependence of fusion energy multiplication factor on reactor size is small. According to model studies, it has been found that the center coil can be protected from heat and radiation damage even if the neutron shielding is optimized to 35 cm instead of 1 m. As a small tokamak with a high-temperature superconducting magnet, ST25 HTS, it demonstrated in 2015 continuous operation for more than 24 hours as a world record. Currently, this company is constructing a slightly larger ST40 type, and it is scheduled to start operation in 2017. ST40 is designed to demonstrate that it can realize a high magnetic field with a compact size and aims at attaining 8-10 keV (reaching the nuclear fusion reaction temperature at about 100 million degrees). This company will verify the startup and heating technology by the coalescence of spherical tokamak expected to have plasma current of 2 MA, and will also use 2 MW of neutral particle beam heating. In parallel with ST40, it is promoting a development program for high-temperature superconducting magnet. (A.O.)

  15. Upgraded wood residue fuels 1995

    International Nuclear Information System (INIS)

    Vinterbaeck, J.

    1995-01-01

    The Swedish market for upgraded residue fuels, i.e. briquettes, pellets and wood powder, has developed considerably during the nineties. The additional costs for the upgrading processes are regained and create a surplus in other parts of the system, e.g. in the form of higher combustion efficiencies, lower investment costs for burning equipment, lower operation costs and a diminished environmental impact. All these factors put together have resulted in a rapid growth of this part of the energy sector. In 1994 the production was 1.9 TWh, an increase of 37 % compared to the previous year. In the forthcoming heating season 1995/96 the production may reach 4 TWh. 57 refs, 11 figs, 6 tabs

  16. Enhancement of confinement in tokamaks

    International Nuclear Information System (INIS)

    Furth, H.P.

    1986-05-01

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

  17. CAT-D-T tokamaks

    International Nuclear Information System (INIS)

    Greenspan, E.; Blue, T.; Miley, G.H.

    1981-01-01

    The domains of plasma fuel cycles bounded by the D-T and Cat-D, and by the D-T and SCD modes of operation are examined. These domains, referred to as, respectively, the Cat-D-T and SCD-T modes of operation, are characterized by the number (γ) of tritons per fusion neutron available from external (to the plasma) sources. Two external tritium sources are considered - the blankets of the Cat-D-T (SCD-T) reactors and fission reactors supported by the Cat-D-T (SCD-T) driven hybrid reactors. It is found that by using 6 Li for the active material of the control elements of the fission reactors, it is possible to achieve γ values close to unity. Cat-D-T tokamaks could be designed to have smaller size, higher power density, lower magnetic field and even lower plasma temperature than Cat-D tokamaks; the difference becomes significant for γ greater than or equal to .75. The SCD-T mode of operation appears to be even more attractive. Promising applications identified for these Cat-D-T and SCD-T modes of operation include hybrid reactors, fusion synfuel factories and fusion reactors which have difficulty in providing all their tritium needs

  18. 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.

  19. 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.

  20. The TEXT upgrade vertical interferometer

    International Nuclear Information System (INIS)

    Hallock, G.A.; Gartman, M.L.; Li, W.; Chiang, K.; Shin, S.; Castles, R.L.; Chatterjee, R.; Rahman, A.S.

    1992-01-01

    A far-infrared interferometer has been installed on TEXT upgrade to obtain electron density profiles. The primary system views the plasma vertically through a set of large (60-cm radialx7.62-cm toroidal) diagnostic ports. A 1-cm channel spacing (59 channels total) and fast electronic time response is used, to provide high resolution for radial profiles and perturbation experiments. Initial operation of the vertical system was obtained late in 1991, with six operating channels

  1. Upgrading Temelin to international standards

    International Nuclear Information System (INIS)

    Anderson, J.J.; Gisoni, G.A.; Stormer, T.D.

    1994-01-01

    The digital technology being installed to upgrade the Instrumentation and Control system at the Temelin twin VVER-1000 station in the Czech Republic will allow the plant to operate more safely because it allows key information to be organised and presented to the operators, enabling them to make better informed decisions about the status of the plant. The new system is based on proven technology previously applied at Beznau in Switzerland and Sizewell B in the UK. (author)

  2. Status of the COMPASS tokamak and characterization of the first H-mode

    Science.gov (United States)

    Pánek, R.; Adámek, J.; Aftanas, M.; Bílková, P.; Böhm, P.; Brochard, F.; Cahyna, P.; Cavalier, J.; Dejarnac, R.; Dimitrova, M.; Grover, O.; Harrison, J.; Háček, P.; Havlíček, J.; Havránek, A.; Horáček, J.; Hron, M.; Imríšek, M.; Janky, F.; Kirk, A.; Komm, M.; Kovařík, K.; Krbec, J.; Kripner, L.; Markovič, T.; Mitošinková, K.; Mlynář, J.; Naydenkova, D.; Peterka, M.; Seidl, J.; Stöckel, J.; Štefániková, E.; Tomeš, M.; Urban, J.; Vondráček, P.; Varavin, M.; Varju, J.; Weinzettl, V.; Zajac, J.; the COMPASS Team

    2016-01-01

    This paper summarizes the status of the COMPASS tokamak, its comprehensive diagnostic equipment and plasma scenarios as a baseline for the future studies. The former COMPASS-D tokamak was in operation at UKAEA Culham, UK in 1992-2002. Later, the device was transferred to the Institute of Plasma Physics of the Academy of Sciences of the Czech Republic (IPP AS CR), where it was installed during 2006-2011. Since 2012 the device has been in a full operation with Type-I and Type-III ELMy H-modes as a base scenario. This enables together with the ITER-like plasma shape and flexible NBI heating system (two injectors enabling co- or balanced injection) to perform ITER relevant studies in different parameter range to the other tokamaks (ASDEX-Upgrade, DIII-D, JET) and to contribute to the ITER scallings. In addition to the description of the device, current status and the main diagnostic equipment, the paper focuses on the characterization of the Ohmic as well as NBI-assisted H-modes. Moreover, Edge Localized Modes (ELMs) are categorized based on their frequency dependence on power density flowing across separatrix. The filamentary structure of ELMs is studied and the parallel heat flux in individual filaments is measured by probes on the outer mid-plane and in the divertor. The measurements are supported by observation of ELM and inter-ELM filaments by an ultra-fast camera.

  3. Microwave Tokamak Experiment: An overview of the construction and checkout phase

    International Nuclear Information System (INIS)

    Lang, L.L.; Bell, H.H.

    1989-01-01

    At Lawrence Livermore National Laboratory (LLNL) we constructed and presently operate the Microwave Tokamak Experiment (MTX) to demonstrate the feasibility of using microwave pulses produced from a free electron laser (FEL) to provide electron cyclotron heating (ECH) for use in tokamaks, particularly high-field machines. The MTX consists primarily of the ALCATOR C tokamak and power supplies that were documented and disassembled at the Massachusetts Institute of Technology (MIT) and shipped to LLNL in April 1987. We made many additions, including a new primary power system from the magnetic Fusion Test Facility (MFTF) substation, a new commutation system, substantially upgraded seismic support system for earthquake loading, a fast controls system for use with the FEL, a new data-acquisition system, and a new vault facility. We checked out these systems and put them into operation in October 1988; we achieved the first plasma in November 1988. We have also constructed and installed the microwave transmission system and the local microwave system to be used with the FEL. These systems transmit the microwaves to MTX quasi-optically through an evacuated tube. The ongoing plasma operations, both with and without FEL heating, are described in a companion paper. 12 refs., 2 figs., 2 tabs

  4. 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.

  5. Comparison of runaway electron generation parameters in small, medium-sized and large tokamaks—A survey of experiments in COMPASS, TCV, ASDEX-Upgrade and JET

    Science.gov (United States)

    Plyusnin, V. V.; Reux, C.; Kiptily, V. G.; Pautasso, G.; Decker, J.; Papp, G.; Kallenbach, A.; Weinzettl, V.; Mlynar, J.; Coda, S.; Riccardo, V.; Lomas, P.; Jachmich, S.; Shevelev, A. E.; Alper, B.; Khilkevitch, E.; Martin, Y.; Dux, R.; Fuchs, C.; Duval, B.; Brix, M.; Tardini, G.; Maraschek, M.; Treutterer, W.; Giannone, L.; Mlynek, A.; Ficker, O.; Martin, P.; Gerasimov, S.; Potzel, S.; Paprok, R.; McCarthy, P. J.; Imrisek, M.; Boboc, A.; Lackner, K.; Fernandes, A.; Havlicek, J.; Giacomelli, L.; Vlainic, M.; Nocente, M.; Kruezi, U.; COMPASS Team; TCV Team; ASDEX-Upgrade Team; EUROFusion MST1 Team; contributors, JET

    2018-01-01

    This paper presents a survey of the experiments on runaway electrons (RE) carried out recently in frames of EUROFusion Consortium in different tokamaks: COMPASS, ASDEX-Upgrade, TCV and JET. Massive gas injection (MGI) has been used in different scenarios for RE generation in small and medium-sized tokamaks to elaborate the most efficient and reliable ones for future RE experiments. New data on RE generated at disruptions in COMPASS and ASDEX-Upgrade was collected and added to the JET database. Different accessible parameters of disruptions, such as current quench rate, conversion rate of plasma current into runaways, etc have been analysed for each tokamak and compared to JET data. It was shown, that tokamaks with larger geometrical sizes provide the wider limits for spatial and temporal variation of plasma parameters during disruptions, thus extending the parameter space for RE generation. The second part of experiments was dedicated to study of RE generation in stationary discharges in COMPASS, TCV and JET. Injection of Ne/Ar have been used to mock-up the JET MGI runaway suppression experiments. Secondary RE avalanching was identified and quantified for the first time in the TCV tokamak in RE generating discharges after massive Ne injection. Simulations of the primary RE generation and secondary avalanching dynamics in stationary discharges has demonstrated that RE current fraction created via avalanching could achieve up to 70-75% of the total plasma current in TCV. Relaxations which are reminiscent the phenomena associated to the kinetic instability driven by RE have been detected in RE discharges in TCV. Macroscopic parameters of RE dominating discharges in TCV before and after onset of the instability fit well to the empirical instability criterion, which was established in the early tokamaks and examined by results of recent numerical simulations.

  6. Plasma position control in TCABR Tokamak

    International Nuclear Information System (INIS)

    Galvao, R.M.O.; Kuznetsov, Yu. K.; Nascimento, I.C.; Fonseca, A.M.M.; Silva, R.P. da; Ruchko, L.F.; Tuszel, A.G.; Reis, A.P. dos; Sanada, E.K.

    1998-01-01

    The plasma control position in the TCABR tokamak is described. The TCA tokamak was transferred from the Centre de Recherches en Physique des Plasmas, Lausanne, to the Institute of Physics of University of Sao Paulo, renamed TCABR (α=0.18 m, R = 0.62 m, B = 1 T,I p = 100 kA). The control system was reconstructed using mainly components obtained from the TCA tokamak. A new method of plasma position determination is used in TCABR to improve its accuracy. A more detailed theoretical analysis of the feed forward and feedback control is performed as compared with. (author)

  7. Estimation of Zeff in Novillo Tokamak

    International Nuclear Information System (INIS)

    Valencia, R.; Olayo, G.; Cruz, G.; Lopez, R.; Chavez, E.; Melendez, L.; Flores, A.; Gaytan, E.

    1996-01-01

    We estimated the Z eff in the Novillo Tokamak after having applied a HeGDC process through two different methods: anomaly factor and mass spectrometry. The first one gave a Z eff value of 2.07 for a tokamak discharge of 4350 A plasma current and 3 V of loop voltage. By mass spectrometry 30 s after the discharge had finished a Z eff of 4.19 was obtained for the same discharge. By mass spectrometry we observed that the Z eff value is a time function. Furthermore this method is helpful for evaluating the level of impurities after many discharges in Novillo Tokamak. (orig.)

  8. Global gyrokinetic simulation of tokamak transport

    International Nuclear Information System (INIS)

    Furnish, G.; Horton, W.; Kishimoto, Y.; LeBrun, M.J.; Tajima, T.

    1998-10-01

    A kinetic simulation code based on the gyrokinetic ion dynamics in global general metric (including a tokamak with circular or noncircular cross-section) has been developed. This gyrokinetic simulation is capable of examining the global and semi-global driftwave structures and their associated transport in a tokamak plasma. The authors investigate the property of the ion temperature gradient (ITG) or η i (η i ≡ ∂ ell nT i /∂ ell n n i ) driven drift waves in a tokamak plasma. The emergent semi-global drift wave modes give rise to thermal transport characterized by the Bohm scaling

  9. Fast IR diodes thermometer for tokamak

    International Nuclear Information System (INIS)

    Chen Xiangbo

    2001-01-01

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

  10. Towards novel biogas upgrading processes

    Energy Technology Data Exchange (ETDEWEB)

    Privalova, E.

    2013-06-01

    Biogas production has considerable development possibilities not only in Finland but all over the world since it it the easiest way of creating value out of various waste fractions and represents an alternative source of renewable energy. Development of efficient biogas upgrading technology has become an important issue since it improves the quality of biogas and for example facilitating its injection into the natural gas pipelines. Moreover, such upgrading contributes to resolving the issue of increasing CO{sub 2} emissions and addresses the increasing climate change concerns. Together with traditional CO{sub 2} capturing technologies a new class of recently emerged sorbents such as ionic liquids is claimed as promising media for gas separations. In this thesis, an extensive comparison of the performance of different solvents in terms of CO{sub 2} capture has been performed. The focus of the present study was on aqueous amine solutions and their mixtures, traditional ionic liquids, 'switchable' ionic liquids and poly(ionic liquid)s in order to reveal the best option for biogas upgrading. The CO{sub 2} capturing efficiency for the most promising solvents achieved values around 50-60 L CO{sub 2}/L absorbent. These values are superior to currently widely applied water wash biogas upgrading system. Regeneration of the solvent mixtures appeared to be challenging since the loss of initial efficiency upon CO{sub 2} release was in excess of 20-40 vol %, especially in the case of aqueous amine solutions. In contrast, some of the ionic liquids displayed reversible behavior. Thus, for selected 'switchable' ionic and poly(ionic liquid)s the CO{sub 2} absorption/regeneration cycles were performed 3-4 times without any notable efficiency decrease. The viscosity issue, typical for ionic liquids upon CO{sub 2} saturation, was addressed and the information obtained was evaluated and related to the ionic interactions. The occurrence of volatile organic compounds

  11. CORBA-based solution for remote participation in SST-1 tokamak control and operation

    Energy Technology Data Exchange (ETDEWEB)

    Mahajan, Kirti [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India)]. E-mail: kirti@ipr.res.in; Ravikiran, M. [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Gulati, Hitesh [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Dave, H.J. [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Kumar, Neeraj [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Patel, Kirit [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Kumar, Aveg [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Raju, D. [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Bhandarkar, M. [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Chudasama, H. [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Kulkarni, S.V. [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India); Saxena, Y.C. [Institute for Plasma Research, Near Indira Bridge, Bhat, Gandhinagar 382428 (India)

    2006-07-15

    The steady state superconducting tokamak (SST-1) central control system is a distributed heterogeneous process communication system built on socket programming. It consists of machine, experiment and discharge control plus timing and a database. The software controls and monitors SST-1 subsystems: water-cooling, power supplies, cryogenics and vacuum over a local area network (LAN). The SST-1 control room is the place where all the activities like session announcement, machine control, experiment control, discharge control and monitoring are performed. We have realized that, instead of having a single monitoring place, we should have multiple monitoring points and it should be made possible to control the experiment from any PC over the LAN. In order to meet such requirements for remote participation in tokamak operation, we are upgrading the existing software. The upgraded software is based on Common Object Request Broker Architecture (CORBA) technology. The software is utilizing CORBA-services such as event service, naming services, interface repository and security services. The inherent features of CORBA make the software, platform and language independent. The software supports a variety of communication paradigms including publish-subscribe, peer-to-peer, and request-reply. Based on this software, one can participate in SST-1 tokamak operation from the LAN, or a wide area network (WAN) connection anywhere on the Internet. Each user can customize plasma parameters and diagnostics data that he wants to monitor, at any time without any change in the software and a copy of these parameters will be available to him. This paper focuses on the publish-subscribe communication paradigm and its application for a machine monitoring system.

  12. CORBA-based solution for remote participation in SST-1 tokamak control and operation

    International Nuclear Information System (INIS)

    Mahajan, Kirti; Ravikiran, M.; Gulati, Hitesh; Dave, H.J.; Kumar, Neeraj; Patel, Kirit; Kumar, Aveg; Raju, D.; Bhandarkar, M.; Chudasama, H.; Kulkarni, S.V.; Saxena, Y.C.

    2006-01-01

    The steady state superconducting tokamak (SST-1) central control system is a distributed heterogeneous process communication system built on socket programming. It consists of machine, experiment and discharge control plus timing and a database. The software controls and monitors SST-1 subsystems: water-cooling, power supplies, cryogenics and vacuum over a local area network (LAN). The SST-1 control room is the place where all the activities like session announcement, machine control, experiment control, discharge control and monitoring are performed. We have realized that, instead of having a single monitoring place, we should have multiple monitoring points and it should be made possible to control the experiment from any PC over the LAN. In order to meet such requirements for remote participation in tokamak operation, we are upgrading the existing software. The upgraded software is based on Common Object Request Broker Architecture (CORBA) technology. The software is utilizing CORBA-services such as event service, naming services, interface repository and security services. The inherent features of CORBA make the software, platform and language independent. The software supports a variety of communication paradigms including publish-subscribe, peer-to-peer, and request-reply. Based on this software, one can participate in SST-1 tokamak operation from the LAN, or a wide area network (WAN) connection anywhere on the Internet. Each user can customize plasma parameters and diagnostics data that he wants to monitor, at any time without any change in the software and a copy of these parameters will be available to him. This paper focuses on the publish-subscribe communication paradigm and its application for a machine monitoring system

  13. Upgrader alley : oil sands fever strikes Edmonton

    International Nuclear Information System (INIS)

    Griffiths, M.; Dyer, S.

    2008-01-01

    Large-scale industrial complexes called upgraders are similar to oil refineries. Several upgraders are planned for the area just northeast of Edmonton, known as Upgrader Alley. Concerns have been expressed over the potential congestion and environmental impacts of these upgraders. Upgraders will also attract other industry, and the cumulative effects of development will have major impacts on the region, its people and the natural environment. The report provided an overview of Upgrader Alley, with reference to what is driving development; upgrading issues; what Upgrader Alley will look like; and how much water Upgrader Alley needs. The report also discussed impacts on the land, air quality, and greenhouse gases. Water demand issues were discussed with reference to impacts on the North Saskatchewan River, water levels, water quality, a water management framework, and groundwater resources. Cumulative impacts were also presented. It was concluded that if all the projects for which applications had been submitted were approved, the rapid pace of growth in Upgrader Alley would mimic that of Fort McMurray. If the rate of development were somewhat slower, there would be more time to develop and implement plans to reduce the impacts. 189 refs., 6 tabs., 14 figs

  14. Density profile evolution during dynamic processes in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Nunes, I.; Santos, J.; Salzedas, F.; Manso, M.; Serra, F.; Conway, G.D.; Horton, L.D.; Neuhauser, J.; Suttrop, W.

    2005-01-01

    The current understanding of edge localized modes (ELMs) and the trigger of major disruptions is largely based on phenomenology. The need to better understand the processes underlying these phenomena requires high temporal and spatial resolution diagnostics. Fast diagnostics for the temperature measurements exist, such as the ECE radiometer but, for the plasma density, the existing diagnostics such as Lithium Beam and Thomson Scattering do not have the required high temporal resolution for a period long enough to characterize the entire ELM event. The microwave reflectometry system on ASDEX Upgrade has the capability to measure electron density profiles simultaneously at the low-field and high-field sides, in broadband swept ultrafast (35μs) operation with a spatial resolution of 5mm. In this paper we report on recent results on the effects of type I ELMs on density profiles and on the density pedestal width and ELM affected depth. During the ELM event, three phases are identified: precursor, collapse and recovery. The density pedestal width is found to be approximately constant for all the ELMy H-mode discharges analyzed here, except for high input power discharges, where an increase of the density pedestal width is observed. Major disruptions limit the range of parameters used in the operation of a tokamak, especially density limit disruptions, that limit the maximum usable density. Very abrupt increases of density are observed before the onset of the electron temperature profile erosion, supporting the hypothesis that this erosion is due to convection of the magnetic field. In ITER, during the long steady state flat-top phase of the discharges magnetic measurements may accumulate significant drifts. Plasma position and shape control using reflectometry is being assessed in ASDEX Upgrade for ITER like scenarios with successful results, where it is shown that position measurements from reflectometry compared to magnetic data satisfy the ITER requirements

  15. Submillimeter wave propagation in tokamak plasmas

    International Nuclear Information System (INIS)

    Ma, C.H.; Hutchinson, D.P.; Staats, P.A.; Vander Sluis, K.L.; Mansfield, D.K.; Park, H.; Johnson, L.C.

    1985-01-01

    The propagation of submillimeter-waves (smm) in tokamak plasmas has been investigated both theoretically and experimentally to ensure successful measurements of electron density and plasma current distributions in tokamak devices. Theoretical analyses have been carried out to study the polarization of the smm waves in TFTR and ISX-B tokamaks. A multichord smm wave interferometer/polarimeter system has been employed to simultaneously measure the line electron density and poloidal field-induced Faraday rotation in the ISX-B tokamak. The experimental study on TFTR is under way. Computer codes have been developed and have been used to study the wave propagation and to reconstruct the distributions of plasma current and density from the measured data. The results are compared with other measurements

  16. Power and particle exhaust in tokamaks

    International Nuclear Information System (INIS)

    Stambaugh, R.D.

    1998-01-01

    The status of power and particle exhaust research in tokamaks is reviewed in the light of ITER requirements. There is a sound basis for ITER's nominal design positions; important directions for further research are identified

  17. Definition of total bootstrap current in tokamaks

    International Nuclear Information System (INIS)

    Ross, D.W.

    1995-01-01

    Alternative definitions of the total bootstrap current are compared. An analogous comparison is given for the ohmic and auxiliary currents. It is argued that different definitions than those usually employed lead to simpler analyses of tokamak operating scenarios

  18. Tokamak research in the Soviet Union

    International Nuclear Information System (INIS)

    Strelkov, V.S.

    1981-01-01

    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.)

  19. Neutral beam in ALVAND IIC tokamak

    International Nuclear Information System (INIS)

    Ghrannevisse, M.; Moradshahi, M.; Avakian, M.

    1992-01-01

    Neutral beams have a wide application in tokamak experiments. It used to heat; fuel; adjust electric potentials in plasmas and diagnose particles densities and momentum distributions. It may be used to sustain currents in tokamaks to extend the pulse length. A 5 KV; 500 mA ion source has been constructed by plasma physics group, AEOI and it used to produce plasma and study the plasma parameters. Recently this ion source has been neutralized and it adapted to a neutral beam source; and it used to heat a cylindrical DC plasma and the plasma of ALVAND IIC Tokamak which is a small research tokamak with a minor radius of 12.6 cm, and a major radius of 45.5 cm. In this paper we report the neutralization of the ion beam and the results obtained by injection of this neutral beam into plasmas. (author) 2 refs., 4 figs

  20. Submillimeter wave propagation in tokamak plasmas

    International Nuclear Information System (INIS)

    Ma, C.H.; Hutchinson, D.P.; Staats, P.A.; Vander Sluis, K.L.; Mansfield, D.K.; Park, H.; Johnson, L.C.

    1986-01-01

    Propagation of submillimeter waves (smm) in tokamak plasma was investigated both theoretically and experimentally to ensure successful measurements of electron density and plasma current distributions in tokamak devices. Theoretical analyses were carried out to study the polarization of the smm waves in TFTR and ISX-B tokamaks. A multichord smm wave interferometer/polarimeter system was employed to simultaneously measure the line electron density and poloidal field-induced Faraday rotation in the ISX-B tokamak. The experimental study on TFTR is under way. Computer codes were developed and have been used to study the wave propagation and to reconstruct the distributions of plasma current and density from the measured data. The results are compared with other measurements. 5 references, 2 figures