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Sample records for textor tokamak measured

  1. Spectral measurements of runway electrons in the TEXTOR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Kudyakov, Timur

    2009-07-22

    The generation of multi-MeV runaway electrons is a well known effect related to the plasma disruptions in tokamaks. The runaway electrons can substantially reduce the lifetime of the future tokamak ITER. In this thesis physical properties of runaway electrons and their possible negative effects on ITER have been studied in the TEXTOR tokamak. A new diagnostic, a scanning probe, has been developed to provide direct measurements of the absolute number of runaway electrons coming from the plasma, its energy distribution and the related energy load in the material during low density (runaway) discharges and during disruptions. The basic elements of the probe are YSO crystals which transform the energy of runaway electrons into visible light which is guided via optical fibres to photomultipliers. In order to obtain the energy distribution of runaways, the crystals are covered with layers of stainless steel (or tungsten in two earlier test versions) of different thicknesses. The final probe design has 9 crystals and can temporally and spectrally resolve electrons with energies between 4 MeV and 30 MeV. The probe is tested and absolutely calibrated at the linear electron accelerator ELBE in Rossendorf. The measurements are in good agreement with Monte Carlo simulations using the Geant4 code. The runaway transport in the presence of the internal and externally applied magnetic perturbations has been studied. The diffusion coefficient and the value of the magnetic fluctuation for runaways were derived as a function of B{sub t}. It was found that an increase of runaway losses from the plasma with the decreasing toroidal magnetic field is accompanied with a growth of the magnetic fluctuation in the plasma. The magnetic shielding picture could be confirmed which predicts that the runaway loss occurs predominantly for low energy runaways (few MeV) and considerably less for the high energy ones. In the case of the externally applied magnetic perturbations by means of the dynamic

  2. Fast-ion dynamics in the TEXTOR tokamak measured by collective Thomson scattering

    DEFF Research Database (Denmark)

    Bindslev, H.; Nielsen, S.K.; Porte, L.

    2006-01-01

    Here we present the first measurements by collective Thomson scattering of the evolution of fast-ion populations in a magnetically confined fusion plasma. 150 kW and 110 Ghz radiation from a gyrotron were scattered in the TEXTOR tokamak plasma with energetic ions generated by neutral beam injection...

  3. Measurements of plasma composition in the TEXTOR tokamak by collective Thomson scattering

    DEFF Research Database (Denmark)

    Stejner Pedersen, Morten; Korsholm, Søren Bang; Nielsen, Stefan Kragh

    2012-01-01

    We demonstrate the use of collective Thomson scattering (CTS) for spatially localized measurements of the isotopic composition of magnetically confined fusion plasmas. The experiments were conducted in the TEXTOR tokamak by scattering millimeter-wave probe radiation off plasma fluctuations...... with wave vector components nearly perpendicular to the magnetic field. Under such conditions the sensitivity of the CTS spectrum to plasma composition is enhanced by the spectral signatures of the ion cyclotron motion and of weakly damped ion Bernstein waves. Recent experiments on TEXTOR demonstrated...

  4. Fast-ion dynamics in the TEXTOR tokamak measured by collective Thomson scattering

    DEFF Research Database (Denmark)

    Bindslev, Henrik; Nielsen, Stefan Kragh; Porte, L.

    2007-01-01

    The dynamics of fast ion populations in the TEXTOR tokamak are measured by collective Thomson scattering of millimetre wave radiation generated by a gyrotron operated at 110 GHz and 100-150 kW. Temporal evolution of the energetic ion velocity distribution at switch on of neutral beam injection (NBI......) and the slowdown after switch off of NBI are measured. The turn on phase of the NBI has, furthermore, been measured in plasmas with a range of electron densities and temperatures. All of these measurements are shown to be in good agreement with simple Fokker-Planck modelling. Bulk ion rotation velocity is also...

  5. Dynamics of fast ions during sawtooth oscillations in the TEXTOR tokamak measured by collective Thomson scattering

    DEFF Research Database (Denmark)

    Nielsen, Stefan Kragh; Salewski, Mirko; Bindslev, Henrik

    2011-01-01

    Experimental investigations of sawteeth interaction with fast ions measured by collective Thomson scattering on TEXTOR are presented. Time-resolved measurements of localized 1D fast-ion distribution functions allow us to study fast-ion dynamics during several sawtooth cycles. Sawtooth oscillation...

  6. Supersonic helium beam diagnostic for fluctuation measurements of electron temperature and density at the Tokamak TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Kruezi, U.; Stoschus, H.; Schweer, B.; Sergienko, G.; Samm, U. [Institute of Energy and Climate Research, Plasma Physics, Forschungszentrum Juelich GmbH, Association EURATOM-FZJ, Partner in the Trilateral Euregio Cluster, Juelich (Germany)

    2012-06-15

    A supersonic helium beam diagnostic, based on the line-ratio technique for high resolution electron density and temperature measurements in the plasma edge (r/a > 0.9) was designed, built, and optimised at TEXTOR (Torus Experiment for Technology Oriented Research). The supersonic injection system, based on the Campargue skimmer-nozzle concept, was developed and optimised in order to provide both a high neutral helium beam density of n{sub 0}= 1.5 Multiplication-Sign 10{sup 18} m{sup -3} and a low beam divergence of {+-}1 Degree-Sign simultaneously, achieving a poloidal resolution of {Delta}{sub poloidal}= 9 mm. The setup utilises a newly developed dead volume free piezo valve for operation in a high magnetic field environment of up to 2 T with a maximum repetition rate of 80 Hz. Gas injections are realised for a duration of 120 ms at a repetition rate of 2 Hz (duty cycle 1/3). In combination with a high sensitivity detection system, consisting of three 32 multi-channel photomultipliers (PMTs), measurements of edge electron temperature and density with a radial resolution of {Delta}{sub radial}= 2 mm and a maximum temporal resolution of {Delta}t Asymptotically-Equal-To 2 {mu}s (470 kHz) are possible for the first time. The diagnostic setup at TEXTOR is presented. The newly developed injection system and its theoretical bases are discussed. The applicability of the stationary collisional-radiative model as basis of the line-ratio technique is shown. Finally, an example of a fluctuation analysis demonstrating the unique high temporal and spatial resolution capabilities of this new diagnostic is presented.

  7. Charge exchange recombination in X-ray spectra of He-like argon measured at the tokamak TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Schlummer, Tobias

    2014-06-16

    Charge exchange recombination between ions and atomic hydrogen is an important atomic process in magnetically confined fusion plasmas. Besides radiative cooling of the plasma edge, charge exchange causes modifications of the ionization balance and the population densities of excited ion states. The central goal of this work is to investigate the influence of charge exchange on X-ray spectra measured at the tokamak TEXTOR. A new 2D X-ray spectrometer developed for future use at the stellarator W7-X was recently installed at TEXTOR. The spectrometer is optimized for measuring the K{sub α}-spectrum of He-like argon (1s2l - 1s{sup 2}) at wavelengths close to 4 Aa. K{sub α}-spectroscopy on He-like impurity ions is an established diagnostic for electron and ion temperature measurements in fusion plasmas. Still, up to now the observed intensity ratios of the K{sub α}-lines and their associated satellites are not fully understood. They show significant deviations from the predictions made by basic corona models. In the past charge exchange with the neutral particle background and radial impurity transport have been discussed as likely explanations. Yet a detailed description of the experimental spectra still has not been achieved. To reconstruct the 2D K{sub α}-spectra measured at TEXTOR the radial argon ion distribution is modeled using an impurity transport code. The model accounts for charge exchange and transport on basis of given radial profiles of the neutral particle density n{sub 0}(r) and the diffusion coefficient D {sub perpendicular} {sub to} (r). The theoretical spectrum is then constructed based on the processes relevant for line emission. Within an iterative procedure n{sub 0}(r) and D {sub perpendicular} {sub to} (r) are varied until consistency between the theoretical and the experimental spectra is achieved. It is shown that the 2D K{sub α}-spectra allow a clear distinction of charge exchange and transport effects, ensuring unique solutions for n

  8. Comparison of measured and simulated fast ion velocity distributions in the TEXTOR tokamak

    DEFF Research Database (Denmark)

    Moseev, Dmitry; Meo, Fernando; Korsholm, Søren Bang

    2011-01-01

    Here we demonstrate a comprehensive comparison of collective Thomson scattering (CTS) measurements with steady-state Monte Carlo simulations performed with the ASCOT and VENUS codes. The measurements were taken at a location on the magnetic axis as well as at an off-axis location, using two...... projection directions at each location. The simulations agree with the measurements on-axis, but for the off-axis geometries discrepancies are observed for both projection directions. For the near perpendicular projection direction with respect to the magnetic field, the discrepancies between measurement...... and simulations can be explained by uncertainty in plasma parameters. However, the discrepancies between measurement and simulations for the more parallel projection direction cannot be explained solely by uncertainties in plasma parameters. Here anomalous fast ion transport is a possible explanation...

  9. Ageing of structural materials in tokamaks: TEXTOR liner study

    Science.gov (United States)

    Weckmann, A.; Petersson, P.; Rubel, M.; Fortuna-Zaleśna, E.; Zielinski, W.; Romelczyk-Baishya, B.; Grigore, E.; Ruset, C.; Kreter, A.

    2017-12-01

    After the final shut-down of the tokamak TEXTOR, all of its machine parts became accessible for comprehensive studies. This unique opportunity enabled the study of the Inconel 625 liner by a wide range of methods. The aim was to evaluate eventual alteration of surface and bulk characteristics from recessed wall elements that may influence the machine performance. The surface was covered with stratified layers consisting mainly of boron, carbon, oxygen, and in some cases also silicon. Wall conditioning and limiter materials hence predominantly define deposition on the liner. Deposited layers on recessed wall elements reach micrometre thickness within decades, peel off and may contribute to the dust inventory in tokamaks. Deuterium content was about 4,7 at% on average most probably due to wall conditioning with deuterated gas, and very low concentration in the Inconel substrate. Inconel 625 retained its mechanical strength despite 26 years of cyclic heating, stresses and particle bombardment.

  10. Systematic design and simulation of a tearing mode suppression feedback control system for the TEXTOR tokamak

    NARCIS (Netherlands)

    Hennen, B.A.; Westerhof, E.; Nuij, Pwjm; M.R. de Baar,; Steinbuch, M.

    2012-01-01

    Suppression of tearing modes is essential for the operation of tokamaks. This paper describes the design and simulation of a tearing mode suppression feedback control system for the TEXTOR tokamak. The two main control tasks of this feedback control system are the radial alignment of electron

  11. Statistical description of intermittent events in the plasma edge of the TEXTOR tokamak; Statistische Beschreibung von intermittenten Ereignissen in der Randschicht des Tokamaks TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Schmitz, D.

    2006-07-15

    Within the scope of this work itermittent events in the plasma edge of the tokamak TEXTOR were characterized. For the data of measurements of the density and the poloidal electrical field were analysed. The data was collected by a reciprocating and a fixed probe as well as by a lithium beam. The intermittent behaviour was quantified by the statistical moments of the data. If intermittency is high, coherent structures (also called blobs) can be detected. The detected blobs were described using the statistical method of conditional averaging. The main results can be summarised as follows: Intermittent behavoiur has been detected in the scrap off layer of the tokamak TEXTOR and it is increasing with the radius from the last closed flux surface (LCFS) on. On the midplane the blobs in the limiter geometry have a radial size of up to 8 cm and move onto the wall with velocities as high as (1-7)% of the ion sound speed. It was found that intermittent transport causes 40% of the total perpendicular transport in the investigated discharges. In the upper part of the tokamak there is less intermittency. This is reasonable if intermittency is caused by interchange instabilities which mainly occur on the low field side of the tokamak. With the Dynamic Ergodic Divertor (DED) and the associated formation of tearing modes intermittency is increasing. This can also be due to the steeper gradient of density in the scrap off layer close to the LCFS which is caused by gas puffing used for the regulation of the density. Outside the LCFS the ergodic field does not have any influence on the characteristics of blobs. Within the LCFS density holes have been found which propagate towards the centre of the plasma. The radial transport due to blobs is still the same. In general the velocity of the detected blobs is proportional to the square root of their poloidal size. That confirms the prediction of the blob model in which the nonlinear development of interchange instabilities causes the

  12. Whole-machine material migration studies in the TEXTOR tokamak with molybdenum

    Directory of Open Access Journals (Sweden)

    A. Weckmann

    2017-08-01

    Full Text Available MoF6 injection from a localised source into plasma edge in the TEXTOR tokamak was the last experiment before the final shut-down of the TEXTOR machine. During decommissioning all plasma-facing components (PFCs became available for surface studies. Detailed mapping of Mo deposition was performed in order to determine its migration on global scale. The concentration of Mo on PFC decays exponentially with distance from the source. The decay length is of the order of 0.1m on the main PFC and 1m on the receded components. Also the decay lengths modelled with the ERO code are between 0.15–1.3m, depending on the anomalous cross-field diffusion coefficient. The inner bumper limiter is found to be the major repository for Mo. Material balance measurements show that only up to 22% of the injected Mo was detected on all the PFCs thus indicating that a large fraction of injected Mo may have been pumped out before being deposited.

  13. Exposure of CFC-materials to high transient heat loads in the TEXTOR tokamak

    Science.gov (United States)

    Scholz, T.; Boedo, J.; Bolt, H.; Duwe, R.; Finken, K. H.; Gray, D.; Hassanein, A.

    1997-02-01

    Transient high heat flux events like ELMs, vertical displacement events and disruptions can cause the thermal ablation of plasma facing material. Until now experimental work in this field had been carried out by exposing material specimens to heat loads by electron or laser beam or by tests in pulsed plasma accelerators. In the present work carbon specimens were directly exposed to intense plasma fluxes in the TEXTOR tokamak. The exposure was performed with a fast probe allowing the insertion of the material over a distance of 9 cm into the edge plasma for a duration of 80 ms. The results of in-situ diagnostic measurements and of the post-experiment examination of the specimens are compared with a reference experiment by electron beam and with numerical analyses. Results indicated that the heat flux to the probe surfaces and the probe erosion is much lower than expected.

  14. Overview of wall probes for erosion and deposition studies in the TEXTOR tokamak

    Directory of Open Access Journals (Sweden)

    M. Rubel

    2017-05-01

    Full Text Available An overview of diagnostic tools – test limiters and collector probes – used over the years for material migration studies in the TEXTOR tokamak is presented. Probe transfer systems are shown and their technical capabilities are described. This is accompanied by a brief presentation of selected results and conclusions from the research on material erosion – deposition processes including tests of candidate materials (e.g. W, Mo, carbon-based composites for plasma-facing components in controlled fusion devices. The use of tracer techniques and methods for analysis of materials retrieved from the tokamak are summarized. The impact of research on the reactor wall technology is addressed.

  15. Electron transport in the plasma edge with rotating resonant magnetic perturbations at the TEXTOR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Stoschus, Henning

    2011-10-13

    Small three-dimensional (3D) magnetic perturbations can be used as a tool to control the edge plasma parameters in magnetically confined plasmas in high confinement mode (''H-mode'') to suppress edge instabilities inherent to this regime, the Edge Localized Modes (ELMs). In this work, the impact of rotating 3D resonant magnetic perturbation (RMP) fields on the edge plasma structure characterized by electron density and temperature fields is investigated. We study a low confinement (L-mode) edge plasma (r/a>0.9) with high resistivity (edge electron collisionality {nu}{sup *}{sub e}>4) at the TEXTOR tokamak. The plasma structure in the plasma edge is measured by a set of high resolution diagnostics: a fast CCD camera ({delta}t=20 {mu}s) is set up in order to visualize the plasma structure in terms of electron density variations. A supersonic helium beam diagnostic is established as standard diagnostic at TEXTOR to measure electron density n{sub e} and temperature T{sub e} with high spatial ({delta}r=2 mm) and temporal resolution ({delta}t=20 {mu}s). The measured plasma structure is compared to modeling results from the fluid plasma and kinetic neutral transport code EMC3-EIRENE. A sequence of five new observations is discussed: (1) Imaging of electron density variations in the plasma edge shows that a fast rotating RMP field imposes an edge plasma structure, which rotates with the external RMP rotation frequency of vertical stroke {nu}{sub RMP} vertical stroke =1 kHz. (2) Measurements of the electron density and temperature provide strong experimental evidence that in the far edge a rotating 3D scrape-off layer (SOL) exists with helical exhaust channels to the plasma wall components. (3) Radially inward, the plasma structure at the next rational flux surface is found to depend on the relative rotation between external RMP field and intrinsic plasma rotation. For low relative rotation the plasma structure is dominated by a particle and energy loss

  16. The influence of the dynamic ergodic divertor on the radial electric field at the Tokamak TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Coenen, Jan Willem

    2009-11-06

    In this work the influence of external Resonant Magnetic Perturbations (RMPs) on the radial electric field Er in magnetically confined plasmas is investigated by Charge Exchange Recombination Spectroscopy (CXRS) at the Tokamak TEXTOR. Here, the RMPs are produced with the Dynamic Ergodic Divertor (DED), a set of 16 helical perturbation coils located at the high field side of TEXTOR. Within this work, the base mode number of perturbations has been m/n=6/2. We have first investigated the influence of external torque from neutral heating beams on plasma rotation and E{sub r}. The ergodic zone causes an electron loss, and subsequently a (vector)j x (vector)B force driven by the compensating ion return current. In addition, the DED changes the global confinement properties. Depending on the edge safety factor (''field line twist'') q{sub a}, either increased or decreased particle confinement is observed. In case of the increased particle confinement (IPC) the increase in density (40%) and particle confinement time {tau}{sub p} (30%) is correlated to the connection of field lines at the q=5/2 surface to the DED target, locally changing the transport properties and the E{sub r}. Transport is reduced and the E{sub r} shear is increased locally at q=5/2 up to 1.5 . 10{sup 5}s{sup -1}, while the E{sub r} becomes more positive. (orig.)

  17. Multi scale study of carbon deposits collected in Tore-Supra and TEXTOR tokamaks; Etude multi echelle des depots carbones collectes dans les tokamaks Tore Supra et TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Richou, M

    2007-06-15

    Tokamaks are devices aimed at studying magnetic fusion. They operate with high temperature plasmas containing hydrogen, deuterium or tritium. One of the major issue is to control the plasma-wall interaction. The plasma facing components are most often in carbon. The major drawback of carbon is the existence of carbon deposits and dust, due to erosion. Dust is potentially reactive in case of an accidental opening of the device. These deposits also contain H, D or T and induce major safety problems when tritium is used, which will be the case in ITER. Therefore, the understanding of the deposit formation and structure has become a main issue for fusion researches. To clarify the role of the deposits in the retention phenomenon, we have done different complementary characterizations for deposits collected on similar places (neutralizers) in tokamaks Tore Supra (France) and TEXTOR (Germany). Accessible microporous volume and pore size distribution of deposits has been determined with the analysis of nitrogen and methane adsorption isotherms using the BET, Dubinin-Radushkevich and {alpha}{sub s} methods and the Density Functional Theory (DFT). To understand growth mechanisms, we have studied the deposit structure and morphology. We have shown using Transmission Electron Microscopy (TEM) and Raman micro-spectrometry that these deposits are non amorphous and disordered. We have also shown the presence of nano-particles (diameter between 4 and 70 nm) which are similar to carbon blacks: nano-particle growth occurs in homogeneous phase in the edge plasma. We have emphasised a dual growth process: a homogenous and a heterogeneous one. (author)

  18. Enhanced particle confinement and turbulence reduction due to E × B shear in the TEXTOR tokamak

    Science.gov (United States)

    Boedo, J.; Gray, D.; Jachmich, S.; Conn, R.; Terry, G. P.; Tynan, G.; Van Oost, G.; Weynants, R. R.; TEXTOR Team

    2000-07-01

    Positive radial electric fields have been created at the edge of the TEXTOR tokamak plasma using an electrode. The electric field induces a thin (δr~1.5 cm), E × B driven layer at the edge rotating poloidally at 12-20 km/s and featuring high shear. Concomitant changes in the density and poloidal electric field fluctuations and their cross-phase in the shear layer result in suppression of radial turbulent particle transport, even at low radial electric field strength. Temperature fluctuations are reduced, resulting in diminished turbulent heat flux. As turbulent particle transport is quenched, the particle confinement time τp increases by a factor of 2 and the energy confinement time τE by 20%. Turbulent transport accounts for ~50% of the total particle flux. Both the cross-phase and the density fluctuations are sensitive to the sign of ∇Er.

  19. Collective Thomson scattering measurements with high frequency resolution at TEXTOR

    DEFF Research Database (Denmark)

    Stejner Pedersen, Morten; Nielsen, Stefan Kragh; Korsholm, Søren Bang

    2010-01-01

    We discuss the development and first results of a receiver system for the collective Thomson scattering (CTS) diagnostic at TEXTOR with frequency resolution in the megahertz range or better. The improved frequency resolution expands the diagnostic range and utility of CTS measurements in general...

  20. Microwave Imaging Reflectometer for TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    T. Munsat; E. Mazzucato; H. Park; B.H. Deng; C.W. Domier; N.C. Luhmann, Jr.; J. Wang; Z.G. Xia; A.J.H. Donne; and M. van de Pol

    2002-07-09

    Understanding the behavior of fluctuations in magnetically confined plasmas is essential to the advancement of turbulence-based transport physics. Though microwave reflectometry has proven to be an extremely useful and sensitive tool for measuring small density fluctuations in some circumstances, this technique has been shown to have limited viability for large amplitude, high kq fluctuations and/or core measurements. To this end, a new instrument based on 2-D imaging reflectometry has been developed to measure density fluctuations over an extended plasma region in the TEXTOR tokamak. This technique is made possible by collecting an extended spectrum of reflected waves with large-aperture imaging optics. Details of the imaging reflectometry concept, as well as technical details of the TEXTOR instrument will be presented. Data from roof-of-principle experiments on TEXTOR using a prototype system is presented, as well as results from a systematic off-line study of the advantages and limitations of the imaging reflectometer.

  1. Thermal instability theory analysis of multifaceted asymmetric radiation from the edge (MARFE) in Tokamak Experiment for Technology Oriented Research (TEXTOR)

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, F. A.; Stacey, W. M.; Rapp, J.; Brix, M.

    2001-07-01

    The density limits for a series of shots in TEXTOR [Tokamak Experiment for Technology Oriented Research, E. Hintz, P. Bogen, H. A. Claa{ss}en , in Contributions to High-Temperature Plasma Physics, edited by K. H. Spatschek and J. Uhlenbusch (Akademie Verlag, Berlin, 1994, p. 373)], over a range of heating powers, that ended in multifaceted asymmetric radiation from the edge (MARFE) have been analyzed within the context of thermal instability theory. The prediction of MARFE onset agrees with observation to within the experimental uncertainty.

  2. Temporally resolved plasma composition measurements by collective Thomson scattering in TEXTOR

    DEFF Research Database (Denmark)

    Stejner Pedersen, Morten; Korsholm, Søren Bang; Nielsen, Stefan Kragh

    2012-01-01

    Fusion plasma composition measurements by collective Thomson scattering (CTS) were demonstrated in recent proof-of-principle measurements in TEXTOR [S. B. Korsholm et al., Phys. Rev. Lett. 106, 165004 (2011)]. Such measurements rely on the ability to resolve and interpret ion cyclotron structure...

  3. Operation and upgrade of diagnostic neutral beam injector RUDI at TEXTOR tokamak.

    Science.gov (United States)

    Listopad, A A; Coenen, J W; Davydenko, V I; Deichuli, P P; Ivanov, A A; Mishagin, V V; Savkin, V Ya; Schalt, W; Schweer, B; Shulzhenko, G I; Stupishin, N V; Uhlemann, R

    2010-02-01

    The status and the executing modernization of RUssian Diagnostic Injector (RUDI) are described. The ion source consists of arc plasma emitter and multiaperture four-electrode ion optical system. The present ion optical system with round beamlets is to be replaced by new slit apertures system for the reducing beam angular divergence in one direction. Due to enlarged dimensions and transparency of new ion optical system the extracted ion beam current will be by 50% increased. For the extension of beam pulse duration from 4 s to 8-10 s an optimized metal-ceramic arc-discharge channel is introduced. In the paper, the optical measurements results of beam parameters, including the profile of species distribution, scanned by custom-built multichannel spectroscope, are also presented.

  4. Liquid Scintillation Detectors for Gamma and Neutron Diagnostic at Textor and Results of Runaway and Sawtooth Oscillations

    NARCIS (Netherlands)

    Hoenen, F.; Graffmann, E.; Finken, K.H.; Barrenscheen, H. J.; Klein, H.; R. Jaspers,

    1994-01-01

    Time and energy resolved neutron and gamma measurements are performed at the TEXTOR tokamak with a fast liquid NE-213 scintillator. To distinguish between neutron and gamma (gamma)-ray induced events, pulse shape discrimination is used. To suppress scattered radiation, the detector is installed in

  5. Tokamak Plasmas: Internal magnetic field measurement in tokamak ...

    Indian Academy of Sciences (India)

    The theory of the measurement and a detailed design of the Zeeman polarimeter constructed to measure the poloidal field profile in the ADITYA tokamak are presented. The Fabry-Perot which we have employed in our design, with photodiode arrays followed by lock-in detection of the polarization signal, allows the ...

  6. Electron temperature dynamics of TEXTOR plasmas

    NARCIS (Netherlands)

    Udintsev, Victor Sergeevich

    2003-01-01

    To study plasma properties in the presence of large and small MHD modes, new high-resolution ECE diagnostics have been installed at TEXTOR tokamak, and some of the already existing systems have been upgraded. Two models for the plasma transport properties inside large m/n = 2/1 MHD islands have been

  7. Runaway snakes in TEXTOR-94

    NARCIS (Netherlands)

    Entrop, I.; R. Jaspers,; Cardozo, N. J. L.; Finken, K.H.

    1999-01-01

    Observations of a runaway beam confined in an island-like structure, a so-called runaway snake, are reported. The observations are made in TEXTOR-94 by measurement of synchrotron radiation emitted by these runaways. A full poloidal View allows for the study of the synchrotron pattern of the snake to

  8. Tokamak Plasmas: Electron temperature $(T_ {e}) $ measurements ...

    Indian Academy of Sciences (India)

    Thomson scattering technique based on high power laser has already proved its superoirity in measuring the electron temperature (e) and density (e) in fusion plasma devices like tokamaks. The method is a direct and unambiguous one, widely used for the localised and simultaneous measurements of the above ...

  9. Charge exchange spectroscopy as a fast ion diagnostic on TEXTOR

    NARCIS (Netherlands)

    Delabie, E.; Jaspers, R. J. E.; von Hellermann, M. G.; Nielsen, S.K.; Marchuk, O.

    2008-01-01

    An upgraded charge exchange spectroscopy diagnostic has been taken into operation at the TEXTOR tokamak. The angles of the viewing lines with the toroidal magnetic field are close to the pitch angles at birth of fast ions injected by one of the neutral beam injectors. Using another neutral beam for

  10. Antenna coupling study for ICWC plasma characterization in TEXTOR

    Indian Academy of Sciences (India)

    2013-01-06

    Jan 6, 2013 ... In order to extrap- olate the ICRF wall conditioning to ITER, experimental studies on ion cyclotron wall conditioning (ICWC) were carried out in tokamak TEXTOR, with conventional ICRF antennas, to simulate the scenario of ITER wall conditioning at half-field. The plasma generation technique in the ICRF ...

  11. Spectra polarimetry of the motional Stark effect at TEXTOR-94

    NARCIS (Netherlands)

    R. Jaspers,; Elzendoorn, B. S. Q.; Donne, A. J. H.; Soetens, T.

    2001-01-01

    The Balmer-alpha light emitted by neutral particles injected into tokamaks is polarized with respect to the Lorentz electric field experienced by these atoms: E-l = nu xB, known as the motional Stark effect (MSE). On TEXTOR-94 a new MSE system is under development which exploits the full spectral

  12. Wall reflection modeling for charge exchange recombination spectroscopy (CXRS) measurements on Textor and ITER

    NARCIS (Netherlands)

    Banerjee, S.; Vasu, P.; von Hellermann, M.; Jaspers, R. J. E.

    2010-01-01

    Contamination of optical signals by reflections from the tokamak vessel wall is a matter of great concern. For machines such as ITER and future reactors, where the vessel wall will be predominantly metallic, this is potentially a risk factor for quantitative optical emission spectroscopy. This is,

  13. Suppression of the runaway electron generation by massive gas injection at TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Lvovskiy, Andrey; Koslowski, Hans R. [Institute of Energy and Climate Research - Plasma Physics, Forschungszentrum Juelich GmbH, Juelich (Germany); Zeng, Long [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China)

    2014-07-01

    Runaway electrons (RE) are a serious threat for the first wall of the ITER tokamak. The mitigation of RE may be an insufficient action for the safety of such large tokamak. A safer approach is to completely suppress the generation of RE in ITER. Massive gas injection (MGI) may be one of the possible techniques for the suppression of RE generation. However, there is still no clear evidence that MGI effects so. TEXTOR tokamak is well-equipped for the MGI investigation. A small disruption mitigation valve (DMV) can inject an amount of particles up to 0.25 bar*liter in order to trigger the disruption and reliably generate RE. A larger DMV injects up to 9 bar*liter of Ar, Ne or He to suppress the RE due to collisions. The electron density is measured during disruption by a dispersion interferometer with time resolution of 2 mks for the reference to Connor-Hastie-Rosenbluth density. The aim of MGI experiments at TEXTOR was to determine the influence of species, amount of injected gas and the time delay between DMVs on suppression of RE generation. The suppression of RE generation in case of sufficient MGI before the current quench has been observed.

  14. Charge exchange spectroscopy as a fast ion diagnostic on TEXTOR

    DEFF Research Database (Denmark)

    Delabie, E.; Jaspers, R.J.E.; von Hellermann, M.G.

    2008-01-01

    for active spectroscopy, injected counter the direction in which fast ions injected by the first beam are circulating, we can simultaneously measure a fast ion tail on the blue wing of the D-alpha spectrum while the beam emission spectrum is Doppler shifted to the red wing. An analysis combining the two......An upgraded charge exchange spectroscopy diagnostic has been taken into operation at the TEXTOR tokamak. The angles of the viewing lines with the toroidal magnetic field are close to the pitch angles at birth of fast ions injected by one of the neutral beam injectors. Using another neutral beam...... parts of the spectrum offers possibilities to improve the accuracy of the absolute (fast) ion density profiles. Fast beam modulation or passive viewing lines cannot be used for background subtraction on this diagnostic setup and therefore the background has to be modeled and fitted to the data together...

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

    Energy Technology Data Exchange (ETDEWEB)

    Moseev, D.

    2011-11-15

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

  16. Temporal evolution of confined fast-ion velocity distributions measured by collective Thomson scattering in TEXTOR

    DEFF Research Database (Denmark)

    Nielsen, Stefan Kragh; Bindslev, Henrik; Porte, L.

    2008-01-01

    Fast ions created in the fusion processes will provide up to 70% of the heating in ITER. To optimize heating and current drive in magnetically confined plasmas insight into fast-ion dynamics is important. First measurements of such dynamics by collective Thomson scattering (CTS) were recently...

  17. Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Meglicki, Z

    1995-09-19

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

  18. In situ detection of hydrogen retention in TEXTOR by laser induced desorption

    Energy Technology Data Exchange (ETDEWEB)

    Schweer, B., E-mail: B.Schweer@fz-juelich.d [Forschungszentrum Juelich GmbH, Institute of Energy Research, IEF-4 Plasma Physis, Association EURATOM-FZJ, Trilateral Euregio Cluster, 52425 Juelich (Germany); Irrek, F.; Zlobinski, M.; Huber, A.; Sergienko, G.; Brezinsek, S.; Philipps, V.; Samm, U. [Forschungszentrum Juelich GmbH, Institute of Energy Research, IEF-4 Plasma Physis, Association EURATOM-FZJ, Trilateral Euregio Cluster, 52425 Juelich (Germany)

    2009-06-15

    Long term tritium retention is one of the most critical issues for ITER and future fusion devices. While a global analysis of the T retention can be made by T accountancy in the activated phase of ITER, fuel retention and control must be already addressed in the non- activated phase, to identify the mechanism, location and amount of retention, its dependence on plasma and wall conditions and to qualify T retention mitigation and control techniques. For this purpose a new diagnostic, laser induced desorption spectroscopy of retained fuel has been developed and applied in TEXTOR. Hydrogen isotopes are desorbed from re-deposited layers on graphite plates by rapid heating with laser radiation. The released particles have been quantified in situ by spectroscopic measurements of hydrogen lines in a tokamak plasma. The results were validated by ex situ analysis of the hydrogen content of deposited a-C:H layers.

  19. Measurement of electron density profile by microwave reflectometry on tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Simonet, F.

    1985-05-01

    A new method for measuring the electron density spatial profile has been successfully tested on the tokamak of Fontenay aux Roses (TFR). This method is based on the total reflection experienced by a wave of frequency F on the layer where F = F/sub p/e(r). The experimental results show that the maximum electron density in the discharge is also easily measured and that accurate determination of a density profile can be obtained with a time resolution of 5 ms. This diagnostic is well adapted to all fusion devices where access to the total plasma cross section is limited, particularly for large tokamaks.

  20. Internal magnetic field measurement in tokamak plasmas using a ...

    Indian Academy of Sciences (India)

    There is a growing interest in developing a reliable method for the measurement of the in- ternal magnetic field in high ... This information is essential for understanding confinement, stability and energy balance of the tokamak plasma. .... The instrument measures the difference between the left-hand and right-hand circularly ...

  1. First experience with the Dynamic Ergodic Divertor on TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Finken, K.H.; Abdullaev, S.S.; Jakubowski, M.; Kobayshi, M.; Lehnen, M.; Matsunaga, G.; Pospieszczyk, A.; Schweer, B.; Sergienko, G.; Wolf, R. [Inst. fuer Plasmaphysik, Forschungszentrum Juelich GmbH, EURATOM Association, Partner in the Trilateral Euregio Cluster, Juelich (Germany)

    2004-07-01

    The Dynamic Ergodic Divertor, DED, is a new experiment on the TEXTOR tokamak in Juelich. The DED consists of a set of coils with DC or AC (4 phases) electrical currents flowing parallel to the magnetic field lines. This causes a braiding of the magnetic flux tubes which is called ergodization. The strongly deflected field lines at the plasma edge form the laminar zone. The dynamic operation of the DED (AC current operation) should distribute the heat load to a large surface area and possibly induce a rotation of the plasma. First results are discussed. (orig.)

  2. Toroidal flow measurement in CT injected STOR-M tokamak

    Science.gov (United States)

    Asai, Tomohiko; Morelli, Jordan; Singh, Ajay; Xiao, Chijin; Hirose, Akira; Nagata, Masayoshi; Uyama, Tadao

    2002-11-01

    Compact Torus (CT) injection is a technology being developed for fueling of large tokamak reactors. It has been demonstrated in the STOR-M tokamak that tangential CT injection is capable of inducing an improved confinement mode (H-mode). It has been conjectured that tangential CT injection may enhance the toroidal rotation of the bulk tokamak plasma which is responsible for the H-mode by preventing or reducing microinstabilities[1]. In order to investigate the mechanisms of the L-H transition induced by enhanced toroidal flow (particularly that caused by CT injection), an Ion Doppler Spectroscope (IDS) has been developed. The IDS employs a 0.75 m focal length Czerny-Turner spectrometer with a resolution of 0.1 Åand a 16-channel PMT array. Data of plasma flow measurements will be presented with and without CT injection. Also, the results will be compared with toroidal flow measurement obtained using a 4-sided Mach probe in the plasma edge region. [1] S. Sen et al., Phys. Rev. Lett. 88, 185001 (2002).

  3. Materials analysis of TEXTOR limiter tiles

    Science.gov (United States)

    Doerner, R.; Mills, B. E.; Wallura, E.; Walsh, D. S.; Chevalier, G.; Conn, R. W.; Dippel, K. H.; Doyle, B. L.; Esser, H. G.; Finken, K. H.; Gray, D.; Hirooka, Y.; Koizlik, K.; Miyahara, A.; Moyer, R. A.; Watkins, J. G.; Winter, J.

    1990-12-01

    Graphite tiles from both the ALT-II and inner-bumper limiters were removed from TEXTOR and subjected to materials analysis. Scanning-electron microscopy and energy dispersive X-ray analysis were performed at the Institut für Reaktorwerkstoffe, Forschungszentrum Julich. Deuterium profiles and metallic contamination were examined using external ion beam analysis at Sandia National Laboratory-Albuquerque. The erosion and hydrogen recycling of the tiles, while subjected to plasma bombardment, were studied at University of California, Los Angeles. In-situ analysis of the inner-bumper limiter tiles was performed by Sandia National Laboratory-Livermore using beta backscattering. Results indicate low metallic impurity concentration on the surfaces of both types of tiles. Increased metallic concentration coincides with regions of increased plasma flux to the surface. The ALT-II tiles exhibit a uniformly eroded surface. The inner-bumper limiter tiles show both eroded and redeposited regions, in agreement with power deposition measurements to the tiles in TEXTOR. The redeposited regions show enhanced erosion and recycling when exposed to controlled plasma bombardment.

  4. Helical modulation of the electrostatic plasma potential due to edge magnetic islands induced by resonant magnetic perturbation fields at TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Ciaccio, G., E-mail: giovanni.ciaccio@igi.cnr.it; Spizzo, G. [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete SpA), Corso Stati Uniti 4, 35127 Padova (Italy); Schmitz, O., E-mail: oschmitz@wisc.edu; Frerichs, H. [Department of Engineering Physics, University of Wisconsin–Madison, 1500 Engineering Drive, Madison, Wisconsin 53706 (United States); Abdullaev, S. S. [Institut für Energieforschung-Plasmaphysik, Association EURATOM-FZJ, Jülich (Germany); Evans, T. E. [General Atomics, P.O. Box 85608, San Diego, California 92121 (United States); White, R. B. [Plasma Physics Laboratory, Princeton University, P.O. Box 451, Princeton, New Jersey 08543 (United States)

    2015-10-15

    The electrostatic response of the edge plasma to a magnetic island induced by resonant magnetic perturbations to the plasma edge of the circular limiter tokamak TEXTOR is analyzed. Measurements of plasma potential are interpreted by simulations with the Hamiltonian guiding center code ORBIT. We find a strong correlation between the magnetic field topology and the poloidal modulation of the measured plasma potential. The ion and electron drifts yield a predominantly electron driven radial diffusion when approaching the island X-point while ion diffusivities are generally an order of magnitude smaller. This causes a strong radial electric field structure pointing outward from the island O-point. The good agreement found between measured and modeled plasma potential connected to the enhanced radial particle diffusivities supports that a magnetic island in the edge of a tokamak plasma can act as convective cell. We show in detail that the particular, non-ambipolar drifts of electrons and ions in a 3D magnetic topology account for these effects. An analytical model for the plasma potential is implemented in the code ORBIT, and analyses of ion and electron radial diffusion show that both ion- and electron-dominated transport regimes can exist, which are known as ion and electron root solutions in stellarators. This finding and comparison with reversed field pinch studies and stellarator literature suggest that the role of magnetic islands as convective cells and hence as major radial particle transport drivers could be a generic mechanism in 3D plasma boundary layers.

  5. Fast temperature fluctuation measurements in SOL of tokamak TCV

    DEFF Research Database (Denmark)

    Horacek, J.; Nielsen, Anders Henry; Pitts, R.A.

    coupling both across the plasma sheath and in the probe circuit itself. Comparisons are also made between the results from higher frequency sweeping and the standard values derived from a slower sweep to show that the fast measurement is reliable. Considerable effort has been expended in recent years......A fast scanning assembly has been widely used on the TCV tokamak to insert a probe head equipped with an array of single Langmuir probe tips up to the separatrix at the plasma midplane. Using fast voltage sweeping, we obtain IV-characteristics every 8 μs, allowing an estimate of the electron......-characteristics, some effort is required to demonstrate the credibility of the Te derived from the characteristics. Following the methodology proposed in [3], we use both numerical (5spice code) and lab simulations of the equivalent probe circuit, together with a simplified plasma circuit to study the capacitative...

  6. Diamagnetic loop measurement in Korea Superconducting Tokamak Advanced Research machine.

    Science.gov (United States)

    Bak, J G; Lee, S G; Kim, H S

    2011-06-01

    Diamagnetic loop (DL), which consists of two poloidal loops inside the vacuum vessel, is used to measure the diamagnetic flux during a plasma discharge in the Korea Superconducting Tokamak Advanced Research (KSTAR) machine. The vacuum fluxes in the DL signal can be compensated up to 0.1 mWb by using the coefficients, which are obtained from experimental investigations, in the vacuum flux measurements during vacuum shots under same operational conditions of magnetic coils for plasma experiment in the KSTAR machine. The maximum error in the diamagnetic flux measurement due to the errors of the coefficients was estimated as ∼0.22 mWb. From the diamagnetic flux measurements for the ohmically heated circular plasmas in the KSTAR machine, the stored energy agrees well with the estimated kinetic energy within the discrepancy of 25%. When the electron cyclotron heating, the neutral beam injection, and the ion cyclotron resonance heating are added to the ohmically heated limiter plasmas, the additional heating effects can be clearly observed from the increase of the stored energy evaluated in the DL measurement. © 2011 American Institute of Physics

  7. Commissioning of{sup textor}CC, the new TEXTOR control system and first operating experiences

    Energy Technology Data Exchange (ETDEWEB)

    Lambertz, H.T. [Institut fuer Energieforschung, Plasmaphysik, Forschungszentrum Juelich GmbH, EURATOM Association, D-52425 Juelich (Germany)], E-mail: h.t.lambertz@fz-juelich.de; Krom, J.G.; Kraemer-Flecken, A. [Institut fuer Energieforschung, Plasmaphysik, Forschungszentrum Juelich GmbH, EURATOM Association, D-52425 Juelich (Germany)

    2008-04-15

    The old TEXTOR control systems have successfully been updated. The machine control has replaced by{sup textor}CC, a solution based on the software package WinCC produced by Siemens. WinCC, and therefore{sup textor}CC, can be easily integrated with the already available Siemens S5/S7 hardware components. This new system has the advantage that it is based on industrial soft- and hardware components.Therefore, the lifetime of the control system is extended and the maintenance effort is reduced. The installation and commissioning of the new control system was done in parallel to TEXTOR operation. During this time each function was tested and compared with the actual TEXTOR data. All functionality of the former control system was step-by-step replaced. Special attention was given to the visualization, data and error logging. The machine control timing system has been replaced by an in house development in partnership with Siemens. It consists of transmitters and receivers based on PROFIBUS modules and is fully compatible with the pre-existing timing infrastructure. The old programmable function generator (PFG) has been replaced by compact RIO modules, controlled and programmed by Labview. This new PFG system allows to program up to 84 different time dependent signals. In this paper we intent to present a more detailed overview of our, on WinCC-based work, and a first status report on this new control system for TEXTOR.

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

    DEFF Research Database (Denmark)

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

    2004-01-01

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

  9. Electron cyclotron resonance heating on TEXTOR

    NARCIS (Netherlands)

    Westerhof, E.; Hoekzema, J. A.; Hogeweij, G. M. D.; Jaspers, R. J. E.; Schüller, F. C.; Barth, C. J.; Bongers, W. A.; Donne, A. J. H.; Dumortier, P.; van der Grift, A. F.; van Gorkom, J. C.; Kalupin, D.; Koslowski, H. R.; Kramer-Flecken, A.; Kruijt, O. G.; Cardozo, N. J. L.; Mantica, P.; van der Meiden, H. J.; Merkulov, A.; Messiaen, A.; Oosterbeek, J. W.; Oyevaar, T.; Poelman, A. J.; Polman, R. W.; Prins, P. R.; Scholten, J.; Sterk, A. B.; Tito, C. J.; Udintsev, V.S.; Unterberg, B.; Vervier, M.; van Wassenhove, G.

    2003-01-01

    The 110 GHz and the new 140 GHz gyrotron systems for electron cyclotron resonance heating (ECRH) and ECCD on TEXTOR are described and results of ECRH experiments with the 110 GHz system are reported. Central ECRH on Ohmic plasmas shows the presence of an internal electron transport barrier near q =

  10. Measurements of Intrinsic Ion Bernstein Waves in a Tokamak by Collective Thomson Scattering

    DEFF Research Database (Denmark)

    Korsholm, Søren Bang; Stejner Pedersen, Morten; Bindslev, Henrik

    2011-01-01

    In this Letter we report measurements of collective Thomson scattering (CTS) spectra with clear signatures of ion Bernstein waves and ion cyclotron motion in tokamak plasmas. The measured spectra are in accordance with theoretical predictions and show clear sensitivity to variation in the density...

  11. Plasma Position Measurements in a Tokamak with an Iron Core Transformer

    Science.gov (United States)

    Kwon, Gi-Chung; Choe, W.; Kim, Jayhyun; Yi, Hyo-Suk; Jeon, Sang-Jean; Huh, Songwhe; Chang, Hong-Young; Choi, Duk-In

    2000-07-01

    Two simple methods of estimating the plasma position in a large-aspect-ratio, low-βp tokamak with an iron core transformer are demonstrated: a magnetic diagnostic method and an optical method. The magnetic diagnostic method utilizes an array of magnetic pickup coils to measure the poloidal magnetic field produced by the plasma current. To include the effects of toroidicity and an iron core transformer, the correction factor was calculated with the magnetic material (or iron core) inside the calculation domain and incorporated in the analysis. The evolution of horizontal and vertical displacement of the plasma center obtained in this way is used to control the KAIST-Tokamak plasmas. To compare the plasma position estimated using the magnetic pickup coils, a simple optical method is also demonstrated on KAIST-TOKAMAK using a composite video signal from a charge-coupled device (CCD) camera. The two results are in good agreement.

  12. Antenna coupling study for ICWC plasma characterization in TEXTOR

    Indian Academy of Sciences (India)

    ... Pramana – Journal of Physics; Volume 80; Issue 1. Antenna coupling study for ICWC plasma characterization in TEXTOR. Manash Kumar Paul A Lyssoivan R Koch G Van Wassenhove M Vervier G Bertschinger R Laengner B Unterberg G Sergienko V Philipps T Wauters the TEXTOR Team. Research Articles Volume 80 ...

  13. Measurement of high-energy electrons by means of a Cherenkov detector in ISTTOK tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Jakubowski, L., E-mail: lech.Jjakubowski@ipj.gov.p [Andrzej Soltan Institute for Nuclear Studies (IPJ), 05-400 Otwock-Swierk (Poland); Zebrowski, J. [Andrzej Soltan Institute for Nuclear Studies (IPJ), 05-400 Otwock-Swierk (Poland); Plyusnin, V.V. [Association Euratom/IST, Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Av. Rovisco Pais, 1049 - 001 Lisboa (Portugal); Malinowski, K.; Sadowski, M.J.; Rabinski, M. [Andrzej Soltan Institute for Nuclear Studies (IPJ), 05-400 Otwock-Swierk (Poland); Fernandes, H.; Silva, C.; Duarte, P. [Association Euratom/IST, Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Av. Rovisco Pais, 1049 - 001 Lisboa (Portugal)

    2010-10-15

    The paper concerns detectors of the Cherenkov radiation which can be used to measure high-energy electrons escaping from short-living plasma. Such detectors have high temporal (about 1 ns) and spatial (about 1 mm) resolution. The paper describes a Cherenkov-type detector which was designed, manufactured and installed in the ISTTOK tokamak in order to measure fast runaway electrons. The radiator of that detector was made of an aluminium nitride (AlN) tablet with a light-tight filter on its front surface. Cherenkov signals from the radiator were transmitted through an optical cable to a fast photomultiplier. It made possible to perform direct measurements of the runaway electrons of energy above 80 keV. The measured energy values and spatial characteristics of the recorded electrons appeared to be consistent with results of numerical modelling of the runaway electron generation process in the ISTTOK tokamak.

  14. Effects of turbulent fluctuations on density measurements with microwave reflectometry in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Mazzucato, E.; Nazikian, R.

    1994-08-01

    The short-scale turbulence of tokamak plasmas has deleterious effects on the measurement of plasma density with microwave reflectometry. Density fluctuations may lead to large amplitude and phase modulations of the reflected wave which can impair the measurement of the wave group delay, and hence the determination of the plasma density. The role played by different types of turbulent fluctuations and the limitations imposed on microwave reflectometry are discussed in this paper.

  15. 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...

  16. Main Physical Factors Limiting the Accuracy of Polarimetric Measurements in Tokamak Plasma

    Science.gov (United States)

    Bieg, Bohdan; Chrzanowski, Janusz; Kravtsov, Yury A.; Orsitto, Francesco

    The paper reviews and discusses the main factors, limiting the accuracy of polarimetric measurements in tokamak plasma. Theoretical methods, describing evolution of polarimetry state in tokamak plasma, are demonstrated not to contribute noticeably to inaccuracy at sufficiently short beam wavelengths. Based on the literature data as well as on our preliminary estimates it is possible to conclude that the following factors dominate: i) calibration procedure; ii) refraction in the inhomogeneous plasma; iii) influence of weak relativistic effects on plasma dielectric permittivity. The contribution of these factors to is within the range of several per cent. Other causes of measurement inaccuracies (absorption in plasma, diffraction of sounding beam, ray torsion, nonstationary processes in plasma) seem to be less significant.

  17. Diamagnetic measurements in the STOR-M tokamak

    Science.gov (United States)

    Trembach, Dallas

    2008-11-01

    Diamagnetic measurements of poloidal beta have been successfully performed on the Saskatchewan Torus-Modified (STOR-M) using a compensated coil system mounted exterior to the vacuum chamber wall. A significant challenge in performing these measurements on STOR-M is the presence of a decaying toroidal magnetic field over the duration of the discharge. A simple method for compensating these measurements based on independently measuring the vacuum field signal and correcting during post-processing is presented. Measurements of poloidal beta using the diamagnetic coil arrangement are compared to calculations of poloidal beta based on the Spitzer conductivity corrected for trapped electrons.

  18. Heterodyne detector for measuring the characteristic of elliptically polarized microwaves

    DEFF Research Database (Denmark)

    Leipold, Frank; Nielsen, Stefan Kragh; Michelsen, Susanne

    2008-01-01

    be calculated. Results from measured and calculated wave characteristics of an elliptically polarized 110 GHz microwave beam for plasma heating launched into the TEXTOR-tokamak experiment are presented. Measurement and calculation are in good agreement. ©2008 American Institute of Physics......In the present paper, a device is introduced, which is capable of determining the three characteristic parameters of elliptically polarized light (ellipticity, angle of ellipticity, and direction of rotation) for microwave radiation at a frequency of 110 GHz. The device consists of two...

  19. Visible imaging measurement of position and displacement of the last closed flux surface in EAST tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Y.F. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Xu, G.S., E-mail: gsxu@ipp.ac.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Li, Y.L.; Yang, J.H.; Yan, N.; Liu, L.; Yuan, S.; Luo, Z.P. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Sang, C.F. [School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China); Gu, S.; Xu, J.C.; Hu, G.H.; Wang, Y.S. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Peng, Y.K.M.; Wan, B.N. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2017-06-15

    Highlights: • A new method for measuring the position and displacement of the LCFS has been developed in EAST tokamak. • This method is based on the visible imaging diagnostic and shown to be an effective and convenient approach. • This method can be applied to measure displacements of the LCFS during application of resonant magnetic perturbation fields. - Abstract: A new method for measuring the position and displacement of the last closed flux surface (LCFS) with visible imaging diagnostics has been developed in EAST. By measuring the relative intensity profiles of the green visible Li-II emission in the tangential planes of the optical systems, it is possible to infer the positions of certain points on the LCFS. This emission line is readily available in discharges with Li-coating wall routinely employed to improve the plasma performance. We describe the measuring method, giving results which are compared with those obtained by EFIT, and showing this as an effective and convenient approach to determine the position of the LCFS. This method is further applied to measure the displacements of the LCFS during application of resonant magnetic perturbation fields in the EAST tokamak.

  20. LETTER: Radial electric field measurement in a tokamak with magnetic field ripple

    Science.gov (United States)

    Trier, E.; Eriksson, L.-G.; Hennequin, P.; Fenzi, C.; Bourdelle, C.; Falchetto, G.; Garbet, X.; Aniel, T.; Clairet, F.; Sabot, R.

    2008-09-01

    In the regions of the Tore Supra tokamak with significant ripple it is expected that a radial electric field (Er) ensures the ambipolarity of fluxes of thermal particles trapped in ripple wells. A neoclassical calculation (Connor and Hastie 1973 Nucl. Fusion 13 221, Stringer 1972 Nucl. Fusion 12 689) shows that Er is related to ion temperature and density gradients. The validity of this relation is investigated in a series of Tore Supra L-mode discharges without external momentum input. Doppler reflectometry measurements of fluctuations perpendicular velocity, which is dominated by the Er × B drift, are found to be in good agreement with the predicted neoclassical Er.

  1. Li-BES detection system for plasma turbulence measurements on the COMPASS tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Berta, M. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Széchenyi István University, Győr (Hungary); Anda, G.; Bencze, A.; Dunai, D. [Wigner – RCP, HAS, Budapest (Hungary); Háček, P., E-mail: hacek@ipp.cas.cz [Institute of Plasma Physics AS CR, Prague (Czech Republic); Faculty of Mathematics and Physics, Charles University in Prague, Prague (Czech Republic); Hron, M. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Kovácsik, A. [Wigner – RCP, HAS, Budapest (Hungary); Department of Nuclear Techniques, Budapest University of Technology and Economics, Budapest (Hungary); Krbec, J. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, Prague (Czech Republic); Pánek, R. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Réfy, D.; Veres, G. [Wigner – RCP, HAS, Budapest (Hungary); Weinzettl, V. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Zoletnik, S. [Wigner – RCP, HAS, Budapest (Hungary)

    2015-10-15

    Highlights: • Li-BES detection system on the COMPASS tokamak is optimized observation system with high temporal resolution. • High sensitivity to low level light fluctuations. • Optics and detectors with electronics are placed in thermally stabilized compact box. • Fast deflection system allows us to measure background corrected electron density profiles on microsecond time-scale. - Abstract: A new Li beam emission spectroscopy (Li-BES) diagnostic system with a ∼ cm spatial resolution, and with beam energy ranging from 10 keV up to 120 keV and a 18 channel Avalanche photo diode (APD) detector system sampled at 2 MHz has been recently installed and tested on the COMPASS tokamak. This diagnostic allows to reconstruct density profile based on directly measured light profiles, and to follow turbulent behaviour of the edge plasma. The paper reports technical capabilities of this new system designed for fine spatio-temporal measurements of plasma electron density. Focusing on turbulence-induced fluctuation measurements, we demonstrate how physically relevant information can be extracted using the COMPASS Li-BES system.

  2. Hydrogen recycling and transport in the helical divertor of TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Clever, Meike

    2010-07-01

    The aim of this thesis was to investigate the hydrogen recycling at the target plates of the helical divertor in TEXTOR and by this the capability of this divertor configuration to access such favourable operational regimes. In order to study the different divertor density regimes in TEXTOR, discharges were performed in which the total plasma density was increased continuously up to the density limit. The recycling was investigated in a fixed helical divertor structure where four helical strike points with a poloidal width of about 8-10 cm are created at the divertor target plates. The experimental investigation of the hydrogen recycling was carried out using mainly spectroscopic methods supplemented by Langmuir probe, interferometric and atomic beam measurements. In the framework of this thesis a spectroscopic multi camera system has been built that facilitates the simultaneous observation of four different spectral lines, recording images of the divertor target plates and the plasma volume close to the target. The system facilitates the simultaneous measurement of the poloidal and toroidal pattern of the recycling flux at the divertor target without the need for sweeping the plasma structure. The simultaneous observation of different spectral lines reduces the uncertainty in the analysis based on several lines, as the contribution from uncertainties in the reproducibility of plasma parameters in different discharges are eliminated and only the uncertainty of the measurement method limits the accuracy. The spatial resolution of the system in poloidal and toroidal direction (0.8 mm{+-}0.01 mm) is small compared to the separation of the helical strike points, the capability of the measurement method to resolve these structures is therefore limited by the line-of-sight integration and the penetration depth of the light emitting species. The measurements showed that the recycling flux increases linearly with increasing plasma density, a high recycling regime is not

  3. Temperature measurement of plasma-facing surfaces in tokamaks by active pyrometry

    Energy Technology Data Exchange (ETDEWEB)

    Grigorova, V.; Semerok, A.; Farcage, D.; Weulersse, J.M. [CEA Saclay, DEN/DPC/SCP/LILM, Bat. 467, 91191 Gif-sur-Yvette (France); Thro, P.Y., E-mail: pierre-yves.thro@cea.f [CEA Saclay, DEN/DPC/SCP/LILM, Bat. 467, 91191 Gif-sur-Yvette (France); Gauthier, E.; Roche, H.; Loarer, Th.; Grisolia, Ch. [CEA Cadarache, DSM/ IRFM/SIPP, 13108 Saint Paul Lez Durance (France)

    2009-06-15

    This paper discusses feasibility and tests of a new method for in situ temperature measurement of tokamak plasma-facing metallic surfaces under plasma presence. In such conditions, the other temperature-measurement methods are not applicable due to the perturbing thermal radiation reflected by the walls. Our approach overcomes this limitation by looking with two pyrometers to the measured surface while thermally perturbed. Because of the thermal perturbation each pyrometer records a signal modulation. The temperature, deduced by the ratio between the two signal modulations is dependent neither on the environmental reflecting fluxes nor on the surface emissivity. Originally, the measured temperature is linked to the signals ratio via the experimental set-up parameters. Here, we proposed an alternative way to deduce it from the pyrometers calibration data only. With this method we obtained temperature measurements with accuracy better than 90%.

  4. A high speed compact microwave interferometer for density fluctuation measurements in Sino-UNIted Spherical Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Zhong, H., E-mail: zhongh14@126.com; Tan, Y.; Liu, Y. Q.; Xie, H. Q.; Gao, Z. [Department of Engineering Physics, Tsinghua University, Beijing 100084 (China)

    2016-11-15

    A single-channel 3 mm interferometer has been developed for plasma density diagnostics in the Sino-UNIted Spherical Tokamak (SUNIST). The extremely compact microwave interferometer utilizes one corrugated feed horn antenna for both emitting and receiving the microwave. The beam path lies on the equatorial plane so the system would not suffer from beam path deflection problems due to the symmetry of the cross section. A focusing lens group and an oblique vacuum window are carefully designed to boost the signal to noise ratio, which allows this system to show good performance even with the small-diameter central column itself as a reflector, without a concave mirror. The whole system discards the reference leg for maximum compactness, which is particularly suitable for the small-sized tokamak. An auto-correcting algorithm is developed to calculate the phase evolution, and the result displays good phase stability of the whole system. The intermediate frequency is adjustable and can reach its full potential of 2 MHz for best temporal resolution. Multiple measurements during ohmic discharges proved the interferometer’s capability to track typical density fluctuations in SUNIST, which enables this system to be utilized in the study of MHD activities.

  5. Plasma flow measurements in improved modes on STOR-M and CASTOR tokamaks

    Science.gov (United States)

    Germaine, G. S.; Xiao, C.; Hirose, A.

    2010-02-01

    A Gundestrup probe, a Mach probe array, is used to measure both the parallel and perpendicular flow velocities in the Saskatchewan Torus-Modified (STOR-M) tokamak during several discharge conditions. It is observed that during ohmic discharges there is no velocity shear and the direction of the parallel flow is independent of the direction of the toroidal magnetic field. During H-mode induced by a turbulent heating current pulse, a region of strong velocity shear develops in the plasma edge and an edge transport barrier develops. This results in a short period of improved particle and energy confinement with reduced fluctuation amplitudes. During electrode biasing experiments, a stainless steel biasing electrode is inserted into the plasma up to r=82 mm and biased to+500 V relative to the vacuum chamber. It is observed that the particle confinement improves during the biasing phase while the energy confinement is degraded. A region of weak shear in the poloidal flow is observed in the plasma scrape-off layer (SOL). The results from STOR-M are compared with results from data taken in the Czech Academy of Sciences Torus (CASTOR) tokamak during both ohmic discharges and discharges with electrode biasing.

  6. Mach Probe and Limiter Current Measurements in the STOR-M Tokamak

    Science.gov (United States)

    Morelli, Jordan; Singh, Ajay; Xiao, Chijin; Hirose, Akira

    2001-10-01

    We report plasma flow measurements conducted with a 4-sided Mach probe and the limiter current in a segmented limiter in both ac and normal modes of operation in the STOR-M tokamak. The Mach probe was biased in the ion saturation region and revealed the radial profiles of the toroidal and poloidal velocity structure in the plasma edge region. It was reported previously, from microwave interferometer measurements, that while the current goes to zero there is still a finite plasma density in the device. There are efforts to explain the equilibrium of this plasma, but to the best of our knowledge these are the first independent measurements from two other diagnostics indicating the finite density in the device during the current reversal with negligible rotational transform. Short circuiting through a limiter may provide a mechanism for decelerating plasma loss due to toroidal drifts.

  7. Spectroscopic measurements of ion temperature in ATC Tokamak with RF and neutral beam heating

    Energy Technology Data Exchange (ETDEWEB)

    Suckewer, S.; Hinnov, E.

    1977-03-01

    Measurements of ion temperatures in the ATC Tokamak by means of Doppler broadening of various ion lines are described, and typical results presented for the various auxiliary heating experiments: compression, neutral beam, lower hybrid and ion cyclotron frequency heating. Radial resolution of the temperature measurements is achieved by utilizing spectrum lines of ions of different ionization potentials: OVII lambda 1623A, CV lambda 2271A and CIV lambda 1548A, which are emitted from regions of different electron temperature. Measurement at a given radial location is performed as a function of time by repeated scanning of the line contour in times 1.5 to 3.0 msec. The results indicate variations of heating efficiency with location and with power input level.

  8. Measurement of Plasma Rotation Velocities in the STOR-M Tokamak

    Science.gov (United States)

    Morelli, Jordan; Xiao, Chijin; McColl, David; Hirose, Akira; Mitarai, Osamu

    2000-10-01

    Measurements of the plasma rotation velocities in the edge region of the Saskatchewan Torus-Modified (STOR-M) tokamak during one full cycle of alternating current operation and CT injection will be presented. In these experiments, a four sided Mach probe is used to measure the radial profile of the plasma poloidal and toroidal rotation velocities in the edge region. It has long been suspected that changes in the plasma edge region of both the velocity structure, and the radial electric field and its gradient are responsible for the transition to the ohmic high-confinement mode (H-mode). Furthermore, the results will help to check a recent theoretical model in which the confinement improvement is based on the toroidal velocity CURVATURE, consistent with the expectation that the tangential CT injection speeds up the toroidal flow.

  9. 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.

  10. High Resolution Transmission Grating Spectrometer for Edge Toroidal Rotation Measurements of Tokamak Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Graf, A; May, M; Beiersdorfer, P; Magee, E; Lawrence, M; Terry, J; Rice, J

    2004-04-29

    We present a high throughput (f/3) visible (3500 - 7000 Angstrom) Doppler spectrometer for toroidal rotation velocity measurements of the Alcator C-Mod tokamak plasma. The spectrometer has a temporal response of 1 ms and a rotation velocity sensitivity of {approx}10{sup 5} cm/s. This diagnostic will have a tangential view and map out the plasma rotation at several locations along the outer half of the minor radius (r/a > 0.5). The plasma rotation will be determined from the Doppler shifted wavelengths of D{sub alpha} and magnetic and electric dipole transitions of highly ionized impurities in the plasma. The fast time resolution and high spectral resolving power are possible due to a 6' diameter circular transmission grating that is capable of {lambda}/{Delta}{lambda} {approx} 15500 at 5769 Angstrom in conjunction with a 50 {micro}m slit.

  11. Microwave Imaging Reflectometry for the Measurement of Turbulent Fluctuations in Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    E. Mazzucato

    2004-02-19

    This article describes a numerical study of microwave reflectometry for the measurement of turbulent fluctuations in tokamak-like plasmas with a cylindrical geometry. Similarly to what was found previously in plane-stratified plasmas, the results indicate that the characteristics of density fluctuations cannot be uniquely determined from the reflected waves if the latter are allowed to propagate freely to the point of detection, as in standard reflectometry. Again, we find that if the amplitude of fluctuations is below a threshold that is set by the spectrum of poloidal wave numbers, the local characteristics of density fluctuations can be obtained from the phase of reflected waves when these are collected with a wide aperture antenna, and an image of the cutoff is formed onto an array of phase-sensitive detectors.

  12. Experimental Investigation of Runaway Electron Generation in Textor

    NARCIS (Netherlands)

    R. Jaspers,; Finken, K.H.; Mank, G.; Hoenen, F.; Boedo, J. A.; Cardozo, N. J. L.; Schüller, F. C.

    1993-01-01

    An experimental study of the generation of runaway electrons in TEXTOR has been performed. From the infrared synchrotron radiation emitted by relativistic electrons, the number of runaway electrons can be obtained as a function of time. In low density discharges (n(e)BAR < 1 X 10(19) m-3)

  13. Runaway generation during disruptions in JET and TEXTOR

    NARCIS (Netherlands)

    Lehnen, M.; Abdullaev, S. S.; Arnoux, G.; Bozhenkov, S. A.; Jakubowski, M. W.; R. Jaspers,; Plyusnin, V. V.; Riccardo, V.; Samm, U.

    2009-01-01

    Runaway electrons generated during ITER disruptions are of concern for the integrity of the plasma facing components. It is expected that a power of up to 8 GW is exposed to ITER PFCs. We present ill this article observations from JET and TEXTOR on the generation of runaways and the heat load

  14. Vertical Position and Current Profile Measurements by Faraday-effect Polarimetry On EAST tokamak

    Science.gov (United States)

    Ding, Weixing; Liu, H. Q.; Jie, Y. X.; Brower, D. L.; Qian, J. P.; Zou, Z. Y.; Lian, H.; Wang, S. X.; Luo, Z. P.; Xiao, B. J.; Ucla Team; Asipp Team

    2017-10-01

    A primary goal for ITER and prospective fusion power reactors is to achieve controlled long-pulse/steady-state burning plasmas. For elongated divertor plasmas, both the vertical position and current profile have to be precisely controlled to optimize performance and prevent disruptions. An eleven-channel laser-based POlarimeter-INTerferometer (POINT) system has been developed for measuring the internal magnetic field in the EAST tokamak and can be used to obtain the plasma current profile and vertical position. Current profiles are determined from equilibrium reconstruction including internal magnetic field measurements as internal constraints. Horizontally-viewing chords at/near the mid-plane allow us to determine plasma vertical position non-inductively with subcentimeter spatial resolution and time response up to 1 s. The polarimeter-based position measurement, which does not require equilibrium reconstruction, is benchmarked against conventional flux loop measurements and can be exploited for feedback control. Work supported by US DOE through Grants No. DE-FG02-01ER54615 and No. DC-SC0010469.

  15. On the harmonic technique to measure electron temperature with high time resolution

    Science.gov (United States)

    Boedo, J. A.; Gray, D.; Conn, R. W.; Luong, P.; Schaffer, M.; Ivanov, R. S.; Chernilevsky, A. V.; Van Oost, G.

    1999-07-01

    A detailed study of the harmonic technique, which exploits the generation of harmonics resulting from excitation of the nonlinearity of the single Langmuir probe characteristic, is presented. The technique is used to measure electron temperature and its fluctuations in tokamak plasmas and the technical issues relevant to extending the technique to high bandwidth (200 kHz) are discussed. The technique has been implemented in a fast reciprocating probe in the TEXTOR tokamak, gaining the ability to study denser and hotter plasmas than previously possible. A corrected analytical expression is derived for the harmonic currents. Measurement of the probe current by inductive pickup is introduced to improve electrical isolation and bandwidth. The temperature profiles in the boundary plasma of TEXTOR have been measured with high spatial (˜2 mm) and temporal (200 kHz) resolution and compared to those obtained with a double probe. The exact expansion of the probe characteristic in terms of Bessel functions is compared to a computationally efficient power series. Various aspects of the interpretation of the measurement are discussed such as the influence of plasma potential and density fluctuations. The technique is well suited to study fast phenomena such as transient plasma discharges or turbulence and turbulent transport in plasmas.

  16. First measurement of the edge charge exchange recombination spectroscopy on EAST tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Li, Y. Y., E-mail: liyy@ipp.ac.cn; Fu, J.; Jiang, D.; Lyu, B.; Hu, C. D.; Wan, B. N. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Yin, X. H.; Feng, S. Y. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei (China); Shi, Y. J. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei (China); Department of Nuclear Engineering, Seoul National University, Seoul 151-742 (Korea, Republic of); Yi, Y.; Ye, M. Y. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei (China); Zhou, X. J. [Anhui Institute of Optics and Fine Mechanics, Chinese Academy of Sciences, Hefei 230031 (China)

    2016-11-15

    An edge toroidal charge exchange recombination spectroscopy (eCXRS) diagnostic, based on a heating neutral beam injection (NBI), has been deployed recently on the Experimental Advanced Superconducting Tokamak (EAST). The eCXRS, which aims to measure the plasma ion temperature and toroidal rotation velocity in the edge region simultaneously, is a complement to the exiting core CXRS (cCXRS). Two rows with 32 fiber channels each cover a radial range from ∼2.15 m to ∼2.32 m with a high spatial resolution of ∼5-7 mm. Charge exchange emission of Carbon VI CVI at 529.059 nm induced by the NBI is routinely observed, but can be tuned to any interested wavelength in the spectral range from 400 to 700 nm. Double-slit fiber bundles increase the number of channels, the fibers viewing the same radial position are binned on the CCD detector to improve the signal-to-noise ratio, enabling shorter exposure time down to 5 ms. One channel is connected to a neon lamp, which provides the real-time wavelength calibration on a shot-to-shot basis. In this paper, an overview of the eCXRS diagnostic on EAST is presented and the first results from the 2015 experimental campaign will be shown. Good agreements in ion temperature and toroidal rotation are obtained between the eCXRS and cCXRS systems.

  17. Numerical analysis of particle recycling in the TEXTOR helical divertor

    Science.gov (United States)

    Frerichs, H.; Clever, M.; Feng, Y.; Lehnen, M.; Reiter, D.; Schmitz, O.

    2012-02-01

    The TEXTOR helical divertor is a magnetic configuration created by the application of external resonant magnetic perturbations with the intention to control plasma edge transport and the resulting particle and heat fluxes to the divertor target. It is confirmed by 3D computer simulations that no high-recycling-like regime is established under TEXTOR relevant conditions, despite the fact that a transition to detachment (i.e. a saturation or even a roll-over of the recycling flux) is observed at high densities. The driving mechanisms are, distinct from apparently similar observations in poloidal divertors and stellarator divertors, a combination of volumetric power losses and enhanced upstream-to-downstream heat transport, but with no significant role of the momentum balance.

  18. Runaway generation during disruptions in JET and TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Lehnen, M., E-mail: m.lehnen@fz-juelich.d [Institute of Energy Research - Plasma Physics, Forschungszentrum Juelich GmbH, Association EURATOM-FZJ, Juelich (Germany); Abdullaev, S.S. [Institute of Energy Research - Plasma Physics, Forschungszentrum Juelich GmbH, Association EURATOM-FZJ, Juelich (Germany); Arnoux, G. [EURATOM/UKAEA Fusion Association, Culham Science Centre, OX14 3DB (United Kingdom); Bozhenkov, S.A. [Institute of Energy Research - Plasma Physics, Forschungszentrum Juelich GmbH, Association EURATOM-FZJ, Juelich (Germany); Jakubowski, M.W. [Max-Planck-Institut fuer Plasmaphysik, IPP-EURATOM Association, Teilinstitut Greifswald, Wendelsteinstr. 1, 17491 Greifswald (Germany); Jaspers, R. [FOM-Rijnhuizen, Association EURATOM-FOM, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Plyusnin, V.V. [Instituto de Plasmas e Fusao Nuclear/IST, Associacao EURATOM-IST, Av. Rovisco Pais, 1049-001 Lisbon (Portugal); Riccardo, V. [EURATOM/UKAEA Fusion Association, Culham Science Centre, OX14 3DB (United Kingdom); Samm, U. [Institute of Energy Research - Plasma Physics, Forschungszentrum Juelich GmbH, Association EURATOM-FZJ, Juelich (Germany)

    2009-06-15

    Runaway electrons generated during ITER disruptions are of concern for the integrity of the plasma facing components. It is expected that a power of up to 8 GW is exposed to ITER PFCs. We present in this article observations from JET and TEXTOR on the generation of runaways and the heat load deposition. Suppression techniques like massive gas injection and resonant magnetic perturbations are discussed.

  19. A study of impurity transport in the plasma boundary of TEXTOR using gas puffing

    Science.gov (United States)

    McCracken, G. M.; Samm, U.; Fielding, S. J.; Matthews, G. F.; Pitts, R. A.; Pitcher, C. S.; Gray, D.; Lie, Y. T.; Moyer, R. A.; Bertschinger, G.; Pospieszczyk, A.; Rusbuldt, D.; Stangeby, P. C.; Elder, D.; Schweer, B.

    1990-12-01

    The transport of carbon and oxygen impurities has been studied in TEXTOR by introducing the gases CH 4 and CO through a small hole in a test limiter. The toroidal distributions of different charge states of the impurities have been measured using a CCD camera with optical filters. Local impurity ion temperatures have been calculated from the Doppler broadening of line emission measured with a high resolution spectrometer. The spatial distributions and the ion temperatures have been modelled using the LIM Monte Carlo impurity code, with experimentally measured plasma profiles. Good agreement is obtained for both sets of measurements. The comparison shows the breakup energies of the atomic fragments to be ≪ 1 eV. The fuelling efficiency of different gas species is discussed.

  20. Fluctuation BES measurements with the ITER core CXRS prototype spectrometer

    Energy Technology Data Exchange (ETDEWEB)

    Pokol, G.I., E-mail: pokol@reak.bme.hu [Institute of Nuclear Techniques, Budapest University of Technology and Economics, EURATOM Association, PO Box 91, H-1521, Budapest (Hungary); Zoletnik, S.; Dunai, D. [WIGNER RCP, RMKI, EURATOM Association, PO Box 91, H-1521, Budapest (Hungary); Marchuk, O. [Institut für Energieforschung – Plasmaphysik, Forschungszentrum Jülich Gmbh, Association EURATOM-FZJ, member of Trilateral Euregio Cluster, 52425 Jülich (Germany); Baross, T. [WIGNER RCP, RMKI, EURATOM Association, PO Box 91, H-1521, Budapest (Hungary); Erdei, G. [Department of Atomic Physics, Budapest University of Technology and Economics, EURATOM Association, PO Box 91, H-1521, Budapest (Hungary); Grunda, G.; Kiss, I.G. [WIGNER RCP, RMKI, EURATOM Association, PO Box 91, H-1521, Budapest (Hungary); Kovacsik, A. [Institute of Nuclear Techniques, Budapest University of Technology and Economics, EURATOM Association, PO Box 91, H-1521, Budapest (Hungary); Hellermann, M. von; Lischtschenko, O. [Dutch-Institute for Fundamental Energy Research, Association EURATOM-FOM, Partner in the Trilateral Euregio Cluster and ITER-NL, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Biel, W. [Institut für Energieforschung – Plasmaphysik, Forschungszentrum Jülich Gmbh, Association EURATOM-FZJ, member of Trilateral Euregio Cluster, 52425 Jülich (Germany); Jaspers, R.J.E. [Science and Technology of Nuclear Fusion, Eindhoven University of Technology (Netherlands); Durkut, M. [TNO Science and Industry, Partner in ITER-NL, PO Box 155, 2600 AD Delft (Netherlands)

    2013-10-15

    Highlights: • We integrated a fluctuation beam emission measurement into the ITER CXRS prototype spectrometer. • The fluctuation BES measurement provided data at TEXTOR that agree well with the simulation based on the Simulation Of Spectra package. • The same simulation method has been used to evaluate the feasibility of a fluctuation BES measurement on the ITER DNB using the CXRS periscopes. -- Abstract: The ITER core CXRS diagnostic system collects the light emitted from the interaction of the diagnostic neutral beam with the core plasma and guides it via a mirror labyrinth through the upper port plug no. 3 towards a fiber bundle, which then transmits the light into a set of spectrometers for spectral analysis. In order to test the accessibility of the special parameter range required for the ITER measurement, a prototype spectrometer was built and operated successfully at the TEXTOR tokamak. In addition to the He/Be, C/Ne and H/D/T regular spectral channels, a fluctuation beam emission spectroscopy (BES) system has been integrated to measure core MHD activity, and validate corresponding ITER simulation results. The fluctuation system can be operated as an alternative to the spectral BES measurement, and has 8 spatial channels sampled at 2 MHz. In this paper, we present details of the fluctuation BES system and its interface to the ITER prototype spectrometer along with simulation and measurement results at TEXTOR. We show that the measurement fully confirms the simulation results on achievable photon current at the detector and on the signal to noise ratio.

  1. Measurement of the electron and ion temperatures by the x-ray imaging crystal spectrometer on joint Texas experimental tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Yan, W.; Chen, Z. Y., E-mail: zychen@hust.edu.cn; Huang, D. W.; Tong, R. H.; Wang, S. Y.; Wei, Y. N.; Ma, T. K.; Zhuang, G. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, School of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan (China); Jin, W. [Center of Interface Dynamics for Sustainability, China Academy of Engineering Physics, Chengdu, Sichuan 610200 (China); Lee, S. G. [National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Shi, Y. J. [Department of Nuclear Engineering, Seoul National University, Seoul 08826 (Korea, Republic of)

    2016-11-15

    An x-ray imaging crystal spectrometer has been developed on joint Texas experimental tokamak for the measurement of electron and ion temperatures from the K{sub α} spectra of helium-like argon and its satellite lines. A two-dimensional multi-wire proportional counter has been applied to detect the spectra. The electron and ion temperatures have been obtained from the Voigt fitting with the spectra of helium-like argon ions. The profiles of electron and ion temperatures show the dependence on electron density in ohmic plasmas.

  2. Development of a Method for Local Electron Temperature and Density Measurements in the Divertor of the JET Tokamak

    Science.gov (United States)

    Jupen, C.; Meigs, A.; Bhatia, A. K.; Brezinsek, S.; OMullane, M.

    2004-01-01

    Plasma volume recombination in the divertor, a process in which charged particles recombine to neutral atoms, contributes to plasma detachment and hence cooling at the divertor target region. Detachment has been observed at JET and other tokamaks and is known to occur at low electron temperatures (T(sub e)10(exp 20)/m(exp 3)). The ability to measure such low temperatures is therefore of interest for modelling the divertor. In present work we report development of a new spectroscopic technique for investigation of local electron density (n(sub e)) and temperature (T,) in the outer divertor at JET.

  3. Development of plasma diagnostics technologies - Measurement of transport= parameters in tokamak edge plasma by using electric transport probes

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Kyu Sun; Chang, Do Hee; Sim, Yeon Gun; Kim, Jin Hee [Hanyang University, Seoul (Korea, Republic of)

    1995-08-01

    Electric transport probe system is developed for the measurement of electron temperature, floating potential, plasma density and flow velocity of= edge plasmas in the KT-2 medium size tokamak. Experiments have been performed in KT-1 small size tokamak. Electric transport probe is composed of a single probe(SP) and a Mach probe (MP). SP is used for the measurements of electron density, floating potential, and plasma density and measured values are {approx} 3*10{sup 11}/cm{sup -3}, -20 volts, 15 {approx} 25 eV. For the most discharges, respectively. MP is for the measurements of toroidal(M{sub T}) and poloidal(M{sub P}) flow velocities, and density, which are M{sub T} {approx_equal} .0.85, M{sub P} {approx_equal}. 0.17, n. {approx_equal} 2.1*10{sup 11} cm{sup -3}, respectively. A triple probe is also developed for the direct reading of T{sub e} and n{sub e}, and is used for DC, RF, and RF+DC plasma in APL of Hanyang university. 38 refs., 36 figs. (author)

  4. Design and fabrication of a new compound probe for plasma flux measurement in IR-T1 tokamak.

    Science.gov (United States)

    Alipour, R; Ghoranneviss, M; Salar Elahi, A

    2017-09-01

    A new compound probe is designed, built, and installed on an IR-T1 tokamak to flow measurements in the plasma edge region. The first results of using this probe on the IR-T1 tokamak are presented. The plasma parameters such as plasma current, loop voltage, floating potential, ion and electron saturation currents, electron temperature, plasma potential, and plasma flow velocities are measured in this work. The results show that the electron temperature and the plasma potential in the edge area are 14 eV and 44 V, respectively. The results indicate that the mean value of a parallel Mach number is 0.5 while the mean value of a perpendicular Mach number is almost zero. The large parallel flow velocity (about 17 km/s) and the negligible perpendicular flow velocity are also seen in this work. The most important advantage of using this compound probe is that it can not only save space and vacuum ports but also measure more physical quantities at the same time, contributing to further physical analysis.

  5. Material deposition and migration processes with resonant magnetic perturbation fields at TEXTOR

    Science.gov (United States)

    Laengner, Ruth; Schmitz, O.; Brezinsek, S.; Coenen, J. W.; Eich, T.; Freisinger, M.; Kirschner, A.; Kreter, A.; Möller, S.; Laengner, M.; Philipps, V.; Pospieszczyk, A.; Reimer, H.; Samm, U.; Wienhold, P.; Textor Team

    2013-07-01

    Resonant Magnetic Perturbations (RMPs) are applied with the Dynamic Ergodic Divertor (DED) at TEXTOR to control the plasma edge transport and the plasma surface interaction. This leads to the formation of a three-dimensional (3D) topology of the scrape-off layer (SOL). To quantify the erosion/deposition balance and the material migration in this 3D boundary, spherical test limiters were exposed to plasmas with and without RMP fields applied. Methane doped with 13C as tracer element was injected through a gas inlet in the test limiter. The local gas source was monitored by spatially resolving spectroscopy and the resulting deposition patterns on the limiters were analysed with colourimetry and nuclear reaction analysis. These measurements were compared to simulations of the magnetic field topology simulations. The data provide evidence of a particle migration dominated by an ExB drift within stochastic zones of the 3D plasma boundary.

  6. LIBS for tokamak plasma facing components characterisation: Perspectives on in situ tritium cartography

    Energy Technology Data Exchange (ETDEWEB)

    Semerok, A., E-mail: alexandre.semerok@cea.fr [CEA, DEN, DPC/SEARS/LISL, F-91191 Gif-sur-Yvette (France); Grisolia, C. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France)

    2013-08-21

    Feasibility of in situ LIBS remote measurements with the plasma facing components (PFCs) from the European tokamaks (TORE SUPRA, CEA Cadarache, France and TEXTOR, Julich, Germany) has been studied in laboratory using Q-switched nanosecond Nd–YAG lasers. LIBS particular properties and optimal parameters were determined for in-depth PFCs characterisation. The LIBS method was in situ tested on the Joint European Torus (JET) in the UK with the EDGE LIDAR Laser System (Ruby laser, 3 J, 690 nm wavelength, 300 ps pulse duration, intensity up to 70 GW/cm{sup 2}). Several analytical spectral lines of H, CII, CrI, and BeII in plasma were observed and identified in 400–600 nm spectral range with the optimised LIBS and detection system. The LIBS in-depth cartography is in agreement with the surface properties of the tile under analysis, thus confirming feasibility of in situ LIBS. Further LIBS technique improvements required to provide tritium concentration measurements more accurately are discussed.

  7. Dust Studies in DIII-D and TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Rudakov, D L; Litnovsky, A; West, W P; Yu, J H; Boedo, J A; Bray, B D; Brezinsek, S; Brooks, N H; Fenstermacher, M E; Groth, M; Hollmann, E M; Huber, A; Hyatt, A W; Krasheninnikov, S I; Lasnier, C J; Moyer, R A; Pigarov, A Y; Philipps, V; Pospieszczyk, A; Smirnov, R D; Sharpe, J P; Solomon, W M; Watkins, J G; Wong, C C

    2009-02-17

    Studies of naturally occurring and artificially introduced carbon dust are conducted in DIII-D and TEXTOR. In DIII-D, dust does not present operational concerns except immediately after entry vents. Submicron sized dust is routinely observed using Mie scattering from a Nd:Yag laser. The source is strongly correlated with the presence of Type I edge localized modes (ELMs). Larger size (0.005-1 mm diameter) dust is observed by optical imaging, showing elevated dust levels after entry vents. Inverse dependence of the dust velocity on the inferred dust size is found from the imaging data. Direct heating of the dust particles by the neutral beam injection (NBI) and acceleration of dust particles by the plasma flows are observed. Energetic plasma disruptions produce significant amounts of dust. Large flakes or debris falling into the plasma may result in a disruption. Migration of pre-characterized carbon dust is studied in DIII-D and TEXTOR by introducing micron-size dust in plasma discharges. In DIII-D, a sample holder filled with {approx}30 mg of dust is introduced in the lower divertor and exposed to high-power ELMing H-mode discharges with strike points swept across the divertor floor. After a brief exposure ({approx}0.1 s) at the outer strike point, part of the dust is injected into the plasma, raising the core carbon density by a factor of 2-3 and resulting in a twofold increase of the radiated power. In TEXTOR, instrumented dust holders with 1-45 mg of dust are exposed in the scrape-off layer 0-2 cm radially outside of the last closed flux surface in discharges heated with neutral beam injection (NBI) power of 1.4 MW. At the given configuration of the launch, the dust did not penetrate the core plasma and only moderately perturbed the edge plasma, as evidenced by an increase of the edge carbon content.

  8. Non-resonant magnetic braking on JET and TEXTOR

    DEFF Research Database (Denmark)

    Sun, Y.; Liang, Y.; Shaing, K.C.

    2012-01-01

    The non-resonant magnetic braking effect induced by a non-axisymmetric magnetic perturbation is investigated on JET and TEXTOR. The collisionality dependence of the torque induced by the n = 1, where n is the toroidal mode number, magnetic perturbation generated by the error field correction coils...... in the 1/ν regime. The strongest NTV torque on JET is also located near the plasma core. The magnitude of the NTV torque strongly depends on the plasma response, which is also discussed in this paper. There is no obvious braking effect with n = 2 magnetic perturbation generated by the dynamic ergodic...

  9. Spectroscopic system for impurity measurements in the TJ-1 Tokamak of JEN; Un sistema espectroscopico para medidas de impurezas en el Tokamak TJ-1 de la JEN

    Energy Technology Data Exchange (ETDEWEB)

    Navas, G.; Zurro, B.

    1982-07-01

    we describe a spectroscopic system with spatial resolution capability that has been configured for plasma diagnostic in the TJ-1 Tokamak of JEN. The experimental system, based on a one meter monochromator, has been absolutely calibrated using a tungsten-halogen lamp. The calibration procedures and the absolute spectral sensitivity are presented as well as its dependence with the polarization. A simplified spectroscopic model of the radiation emitted by the intrinsic plasma impurities (C, 0, . . . ) has been developed. A one dimensional model of the temporal evolution of various ionization stages in coronal equilibrium is used to predict the electron temperature and impurity concentration. This model has been applied to experimental data from several Tokamaks. (Author) 23 refs.

  10. Polarized fusion, its Implications and plans for Direct Measurements in a Tokamak

    OpenAIRE

    Sandorfi, A. M.; Deur, A.; Hanretty, C.; Jackson, G. L.; Lanctot, M.; Liu, J.; Lowry, M. M.; Miller, G. W.; Pace, D.; Smith, S. P.; Wei, K.; Wei, X.; Zheng, X.

    2017-01-01

    A long-term energy option that is just approaching the horizon after decades of struggle, is fusion. Recent developments allow us to apply techniques from spin physics to advance its viability. The cross section for the primary fusion fuel in a tokamak reactor, D+T=>alpha+n, would be increased by a factor of 1.5 if the fuels were polarized. Simulations predict further non-linear power gains in large-scale machines such as ITER, due to increased alpha heating. These are significant enhancement...

  11. Characteristics of four-channel Cherenkov-type detector for measurements of runaway electrons in the ISTTOK tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Plyusnin, V. V.; Duarte, P.; Fernandes, H.; Silva, C. [Instituto de Plasmas e FuSao Nuclear - Laboratorio Associado, Association Euratom/IST, Instituto Superior Tecnico, Av. Rovisco Pais, 1049-001 Lisboa (Portugal); Jakubowski, L.; Zebrowski, J.; Malinowski, K.; Rabinski, M.; Sadowski, M. J. [The Andrzej Soltan Institute for Nuclear Studies (IPJ), 05-400 Otwock-Swierk (Poland)

    2010-10-15

    A diagnostics capable of characterizing the runaway and superthermal electrons has been developing on the ISTTOK tokamak. In previous paper, a use of single-channel Cherenkov-type detector with titanium filter for runaway electron studies in ISTTOK was reported. To measure fast electron populations with different energies, a prototype of a four-channel detector with molybdenum filters was designed. Test-stand studies of filters with different thicknesses (1, 3, 7, 10, 20, 50, and 100 {mu}m) have shown that they should allow the detection of electrons with energies higher than 69, 75, 87, 95, 120, 181, and 260 keV, respectively. First results of measurements with the four-channel detector revealed the possibility to measure reliably different fast electrons populations simultaneously.

  12. Core-ion temperature measurement of the ADITYA tokamak using passive charge exchange neutral particle energy analyzer

    Energy Technology Data Exchange (ETDEWEB)

    Pandya, Santosh P.; Ajay, Kumar; Mishra, Priyanka; Dhingra, Rajani D.; Govindarajan, J. [Institute for Plasma Research, Bhat, Gandhinagar 382 428, Gujarat (India)

    2013-02-15

    Core-ion temperature measurements have been carried out by the energy analysis of passive charge exchange (CX) neutrals escaping out of the ADITYA tokamak plasma (minor radius, a= 25 cm and major radius, R= 75 cm) using a 45 Degree-Sign parallel plate electrostatic energy analyzer. The neutral particle analyzer (NPA) uses a gas cell configuration for re-ionizing the CX-neutrals and channel electron multipliers (CEMs) as detectors. Energy calibration of the NPA has been carried out using ion-source and {Delta}E/E of high-energy channel has been found to be {approx}10%. Low signal to noise ratio (SNR) due to VUV reflections on the CEMs was identified during the operation of the NPA with ADITYA plasma discharges. This problem was rectified by upgrading the system by incorporating the additional components and arrangements to suppress VUV radiations and improve its VUV rejection capabilities. The noise rejection capability of the NPA was experimentally confirmed using a standard UV-source and also during the plasma discharges to get an adequate SNR (>30) at the energy channels. Core-ion temperature T{sub i}(0) during flattop of the plasma current has been measured to be up to 150 eV during ohmically heated plasma discharges which is nearly 40% of the average core-electron temperature (typically T{sub e}(0) {approx} 400 eV). The present paper describes the principle of tokamak ion temperature measurement, NPA's design, development, and calibration along with the modifications carried out for minimizing the interference of plasma radiations in the CX-spectrum. Performance of the NPA during plasma discharges and experimental results on the measurement of ion-temperature have also been reported here.

  13. Study on sawtooth and transport in part of Japan-TEXTOR collaboration 1995

    Energy Technology Data Exchange (ETDEWEB)

    Itoh, K. [ed.

    1996-02-01

    A collaboration programme `physics of sawtooth and transport` has been performed in the frame work of the Japan-TEXTOR collaboration. The summary of the workshops and collaborations in 1995 is reported. (author)

  14. Equilibrium reconstruction based on core magnetic measurement and its applications on equilibrium transition in Joint-TEXT tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Chen, J.; Zhuang, G., E-mail: ge-zhuang@hust.edu.cn; Jian, X.; Li, Q.; Liu, Y.; Gao, L.; Wang, Z. J. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China)

    2014-10-15

    Evaluation and reconstruction of plasma equilibrium, especially to resolve the safety factor profile, is imperative for advanced tokamak operation and physics study. Based on core magnetic measurement by the high resolution laser polarimeter-interferometer system (POLARIS), the equilibrium of Joint-TEXT (J-TEXT) plasma is reconstructed and profiles of safety factor, current density, and electron density are, therefore, obtained with high accuracy and temporal resolution. The equilibrium reconstruction procedure determines the equilibrium flux surfaces essentially from the data of POLARIS. Refraction of laser probe beam, a major error source of the reconstruction, has been considered and corrected, which leads to improvement of accuracy more than 10%. The error of reconstruction has been systematically assessed with consideration of realistic diagnostic performance and scrape-off layer region of plasma, and its accuracy has been verified. Fast equilibrium transitions both within a single sawtooth cycle and during the penetration of resonant magnetic perturbation have been investigated.

  15. Design and characterization of a 32-channel heterodyne radiometer for electron cyclotron emission measurements on experimental advanced superconducting tokamak.

    Science.gov (United States)

    Han, X; Liu, X; Liu, Y; Domier, C W; Luhmann, N C; Li, E Z; Hu, L Q; Gao, X

    2014-07-01

    A 32-channel heterodyne radiometer has been developed for the measurement of electron cyclotron emission (ECE) on the experimental advanced superconducting tokamak (EAST). This system collects X-mode ECE radiation spanning a frequency range of 104-168 GHz, where the frequency coverage corresponds to a full radial coverage for the case with a toroidal magnetic field of 2.3 T. The frequency range is equally spaced every 2 GHz from 105.1 to 167.1 GHz with an RF bandwidth of ~500 MHz and the video bandwidth can be switched among 50, 100, 200, and 400 kHz. Design objectives and characterization of the system are presented in this paper. Preliminary results for plasma operation are also presented.

  16. Real time magnetic field and flux measurements for tokamak control using a multi-core PCI Express system

    Energy Technology Data Exchange (ETDEWEB)

    Giannone, L. [Max-Planck-Institut fuer Plasmaphysik, IPP-EURATOM Association, 85748 Garching (Germany)], E-mail: Louis.Giannone@ipp.mpg.de; Schneider, W.; McCarthy, P.J.; Sips, A.C.C.; Treutterer, W.; Behler, K.; Eich, T.; Fuchs, J.C.; Hicks, N.; Kallenbach, A.; Maraschek, M.; Mlynek, A.; Neu, G.; Pautasso, G.; Raupp, G.; Reich, M.; Schuhbeck, K.H.; Stober, J.; Volpe, F.; Zehetbauer, T. [Max-Planck-Institut fuer Plasmaphysik, IPP-EURATOM Association, 85748 Garching (Germany)] (and others)

    2009-06-15

    The existing real time system for the position and shape control in ASDEX Upgrade has been extended to calculate magnetic flux surfaces in real time using a multi-core PCI Express system running LabVIEW RT. The availability of reflective memory for LabVIEW RT will allow this system to be connected to the control system and other diagnostics in a multi-platform real time network. The measured response of each magnetic probe to the individual poloidal field coil currents in the absence of plasma current is compared to the calculated value. Prior to a tokamak discharge this comparison can be used to check for failure of the magnetic probe, flux loop or integrator.

  17. The ten-channel pulsed radar reflectometer at the TEXTOR-94 tokamak

    NARCIS (Netherlands)

    van Gorkom, J. C.; van de Pol, M.J.; Donne, A. J. H.

    2001-01-01

    A new ten-channel pulsed radar reflectometer has been taken into operation at the Torus Experiment for Technology Oriented Research-94. The system will be used simultaneously as a density profile and as a density fluctuation diagnostic. Ten density layers from 0.4 x 10(19) to 4 x 10(19) m(-3) can be

  18. Surface temperature measurement of the plasma facing components with the multi-spectral infrared thermography diagnostics in tokamaks

    Science.gov (United States)

    Zhang, C.; Gauthier, E.; Pocheau, C.; Balorin, C.; Pascal, J. Y.; Jouve, M.; Aumeunier, M. H.; Courtois, X.; Loarer, Th.; Houry, M.

    2017-03-01

    For the long-pulse high-confinement discharges in tokamaks, the equilibrium of plasma requires a contact with the first wall materials. The heat flux resulting from this interaction is of the order of 10 MW/m2 for steady state conditions and up to 20 MW/m2 for transient phases. The monitoring on surface temperatures of the plasma facing components (PFCs) is a major concern to ensure safe operation and to optimize performances of experimental operations on large fusion facilities. Furthermore, this measurement is also required to study the physics associated to the plasma material interactions and the heat flux deposition process. In tokamaks, infrared (IR) thermography systems are routinely used to monitor the surface temperature of the PFCs. This measurement requires an accurate knowledge of the surface emissivity. However, and particularly for metallic materials such as tungsten, this emissivity value can vary over a wide range with both the surface condition and the temperature itself, which makes instantaneous measurement challenging. In this context, the multi-spectral infrared method appears as a very promising alternative solution. Indeed, the system has the advantage to carry out a non-intrusive measurement on thermal radiation while evaluating surface temperature without requiring a mandatory surface emissivity measurement. In this paper, a conceptual design for the multi-spectral infrared thermography is proposed. The numerical study of the multi-channel system based on the Levenberg-Marquardt (LM) nonlinear curve fitting is applied. The numerical results presented in this paper demonstrate the design allows for measurements over a large temperature range with a relative error of less than 10%. Furthermore, laboratory experiments have been performed from 200 °C to 740 °C to confirm the feasibility for temperature measurements on stainless steel and tungsten. In these experiments, the unfolding results from the multi-channel detection provide good

  19. Transport modelling of TEXTOR-DED laminar zone

    Energy Technology Data Exchange (ETDEWEB)

    Eich, Th. E-mail: th.eich@fz-juelich.de; Reiser, D.; Finken, K.H

    2001-03-01

    In the case of a strong ergodisation of the plasma edge of TEXTOR-DED, the edge magnetic field forms an extended laminar zone, which is established by magnetic field lines with short connection lengths (open ergodic system). In the laminar zone the parallel transport can compete with the cross-field transport and the situation is similar to that in a regular divertor. For an analysis of the generic effects of the laminar zone on the plasma transport, the LUPUS code is developed taking flux tubes with short connection lengths into account. The ergodic zone with rather high connection lengths is described by enhanced perpendicular diffusion coefficients. As important results, which differ significantly from common SOL's, the expected power load and the flow pattern to the plasma facing components are presented.

  20. Diamagnetic measurements in the STOR-M tokamak by a flux loop system exterior to the vacuum vessel

    Science.gov (United States)

    Trembach, Dallas; Xiao, Chijin; Dreval, Mykola; Hirose, Akira

    2009-05-01

    Diamagnetic measurements of poloidal beta have been performed in the STOR-M tokamak by a flux loop placed exterior to the vacuum chamber with compensation for the vacuum toroidal field using a nonenclosing coplanar coil, and vibrational compensation from auxiliary coils. It was found that in STOR-M conditions (20% toroidal magnetic field decay over discharge) there is significant influence on the diamagnetic flux measurements from strong residual signals, presumably from image currents being induced by the toroidal field coils, requiring further compensation. A blank (nonplasma) shot is used specifically to eliminate the residual component which is not proportional to the toroidal magnetic field. Data from normal Ohmic discharge operation is presented and calculations of poloidal beta from coil data (βθ˜0.5) is found to be in reasonable agreement with the values of poloidal beta obtained from measurements of electron density and Spitzer temperature with neoclassical corrections for trapped electrons. Contributions present in the blank shot (residual) signal and the limitations of this method are discussed.

  1. Direct measurements of safety factor profiles with motional Stark effect for KSTAR tokamak discharges with internal transport barriers

    Science.gov (United States)

    Ko, J.; Chung, J.

    2017-06-01

    The safety factor profile evolutions have been measured from the plasma discharges with the external current drive mechanism such as the multi-ion-source neutral beam injection for the Korea Superconducting Tokamak Advanced Research (KSTAR) for the first time. This measurement has been possible by the newly installed motional Stark effect (MSE) diagnostic system that utilizes the polarized Balmer-alpha emission from the energetic neutral deuterium atoms induced by the Stark effect under the Lorentz electric field. The 25-channel KSTAR MSE diagnostic is based on the conventional photoelastic modulator approach with the spatial and temporal resolutions less than 2 cm (for the most of the channels except 2 to 3 channels inside the magnetic axis) and about 10 ms, respectively. The strong Faraday rotation imposed on the optical elements in the diagnostic system is calibrated out from a separate and well-designed polarization measurement procedure using an in-vessel reference polarizer during the toroidal-field ramp-up phase before the plasma experiment starts. The combination of the non-inductive current drive during the ramp-up and shape control enables the formation of the internal transport barrier where the pitch angle profiles indicate flat or slightly hollow profiles in the safety factor.

  2. Resonant Character of Edge Plasma Parameters in Stochastic Boundary Experiments at DIII-D and TEXTOR

    Science.gov (United States)

    Schmitz, O.; Bray, B. D.; Brooks, N. H.; Evans, T. E.; Leonard, A. W.; Osborne, T. H.; West, W. P.; Fenstermacher, M. E.; Groth, M.; Lasnier, C. J.; Frerichs, H.; Lehnen, M.; Unterberg, B.; Jakubowski, M. W.; Moyer, R. A.; Watkins, J. G.

    2008-11-01

    Dependence of electron pressure pe profiles on the edge safety factor during resonant magnetic perturbations (RMPs) is analyzed and compared to heat and particle fluxes. For TEXTOR, a strong reduction of pe and an increase of target fluxes is measured when the inward penetration of the vacuum stochastic layer is maximized. For DIII-D, target heat and particle fluxes follow the 3-D perturbed separatrix due to a stochastic layer of open, perturbed field lines with a minimum penetration to ψN=0.95 in normalized poloidal flux. Experimental measurements show the toroidally spiraling structure of perturbed target plate separatrix lobes depend on q95 and that there is a clear q95 dependent reduction of ne(ψN), Te(ψN) and pe(ψN) which follows the toroidal phase of the RMP field. The measurements provide evidence for pitch resonant edge stochastisation as a mechanism leading to peeling-ballooning stabilized RMP H-modes at DIII-D.

  3. Boundary plasma heat flux width measurements for poloidal magnetic fields above 1 Tesla in the Alcator C-Mod tokamak

    Science.gov (United States)

    Brunner, Dan; Labombard, Brian; Kuang, Adam; Terry, Jim; Alcator C-Mod Team

    2017-10-01

    The boundary heat flux width, along with the total power flowing into the boundary, sets the power exhaust challenge for tokamaks. A multi-machine boundary heat flux width database found that the heat flux width in H-modes scaled inversely with poloidal magnetic field (Bp) and was independent of machine size. The maximum Bp in the database was 0.8 T, whereas the ITER 15 MA, Q =10 scenario will be 1.2 T. New measurements of the boundary heat flux width in Alcator C-Mod extend the international database to plasmas with Bp up to 1.3 T. C-Mod was the only experiment able to operate at ITER-level Bp. These new measurements are from over 300 plasma shots in L-, I-, and EDA H-modes spanning essentially the whole operating space in C-Mod. We find that the inverse-Bp dependence of the heat flux width in H-modes continues to ITER-level Bp, further reinforcing the empirical projection of 500 μm heat flux width for ITER. We find 50% scatter around the inverse-Bp scaling and are searching for the `hidden variables' causing this scatter. Supported by USDoE award DE-FC02-99ER54512.

  4. Comparison of 2D simulations of detached divertor plasmas with divertor Thomson measurements in the DIII-D tokamak

    Directory of Open Access Journals (Sweden)

    T.D. Rognlien

    2017-08-01

    Full Text Available A modeling study is reported using new 2D data from DIII-D tokamak divertor plasmas and improved 2D transport model that includes large cross-field drifts for the numerically difficult low anomalous transport regime associated with the H-mode. The data set, which spans a range of plasma densities for both forward and reverse toroidal magnetic field (Bt, is provided by divertor Thomson scattering (DTS. Measurements utilizing X-point sweeping give corresponding 2D profiles of electron temperature (Te and density (ne across both divertor legs for individual discharges. The simulations focus on the open magnetic field-line regions, though they also include a small region of closed field lines. The calculations show the same features of in/out divertor plasma asymmetries as measured in the experiment, with the normal Bt direction (ion ∇B drift toward the X-point having higher ne and lower Te in the inner divertor leg than outer. Corresponding emission data for total radiated power shows a strong inner-divertor/outer-divertor asymmetry that is reproduced by the simulations. These 2D UEDGE transport simulations are enabled for steep-gradient H-mode conditions by newly implemented algorithms to control isolated grid-scale irregularities.

  5. Measurements of Soft-X-Ray Spectra in Ecr-Heated Tokamak Plasmas and a Comparison with Fokker-Planck Simulations

    NARCIS (Netherlands)

    Da Cruz, D. F.; Peeters, A.G.; Donne, A. J. H.; Cardozo, N. J. L.; Westerhof, E.

    1993-01-01

    Soft x-ray spectra have been measured in the RTP tokamak both during the ohmic phase for densities 2 X 10(19) less-than-or-equal-to n(e)(0) less-than-or-equal-to 7 X 10(19) m-3, and during the ECRH phase at a density n(e)(0) = 2 X 10(19) m-3 for several power levels. Large deformations of the soft

  6. Modelling of deposition and erosion of injected WF6 and MoF6 in TEXTOR

    Directory of Open Access Journals (Sweden)

    A. Kirschner

    2017-08-01

    Full Text Available Tracer injection experiments in TEXTOR with MoF6 and WF6 lead to local deposition of about 6% for Mo and about 1% for W relative to the injected amount of Mo and W atoms. Modelling of these experiments has been done with ERO applying updated data for physical sputtering. The dissociation of the injected molecules has been treated in a simplified manner due to the lack of dissociation rate coefficients. However, with this it was possible to reproduce the observed radial penetration of Mo and W atoms into the plasma. The modelled local deposition efficiencies are about 50% for Mo and 60% for W assuming typical plasma parameters for the experimental conditions used. To reproduce the measured deposition efficiencies an enhancement factor for the erosion of deposited Mo and W has to be assumed (∼10 for Mo and ∼25 for W. Due to the rather low electron temperature Te of these plasma conditions (Te∼15eV at the location of injection, Mo and W are mostly sputtered by impurities whereas sputtering due to deuterium is negligible. A parameter study applying larger electron temperature leads to increased sputtering and thus to reduced local deposition efficiencies of about 30% for Mo and 5% for W. Though, even under these conditions enhanced erosion, albeit with reduced enhancement factors, is needed in the modelling to obtain the small measured deposition efficiencies.

  7. One-dimensional linear calculation of the heat flux from infrared and thermocouple measurements at Jet tokamak; Calcul 1D lineaire du flux de chaleur par inversion des mesures de temperatures infrarouges et des thermocouples du Tokamak Jet

    Energy Technology Data Exchange (ETDEWEB)

    Poyet, M

    2005-07-01

    Our work is dedicated to the assessment of the heat released in the Jet tokamak divertor tiles. We have performed the computation of the heat flux from temperature data collected by thermo-couples through a 1 dimensional linear model. This method has implied solving an inverse problem whose matrix is singular, we have succeeded in using Tikhonov's regularization technique. Then we have compared these values of the heat flux with those deduced from infra-red measurements. Infra-red measurements are impaired by the deposition of particles on the surface. Both methods give unrealistic negative values at the end of the plasma discharge. The use of a non-linear 1-dimensional model that would allow the diffusion coefficient to vary is expected to improve the calculation. (A.C.)

  8. Surface thermocouples for measurement of pulsed heat flux in the divertor of the Alcator C-Mod tokamak.

    Science.gov (United States)

    Brunner, D; LaBombard, B

    2012-03-01

    A novel set of thermocouple sensors has been developed to measure heat fluxes arriving at divertor surfaces in the Alcator C-Mod tokamak, a magnetic confinement fusion experiment. These sensors operate in direct contact with the divertor plasma, which deposits heat fluxes in excess of ~10 MW/m(2) over an ~1 s pulse. Thermoelectric EMF signals are produced across a non-standard bimetallic junction: a 50 μm thick 74% tungsten-26% rhenium ribbon embedded in a 6.35 mm diameter molybdenum cylinder. The unique coaxial geometry of the sensor combined with its single-point electrical ground contact minimizes interference from the plasma/magnetic environment. Incident heat fluxes are inferred from surface temperature evolution via a 1D thermal heat transport model. For an incident heat flux of 10 MW/m(2), surface temperatures rise ~1000 °C/s, corresponding to a heat flux flowing along the local magnetic field of ~200 MW/m(2). Separate calorimeter sensors are used to independently confirm the derived heat fluxes by comparing total energies deposited during a plasma pulse. Langmuir probes in close proximity to the surface thermocouples are used to test plasma-sheath heat transmission theory and to identify potential sources of discrepancies among physical models.

  9. Measurement of thickness of film deposited on the plasma-facing wall in the QUEST tokamak by colorimetry

    Science.gov (United States)

    Wang, Z.; Hanada, K.; Yoshida, N.; Shimoji, T.; Miyamoto, M.; Oya, Y.; Zushi, H.; Idei, H.; Nakamura, K.; Fujisawa, A.; Nagashima, Y.; Hasegawa, M.; Kawasaki, S.; Higashijima, A.; Nakashima, H.; Nagata, T.; Kawaguchi, A.; Fujiwara, T.; Araki, K.; Mitarai, O.; Fukuyama, A.; Takase, Y.; Matsumoto, K.

    2017-09-01

    After several experimental campaigns in the Kyushu University Experiment with Steady-state Spherical Tokamak (QUEST), the originally stainless steel plasma-facing wall (PFW) becomes completely covered with a deposited film composed of mixture materials, such as iron, chromium, carbon, and tungsten. In this work, an innovative colorimetry-based method was developed to measure the thickness of the deposited film on the actual QUEST wall. Because the optical constants of the deposited film on the PFW were position-dependent and the extinction coefficient k1 was about 1.0-2.0, which made the probing light not penetrate through some thick deposited films, the colorimetry method developed can only provide a rough value range of thickness of the metal-containing film deposited on the actual PFW in QUEST. However, the use of colorimetry is of great benefit to large-area inspections and to radioactive materials in future fusion devices that will be strictly prohibited from being taken out of the limited area.

  10. Spectroscopic determination of inverse photon efficiencies of W atoms in the scrape-off layer of TEXTOR

    Science.gov (United States)

    Brezinsek, S.; Laengner, M.; Coenen, J. W.; O’Mullane, M. G.; Pospieszczyk, A.; Sergienko, G.; Samm, U.

    2017-12-01

    Optical emission spectroscopy can be applied to determine in situ tungsten particle fluxes from erosion processes at plasma-facing materials. Inverse photon efficiencies convert photon fluxes of WI and WII line transitions into W and {{{W}}}+ particle fluxes, respectively, dependening on the local plasma conditions. Experiments in TEXTOR were carried out to determine effective conversion factors for different WI and WII transitions with the aid of WF6 injection into deuterium scrape-off layer plasmas in the electron temperature T e range between {T}{e}=20 {eV} and {T}{e}=82 {eV}. The inverse photon efficiencies or so-called effective \\tfrac{S}{{XB}}-values have been determined for WI lines at λ =400.9 {nm}, 429.5 nm, 488.7 nm, 498.3 nm, and 522.5 nm as well as for WII at λ =434.6 {nm} and compared with theoretical calculations from the ADAS data base. Moreover, a multi-machine scaling for the \\tfrac{S}{{XB}}-value in the range of T e between 2...100 {eV} has been determined for the most prominent WI line at λ =400.9 {nm} to \\tfrac{S}{{XB}}({T}{e})=53.63-56.07× {e}(0.045× {T{e}[{eV}])} considering experimental data from TEXTOR, ASDEX Upgrade, PSI and PISCES. Comparison with ADAS calculations for the same transition reveal a good qualitative agreement with the dependence on T e , but an underestimation of ADAS calculations of less than 25% over the full covered range of experimentally accessible T e in the multi-machine scaling. A good agreement within the experimental uncertainties is found between TEXTOR and ADAS \\tfrac{S}{{XB}}-values for WI at λ =429.5 {nm} and λ =488.7 {nm} whereas an underestimation of up to a factor two of ADAS values for WI at λ =522.5 {nm} and λ =498.3 {nm} was measured. Potentially, reasons for the discrepancy are an overestimation of applied ionisation rate coefficients in ADAS for neutral W and a stronger electron dependence n e for these transitions.

  11. ALT-I pump limiter experiments with ICRF heating on TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Leung, W.K.; Goebel, D.M.; Conn, R.W.; Dippel, K.H.; Finken, K.H.; Thomas, G.J.

    1986-05-01

    The ALT-I (Advanced Limiter Test-I) was installed on TEXTOR to benchmark the ability of a pump limiter as an efficient particle collector and to determine the physics of pump limiter operation. Experiments continue to show its capability of removing particles from the plasma edge under different operating conditions. In this paper we report first experimental results using ALT-I in conjunction with high power ICRF heating. The particle removal rate increases as the edge flux and density increase during the ICRF pulse. For a head geometry that collects flux from both electron and ion drift sides, the plasma temperature rise is asymmetric with electron temperature on the electron side increasing more than on the ion side during the ICRF pulse. When ALT-I is the major limiter, the particle fluxes on both sides increase by about the same factor and the particle flux on the ion side is always larger, by a factor of 1.5 to 2 than on the electron side during both ohmic and ICRF periods. The degradation of particle confinement inferred from Langmuir probe measurement is more than a factor of two at a maximum achieved power of 2 MW.

  12. Paleoclassical transport explains electron transport barriers in RTP and TEXTOR

    NARCIS (Netherlands)

    Hogeweij, G. M. D.; Callen, J.D.

    2008-01-01

    The recently developed paleoclassical transport model sets the minimum level of electron thermal transport in a tokamak. This transport level has proven to be in good agreement with experimental observations in many cases when fluctuation-induced anomalous transport is small, i.e. in (near-) ohmic

  13. Optical-Thickness Corrections to Transient Ece Temperature-Measurements in Tokamak and Stellarator Plasmas

    NARCIS (Netherlands)

    Peters, M.; Gorini, G.; Mantica, P.

    1995-01-01

    The conditions are examined under which optical thickness (tau) corrections to electron cyclotron emission (ECE) measurements of electron temperature (T-e) can be neglected. By means of simple algebra it is demonstrated that for measurements of T-e transients the ECE radiation temperature (T-rad)

  14. Observations of Infrared Radiation During Disruptions in Textor - Heat Pulses and Runaway Electrons

    NARCIS (Netherlands)

    R. Jaspers,; Grewe, T.; Finken, K.H.; KramerFlecken, A.; Cardozo, N. J. L.; Mank, G.; Waidmann, G.

    1995-01-01

    Disruptions are studied in TEXTOR using two infrared cameras. In the thermal quench phase, fast changing heat fluxes are observed, each delivering energies larger than 1 kJ/m(2) to the limiter. These bursts are correlated with an electron temperature pulse near the limiter and an increased release

  15. Experimental characterization of anomalous strong scattering of mm-waves in TEXTOR plasmas with rotating islands

    DEFF Research Database (Denmark)

    Nielsen, Stefan Kragh; Salewski, Mirko; Westerhof, E

    2013-01-01

    Anomalous scattering of high power millimetre waves from gyrotrons at 140 and 110 GHz is investigated for plasma with rotating islands at TEXTOR. The magnetic field and plasma density influence the spectral content of the scattered waves and their power levels significantly. Anomalous strong...

  16. Modification of the collective Thomson scattering radiometer in the search for parametric decay on TEXTOR

    DEFF Research Database (Denmark)

    Nielsen, Stefan Kragh; Salewski, Mirko; Bongers, W.

    2012-01-01

    Strong scattering of high-power millimeter waves at 140 GHz has been shown to take place in heating and current-drive experiments at TEXTOR when a tearing mode is present in the plasma. The scattering signal is at present supposed to be generated by the parametric decay instability. Here we descr...... with independent backscattering radiometer data....

  17. Measurement of the energy content of the JET tokamak plasma with a diamagnetic loop

    Science.gov (United States)

    Tonetti, G.; Christiansen, J. P.; de Kock, L.

    1986-08-01

    An accurate and reliable measurement of poloidal β is essential to assess the performances of Joint European Torus (JET). The diamagnetic loop can measure β values as low as 0.1 in JET discharges with a plasma current larger than 2×106 A. The instrumentation used includes a flux loop rigidly fitted on a toroidal field (TF) coil, a large Rogowski coil measuring the TF busbar current, and a displacement gauge measuring the TF coil expansion. The fluxes to be compensated originate, in order of importance, from the TF current, the eddy current in the vessel, the TF coil expansion, and the stray coupling with the poloidal fields. The TF and eddy currents must be particularly well compensated on JET since the plasma current starts before the toroidal field has reached its plateau value. Comparison between the diamagnetic and other evaluations of β shows a good agreement.

  18. Density profile measurements by amplitude modulation reflectometry on the TJ-1 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Luna, E. de la; Zhuravlev, V.; Branas, B.; Sanchez, J.; Estrada, T. [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas (CIEMAT), Madrid (Spain); Segovia, J.; Oramas, J.L. [Universidad Politecnica de Madrid (Spain)

    1993-12-31

    Amplitude Modulation (AM) reflectometry has been proposed as an alternative method to the more traditional swept frequency reflectometry systems which can be strongly affected by the presence of plasma density fluctuations. Compared to the time domain systems, the time measurements are replaced in AM reflectometry by much simpler phase delay measurements. In AM reflectometry, the time delay of the microwave beam propagating to the reflecting layer and back is directly obtained through the phase delay of the modulating envelope, which is directly measured in a linear (0-2{pi}) phase meter. This method avoids operation with multifringe counters and is a very attractive alternative for real time determination of the plasma position and density profile. In order to achieve real time density profile monitoring, fast analysis methods are necessary, first results of Neural Networks application to the problem of fast inversion of the density profile are shown. (author) 3 refs., 5 figs.

  19. Comparison of two models for the X-ray dispersion produced in a Novillo Tokamak with measurements make with thermoluminescent dosemeters; Comparacion de dos modelos para la dispersion de rayos X producidos en un Novillo Tokamak con mediciones efectuadas con dosimetros termoluminiscentes

    Energy Technology Data Exchange (ETDEWEB)

    Flores O, A.; Castillo, A.; Barocio, S.R.; Melendez L, L.; Chavez A, E.; Cruz C, G.J.; Lopez, R.; Olayo, M.G.; Gonzalez M, P. [Instituto Nacional de Investigaciones Nucleares, 52045 Salazar, Estado de Mexico (Mexico); Azorin N, J. [Universidad Autonoma Metropolitana Iztapalapa, 09340 Mexico D.F. (Mexico)

    1999-07-01

    It was presented the results to study about the X-ray dispersion produced in the Novillo Tokamak using thermoluminescent dosemeters (DTL). The measurements were make in the equatorial plane of Tokamak, along twelve radial directions. The dispersion is observed due to the radiation interaction with walls surrounding the machine. It was proposed two types of heuristic mathematical methods for describing the X-ray dispersion, comparing them with the experimental data obtained with Dtl. The predictions of both models are adjusted well to the experimental data. (Author)

  20. A method for measuring plasma position in TJ-I Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Quin, J.; TJ-I, Team

    1993-07-01

    A method using pairs of Mirnov coils to measure the plasma position in TJ-I is presented. The simple toroidal filament model which neglects the effect of plasma current density profile has proven to be acceptable within the experimental accuracy. The effect of plasma current density profile remains to be small, if the plasma current density profile has a quadratic form. (Author) 5 refs.

  1. Effect of 3D magnetic perturbations on divertor conditions and detachment in tokamak and stellarator

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, J-W; Briesemester, A. R.; Kobayashi, M.; Lore, J. D.; Schmitz, O.; Diallo, A.; Gray, T. K.; Lasnier, C. J.; LeBlanc, B. P.; Maingi, R.; McLean, A. G.; Sabbagh, S. A.; Soukhanovskii, V. A.

    2017-06-22

    Enhanced perpendicular heat and momentum transport induces parallel pressure loss leading to divertor detachment, which can be produced by the increase of density in 2D tokamaks. However, in the 3D configurations such as tokamaks with 3D fields and stellarators, the fraction of perpendicular transport can be higher even in a lower density regime, which could lead to the early transition to detachment without passing through the high-recycling regime. 3D fields applied to the limiter tokamak plasmas produce edge stochastic layers close to the last closed flux surface (LCFS), which can allow for enhanced perpendicular transport and indeed the absence of high recycling regime and early detachment have been observed in TEXTOR and Tore Supra. However, in the X-point divertor tokamaks with the applied 3D fields, the parallel transport is still dominant and the detachment facilitation has not been observed yet. Rather, 3D fields affected detachment adversely under certain conditions, either by preventing detachment onset as seen in DIII-D or by re-attaching the existing detached plasma as shown in NSTX. The possible way for strong 3D effects to induce access to the early detachment in divertor tokamaks appears to be via significant perpendicular loss of parallel momentum by frictional force for the counter-streaming flows between neighboring flow channels in the divertor. In principle, the adjacent lobes in the 3D divertor tokamak may generate the counter-streaming flow channels. However, an EMC3-EIRENE simulation for ITER H-mode plasmas demonstrated that screened RMP leads to significantly reduced counter-flows near the divertor target, therefore the momentum loss effect leading to detachment facilitation is expected to be small. This is consistent with the observation in LHD, which showed screening (amplification) of RMP fields in the attachment (stable detachment) case. Work for optimal parameter window for best divertor operation scenario is needed particularly for

  2. Effect of 3D magnetic perturbations on divertor conditions and detachment in tokamak and stellarator

    Science.gov (United States)

    Ahn, J.-W.; Briesemester, A. R.; Kobayashi, M.; Lore, J. D.; Schmitz, O.; Diallo, A.; Gray, T. K.; Lasnier, C. J.; LeBlanc, B. P.; Maingi, R.; McLean, A. G.; Sabbagh, S. A.; Soukhanovskii, V. A.

    2017-08-01

    Enhanced perpendicular heat and momentum transport induces parallel pressure loss leading to divertor detachment, which can be produced by the increase of density in 2D tokamaks. However, in the 3D configurations such as tokamaks with 3D fields and stellarators, the fraction of perpendicular transport can be higher even in a lower density regime, which could lead to the early transition to detachment without passing through the high-recycling regime. 3D fields applied to the limiter tokamak plasmas produce edge stochastic layers close to the last closed flux surface (LCFS), which can allow for enhanced perpendicular transport and indeed the absence of high recycling regime and early detachment have been observed in TEXTOR and Tore Supra. However, in the X-point divertor tokamaks with the applied 3D fields, the parallel transport is still dominant and the detachment facilitation has not been observed yet. Rather, 3D fields affected detachment adversely under certain conditions, either by preventing detachment onset as seen in DIII-D or by re-attaching the existing detached plasma as shown in NSTX. The possible way for strong 3D effects to induce access to the early detachment in divertor tokamaks appears to be via significant perpendicular loss of parallel momentum by frictional force for the counter-streaming flows between neighboring flow channels in the divertor. In principle, the adjacent lobes in the 3D divertor tokamak may generate the counter-streaming flow channels. However, an EMC3-EIRENE simulation for ITER H-mode plasmas demonstrated that screened RMP leads to significantly reduced counter-flows near the divertor target, therefore the momentum loss effect leading to detachment facilitation is expected to be small. This is consistent with the observation in LHD, which showed screening (amplification) of RMP fields in the attachment (stable detachment) case. Work for optimal parameter window for best divertor operation scenario is needed particularly for

  3. Confocal microscopy: A new tool for erosion measurements on large scale plasma facing components in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Gauthier, E., E-mail: eric.gauthier@cea.fr [CEA/DSM/IRFM, CEA Cadarache, Saint-Paul-lez-Durance (France); Brosset, C.; Roche, H.; Tsitrone, E.; Pégourié, B.; Martinez, A. [CEA/DSM/IRFM, CEA Cadarache, Saint-Paul-lez-Durance (France); Languille, P. [PIIM, CNRS-Université de Provence, Centre de St Jérôme, 13397 Marseille, Cedex 20 (France); Courtois, X.; Lallier, Y. [CEA/DSM/IRFM, CEA Cadarache, Saint-Paul-lez-Durance (France); Salami, M. [AVANTIS CONCEPT, 75 Rue Marcelin Berthelot, 13858 Aix en Provence (France)

    2013-07-15

    A diagnostic based on confocal microscopy was developed at CEA Cadarache in order to measure erosion on large plasma facing components during shutdown in situ in Tore Supra. This paper describes the diagnostic and presents results obtained on Beryllium and Carbon Fibre Composite (CFC) materials. Erosion in the range of 800 μm was found on one sector of the Toroidal Pumped Limiter (TPL) which provides, by integration to the full limiter a net carbon erosion of about 900 g over the period 2002–2007.

  4. FPGA based phase detection technique for electron density measurement in SST-1 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Pramila, E-mail: pramila@ipr.res.in; Mandaliya, Hitesh; Rajpal, Rachana; Kaur, Rajwinder

    2016-11-15

    A multi-channel signal-conditioning and phase-detection concept is implemented in the prototype design using the high-precision OPAMP, high-speed comparators, high Q filters, high-density FPGA (Field Programmable Gate array), 10 MHz parallel-multiplying DACs (Digital to Analog Converter), etc. The complete digital-logic for the phase-detection is implemented inside the logic cells of FPGA using VHDL code, with high speed 100 MHz clock generated from Digital Clock Manager (DCM), which is used to measure the time elapsed between zero crossings of the two signals coming from reference and probe paths of the diagnostics. The logic is implemented to measure either leading or lagging phase and also to accumulate the total phase difference throughout the shot duration with the maximum value of accumulated phase of 5760 (16 cycles × 360°) degree and a resolution of 3.6 °. A precision high speed and high bandwidth (80 MHz) operational amplifiers are used as the front end-electronics component for conditioning the high-frequency (1 MHz) and low amplitude signal (μV). The hardware detail, implementation concept in FPGA and testing results will be presented in the paper.

  5. A fast-time-response extreme ultraviolet spectrometer for measurement of impurity line emissions in the Experimental Advanced Superconducting Tokamak.

    Science.gov (United States)

    Zhang, Ling; Morita, Shigeru; Xu, Zong; Wu, Zhenwei; Zhang, Pengfei; Wu, Chengrui; Gao, Wei; Ohishi, Tetsutarou; Goto, Motoshi; Shen, Junsong; Chen, Yingjie; Liu, Xiang; Wang, Yumin; Dong, Chunfeng; Zhang, Hongmin; Huang, Xianli; Gong, Xianzu; Hu, Liqun; Chen, Junlin; Zhang, Xiaodong; Wan, Baonian; Li, Jiangang

    2015-12-01

    A flat-field extreme ultraviolet (EUV) spectrometer working in the 20-500 Å wavelength range with fast time response has been newly developed to measure line emissions from highly ionized tungsten in the Experimental Advanced Superconducting Tokamak (EAST) with a tungsten divertor, while the monitoring of light and medium impurities is also an aim in the present development. A flat-field focal plane for spectral image detection is made by a laminar-type varied-line-spacing concave holographic grating with an angle of incidence of 87°. A back-illuminated charge-coupled device (CCD) with a total size of 26.6 × 6.6 mm(2) and pixel numbers of 1024 × 255 (26 × 26 μm(2)/pixel) is used for recording the focal image of spectral lines. An excellent spectral resolution of Δλ0 = 3-4 pixels, where Δλ0 is defined as full width at the foot position of a spectral line, is obtained at the 80-400 Å wavelength range after careful adjustment of the grating and CCD positions. The high signal readout rate of the CCD can improve the temporal resolution of time-resolved spectra when the CCD is operated in the full vertical binning mode. It is usually operated at 5 ms per frame. If the vertical size of the CCD is reduced with a narrow slit, the time response becomes faster. The high-time response in the spectral measurement therefore makes possible a variety of spectroscopic studies, e.g., impurity behavior in long pulse discharges with edge-localized mode bursts. An absolute intensity calibration of the EUV spectrometer is also carried out with a technique using the EUV bremsstrahlung continuum at 20-150 Å for quantitative data analysis. Thus, the high-time resolution tungsten spectra have been successfully observed with good spectral resolution using the present EUV spectrometer system. Typical tungsten spectra in the EUV wavelength range observed from EAST discharges are presented with absolute intensity and spectral identification.

  6. Measurements and modelling of plasma response field to RMP on the COMPASS tokamak

    Czech Academy of Sciences Publication Activity Database

    Markovič, Tomáš; Liu, Y.Q.; Cahyna, Pavel; Pánek, Radomír; Peterka, Matěj; Aftanas, Milan; Bílková, Petra; Böhm, Petr; Imríšek, Martin; Háček, Pavel; Havlíček, Josef; Havránek, Aleš; Komm, Michael; Urban, Jakub; Weinzettl, Vladimír

    2016-01-01

    Roč. 56, č. 9 (2016), č. článku 092010. ISSN 0029-5515. [Joint Meeting of the 597th Wilhelm and Else Heraeus Seminar / 7th International Workshop on Stochasticity in Fusion Plasmas. Greifswald, 10.09.2015-12.09.2015] R&D Projects: GA MŠk(CZ) 8D15001; GA MŠk(CZ) LM2015045; GA ČR(CZ) GA14-35260S; GA AV ČR(CZ) GA16-24724S EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : RMP * magnetic measurements * MARS -F Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.307, year: 2016 http://iopscience.iop.org/article/10.1088/0029-5515/56/9/092010/meta

  7. Stability analysis of tokamak plasmas; Analyse de stabilite de plasmas de tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bourdelle, C

    2000-10-01

    In a tokamak plasma, the energy transport is mainly turbulent. In order to increase the fusion reactions rate, it is needed to improve the energy confinement. The present work is dedicated to the identification of the key parameters leading to plasmas with a better confined energy in order to guide the future experiments. For this purpose, a numerical code has been developed. It calculates the growth rates characterizing the instabilities onset. The stability analysis is completed by the evaluation of the shearing rate of the rotation due to the radial electric field. When this shearing rate is greater than the growth rate the ion turbulence is fully stabilised. The shearing rate and the growth rate are determined from the density, temperature and security factor profiles of a given plasma. Three types of plasmas have been analysed. In the Radiative Improved modes of TEXTOR, high charge number ions seeding lowers the growth rates. In Tore Supra-high density plasmas, a strong magnetic shear and/or a more efficient ion heating linked to a bifurcation of the toroidal rotation direction (which is not understood) trigger the improvement of the confinement. In other Tore Supra plasmas, locally steep electron pressure gradients have been obtained following magnetic shear reversal. This locally negative magnetic shear has a stabilizing effect. In these three families of plasmas, the growth rates decrease, the confinement improves, the density and temperature profiles are steeper. This steepening induces an increase of the rotation shearing rate, which then maintains the confinement high quality. (author)

  8. Experimental measurement of magnetic field null in the vacuum chamber of KTM tokamak based on matrix of 2D Hall sensors

    Energy Technology Data Exchange (ETDEWEB)

    Shapovalov, G.; Chektybayev, B., E-mail: chektybaev@nnc.kz; Sadykov, A.; Skakov, M.; Kupishev, E.

    2016-11-15

    Experimental technique of measurement of magnetic field null region inside of the KTM tokamak vacuum chamber has been developed. Square matrix of 36 2D Hall sensors, which used in the technique, allows carrying out direct measurements of poloidal magnetic field dynamics in the vacuum chamber. To better measuring accuracy, Hall sensor’s matrix was calibrated with commercial Helmholtz coils and in situ measurement of defined magnetic field from poloidal and toroidal coils. Standard KTM Data-Acquisition System has been used to collect data from Hall sensors. Experimental results of measurement of magnetic field null in the vacuum chamber of KTM are shown in the paper. Additionally results of the magnetic field null reconstruction from signals of inductive total flux loops are shown in the paper.

  9. Measurement of electron density of the plasma in the Tokamak TCABR, through Thomson scattering diagnostic; Medida da densidade eletronica do plasma no Tokamak TCABR, atraves do diagnostico Espalhamento Thomson

    Energy Technology Data Exchange (ETDEWEB)

    Jeronimo, Leonardo Cunha

    2013-07-01

    Over the last few years is remarkable, so increasingly evident the need for a new source of energy for mankind. One promising option is through nuclear fusion, where the plasma produced in the reactor can be converted into electrical energy. Therefore, knowing the characteristics of this plasma is very important to control it and understand it so desirable. One of the diagnostic options is called Thomson scattering . This is considered the most reliable method for the determination of important plasma parameters such as temperature and electron density, and may also help in the study and explanation of various internal mechanisms. The great advantage lies in the tact that they consist of a direct measurement and nonperturbative. But it is a diagnosis whose installation and execution is admittedly complex, limiting it only a few laboratories in the fíeld of fusion for the world. Among the main difficulties, wc can highlight the fact that the scattered signal is very small, thus requiring a large increase of the incident power. Moreover, the external physical conditions can cause mechanical vibrations that eliminate or minimize them as much as possible, is a great challenge, considering the optical micrometrically very sensitive and needs involved in the system. This work describes the entire process of installation and operation of Thomson scattering diagnostic in tokamak TCABR and through this diagnosis, we work on results of electron temperature, to finally be able to calculate the electron density of the plasma. (author)

  10. 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

  11. Measurement of plasma rotation velocities with electrode biasing in the Saskatchewan Torus-Modified (STOR-M) tokamak

    Science.gov (United States)

    Xiao, C.; Jain, K. K.; Zhang, W.; Hirose, A.

    1994-07-01

    In the Saskatchewan Torus-Modified (STOR-M) tokamak [Phys. Fluids B 4, 3277 (1992)], application of a negative bias results in large negative radial electric field, Er, at the plasma edge, reduced plasma toroidal rotation velocity, and a large poloidal rotation in the electron diamagnetic drift direction. Conversely, a positive bias leads to a relatively small negative Er at the plasma edge, a positive Er in the scrape-off layer, increased toroidal rotation, and an increased poloidal rotation speed in the ion diamagnetic drift direction. Increases in edge plasma density and steepening of its radial profile have also been observed for both polarities.

  12. Relativistic down-shift frequency effect on the application of electron cyclotron emission measurements to JT-60U tokamak plasmas. Second harmonics

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Masayasu; Isei, Nobuaki; Ishida, Sinichi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1995-11-01

    Effect of relativistic frequency down-shift on the determination of the electron temperature profile from electron cyclotron emission(ECE) in JT-60U tokamak plasmas is studied. The radial shift of the electron temperature profile due to the effects is not negligible, compared with the spatial resolution of ECE measurement systems of JT-60U. Therefore it is necessary to correct the effect for precise measurement of the electron temperature profile. Dependencies of the shifted frequency on the electron density, electron temperature and toroidal magnetic field are studied for the uniform electron density and parabolic electron temperature profile in JT-60U. It is revealed to be necessary for the estimation of shift due to the relativistic down-shift frequency to take into account of the optical thickness. (author).

  13. The electron cyclotron absorption diagnostic at the Rijnhuizen tokamak project

    NARCIS (Netherlands)

    van Gelder, J. F. M.; Miedema, H. S.; Donne, A. J. H.; Oomens, A. A. M.; Schüller, F. C.

    1997-01-01

    A new 20-channel electron cyclotron absorption diagnostic has been developed at the Rijnhuizen tokamak project. It is the first time the electron pressure profile in a tokamak plasma can be measured directly with a time resolution of 1 ms. The diagnostic measures simultaneously the emission and

  14. Experimental measurements of energy transfer and nonlinear interaction in turbulence at the sino-united spherical tokamak

    Science.gov (United States)

    Chai, Song; Xu, Yuhong; Gao, Zhe; Wang, Wenhao; Liu, Yangqing; Tan, Yi

    2017-03-01

    The characteristics of the energy transfer and nonlinear coupling among edge electromagnetic turbulence have been dedicatedly studied in various discharge stages at the sino-united spherical tokamak using multiple Langmuir and magnetic probe arrays. The wavelet bispectral analysis and the modified Kim's method are applied to investigate turbulence properties and their linear growth/damping and nonlinear energy transfer rates, along with multi-field turbulence interactions. The results show diverse features in the linear growth and nonlinear energy transfer between multi-field fluctuations during the current ramp-up, stationary, and internal connection event discharge phases. The diversity implies the importance to develop more sophisticated multi-field models to directly estimate the energy transfer rate among multiple turbulent fields.

  15. Infrared surface temperature measurements for long pulse operation, and real time feedback control in Tore-Supra, an actively cooled Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Guilhem, D.; Adjeroud, B.; Balorin, C.; Buravand, Y.; Bertrand, B.; Bondil, J.L.; Desgranges, C.; Gauthier, E.; Lipa, M.; Messina, P.; Missirlian, M.; Mitteau, R.; Moulin, D.; Pocheau, C.; Portafaix, C.; Reichle, R.; Roche, H.; Saille, A.; Vallet, S

    2004-07-01

    Tore-Supra has a steady-state magnetic field using super-conducting magnets and water-cooled plasma facing components for high performances long pulse plasma discharges. When not actively cooled, plasma-facing components can only accumulate a limited amount of energy since the temperature increase continuously (T proportional to {radical}(t)) during the discharge until radiation cooling is equal to the incoming heat flux (T > 1800 K). Such an environment is found in most today Tokamaks. In the present paper we report the recent results of Tore-Supra, especially the design of the new generation of infrared endoscopes to measure the surface temperature of the plasma facing components. The Tore-Supra infrared thermography system is composed of 7 infrared endoscopes, this system is described in details in the paper, the new JET infrared thermography system is presented and some insights of the ITER set of visible/infrared endoscope is given. (authors)

  16. Testing the {rho}* scaling of thermal transport models: predicted and measured temperatures in the Tokamak Fusion Test Reactor dimensionless scaling experiments

    Energy Technology Data Exchange (ETDEWEB)

    Mikkelsen, D.R.; Scott, S.D. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Dorland, W. [Texas Univ., Austin, TX (United States). Inst. for Fusion Studies

    1997-04-01

    Theoretical predictions of ion and electron thermal diffusivities are tested by comparing calculated and measured temperatures in low (L) mode plasmas from the Tokamak Fusion Test Reactor [D. J. Grove and D. M. Meade, Nucl. Fusion 25 , 1167 (1985)] nondimensional scaling experiments. The DIII-D [J. L. Luxon and L. G. Davis, Fusion Technol. 8 , 441 (1985)] L-mode {rho}* scalings, the transport models of Rebut-Lallia-Watkins (RLW), Boucher`s modification of RLW, and the Institute for Fusion Studies-Princeton Plasma Physics Laboratory (IFS-PPPL) model for transport due to ion temperature gradient modes are tested. The predictions use the measured densities in order to include the effects of density profile shape variations on the transport models. The uncertainties in the measured and predicted temperatures are discussed. The predictions based on the DIII- D scalings are within the measurement uncertainties. All the theoretical models predict a more favorable {rho}* dependence for the ion temperatures than is seen. Preliminary estimates indicate that sheared ow stabilization is important for some discharges, and that inclusion of its effects may bring the predictions of the IFS-PPPL model into agreement with the experiments.

  17. Stark-effect measurement of high FEL (free-electron laser) electric fields in MTX (Microwave Tokamak Experiment) by laser-aided particle-probe spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Oda, T.; Takiyama, K. (Hiroshima Univ. (Japan)); Odajima, K.; Ohasa, K.; Shiho, M. (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)); Mizuno, K. (California Univ., Davis, CA (USA) Sandia National Labs., Livermore, CA (USA)); Foote, J.H.; Nilson, D.G. (Sandia National Labs., Livermore, CA (USA))

    1990-05-04

    We are constructing a diagnostic system to measure the electric field (>100 kV/cm) of a free-electron laser (FEL) beam when injected into the plasma of the Microwave Tokamak Experiment (MTX). The apparatus allows a crossed-beam measurement, with 2-cm spatial resolution in the plasma, involving the FEL beam (with 140-GHz, {approx}1-GW ECH pulses), a neutral-helium beam, and a dye-laser beam. After the laser beam pumps metastable helium atoms to higher excited states, their decay light is detected by a collimated optical system. Because of the Stark effect due to the FEL electric field ({rvec E}), a forbidden transition can be strongly induced. The intensity of emitted light resulting from the forbidden transition is proportional to E{sup 2}. Because photon counting rates are calculated to be low, extra effort is made to minimize background and noise levels. It is possible that the lower {rvec E} of an MTX gyrotron-produced ECH beam with its longer-duration pulses also can be measured using this method. Other applications may include measurements of ion temperature (using charge-exchange recombination), edge-density fluctuations, and core impurity concentrations. 11 refs., 2 figs., 2 tabs.

  18. Start-effect measurement of high FEL (free-electron laser) electric fields in MTX (Microwave Tokamak Experiment) by laser-aided particle-probe spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Oda, T.; Takiyama, K. (Hiroshima Univ. (Japan)); Odajima, K.; Ohasa, K.; Shiho, M. (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)); Mizuno, K. (California Univ., Davis, CA (USA) Lawrence Livermore National Lab., CA (USA)); Foote, J.H.; Nilson, D.G. (Lawrence Livermore National Lab., CA (USA))

    1990-05-10

    We are constructing a diagnostic system to measure the electric field (>100 kV/cm) of a free-electron laser (FEL) beam when injected into the plasma of the Microwave Tokamak Experiment (MTX). The apparatus allows a crossed-beam measurement, with 2-cm spatial resolution in the plasma, involving the FEL beam (with 140-GHz, {approx}1-GW ECH pulses), a neutral-helium beam, and a dye-laser beam. After the laser beam pumps metastable helium atoms to higher excited states, their decay light is detected by an efficient optical system. Because of the Stark effect arising from the FEL electric field ({rvec E}), a forbidden transition can be strongly induced. The intensity of emitted light resulting from the forbidden transition is proportional to E{sup 2}. Because photon counting rates are estimated to be low, extra effort is made to minimize background and noise levels. It is possible that the lower {rvec E} of an MTX gyrotron-produced ECH beam with its longer-duration pulses can also be measured using this method. Other applications of the apparatus described here may include measurements of ion temperature (using charge-exchange recombination), edge-density fluctuations, and core impurity concentrations.

  19. Weaving through a cryptic species: Comparing the Neotropical ants Camponotus senex and Camponotus textor (Hymenoptera: Formicidae).

    Science.gov (United States)

    Fox, Eduardo Gonçalves Paterson; Solis, Daniel Russ; Lazoski, Cristiano; Mackay, William

    2017-08-01

    Camponotus senex (Fr. Smith 1858) and Camponotus textor Forel, 1899 are commonly confused species in the New World tropics. We provide morphological characteristics based on the larvae and adults, behavioural differences, together with evidence from molecular markers (cuticular hydrocarbon profiles, venom differences, nuclear ribosomal ITS-1, and mtDNA COI sequence comparisons) to separate the two species, demonstrating they are not immediately closely related. In conclusion we suggest new reliable morphological characters which can benefit from deeper phenetic analysis, and support the contextual usefulness of non-morphological tools in resolving sibling ant species. Copyright © 2017 Elsevier Ltd. All rights reserved.

  20. Optimal control of a coupled partial and ordinary differential equations system for the assimilation of polarimetry Stokes vector measurements in tokamak free-boundary equilibrium reconstruction with application to ITER

    Science.gov (United States)

    Faugeras, Blaise; Blum, Jacques; Heumann, Holger; Boulbe, Cédric

    2017-08-01

    The modelization of polarimetry Faraday rotation measurements commonly used in tokamak plasma equilibrium reconstruction codes is an approximation to the Stokes model. This approximation is not valid for the foreseen ITER scenarios where high current and electron density plasma regimes are expected. In this work a method enabling the consistent resolution of the inverse equilibrium reconstruction problem in the framework of non-linear free-boundary equilibrium coupled to the Stokes model equation for polarimetry is provided. Using optimal control theory we derive the optimality system for this inverse problem. A sequential quadratic programming (SQP) method is proposed for its numerical resolution. Numerical experiments with noisy synthetic measurements in the ITER tokamak configuration for two test cases, the second of which is an H-mode plasma, show that the method is efficient and that the accuracy of the identification of the unknown profile functions is improved compared to the use of classical Faraday measurements.

  1. Advanced commercial tokamak study

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. Electron cyclotron emission diagnostics on KSTAR tokamak.

    Science.gov (United States)

    Jeong, S H; Lee, K D; Kogi, Y; Kawahata, K; Nagayama, Y; Mase, A; Kwon, M

    2010-10-01

    A new electron cyclotron emission (ECE) diagnostics system was installed for the Second Korea Superconducting Tokamak Advanced Research (KSTAR) campaign. The new ECE system consists of an ECE collecting optics system, an overmode circular corrugated waveguide system, and 48 channel heterodyne radiometer with the frequency range of 110-162 GHz. During the 2 T operation of the KSTAR tokamak, the electron temperatures as well as its radial profiles at the high field side were measured and sawtooth phenomena were also observed. We also discuss the effect of a window on in situ calibration.

  3. Texas Experimental Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. Application of Thomson scattering at 1.06{mu}m as a diagnostic for spatial profile measurements of electron temperature and density on the TCV tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Franke, S. [Ecole Polytechnique Federale, Lausanne (Switzerland). Centre de Recherche en Physique des Plasma (CRPP)

    1997-04-01

    The variable configuration tokamak, TCV, in operation at CRPP since the end of 1991, is a particularly challenging machine with regard to the experimental system that must provide essential information regarding properties of confined plasmas with strongly shaped, non-circular cross-sections. The importance of the energy confinement issue in a machine designed specifically for the investigation of the effect of plasma shape on confinement and stability is self-evident, as is the necessity for a diagnostic capable of providing the profiles of electron temperature and density required for evaluation of this confinement. For TCV, a comprehensive Thomson Scattering (TS) diagnostic was the natural choice, specifically owing to the resulting spatially localized and time resolved measurement. The details of the system installed on TCV, together with the results obtained from the diagnostic comprise the subject matter of this thesis. A first version of the diagnostic was equipped with only ten observation volumes. In this case, adequate spatial resolution can only be maintained if measurements are limited to plasmas located in the upper half of the highly elongated TCV vacuum vessel. The system has recently been upgraded through the addition of a further fifteen observation volumes, together with major technical improvements in the scattered light detection system. This new version now permits TS observations in all TCV plasma configurations, including equilibria produced in the lower and upper halves of the vacuum vessel and the highly elongated plasmas now routinely created. Whilst a description of the new detection system along with some results obtained using the extended set of observation volumes are included, this thesis reports principally on the hardware details of and the interpretation of data from the original, ten observation volume system. (author) figs., tabs., 75 refs.

  5. Magnetic sensor for steady state tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-06-01

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

  6. Measurement of the surface morphology of plasma facing components on the EAST tokamak by a laser speckle interferometry approach

    Science.gov (United States)

    Hongbei, WANG; Xiaoqian, CUI; Yuanbo, LI; Mengge, ZHAO; Shuhua, LI; Guangnan, LUO; Hongbin, DING

    2018-03-01

    The laser speckle interferometry approach provides the possibility of an in situ optical non-contacted measurement for the surface morphology of plasma facing components (PFCs), and the reconstruction image of the PFC surface morphology is computed by a numerical model based on a phase unwrapping algorithm. A remote speckle interferometry measurement at a distance of three meters for real divertor tiles retired from EAST was carried out in the laboratory to simulate a real detection condition on EAST. The preliminary surface morphology of the divertor tiles was well reproduced by the reconstructed geometric image. The feasibility and reliability of this approach for the real-time measurement of PFCs have been demonstrated.

  7. Measurement of the passive fast-ion D-alpha emission on the NSTX-U tokamak

    Science.gov (United States)

    Hao, G. Z.; Heidbrink, W. W.; Liu, D.; Podesta, M.; Stagner, L.; Bell, R. E.; Bortolon, A.; Scotti, F.

    2018-02-01

    On National Spherical Torus Experiment Upgrade, the passive fast-ion D-alpha (passive-FIDA) spectra from charge exchange (CX) between the beam ions and the background neutrals are measured and simulated. The results indicate that the passive-FIDA signal is measurable and comparable to the active-FIDA on several channels, such as at the major radius R = 117 cm. Here, active-FIDA means the active D-alpha emission from the fast ions that CX with the injected neutrals. The shapes of measured spectra are in agreement with FIDASIM simulations on many fibers. Furthermore, the passive-FIDA spatial profile agrees with the simulation. When making measurements of active-FIDA in the edge region using time-slice subtraction, variations in the passive-FIDA contribution to the signal should be considered.

  8. Tokamak turbulence with stochastic field lines

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, P.; Garbet, X.; Ghendrih, Ph

    1998-03-01

    Three-dimensional numerical simulations of ballooning turbulence in a tokamak plasma with stochastic magnetic field lines are presented. Three main features are observed. First, the level of pressure fluctuations decreases in the ergodic layer. Secondly, this is essentially due to a suppression of large scale structures. Finally, the turbulent heat diffusivity does not decrease in the stochastic are due to an increase of electric fluctuations. These observations are in agreement with turbulence measurements on Tore Supra. (author) 27 refs.

  9. Variability of carbonaceous aerosols, ozone and radon at Piton Textor, a mountain site on Reunion island (southwestern Indian Ocean)

    Energy Technology Data Exchange (ETDEWEB)

    Bhugwant, C.; Riviere, E.; Leveau, J. [Univ. de la La Reunion, Saint Denis (France). Lab. de Physique de l' Atmosphere; Keckhut, P. [CNRS, Verrieres-Le-Buisson (France). Service d' Aeronomie

    2001-11-01

    Black carbon (BC) was monitored during 1997-1999 in the lower troposphere of the southern Indian Ocean at La Reunion island (21.5 deg S, 55.5 deg E). BC concentrations obtained at Piton Textor, an altitude site (2150 m) representative of free troposphere, exhibited diurnal patterns and concentrations different from urban locations on the island, with maximum concentrations observed at daytime ({approx} 50-150 ng/m{sup 3}) and minimum levels ({approx} 10-70 ng/m{sup 3}) at night-time. BC diurnal variation is anti-correlated with diurnal ozone measured semi-continuously in parallel during 1998-1999, suggesting possible interaction of ozone and precursors (NO{sub x}, VOC, etc.) on carbonaceous aerosols, especially at night-time. Daytime BC enhancement may be explained by dynamical processes, due to up draught of air masses from lower levels to the troposphere, while at night-time, this process is reversed. Daytime ozone depletion is governed by photochemical processes, due to low precursor levels, while night-time ozone recovery is mainly driven by dynamical processes from upper tropospheric layers. Night-time BC and ozone in the lower troposphere show a marked seasonal pattern too, with minimum levels during austral summer ({approx} 15 ng/m{sup 3}, 22 ppbv), secondary peaks in autumn and spring ({approx} 35 ng/m{sup 3}, 36 ppbv) and maximum values during austral winter ({approx} 70 ng/m{sup 3}, 41 ppbv) respectively. Night-time BC and ozone seasonalities are concordant with night-time radon seasonal trend in the lower troposphere, indicating that sampled air masses have mainly a marine origin in summer, off the African biomass burning season, and a continental origin in austral winter and spring. Winter and spring BC and ozone enhancement corroborate with fire-count maximum peaks observed over Africa and Madagascar, suggesting that the main cause is combustion products long-range transported in stable layers evidenced by thermodynamic analysis using 1996-1999 PTU

  10. A closed-loop control system for stabilization of MHD events on TEXTOR

    NARCIS (Netherlands)

    Hennen, B.A.; Westerhof, E.; Oosterbeek, J. W.; Nuij, Pwjm; De Lazzari, D.; Spakman, G. W.; de M. Baar,; Steinbuch, M.

    2009-01-01

    This paper presents an integrated installation that facilitates closed-loop control of magnetohydrodynamic (MHD) events in a tokamak by means of electron cyclotron resonance heating and current drive. Model-based control of an elect ro-mechanical launcher, diagnosis and identification of mode

  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. 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)

  13. The ETE spherical Tokamak project

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

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

  14. 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)

  15. The ARIES tokamak reactor study

    Energy Technology Data Exchange (ETDEWEB)

    1989-10-01

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

  16. Bibliography of fusion product physics in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Hively, L. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sigmar, D. J. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    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.

  17. Modeling of plasma distortions by laser-induced ablation spectroscopy (LIAS) and implications for the interpretation of LIAS measurements

    Science.gov (United States)

    Tokar, M. Z.; Gierse, N.; Philipps, V.; Samm, U.

    2015-09-01

    For the interpretation of the line radiation observed from laser induced ablation spectroscopy (LIAS) such parameters as the density and temperature of electrons within very compact clouds of atoms and singly charged ions of ablated material have to be known. Compared to the local plasma conditions prior to the laser pulse, these can be strongly changed during LIAS since new electrons are generated by the ionisation of particles ejected from the irradiated target. Because of their transience and spatial inhomogeneity it is technically difficult to measure disturbances induced in the plasma by LIAS. To overcome this uncertainty a numerical model has been elaborated, providing a self-consistent description for the spreading of ablated particles and accompanying modifications in the plasma. The results of calculations for LIAS performed on carbon-containing targets in Ohmic and additionally heated discharges in the tokamak TEXTOR are presented. Due to the increase in the electron density the ‘ionisation per photon’ ratio, S/XB factor, is significantly enhanced compared to unperturbed plasma conditions. The impact of the amount of material ablated and of the plasma conditions before LIAS on the level of the S/XB-enhancement is investigated.

  18. Equilibrium reconstruction in the TCA/Br tokamak; Reconstrucao do equilibrio no tokamak TCA/BR

    Energy Technology Data Exchange (ETDEWEB)

    Sa, Wanderley Pires de

    1996-12-31

    The accurate and rapid determination of the Magnetohydrodynamic (MHD) equilibrium configuration in tokamaks is a subject for the magnetic confinement of the plasma. With the knowledge of characteristic plasma MHD equilibrium parameters it is possible to control the plasma position during its formation using feed-back techniques. It is also necessary an on-line analysis between successive discharges to program external parameters for the subsequent discharges. In this work it is investigated the MHD equilibrium configuration reconstruction of the TCA/BR tokamak from external magnetic measurements, using a method that is able to fast determine the main parameters of discharge. The thesis has two parts. Firstly it is presented the development of an equilibrium code that solves de Grad-Shafranov equation for the TCA/BR tokamak geometry. Secondly it is presented the MHD equilibrium reconstruction process from external magnetic field and flux measurements using the Function Parametrization FP method. this method. This method is based on the statistical analysis of a database of simulated equilibrium configurations, with the goal of obtaining a simple relationship between the parameters that characterize the equilibrium and the measurements. The results from FP are compared with conventional methods. (author) 68 refs., 31 figs., 16 tabs.

  19. 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.

  20. 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.

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

    DEFF Research Database (Denmark)

    Moseev, Dmitry

    Fast ions are an essential ingredient in burning nuclear fusion plasmas: they are responsible for heating the bulk plasma, carry a significant amount of plasma current and moreover interact with various magnetohydrodynamic (MHD) instabilities. The collective Thomson scattering (CTS) diagnostic...

  2. STARFIRE: a commercial tokamak reactor

    Energy Technology Data Exchange (ETDEWEB)

    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.

  3. 2D High-Resolution Measurement of High Guide-Field Magnetic Reconnection in TS-3U Spherical Tokamak Merging Experiment

    Science.gov (United States)

    Cao, Qinghong; Akimitsu, Moe; Sawada, Asuka; Tanabe, Hiroshi; Ono, Yasushi

    2017-10-01

    The TS-3U experiment performs magnetic reconnection with a strong guide field by merging two spherical tokamak plasmas. To observe the 2D configuration of the current sheet, we developed a high-resolution 2D magnetic probe array with 260 channels, arranged into 13x10 Bz components and 13x10 Br components, with up to 5 mm spatial resolution spread over a 40 cm x 30 cm poloidal area. The current density Jt, the electric field Et, and the current sheet's effective resistivity ηeff(=Et/Jt) will therefore be followed during the reconnection process. Under a strong guide magnetic field, the sheet resistivity is expected to be almost classical because the sheet thickness is much larger than the ion gyroradius. But resistivity is observed to be anomalous with pileup and plasmoid formation appearing to regulate the reconnection speed. The anomalous increase in resistivity is being studied as a possible cause for the high power heating of fast magnetic reconnection.

  4. Electron cyclotron emission imaging diagnostic system for Rijnhuizen Tokamak Project

    NARCIS (Netherlands)

    Deng, B.H.; Hsia, R. P.; Domier, C.W.; Burns, S. R.; Hillyer, T. R.; N C Luhmann Jr.,; Oyevaar, T.; Donne, A. J. H.; R. T. P. Team,

    1999-01-01

    A 16-channel electron cyclotron emission (ECE) imaging diagnostic system has been developed and installed on the Rijnhuizen Tokamak Project for measuring plasma electron cyclotron emission with a temporal resolution of 2 mu s. The high spatial resolution of the system is achieved by utilizing a low

  5. Loop-voltage tomography in tokamaks using transient synchrotron radiation

    Energy Technology Data Exchange (ETDEWEB)

    Fisch, N.J.; Kritz, A.H. (Princeton Univ., NJ (USA). Plasma Physics Lab.; Hunter Coll., New York, NY (USA). Dept. of Physics)

    1989-07-01

    The loop voltage in tokamaks is particularly difficult to measure anywhere but at the plasma periphery. A brief, deliberate, perturbation of hot plasma electrons, however, produces a transient radiation response that is sensitive to this voltage. We investigate how such a radiation response can be used to diagnose the loop voltage. 24 refs., 6 figs.

  6. 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 ...

  7. Development of atomic beam probe for tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Berta, M., E-mail: bertam@sze.hu [Széchenyi István University, EURATOM Association, Győr (Hungary); Institute of Plasma Physics AS CR, v.v.i., Prague (Czech Republic); Anda, G.; Aradi, M.; Bencze, A.; Buday, Cs.; Kiss, I.G.; Tulipán, Sz.; Veres, G.; Zoletnik, S. [Wigner – RCP, HAS, EURATOM Association, Budapest (Hungary); Havlícek, J.; Háček, P. [Institute of Plasma Physics AS CR, v.v.i., Prague (Czech Republic); Charles University in Prague, Faculty of Mathematics and Physics (Czech Republic)

    2013-11-15

    Highlights: • ABP is newly developed diagnostic. • Unique measurement method for the determination of plasma edge current variations caused by different transient events such as ELMs. • The design process has been fruitfully supported by the physically motivated computer simulations. • Li-BES system has been modified accordingly to the needs of the ABP. -- Abstract: The concept and development of a new detection method for light alkali ions stemming from diagnostic beams installed on medium size tokamak is described. The method allows us the simultaneous measurement of plasma density fluctuations and fast variations in poloidal magnetic field, therefore one can infer the fast changes in edge plasma current. The concept has been worked out and the whole design process has been done at Wigner RCP. The test detector with appropriate mechanics and electronics is already installed on COMPASS tokamak. General ion trajectory calculation code (ABPIons) has also been developed. Detailed calculations show the possibility of reconstruction of edge plasma current density profile changes with high temporal resolution, and the possibility of density profile reconstruction with better spatial resolution compared to standard Li-BES measurement, this is important for pedestal studies.

  8. ICPP: Results from the MAST Spherical Tokamak

    Science.gov (United States)

    Sykes, Alan

    2000-10-01

    The MAST (Mega-Amp Spherical Tokamak) experiment is now fully operational, producing 1MA plasmas with MW level auxiliary heating from Neutral Beam Injection and 60GHz Electron Cyclotron Resonance Heating. Central electron and ion temperatures are both of order 1keV (measured by 30-point Thomson Scattering, Neutral Particle Analyzer and Charge-Exchange spectroscopy respectively). Following boronisation, the Greenwald density limit has been exceeded in double-null divertor discharges by 50operation has been achieved in both Ohmic and NBI heated plasmas. In addition to conventional plasma induction, MAST can employ the `merging-compression' scheme (pioneered on START) producing initial spherical tokamak plasmas of up to 0.5MA without use of flux from the central solenoid. The central solenoid can then be applied to further increase the current at ramp rates of up to 13MA/s; plasma current of 1MA is reached at only one-half of the full solenoid swing. Studies of strike point power loading in both Ohmic and beam heated plasmas have confirmed the result from START that the fraction of power loading on the inboard strike point is lower than predicted from simple models. Comprehensive arrays of halo detectors indicate tolerable levels of halo currents with low asymmetries; an encouraging result for the ST concept, and providing key data to test models. Results from MAST will be used both to extend the conventional tokamak database, and to determine the potential of the ST as a route to fusion power in its own right. Acknowledgement: this work is funded jointly by the UK Department of Trade and Industry and EURATOM. The NBI equipment is on loan from ORNL, the NPA from PPPL.

  9. Degraded Confinement in Tokamak Experiments

    NARCIS (Netherlands)

    Schüller, F. C.

    1994-01-01

    After a review on the state of tokamak transport theory, the methodology to derive experimental results will be described. Examples of confinement in ohmic plasmas and the deterioration with additional heating will be given. Some examples of improved confinement modes will be discussed. Fluctuation

  10. 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.

  11. Prospects for Tokamak Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sheffield, J.; Galambos, J.

    1995-04-01

    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.

  12. Development of 3D ferromagnetic model of tokamak core with strong toroidal asymmetry

    DEFF Research Database (Denmark)

    Markovič, Tomáš; Gryaznevich, Mikhail; Ďuran, Ivan

    2015-01-01

    Fully 3D model of strongly asymmetric tokamak core, based on boundary integral method approach (i.e. characterization of ferromagnet by its surface) is presented. The model is benchmarked on measurements on tokamak GOLEM, as well as compared to 2D axisymmetric core equivalent for this tokamak......, presented in previous work. Linearized model well describes quantitative characteristics of BR field, generated by poloidal field coils located close to core central column, and distorted by ferromagnet. A discrepancy is seen between linearized form of model for BR field generated by coils under...

  13. Microtearing mode (MTM) turbulence in JIPPT-IIU tokamak plasmas

    Science.gov (United States)

    Hamada, Y.; Watari, T.; Nishizawa, A.; Yamagishi, O.; Narihara, K.; Ida, K.; Kawasumi, Y.; Ido, T.; Kojima, M.; Toi, K.; the JIPPT-IIU Group

    2015-04-01

    Magnetic, density and potential fluctuations up to 500 kHz at several spatial points have been observed in the core region of JIPPT-IIU tokamak plasmas using a heavy ion beam probe. The frequency spectra of the density and magnetic oscillations are found to be similar, whereas there are large differences in the phase, coherence and frequency dependences deduced from signals at adjacent sample volumes. These differences allow us to ascribe the detected magnetic fluctuations to the microtearing mode (MTM) by simple dispersion relations of the MTM in collisionless and intermediate regimes. The frequency-integrated level of magnetic fluctuations around 150 kHz (100-200 kHz) is \\tilde{{B}}r /Bt ≈ 1× 10-4 , a level high enough for the ergodization of the magnetic surface and enhanced electron heat loss as derived by Rechester and Rosenbluth (1978 Phys. Rev. Lett. 40 38). This level is consistent with the measurements performed using cross-polarization scattering of microwaves in the Tore Supra tokamak. Our results are the first direct experimental verification of the MTM in the core region of tokamak plasmas, which has been recently observed in gyrokinetic simulations using a very fine mesh in tokamak and ST plasmas.

  14. Transport of dust particles in tokamak devices

    Energy Technology Data Exchange (ETDEWEB)

    Pigarov, A.Yu. [University of California at San Diego, La Jolla, CA (United States)]. E-mail: apigarov@uscd.edu; Smirnov, R.D. [University of California at San Diego, La Jolla, CA (United States); Krasheninnikov, S.I. [University of California at San Diego, La Jolla, CA (United States); Rognlien, T.D. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Rosenberg, M. [University of California at San Diego, La Jolla, CA (United States); Soboleva, T.K. [UNAM, Mexico, DistritoFederal (Mexico)

    2007-06-15

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

  15. Abel inversion of asymmetric plasma density profile at Aditya tokamak

    Science.gov (United States)

    Joshi, N. Y.; Atrey, P. K.; Pathak, S. K.

    2010-02-01

    In Aditya tokamak, at Institute for Plasma Research, till now, multi-channel microwave interferometer system is used to measure the cord averaged plasma density at predefined radial position. An inversion code is developed to determine the local density profile from the chord average density measurement of radially asymmetric plasma. The radial density profile is interpolated using Spline interpolation analytical technique for symmetric plasma density profile. Code implements the Slice and Stack method to determine localized density from asymmetric averaged plasma density measurement from interferometer. Inverted results are tested with various monotonically varying asymmetric radial density profiles of the plasma shots. It also provides the poloidal picture of plasma density distribution with circular constant density surfaces. Localized density measurements, which is very important for successful operation of tokamak, is in agreement with observation of other diagnostics.

  16. Heavy Neutral Beam Probe for Edge Plasma Analysis in Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Castracane, J.

    2001-01-04

    The Heavy Neutral Beam Probe (HNBP) developed initially with DOE funding under the Small Business Innovation Research (SBIR) program was installed on the Tokamak de Varennes (TdeV) at the CCFM. This diagnostic was designed to perform fundamental measurements of edge plasma properties. The hardware was capable of measuring electron density and potential profiles with high spatial and temporal resolution. Fluctuation spectra for these parameters were obtained with HNBP for transport studies.

  17. Development of frequency modulation reflectometer for Korea Superconducting Tokamak Advanced Research tokamak.

    Science.gov (United States)

    Seo, Seong-Heon; Park, Jinhyung; Wi, H M; Lee, W R; Kim, H S; Lee, T G; Kim, Y S; Kang, Jin-Seob; Bog, M G; Yokota, Y; Mase, A

    2013-08-01

    Frequency modulation reflectometer has been developed to measure the plasma density profile of the Korea Superconducting Tokamak Advanced Research tokamak. Three reflectometers are operating in extraordinary polarization mode in the frequency range of Q band (33.6-54 GHz), V band (48-72 GHz), and W band (72-108 GHz) to measure the density up to 7 × 10(19) m(-3) when the toroidal magnetic field is 2 T on axis. The antenna is installed inside of the vacuum vessel. A new vacuum window is developed by using 50 μm thick mica film and 0.1 mm thick gold gasket. The filter bank of low pass filter, notch filter, and Faraday isolator is used to reject the electron cyclotron heating high power at attenuation of 60 dB. The full frequency band is swept in 20 μs. The mixer output is directly digitized with sampling rate of 100 MSamples/s. The phase is obtained by using wavelet transform. The whole hardware and software system is described in detail and the measured density profile is presented as a result.

  18. 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

  19. The disruptive instability in Tokamak plasmas

    NARCIS (Netherlands)

    Salzedas, F.J.B.

    2000-01-01

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

  20. Safety factor profile control in a tokamak

    CERN Document Server

    Bribiesca Argomedo, Federico; Prieur, Christophe

    2014-01-01

    Control of the Safety Factor Profile in a Tokamak uses Lyapunov techniques to address a challenging problem for which even the simplest physically relevant models are represented by nonlinear, time-dependent, partial differential equations (PDEs). This is because of the  spatiotemporal dynamics of transport phenomena (magnetic flux, heat, densities, etc.) in the anisotropic plasma medium. Robustness considerations are ubiquitous in the analysis and control design since direct measurements on the magnetic flux are impossible (its estimation relies on virtual sensors) and large uncertainties remain in the coupling between the plasma particles and the radio-frequency waves (distributed inputs). The Brief begins with a presentation of the reference dynamical model and continues by developing a Lyapunov function for the discretized system (in a polytopic linear-parameter-varying formulation). The limitations of this finite-dimensional approach motivate new developments in the infinite-dimensional framework. The t...

  1. Natural current profiles in a tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, J.B.

    1990-08-01

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

  2. Phase Contrast Imaging on the HL-2A Tokamak

    Science.gov (United States)

    Yu, Yi; Gong, Shaobo; Xu, Min; Jiang, Wei; Zhong, Wulv; Shi, Zhongbin; Wang, Huajie; Wu, Yifan; Yuan, Boda; Lan, Tao; Ye, Minyou; Duan, Xuru; HL-2A Team

    2016-10-01

    In this article we present the design of a phase contrast imaging (PCI) system on the HL-2A tokamak. This diagnostic is developed to infer line integrated plasma density fluctuations by measuring the phase shift of an expanded CO2 laser beam passing through magnetically confined high temperature plasmas. This system is designed to diagnose plasma density fluctuations with the maximum wavenumber of 66 cm-1. The designed wavenumber resolution is 2.09cm-1, and the time resolution is higher than 0.2 μs. The broad kρs ranging from 0.34 to 13.37 makes it suitable for turbulence measurement. An upgraded PCI system is also discussed, which is designed for the HL-2M tokamak. Supported by the National Magnetic Confinement Fusion Energy Research Project (Grant No. 2015GB120002), the National Natural Science Foundation of China (Grant No. 11375053, 11105144, 10905057, 11535013).

  3. 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

  4. Deposit of thin films for Tokamaks conditioning; Deposito de peliculas delgadas para acondicionar Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Valencia A, R

    2006-07-01

    spectroscopy and plasma resistivity measurement. Such mechanism depends fundamentally on the mass of the ions that interact with the wall during the plasma current formation phase. The reaction products generated by the glow discharge plasma, created in a calibrated mixture of methane-hydrogen during the hydrogenated amorphous carbon film deposit on the vessel wall of Novillo tokamak, were determined by mass spectrometry. By way of measuring the emission lines of the carbon and oxygen impurities in intense discharges, the time required by the plasma to interact with the wall was estimated. In addition to it, the temporal conduct of the emission line intensity of these impurities was observed by means of an intensified CCD detector. Once an {approx} 10 % of helium was introduced in the operating gas of the tokamak discharges, a 25-42 eV time variation of the electron temperature was measured using the intensity ratio technique. (Author)

  5. Application of avalanche photodiode for soft X-ray pulse-height analyses in the Ht-7 tokamak

    CERN Document Server

    Shi Yue Jiang; Hu Li Qun; Sun Yan Jun; LiuSheng; Ling Bil

    2002-01-01

    An avalanche photodiode (APD) has been used as soft X-ray energy pulse-height analysis system for the measurement of the electron temperature on the HT-7 tokamak. The experimental results obtained with the APD with its inferior energy resolution show a little difference compared to the conventional high energy-resolution Si (Li) detector. Both numerical analysis and experimental results prove that the APD is good enough for application of the electron temperature measurement in tokamaks.

  6. Feedback system for divertor impurity seeding based on real-time measurements of surface heat flux in the Alcator C-Mod tokamak

    Science.gov (United States)

    Brunner, D.; Burke, W.; Kuang, A. Q.; LaBombard, B.; Lipschultz, B.; Wolfe, S.

    2016-02-01

    Mitigation of the intense heat flux to the divertor is one of the outstanding problems in fusion energy. One technique that has shown promise is impurity seeding, i.e., the injection of low-Z gaseous impurities (typically N2 or Ne) to radiate and dissipate the power before it arrives to the divertor target plate. To this end, the Alcator C-Mod team has created a first-of-its-kind feedback system to control the injection of seed gas based on real-time surface heat flux measurements. Surface thermocouples provide real-time measurements of the surface temperature response to the plasma heat flux. The surface temperature measurements are inputted into an analog computer that "solves" the 1-D heat transport equation to deliver accurate, real-time signals of the surface heat flux. The surface heat flux signals are sent to the C-Mod digital plasma control system, which uses a proportional-integral-derivative (PID) algorithm to control the duty cycle demand to a pulse width modulated piezo valve, which in turn controls the injection of gas into the private flux region of the C-Mod divertor. This paper presents the design and implementation of this new feedback system as well as initial results using it to control divertor heat flux.

  7. (Injection of compact toroids for tokamak fueling and current drive)

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, D.Q.; Rogers, J.H.; Thomas, J.C.; Evans, R.; Foley, R.; Hillyer, T.

    1991-01-01

    The experimental goals for the 1990--1991 period were the operation of the Davis Diverted Tokamak(DDT), the beat wave experiment, and the construction of the compact toroid injection experiment(CTIX). The experiment results from these areas are summarized in the posters given in the APS meeting past November. Here we shall describe the technical progress of the development of the diagnostic system for beat wave experiment, and CT injection especially in relation to the up coming injection experiments into DDT tokamak. The tokamak operation of DDT over the past year has been focused in two parameter ranges. The long pulse discharges (over 100 msec), and the low q short pulse discharges (about 10 msec). We found that the long pulse discharges required a position feedback more sophisticated than the simple passive program that we have. We are in the process of assembling this system. We also found an interesting low q(a) operating regime. Here an equilibrium can be established for a toroidal field between .5 and 1 kG. The typical plasma current is > 5kA. The density of the plasma is between 10{sup 12} and 10{sup 13} cm{sup {minus}3}. The plasma condition in these discharge are sufficiently mild that diagnostic probes can be used to measure various plasma fluctuations. We believe that this will be the regime best suited to study the interaction between the tokamak plasma and the compact toroid. A sophisticated probe system of both electrostatic and electromagnetic types similar to those used in the beat wave experiment has been designed for the up coming experiments.

  8. Robust Sliding Mode Control for Tokamaks

    Directory of Open Access Journals (Sweden)

    I. Garrido

    2012-01-01

    Full Text Available Nuclear fusion has arisen as an alternative energy to avoid carbon dioxide emissions, being the tokamak a promising nuclear fusion reactor that uses a magnetic field to confine plasma in the shape of a torus. However, different kinds of magnetohydrodynamic instabilities may affect tokamak plasma equilibrium, causing severe reduction of particle confinement and leading to plasma disruptions. In this sense, numerous efforts and resources have been devoted to seeking solutions for the different plasma control problems so as to avoid energy confinement time decrements in these devices. In particular, since the growth rate of the vertical instability increases with the internal inductance, lowering the internal inductance is a fundamental issue to address for the elongated plasmas employed within the advanced tokamaks currently under development. In this sense, this paper introduces a lumped parameter numerical model of the tokamak in order to design a novel robust sliding mode controller for the internal inductance using the transformer primary coil as actuator.

  9. The ETE spherical Tokamak project. IAEA report

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, Gerson Otto; Del Bosco, E.; Berni, L.A.; Ferreira, J.G.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Barroso, J.J.; Castro, P.J.; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma]. E-mail: ludwig@plasma.inpe.br

    2002-07-01

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

  10. Benchmarking Tokamak edge modelling codes

    Science.gov (United States)

    Contributors To The Efda-Jet Work Programme; Coster, D. P.; Bonnin, X.; Corrigan, G.; Kirnev, G. S.; Matthews, G.; Spence, J.; Contributors to the EFDA-JET work programme

    2005-03-01

    Tokamak edge modelling codes are in widespread use to interpret and understand existing experiments, and to make predictions for future machines. Little direct benchmarking has been done between the codes, and the users of the codes have tended to concentrate on different experimental machines. An important validation step is to compare the codes for identical scenarios. In this paper, two of the major edge codes, SOLPS (B2.5-Eirene) and EDGE2D-NIMBUS are benchmarked against each other. A set of boundary conditions, transport coefficients, etc. for a JET plasma were chosen, and the two codes were run on the same grid. Initially, large differences were seen in the resulting plasmas. These differences were traced to differing physics assumptions with respect to the parallel heat flux limits. Once these were switched off in SOLPS, or implemented and switched on in EDGE2D-NIMBUS, the remaining differences were small.

  11. Economic considerations of commercial tokamak options

    Energy Technology Data Exchange (ETDEWEB)

    Dabiri, A.E.

    1986-05-01

    Systems studies have been performed to assess commercial tokamak options. Superconducting, as well as normal, magnet coils in either first or second stability regimes have been considered. A spherical torus (ST), as well as an elongated tokamak (ET), is included in the study. The cost of electricity (COE) is selected as the figure of merit, and beta and first-wall neutron wall loads are selected to represent the physics and technology characteristics of various options. The results indicate that an economical optimum for tokamaks is predicted to require a beta of around 10%, as predicted to be achieved in the second stability regime, and a wall load of about 5 MW/m/sup 2/, which is assumed to be optimum technologically. This tokamak is expected to be competitive with fission plants if efficient, noninductive current drive is developed. However, if this regime cannot be attained, all other tokamaks operating in the first stability regime, including spherical torus and elongated tokamak and assuming a limiting wall load of 5 MW/m/sup 2/, will compete with one another with a COE of about 50 mill/kWh. This 40% higher than the COE for the optimum reactor in the second stability regime with fast-wave current drive. The above conclusions pertain to a 1200-MW(e) net electric power plant. A comparison was also made between ST, ET, and superconducting magnets in the second stability regime with fast-wave current drive at 600 MW(e).

  12. Optimization of magnetic field system for glass spherical tokamak GLAST-III

    Science.gov (United States)

    Ahmad, Zahoor; Ahmad, S.; Naveed, M. A.; Deeba, F.; Aqib Javeed, M.; Batool, S.; Hussain, S.; Vorobyov, G. M.

    2017-04-01

    GLAST-III (Glass Spherical Tokamak) is a spherical tokamak with aspect ratio A = 2. The mapping of its magnetic system is performed to optimize the GLAST-III tokamak for plasma initiation using a Hall probe. Magnetic field from toroidal coils shows 1/R dependence which is typical with spherical tokamaks. Toroidal field (TF) coils can produce 875 Gauss field, an essential requirement for electron cyclotron resonance assisted discharge. The central solenoid (CS) of GLAST-III is an air core solenoid and requires compensation coils to reduce unnecessary magnetic flux inside the vessel region. The vertical component of magnetic field from the CS in the vacuum vessel region is reduced to 1.15 Gauss kA-1 with the help of a differential loop. The CS of GLAST can produce flux change up to 68 mVs. Theoretical and experimental results are compared for the current waveform of TF coils using a combination of fast and slow capacitor banks. Also the magnetic field produced by poloidal field (PF) coils is compared with theoretically predicted values. It is found that calculated results are in good agreement with experimental measurement. Consequently magnetic field measurements are validated. A tokamak discharge with 2 kA plasma current and pulse length 1 ms is successfully produced using different sets of coils.

  13. Electron density and temperature determination in a Tokamak plasma using light scattering; Determinacion de la densidad y temperatura electronicas en un Tokamak mediante difusion luminosa

    Energy Technology Data Exchange (ETDEWEB)

    Perez-Navarro Gomerz, A.; Zurro Hernandez, B.

    1976-07-01

    A theoretical foundation review for light scattering by plasmas is presented. Furthermore, we have included a review of the experimental methods for electron density and temperature measurements, with spatial and time resolution, in a Tokamak plasma using spectral analysis of the scattered radiation. (Author) 13 refs.

  14. MHD stability limits in the TCV Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Reimerdes, H. [Ecole Polytechnique Federale de Lausanne, Centre de Recherches en Physique des Plasmas (CRPP), CH-1015 Lausanne (Switzerland)

    2001-07-01

    observed decrease of this limit with elongation is also in qualitative agreement with ideal MHD theory. Edge localised modes (ELMs), occurring in TCV Ohmic high-confinement mode discharges, were observed to be preceded by coherent magnetic oscillations. The detected poloidal and toroidal mode structures are consistent with a resonant flux surface close to the plasma edge. Unlike conventional MHD modes, these precursors start at a random toroidal location and then grow in amplitude and toroidal extent until they encompass the whole toroidal circumference. Thus, the asymmetry causing and maintaining the toroidal localisation of the ELM precursor must be intrinsic to the plasma. Soft X-ray measurements show that the localised precursor always coincides with a central m = 1 mode, which can usually be associated with the sawtooth pre- or postcursor mode. A comparison of the phases indicates a correlation with the maximum of the central mode preceding the toroidal location of the ELM precursor and, therefore, a hitherto unobserved coupling between central modes and ELMs. Highly elongated plasmas promise several advantages, among them higher current and beta limits. During TCV experiments dedicated to an increasing of the plasma elongation, a new disruptive current limit, at values well below the conventional current limit corresponding to q{sub a} > 2, was encountered for {kappa} > 2.3. This limit, which is preceded by a kink-type mode, is found to be consistent with ideal MHD stability calculations. The TCV observations, therefore, provide the first experimental confirmation of a deviation of the linear Troyon-scaling of the ideal beta limit with normalised current at high elongation, which was predicted over 10 years ago. Neoclassical tearing modes (NTMs), which have been observed to limit the achievable beta in a number of tokamaks, arise from a helical perturbation of the bootstrap current caused by an existing seed island. Neoclassical m/n = 2/1 tearing modes have been

  15. Remote operation of the vertical plasma stabilization @ the GOLEM tokamak for the plasma physics education

    Energy Technology Data Exchange (ETDEWEB)

    Svoboda, V., E-mail: svoboda@fjfi.cvut.cz [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Kocman, J.; Grover, O. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Krbec, J.; Stöckel, J. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Institute of Plasma Physics AS CR, CZ-182 21 Prague (Czech Republic)

    2015-10-15

    Graphical abstract: * Understandable remote operation of a vertical plasma position control system in the tokamak GOLEM for educational purposes.* Two combinable modes of real-time plasma position control: position based feedback and a pre-defined waveform.* More than 20% plasma life prolongation with plasma position control in feedback mode. - Highlights: • Understandable remote operation of a vertical plasma position control system in the tokamak GOLEM for educational purposes. • Two combinable modes of real-time plasma position control: position based feedback and a pre-defined waveform. • More than 20% plasma life prolongation with plasma position control in feedback mode. - Abstract: The GOLEM tokamak at the Czech Technical University has been established as an educational tokamak device for domestic and foreign students. Remote participation in the scope of several laboratory practices, plasma physics schools and workshops has been successfully performed from abroad. A new enhancement allowing understandable remote control of vertical plasma position in two modes (i) predefined and (ii) feedback control is presented. It allows to drive the current in the stabilization coils in any time-dependent scenario, which can include as a parameter the actual plasma position measured by magnetic diagnostics. Arbitrary movement of the plasma column in a vertical direction, stabilization of the plasma column in the center of the tokamak vessel as well as prolongation/shortening of plasma life according to the remotely defined request are demonstrated.

  16. Exploration of turbulent optimization in stellarators & tokamaks

    Science.gov (United States)

    Mynick, H.; Pomphrey, N.; Xanthopoulos, P.; Lucia, M.

    2012-03-01

    A methodfootnotetextH.E. Mynick, N. Pomphrey, P. Xanthopoulos, Phys. Rev. Letters, 105, 095004 (2010).^,footnotetextH.E. Mynick, N. Pomphrey, P. Xanthopoulos, Phys. Plasmas, 18, 056101 (2011). recently developed for evolving toroidal configurations to ones with reduced turbulent transport, using the STELLOPT optimization codes and the GENE gyrokinetic code, is being applied and extended. The growing body of results has found that the effectiveness of the current proxy measure Qprox used by STELLOPT to estimate transport levels depends on the class of toroidal device considered. The present proxy works well for quasi-axisymmetric stellarators and tokamaks, modestly for quasi-helically symmetric designs, but not for the W7X quasi-omnigenous/quasi-isodynamic design. We are exploring the origin of this variation, and improving the dependence of the proxy on key geometric factors, extending the proxy to apply to transport channels other than the ITG turbulence it was originally developed for, and are also examining the relative effectiveness of different search algorithms. To help in these efforts, we have adapted STELLOPT to provide a new capability for mapping the topography of the cost function in the search space.

  17. Detachment evolution on the TCV tokamak

    Directory of Open Access Journals (Sweden)

    J.R. Harrison

    2017-08-01

    Full Text Available Divertor detachment in the TCV tokamak has been investigated through experiments and modelling. Density ramp experiments were carried out in ohmic heated L-mode pulses with the ion ∇B drift directed away from the primary X-point, similar to previous studies [1]. Before the roll-over in the ion current to the outer strike point, C III and Dα emission from the outer leg recede slowly from the strike point toward the X-point, at a rate of ∼2.0 × 10−19m/m−3 along the magnetic field as the electron temperature along the leg reduces with increasing density. Around the onset of detachment, the upstream density profile and outer target Dα profiles broaden, possibly leading to an increase in radiation in the SOL by increased interaction between the SOL and the carbon tiles lining the outer wall. The plasma conditions upstream and at various locations along the detached outer divertor leg have been characterised, and the consistency of this data has been checked with the interpretive OSM-EIRENE-DIVIMP suite of codes [2] and are broadly found to be consistent with measured Dγ/Dα emissivity profiles along the detached outer divertor leg.

  18. Multi-channel H_α Diagnostics for Position Determination of a Tokamak Plasma

    Science.gov (United States)

    Kim, Jayhyun; Yi, H. S.; Kwon, G. C.; Kim, J. S.; Choe, W.

    1999-11-01

    A multi-channel H_α spectroscopic diagnostic system was developed on KAIST-Tokamak to be utilized for diagnosing the plasma position in the early phase of ohmic discharges. Since the measured intensity is line-integrated along the line of sight, an Abel inversion computer program was developed. Using the inversion program, the vertical (similar to minor-radial) H_α intensity profile was obtained at several different time steps in the start-up phase of KAIST-Tokamak ohmic discharges. The center position of the intensity is an indirect indication of the plasma center. Comparison with the magnetics data shows reasonable agreement. This suggests that the multi-channel H_α diagnostic may be a good candidate to determine the plasma position which can be useful especially for plasma position control in the tokamak start-up phase.

  19. Electron thermal transport in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Konings, J.A.

    1994-11-30

    The process of fusion of small nuclei thereby releasing energy, as it occurs continuously in the sun, is essential for the existence of mankind. The same process applied in a controlled way on earth would provide a clean and an abundant energy source, and be the long term solution of the energy problem. Nuclear fusion requires an extremely hot (10{sup 8} K) ionized gas, a plasma, that can only be maintained if it is kept insulated from any material wall. In the so called `tokamak` this is achieved by using magnetic fields. The termal insulation, which is essential if one wants to keep the plasma at the high `fusion` temperature, can be predicted using basic plasma therory. A comparison with experiments in tokamaks, however, showed that the electron enery losses are ten to hundred times larger than this theory predicts. This `anomalous transport` of thermal energy implies that, to reach the condition for nuclear fusion, a fusion reactor must have very large dimensions. This may put the economic feasibility of fusion power in jeopardy. Therefore, in a worldwide collaboration, physicists study tokamak plasmas in an attempt to understand and control the energy losses. From a scientific point of view, the mechanisms driving anomalous transport are one of the challenges in fudamental plasma physics. In Nieuwegein, a tokamak experiment (the Rijnhuizen Tokamak Project, RTP) is dedicated to the study of anomalous transport, in an international collaboration with other laboratories. (orig./WL).

  20. Up-down asymmetric tokamaks

    CERN Document Server

    Ball, Justin

    2016-01-01

    Bulk toroidal rotation has proven capable of stabilising both dangerous MHD modes and turbulence. In this thesis, we explore a method to drive rotation in large tokamaks: up-down asymmetry in the magnetic equilibrium. We seek to maximise this rotation by finding optimal up-down asymmetric flux surface shapes. First, we use the ideal MHD model to show that low order external shaping (e.g. elongation) is best for creating up-down asymmetric flux surfaces throughout the device. Then, we calculate realistic up-down asymmetric equilibria for input into nonlinear gyrokinetic turbulence analysis. Analytic gyrokinetics shows that, in the limit of fast shaping effects, a poloidal tilt of the flux surface shaping has little effect on turbulent transport. Since up-down symmetric surfaces do not transport momentum, this invariance to tilt implies that devices with mirror symmetry about any line in the poloidal plane will drive minimal rotation. Accordingly, further analytic investigation suggests that non-mirror symmetri...

  1. Liquid nitrogen cooling considerations of the Compact Ignition Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Dabiri, A.E.

    1986-10-01

    A simple model was developed to estimate the cooldown time between pulses of toroidal field (TF) coils of the Compact Ignition Tokamak (CIT) using liquid nitrogen. Good agreement was obtained between the analysis results and those measured in the early fusion experimental devices. A cooldown time of about 1 h would reduce the TF coil temperature to about 80 K. An R and D experimental program is required to determine the actual cooldown time between pulses, an issue in the conceptual design of the CIT.

  2. Tokamak power systems studies, FY 1985

    Energy Technology Data Exchange (ETDEWEB)

    Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Smith, D.L.; Sze, D.K.

    1985-12-01

    The Tokamak Power System Studies (TPSS) at ANL in FY-1985 were devoted to exploring innovative design concepts which have the potential for making substantial improvements in the tokamak as a commercial power reactor. Major objectives of this work included improved reactor economics, improved environmental and safety features, and the exploration of a wide range of reactor plant outputs with emphasis on reduced plant sizes compared to STARFIRE. The activities concentrated on three areas: plasma engineering, impurity control, and blanket/first wall/shield technology. 205 refs., 125 figs., 107 tabs.

  3. Overview of spherical tokamak research in Japan

    Science.gov (United States)

    Takase, Y.; Ejiri, A.; Fujita, T.; Fukumoto, N.; Fukuyama, A.; Hanada, K.; Idei, H.; Nagata, M.; Ono, Y.; Tanaka, H.; Uchida, M.; Horiuchi, R.; Kamada, Y.; Kasahara, H.; Masuzaki, S.; Nagayama, Y.; Oishi, T.; Saito, K.; Takeiri, Y.; Tsuji-Iio, S.

    2017-10-01

    Nationally coordinated research on spherical tokamak is being conducted in Japan. Recent achievements include: (i) plasma current start-up and ramp-up without the use of the central solenoid by RF waves (in electron cyclotron and lower hybrid frequency ranges), (ii) plasma current start-up by AC Ohmic operation and by coaxial helicity injection, (iii) development of an advanced fuelling technique by compact toroid injection, (iv) ultra-long-pulse operation and particle control using a high temperature metal wall, (v) access to the ultra-high-β regime by high-power reconnection heating, and (vi) improvement of spherical tokamak plasma stability by externally applied helical field.

  4. A Fast Shutdown Technique for Large Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    E. Fredrickson; G.L. Schmidt; K. Hill; S.C. Jardin; et al

    1999-09-01

    A practical method is proposed for the fast shutdown of a large ignited tokamak. The method consists of injecting a rapid series of 30-50 deuterium pellets doped with a small ( 0.0005%) concentration of Krypton impurity, and simultaneously ramping the plasma current and shaping fields down over a period of several seconds using the poloidal field system. Detailed modeling with the Tokamak Simulation Code using a newly developed pellet mass deposition model shows that this method should terminate the discharge in a controlled and stable way without producing significant numbers of runaway electrons. A partial prototyping of this technique was accomplished in TFTR.

  5. Boundary perturbations coupled to core 3/2 tearing modes on the DIII-D tokamak

    NARCIS (Netherlands)

    Tobias, B.; Yu, L.; Domier, C.W.; N C Luhmann Jr.,; Austin, M. E.; Paz-Soldan, C.; Turnbull, A. D.; Classen, I.G.J.

    2013-01-01

    High confinement (H-mode) discharges on the DIII-D tokamak are routinely subject to the formation of long-lived, non-disruptive magnetic islands that degrade confinement and limit fusion performance. Simultaneous, 2D measurement of electron temperature fluctuations in the core and edge regions

  6. 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...

  7. A charged fusion product diagnostic for a spherical tokamak

    Science.gov (United States)

    Perez, Ramona Leticia Valenzuela

    Designs for future nuclear fusion power reactors rely on the ability to create a stable plasma (hot ionized gas of hydrogen isotopes) as a medium with which to sustain nuclear fusion reactions. My dissertation work involves designing, constructing, testing, installing, operating, and validating a new diagnostic for spherical tokamaks, a type of reactor test facility. Through detecting charged particles emitted from the plasma, this instrument can be used to study fusion reaction rates within the plasma and how they are affected by plasma perturbations. Quantitatively assessing nuclear fusion reaction rates at specific locations inside the plasma and as a function of time can provide valuable data that can be used to evaluate theory-based simulations related to energy transport and plasma stability. The Proton Detector (PD), installed in the Mega Amp Spherical Tokamak (MAST) at the Culham Centre for Fusion Energy (CCFE) in Abingdon, England, was the first instrument to experimentally detect 3 MeV Protons and 1 MeV Tritons created from deuterium- deuterium (hydrogen isotopes) nuclear fusion reactions inside a spherical tokamak's plasma. The PD consists of an array of particle detectors with a protective housing and the necessary signal conditioning electronics and readout. After several years of designing (which included simulations for detector orientations), fabricating, and testing the PD, it was installed in MAST and data were collected over a period of two months in the summer of 2013. Proton and triton rates as high as 200 kHz were measured and an initial radial profile of these fusion reaction rates inside the plasma was extracted. These results will be compared to a complementary instrument at MAST as well as theory-based simulations and form the knowledge basis for developing a larger future instrument. The design and performance of all instrument components (electrical, computational, mechanical), and subsequent data analysis methods and results are

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

    DEFF Research Database (Denmark)

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

    Moving towards the era of burning fusion plasmas, a better knowledge of the physics of highly energetic particles, such as fusion born alpha particles, becomes essential. Diagnosing the fast ions in a fusion plasma is a challenging task, but the technique of collective Thomson scattering (CTS...... perpendicular to the magnetic field. The feasibility study and conceptual design of this diagnostic was provided by the CTS group at Risø DTU. The development of the ITER CTS diagnostic builds on the experiences and expertise gained from the construction and current operation of the CTS diagnostic systems...

  9. UCLA Tokamak Program Close Out Report.

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, Robert John [UCLA/retired

    2014-02-04

    The results of UCLA experimental fusion program are summarized. Starting with smaller devices like Microtor, Macrotor, CCT and ending the research on the large (5 m) Electric Tokamak. CCT was the most diagnosed device for H-mode like physics and the effects of rotation induced radial fields. ICRF heating was also studied but plasma heating of University Type Tokamaks did not produce useful results due to plasma edge disturbances of the antennae. The Electric Tokamak produced better confinement in the seconds range. However, it presented very good particle confinement due to an "electric particle pinch". This effect prevented us from reaching a quasi steady state. This particle accumulation effect was numerically explained by Shaing's enhanced neoclassical theory. The PI believes that ITER will have a good energy confinement time but deleteriously large particle confinement time and it will disrupt on particle pinching at nominal average densities. The US fusion research program did not study particle transport effects due to its undue focus on the physics of energy confinement time. Energy confinement time is not an issue for energy producing tokamaks. Controlling the ash flow will be very expensive.

  10. Spontaneous generation of rotation in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Parra Diaz, Felix [Oxford University

    2013-12-24

    Three different aspects of intrinsic rotation have been treated. i) A new, first principles model for intrinsic rotation [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has been implemented in the gyrokinetic code GS2. The results obtained with the code are consistent with several experimental observations, namely the rotation peaking observed after an L-H transition, the rotation reversal observed in Ohmic plasmas, and the change in rotation that follows Lower Hybrid wave injection. ii) The model in [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has several simplifying assumptions that seem to be satisfied in most tokamaks. To check the importance of these hypotheses, first principles equations that do not rely on these simplifying assumptions have been derived, and a version of these new equations has been implemented in GS2 as well. iii) A tokamak cross-section that drives large intrinsic rotation has been proposed for future large tokamaks. In large tokamaks, intrinsic rotation is expected to be very small unless some up-down asymmetry is introduced. The research conducted under this contract indicates that tilted ellipticity is the most efficient way to drive intrinsic rotation.

  11. Detachment evolution on the TCV tokamak

    NARCIS (Netherlands)

    Harrison, J. R.; Vijvers, W. A. J.; Theiler, C.; Duval, B. P.; Elmore, S.; Labit, B.; Lipschultz, B.; van Limpt, S. H. M.; Lisgo, S. W.; Tsui, C. K.; Reimerdes, H.; Sheikh, U.; Verhaegh, K. H. A.; Wischmeier, M.

    2017-01-01

    Divertor detachment in the TCV tokamak has been investigated through experiments and modelling. Density ramp experiments were carried out in ohmic heated L-mode pulses with the ion ∇B drift directed away from the primary X-point, similar to previous studies [1]. Before the roll-over in the ion

  12. Degraded confinement and turbulence in tokamak experiments

    NARCIS (Netherlands)

    Schüller, F. C.

    1996-01-01

    After a review on the state of tokamak transport theory, the methodology to derive experimental results will be described. Examples of confinement in ohmic plasmas and the deterioration with additional heating will be given. Some examples of improved confinement modes will be discussed. Fluctuation

  13. Degraded confinement and turbulence in tokamak experiments

    NARCIS (Netherlands)

    Hogeweij, G. M. D.

    2012-01-01

    After a review on the state of tokamak transport theory, the methodology to derive experimental results will be described. Examples of confinement in ohmic plasmas and the deterioration with additional healing will be given. Some examples of improved confinement; modes will be discussed.

  14. Tokamak Transport Studies Using Perturbation Analysis

    NARCIS (Netherlands)

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

    1990-01-01

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

  15. On dust in tokamak edge plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Krasheninnikov, S.I. [Jacobs School of Engineering, Department of Mechanical and Aerospace Engineering, University of California at San Diego, Engineering Building II, room 474, 9500 Gilman Drive, La Jolla, CA 92093-0411 (United States)]. E-mail: skrash@mae.ucsd.edu; Soboleva, T.K. [UNAM, Mexico, DF (Mexico); Kurchatov Institute, Moscow (Russian Federation); Tomita, Y. [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Smirnov, R.D. [Graduate University for Advanced Studies, Toki, Gifu 509-5292 (Japan); Janev, R.K. [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan)

    2005-03-01

    We study the dust particle dynamics in tokamak edge plasmas, with special emphasis on dust particle transport in the sheath and plasma recycling regions. The characteristics of this transport have been examined for both smooth and corrugated wall surfaces. The implications of dust particle transport in the divertor region on the core plasma contamination with impurities have also been examined.

  16. Tokamak startup with electron cyclotron heating

    Energy Technology Data Exchange (ETDEWEB)

    Holly, D J; Prager, S C; Shepard, D A; Sprott, J C

    1980-04-01

    Experiments are described in which the startup voltage in a tokamak is reduced by approx. 60% by the use of a modest amount of electron cyclotron resonance heating power for preionization. A 50% reduction in volt-second requirement and impurity reflux are also observed.

  17. Lithium beam diagnostic system on the COMPASS tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Anda, G.; Bencze, A. [Wigner – RCP, HAS, Budapest (Hungary); Berta, M., E-mail: bertam@sze.hu [Institute of Plasma Physics AS CR, Prague (Czech Republic); Széchenyi István University, Győr (Hungary); Dunai, D. [Wigner – RCP, HAS, Budapest (Hungary); Hacek, P. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Faculty of Mathematics and Physics, Charles University in Prague, Prague (Czech Republic); Krbec, J. [Institute of Plasma Physics AS CR, Prague (Czech Republic); Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, Prague (Czech Republic); Réfy, D.; Krizsanóczi, T.; Bató, S.; Ilkei, T.; Kiss, I.G.; Veres, G.; Zoletnik, S. [Wigner – RCP, HAS, Budapest (Hungary)

    2016-10-15

    Highlights: • Li-beam diagnostic system on the COMPASS tokamak is an improved and compact system to allow testing of Atomic Beam Probe. • The possibility to measure background corrected density profiles on the few microseconds time scale. • First Li-beam diagnostic system with recirculating neutralizer. • The system includes the redesigned ion source with longer lifetime. - Abstract: An improved lithium beam based beam emission spectroscopy system – installed on COMPASS tokamak – is described. The beam energy enhanced up to 120 keV for Atomic Beam Probe measurement. The size of the ion source is doubled, using a newly developed thermionic heater instead of the conventionally used heating (tungsten or molybdenum) filament. The neutralizer is also improved. It produces the same sodium vapor in a cell but minimize the loss condensing the vapor on a cold surface which is led back (in fluid state) into the sodium oven. This way we call it recirculating neutralizer. The observation system consists of a CCD camera and an avalanche photodiode array.

  18. Magnetic flux reconstruction methods for shaped tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Tsui, Chi-Wa [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    1993-12-01

    The use of a variational method permits the Grad-Shafranov (GS) equation to be solved by reducing the problem of solving the 2D non-linear partial differential equation to the problem of minimizing a function of several variables. This high speed algorithm approximately solves the GS equation given a parameterization of the plasma boundary and the current profile (p` and FF` functions). The author treats the current profile parameters as unknowns. The goal is to reconstruct the internal magnetic flux surfaces of a tokamak plasma and the toroidal current density profile from the external magnetic measurements. This is a classic problem of inverse equilibrium determination. The current profile parameters can be evaluated by several different matching procedures. Matching of magnetic flux and field at the probe locations using the Biot-Savart law and magnetic Green`s function provides a robust method of magnetic reconstruction. The matching of poloidal magnetic field on the plasma surface provides a unique method of identifying the plasma current profile. However, the power of this method is greatly compromised by the experimental errors of the magnetic signals. The Casing Principle provides a very fast way to evaluate the plasma contribution to the magnetic signals. It has the potential of being a fast matching method. The performance of this method is hindered by the accuracy of the poloidal magnetic field computed from the equilibrium solver. A flux reconstruction package has been implemented which integrates a vacuum field solver using a filament model for the plasma, a multi-layer perception neural network as an interface, and the volume integration of plasma current density using Green`s functions as a matching method for the current profile parameters. The flux reconstruction package is applied to compare with the ASEQ and EFIT data. The results are promising.

  19. Plasma density determination by microwave interferometry .- The 2 mm interferometer of the TJ-1 Tokamak; Determinacion de la densidad de un plasma por interferometria de microondas. El interferometro de 2 mm del Tokamak TJ-1

    Energy Technology Data Exchange (ETDEWEB)

    Martin, R.; Manero, F.

    1984-07-01

    In this paper a description is given of the microwave interferometer used for measuring the plasma electronic density in the TJ-1 Tokamak of Fusion Division of JEN. The principles of the electronic density measurement are discussed in detail, as well as those concerning the determination of density pro files from experimental data. A description of the interferometer used in the TJ-1 Tokamak is given, together with a detailed analysis of the circuits which constitute the measuring chain. The working principles of the klystron reflex and hybrid rings are also presented. (Author) 23 refs.

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

  1. Atomic physics studies of highly charged ions on tokamaks using x-ray spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Beiersdorfer, P.; von Goeler, S.; Bitter, M.; Hill, K.W.

    1989-07-01

    An overview is given of atomic physics issues which have been studied on tokamaks with the help resolution x-ray spectroscopy. The issues include the testing of model calculations predicting the excitation of line radiation, the determination of rate coefficients, and accurate atomic structure measurements. Recent research has focussed primarily on highly charged heliumlike (22 less than or equal to Z less than or equal to 28) and neonlike (34 less than or equal to Z less than or equal to 63) ions, and results are presented from measurements on the PLT and TFTR tokamaks. Many of the measurements have been aided by improved instrumental design and new measuring techniques. Remarkable agreement has been found between measurements and theory in most cases. However, in this review those areas are stressed where agreement is worst and where further investigations are needed. 19 refs., 13 figs., 2 tabs.

  2. Advanced Tokamak Scenarios for the FIRE Burning Plasma Experiment

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Kessel; D. Ignat; T.K. Mau

    2001-10-12

    The advanced tokamak (AT) capability of the Fusion Ignition Research Experiment (FIRE) burning plasma experiment is examined with 0-D systems analysis, equilibrium and ideal-MHD stability, radio-frequency current-drive analysis, and full discharge dynamic simulations. These analyses have identified the required parameters for attractive burning advanced tokamak plasmas, and indicate that these are feasible with the present progress on existing experimental tokamaks.

  3. Simulations of burn dynamics in tokamak fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mantsinen, M.J.; Salomaa, R.R.E.

    1997-10-01

    The global dynamics of tokamak reactors is investigated with the time-dependent, volume-averaged (0D) particle and power balance code FRESCO (Fusion REactor Simulation COde). The main emphasis is on studies of reactivity transients during tokamak start-up and shut down, as well as after sudden changes in plasma and tokamak parameters. In particular, the plasma responses to changes in the confinement, fuelling rates and impurity concentrations are considered. 76 refs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-02-15

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

  5. Study of energy transport in Tore Supra Tokamak; Etude du transport de l`energie sur le Tokamak Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Guiziou, L.

    1995-12-18

    The goal of this thesis is to characterize the energy confinement and the heat transport in Tore Supra tokamak. The first chapter is an introduction to the different plasma confinement regimes: ohmic, low confinement and improved confinement regimes. The second chapter is devoted to the presentation of the different theoretical and empirical approaches about energy confinement and heat transport. In the third chapter an attempt of explanations for non-local transport phenomenons is given. A turbulence correlation length greater than the ionic Larmor radius seams to be a reasonable explanation. This theoretical study focusses on the possibility for modes coupling in a tokamak. This study tries to determine a radial correlation length considering the two principal coupling modes: toroidal and non-linear. Different transport regimes are discussed using an analytical model and considering the influence of one coupling with respect to the other. In chapter four, the measurements of current profiles and transport coefficients are presented. The codes used for the reconstruction of equilibrium and for the experimental determination of the diffusivity are briefly presented. In chapter five, experimental results of energy transport studies for Tore Supra plasmas are presented. The different modes are analysed in detail and the study focusses on the influence of magnetic shear in the improved confinement regime. Finally, the different parametric dependences of the electronic thermal diffusivity are compared to local transport models. 165 refs., 57 figs., 2 tabs., 2 appendix.

  6. Module description of TOKAMAK equilibrium code MEUDAS

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-01-01

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

  7. Rapidly Moving Divertor Plates In A Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S. Zweben

    2011-05-16

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

  8. MHD stable regime of the tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, C.Z.; Furth, H.P.; Boozer, A.H.

    1986-10-01

    A broad family of tokamak current profiles is found to be stable against ideal and resistive MHD kink modes for 1 less than or equal to q(0), with q(a) as low 2. For 0.5 less than or equal to q(0) < and q(a) > 1, current profiles can be found that are unstable only to the m = 1, n = 1 mode. A specific ''optimal'' tokamak profile can be selected from the range of stable solutions, by imposing a common upper limit on dj/dr - corresponding in ohmic equilibrium to a limitation of dT/sub e//dr by anomalous transport.

  9. Boundary Plasma Turbulence Simulations for Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-05-15

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

  10. Tokamaks: from A D Sakharov to the present (the 60-year history of tokamaks)

    Science.gov (United States)

    Azizov, E. A.

    2012-02-01

    The paper is prepared on the basis of the report presented at the session of the Physical Sciences Division of the Russian Academy of Sciences (RAS) at the Lebedev Physical Institute, RAS on 25 May 2011, devoted to the 90-year jubilee of Academician Andrei D Sakharov - the initiator of controlled nuclear fusion research in the USSR. The 60-year history of plasma research work in toroidal devices with a longitudinal magnetic field suggested by Andrei D Sakharov and Igor E Tamm in 1950 for the confinement of fusion plasma and known at present as tokamaks is described in brief. The recent (2006) agreement among Russia, the EU, the USA, Japan, China, the Republic of Korea, and India on the joint construction of the international thermonuclear experimental reactor (ITER) in France based on the tokamak concept is discussed. Prospects for using the tokamak as a thermonuclear (14 MeV) neutron source are examined.

  11. Microinstabilities in weak density gradient tokamak systems

    Energy Technology Data Exchange (ETDEWEB)

    Tang, W.M.; Rewoldt, G.; Chen, L.

    1986-04-01

    A prominent characteristic of auxiliary-heated tokamak discharges which exhibit improved (''H-mode type'') confinement properties is that their density profiles tend to be much flatter over most of the plasma radius. Depsite this favorable trend, it is emphasized here that, even in the limit of zero density gradient, low-frequency microinstabilities can persist due to the nonzero temperature gradient.

  12. Electron cyclotron emission imaging diagnostic system for Rijnhuizen Tokamak Project

    Energy Technology Data Exchange (ETDEWEB)

    Deng, B.H.; Hsia, R.P.; Domier, C.W.; Burns, S.R.; Hillyer, T.R.; Luhmann, N.C. Jr. [University of California at Davis, 228 Walker Hall, Davis, California 95616 (United States); Oyevaar, T.; Donne, A.J. [FOM-Inst. voor Plasmafysica Rijnhuizen, Association Euratom-FOM (International organizations without location); RTP team

    1999-01-01

    A 16-channel electron cyclotron emission (ECE) imaging diagnostic system has been developed and installed on the Rijnhuizen Tokamak Project for measuring plasma electron cyclotron emission with a temporal resolution of 2 {mu}s. The high spatial resolution of the system is achieved by utilizing a low cost linear mixer/receiver array. Unlike conventional ECE diagnostics, the sample volumes of the ECE imaging system are aligned vertically, and can be shifted across the plasma cross-section by varying the local oscillator frequency, making possible 2D measurements of electron temperature profiles and fluctuations. The poloidal/radial wavenumber spectra and correlation lengths of T{sub e} fluctuations in the plasma core can also be obtained by properly positioning the focal plane of the imaging system. Due to these unique features, ECE imaging is an ideal tool for plasma transport study. Technical details of the system are described, together with preliminary experimental results. {copyright} {ital 1999 American Institute of Physics.}

  13. Initial DEMO tokamak design configuration studies

    Energy Technology Data Exchange (ETDEWEB)

    Bachmann, Christian, E-mail: christian.bachmann@efda.org [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Aiello, G. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Albanese, R.; Ambrosino, R. [ENEA/CREATE, Universita di Napoli Federico II, Naples (Italy); Arbeiter, F. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Aubert, J. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Boccaccini, L.; Carloni, D. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Federici, G. [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Fischer, U. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Kovari, M. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Li Puma, A. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Loving, A. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Maione, I. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Mattei, M. [ENEA/CREATE, Universita di Napoli Federico II, Naples (Italy); Mazzone, G. [ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Meszaros, B. [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Riccardo, V. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); and others

    2015-10-15

    Highlights: • A definition of main DEMO requirements. • A description of the DEMO tokamak design configuration. • A description of issues yet to be solved. - Abstract: To prepare the DEMO conceptual design phase a number of physics and engineering assessments were carried out in recent years in the frame of EFDA concluding in an initial design configuration of a DEMO tokamak. This paper gives an insight into the identified engineering requirements and constraints and describes their impact on the selection of the technologies and design principles of the main tokamak components. The EU DEMO program aims at making best use of the technologies developed for ITER (e.g., magnets, vessel, cryostat, and to some degree also the divertor). However, other systems in particular the breeding blanket require design solutions and advanced technologies that will only partially be tested in ITER. The main differences from ITER include the requirement to breed, to extract, to process and to recycle the tritium needed for plasma operation, the two orders of magnitude larger lifetime neutron fluence, the consequent radiation dose levels, which limit remote maintenance options, and the requirement to use low-activation steel for in-vessel components that also must operate at high temperature for efficient energy conversion.

  14. Electron cyclotron resonance heating in the microwave tokamak experiment

    Energy Technology Data Exchange (ETDEWEB)

    Allen, S.L.; Casper, T.A.; Fenstermacher, M.E. [and others

    1992-09-01

    This paper presents the results from a series of Electron Cyclotron Resonance Heating (ECRH) experiments on the Microwave Tokamak Experiment (MTX). On-axis heating at B{sub T} = 5T (f{sub ce} = 140 GHz) has been performed at electron densities up to cutoff. We have used both a long-pulse gryotron ({approximately}200 kW, {approximately}0.1s) and a pulsed Free Electron Laser (FEL) as microwave sources. Gyrotron experiments with power densities corresponding to 4 MW m{sup {minus}3}. A far infrared (FIR) polarimeter measured peaking of plasma current profiles in some discharges during the ECRH pulse. During high-power single-pulse FEL experiments, single-pass microwave !transmission measurements show nonlinear effects; i.e., higher transmission than predicted by linear theory. A corrugated-wall duct was used in the tokamak port to increase the gradient of the parallel refractive index n{sub parallel} of the incident wave, and increased absorption was observed. Evidence of electron tail heating during FEL pulses was observed on soft x-ray and ECE diagnostics. These results are in agreement with predictions of nonlinear theory; extrapolation of this theory to reactor-like conditions indicates efficient absorption and heating. A Laser Assisted Particle Probe Spectroscopy (LAPPS) diagnostic provided estimates of the vacuum electric field of the FEL which were consistent with the measured power. Multiple pulse operation of the ETA-II accelerator for the FEL has also been demonstrated, indicating the feasibility of high-average power FEL operation.

  15. Magnetohydrodynamic Waves and Instabilities in Rotating Tokamak Plasmas

    NARCIS (Netherlands)

    J.W. Haverkort (Willem)

    2013-01-01

    htmlabstractOne of the most promising ways to achieve controlled nuclear fusion for the commercial production of energy is the tokamak design. In such a device, a hot plasma is confined in a toroidal geometry using magnetic fields. The present generation of tokamaks shows significant plasma

  16. Soft-X-Ray Tomography Diagnostic at the Rtp Tokamak

    NARCIS (Netherlands)

    Da Cruz, D. F.; Donne, A. J. H.

    1994-01-01

    An 80-channel soft x-ray tomography system has been constructed for diagnosing the RTP (Rijnhuizen Tokamak Project) tokamak plasma. Five pinhole cameras, each with arrays of 16 detectors are distributed more or less homogeneously around a poloidal plasma cross section. The cameras are positioned

  17. Tokamak plasma self-organization-synergetics of magnetic trap plasmas

    NARCIS (Netherlands)

    Razumova, K. A.; Andreev, V. F.; Eliseev, L. G.; Kislov, A. Y.; La Haye, R. J.; Lysenko, S. E.; Melnikov, A. V.; Notkin, G. E.; Pavlov, Y. D.; Kantor, M. Y.

    2011-01-01

    Analysis of a wide range of experimental results in plasma magnetic confinement investigations shows that in most cases, plasmas are self-organized. In the tokamak case, it is realized in the self-consistent pressure profile, which permits the tokamak plasma to be macroscopically MHD stable.

  18. Stability-transport modeling of the SINP tokamak discharges

    Indian Academy of Sciences (India)

    2015-11-27

    Nov 27, 2015 ... The code has been applied to follow the evolution of tokamak plasma discharges obtained in the Saha Institute of Nuclear Physics (SINP) tokamak. From these simulations, we have been able to identify the possible models of thermal conductivity, diffusion and impurity contents in these discharges. Effects ...

  19. Equilibrium and stability of tokamak plasmas and accretion disks

    NARCIS (Netherlands)

    Blokland, J.W.S.

    2007-01-01

    In both fusion research as well in astrophysics, plasmas are widely studied. These plasmas can be found in different geometric configurations, such as in a tokamak, stellarator or in astrophysical jets, accretion disks, etc. In this thesis we focus on plasmas found in tokamaks or accretion disks. In

  20. Poloidal magnetics of a divertor compact ignition tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Strickler, D.J.; Peng, Y.K.M.; Jardin, S.C.

    1987-10-01

    A technique is presented for calculating bounds on the poloidal field (PF) coil currents required to constrain critical plasma shape parameters when plasma pressure and current density profiles are changed. Such considerations are important in the conceptual design of the PF coils for the Compact Ignition Tokamak (CIT) and their electrical power systems in view of the uncertainty in plasma profiles and operating scenarios. Four relatively independent coil groups are sufficient to find a coil current distribution and equilibrium satisfying a prescribed plasma major radius, minor radius, and divertor strike point coordinates. The variation in the coil current distribution with plasma profiles tends to be large for external PF systems and provides a measure by which coil configurations may be compared. 6 refs., 7 figs., 4 tabs.

  1. Predicting high harmonic ion cyclotron heating efficiency in Tokamak plasmas.

    Science.gov (United States)

    Green, D L; Berry, L A; Chen, G; Ryan, P M; Canik, J M; Jaeger, E F

    2011-09-30

    Observations of improved radio frequency (rf) heating efficiency in ITER relevant high-confinement (H-)mode plasmas on the National Spherical Tokamak Experiment are investigated by whole-device linear simulation. The steady-state rf electric field is calculated for various antenna spectra and the results examined for characteristics that correlate with observations of improved or reduced rf heating efficiency. We find that launching toroidal wave numbers that give fast-wave propagation in the scrape-off plasma excites large amplitude (∼kV m(-1)) coaxial standing modes between the confined plasma density pedestal and conducting vessel wall. Qualitative comparison with measurements of the stored plasma energy suggests that these modes are a probable cause of degraded heating efficiency.

  2. Status of neutron diagnostics on the experimental advanced superconducting tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Zhong, G. Q.; Hu, L. Q., E-mail: lqhu@ipp.ac.cn; Pu, N.; Zhou, R. J.; Xiao, M.; Cao, H. R.; Li, K.; Huang, J.; Xu, G. S.; Wan, B. N. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Zhu, Y. B. [Department of Physics and Astronomy, University of California, Irvine, California 92697-4575 (United States); Fan, T. S.; Peng, X. Y.; Du, T. F.; Ge, L. J. [School of Physics and State Key Laboratory of Nuclear Physics and Technology, Peking University, Chengfu Road 201, 100871 Beijing (China)

    2016-11-15

    Neutron diagnostics have become a significant means to study energetic particles in high power auxiliary heating plasmas on the Experimental Advanced Superconducting Tokamak (EAST). Several kinds of neutron diagnostic systems have been implemented for time-resolved measurements of D-D neutron flux, fluctuation, emission profile, and spectrum. All detectors have been calibrated in laboratory, and in situ calibration using {sup 252}Cf neutron source in EAST is in preparation. A new technology of digitized pulse signal processing is adopted in a wide dynamic range neutron flux monitor, compact recoil proton spectrometer, and time of flight spectrometer. Improvements will be made continuously to the system to achieve better adaptation to the EAST’s harsh γ-ray and electro-magnetic radiation environment.

  3. Migration of tungsten dust in tokamaks: role of dust–wall collisions

    NARCIS (Netherlands)

    Ratynskaia, S.; Vignitchouk, L.; Tolias, P.; I. Bykov,; H. Bergsåker,; Litnovsky, A.; N. den Harder,; Lazzaro, E.

    2013-01-01

    The modelling of a controlled tungsten dust injection experiment in TEXTOR by the dust dynamics code MIGRAINe is reported. The code, in addition to the standard dust–plasma interaction processes, also encompasses major mechanical aspects of dust–surface collisions. The use of analytical expressions

  4. Study of heat flux deposition in the Tore Supra Tokamak; Etude des depots de chaleur dans le tokamak Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Carpentier, S.

    2009-02-15

    Accurate measurements of heat loads on internal tokamak components is essential for protection of the device during steady state operation. The optimisation of experimental scenarios also requires an in depth understanding of the physical mechanisms governing the heat flux deposition on the walls. The objective of this study is a detailed characterisation of the heat flux to plasma facing components (PFC) of the Tore Supra tokamak. The power deposited onto Tore Supra PFCs is calculated using an inverse method, which is applied to both the temperature maps measured by infrared thermography and to the enthalpy signals from calorimetry. The derived experimental heat flux maps calculated on the toroidal pumped limiter (TPL) are then compared with theoretical heat flux density distributions from a standard SOL-model. They are two experimental observations that are not consistent with the model: significant heat flux outside the theoretical wetted area, and heat load peaking close to the tangency point between the TPL and the last closed field surface (LCFS). An experimental analysis for several discharges with variable security factors q is made. In the area consistent with the theoretical predictions, this parametric study shows a clear dependence between the heat flux length lambda{sub q} (estimated in the SOL (scrape-off layer) from the IR measurements) and the magnetic configuration. We observe that the spreading of heat fluxes on the component is compensated by a reduction of the power decay length lambda{sub q} in the SOL when q decreases. On the other hand, in the area where the derived experimental heat loads are not consistent with the theoretical predictions, we observe that the spreading of heat fluxes outside the theoretical boundary increases when q decreases, and is thus not counterbalanced. (author)

  5. Selected methods of electron-and ion-diagnostics in tokamak scrape-off-layer

    Directory of Open Access Journals (Sweden)

    Sadowski Marek J.

    2015-06-01

    Full Text Available This invited paper considers reasons why exact measurements of fast electron and ion losses in tokamaks, and particularly i n a scrape-off-layer and near a divertor region, are necessary in order to master nuclear fusion energy production. Attention is also paid to direct measurements of escaping fusion products from D-D and D-T reactions, and in particular of fast alphas which might be used for plasma heating. The second part describes the generation of so-called runaway and ripple-born electrons which might induce high energy losses and cause severe damages of internal walls in fusion facilities. Advantages and disadvantages of different diagnostic methods applied for studies of such fast electrons are discussed. Particular attention is paid to development of a direct measuring technique based on the Cherenkov effect which might be induced by fast electrons in appropriate radiators. There are presented various versions of Cherenkov-type probes which have been developed by the NCBJ team and applied in different tokamak experiments. The third part is devoted to direct measurements of fast ions (including those produced by the nuclear fusion reactions which can escape from a high-temperature plasma region. Investigation of fast fusion-produced protons from tokamak discharges is reported. New ion probes, which were developed by the NCBJ team, are also presented. For the first time there is given a detailed description of an ion pinhole camera, which enables irradiation of several nuclear track detectors during a single tokamak discharge, and a miniature Thomson-type mass-spectrometer, which can be used for ion measurements at plasma borders.

  6. Observation of Energetic Particle Driven Modes Relevant to Advanced Tokamak Regimes

    Energy Technology Data Exchange (ETDEWEB)

    R. Nazikian; B. Alper; H.L. Berk; D. Borba; C. Boswell; R.V. Budny; K.H. Burrell; C.Z. Cheng; E.J. Doyle; E. Edlund; R.J. Fonck; A. Fukuyama; N.N. Gorelenkov; C.M. Greenfield; D.J. Gupta; M. Ishikawa; R.J. Jayakumar; G.J. Kramer; Y. Kusama; R.J. La Haye; G.R. McKee; W.A. Peebles; S.D. Pinches; M. Porkolab; J. Rapp; T.L. Rhodes; S.E. Sharapov; K. Shinohara; J.A. Snipes; W.M. Solomon; E.J. Strait; M. Takechi; M.A. Van Zeeland; W.P. West; K.L. Wong; S. Wukitch; L. Zeng

    2004-10-21

    Measurements of high-frequency oscillations in JET [Joint European Torus], JT-60U, Alcator C-Mod, DIII-D, and TFTR [Tokamak Fusion Test Reactor] plasmas are contributing to a new understanding of fast ion-driven instabilities relevant to Advanced Tokamak (AT) regimes. A model based on the transition from a cylindrical-like frequency-chirping mode to the Toroidal Alfven Eigenmode (TAE) has successfully encompassed many of the characteristics seen in experiments. In a surprising development, the use of internal density fluctuation diagnostics has revealed many more modes than has been detected on edge magnetic probes. A corollary discovery is the observation of modes excited by fast particles traveling well below the Alfven velocity. These observations open up new opportunities for investigating a ''sea of Alfven Eigenmodes'' in present-scale experiments, and highlight the need for core fluctuation and fast ion measurements in a future burning-plasma experiment.

  7. Generation of two-column helicon plasma on KAIST-TOKAMAK

    Science.gov (United States)

    Jeon, S. J.; Huh, S. W.; Kim, J.; Lee, T. S.; Moon, S. Y.; Choe, W.; Choi, D. I.

    2000-10-01

    Industrial plasma application studies reveal that helicon waves provide high ionization rate even at modest rf input power. This suggests that helicon waves be effectively used for plasma pre-ionization/startup, and plasma heating in a tokamak. The two-column helicon plasma was produced with a Nagoya type ¥2 antenna which was modified for toroidal geometry of KAIST-TOKAMAK. The observed two columns locate at the same major radius and they move outward as toroidal magnetic field increases. In addition to the 2D image captured by a CCD camera, an 8-channel Langmuir probe array is used to measure the density profile. Parallel wave number is measured by magnetic pickup probes and a phase detector in order to study wave generation and propagation inside the plasma.

  8. Development of a Cherenkov-type diagnostic system to study runaway electrons within the COMPASS tokamak

    Science.gov (United States)

    Rabinski, M.; Jakubowski, L.; Malinowski, K.; Sadowski, M. J.; Zebrowski, J.; Jakubowski, M. J.; Mirowski, R.; Weinzettl, V.; Ficker, O.; Mlynar, J.; Panek, R.; Paprok, R.; Vlainic, M.

    2017-10-01

    Direct measurements of fast electrons, which are produced in high-temperature plasma and escape from tokamak-type facilities, are of particular interest for ITER and future fusion devices, where intense runaway electrons (RE) can significantly damage the first wall components. Therefore, the runaway control and mitigation based on credible measuring methods should be developed already in present devices. A team from the National Centre for Nuclear Research (NCBJ), Poland, developed special probes equipped with Cherenkov-type detectors for measurements of the fast electrons within edge plasmas of tokamaks. Studies of the fast runaway electrons were extensively carried out at the COMPASS tokamak at the Institute of Plasma Physics (IPP) in Prague during experimental campaigns in 2014–2016. In order to investigate an electron-beam energy distribution a three-channel probe equipped with the Cherenkov-type detectors sensitive to electrons of different energies has been constructed. The measurements performed by means of these detectors showed that the first fast electron peak appears usually in the current ramp-up phase, even before the hard X-rays (HXR) pulse. Some electron signals can also be observed during subsequent HXR emissions. However, the most distinct electron peaks in all energy channels appear mainly during the plasma disruption. A correlation of Cherenkov signals with the MHD activity was also studied.

  9. New Host Record for Camponotophilus delvarei (Hymenoptera: Eurytomidae, a Parasitoid of Microdontine Larvae (Diptera: Syrphidae, Associated with the Ant Camponotus sp. aff. textor

    Directory of Open Access Journals (Sweden)

    Gabriela Pérez-Lachaud

    2013-01-01

    Full Text Available Microdontine syrphid flies are obligate social parasites of ants. Larvae prey on ant brood whereas adults live outside the nests. Knowledge of their interaction with their host is often scarce, as it is information about their natural enemies. Here we report the first case of parasitism of a species of microdontine fly by a myrmecophilous eurytomid wasp. This is also the first host record for Camponotophilus delvarei Gates, a recently described parasitic wasp discovered in Chiapas, Mexico, within the nests of the weaver ant, Camponotus sp. aff. textor Forel. Eleven pupal cases of a microdontine fly were found within a single nest of this ant, five of them being parasitized. Five adult C. delvarei females were reared from a puparium and 29 female and 2 male pupae were obtained from another one. The eurytomid is a gregarious, primary ectoparasitoid of larvae and pupae of Microdontinae, its immature stages developing within the protective puparium of the fly. The species is synovigenic. Adult females likely locate and parasitize their hosts within the ant nest. As some species of Microdontinae are considered endangered, their parasitoids are likewise threatened and in need of accurate and urgent surveys in the future.

  10. Diagnostics and control for the steady state and pulsed tokamak DEMO

    Czech Academy of Sciences Publication Activity Database

    Orsitto, F.P.; Villari, R.; Moro, F.; Todd, T.N.; Lilley, S.; Jenkins, I.; Felton, R.; Biel, W.; Silva, A.; Scholz, M.; Rzadkiewicz, J.; Ďuran, Ivan; Tardocchi, M.; Gorini, G.; Morlock, C.; Federici, G.; Litnovsky, A.

    2016-01-01

    Roč. 56, č. 2 (2016), č. článku 026009. ISSN 0029-5515 Institutional support: RVO:61389021 Keywords : measurement systems, fusion reactor, fusion plasma diagnostics * fusion reactor * fusion plasma diagnostics * DEMO * Hall sensors * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.307, year: 2016 http://iopscience.iop.org/article/10.1088/0029-5515/56/2/026009

  11. Digital controlled pulsed electric system of the ETE tokamak. First report; Sistema eletrico pulsado com controle digital do Tokamak ETE (experimento Tokamak esferico). Primeiro relatorio

    Energy Technology Data Exchange (ETDEWEB)

    Barbosa, Luis Felipe de F.P.W.; Del Bosco, Edson

    1997-12-31

    This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author) 55 refs., 40 figs.

  12. Energetic particles in spherical tokamak plasmas

    Science.gov (United States)

    McClements, K. G.; Fredrickson, E. D.

    2017-05-01

    Spherical tokamaks (STs) typically have lower magnetic fields than conventional tokamaks, but similar mass densities. Suprathermal ions with relatively modest energies, in particular beam-injected ions, consequently have speeds close to or exceeding the Alfvén velocity, and can therefore excite a range of Alfvénic instabilities which could be driven by (and affect the behaviour of) fusion α-particles in a burning plasma. STs heated with neutral beams, including the small tight aspect ratio tokamak (START), the mega amp spherical tokamak (MAST), the national spherical torus experiment (NSTX) and Globus-M, have thus provided an opportunity to study toroidal Alfvén eigenmodes (TAEs), together with higher frequency global Alfvén eigenmodes (GAEs) and compressional Alfvén eigenmodes (CAEs), which could affect beam current drive and channel fast ion energy into bulk ions in future devices. In NSTX GAEs were correlated with a degradation of core electron energy confinement. In MAST pulses with reduced magnetic field, CAEs were excited across a wide range of frequencies, extending to the ion cyclotron range, but were suppressed when hydrogen was introduced to the deuterium plasma, apparently due to mode conversion at ion-ion hybrid resonances. At lower frequencies fishbone instabilities caused fast particle redistribution in some MAST and NSTX pulses, but this could be avoided by moving the neutral beam line away from the magnetic axis or by operating the plasma at either high density or elevated safety factor. Fast ion redistribution has been observed during GAE avalanches on NSTX, while in both NSTX and MAST fast ions were transported by saturated kink modes, sawtooth crashes, resonant magnetic perturbations and TAEs. The energy dependence of fast ion redistribution due to both sawteeth and TAEs has been studied in Globus-M. High energy charged fusion products are unconfined in present-day STs, but have been shown in MAST to provide a useful diagnostic of beam ion

  13. Differential and Integral Models of TOKAMAK

    Directory of Open Access Journals (Sweden)

    Ivo Dolezel

    2004-01-01

    Full Text Available Modeling of 3D electromagnetic phenomena in TOKAMAK with typically distributed main and additional coils is not an easy business. Evaluated must be not only distribution of the magnetic field, but also forces acting in particular coils. Use of differential methods (such as FDM or FEM for this purpose may be complicated because of geometrical incommensurability of particular subregions in the investigated area or problems with the boundary conditions. That is why integral formulation of the problem may sometimes be an advantages. The theoretical analysis is illustrated on an example processed by both methods, whose results are compared and discussed.

  14. Observation of short time-scale spectral emissions at millimeter wavelengths with the new CTS diagnostic on the FTU tokamak

    DEFF Research Database (Denmark)

    Bruschi, A.; Alessi, E.; Bin, W.

    2017-01-01

    On the FTU tokamak, the collective Thomson scattering (CTS) diagnostic was renewed for investigating the possible excitation of parametric decay instabilities (PDI) by electron cyclotron (EC) or CTS probe beams in presence of magnetic islands and measure their effects on the EC power absorption...

  15. Correlation of electron beams and hard x-ray emissions in ISTTOK Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Jakubowski, L.; Malinowski, K.; Sadowski, M.J.; Zebrowski, J.; Rabinski, M.; Jakubowski, M.J. [National Centre for Nuclear Research (NCBJ), Otwock (Poland); Plyusnin, V.V.; Fernandes, H.; Silva, C.; Duarte, P. [Association Euratom/IST, Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Lisboa (Portugal)

    2013-11-15

    The paper reports on experimental studies of electron beams in the ISTTOK tokamak, those were performed by means of an improved four-channel detector. The Cherenkov-type detector measuring head was equipped with four radiators made of two types of alumina-nitrate (AlN) poly-crystals: machinable and translucent ones, both of 10 mm in diameter and 2.5 mm in thickness. The movable support that enabled the whole detectors to be placed inside the tokamak vacuum chamber, at chosen positions along the ISTTOK minor radius. Since the electron energy distribution is one of the most important characteristics of tokamak plasmas, the main aim of the study was to perform estimations of an energy spectrum of the recorded electrons. For this purpose the radiators were coated with molybdenum (Mo) layers of different thickness. The technique based on the use of Cherenkov-type detectors enabled the detection of fast electrons (of energy above 66 keV) and determination of their spatial and temporal characteristics in the ISTTOK experiment. Measurements of hard X-rays (HXR), which were emitted during ISTTOK discharges, have also been performed. Particular attention was paid to the correlation measurements of HXR pulses with run-away electron beams. (copyright 2013 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  16. Bulk ion heating with ICRF waves in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-10

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

  17. Continuous, saturation, and discontinuous tokamak plasma vertical position control systems

    Energy Technology Data Exchange (ETDEWEB)

    Mitrishkin, Yuri V., E-mail: y_mitrishkin@hotmail.com [M. V. Lomonosov Moscow State University, Faculty of Physics, Moscow 119991 (Russian Federation); Pavlova, Evgeniia A., E-mail: janerigoler@mail.ru [M. V. Lomonosov Moscow State University, Faculty of Physics, Moscow 119991 (Russian Federation); Kuznetsov, Evgenii A., E-mail: ea.kuznetsov@mail.ru [Troitsk Institute for Innovation and Fusion Research, Moscow 142190 (Russian Federation); Gaydamaka, Kirill I., E-mail: k.gaydamaka@gmail.com [V. A. Trapeznikov Institute of Control Sciences of the Russian Academy of Sciences, Moscow 117997 (Russian Federation)

    2016-10-15

    Highlights: • Robust new linear state feedback control system for tokamak plasma vertical position. • Plasma vertical position relay control system with voltage inverter in sliding mode. • Design of full models of multiphase rectifier and voltage inverter. • First-order unit approximation of full multiphase rectifier model with high accuracy. • Wider range of unstable plant parameters of stable control system with multiphase rectifier. - Abstract: This paper is devoted to the design and comparison of unstable plasma vertical position control systems in the T-15 tokamak with the application of two types of actuators: a multiphase thyristor rectifier and a transistor voltage inverter. An unstable dynamic element obtained by the identification of plasma-physical DINA code was used as the plasma model. The simplest static feedback state space control law was synthesized as a linear combination of signals accessible to physical measurements, namely the plasma vertical displacement, the current, and the voltage in a horizontal field coil, to solve the pole placement problem for a closed-loop system. Only one system distinctive parameter was used to optimize the performance of the feedback system, viz., a multiple real pole. A first-order inertial unit was used as the rectifier model in the feedback. A system with a complete rectifier model was investigated as well. A system with the voltage inverter model and static linear controller was brought into a sliding mode. As this takes place, real time delays were taken into account in the discontinuous voltage inverter model. The comparison of the linear and sliding mode systems showed that the linear system enjoyed an essentially wider range of the plant model parameters where the feedback system was stable.

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

    Directory of Open Access Journals (Sweden)

    Peter Ondac

    2015-04-01

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

  19. MDSplus integration at TCABR tokamak: Current status

    Energy Technology Data Exchange (ETDEWEB)

    Sá, W.P. de, E-mail: pires@if.usp.br; Ronchi, G., E-mail: gronchi@if.usp.br

    2016-11-15

    Highlights: • The implementation of MDSplus in TCABR tokamak, current status. • Interfaces between the system already installed and the MDSplus. • Web MDSplus interface. - Abstract: Experimental data for the TCABR tokamak is currently stored in MDSplus (Model Driven System Plus) database. The access to the data recorded during the experiments is performed using tools and libraries available by MDSplus system. The MDSplus system is widely used in different physics experiments, especially in plasmas physics and nuclear fusion. This standardized environment enables easy interaction among scientists of different experiments in different countries without the need to understand the particular characteristics of control, data acquisition and analysis, and remote access (CODAS) customized in each laboratory. In the first phase of implementation, intermediate interfaces had been developed between the legacy proprietary system and the MDSplus. In a second phase, the new diagnostic systems had been directly included in the created MDSplus system in the laboratory. After three years of use, the system installed on TCABR proved extremely efficient and significantly increased productivity in data analysis by involved scientists, regardless of whether they are locally at the TCABR, or accessing the system remotely from their home laboratories. The third phase, and subject of this article, are the development and implementation of the following systems: (i) web tools for the visualization of data, integrated with the experiment logbook, (ii) integration of MDSplus with applications (LabVIEW + MDSplus) and newer data acquisition hardware.

  20. System studies of compact ignition tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Galambos, J.D.; Blackfield, D.T.; Peng, Y.K.M.; Reid, R.L.; Strickler, D.J.; Selcow, E.

    1987-08-01

    The new Tokamak Systems Code, used to investigate Compact Ignition Tokamaks (CITs), can simultaneously vary many parameters, satisfy many constraints, and minimize or maximize a figure of merit. It is useful in comparing different CIT design configurations over wide regions of parameter space and determining a desired design point for more detailed physics and engineering analysis, as well as for performing sensitivity studies for physics or engineering issues. Operational windows in major radius (R) and toroidal field (B) space for fixed ignition margin are calculated for the Ignifed and Inconel candidate CITs. The minimum R bounds are predominantly physics limited, and the maximum R portions of the windows are engineering limited. For a modified Kaye-Goldston plasma-energy-confinement scaling, the minimum size is 1.15 m for the Ignifed device and 1.25 m for the Inconel device. With the Ignition Technical Oversight Committee (ITOC) physics guidance of B/sup 2/a/q and I/sub p/ >10 MA, the Ignifed and Base-line Inconel devices have a minimum size of 1.2 and 1.25 m and a toroidal field of 11 and 10.4 T, respectively. Sensitivity studies show Ignifed to be more sensitive to coil temperature changes than the Inconel device, whereas the Inconel device is more sensitive to stress perturbations.

  1. The Spherical Tokamak MEDUSA for Costa Rica

    Science.gov (United States)

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

    2012-10-01

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

  2. Interaction of candidate plasma facing materials with tokamak plasma in COMPASS

    Science.gov (United States)

    Matějíček, Jiří; Weinzettl, Vladimír; Macková, Anna; Malinský, Petr; Havránek, Vladimír; Naydenkova, Diana; Klevarová, Veronika; Petersson, Per; Gasior, Pawel; Hakola, Antti; Rubel, Marek; Fortuna, Elzbieta; Kolehmainen, Jukka; Tervakangas, Sanna

    2017-09-01

    The interaction of tokamak plasma with several materials considered for the plasma facing components of future fusion devices was studied in a small-size COMPASS tokamak. These included mainly tungsten as the prime candidate and chromium steel as an alternative whose suitability was to be assessed. For the experiments, thin coatings of tungsten, P92 steel and nickel on graphite substrates were prepared by arc-discharge sputtering. The samples were exposed to hydrogen and deuterium plasma discharges in the COMPASS tokamak in two modes: a) short exposure (several discharges) on a manipulator in the proximity of the separatrix, close to the central column, and b) long exposure (several months) at the central column, aligned with the other graphite tiles. During the discharges, standard plasma diagnostics were used and a local emission of spectral lines in the visible near ultraviolet regions, corresponding to the material erosion, was monitored. Before and after the plasma exposures, the sample surfaces were observed using scanning electron microscopy, the coatings thickness was measured using Rutherford backscattering spectroscopy, and the concentration profiles of hydrogen and deuterium were measured by elastic recoil detection analysis. The uniformity of the coatings and their thickness was verified before the exposure. After the exposure, no reduction of the thickness was observed, indicating the absence of 'global' erosion. Erosion was observed only in isolated spots, and attributed to unipolar arcing. Slightly larger erosion was found on the steel coatings compared to the tungsten ones. Incorporation of deuterium in a thin surface layer was observed, in dependence on the exposure mode. Additionally, boron enrichment of the long-exposure samples was observed, as a result of the tokamak chamber boronization.

  3. Dust characterization in FTU tokamak

    Science.gov (United States)

    De Angeli, M.; Maddaluno, G.; Laguardia, L.; Ripamonti, D.; Perelli Cippo, E.; Apicella, M. L.; Conti, C.; Giacomi, G.; Grosso, G.

    2015-08-01

    Dust present in the vessel of FTU has been collected and analysed. Being FTU a device with full metal plasma facing components for the whole life and equipped with a liquid lithium limiter (LLL) make FTU of special interest from a point of view of dust studies. Analyses were conducted by standard dust analysis methods and by dedicated analysis, as X-rays and neutron diffraction, to investigate the presence of lithium compounds due the presence of the LLL in FTU. Dust collected near the LLL presents a different elemental composition, namely Li compounds, compared to the dust collected in the rest of the vessel; in particular LiO2, LiOH, and Li2CO3. On the basis of these results, the formation of Li2CO3 is proposed via a two steps process. Results of fuel retention measured by thermal desorption spectroscopy (TDS) method show that fuel retention should not be an issue for FTU.

  4. Soft-X Spectroscopy on the Phaedrus-T Tokamak

    Science.gov (United States)

    Regan, Sean Patrick

    Soft-x-ray spectroscopic techniques were used to estimate the electron temperature, to study the impurity transport in ohmic and H-mode plasmas, and to measure the vertical poloidal asymmetry of the impurity distribution in the central region of the Phaedrus-T tokamak plasma. The spectroscopic measurements were performed with a soft-x -ray polychromator that utilized multilayer mirrors (MLM) as the dispersive elements. This MLM based device, which is referred to as the MLM Polychromator, was designed and constructed at the Johns Hopkins University, and then mounted and operated on the Phaedrus-T tokamak. The purpose of this instrument was to resolve spectrally the soft-x-ray emissions of the intrinsic impurities--carbon, oxygen, and iron--and spatially resolve their distributions in the plasma with a temporal resolution of 1 ms. The MLM Polychromator has four spectral channels, each of which is tuned for a particular soft-x-ray spectral emission line of an intrinsic impurity from the Phaedrus -T tokamak. The selected combinations of emissions defined two spatial scanning modes of operation for the MLM Polychromator: the C-O mode and the C-Fe mode. In the C-O mode the MLM Polychromator simultaneously monitors the Lyman alpha and Lyman beta emissions of H I-like O at 19.0 A and 16.0 A, respectively, the Lyman alpha emission of H I-like C at 33.7 A, and the blended singlet and triplet transitions of He I-like C at 40.3 A and 40.7 A. In the C-Fe mode it simultaneously monitors the Lyman alpha and Lyman beta emissions of H I-like C at 33.7 A and 28.5 A, respectively, as well as the 15.6 A and 94 A emissions of F I-like Fe. The wavelength resolution of the MLM Polychromator varies from 0.3 A at 16.0 A to 3.6 A at 94 A. Calibrated soft-x-ray spectra and emission profiles are presented along with the spectral fitting and the spatial inversion procedures that were developed in order to analyze the impurity emissions. A presentation of the photometric calibrations of the MLM

  5. Dust characterization in FTU tokamak

    Energy Technology Data Exchange (ETDEWEB)

    De Angeli, M., E-mail: deangeli@ifp.cnr.it [Istituto di Fisica del Plasma – Consiglio Nazionale delle Ricerche, Milan (Italy); Maddaluno, G. [ENEA Unità Tecnica Fusione, C.R. ENEA Frascati, CP65, 00044 Frascati (Italy); Laguardia, L. [Istituto di Fisica del Plasma – Consiglio Nazionale delle Ricerche, Milan (Italy); Ripamonti, D. [Istituto per l’Energetica e le Interfasi – Consiglio Nazionale delle Ricerche, Milan (Italy); Perelli Cippo, E. [Istituto di Fisica del Plasma – Consiglio Nazionale delle Ricerche, Milan (Italy); Apicella, M.L. [ENEA Unità Tecnica Fusione, C.R. ENEA Frascati, CP65, 00044 Frascati (Italy); Conti, C. [Istituto per la Conservazione e la Valorizzazione dei Beni Culturali – CNR, Milan (Italy); Giacomi, G. [ENEA Unità Tecnica Fusione, C.R. ENEA Frascati, CP65, 00044 Frascati (Italy); Grosso, G. [Istituto di Fisica del Plasma – Consiglio Nazionale delle Ricerche, Milan (Italy)

    2015-08-15

    Dust present in the vessel of FTU has been collected and analysed. Being FTU a device with full metal plasma facing components for the whole life and equipped with a liquid lithium limiter (LLL) make FTU of special interest from a point of view of dust studies. Analyses were conducted by standard dust analysis methods and by dedicated analysis, as X-rays and neutron diffraction, to investigate the presence of lithium compounds due the presence of the LLL in FTU. Dust collected near the LLL presents a different elemental composition, namely Li compounds, compared to the dust collected in the rest of the vessel; in particular LiO{sub 2}, LiOH, and Li{sub 2}CO{sub 3}. On the basis of these results, the formation of Li{sub 2}CO{sub 3} is proposed via a two steps process. Results of fuel retention measured by thermal desorption spectroscopy (TDS) method show that fuel retention should not be an issue for FTU.

  6. [Injection of compact toroids for tokamak fueling and current drive]. Progress report, 1990--1991

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, D.Q.; Rogers, J.H.; Thomas, J.C.; Evans, R.; Foley, R.; Hillyer, T.

    1991-12-31

    The experimental goals for the 1990--1991 period were the operation of the Davis Diverted Tokamak(DDT), the beat wave experiment, and the construction of the compact toroid injection experiment(CTIX). The experiment results from these areas are summarized in the posters given in the APS meeting past November. Here we shall describe the technical progress of the development of the diagnostic system for beat wave experiment, and CT injection especially in relation to the up coming injection experiments into DDT tokamak. The tokamak operation of DDT over the past year has been focused in two parameter ranges. The long pulse discharges (over 100 msec), and the low q short pulse discharges (about 10 msec). We found that the long pulse discharges required a position feedback more sophisticated than the simple passive program that we have. We are in the process of assembling this system. We also found an interesting low q(a) operating regime. Here an equilibrium can be established for a toroidal field between .5 and 1 kG. The typical plasma current is > 5kA. The density of the plasma is between 10{sup 12} and 10{sup 13} cm{sup {minus}3}. The plasma condition in these discharge are sufficiently mild that diagnostic probes can be used to measure various plasma fluctuations. We believe that this will be the regime best suited to study the interaction between the tokamak plasma and the compact toroid. A sophisticated probe system of both electrostatic and electromagnetic types similar to those used in the beat wave experiment has been designed for the up coming experiments.

  7. Calculations of axisymmetric stability of tokamak plasmas with active and passive feedback

    Energy Technology Data Exchange (ETDEWEB)

    Ward, D.J.; Jardin, S.C.; Cheng, C.Z.

    1991-07-01

    A new linear MHD stability code, NOVA-W, has been developed in order to study feedback stabilization of the axisymmetric mode in deformable tokamak plasmas. The NOVA-W code is a modification of the non-variational MHD stability code NOVA that includes the effects of resistive passive conductors and active feedback circuits. The vacuum calculation has been reformulated in terms of the perturbed poloidal flux to allow the inclusion of perturbed toroidal currents outside the plasma. The boundary condition at the plasma-vacuum interface relates the instability displacement to the perturbed poloidal flux. This allows a solution of the linear MHD stability equations with the feedback effects included. The passive stability predictions of the code have been tested both against a simplified analytic model and against a different numerical calculation for a realistic tokamak configuration. The comparisons demonstrate the accuracy of the NOVA-W results. Active feedback calculations are performed for the CIT tokamak design demonstrating the effect of varying the position of the flux loops that provide the measurements of vertical displacement. The results compare well with those computed earlier using a less efficient nonlinear code. 37 refs., 13 figs.

  8. Neutral beam injector performance on the PLT and PDX tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Schilling, G.; Ashcroft, D.L.; Eubank, H.P.; Grisham, L.R.; Kozub, T.A.; Kugel, H.W.; Rossmassler, J.; Williams, M.D.

    1981-02-01

    An overall injector system description is presented first, and this will be followed by a detailed discussion of those problems unique to multiple injector operation on the tokamaks, i.e., power transmission, conditioning, reliability, and failures.

  9. Tokamak reactor cost model based on STARFIRE/WILDCAT costing

    Energy Technology Data Exchange (ETDEWEB)

    Evans, K. Jr.

    1983-03-01

    A cost model is presented which is useful for survey and comparative studies of tokamak reactors. The model is heavily based on STARFIRE and WILDCAT costing guidelines, philosophies, and procedures and reproduces the costing for these devices quite accurately.

  10. Three-Dimensional Analysis of Tokamaks and Stellarators

    National Research Council Canada - National Science Library

    Paul R. Garabedian

    2008-01-01

    The NSTAB equilibrium and stability code and the TRAN Monte Carlo transport code furnish a simple but effective numerical simulation of essential features of present tokamak and stellarator experiments...

  11. Plasma Current Start-up in a Spherical Tokamak

    Science.gov (United States)

    Mitarai, Osamu; Kessel, Charles; Hirose, Akira

    The various plasma current start-up techniques and related topics in a spherical tokamak (ST) device are described. The Ohmic heating coil current clamp experiments in NSTX are described and discussed, and the plasma current start-up experiments in the STOR-M tokamak with iron core and the outer vertical field coil is presented as one of technique for a plasma current start-up in a ST.

  12. Study of neutron generation in the compact tokamak TUMAN-3M in support of a tokamak-based fusion neutron source

    Science.gov (United States)

    Kornev, V. A.; Askinazi, L. G.; Belokurov, A. A.; Chernyshev, F. V.; Lebedev, S. V.; Melnik, A. D.; Shabelsky, A. A.; Tukachinsky, A. S.; Zhubr, N. A.

    2017-12-01

    The paper presents DD neutron flux measurements in neutron beam injection (NBI) experiments aimed at the optimization of target plasma and heating beam parameters to achieve maximum neutron flux in the TUMAN-3M compact tokamak. Two ion sources of different design were used, which allowed the separation of the beam’s energy and power influence on the neutron rate. Using the database of experiments performed with the two ion sources, an empirical scaling was derived describing the neutron rate dependence on the target plasma and heating beam parameters. Numerical modeling of the neutron rate in the NBI experiments performed using the ASTRA transport code showed good agreement with the scaling.

  13. Cooldown of the Compact Ignition Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Keeton, D.C.

    1987-08-01

    Cooldown of the Compact Ignition Tokamak (CIT) with the baseline liquid nitrogen cooling system was analyzed. On the basis of this analysis and present knowledge of the two-phase heat transfer, the current baseline CIT can be cooled down in about 1.5 h. An extensive heat transfer test program is recommended to reduce uncertainty in the heat transfer performance and to explore methods for minimizing the cooldown time. An alternate CIT cooldown system is described which uses a pressurized gaseous helium coolant in a closed-loop system. It is shown analytically that this system will cool down the CIT well within 1 h. Confidence in this analysis is sufficiently high that a heat transfer test program would not be necessary. The added cost of this alternate system is estimated to be about $5.3 million. This helium cooling system represents a reasonable backup approach to liquid nitrogen cooling of the CIT. 3 refs., 12 figs., 3 tabs.

  14. Decommissioning of the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    E. Perry; J. Chrzanowski; C. Gentile; R. Parsells; K. Rule; R. Strykowsky; M. Viola

    2003-10-28

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D&D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D&D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D&D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget.

  15. Sliding Mode Control of a Tokamak Transformer

    Energy Technology Data Exchange (ETDEWEB)

    Romero, J. A.; Coda, S.; Felici, F.; Moret, J. M.; Paley, J.; Sevillano, G.; Garrido, I.; Le, H. B.

    2012-06-08

    A novel inductive control system for a tokamak transformer is described. The system uses the flux change provided by the transformer primary coil to control the electric current and the internal inductance of the secondary plasma circuit load. The internal inductance control is used to regulate the slow flux penetration in the highly conductive plasma due to the skin effect, providing first-order control over the shape of the plasma current density profile. Inferred loop voltages at specific locations inside the plasma are included in a state feedback structure to improve controller performance. Experimental tests have shown that the plasma internal inductance can be controlled inductively for a whole pulse starting just 30ms after plasma breakdown. The details of the control system design are presented, including the transformer model, observer algorithms and controller design. (Author) 67 refs.

  16. Tokamak foundation in USSR/Russia 1950-1990

    Science.gov (United States)

    Smirnov, V. P.

    2010-01-01

    In the USSR, nuclear fusion research began in 1950 with the work of I.E. Tamm, A.D. Sakharov and colleagues. They formulated the principles of magnetic confinement of high temperature plasmas, that would allow the development of a thermonuclear reactor. Following this, experimental research on plasma initiation and heating in toroidal systems began in 1951 at the Kurchatov Institute. From the very first devices with vessels made of glass, porcelain or metal with insulating inserts, work progressed to the operation of the first tokamak, T-1, in 1958. More machines followed and the first international collaboration in nuclear fusion, on the T-3 tokamak, established the tokamak as a promising option for magnetic confinement. Experiments continued and specialized machines were developed to test separately improvements to the tokamak concept needed for the production of energy. At the same time, research into plasma physics and tokamak theory was being undertaken which provides the basis for modern theoretical work. Since then, the tokamak concept has been refined by a world-wide effort and today we look forward to the successful operation of ITER.

  17. Effects of electrode biasing in STOR-M Tokamak

    Science.gov (United States)

    Basu, Debjyoti; Nakajima, Masaru; Rohollahi, Akbar; McColl, David; Adegun, Joseph; Xiao, Chijin; Hirose, Akira

    2015-11-01

    STOR-M is an iron-core, limiter based tokamak with major and minor radii of 46cm and 12 cm, respectively. Recently, electrode biasing experiments have been carried to study the improved confinement. For this purpose we have developed a DC power supply which can be gated by a high power SCR. The rectangular SS electrode has a height of 10 cm, a width of 2 cm and a thickness of 0.2 cm. The radial position of the electrode throughout the experiments is kept around 4mm inside the limiter in the plasma edge region. After application of positive bias with voltages between +90 V to +110 V during the plasma discharge current flat top with slightly higher edge-qa (nearly 5 to 6), noticeable increment of average plasma density and soft x-ray intensity along the central chord have been observed. No distinguishable change in H α emission has been measured. These phenomena may be attributed to improved confinement formed at the inner region but not at the edge. In the upcoming experimental campaign, Ion Doppler spectroscopy will be used to measure possible velocity shear inside the inner plasma region. Edge plasma pressure gradient will also be measured using Langmuir probes. Detailed experimental results will be presented.

  18. Structural properties of resonant electric and magnetic fields correlation with X-ray generation and MHD activity in tokamak

    Science.gov (United States)

    Salar Elahi, A.; Ghoranneviss, M.

    In this research we have investigated on a Runaway electron generation in IR-T1 tokamak. For this purpose we used the hard X-ray spectroscopy and magnetic diagnostic. Hard X-ray emission produces due to collision of the Runaway electrons with the plasma particles or tokamak limiters. Runaway electrons in tokamaks can cause serious damage to the first wall of the reactor and decrease its life time. Also, hard X-ray emission generated from high energy Runaway electrons lead to the plasma energy loss. Therefore, suggesting methods to minimize Runaway electrons in tokamaks are very important. Applying external resonant field is one of the methods for controlling the Magnetohydrodynamic (MHD) activity. Present study attempts to investigate the effects of limiter biasing and Resonant Helical magnetic Field (RHF) on the generation of Runaway electrons. For this purpose, plasma parameters such as plasma current, MHD oscillation, loop voltage, emitted hard X-ray intensity, Hα impurity, safety factor in the presence and absence of external fields, were measured. Frequency activity was investigated with FFT analysis. The results show that applying resonant fields can control the MHD activity, and then hard X-ray emitted from the Runaway electrons.

  19. Self-organized stationary states of tokamaks

    Science.gov (United States)

    Jardin, Stephen

    2015-11-01

    We report here on a nonlinear mechanism that forms and maintains a self-organized stationary (sawtooth free) state in tokamaks. This process was discovered by way of extensive long-time simulations using the M3D-C1 3D extended MHD code in which new physics diagnostics have been added. It is well known that most high-performance modes of tokamak operation undergo ``sawtooth'' cycles, in which the peaking of the toroidal current density triggers a periodic core instability which redistributes the current density. However, certain modes of operation are known, such as the ``hybrid'' mode in DIII-D, ASDEX-U, JT-60U and JET, and the long-lived modes in NSTX and MAST, which do not experience this cycle of instability. Empirically, it is observed that these modes maintain a non-axisymmetric equilibrium which somehow limits the peaking of the toroidal current density. The physical mechanism responsible for this has not previously been understood, but is often referred to as ``flux-pumping,'' in which poloidal flux is redistributed in order to maintain q0 >1. In this talk, we show that in long-time simulations of inductively driven plasmas, a steady-state magnetic equilibrium may be obtained in which the condition q0 >1 is maintained by a dynamo driven by a stationary marginal core interchange mode. This interchange mode, unstable because of the pressure gradient in the ultra-low shear region in the center region, causes a (1,1) perturbation in both the electrostatic potential and the magnetic field, which nonlinearly cause a (0,0) component in the loop voltage that acts to sustain the configuration. This hybrid mode may be a preferred mode of operation for ITER. We present parameter scans that indicate when this sawtooth-free operation can be expected.

  20. Transport Barriers in Bootstrap Driven Tokamaks

    Science.gov (United States)

    Staebler, Gary

    2017-10-01

    Maximizing the bootstrap current in a tokamak, so that it drives a high fraction of the total current, reduces the external power required to drive current by other means. Improved energy confinement, relative to empirical scaling laws, enables a reactor to more fully take advantage of the bootstrap driven tokamak. Experiments have demonstrated improved energy confinement due to the spontaneous formation of an internal transport barrier in high bootstrap fraction discharges. Gyrokinetic analysis, and quasilinear predictive modeling, demonstrates that the observed transport barrier is due to the suppression of turbulence primarily due to the large Shafranov shift. ExB velocity shear does not play a significant role in the transport barrier due to the high safety factor. It will be shown, that the Shafranov shift can produce a bifurcation to improved confinement in regions of positive magnetic shear or a continuous reduction in transport for weak or negative magnetic shear. Operation at high safety factor lowers the pressure gradient threshold for the Shafranov shift driven barrier formation. The ion energy transport is reduced to neoclassical and electron energy and particle transport is reduced, but still turbulent, within the barrier. Deeper into the plasma, very large levels of electron transport are observed. The observed electron temperature profile is shown to be close to the threshold for the electron temperature gradient (ETG) mode. A large ETG driven energy transport is qualitatively consistent with recent multi-scale gyrokinetic simulations showing that reducing the ion scale turbulence can lead to large increase in the electron scale transport. A new saturation model for the quasilinear TGLF transport code, that fits these multi-scale gyrokinetic simulations, can match the data if the impact of zonal flow mixing on the ETG modes is reduced at high safety factor. This work was supported by the U.S. Department of Energy under DE-FG02-95ER54309 and DE-FC02

  1. Overview of experimental results on the HL-2A tokamak

    Science.gov (United States)

    Yan, L. W.; Duan, X. R.; Ding, X. T.; Dong, J. Q.; Yang, Q. W.; Liu, Yi; Zou, X. L.; Liu, D. Q.; Xuan, W. M.; Chen, L. Y.; Rao, J.; Song, X. M.; Huang, Y.; Mao, W. C.; Wang, Q. M.; Li, Q.; Cao, Z.; Li, B.; Cao, J. Y.; Lei, G. J.; Zhang, J. H.; Li, X. D.; Chen, W.; Cheng, J.; Cui, C. H.; Cui, Z. Y.; Deng, Z. C.; Dong, Y. B.; Feng, B. B.; Gao, Q. D.; Han, X. Y.; Hong, W. Y.; Huang, M.; Ji, X. Q.; Kang, Z. H.; Kong, D. F.; Lan, T.; Li, G. S.; Li, H. J.; Li, Qing; Li, W.; Li, Y. G.; Liu, A. D.; Liu, Z. T.; Luo, C. W.; Mao, X. H.; Pan, Y. D.; Peng, J. F.; Shi, Z. B.; Song, S. D.; Song, X. Y.; Sun, H. J.; Wang, A. K.; Wang, M. X.; Wang, Y. Q.; Xiao, W. W.; Xie, Y. F.; Yao, L. H.; Yao, L. Y.; Yu, D. L.; Yuan, B. S.; Zhao, K. J.; Zhong, G. W.; Zhou, J.; Zhou, Y.; Yan, J. C.; Yu, C. X.; Pan, C. H.; Liu, Yong; HL-2A Team

    2011-09-01

    The physics experiments on the HL-2A tokamak have been focused on confinement improvement, particle and thermal transport, zonal flow and turbulence, filament characteristics, energetic particle induced modes and plasma fuelling efficiency since 2008. ELMy H-mode discharges are achieved in a lower density regime using a combination of NBI heating with ECRH. The power threshold is found to increase with a decrease in density, almost independent of the launching order of the ECRH and NBI heating power. The pedestal density profiles in the H-mode discharges are measured. The particle outward convection is observed during the pump-out transient phase with ECRH. The negative density perturbation (pump-out) is observed to propagate much faster than the positive one caused by out-gassing. The core electron thermal transport reduction triggered by far off-axis ECRH switch-off is investigated. The coexistence of low frequency zonal flow (LFZF) and geodesic acoustic mode (GAM) is observed. The dependence of the intensities of LFZFs and GAMs on the safety factor and ECRH power is identified. The 3D spatial structures of plasma filaments are measured in the boundary plasma and large-scale structures along a magnetic field line analysed for the first time. The beta-induced Alfvén eigenmodes (BAEs), excited by large magnetic islands (m-BAE) and by energetic electrons (e-BAE), are observed. The results for the study of fuelling efficiency and penetration characteristics of supersonic molecular beam injection (SMBI) are described.

  2. Edge Turbulence Imaging in the Alcator C-Mod Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S.J. Zweben; D.P. Stotler; J.L. Terry; B. LaBombard; M. Greenwald; M. Muterspaugh; C.S. Pitcher; the Alcator C-Mod Group; K. Hallatschek; R.J. Maqueda; B. Rogers; J.L. Lowrance; V.J. Mastrocola; G.F. Renda

    2001-11-26

    The 2-D radial vs. poloidal structure of edge turbulence in the Alcator C-Mod tokamak [I.H. Hutchinson, R. Boivin, P.T. Bonoli et al., Nuclear Fusion 41(2001) 1391] was measured using fast cameras and compared with 3-D numerical simulations of edge plasma turbulence. The main diagnostic is Gas Puff Imaging (GPI), in which the visible D(subscript alpha) emission from a localized D(subscript 2) gas puff is viewed along a local magnetic field line. The observed D(subscript alpha) fluctuations have a typical radial and poloidal scale of approximately 1 cm, and often have strong local maxima (''blobs'') in the scrape-off layer. The motion of this 2-D structure motion has also been measured using an ultra-fast framing camera with 12 frames taken at 250,000 frames/sec. Numerical simulations produce turbulent structures with roughly similar spatial and temporal scales and transport levels as that observed in the experiment; however, some differences are also noted, perhaps requiring diagnostic improvement and/or additional physics in the numerical model.

  3. Development of a helium-beam diagnostic for the measurement of the electron density and temperature with high space and time resolution; Entwicklung einer Heliumstrahldiagnostik zur Messung der Elektronendichte und -temperatur mit hoher raeumlicher und zeitlicher Aufloesung

    Energy Technology Data Exchange (ETDEWEB)

    Kruezi, U.

    2006-11-15

    A cvoncept for the control of teh particle and energy removal is available with the Dynamic Ergodic Divertor (DED) at the TEXTOR tokamak and is studied there. In the framework of this thesis a new diagnostic fot the study of short-time events in the plasma boundary layer was developed and constructed. It allows spatially (2 mm) and timely (10 {mu}s) highly resolved measurements of the electron density n{sub e} and electron temperaturew T{sub e}. This occurs by spectroscopy on helium atoms injected into the plasma, for whose measured line intensities respectively intensity ratios by means of a collision-radiation model n{sub e} and T{sub e} can be determined. In order to fulfil the requirements for the measurement of the plasma fluctuations up to 100 kHz, an injection system was developed, which can produce a supersonic helium beam of high particle density (1.5.10{sup 18} m{sup -3}) and simulataneously low deivergence {+-}1 . Parallely for this an observation system consisting of many-channel photomultipliers (PMT) with high and a CCD camera with lower time resolution. The signals of the different MT channels are calibrated on the intensities of the comparable spatial channels of the CCD camera. The first spectroscopic measurement of T{sub e} fluctuations resulted for the characterizing parameters: velocity v{sub r}=(380{+-}60) m/s, correlation length L{sub r}{approx}(5{+-}1) mm, and lifetime {tau}{sub L}{approx}(10{+-}1.25) {mu}s. Under the influence of resonant disturbing magnetic fields by the DED because of the not negligible photon noise no quantitative fluctuation characteristics could be determined. Furthermore during the dynamic AC operation of the DED with rotating disturbing field (974 Hz) n{sub e} and T{sub e} could be spatially and timely resolved and showed because of dynamically co-moved plasma structures a strong modulation by a factor 3 respectively 2. Beside an expected pressure decreasement in the laminar flux tube a hitherto unknown increasement

  4. Vertical compact torus injection into the STOR-M tokamak

    Science.gov (United States)

    Liu, Dazhi

    Central fuelling is a fundamental issue in the next generation tokamak-ITER (International Thermonuclear Experimental Reactor). It is essential for optimization of the bootstrap current which is proportional to the pressure gradient of trapped particles. The conventional fusion reactor fuelling techniques, such as gas puffing and cryogenic pellet injection, are considered inadequate to fulfill this goal due to premature ionization caused by high plasma temperature and density. Compact Torus (CT) injection is a promising fuelling technique for central fuelling a reactor-grade tokamak. An accelerated CT is expected to penetrate into the core region and deposit fuel there provided the CT kinetic energy density exceeds the magnetic energy density in a target plasma. This process is complicated and involves CT penetration into an external magnetic field, a CT stopping mechanism, magnetic reconnection, and excitation of plasma waves. CTs can be injected at different angles with respect to the tokamak toroidal magnetic field, either horizontally or vertically. Normally, CTs are injected radially in the mid-plane of a tokamak. In this configuration, CTs will undergo a decelerating force due to the gradient of the tokamak toroidal magnetic field. CTs will stop inside the tokamak chamber or bunce back depending on the relation between kinetic energy density of injected CTs and the tokamak toroidal magnetic field energy density. In the case of vertical injection, deeper penetration is expected due to the absence of the gradient of the tokamak toroidal field in that direction. Experimental investigations on vertical CT injection into a tokamak will be of great significance. The aim of this thesis is to experimentally investigate the feasibility of vertical CT injection into a tokamak and effects of CTs on tokamak plasma confinements. The Saskatchewan Torus-Modified (STOR-M) tokamak is currently the only tokamak equipped with a CT injector in the world. Vertical CT injection

  5. Tritium Removal by Laser Heating and Its Application to Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    C.H. Skinner; C.A. Gentile; G. Guttadora; A. Carpe; S. Langish; K.M. Young; M. Nishi; W. Shu

    2001-11-16

    A novel laser heating technique has recently been applied to removing tritium from carbon tiles that had been exposed to deuterium-tritium (DT) plasmas in the Tokamak Test Fusion Reactor (TFTR). A continuous wave neodymium laser, of power up to 300 watts, was used to heat the surface of the tiles. The beam was focused to an intensity, typically 8 kW/cm{sup 2}, and rapidly scanned over the tile surface by galvanometer-driven scanning mirrors. Under the laser irradiation, the surface temperature increased dramatically, and temperatures up to 2,300 degrees C were recorded by an optical pyrometer. Tritium was released and circulated in a closed-loop system to an ionization chamber that measured the tritium concentration. Most of the tritium (up to 84%) could be released by the laser scan. This technique appears promising for tritium removal in a next-step DT device as it avoids oxidation, the associated deconditioning of the plasma facing surfaces, and the expense of processing large quantities of tritium oxide. Some engineering aspects of the implementation of this method in a next-step fusion device will be discussed.

  6. Predicting high harmonic ion cyclotron heating efficiency in Tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Green, David L [ORNL; Jaeger, E. F. [XCEL; Berry, Lee A [ORNL; Chen, Guangye [ORNL; Ryan, Philip Michael [ORNL; Canik, John [ORNL

    2011-01-01

    Observations of improved radio frequency (RF) heating efficiency in high-confinement (H-) mode plasmas on the National Spherical Tokamak Experiment (NSTX) are investigated by whole-device linear simulation. We present the first full-wave simulation to couple kinetic physics of the well confined core plasma to the poorly confined scrape-off plasma. The new simulation is used to scan the launched fast-wave spectrum and examine the steady-state electric wave field structure for experimental scenarios corresponding to both reduced, and improved RF heating efficiency. We find that launching toroidal wave-numbers that required for fast-wave propagation excites large amplitude (kVm 1 ) coaxial standing modes in the wave electric field between the confined plasma density pedestal and conducting vessel wall. Qualitative comparison with measurements of the stored plasma energy suggest these modes are a probable cause of degraded heating efficiency. Also, the H-mode density pedestal and fast-wave cutoff within the confined plasma allow for the excitation of whispering gallery type eigenmodes localised to the plasma edge.

  7. Control of MHD instabilities in the STOR-M tokamak

    Science.gov (United States)

    Xiao, Chijin; Elgriw, Sayf; Hirose, Akira; STOR-M Team

    2011-10-01

    Experiments to control the MHD activities have been carried out through compact torus injection (CTI) and resonant helical coils (RHC) on the STOR-M tokamak. The MHD instabilities have been measured by Mirnov coil arrays and miniature soft X-ray (SXR) pin-hole cameras. The data have been analyzed by singular value decomposition algorithm and the spatial Fourier harmonic analysis. Injection of a high density compact torus into STOR-M induced a transient phase with reduced m = 2 Mirnov oscillation amplitude. After appearance of an m = 1 gong mode burst the m = 2 oscillation amplitude returned to its nominal level before CTI. In the RHC experiments, an m = 2 helical coil was wound outside the vacuum chamber and powered by a capacitor bank through an IGBT switch. A current pulse of a few milliseconds was applied to RHC during the plasma current plateau. Once the current amplitude reaches a threshold level, the m = 2 MHD oscillation level was significantly reduced. Addition of equilibrium poloidal magnetic field calculated by TOSCA code, an assumed magnetic island perturbation, and the vacuum magnetic field produced by RHC also showed that the island can be eliminated when the RHC current reached a certain level. NSERC and the Canada Research Chair Program

  8. Progress of Thomson scattering diagnostic on HL-2A tokamak

    Science.gov (United States)

    Feng, Z.; Wang, Y. Q.; Hou, Z. P.; Ren, L. L.; Liu, C. H.; Luo, C. W.; Huang, Y.

    2017-11-01

    Some efforts have been made to promote the performance of incoherent Thomson scattering (TS) diagnostic on HL-2A tokamak. Motorized stages are used to adjust the reflecting mirrors and focusing lens of the input laser beam optics, by which it is easy to control the laser beam pass through the narrow throats of the lower and upper closed divertors. Spectral calibration has been refined. Hardware of Si-APD detector electronics is improved, which provides two output signal channels. In one channel, only the rapid TS signal is output after deducting the influence of plasma light. In the other, both the rapid TS signal and the background signal of slow-varying plasma light are output. In this 2017 experiment campaign, the new developed electronics are tested and TS signals can be obtained from the two channels, which are digitized by 1GS-12bit transient recorders. In data processing, the TS pulse shape is fitted with different functions output from the two different channels. The statistical estimation of Te data is also optimized. More channels of high-speed digitizers and more positions of Te and ne measurement are planned and are in constructions.

  9. Basic Physics of Tokamak Transport Final Technical Report.

    Energy Technology Data Exchange (ETDEWEB)

    Sen, Amiya K.

    2014-05-12

    The goal of this grant has been to study the basic physics of various sources of anomalous transport in tokamaks. Anomalous transport in tokamaks continues to be one of the major problems in magnetic fusion research. As a tokamak is not a physics device by design, direct experimental observation and identification of the instabilities responsible for transport, as well as physics studies of the transport in tokamaks, have been difficult and of limited value. It is noted that direct experimental observation, identification and physics study of microinstabilities including ITG, ETG, and trapped electron/ion modes in tokamaks has been very difficult and nearly impossible. The primary reasons are co-existence of many instabilities, their broadband fluctuation spectra, lack of flexibility for parameter scans and absence of good local diagnostics. This has motivated us to study the suspected tokamak instabilities and their transport consequences in a simpler, steady state Columbia Linear Machine (CLM) with collisionless plasma and the flexibility of wide parameter variations. Earlier work as part of this grant was focused on both ITG turbulence, widely believed to be a primary source of ion thermal transport in tokamaks, and the effects of isotope scaling on transport levels. Prior work from our research team has produced and definitively identified both the slab and toroidal branches of this instability and determined the physics criteria for their existence. All the experimentally observed linear physics corroborate well with theoretical predictions. However, one of the large areas of research dealt with turbulent transport results that indicate some significant differences between our experimental results and most theoretical predictions. Latter years of this proposal were focused on anomalous electron transport with a special focus on ETG. There are several advanced tokamak scenarios with internal transport barriers (ITB), when the ion transport is reduced to

  10. Power supplies and quench protection for the Tokamak Physics Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Neumeyer, C.L. [Raytheon Engineers & Constructors, Princeton, NJ (United States). EBASCO Div.

    1994-07-01

    The Tokamak Physics Experiment (TPX) is an advanced tokamak project aimed at the production of quasi-steady state plasmas with advanced shape, heating, and particle control. TPX is to be built at the Princeton Plasma Physics Laboratory (PPPL) using many of the facilities from the Tokamak Fusion Test Reactor (TFTR). First plasma is scheduled for the year 2000. TPX will be the first tokamak to utilize superconducting (SC) magnets in both the toroidal field (TF) and poloidal field (PF) systems. This is a new feature which requires not only a departure from the traditional tokamak power supply schemes but also that ultra-reliable quench protection devices be used to rapidly discharge the stored energy from the magnets in the event of a quench. This paper describes the plan and basis for the adaptation and augmentation of the PPPL/TFTR power system facilities to supply TPX. Following a description of the basic operational requirements, four major areas are addressed, namely the AC power system, the TF power supply, the PF power supply, and quench protection for the TF and PF systems.

  11. Aspects of Tokamak toroidal magnet protection

    Energy Technology Data Exchange (ETDEWEB)

    Green, R.W.; Kazimi, M.S.

    1979-07-01

    Simple but conservative geometric models are used to estimate the potential for damage to a Tokamak reactor inner wall and blanket due to a toroidal magnet field collapse. The only potential hazard found to exist is due to the MHD pressure rise in a lithium blanket. A survey is made of proposed protection methods for superconducting toroidal magnets. It is found that the two general classifications of protection methods are thermal and electrical. Computer programs were developed which allow the toroidal magnet set to be modeled as a set of circular filaments. A simple thermal model of the conductor was used which allows heat transfer to the magnet structure and which includes the effect of temperature dependent properties. To be effective in large magnets an electrical protection system should remove at least 50% of the stored energy in the protection circuit assuming that all of the superconductor in the circuit quenches when the circuit is activated. A protection system design procedure based on this criterion was developed.

  12. Tokamak blanket design study, final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

    A cylindrical module concept was developed, analyzed, and incorporated in a tokamak blanket system that includes piping systems, vacuum boundary sealing, and support structures. The design is based on the use of state-of-the-art structural materials (20% cold-worked type 316 stainless steel), lithium as the breeding material, and pressurized helium as the coolant. The module design consists of nested concentric cylinders (with an outer diameter of 10 cm) and features direct wall cooling by helium flowing between the outer (first-wall) cylinder and the inner (lithium-containing) cylinder. Each cylinder can withstand full coolant pressure, thus enhancing reliability. Results show that stainless steel is a viable material for a first wall subjected to a neutron wall loading of 4 MW/m/sup 2/ and a particle heat flux of 1 MW/m/sup 2/. Lifetime analysis shows that the first-wall design meets the goal of operating at 20-min cycles with 95% duty for 100,000 cycles. To reduce system complexity, a larger 20-cm-diam module also was analyzed for incorporation in the blanket assembly. Reliability assessment indicates that it may be possible to double the module in size from 10 to 20 cm in diameter. With a modest increase in coolant pumping power, a blanket assembly comprising 20-cm-diam modules can still achieve 100,000 operating cycles - equivalent to a 3.6-year design lifetime - with only one or two helium coolant leaks into the plasma.

  13. Lower hybrid current drive in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Ushigusa, Kenkichi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1999-03-01

    Past ten years progress on Lower Hybrid Current Drive (LHCD) experiments have demonstrated the largest non-inductive current (3.6 MA, JT-60U), the longest current sustainment (2 hours, TRIAM-1M), non-inductive current drive at the highest density (n-bar{sub e} - 10{sup 20}m{sup -3}, ALCATOR-C) and the highest current drive efficiency ({eta}{sub CD} = 3.5x10{sup 19} m{sup -2}A/W, JT-60). These results indicate that LHCD is one of the most promising methods to drive non-inductive current in the present tokamak plasmas. This paper presents recent experimental results on LHCD experiments. Basic theories of LH waves, the wave propagation and the current drive are briefly summarized. The main part of this paper describes several important results and their physical pictures on recent LHCD experiments; 1) the experimental set-up, 2) the current drive efficiency, 3) the control of current profile and MHD activities, 4) the global energy confinement, 5) the global power flow, 6) fast electron behavior, 7) interaction between LH waves and thermal/fast ions, 8) combination with other CD method. (author)

  14. Fast scanning probe for tokamak plasmas

    Science.gov (United States)

    Boedo, J.; Gray, D.; Chousal, L.; Conn, R.; Hiller, B.; Finken, K. H.

    1998-07-01

    We describe a fast reciprocating probe drive, which has three main new features: (1) a detachable and modular probe head for easy maintenance, (2) a combination of high heat flux capability, high bandwidth, and low-Z materials construction, and (3) low weight, compact, inexpensive construction. The probe is mounted in a fast pneumatic drive in order to reach plasma regions of interest and remain inserted long enough to obtain good statistics while minimizing the heat flux to the tips and head. The drive is pneumatic and has been designed to be compact and reliable to comply with space and maintenance requirements of tokamaks. The probe described here has five tips which obtain a full spectrum of plasma parameters: electron temperature profile Te(r), electron density profile ne(r), floating potential profile Vf(r), poloidal electric field profile Eθ(r), saturation current profile Isat(r), and their fluctuations up to 3 MHz. We describe the probe show radial profiles of various parameters. We compare the density and temperature data to that obtained with a helium beam. We also discuss the techniques to process the data optimally, particularly double probe data and profile fits.

  15. Modelling and control of a tokamak plasma; Modelisation et commande d`un plasma de tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bremond, S.

    1995-10-18

    Vertically elongated tokamak plasmas, while attractive as regards Lawson criteria, are intrinsically instable. It is found that the open-loop instability dynamics is characterised by the relative value of two dimensionless parameters: the coefficient of inductive coupling between the vessel and the coils, and the coil damping efficiency on the plasma displacement relative to that of the vessel. Applications to Tore Supra -where the instability is due to the iron core attraction- and DIII-D are given. A counter-effect of the vessel, which temporarily reverses the effect of coil control on the plasma displacement, is seen when the inductive coupling is higher than the damping ratio. Precise control of the plasma boundary is necessary if plasma-wall interaction and/or coupling to heating antennas are to be monitored. A positional drift, of a few mm/s, which had been observed in the Tore Supra tokamak, is explained and corrected. A linear plasma shape response model is then derived from magnetohydrodynamic equilibrium calculation, and proved to be in good agreement with experimental data. An optimal control law is derived, which minimizes an integral quadratic criteria on tracking errors and energy expenditure. This scheme avoids compensating coil currents, and could render local plasma shaping more precise. (authors). 123 refs., 77 figs., 6 tabs., 4 annexes.

  16. Prospects for pilot plants based on the tokamak, spherical tokamak and stellarator

    Science.gov (United States)

    Menard, J. E.; Bromberg, L.; Brown, T.; Burgess, T.; Dix, D.; El-Guebaly, L.; Gerrity, T.; Goldston, R. J.; Hawryluk, R. J.; Kastner, R.; Kessel, C.; Malang, S.; Minervini, J.; Neilson, G. H.; Neumeyer, C. L.; Prager, S.; Sawan, M.; Sheffield, J.; Sternlieb, A.; Waganer, L.; Whyte, D.; Zarnstorff, M.

    2011-10-01

    A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a single facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.

  17. Combined hydrogen and lithium beam emission spectroscopy observation system for Korea Superconducting Tokamak Advanced Research

    Energy Technology Data Exchange (ETDEWEB)

    Lampert, M. [Wigner RCP, Euratom Association-HAS, Budapest (Hungary); BME NTI, Budapest (Hungary); Anda, G.; Réfy, D.; Zoletnik, S. [Wigner RCP, Euratom Association-HAS, Budapest (Hungary); Czopf, A.; Erdei, G. [Department of Atomic Physics, BME IOP, Budapest (Hungary); Guszejnov, D.; Kovácsik, Á.; Pokol, G. I. [BME NTI, Budapest (Hungary); Nam, Y. U. [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-07-15

    A novel beam emission spectroscopy observation system was designed, built, and installed onto the Korea Superconducting Tokamak Advanced Research tokamak. The system is designed in a way to be capable of measuring beam emission either from a heating deuterium or from a diagnostic lithium beam. The two beams have somewhat complementary capabilities: edge density profile and turbulence measurement with the lithium beam and two dimensional turbulence measurement with the heating beam. Two detectors can be used in parallel: a CMOS camera provides overview of the scene and lithium beam light intensity distribution at maximum few hundred Hz frame rate, while a 4 × 16 pixel avalanche photo-diode (APD) camera gives 500 kHz bandwidth data from a 4 cm × 16 cm region. The optics use direct imaging through lenses and mirrors from the observation window to the detectors, thus avoid the use of costly and inflexible fiber guides. Remotely controlled mechanisms allow adjustment of the APD camera’s measurement location on a shot-to-shot basis, while temperature stabilized filter holders provide selection of either the Doppler shifted deuterium alpha or lithium resonance line. The capabilities of the system are illustrated by measurements of basic plasma turbulence properties.

  18. Investigation of the plasma radiation power in the Globus-M tokamak by means of SPD silicon photodiodes

    Energy Technology Data Exchange (ETDEWEB)

    Iblyaminova, A. D., E-mail: a.iblyaminova@mail.ioffe.ru; Avdeeva, G. F.; Aruev, P. N.; Bakharev, N. N.; Gusev, V. K.; Zabrodsky, V. V.; Kurskiev, G. S.; Minaev, V. B.; Miroshnikov, I. V.; Patrov, M. I.; Petrov, Yu. V.; Sakharov, N. V.; Tolstyakov, S. Yu.; Shchegolev, P. B. [Russian Academy of Sciences, Ioffe Institute (Russian Federation)

    2016-10-15

    Radiation losses from the plasma of the Globus-M tokamak are studied by means of SPD silicon photodiodes developed at the Ioffe Institute, Russian Academy of Sciences. The results from measurements of radiation losses in regimes with ohmic and neutral beam injection heating of plasmas with different isotope compositions are presented. The dependence of the radiation loss power on the plasma current and plasma–wall distance is investigated. The radiation power in different spectral ranges is analyzed by means of an SPD spectrometric module. Results of measurements of radiation losses before and after tokamak vessel boronization are presented. The time evolution of the sensitivity of the SPD photodiode during its two-year exploitation in Globus-M is analyzed.

  19. Plasma Turbulence in the Scrape-off Layer of the ISTTOK Tokamak

    CERN Document Server

    Jorge, Rogerio; Halpern, Federico D; Loureiro, Nuno F; Silva, Carlos

    2016-01-01

    The properties of plasma turbulence in a poloidally limited scrape-off layer (SOL) are addressed, with focus on ISTTOK, a large aspect ratio tokamak with a circular cross section. Theoretical investigations based on the drift-reduced Braginskii equations are carried out through linear calculations and non-linear simulations, in two- and three-dimensional geometries. The linear instabilities driving turbulence and the mechanisms that set the amplitude of turbulence as well as the SOL width are identified. A clear asymmetry is shown to exist between the low-field and the high-field sides of the machine. A comparison between experimental measurements and simulation results is presented.

  20. Density modulation experiment to determine transport coefficients on Joint-TEXT Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Chen, W.; Zhuang, G.; Gao, L., E-mail: gaoli@mail.hust.edu.cn; Chen, J.; Shi, P.; Liu, Y.; Li, Q.; Wang, Z. J. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, School of Electrical and Electronics Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Gentle, K. W. [Institute of Fusion Studies, University of Texas at Austin, Austin, Texas 78712 (United States)

    2015-02-15

    Density modulation experiments have been conducted on Joint-TEXT (J-TEXT) Tokamak Ohmic discharge to investigate particle transport based on a model with constant diffusion plus inward convection. Like the HCN interferometer, the newly developed three-wave polarimeter-interferometer system (POLARIS) is used to measure the perturbed density. The comparison of results between the HCN interferometer and POLARIS is given. The consistent results indicate the validity of the analysis scheme. At lower densities, the typical particle confinement time τ{sub p} is found to increase with electron density, while it saturates at higher densities.

  1. Contribution to the multi-machine pedestal scaling from the COMPASS tokamak

    Science.gov (United States)

    Komm, M.; Bílková, P.; Aftanas, M.; Berta, M.; Böhm, P.; Bogár, O.; Frassinetti, L.; Grover, O.; Háček, P.; Havlicek, J.; Hron, M.; Imríšek, M.; Krbec, J.; Mitošínková, K.; Naydenkova, D.; Pánek, R.; Peterka, M.; Snyder, P. B.; Stefanikova, E.; Stöckel, J.; Sos, M.; Urban, J.; Varju, J.; Vondráček, P.; Weinzettl, V.; the COMPASS Team

    2017-05-01

    First systematic measurements of pedestal structure during Ohmic and NBI-assisted Type I ELMy H-modes were performed on the COMPASS tokamak in two dedicated experimental campaigns during 2015 and 2016. By adjusting the NBI heating and a toroidal magnetic field, the electron pedestal temperature was increased from 200 eV up to 300 eV, which allowed reaching pedestal collisionality ν \\text{ped}\\ast   text{ped}\\ast . The pedestal pressure was successfully reproduced by the EPED model. The dependence of pedestal pressure width on ν \\text{ped}\\ast and β \\text{ped ~ }\\text{pol} is discussed.

  2. Influence of helium puff on divertor asymmetry in experimental advanced superconducting tokamak

    DEFF Research Database (Denmark)

    Liu, S. C.; Guo, H. Y.; Xu, G. S.

    2014-01-01

    Divertor asymmetries with helium puffing are investigated in various divertor configurations on Experimental Advanced Superconducting Tokamak (EAST). The outer divertor electron temperature decreases significantly during the gas injection at the outer midplane. As soon as the gas is injected...... parameters are measured by reciprocating probes at the outer midplane, showing that the electron temperature and density increase but the parallel Mach number decreases significantly due to the gas injection. Effects of poloidal E × B drifts and parallel SOL flows on the divertor asymmetry observed in EAST...

  3. Preliminary conceptual design of a medium sized tokamak (IST-1)

    Science.gov (United States)

    Bagerpour, M.; Alinejad, N.; Sobhanian, S.

    2015-08-01

    In this paper an attempt is made to estimate the main parameters of the Iranian superconducting tokamak as a medium sized tokamak. In the first stage, the production and confinement of ohmically heated plasma is considered. Considering the aim of the design and the kink stability limit, three main parameters are assumed to be known. Using the known theoretical, empirical scale laws and numerical solution of Grad-Shafranov equation for a D-shaped plasmas and also considering the correction terms due to triangularity of the torus cross section, other physical and geometrical parameters have been estimated. The magnetic flux surfaces, plasma pressure and toroidal current density profiles are found by solving of Grad-Shafranov equation as an eigenvalue problem using finite element method. The preliminary results are compared with some recent tokamaks now in operation in different research centers.

  4. Particle transport in tokamak plasmas, theory and experiment

    Energy Technology Data Exchange (ETDEWEB)

    Angioni, C [Max-Planck Institut fuer Plasmaphysik, IPP-EURATOM Association, D-85748 Garching (Germany); Fable, E; Maslov, M; Weisen, H [Centre de Recherches en Physique des Plasmas, Association EURATOM-Confederation Suisse, EPFL, 1015 Lausanne (Switzerland); Greenwald, M [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA (United States); Peeters, A G [Centre for Fusion, Space and Astrophysics, University of Warwick, CV4 7AL, Coventry (United Kingdom); Takenaga, H [Japan Atomic Energy Agency, 801-1 Mukouyama, Naka, Ibaraki 311-0193 (Japan)

    2009-12-15

    The physical processes producing electron particle transport in the core of tokamak plasmas are described. Starting from the gyrokinetic equation, a simple analytical derivation is used as guidance to illustrate the main mechanisms driving turbulent particle convection. A review of the experimental observations on particle transport in tokamaks is presented and the consistency with the theoretical predictions is discussed. An overall qualitative agreement, and in some cases even a specific quantitative agreement, emerges between complex theoretical predictions and equally complex experimental observations, exhibiting different dependences on plasma parameters under different regimes. By these results, the direct connection between macroscopic transport properties and the character of microscopic turbulence is pointed out, and an important confirmation of the paradigm of microinstabilities and turbulence as the main cause of transport in the core of tokamaks is obtained. Finally, the impact of these results on the prediction of the peaking of the electron density profile in a fusion reactor is illustrated.

  5. A control approach for plasma density in tokamak machines

    Energy Technology Data Exchange (ETDEWEB)

    Boncagni, Luca, E-mail: luca.boncagni@enea.it [EURATOM – ENEA Fusion Association, Frascati Research Center, Division of Fusion Physics, Rome, Frascati (Italy); Pucci, Daniele; Piesco, F.; Zarfati, Emanuele [Dipartimento di Ingegneria Informatica, Automatica e Gestionale ' ' Antonio Ruberti' ' , Sapienza Università di Roma (Italy); Mazzitelli, G. [EURATOM – ENEA Fusion Association, Frascati Research Center, Division of Fusion Physics, Rome, Frascati (Italy); Monaco, S. [Dipartimento di Ingegneria Informatica, Automatica e Gestionale ' ' Antonio Ruberti' ' , Sapienza Università di Roma (Italy)

    2013-10-15

    Highlights: •We show a control approach for line plasma density in tokamak. •We show a control approach for pressure in a tokamak chamber. •We show experimental results using one valve. -- Abstract: In tokamak machines, chamber pre-fill is crucial to attain plasma breakdown, while plasma density control is instrumental for several tasks such as machine protection and achievement of desired plasma performances. This paper sets the principles of a new control strategy for attaining both chamber pre-fill and plasma density regulation. Assuming that the actuation mean is a piezoelectric valve driven by a varying voltage, the proposed control laws ensure convergence to reference values of chamber pressure during pre-fill, and of plasma density during plasma discharge. Experimental results at FTU are presented to discuss weaknesses and strengths of the proposed control strategy. The whole system has been implemented by using the MARTe framework [1].

  6. Texas Experimental Tokamak. Technical progress report, April 1990--April 1993

    Energy Technology Data Exchange (ETDEWEB)

    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.

  7. Scrape-off-layer current and EUV diagnostics and control on the HBT-EP tokamak

    Science.gov (United States)

    Levesque, J. P.; Mauel, M. E.; Bialek, J.; Navratil, G. A.; Delgado-Aparicio, L.; Hansen, C. J.

    2015-11-01

    Non-axisymmetric currents in the scrape-off-layer (SOL) and conducting structure of a tokamak can produce severe forces at high plasma performance, compromising the device's structural integrity. Diagnosing these currents during disruptions is important for extrapolating forces in future machines including ITER. Progress on designing components to measure and control SOL and vessel currents in the HBT-EP tokamak is presented. Movable tiles positioned around limiting surfaces will measure SOL and vessel currents during mode activity and disruptions. Biasable plates at divertor strike points will allow control of field-aligned SOL currents for kink mode control studies and will drive convection in the plasma edge. In-vessel Rogowski coils will measure currents in wall components with high spatial resolution. A planned EUV diagnostic upgrade is also presented. Four sets of 16 poloidal views will allow tomographic reconstruction of plasma emissivity and internal kink mode structure. A separate two-color, 16-chord tangential system will allow reconstruction of temperature profiles versus time. Measurements will be input to HBT-EP's GPU-based feedback system, providing active feedback for kink modes using only optical sensors and both magnetic and edge current actuators. Supported by U.S. DOE Grant DE-FG02-86ER53222.

  8. Impurity control in near-term tokamak reactors

    Energy Technology Data Exchange (ETDEWEB)

    Stacey, W.M. Jr.; Smith, D.L.; Brooks, J.N.

    1976-10-01

    Several methods for reducing impurity contamination in near-term tokamak reactors by modifying the first-wall surface with a low-Z or low-sputter material are examined. A review of the sputtering data and an assessment of the technological feasibility of various wall modification schemes are presented. The power performance of a near-term tokamak reactor is simulated for various first-wall surface materials, with and without a divertor, in order to evaluate the likely effect of plasma contamination associated with these surface materials.

  9. Geodesic acoustic modes in noncircular cross section tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Sorokina, E. A., E-mail: sorokina.ekaterina@gmail.com; Lakhin, V. P. [National Research Center “Kurchatov Institute,” (Russian Federation); Konovaltseva, L. V. [People’s Friendship University of Russia (Russian Federation); Ilgisonis, V. I. [National Research Center “Kurchatov Institute,” (Russian Federation)

    2017-03-15

    The influence of the shape of the plasma cross section on the continuous spectrum of geodesic acoustic modes (GAMs) in a tokamak is analyzed in the framework of the MHD model. An expression for the frequency of a local GAM for a model noncircular cross section plasma equilibrium is derived. Amendments to the oscillation frequency due to the plasma elongation and triangularity and finite tokamak aspect ratio are calculated. It is shown that the main factor affecting the GAM spectrum is the plasma elongation, resulting in a significant decrease in the mode frequency.

  10. TIBER: Tokamak Ignition/Burn Experimental Research. Final design report

    Energy Technology Data Exchange (ETDEWEB)

    Henning, C.D.; Logan, B.G.; Barr, W.L.; Bulmer, R.H.; Doggett, J.N.; Johnson, B.M.; Lee, J.D.; Hoard, R.W.; Miller, J.R.; Slack, D.S.

    1985-11-01

    The Tokamak Ignition/Burn Experimental Research (TIBER) device is the smallest superconductivity tokamak designed to date. In the design plasma shaping is used to achieve a high plasma beta. Neutron shielding is minimized to achieve the desired small device size, but the superconducting magnets must be shielded sufficiently to reduce the neutron heat load and the gamma-ray dose to various components of the device. Specifications of the plasma-shaping coil, the shielding, coaling, requirements, and heating modes are given. 61 refs., 92 figs., 30 tabs. (WRF)

  11. Advanced tokamak physics scenarios in Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Porkolab, M.; Bonoli, P.T.; Golovato, S.; Ramos, J.; Sugiyama, L.; Takase, Y. [Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Kessel, C. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Nevins, W.M. [LLNL, Livermore, California 94550 (United States)

    1996-02-01

    Several advanced tokamak modes of operation have been identified in the Alcator C-Mod tokamak. Of particular interest are (i) Reversed shear mode with high bootstrap fraction using on-axis FW current drive and off-axis mode-conversion current drive and/or lower hybrid current drive; (ii) High performance plasmas ({ital Q}{approximately}0.1{endash}1) which may be accessed by the PEP (pellet enhanced performance) mode of operation with intense ICRF heating. {copyright} {ital 1996 American Institute of Physics.}

  12. Study of electron beams within ISTTOK tokamak by means of a multi-channel Cherenkov detector; their correlation with hard X-rays

    Energy Technology Data Exchange (ETDEWEB)

    Jakubowski, L., E-mail: Lech.Jakubowski@ipj.gov.p [Andrzej Soltan Institute for Nuclear Studies, 05-400 Otwock-Swierk (Poland); Malinowski, K.; Sadowski, M.J.; Zebrowski, J. [Andrzej Soltan Institute for Nuclear Studies, 05-400 Otwock-Swierk (Poland); Plyusnin, V.V. [Association Euratom/IST, Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Av. Rovisco Pais, 1049-001 Lisboa (Portugal); Rabinski, M. [Andrzej Soltan Institute for Nuclear Studies, 05-400 Otwock-Swierk (Poland); Fernandes, H.; Silva, C.; Duarte, P. [Association Euratom/IST, Instituto de Plasmas e Fusao Nuclear, Instituto Superior Tecnico, Av. Rovisco Pais, 1049-001 Lisboa (Portugal); Jakubowski, M.J. [Andrzej Soltan Institute for Nuclear Studies, 05-400 Otwock-Swierk (Poland)

    2010-11-11

    The paper describes experimental studies of electron beams emitted from a plasma torus within the ISTTOK tokamak, which were performed by means of a new four-channel detector of the Cherenkov type. A range of electron energy was estimated. There were also measured hard X-rays, and their correlation with the fast run-away electron beams was investigated experimentally.

  13. Parasitic momentum flux in the tokamak core

    Science.gov (United States)

    Stoltzfus-Dueck, T.

    2017-10-01

    Tokamak plasmas rotate spontaneously without applied torque. This intrinsic rotation is important for future low-torque devices such as ITER, since rotation stabilizes certain instabilities. In the mid-radius `gradient region,' which reaches from the sawtooth inversion radius out to the pedestal top, intrinsic rotation profiles may be either flat or hollow, and can transition suddenly between these two states, an unexplained phenomenon referred to as rotation reversal. Theoretical efforts to explain the mid-radius rotation shear have largely focused on quasilinear models, in which the phase relationships of some selected instability result in a nondiffusive momentum flux (``residual stress''). In contrast, the present work demonstrates the existence of a robust, fully nonlinear symmetry-breaking momentum flux that follows from the free-energy flow in phase space and does not depend on any assumed linear eigenmode structure. The physical origin is an often-neglected portion of the radial ExB drift, which is shown to drive a symmetry-breaking outward flux of co-current momentum whenever free energy is transferred from the electrostatic potential to ion parallel flows. The fully nonlinear derivation relies only on conservation properties and symmetry, thus retaining the important contribution of damped modes. The resulting rotation peaking is counter-current and scales as temperature over plasma current. As first demonstrated by Landau, this free-energy transfer (thus also the corresponding residual stress) becomes inactive when frequencies are much higher than the ion transit frequency, which allows sudden transitions between hollow and flat profiles. Simple estimates suggest that this mechanism may be consistent with experimental observations. This work was funded in part by the Max-Planck/Princeton Center for Plasma Physics and in part by the U.S. Dept. of Energy, Office of Science, Contract No. DE-AC02-09CH11466.

  14. Non-perturbative measurement of cross-field thermal diffusivity reduction at the O-point of 2/1 neoclassical tearing mode islands in the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Bardóczi, L.; Rhodes, T. L.; Carter, T. A.; Crocker, N. A.; Peebles, W. A. [University of California Los Angeles, Los Angeles, California 90095 (United States); Grierson, B. A. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

    2016-05-15

    Neoclassical tearing modes (NTMs) often lead to the decrease of plasma performance and can lead to disruptions, which makes them a major impediment in the development of operating scenarios in present toroidal fusion devices. Recent gyrokinetic simulations predict a decrease of plasma turbulence and cross-field transport at the O-point of the islands, which in turn affects the NTM dynamics. In this paper, a heat transport model of magnetic islands employing spatially non-uniform cross-field thermal diffusivity (χ{sub ⊥}) is presented. This model is used to derive χ{sub ⊥} at the O-point from electron temperature data measured across 2/1 NTM islands in DIII-D. It was found that χ{sub ⊥} at the O-point is 1 to 2 orders of magnitude smaller than the background plasma transport, in qualitative agreement with gyrokinetic predictions. As the anomalously large values of χ{sub ⊥} are often attributed to turbulence driven transport, the reduction of the O-point χ{sub ⊥} is consistent with turbulence reduction found in recent experiments. Finally, the implication of reduced χ{sub ⊥} at the O-point on NTM dynamics was investigated using the modified Rutherford equation that predicts a significant effect of reduced χ{sub ⊥} at the O-point on NTM saturation.

  15. Data processing system for spectroscopy at Novillo Tokamak; Sistema de procesamiento de datos para espectroscopia en el Tokamak Novillo

    Energy Technology Data Exchange (ETDEWEB)

    Ortega C, G.; Gaytan G, E. [Instituto Tecnologico de Toluca, Instituto nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1998-07-01

    Taking as basis some proposed methodologies by software engineering it was designed and developed a data processing system coming from the diagnostic equipment by spectroscopy, for the study of plasma impurities, during the cleaning discharges. the data acquisition is realized through an electronic interface which communicates the computer with the spectroscopy system of Novillo Tokamak. The data were obtained starting from files type text and processed for their subsequently graphic presentation. For development of this system named PRODATN (Processing of Data for Spectroscopy in Novillo Tokamak) was used the LabVIEW graphic programming language. (Author)

  16. Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Federici, G.; Skinner, C.H.; Brooks, J.N.; Coad, J.P.; Grisolia, C. [and others

    2001-01-10

    The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D [Research and Development] avenues for their resolution are presented.

  17. Ion cyclotron emission in tokamak plasmas; Emission cyclotronique ionique dans les plasmas de tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Fraboulet, D.

    1996-09-17

    Detection of {alpha}(3.5 MeV) fusion products will be of major importance for the achievement of self sustained discharges in fusion thermonuclear reactors. Due to their cyclotronic gyration in the confining magnetic field of a tokamak, {alpha} particles are suspected to radiate in the radio-frequency band [RF: 10-500 MHz]. Our aim is to determine whether detection of RF emission radiated from a reactor plasma can provide information concerning those fusion products. We observed experimentally that the RF emission radiated from fast ions situated in the core of the discharge is detectable with a probe located at the plasma edge. For that purpose, fast temporal acquisition of spectral power was achieved in a narrow frequency band. We also propose two complementary models for this emission. In the first one, we describe locally the energy transfer between the photon population and the plasma and we compute the radiation equilibrium taking place in the tokamak. {alpha} particles are not the unique species involved in the equilibrium and it is necessary to take into account all other species present in the plasma (Deuterium, Tritium, electrons,...). Our second model consists in the numerical resolution of the Maxwell-Vlasov with the use of a variational formulation, in which all polarizations are considered and the 4 first cyclotronic harmonics are included in a 1-D slab geometry. The development of this second model leads to the proposal for an experimental set up aiming to the feasibility demonstration of a routine diagnostic providing the central {alpha} density in a reactor. (author). 166 refs.

  18. Multi-channel control circuit for real-time control of events in Aditya tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Edappala, Praveenlal, E-mail: praveen@ipr.res.in; Shah, Minsha; Rajpal, Rachana; Tanna, R.L.; Ghosh, Joydeep; Chattopadhyay, P.K.; Jha, R.

    2016-11-15

    Highlights: • Low cost microcontroller based control circuit. • The control hardware can be programmed/configured very easily for different applications. • Microcontroller programming is done in assembly language so that precise timing can be achieved with micro seconds resolution. • Successful implementation of this circuit in noisy tokamak environment. • Efficient noise and burst elimination. • Can be integrated in to the other subsystems. • Low cost solution for implementing feedback control in small and medium size tokamaks and other experiments requiring feedback control. - Abstract: Tokamak plasma is prone to many random events having potential for causing severe damages to the machine, such as disruptions, production and elimination of high-energy runaway electrons etc. These events can be mitigated by obtaining pre-cursor signal leading to these events and then taking proper measures just before their onset to avoid their happenings, like disruptions can be mitigated by massive gas injection or putting a bias voltage on an electrode placed inside the plasma, the runaways can be mitigated by gas injection and by applying specific magnetic fields. Hence for real time control of these events, the pre-cursors should be electronically recorded and the mitigation techniques should be initiated by sending triggers to their individual operational systems. To implement these methodologies of real-time controlling of events in Aditya Tokamak, a low cost multi-channel Micro-Controller based timing circuit is designed and developed in-house. This circuit first compares the precursor signals fed into it with the pre-set values and gives a trigger output whenever the signals overshoot the pre-set values. The circuit readies itself for operation along with start of the tokamak discharge and waits up to an initial pre-determined delay and then initiates a trigger at the time of overshooting of precursor signal. The circuit is fully integrated and assembled in

  19. Analysis of tokamak plasma confinement modes using the fast ...

    Indian Academy of Sciences (India)

    2016-10-20

    Oct 20, 2016 ... absence of the outer field, and then compared with each other. The number of plasma modes and the safety factor q were determined using the FFT method in the presence and absence of the outer field. The safety factor q plays a significant role in determin- ing the stability of tokamak plasma and seems to.

  20. Bulk Ion Heating with ICRF Waves in Tokamaks

    DEFF Research Database (Denmark)

    Mantsinen, M. J.; Bilato, R.; Bobkov, V. V.

    2015-01-01

    Heating with ICRF waves is a well-established method on present-day tokamaks and one of the heating systems foreseen for ITER. However, further work is still needed to test and optimize its performance in fusion devices with metallic high-Z plasma facing components (PFCs) in preparation of ITER a...

  1. High-pressure, flux-conserving tokamak equilibria

    Energy Technology Data Exchange (ETDEWEB)

    Dory, R.A.; Peng, Y.K.M.

    1976-08-01

    Magnetohydrodynamic (MHD) tokamak equilibria are found with values of ..beta.. up to 20 percent and prescribed MHD safety factor values (e.g., q(axis) = 1 and q(edge) = 4.8) for tokamaks with aspect ratio A = 4 and D-shaped cross section. If such equilibria could be attained experimentally, they would be very attractive for decreasing the projected costs of tokamak power reactors substantially. In the flux-conserving tokamak (FCT) model, where rapid heating is applied to an already relatively hot plasma, these high ..beta.. equilibria are achievable. We study the quasi-static evolution of FCT equilibria as ..beta.. increases. An operating window is found in the pressure profile width w/sub p/: for high ..beta.. the values of w/sub p/ must lie between 0.40 and 0.55 of the plasma minor width. Within this window, plasma current and poloidal ..beta.. increase monotonically with ..beta... For fixed plasma boundary, significant poloidal surface currents are induced, but these can be eliminated by small increases in the plasma minor radius, the pressure profile width, and the vacuum toroidal field.

  2. Conceptual design of Remote Control System for EAST tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Sun, X.Y., E-mail: xysun@ipp.ac.cn; Wang, F.; Wang, Y.; Li, S.

    2014-05-15

    Highlights: • A new design conception for remote control for EAST tokamak is proposed. • Rich Internet application (RIA) was selected to implement the user interface. • Some security mechanism was used to fulfill security requirement. - Abstract: The international collaboration becomes popular in tokamak research like in many other fields of science, because the experiment facilities become larger and more expensive. The traditional On-site collaboration Model that has to spend much money and time on international travel is not fit for the more frequent international collaboration. The Remote Control System (RCS), as an extension of the Central Control System for the EAST tokamak, is designed to provide an efficient and economical way to international collaboration. As a remote user interface, the RCS must integrate with the Central Control System for EAST tokamak to perform discharge control function. This paper presents a design concept delineating a few key technical issues and addressing all significant details in the system architecture design. With the aim of satisfying system requirements, the RCS will select rich Internet application (RIA) as a user interface, Java as a back-end service and Secure Socket Layer Virtual Private Network (SSL VPN) for securable Internet communication.

  3. Evidence of Inward Toroidal Momentum Convection in the JET Tokamak

    DEFF Research Database (Denmark)

    Tala, T.; Zastrow, K.-D.; Ferreira, J.

    2009-01-01

    Experiments have been carried out on the Joint European Torus tokamak to determine the diffusive and convective momentum transport. Torque, injected by neutral beams, was modulated to create a periodic perturbation in the toroidal rotation velocity. Novel transport analysis shows the magnitude...

  4. Sensitivity of transient synchrotron radiation to tokamak plasma parameters

    Energy Technology Data Exchange (ETDEWEB)

    Fisch, N.J.; Kritz, A.H.

    1988-12-01

    Synchrotron radiation from a hot plasma can inform on certain plasma parameters. The dependence on plasma parameters is particularly sensitive for the transient radiation response to a brief, deliberate, perturbation of hot plasma electrons. We investigate how such a radiation response can be used to diagnose a variety of plasma parameters in a tokamak. 18 refs., 13 figs.

  5. Recording non-local temperature rise in the RTP tokamak

    NARCIS (Netherlands)

    Hogeweij, G. M. D.; Mantica, P.; Gorini, G.; de Kloe, J.; Cardozo, N. J. L.; R. T. P. Team,

    2000-01-01

    In the Rijnhuizen Tokamak Project (RTP) plasmas with electron cyclotron heating (ECH), a transient rise of the core T-e is observed when hydrogen pellets are injected tangentially to induce fast cooling of the peripheral region. The core T-e rise is a sharp function of the normalized power

  6. Stability of localized modes in rotating tokamak plasmas

    NARCIS (Netherlands)

    J.W. Haverkort (Willem); H.J. de Blank

    2011-01-01

    textabstractThe ideal magnetohydrodynamic stability is investigated of localized interchange modes in a large-aspect ratio tokamak plasma. The resulting stability criterion includes the effects of toroidal rotation and rotation shear and contains various well-known limiting cases. The analysis

  7. Design of tangential multi-energy SXR cameras for tokamak plasmas

    Science.gov (United States)

    Yamazaki, H.; Delgado-Aparicio, L. F.; Pablant, N.; Hill, K.; Bitter, M.; Takase, Y.; Ono, M.; Stratton, B.

    2017-10-01

    A new synthetic diagnostic capability has been built to study the response of tangential multi-energy soft x-ray pin-hole cameras for arbitrary plasma densities (ne , D), temperature (Te) and ion concentrations (nZ). For tokamaks and future facilities to operate safely in a high-pressure long-pulse discharge, it is imperative to address key issues associated with impurity sources, core transport and high-Z impurity accumulation. Multi-energy soft xray imaging provides a unique opportunity for measuring, simultaneously, a variety of important plasma properties (e.g. Te, nZ and ΔZeff). These systems are designed to sample the continuum- and line-emission from low- to high-Z impurities (e.g. C, O, Al, Si, Ar, Ca, Fe, Ni and Mo) in multiple energy-ranges. These x-ray cameras will be installed in the MST-RFP, as well as NSTX-U and DIII-D tokamaks, measuring the radial structure of the photon emissivity with a radial resolution below 1 cm at a 500 Hz frame rate and a photon-energy resolution of 500 eV. The layout and response expected for the new systems will be shown for different plasma conditions and impurity concentrations. The effect of toroidal rotation driving poloidal asymmetries in the core radiation is also addressed for the case of NSTX-U.

  8. Fuelling and plasma flow change by compact torus injection into the STOR-M Tokamak

    Science.gov (United States)

    Onchi, Takumi; Liu, Yelu; Dreval, Mykola; McColl, David; Xiao, Chijin; Hirose, Akira; Asai, Tomohiko; Wolfe, Sean

    2012-10-01

    The Saskatchewan TORus Modified (STOR-M) tokamak is equipped with a Compact Torus (CT) injector for tangential (toroidal) injection of a high density plasmoid at a velocity of 150 km/s. The objectives of CT injection (CTI) are to fuel the core region of tokamak and optimize the bootstrap current in future reactors by control of the plasma pressure gradient. After CTI, the line averaged density along central chord increases from ne˜x 10^12 to 1.5 x 10^13 [cm-3]. Measurement of soft X-ray bremsstrahlung emission profile indicates a steeper density gradient is generated after the asymmetric density profile is formed and the profile become symmetry again in STOR-M. Intrinsic impurity ion flows have been measured with ion Doppler spectroscopy. Significant radial velocity shear from center to edge region is observed even in Ohmic discharges. The toroidal flow direction is found to depend on the plasma current direction. CTI also modifies toroidal plasma flow. The edge plasma flow increases by 5 km/s 1millisecond after CTI. During these milliseconds of time, toroidal flow shear is also increased from 214.3 to 285.7 [10^3 x1/s]. A few milliseconds later than that time, plasma flow slows down, but plasma confinement is improved. Hα emission decreases by 50%.

  9. Data Acquisition and Automation for Plasma Rotation Diagnostic in the TCABR Tokamak

    Science.gov (United States)

    Ronchi, G.; Severo, J. H. F.; de Sá, W. P.; Galvão, R. M. O.

    2015-03-01

    In this work we describe the implementation of a full modular system of data acquisition and processing for the plasma rotation diagnostic in the TCABR tokamak. The experimental setup uses a single monochromator and six photomultipliers (PMT), in which pair of PMTs measures the light at slightly different wavelengths. Thus, it can measure the time evolution of the Doppler shift of the impurities emission lines coming from three spatial positions (one for toroidal rotation and two for poloidal rotation). The data acquisition and preanalysis program were written with LabVIEW software and is capable of controlling the spectrometer wavelength, PMTs power supplies, data acquisition, and storage. All data are recorded in MDSplus trees that easily allow data visualization and post-processing analysis (both locally and remotely) via MATLAB, Python, Java and others programming languages. This system can run independently from other diagnostics and machine systems and can be integrated with the main tokamak control system by means of TCP/IP messages.

  10. Nonlinear effects of energetic particle driven instabilities in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Bruedgam, Michael

    2010-03-25

    In a tokamak plasma, a population of superthermal particles generated by heating methods can lead to a destabilization of various MHD modes. Due to nonlinear wave-particle interactions, a consequential fast particle redistribution reduces the plasma heating and can cause severe damages to the wall of the fusion device. In order to describe the wave-particle interaction, the drift-kinetic perturbative HAGIS code is applied which evolves the particle trajectories and the waves nonlinearly. For a simulation speed-up, the 6-d particle phase-space is reduced by the guiding centre approach to a 5-d description. The eigenfunction of the wave is assumed to be invariant, but its amplitude and phase is altered in time. A sophisticated {delta}/f-method is employed to model the change in the fast particle distribution so that numerical noise and the excessive number of simulated Monte-Carlo points are reduced significantly. The original code can only calculate the particle redistribution inside the plasma region. Therefore, a code extension has been developed during this thesis which enlarges the simulation region up to the vessel wall. By means of numerical simulations, this thesis addresses the problem of nonlinear waveparticle interactions in the presence of multiple MHD modes with significantly different eigenfrequencies and the corresponding fast particle transport inside the plasma. In this context, a new coupling mechanism between resonant particles and waves has been identified that leads to enhanced mode amplitudes and fast particle losses. The extension of the code provides for the first time the possibility of a quantitative and qualitative comparison between simulation results and recent measurements in the experiment. The findings of the comparison serve as a validation of both the theoretical model and the interpretation of the experimental results. Thus, a powerful interface tool has been developed for a deeper insight of nonlinear wave-particle interaction

  11. Tokamak Plasmas: Measurement of temperature fluctuations and ...

    Indian Academy of Sciences (India)

    Keywords. Temperature fluctuations; anomalous transport; plasma rotation. ... S K Saha1. Plasma Physics Division, Saha Institute of Nuclear Physics, 1/AF Bidhannagar, Calcutta 700 064, India ... Proceedings of the International Workshop/Conference on Computational Condensed Matter Physics and Materials Science

  12. Development of fast video recording of plasma interaction with a lithium limiter on T-11M tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lazarev, V.B., E-mail: v_lazarev@triniti.ru [SSC RF TRINITI Troitsk, Moscow (Russian Federation); Dzhurik, A.S.; Shcherbak, A.N. [SSC RF TRINITI Troitsk, Moscow (Russian Federation); Belov, A.M. [NRC “Kurchatov Institute”, Moscow (Russian Federation)

    2016-11-15

    Highlights: • The paper presents the results of the study of tokamak plasma interaction with lithium capillary-porous system limiters and PFC by high-speed color camera. • Registration of emission near the target in SOL in neutral lithium light and e-folding length for neutral Lithium measurements. • Registration of effect of MHD instabilities on CPS Lithium limiter. • A sequence of frames shows evolution of lithium bubble on the surface of lithium limiter. • View of filament structure near the plasma edge in ohmic mode. - Abstract: A new high-speed color camera with interference filters was installed for fast video recording of plasma-surface interaction with a Lithium limiter on the base of capillary-porous system (CPS) in T-11M tokamak vessel. The paper presents the results of the study of tokamak plasma interaction (frame exposure time up to 4 μs) with CPS Lithium limiter in a stable stationary phase, unstable regimes with internal disruption and results of processing of the image of the light emission around the probe, i.e. e-folding length for neutral Lithium penetration and e-folding length for Lithium ion flux in SOL region.

  13. 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....

  14. Maximum entropy reconstruction of poloidal magnetic field and radial electric field profiles in tokamaks

    Science.gov (United States)

    Chen, Yihang; Xiao, Chijie; Yang, Xiaoyi; Wang, Tianbo; Xu, Tianchao; Yu, Yi; Xu, Min; Wang, Long; Lin, Chen; Wang, Xiaogang

    2017-10-01

    The Laser-driven Ion beam trace probe (LITP) is a new diagnostic method for measuring poloidal magnetic field (Bp) and radial electric field (Er) in tokamaks. LITP injects a laser-driven ion beam into the tokamak, and Bp and Er profiles can be reconstructed using tomography methods. A reconstruction code has been developed to validate the LITP theory, and both 2D reconstruction of Bp and simultaneous reconstruction of Bp and Er have been attained. To reconstruct from experimental data with noise, Maximum Entropy and Gaussian-Bayesian tomography methods were applied and improved according to the characteristics of the LITP problem. With these improved methods, a reconstruction error level below 15% has been attained with a data noise level of 10%. These methods will be further tested and applied in the following LITP experiments. Supported by the ITER-CHINA program 2015GB120001, CHINA MOST under 2012YQ030142 and National Natural Science Foundation Abstract of China under 11575014 and 11375053.

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

    Science.gov (United States)

    Xu, Liqing; Zhang, Jizong; Chen, Kaiyun; Hu, Liqun; Li, Erzhong; Lin, Shiyao; Shi, Tonghui; Duan, Yanmin; Zhu, Yubao

    2015-12-01

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

  16. Nonlinear coupling of tearing modes with self-consistent resistivity evolution in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Carreras, B.; Hicks, H.R.; Holmes, J.A.; Waddell, B.V.

    1980-02-01

    The nonlinear interaction of tearing modes of different helicity is studied for realistic values of the tokamak parameters of resistivity and parallel heat conduction. The self-consistent evolution of the resistivity is taken into account through the electron heat conduction equation. For equilibrium q profiles inferred from electron temperature profile measured before a tokamak disruption, the essential result is that the (m = 2; n = 1) model nonlinearly destabilizes other modes on a rapid time scale. Because of the development of magnetic islands of different helicity, the toroidal current density is severely deformed. These islands overlap and field lines become stochastic in a sizable plasma volume, flattening the temperature profile in this region through parallel heat transport. The deformation of the toroidal current produces a rapid decrease in the self-inductance of the plasma, and the voltage at the limiter decreases, becoming increasingly negative. An extensive survey of equilibria and initial conditions has been conducted, and a simple prescription for their nonlinear stability properties is given.

  17. Density profile control using compact toroid injection in STOR-M Tokamak

    Science.gov (United States)

    Onchi, Takumi; Liu, Dazhi; Xiao, Chijin; Hirose, Akira; Asai, Tomohiko; Wolfe, Sean

    2011-10-01

    The Saskatchewan TORus Modified (STOR-M) tokamak has a Compact Torus (CT) injector which allows tangential injection of high density plasmoid. The objectives of CT injection (CTI) into the core of plasma are to fuel tokamaks and also optimize the bootstrap current in the future reactors by control of the plasma pressure gradient. Measurement of soft X-ray bremsstrahlung emission profile have verified that CT particles are deposited in the core region from outside and steeper density gradient is generated via a balancing process after the asymmetric density profile is formed in STOR-M. The major radius of the core plasma is shifted outward and stays in equilibrium until the end of discharge. H alpha line-emission considerably decreases in the core region and the high emitting area with low temperature plasma exists in the edge region. A few milliseconds seconds after these altered profiles of the density and the emission by CTI are generated, stronger edge radial electric field as well as H-mode appears and the average electron density peaks. This work has been sponsored by the CRC program and NSERC of Canada.

  18. Suppression of high-energy electrons generated in both disrupting and sustained MST tokamak plasmas

    Science.gov (United States)

    Pandya, M. D.; Chapman, B. E.; Munaretto, S.; Cornille, B. S.; McCollam, K. J.; Sovinec, C. R.; Dubois, A. M.; Almagri, A. F.; Goetz, J. A.

    2017-10-01

    High-energy electrons appearing during MST tokamak plasma disruptions are rapidly lost from the plasma due apparently to internal MHD activity. Work has just recently begun on generating and diagnosing disruptions in MST tokamak plasmas. Initial measurements show the characteristic drop in central temperature and density preceding a quench of the plasma current. This corresponds to a burst of dominantly n=1 MHD activity, which is accompanied by a short-lived burst of high-energy electrons. The short-lived nature of these electrons is suspected to be due to stochastic transport associated with the increased MHD. Earlier work shows that runaway electrons generated in low density, sustained plasmas are suppressed by a sufficiently large m=3 RMP in plasmas with q(a) MST's thick conducting shell. With an m=3 RMP, the degree of runaway suppression increases with RMP amplitude, while an m=1 RMP has little effect on the runaways. Nonlinear MHD modeling with NIMROD of these MST plasmas indicates increased stochasticity with an m=3 RMP, while no such increase in stochasticity is observed with an m=1 RMP. Work supported by US DOE.

  19. Energy composition of high-energy neutral beams on the COMPASS tokamak

    Directory of Open Access Journals (Sweden)

    Mitosinkova Klara

    2016-12-01

    Full Text Available The COMPASS tokamak is equipped with two identical neutral beam injectors (NBI for additional plasma heating. They provide a beam of deuterium atoms with a power of up to ~(2 × 300 kW. We show that the neutral beam is not monoenergetic but contains several energy components. An accurate knowledge of the neutral beam power in each individual energy component is essential for a detailed description of the beam- -plasma interaction and better understanding of the NBI heating processes in the COMPASS tokamak. This paper describes the determination of individual energy components in the neutral beam from intensities of the Doppler-shifted Dα lines, which are measured by a high-resolution spectrometer viewing the neutral beam-line at the exit of NBI. Furthermore, the divergence of beamlets escaping single aperture of the last accelerating grid is deduced from the width of the Doppler-shifted lines. Recently, one of the NBI systems was modified by the removal of the Faraday copper shield from the ion source. The comparison of the beam composition and the beamlet divergence before and after this modification is also presented.

  20. Selected highlights of ECH/ECCD physics studies in the TCV tokamak

    Directory of Open Access Journals (Sweden)

    Goodman T.P.

    2015-01-01

    Full Text Available The Tokamak a Configuration Variable, TCV, has used Electron Cyclotron Heating and Current Drive as its only auxiliary heating system for nearly two decades. In addition to basic plasma heating and current profiling, ECH and ECCD under either feedforward or real-time (feedback control allows control of plasma parameters and MHD behaviour to aid in physics studies and measurements. This paper describes four such studies in which EC control has proved crucial – increased resolution Thomson Scattering measurements in the plasma edge, time-resolved plasma rotation modification during the sawtooth cycle, robust neoclassical tearing mode (NTM suppression, and double pass transmission measurements of EC waves for scattering and polarization studies. The relative merits of feedforward and feedback methods for recent TCV experiments are discussed.

  1. Concept definition of KT-2, a large-aspect-ratio diverter tokamak with FWCD

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Kyoo; Chang, In Soon; Chung, Moon Kyoo; Hwang, Chul Kyoo; Lee, Kwang Won; In, Sang Ryul; Choi, Byung Ho; Hong, Bong Keun; Oh, Byung Hoon; Chung, Seung Ho; Yoon, Byung Joo; Yoon, Jae Sung; Song, Woo Sub [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Chang, Choong Suk; Chang, Hong Yung; Choi, Duk In; Nam, Chang Heui [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of); Chung, Kyoo Sun [Hanyang Univ., Seoul (Korea, Republic of); Hong, Sang Heui [Seoul National Univ., Seoul (Korea, Republic of); Kang, Heui Dong [Kyungpook National Univ., Taegu (Korea, Republic of); Lee, Jae Koo [Pohang Inst. of Science and Technology, Kyungnam (Korea, Republic of)

    1994-11-01

    A concept definition of the KT-2 tokamak is made. The research goal of the machine is to study the `advanced tokamak` physics and engineering issues on the mid size large-aspect-ratio diverter tokamak with intense RF heating (>5 MW). Survey of the status of the research fields, the physics basis for the concept, operation scenarios, as well as machine design concept are presented. (Author) 86 refs., 17 figs., 22 tabs.

  2. Structural effects of plasma instabilities on the JET tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Buzio, M

    1999-07-01

    The subject of this work is a novel approach to the analysis of the mechanical response of the main structural components of the JET Tokamak to the JxB forces generated by MHD plasma instabilities. The proposed method is based on the creation of simplified lumped-parameter models representing the essential mechanical and electromagnetic characteristics of the interacting components and consists basically in fitting their output to experimental measurements in order to infer information on induced currents and related forces by means of statistical parameter estimation techniques. First of all, the general time history and space distribution of disruption loads are described and the path of the reaction forces throughout the machine is analysed in detail, with particular reference to the recently observed non-axisymmetric cases. For this purpose, a magneto-static model of the interaction between a kinked plasma and the coil system has been developed. This is based on semi-analytical integration of Biot-Savart's law and makes use of a representation of the conductors in terms of Bezier curves. The method implemented, which pen-nits efficient and detailed calculation of magnetic load distributions in complex geometries, is used to analyse non-axisymmetric events and to point out the most critical components under this kind of loading. The attention is focused next on lumped-parameter models which have been created to represent the basic response modes of the Vessel. These models are represented by a combination of concentrated masses, springs and dampers and include additional parameters describing the magnetic loads. The output of these models is fitted to experimental measurements of displacements and support forces in order to obtain estimates for the magnitude, position and timing of induced currents and related forces. A general procedure for model-based maximum likelihood parameters estimation has been implemented as an interactive PC-Windows program in

  3. Features of self-organized plasma physics in tokamaks

    Science.gov (United States)

    Razumova, K. A.

    2018-01-01

    The history of investigations the role of self-organization processes in tokamak plasma confinement is presented. It was experimentally shown that the normalized pressure profile is the same for different tokamaks. Instead of the conventional Fick equation, where the thermal flux is proportional to a pressure gradient, processes in the plasma are well described by the Dyabilanin’s energy balance equation, in which the heat flux is proportional to the difference of normalized gradients for self-consistent and real pressure profiles. The transport coefficient depends on the values of heat flux, which compensates distortion of the pressure profile with external impacts. Radiative cooling of the plasma edge decreases the heat flux and improves the confinement.

  4. Statistical study of density fluctuations in the tore supra tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Devynck, P.; Fenzi, C.; Garbet, X.; Laviron, C. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Antar, G.; Gervais, F.; Hennequin, P.; Quemeneur, A.; Sabot, R.; Truc, A. [LPMI, CNRS UPR-287, Ecole Polytechnique, 91 - Palaiseau (France)

    1998-03-01

    It is believed that the radial anomalous transport in tokamaks is caused by plasma turbulence. Using infra-red laser scattering technique on the Tore Supra tokamak, statistical properties of the density fluctuations are studied as a function of the scales in ohmic as well as additional heating regimes using the lower hybrid or the ion cyclotron frequencies. The probability distributions are compared to a Gaussian in order to estimate the role of intermittency which is found to be negligible. The temporal behaviour of the three-dimensional spectrum is thoroughly discussed; its multifractal character is reflected in the singularity spectrum. The autocorrelation coefficient as well as their long-time incoherence and statistical independence. We also put forward the existence of fluctuations transfer between two distinct but close wavenumbers. A rather clearer image is thus obtained about the way energy is transferred through the turbulent scales. (author) 28 refs.

  5. The simple map for a single-null divertor tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Punjabi, A.; Verma, A.; Boozer, A. [Hampton Univ. (Vatican City State, Holy See). Center for Fusion Research and Training

    1996-12-01

    We present the simple map for a single-null divertor tokamak. The simple map is an area-preserving map based on the idea that magnetic field lines are a single-degree-of-freedom time-dependent Hamiltonian system, and that the basic features of such systems near the X-point are generic. We obtain the properties of this map and the resulting footprints of field lines on the divertor plate. These include the width of the stochastic layer, the edge safety factor, the area of the footprint and the amount of magnetic flux diverted. We give the safety factor profile, the average and median values of strike angles, lengths and the Liapunov exponents. We describe how the effects of magnetic perturbations can be included in the simple map. We show how the map can be applied to the problem of the determination of heat flux on the divertor plate in tokamaks. (Author).

  6. Stability and heating of a poloidal divertor tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Biddle, A. P.; Dexter, R. N.; Holly, D. T.; Lipschultz, B.; Osborne, T. H.; Prager, S. C.; Shepard, D.A., Sprott, J.C.; Witherspoon, F. D.

    1980-06-01

    Five experimental studies - two stability and three heating investigations - have been carried out on Tokapole II, a Tokamak with a four node poloidal divertor. First, discharges have been attained with safety factor q as low as 0.6 over most of the column without degradation of confinement, and correlation of helical instability onset with current profile shape is being studied. Second, the axisymmetric instability has been investigated in detail for various noncircular cross-sectional shapes, and results have been compared with a numerical stability code adapted to the Tokapole machine. Third, application of high power fast wave ion cyclotron resonance heating doubles the ion temperature and permits observation of heating as a function of harmonic number and spatial location of the resonance. Fourth, low power shear Alfven wave propagation is underway to test the applicability of this heating method to tokamaks. Fifth, preionization by electron cyclotron heating has been employed to reduce the startup loop voltage by approx. 60%.

  7. Remote operation of the GOLEM tokamak for Fusion Education

    Energy Technology Data Exchange (ETDEWEB)

    Grover, O.; Kocman, J. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Odstrcil, M. [University of Southampton, Southampton SO17 1BJ (United Kingdom); Odstrcil, T. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Matusu, M. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Stöckel, J. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Institute of Plasma Physics AS CR, Prague CZ-182 21 (Czech Republic); Svoboda, V., E-mail: svoboda@fjfi.cvut.cz [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Vondrasek, G. [Faculty of Nuclear Sciences and Physical Engineering CTU Prague, CZ-115 19 (Czech Republic); Zara, J. [Faculty of Electrical Engineering CTU Prague, CZ-166 27 (Czech Republic)

    2016-11-15

    Highlights: • The remote operation of the tokamak GOLEM for educational purposes. - Abstract: Practically oriented education in the field of thermonuclear fusion is highly requested. However, the high complexity of appropriate experiments makes it difficult to develop and maintain laboratories where students can take part in hands-on experiments in this field of study. One possible solution is to establish centres with specific high temperature plasma experiments where students can visit such a laboratory and perform their experiments in-situ. With the advancements of IT technologies it naturally follows to make a step forward and connect these with necessary plasma physics technologies and thus allow to access even sophisticated experiments remotely. Tokamak GOLEM is a small, modest device with its infrastructure linked to web technologies allowing students to set-up necessary discharge parameters, submit them into a queue and within minutes obtain the results in the form of a discharge homepage.

  8. Gyrokinetic Simulation of Global Turbulent Transport Properties in Tokamak Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Wang, W.X.; Lin, Z.; Tang, W.M.; Lee, W.W.; Ethier, S.; Lewandowski, J.L.V.; Rewoldt, G.; Hahm, T.S.; Manickam, J.

    2006-01-01

    A general geometry gyro-kinetic model for particle simulation of plasma turbulence in tokamak experiments is described. It incorporates the comprehensive influence of noncircular cross section, realistic plasma profiles, plasma rotation, neoclassical (equilibrium) electric fields, and Coulomb collisions. An interesting result of global turbulence development in a shaped tokamak plasma is presented with regard to nonlinear turbulence spreading into the linearly stable region. The mutual interaction between turbulence and zonal flows in collisionless plasmas is studied with a focus on identifying possible nonlinear saturation mechanisms for zonal flows. A bursting temporal behavior with a period longer than the geodesic acoustic oscillation period is observed even in a collisionless system. Our simulation results suggest that the zonal flows can drive turbulence. However, this process is too weak to be an effective zonal flow saturation mechanism.

  9. Analytical solutions for Tokamak equilibria with reversed toroidal current

    Energy Technology Data Exchange (ETDEWEB)

    Martins, Caroline G. L.; Roberto, M.; Braga, F. L. [Departamento de Fisica, Instituto Tecnologico de Aeronautica, Sao Jose dos Campos, Sao Paulo 12228-900 (Brazil); Caldas, I. L. [Instituto de Fisica, Universidade de Sao Paulo, 05315-970 Sao Paulo, SP (Brazil)

    2011-08-15

    In tokamaks, an advanced plasma confinement regime has been investigated with a central hollow electric current with negative density which gives rise to non-nested magnetic surfaces. We present analytical solutions for the magnetohydrodynamic equilibria of this regime in terms of non-orthogonal toroidal polar coordinates. These solutions are obtained for large aspect ratio tokamaks and they are valid for any kind of reversed hollow current density profiles. The zero order solution of the poloidal magnetic flux function describes nested toroidal magnetic surfaces with a magnetic axis displaced due to the toroidal geometry. The first order correction introduces a poloidal field asymmetry and, consequently, magnetic islands arise around the zero order surface with null poloidal magnetic flux gradient. An analytic expression for the magnetic island width is deduced in terms of the equilibrium parameters. We give examples of the equilibrium plasma profiles and islands obtained for a class of current density profile.

  10. Three-dimensional analysis of tokamaks and stellarators.

    Science.gov (United States)

    Garabedian, Paul R

    2008-09-16

    The NSTAB equilibrium and stability code and the TRAN Monte Carlo transport code furnish a simple but effective numerical simulation of essential features of present tokamak and stellarator experiments. When the mesh size is comparable to the island width, an accurate radial difference scheme in conservation form captures magnetic islands successfully despite a nested surface hypothesis imposed by the mathematics. Three-dimensional asymmetries in bifurcated numerical solutions of the axially symmetric tokamak problem are relevant to the observation of unstable neoclassical tearing modes and edge localized modes in experiments. Islands in compact stellarators with quasiaxial symmetry are easier to control, so these configurations will become good candidates for magnetic fusion if difficulties with safety and stability are encountered in the International Thermonuclear Experimental Reactor (ITER) project.

  11. Three-dimensional analysis of tokamaks and stellarators

    Science.gov (United States)

    Garabedian, Paul R.

    2008-01-01

    The NSTAB equilibrium and stability code and the TRAN Monte Carlo transport code furnish a simple but effective numerical simulation of essential features of present tokamak and stellarator experiments. When the mesh size is comparable to the island width, an accurate radial difference scheme in conservation form captures magnetic islands successfully despite a nested surface hypothesis imposed by the mathematics. Three-dimensional asymmetries in bifurcated numerical solutions of the axially symmetric tokamak problem are relevant to the observation of unstable neoclassical tearing modes and edge localized modes in experiments. Islands in compact stellarators with quasiaxial symmetry are easier to control, so these configurations will become good candidates for magnetic fusion if difficulties with safety and stability are encountered in the International Thermonuclear Experimental Reactor (ITER) project. PMID:18768807

  12. Three-dimensional equilibria in axially symmetric tokamaks

    Science.gov (United States)

    Garabedian, Paul R.

    2006-01-01

    The NSTAB and TRAN computer codes have been developed to study equilibrium, stability, and transport in fusion plasmas with three-dimensional (3D) geometry. The numerical method that is applied calculates islands in tokamaks like the Doublet III-D at General Atomic and the International Thermonuclear Experimental Reactor. When bifurcated 3D solutions are used in Monte Carlo computations of the energy confinement time, a realistic simulation of transport is obtained. The significance of finding many 3D magnetohydrodynamic equilibria in axially symmetric tokamaks needs attention because their cumulative effect may contribute to the prompt loss of α particles or to crashes and disruptions that are observed. The 3D theory predicts good performance for stellarators. PMID:17159158

  13. Progress in application of high temperature superconductor in tokamak magnets

    Czech Academy of Sciences Publication Activity Database

    Gryaznevich, M.; Svoboda, V.; Stöckel, Jan; Sykes, A.; Sykes, N.; Kingham, D.; Hammond, G.; Apte, P.; Todd, T.N.; Ball, S.; Chappell, S.; Melhem, D.; Ďuran, Ivan; Kovařík, Karel; Grover, O.; Markovič, T.; Odstrčil, M.; Odstrčil, T.; Šindlery, A.; Vondrášek, G.; Kocman, J.; Lilley, M.K.; de Grouchy, P.; Kim, H.-T.

    2013-01-01

    Roč. 88, 9-10 (2013), s. 1593-1596 ISSN 0920-3796. [Symposium on Fusion Technology (SOFT-27)/27./. Liège, 24.09.2012-28.09.2012] Institutional support: RVO:61389021 Keywords : tokamaks * HTS * magnet s Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.149, year: 2013 http://www.sciencedirect.com/science/article/pii/S0920379613001117#

  14. Advanced Tokamak Plasmas in the Fusion Ignition Research Experiment

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Kessel; D. Meade; D.W. Swain; P. Titus; M.A. Ulrickson

    2003-10-13

    The Advanced Tokamak (AT) capability of the Fusion Ignition Research Experiment (FIRE) burning plasma experiment is examined with 0-D systems analysis, equilibrium and ideal-MHD stability, radio-frequency current-drive analysis, and full discharge dynamic simulations. These analyses have identified the required parameters for attractive burning AT plasmas, and indicate that these are feasible within the engineering constraints of the device.

  15. Structural materials for large superconducting magnets for tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Long, C.J.

    1976-12-01

    The selection of structural materials for large superconducting magnets for tokamak-type fusion reactors is considered. The important criteria are working stress, radiation resistance, electromagnetic interaction, and general feasibility. The most advantageous materials appear to be face-centered-cubic alloys in the Fe-Ni-Cr system, but high-modulus composites may be necessary where severe pulsed magnetic fields are present. Special-purpose structural materials are considered briefly.

  16. A moving finite element model of the tokamak scrapeoff layer

    Energy Technology Data Exchange (ETDEWEB)

    Glasser, A.H.; Kuprat, A.P.

    1993-10-01

    Most numerical simulations of the tokamak scrapeoff layer use a mapping to flux coordinates and a piecewise equidistributed grid in those coordinates to resolve the multiple length scales and anisotropy characteristic of this problem. We have developed an alternative numerical method using simple cylindrical coordinates with a complex adaptive grid scheme. It is based on an understructured grid of traingles which move adaptively, aligning themselves with the magnetic field and concentrating in regions of sharp gradients.

  17. Application of advanced composites in tokamak magnet systems

    Energy Technology Data Exchange (ETDEWEB)

    Long, C. J.

    1977-11-01

    The use of advanced (high-modulus) composites in superconducting magnets for tokamak fusion reactors is discussed. The most prominent potential application is as the structure in the pulsed poloidal-field coil system, where a significant reduction in eddy currents could be achieved. Present low-temperature data on the advanced composites are reviewed briefly; they are too meager to do more than suggest a broad class of composites for a particular application.

  18. Implementation of rapid imaging system on the COMPASS tokamak.

    Czech Academy of Sciences Publication Activity Database

    Havránek, Aleš; Weinzettl, Vladimír; Fridrich, David; Cavalier, Jordan; Urban, Jakub; Komm, Michael

    2017-01-01

    Roč. 123, November (2017), s. 857-860 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 : Camera * Data acquisition * Video processing * Tokamak Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S092037961730354X

  19. Remote operation of the GOLEM tokamak for Fusion Education

    Czech Academy of Sciences Publication Activity Database

    Grover, O.; Kocman, J.; Odstrčil, M.; Odstrčil, T.; Matušů, M.; Stöckel, Jan; Svoboda, V.; Vondrášek, G.; Žára, J.

    2016-01-01

    Roč. 112, November (2016), s. 1038-1044 ISSN 0920-3796. [Technical Meeting on Control, Data Acquisition, and Remote Participation for Fusion Research IAEA /10./. Ahmedabad, 20.04.2015-24.04.2015] Institutional support: RVO:61389021 Keywords : Tokamak technology * Remote participation * Education * Nuclear fusion Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379616303441

  20. Automation of Aditya tokamak plasma position control DC power supply

    Energy Technology Data Exchange (ETDEWEB)

    Arambhadiya, Bharat, E-mail: bharat@ipr.res.in; Raj, Harshita; Tanna, R.L.; Edappala, Praveenlal; Rajpal, Rachana; Ghosh, Joydeep; Chattopadhyay, P.K.; Kalal, M.B.

    2016-11-15

    Highlights: • Plasma position control is very essential for obtaining repeatable high temperature, high-density discharges of longer durations in tokomak. • The present capacitor bank has limitations of maximum current capacity and position control beyond 200 ms. • The installation of a separate set of coils and a DC power supply can control the plasma position beyond 200 ms. • A high power thyristor (T588N1200) triggers for DC current pulse of 300 A fires precisely at required positions to modify plasma position. • The commissioning is done for the automated in-house, quick and reliable solution. - Abstract: Plasma position control is essential for obtaining repeatable high temperature, high-density discharges of longer duration in tokamaks. Recently, a set of external coils is installed in the vertical field mode configuration to control the radial plasma position in ADITYA tokamak. The existing capacitor bank cannot provide the required current pulse beyond 200 ms for position control. This motivated to have a DC power supply of 500 A to provide current pulse beyond 200 ms for the position control. The automatization of the DC power supply mandated interfaces with the plasma control system, Aditya Pulse Power supply, and Data acquisition system for coordinated discharge operation. A high current thyristor circuit and a timer circuit have been developed for controlling the power supply automatically for charging vertical field coils of Aditya tokamak. Key protection interlocks implemented in the development ensure machine and occupational safety. Fiber-optic trans-receiver isolates the power supply with other subsystems, while analog channel is optically isolated. Commissioning and testing established proper synchronization of the power supply with tokamak operation. The paper discusses the automation of the DC power supply with main circuit components, timing control, and testing results.

  1. Tokamak startup using point-source dc helicity injection.

    Science.gov (United States)

    Battaglia, D J; Bongard, M W; Fonck, R J; Redd, A J; Sontag, A C

    2009-06-05

    Startup of a 0.1 MA tokamak plasma is demonstrated on the ultralow aspect ratio Pegasus Toroidal Experiment using three localized, high-current density sources mounted near the outboard midplane. The injected open field current relaxes via helicity-conserving magnetic turbulence into a tokamaklike magnetic topology where the maximum sustained plasma current is determined by helicity balance and the requirements for magnetic relaxation.

  2. Viscous damping of toroidal angular momentum in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Stacey, W. M. [Georgia Tech Fusion Research Center, Atlanta, Georgia 30332 (United States)

    2014-09-15

    The Braginskii viscous stress tensor formalism was generalized to accommodate non-axisymmetric 3D magnetic fields in general toroidal flux surface geometry in order to provide a representation for the viscous damping of toroidal rotation in tokamaks arising from various “neoclassical toroidal viscosity” mechanisms. In the process, it was verified that the parallel viscosity contribution to damping toroidal angular momentum still vanishes even in the presence of toroidal asymmetries, unless there are 3D radial magnetic fields.

  3. Multi-field plasma sandpile model in tokamaks and applications

    Science.gov (United States)

    Peng, X. D.; Xu, J. Q.

    2016-08-01

    A multi-field sandpile model of tokamak plasmas is formulated for the first time to simulate the dynamic process with interaction between avalanche events on the fast/micro time-scale and diffusive transports on the slow/macro time-scale. The main characteristics of the model are that both particle and energy avalanches of sand grains are taken into account simultaneously. New redistribution rules of a sand-relaxing process are defined according to the transport properties of special turbulence which allows the uphill particle transport. Applying the model, we first simulate the steady-state plasma profile self-sustained by drift wave turbulences in the Ohmic discharge of a tokamak. A scaling law as f = a q0 b + c for the relation of both center-density n ( 0 ) and electron (ion) temperatures T e ( 0 ) ( T i ( 0 ) ) with the center-safety-factor q 0 is found. Then interesting work about the nonlocal transport phenomenon observed in tokamak experiments proceeds. It is found that the core electron temperature increases rapidly in response to the edge cold pulse and inversely it decreases in response to the edge heat pulse. The results show that the nonlocal response of core electron temperature depending on the amplitudes of background plasma density and temperature is more remarkable in a range of gas injection rate. Analyses indicate that the avalanche transport caused by plasma drift instabilities with thresholds is a possible physical mechanism for the nonlocal transport in tokamaks. It is believed that the model is capable of being applied to more extensive questions occurring in the transport field.

  4. A new soft x-ray pulse height analysis array in the HL-2A tokamak.

    Science.gov (United States)

    Zhang, Y P; Liu, Yi; Yang, J W; Song, X Y; Liao, M; Li, X; Yuan, G L; Yang, Q W; Duan, X R; Pan, C H

    2009-12-01

    A new soft x-ray pulse height analysis (PHA) array including nine independent subsystems, on basis of a nonconventional software multichannel analysis system and a silicon drift detector (SDD) linear array consisting of nine high performance SDD detectors, has been developed in the HL-2A tokamak. The use of SDD has greatly improved the measurement accuracy and the spatiotemporal resolutions of the soft x-ray PHA system. Since the ratio of peak to background counts obtained from the SDD PHA system is very high, p/b > or = 3000, the soft x-ray spectra measured by the SDD PHA system can approximatively be regarded as electron velocity distribution. The electron velocity distribution can be well derived in the pure ohmic and auxiliary heating discharges. The performance of the new soft x-ray PHA array and the first experimental results with some discussions are presented.

  5. Reconstruction of equilibrium magnetic configurations in the Globus-M spherical tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Sakharov, N. V., E-mail: nikolay.sakharov@mail.ioffe.ru; Voronin, A. V.; Gusev, V. K. [Russian Academy of Sciences, Ioffe Physical Technical Institute (Russian Federation); Kavin, A. A.; Kamenshchikov, S. N.; Lobanov, K. M. [Efremov Research Institute of Electrophysical Apparatus (Russian Federation); Minaev, V. B.; Novokhatsky, A. N.; Patrov, M. I., E-mail: michael.patrov@mail.ioffe.ru; Petrov, Yu. V.; Shchegolev, P. B. [Russian Academy of Sciences, Ioffe Physical Technical Institute (Russian Federation)

    2015-12-15

    The results of reconstruction of equilibrium magnetic configurations in the Globus-M spherical tokamak by means of the EFIT code and by the method of movable filaments with the use of the data from magnetic measurements are compared. The EFIT code allows one to completely reconstruct the magnetic configuration by solving the Grad−Shafranov equation. In the method of movable filaments, the distribution of the toroidal current flowing through the plasma is described by a set of infinitely thin current-carrying rings. In this method, the last closed magnetic surface (LCMS) and the open surfaces lying beyond the LCMS are calculated. Using both methods, the coordinates of the regions where the separatrix strikes the divertor plates were determined. The results obtained agree well with the distributions of the temperature over the tungsten divertor tiles measured using an IR camera.

  6. Stray light analysis for the Thomson scattering diagnostic of the ETE Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Berni, L. A. [Instituto Nacional de Pesquisas Espaciais (INPE), Laboratorio Associado de Sensores e Materiais (LAS), 12.227-010 Sao Jose dos Campos, SP (Brazil); Albuquerque, B. F. C. [Instituto Nacional de Pesquisas Espaciais (INPE), Engenharia e Tecnologia Espaciais, Divisao de Eletronica Aeroespacial, 12.227-010 Sao Jose dos Campos, SP (Brazil)

    2010-12-15

    Thomson scattering is a well-established diagnostic for measuring local electron temperature and density in fusion plasma, but this technique is particularly difficult to implement due to stray light that can easily mask the scattered signal from plasma. To mitigate this problem in the multipoint Thomson scattering system implemented at the ETE (Experimento Tokamak Esferico) a detailed stray light analysis was performed. The diagnostic system was simulated in ZEMAX software and scattering profiles of the mechanical parts were measured in the laboratory in order to have near realistic results. From simulation, it was possible to identify the main points that contribute to the stray signals and changes in the dump were implemented reducing the stray light signals up to 60 times.

  7. First results from solid state neutral particle analyzer on experimental advanced superconducting tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, J. Z.; Zhao, J. L.; Wan, B. N.; Li, J. G. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Zhu, Y. B., E-mail: y.zhu@uci.edu; Heidbrink, W. W. [Department of Physics and Astronomy, University of California, Irvine, California 92697 (United States)

    2016-11-15

    Full function integrated, compact solid state neutral particle analyzers (ssNPA) based on absolute extreme ultraviolet silicon photodiode have been successfully implemented on the experimental advanced superconducting tokamak to measure energetic particle. The ssNPA system has been operated in advanced current mode with fast temporal and spatial resolution capabilities, with both active and passive charge exchange measurements. It is found that the ssNPA flux signals are increased substantially with neutral beam injection (NBI). The horizontal active array responds to modulated NBI beam promptly, while weaker change is presented on passive array. Compared to near-perpendicular beam, near-tangential beam brings more passive ssNPA flux and a broader profile, while no clear difference is observed on active ssNPA flux and its profile. Significantly enhanced intensities on some ssNPA channels have been observed during ion cyclotron resonant heating.

  8. Tokamak operation with safety factor q95 MHD stability.

    Science.gov (United States)

    Piovesan, P; Hanson, J M; Martin, P; Navratil, G A; Turco, F; Bialek, J; Ferraro, N M; La Haye, R J; Lanctot, M J; Okabayashi, M; Paz-Soldan, C; Strait, E J; Turnbull, A D; Zanca, P; Baruzzo, M; Bolzonella, T; Hyatt, A W; Jackson, G L; Marrelli, L; Piron, L; Shiraki, D

    2014-07-25

    Magnetic feedback control of the resistive-wall mode has enabled the DIII-D tokamak to access stable operation at safety factor q(95) = 1.9 in divertor plasmas for 150 instability growth times. Magnetohydrodynamic stability sets a hard, disruptive limit on the minimum edge safety factor achievable in a tokamak, or on the maximum plasma current at a given toroidal magnetic field. In tokamaks with a divertor, the limit occurs at q(95) = 2, as confirmed in DIII-D. Since the energy confinement time scales linearly with current, this also bounds the performance of a fusion reactor. DIII-D has overcome this limit, opening a whole new high-current regime not accessible before. This result brings significant possible benefits in terms of fusion performance, but it also extends resistive-wall mode physics and its control to conditions never explored before. In present experiments, the q(95) < 2 operation is eventually halted by voltage limits reached in the feedback power supplies, not by intrinsic physics issues. Improvements to power supplies and to control algorithms have the potential to further extend this regime.

  9. Optimization study of normal conductor tokamak for commercial neutron source

    Science.gov (United States)

    Fujita, T.; Sakai, R.; Okamoto, A.

    2017-05-01

    The optimum conceptual design of tokamak with normal conductor coils was studied for minimizing the cost for producing a given neutron flux by using a system code, PEC. It is assumed that the fusion neutrons are used for burning transuranics from the fission reactor spent fuel in the blanket and a fraction of the generated electric power is circulated to opearate the tokamak with moderate plasma fusion gain. The plasma performance was assumed to be moderate ones; {β\\text{N}}~∼ ~3{--}4 in the aspect ratio A~=~2{--}3 and {{H}98y2}~=~1 . The circulating power is an important factor affecting the cost. Though decreasing the aspect ratio is useful to raise the plasma beta and decrease the toroidal field, the maximum field in the coil starts to rise in the very low aspect ratio range and then the circulating power increases with decrease in the plasma aspect ratio A below A~∼ ~2 , while the construction cost increases with A . As a result, the cost per neutron has its minimum around A~∼ ~2.2 , namely, between ST and the conventional tokamak. The average circulating power fraction is expected to be ~51%.

  10. TOKAMAK EQUILIBRIA WITH CENTRAL CURRENT HOLES AND NEGATIVE CURRENT DRIVE

    Energy Technology Data Exchange (ETDEWEB)

    CHU, M.S.; PARKS, P.B.

    2002-06-01

    OAK B202 TOKAMAK EQUILIBRIA WITH CENTRAL CURRENT HOLES AND NEGATIVE CURRENT DRIVE. Several tokamak experiments have reported the development of a central region with vanishing currents (the current hole). Straightforward application of results from the work of Greene, Johnson and Weimer [Phys. Fluids, 3, 67 (1971)] on tokamak equilibrium to these plasmas leads to apparent singularities in several physical quantities including the Shafranov shift and casts doubts on the existence of this type of equilibria. In this paper, the above quoted equilibrium theory is re-examined and extended to include equilibria with a current hole. It is shown that singularities can be circumvented and that equilibria with a central current hole do satisfy the magnetohydrodynamic equilibrium condition with regular behavior for all the physical quantities and do not lead to infinitely large Shafranov shifts. Isolated equilibria with negative current in the central region could exist. But equilibria with negative currents in general do not have neighboring equilibria and thus cannot have experimental realization, i.e. no negative currents can be driven in the central region.

  11. Non-Axisymmetric Shaping of Tokamaks Preserving Quasi-Axisymmetry

    Energy Technology Data Exchange (ETDEWEB)

    Long-Poe Ku and Allen H. Boozer

    2009-06-05

    If quasi-axisymmetry is preserved, non-axisymmetric shaping can be used to design tokamaks that do not require current drive, are resilient to disruptions, and have robust plasma stability without feedback. Suggestions for addressing the critical issues of tokamaks can only be validated when presented with sufficient specificity that validating experiments can be designed. The purpose of this paper is provide that specificity for non-axisymmetric shaping. To our knowledge, no other suggestions for the solution of a number of tokamak issues, such as disruptions, have reached this level of specificity. Sequences of three-field-period quasi-axisymmetric plasmas are studied. These sequences address the questions: (1) What can be achieved at various levels of non-axisymmetric shaping? (2) What simplifications to the coils can be achieved by going to a larger aspect ratio? (3) What range of shaping can be achieved in a single experimental facility? The sequences of plasmas found in this study provide a set of interesting and potentially important configurations.

  12. Compact Torus Fueling of the STOR-M Tokamak

    Science.gov (United States)

    Xiao, C.; Hirose, A.; Zawalski, W.; White, D.; Raman, R.; Decoste, R.; Gregory, B. C.; Martin, F.

    1996-11-01

    Tangential injection of accelerated compact torus (CT) has been performed on the STOR-M tokamak (R/a=46/12 cm, B_t<1 T, I_p<= 50 kA, barn_e=(0.5 - 1)×10^13 cm-3) using the University of Saskatchewan Compact Torus Injector (USCTI). The CT parameters are: m~=1 μg, v=120 km/sec, B=0.1 T and n=(2 - 4)×10^15 cm-3. After CT injection, the electron density in tokamak doubles and the poloidal β-value increases. Indications of reduction in the loop voltage and H_α emission level have also been observed. Currently, following efforts are being made: (a) to coat chromium on the electrode surface, (b) to increase the on-line baking temperature, and (c) to reduce the neutral gas load which follows the CT plasma. In addition, numerical calculation of CT motion in a tokamak magnetic field has been carried out. For horizontal injection, the initial CT magnetic dipole direction should be aligned with the CT velocity for deeper penetration. In the case of vertical injection, the CT trajectory is independent of the initial magnetic dipole direction and central penetration is facilitated by off-axis injection.

  13. Dust-Particle Transport in Tokamak Edge Plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Pigarov, A Y; Krasheninnikov, S I; Soboleva, T K; Rognlien, T D

    2005-09-12

    Dust particulates in the size range of 10nm-100{micro}m are found in all fusion devices. Such dust can be generated during tokamak operation due to strong plasma/material-surface interactions. Some recent experiments and theoretical estimates indicate that dust particles can provide an important source of impurities in the tokamak plasma. Moreover, dust can be a serious threat to the safety of next-step fusion devices. In this paper, recent experimental observations on dust in fusion devices are reviewed. A physical model for dust transport simulation, and a newly developed code DUSTT, are discussed. The DUSTT code incorporates both dust dynamics due to comprehensive dust-plasma interactions as well as the effects of dust heating, charging, and evaporation. The code tracks test dust particles in realistic plasma backgrounds as provided by edge-plasma transport codes. Results are presented for dust transport in current and next-step tokamaks. The effect of dust on divertor plasma profiles and core plasma contamination is examined.

  14. Disruption-induced poloidal currents in the tokamak wall

    Energy Technology Data Exchange (ETDEWEB)

    Pustovitov, V.D., E-mail: Pustovitov_VD@nrcki.ru [National Research Centre ‘Kurchatov Institute’, Pl. Kurchatova 1, Moscow 123182 (Russian Federation); National Research Nuclear University MEPhI, Kashirskoe sh. 31, Moscow 115409, Russia (Russian Federation)

    2017-04-15

    Highlights: • Induction effects during disruptions and rapid transient events in tokamaks. • Plasma-wall electromagnetic interaction. • Flux-conserving evolution of plasma equilibrium. • Poloidal current induced in the vacuum vessel wall in a tokamak. • Complete analytical derivations and estimates. - Abstract: The poloidal current induced in the tokamak wall during fast transient events is analytically evaluated. The analysis is based on the electromagnetic relations coupled with plasma equilibrium equations. The derived formulas describe the consequences of both thermal and current quenches. In the final form, they give explicit dependence of the wall current on the plasma pressure and current. A comparison with numerical results of Villone et al. [F. Villone, G. Ramogida, G. Rubinacci, Fusion Eng. Des. 93, 57 (2015)] for IGNITOR is performed. Our analysis confirms the importance of the effects described there. The estimates show that the disruption-induced poloidal currents in the wall should be necessarily taken into account in the studies of disruptions and disruption mitigation in ITER.

  15. Intrinsic momentum transport in up-down asymmetric tokamaks

    CERN Document Server

    Ball, Justin; Barnes, Michael; Dorland, William; Hammett, Gregory W; Rodrigues, Paulo; Loureiro, Nuno F

    2014-01-01

    Recent work demonstrated that breaking the up-down symmetry of tokamak flux surfaces removes a constraint that limits intrinsic momentum transport, and hence toroidal rotation, to be small. We show, through MHD analysis, that ellipticity is most effective at introducing up-down asymmetry throughout the plasma. We detail an extension to GS2, a local $\\delta f$ gyrokinetic code that self-consistently calculates momentum transport, to permit up-down asymmetric configurations. Tokamaks with tilted elliptical poloidal cross-sections were simulated to determine nonlinear momentum transport. The results, which are consistent with experiment in magnitude, suggest that a toroidal velocity gradient, $\\left( \\partial u_{\\zeta i} / \\partial \\rho \\right) / v_{th i}$, of 5% of the temperature gradient, $\\left(\\partial T_{i} / \\partial \\rho \\right) / T_{i}$, is sustainable. Here $v_{th i}$ is the ion thermal speed, $u_{\\zeta i}$ is the ion toroidal mean flow, $\\rho$ is the minor radial coordinate normalized to the tokamak m...

  16. On the computation of the disruption forces in tokamaks

    Science.gov (United States)

    Pustovitov, V. D.; Rubinacci, G.; Villone, F.

    2017-12-01

    The currents and forces induced in the tokamak vacuum vessel (wall) during the disruption are calculated for different values of wall resistivity. Several consequences and new developments are derived from the general result that the global disruption force acting on the perfectly conducting wall must be exactly opposite to the similar force acting on the plasma, which is inherently small in tokamaks. This theoretical prediction is tested and confirmed here for the ITER tokamak with disruption modelled as the fast thermal quench followed by slower current quench that develops into the vertical displacement event. The plasma is simulated by the evolutionary equilibrium code CarMa0NL. One of the results is that the computed integral force on a perfectly conducting wall is zero at each instant during a disruption. This in turn highlights the importance of having good models for the plasma (in which the equilibrium constraint is explicitly imposed) and for the structures (able to correctly describe the induced currents and the resistive effects). The dependence of the disruption force on the magnetic field penetration through the wall is demonstrated. Also the concept of a disruption force damper is proposed, able to ‘absorb’ a significant part of the force that would arise on a resistive wall during a disruption.

  17. Core fueling to produce peaked density profiles in large tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Mikkelsen, D.R.; McGuire, K.M.; Schmidt, G.L.; Zweben, S.J. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Attenberger, S.E.; Houlberg, W.A.; Milora, S.L. [Oak Ridge National Lab., TN (United States)

    1994-06-01

    Peaking the density profile increases the usable bootstrap current and the average fusion power density; this could reduce the current drive power and increase the net output of power producing tokamaks. The use of neutral beams and pellet injection to produce peaked density profiles is assessed. We show that with radially ``hollow`` diffusivity profiles (and no particle pinch) moderately peaked density profiles can be produced by particle source profiles which are peaked off-axis. The fueling penetration requirements can therefore be relaxed and this greatly improves the feasibility of generating peaked density profiles in large tokamaks. In particular, neutral beam fueling does not require MeV particle energy. Even with beam voltages of {approximately}200 keV, however, exceptionally good particle confinement, {tau}{sub p} {much_gt} {tau}{sub E} is required to achieve net electrical power generation. In system with no power production requirement (e.g., neutron sources) neutral beam fueling should be capable of producing peaked density profiles in devices as large as ITER. Fueling systems with low energy cost per particle (such as cryogenic pellet injection) must be used in power producing tokamaks when {tau}{sub p} {approximately} {tau}{sub E}. Simulations with pellet injection speeds of 7 km/sec show the peaking factor, n{sub eo}/{l_angle}n{sub e}{r_angle}, approaching 2.

  18. Effect of Magnetic Islands on Divertors in Tokamaks and Stellarators

    Science.gov (United States)

    Punjabi, Alkesh; Boozer, Allen

    2017-10-01

    Divertors are required for handling the plasma particle and heat exhausts on the walls in fusion plasmas. Relatively simple methods, models, and maps from field line Hamiltonian are developed to better understand the interaction of strong plasma shaping and magnetic islands on the size and behavior of the magnetic flux tubes that go from the plasma edge to the wall in non-axisymmetric system. This approach is applicable not only in tokamaks but also in stellarators. Stellarator diverters in which magnetic islands are dominant are called resonant and when shaping is dominant are called non-resonant. Optimized stellarators generally have sharp edges on their surface, but unlike the case for tokamaks these edges do not encircle the entire plasma, so they do not define an edge value for the rotational transform. The approach is used in the DIII-D tokamak. Computation results are consistent with the predictions of the models. Further simulations are being done to understand why the transition from an effective cubic to a linear increase in loss time and area of footprint occurs and whether this increase is discontinuous or not. This work is supported by the US DOE Grants DE-FG02-01ER54624 and DE-FG02-04ER54793 to Hampton University and DE-FG02-95ER54333 to Columbia University. This research used resources of the NERSC, supported by the Office of Science, US DOE, under Contract No. DE-AC02-05CH11231.

  19. Energy and particle core transport in tokamaks and stellarators compared

    Energy Technology Data Exchange (ETDEWEB)

    Beurskens, Marc; Angioni, Clemente; Beidler, Craig; Dinklage, Andreas; Fuchert, Golo; Hirsch, Matthias; Puetterich, Thomas; Wolf, Robert [Max-Planck-Institut fuer Plasmaphysik, Greifswald/Garching (Germany)

    2016-07-01

    The paper discusses expectations for core transport in the Wendelstein 7-X stellarator (W7-X) and presents a comparison to tokamaks. In tokamaks, the neoclassical trapped-particle-driven losses are small and turbulence dominates the energy and particle transport. At reactor relevant low collisionality, the heat transport is limited by ion temperature gradient limited turbulence, clamping the temperature gradient. The particle transport is set by an anomalous inward pinch, yielding peaked profiles. A strong edge pedestal adds to the good confinement properties. In traditional stellarators the 3D geometry cause increased trapped orbit losses. At reactor relevant low collisionality and high temperatures, these neoclassical losses would be well above the turbulent transport losses. The W7-X design minimizes neoclassical losses and turbulent transport can become dominant. Moreover, the separation of regions of bad curvature and that of trapped particle orbits in W7-X may have favourable implications on the turbulent electron heat transport. The neoclassical particle thermodiffusion is outward. Without core particle sources the density profile is flat or even hollow. The presence of a turbulence driven inward anomalous particle pinch in W7-X (like in tokamaks) is an open topic of research.

  20. Statistical analysis of first period of operation of FTU Tokamak; Analisi statistica del primo periodo di operazioni del Tokamak FTU

    Energy Technology Data Exchange (ETDEWEB)

    Crisanti, F.; Apruzzese, G.; Frigione, D.; Kroegler, H.; Lovisetto, L.; Mazzitelli, G.; Podda, S. [ENEA, Centro Ricerche Frascati, Rome (Italy). Dip. Energia

    1996-09-01

    On the FTU Tokamak the plasma physics operations started on the 20/4/90. The first plasma had a plasma current Ip=0.75 MA for about a second. The experimental phase lasted until 7/7/94, when a long shut-down begun for installing the toroidal limiter in the inner side of the vacuum vessel. In these four years of operations plasma experiments have been successfully exploited, e.g. experiments of single and multiple pellet injections; full current drive up to Ip=300 KA was obtained by using waves at the frequency of the Lower Hybrid; analysis of ohmic plasma parameters with different materials (from the low Z silicon to high Z tungsten) as plasma facing element was performed. In this work a statistical analysis of the full period of operation is presented. Moreover, a comparison with the statistical data from other Tokamaks is attempted.

  1. Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kobori, Hikaru [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hiwatari, Ryoji [Central Research Institute of Electric Power Industry, Tokyo (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)

    2016-11-01

    Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO{sub 2} emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

  2. Characterization of the Novillo Tokamak in main discharge regime; Caracterizacion del Tokamak Novillo en regimen de descarga principal

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-07-15

    The analytical procedure to carry out the establishment of the discharge in a Tokamak including: a) Ionization, b) Diffusion losses, recombination, union, drift speed, spurious fields, and c) Electric field is presented. In an experimental way a procedure settles down by means of which it is characterized the plasma, specially a new characteristic discharge parameter is settled down and it is the plasma current by the duration of the (I{sub p}t) discharge. (Author)

  3. Calculation about a modification to the toroidal magnetic field of the Tokamak Novillo. Part I; Calculo sobre una modificacion al campo magnetico toroidal del Tokamak Novillo. Parte I

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-07-15

    The charged particles that constitute the plasma in the tokamaks are located in magnetic fields that determine its behavior. The poloidal magnetic field of the plasma current and the toroidal magnetic field of the tokamak possess relatively big gradients, which produce drifts on these particles. These drifts are largely the cause of the continuous lost of particles and of energy of the confinement region. In this work the results of numerical calculations of a modification to the 'traditional' toroidal magnetic field that one waits it diminishes the drifts by gradient and improve the confinement properties of the tokamaks. (Author)

  4. Quantification of chemical erosion in the divertor of the DIII-D tokamak

    Science.gov (United States)

    McLean, Adam Gordon

    The International Thermonuclear Experimental Reactor (ITER) is currently designed to use graphite targets in the divertor for power handling and impurity control. Understanding and quantifying chemical sputtering is therefore key to the success of fusion as a clean energy source. The principal goal of this thesis is to design and carry out experiments, then analyze and interpret the results in order to elucidate the role of chemical sputtering in carbon sources in the DIII-D tokamak. A self-contained gas puff system has been designed, constructed, and employed for in-situ study of chemical erosion. The porous plug injector (PPI) releases methane through a porous graphite surface into the divertor plasma at a precisely calibrated rate, minimizing perturbation to local plasma while replicating the immediate environment of methane molecules released from a solid graphite surface more accurately than done previously. For the first time in a tokamak environment, the methane flow rate used in a puffing experiment was the same order of magnitude as that expected from laboratory experiments for intrinsic chemical sputtering. Effective photon efficiencies for CH4 injection are reported; results are found to have significant dependencies on surface conditions and the divertor operating regime. The contribution of sputtering processes to sources of C0 and C+ are assessed through measurement of background and incremental spectroscopic emissions of both physically and chemically-released sputtering products and by CI, 910 nm line profile fitting. Comparison of background and incremental emissions of chemically-released products demonstrate a dramatic drop in production of CH in cold and detached conditions. Finally, the chemical erosion yield is calculated in both attached and cold-divertor conditions and found to be much closer to that measured ex-situ in ion beam experiments than previously determined in DII-D. These observations represent a positive result for ITER which

  5. Including collisions in gyrokinetic tokamak and stellarator simulations

    Energy Technology Data Exchange (ETDEWEB)

    Kauffmann, Karla

    2012-04-10

    Particle and heat transport in fusion devices often exceed the neoclassical prediction. This anomalous transport is thought to be produced by turbulence caused by microinstabilities such as ion and electron-temperature-gradient (ITG/ETG) and trapped-electron-mode (TEM) instabilities, the latter ones known for being strongly influenced by collisions. Additionally, in stellarators, the neoclassical transport can be important in the core, and therefore investigation of the effects of collisions is an important field of study. Prior to this thesis, however, no gyrokinetic simulations retaining collisions had been performed in stellarator geometry. In this work, collisional effects were added to EUTERPE, a previously collisionless gyrokinetic code which utilizes the {delta}f method. To simulate the collisions, a pitch-angle scattering operator was employed, and its implementation was carried out following the methods proposed in [Takizuka and Abe 1977, Vernay Master's thesis 2008]. To test this implementation, the evolution of the distribution function in a homogeneous plasma was first simulated, where Legendre polynomials constitute eigenfunctions of the collision operator. Also, the solution of the Spitzer problem was reproduced for a cylinder and a tokamak. Both these tests showed that collisions were correctly implemented and that the code is suited for more complex simulations. As a next step, the code was used to calculate the neoclassical radial particle flux by neglecting any turbulent fluctuations in the distribution function and the electric field. Particle fluxes in the neoclassical analytical regimes were simulated for tokamak and stellarator (LHD) configurations. In addition to the comparison with analytical fluxes, a successful benchmark with the DKES code was presented for the tokamak case, which further validates the code for neoclassical simulations. In the final part of the work, the effects of collisions were investigated for slab and toroidal

  6. Effect of limiter currents on plasma equilibrium and stability in a tokamak

    Science.gov (United States)

    Belashov, V. I.; Gribov, Yu. V.; Putvinskij, S. V.; Brevnov, N. N.

    The results of theoretical and experimental research of currents between diaphragms limiting plasma cord in tokamak on plasma equilibrium and stability with an arbitrary form of transverse cross section are presented. It is shown that plasma cord behaviour depends on applied voltage polarity. The phenomena considered can be important for tokamaks in which fast plasma compression in a big radius is invisaged.

  7. Tokamak plasma self-organization and the possibility to have the peaked density profile in ITER

    NARCIS (Netherlands)

    Razumova, K. A.; Andreev, V. F.; Kislov, A. Y.; Kirneva, N. A.; Lysenko, S. E.; Pavlov, Y. D.; Shafranov, T. V.; Donne, A. J. H.; Hogeweij, G. M. D.; Spakman, G. W.; R. Jaspers,; Kantor, M.; Walsh, M.

    2009-01-01

    The self-organization of a tokamak plasma is a fundamental turbulent plasma phenomenon, which leads to the formation of a self-consistent pressure profile. This phenomenon has been investigated in several tokamaks with different methods of heating. It is shown that the normalized pressure profile

  8. Prediction of Pressure and Temperature Gradients in the Tokamak Plasma Edge

    Science.gov (United States)

    Stacey, W. M.

    2017-10-01

    An extended plasma fluid theory that takes into account kinetic ion orbit loss and electromagnetic forces in the continuity, momentum and energy balances, as well as atomic physics and radiation, has been used to reveal the explicit dependence of the temperature and pressure gradients in the tokamak edge plasma on these various factors. Combining the ion radial momentum balance and the Ohm's Law expression for Er reveals the dependence of the radial ion pressure gradient on VxB forces driven by radial particle fluxes, which depend on ion orbit loss, and other factors. The strong temperature gradients measured in the H-mode edge pedestal could certainly be associated with radiative and atomic physics edge cooling effects and the strong reduction in ion and energy fluxes due to ion orbit loss, as well as to the possible reductions in thermal diffusivities that is usually assumed to be the cause. Work supported by USDOE under DE-FC02-04ER54698.

  9. Modification of plasma rotation with resonant magnetic perturbations in the STOR-M tokamak

    Science.gov (United States)

    Elgriw, S.; Liu, Y.; Hirose, A.; Xiao, C.

    2016-04-01

    The toroidal plasma flow velocity of impurity ions has been significantly modified in the Saskatchewan Torus-Modified (STOR-M) tokamak by means of resonant magnetic perturbations (RMP). It has been found that the toroidal flow velocities of OV and CVI impurity ions change towards the co-current direction after the application of a current through a set of (l  =  2, n  =  1) RMP field coils. It has been observed that the reduction of the toroidal flow velocity is closely correlated to the reduction of the magnetohydrodynamic (MHD) fluctuation frequency measured by Mirnov coils. Modulation of the flow velocity has been achieved by switching the RMP current pulses. Non-resonant magnetic perturbations have also induced a much smaller change in the toroidal plasma flow. A theoretical model has been adopted to assess the contributions of different drift mechanisms to magnetic islands rotation in STOR-M.

  10. Effect of Resonant Helical Field (RHF) on Runaway Electrons in Tokamaks

    Science.gov (United States)

    Ghanbari, M. R.; Ghoranneviss, M.; Ghanbari, K.; Elahi, A. Salar; Salem, M. K.; Mohammadi, S.; Arvin, R.

    2013-10-01

    The high energy current of runaway electrons during a major disruption in tokamak reactors can cause serious damage to the first wall of the reactor and reduce its life time. Therefore, finding a method to minimize runaway electron is much needed. Resonant helical field (RHF) is one of the methods for controlling the magnetohydrodynamic (MHD) activity. This paper attempts to examine the effect of RHF on the generation of runaway electrons. Main parameters such as plasma current, loop voltage, emitted hard X-ray intensity, MHD oscillation, Hα radiation and MHD activity modes, in the presence and absence of RHF (L = 2 and L = 3), were measured. The results show that applying this system can change runaway electrons generation.

  11. Two-dimensional vacuum ultraviolet images in different MHD events on the EAST tokamak

    Science.gov (United States)

    Zhijun, WANG; Xiang, GAO; Tingfeng, MING; Yumin, WANG; Fan, ZHOU; Feifei, LONG; Qing, ZHUANG; EAST Team

    2018-02-01

    A high-speed vacuum ultraviolet (VUV) imaging telescope system has been developed to measure the edge plasma emission (including the pedestal region) in the Experimental Advanced Superconducting Tokamak (EAST). The key optics of the high-speed VUV imaging system consists of three parts: an inverse Schwarzschild-type telescope, a micro-channel plate (MCP) and a visible imaging high-speed camera. The VUV imaging system has been operated routinely in the 2016 EAST experiment campaign. The dynamics of the two-dimensional (2D) images of magnetohydrodynamic (MHD) instabilities, such as edge localized modes (ELMs), tearing-like modes and disruptions, have been observed using this system. The related VUV images are presented in this paper, and it indicates the VUV imaging system is a potential tool which can be applied successfully in various plasma conditions.

  12. Current profile evolution during fast wave current drive on the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Petty, C.C.; Forest, C.B.; Baity, F.W.; Chiu, S.C.; deGrassie, J.S.; Groebner, R.J.; Lin-Liu, Y.; Luce, T.C.; Pinsker, R.I.; Porkolab, M.; Prater, R.; Rice, B.W. [General Atomics, San Diego, California 92186 (United States)

    1996-02-01

    The effect of co and counter fast wave current drive (FWCD) on the plasma current profile has been measured for neutral beam heated plasmas with reversed magnetic shear on the DIII-D tokamak. Although the response of the loop voltage profile was consistent with the application of co and counter FWCD, little difference was observed between the current profiles for the opposite directions of FWCD. The evolution of the current profile was successfully modeled using the ONETWO transport code. The simulation showed that the small difference between the current profiles for co and counter FWCD was mainly due to an offsetting change in the ohmic current profile. In addition, the time scale for the loop voltage to reach equilibrium (i.e., flatten) was found to be much longer than the FWCD pulse, which limited the ability of the current profile to fully respond to co or counter FWCD. {copyright} {ital 1996 American Institute of Physics.}

  13. Absolute calibration of Phase Contrast Imaging on HL-2A tokamak

    Science.gov (United States)

    Yu, Yi; Gong, Shaobo; Xu, Min; Wu, Yifan; Yuan, Boda; Ye, Minyou; Duan, Xuru; HL-2A team Team

    2017-10-01

    Phase contrast imaging (PCI) has recently been developed on HL-2A tokamak. In this article we present the calibration of this diagnostic. This system is to diagnose chord integral density fluctuations by measuring the phase shift of a CO2 laser beam with a wavelength of 10.6 μm when the laser beam passes through plasma. Sound waves are used to calibrate PCI diagnostic. The signal series in different PCI channels show a pronounced modulation of incident laser beam by the sound wave. Frequency-wavenumber spectrum is achieved. Calibrations by sound waves with different frequencies exhibit a maximal wavenumber response of 12 cm-1. The conversion relationship between the chord integral plasma density fluctuation and the signal intensity is 2.3-1013 m-2/mV, indicating a high sensitivity. Supported by the National Magnetic Confinement Fusion Energy Research Project (Grant No.2015GB120002, 2013GB107001).

  14. First Real-Time Detection of Surface Dust in a Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Skinner, C.; Rais, B.; Roquemore, A. L.; Kugel, H. W.; Marsala, R.; Provost, T.

    2010-05-20

    The first real-time detection of surface dust inside a tokamak was made using an electrostatic dust detector. A fine grid of interlocking circuit traces was installed in the NSTX vessel and biased to 50 v. Impinging dust particles created a temporary short circuit and the resulting current pulse was recorded by counting electronics. The techniques used to increase the detector sensitivity by a factor of x10,000 to match NSTX dust levels while suppressing electrical pickup are presented. The results were validated by comparison to lab measurements, by the null signal from a covered detector that was only sensitive to pickup, and by the dramatic increase in signal when Li particles were introduced for wall conditioning purposes.

  15. High performance gamma-ray spectrometer for runaway electron studies on the FT-2 tokamak

    Science.gov (United States)

    Shevelev, A. E.; Khilkevitch, E. M.; Lashkul, S. I.; Rozhdestvensky, V. V.; Altukhov, A. B.; Chugunov, I. N.; Doinikov, D. N.; Esipov, L. A.; Gin, D. B.; Iliasova, M. V.; Naidenov, V. O.; Nersesyan, N. S.; Polunovsky, I. A.; Sidorov, A. V.; Kiptily, V. G.

    2016-09-01

    A gamma-ray spectrometer based on LaBr3(Ce) scintillator has been used for measurements of hard X-ray emission generated by runaway electrons in the FT-2 tokamak plasmas. Using of the fast LaBr3(Ce) has allowed extending count rate range of the spectrometer by a factor of 10. A developed digital processing algorithm of the detector signal recorded with a digitizer sampling rate of 250 MHz has provided a pulse height analysis at count rates up to 107 s-1. A spectrum deconvolution code DeGaSum has been applied for inferring the energy distribution of runaway electrons escaping from the plasma and interacting with materials of the FT-2 limiter in the vacuum chamber. The developed digital signal processing technique for LaBr3(Ce) spectrometer has allowed studying the evolution of runaways energy distribution in the FT-2 plasma discharges with time resolution of 1-5 ms.

  16. Simulation of W dust transport in the KSTAR tokamak, comparison with fast camera data

    Directory of Open Access Journals (Sweden)

    A. Autricque

    2017-08-01

    Full Text Available In this paper, dust transport in tokamak plasmas is studied through both experimental and modeling aspects. Image processing routines allowing dust tracking on CCD camera videos are presented. The DUMPRO (DUst Movie PROcessing code features a dust detection method and a trajectory reconstruction algorithm. In addition, a dust transport code named DUMBO (DUst Migration in a plasma BOundary is briefly described. It has been developed at CEA in order to simulate dust grains transport in tokamaks and to evaluate the contribution of dust to the impurity inventory of the plasma. Like other dust transport codes, DUMBO integrates the Orbital Motion Limited (OML approach for dust/plasma interactions modeling. OML gives direct expressions for plasma ions and electrons currents, forces and heat fluxes on a dust grain. The equation of motion is solved, giving access to the dust trajectory. An attempt of model validation is made through comparison of simulated and measured trajectories on the 2015 KSTAR dust injection experiment, where W dust grains were successfully injected in the plasma using a gun-type injector. The trajectories of the injected particles, estimated using the DUMPRO routines applied on videos from the fast CCD camera in KSTAR, show two distinct general dust behaviors, due to different dust sizes. Simulations were made with DUMBO to match the measurements. Plasma parameters were estimated using different diagnostics during the dust injection experiment plasma discharge. The experimental trajectories show longer lifetimes than the simulated ones. This can be due to the substitution of a boiling/sublimation point to the usual vaporization/sublimation cooling, OML limitations (eventual potential barriers in the vicinity of a dust grain are neglected and/or to the lack of a vapor shielding model in DUMBO.

  17. Design and optimization of the electrostatic input module for the ISTTOK Tokamak HIBD cylindrical energy analyzer

    Science.gov (United States)

    Sharma, R.; Nedzelskiy, I. S.; Malaquias, A.; Henriques, R. B.

    2017-11-01

    The Heavy Ion Beam Diagnostic of the ISTTOK Tokamak is based on the injection of a single ionized primary beam (Xe+, Cs+) of energy 20–25 keV and on the collection of all doubly ionized ions emerging from the primary beam due to the impact ionization with the plasma electrons. In the present configuration three local plasma parameters can be retrieved: plasma density (ne), electron temperature (Te), and poloidal magnetic field (Bp). A new detection system is being design in order to determine the plasma potential (Vp) profile via the secondary ions energy measurements. To that end an improved 90o cylindrical electrostatic analyzer (CEA) has been designed and optimized to measure the energy of the secondary ions. This paper reports on the design optimization of the input stage of the new detection arrangement to guide and shape 4 secondary beams into the entrance of the CEA. It is composed of a set of four electrostatic cylindrical plates followed by four pairs of steering plates and a multiple aperture Einzel-like electrostatic lens, consisting of four planar vertical electrodes containing four passing slits. Individual secondary beams with dimension of 8 mm × 2.5 mm were obtained without overlapping between the four beams from different slits. By changing the incident beam energy from 20 keV to 20.2 keV (in order to simulate a local plasma potential value of +200 V), without adjusting bias voltage of the input system, results in a vertical beam shift of 1 mm which can be recovered by changing the voltage on the parallel plate arrangement by 40 V. The flexibility and tolerance of the optic system was tested successfully for the expected angular uncertainty of the beam by +/‑0.5o at the input module entrance and with the expected energy spread of the beam corresponding to different plasma positions inside the tokamak.

  18. Transport and turbulence in a magnetized plasma (application to tokamak plasmas); Transport et turbulence dans un plasma magnetise (application aux plasmas de tokamaks)

    Energy Technology Data Exchange (ETDEWEB)

    Sarazin, Y

    2004-03-01

    This document gathers the lectures made in the framework of a Ph.D level physics class dedicated to plasma physics. This course is made up of 3 parts : 1) collisions and transport, 2) transport and turbulence, and 3) study of a few exchange instabilities. More precisely the first part deals with the following issues: thermonuclear fusion, Coulomb collisions, particles trajectories in a tokamak, neo-classical transport in tokamaks, the bootstrap current, and ware pinch. The second part involves: particle transport in tokamaks, quasi-linear transport, resonance islands, resonance in tokamaks, from quasi to non-linear transport, and non-linear saturation of turbulence. The third part deals with: shift velocities in fluid theory, a model for inter-change instabilities, Rayleigh-Benard instability, Hasegawa-Wakatani model, and Hasegawa-Mima model. This document ends with a series of appendices dealing with: particle-wave interaction, determination of the curvature parameter G, Rossby waves.

  19. Resistive reduced MHD modeling of multi-edge-localized-mode cycles in Tokamak X-point plasmas.

    Science.gov (United States)

    Orain, F; Bécoulet, M; Huijsmans, G T A; Dif-Pradalier, G; Hoelzl, M; Morales, J; Garbet, X; Nardon, E; Pamela, S; Passeron, C; Latu, G; Fil, A; Cahyna, P

    2015-01-23

    The full dynamics of a multi-edge-localized-mode (ELM) cycle is modeled for the first time in realistic tokamak X-point geometry with the nonlinear reduced MHD code jorek. The diamagnetic rotation is found to be instrumental to stabilize the plasma after an ELM crash and to model the cyclic reconstruction and collapse of the plasma pressure profile. ELM relaxations are cyclically initiated each time the pedestal gradient crosses a triggering threshold. Diamagnetic drifts are also found to yield a near-symmetric ELM power deposition on the inner and outer divertor target plates, consistent with experimental measurements.

  20. Observation of reversed shear Alfvén eigenmodes between sawtooth crashes in the Alcator C-Mod tokamak.

    Science.gov (United States)

    Edlund, E M; Porkolab, M; Kramer, G J; Lin, L; Lin, Y; Wukitch, S J

    2009-04-24

    Groups of frequency chirping modes observed between sawtooth crashes in the Alcator C-Mod tokamak are interpreted as reversed shear Alfvén eigenmodes near the q=1 surface. These modes indicate that a reversed shear q profile is generated during the relaxation phase of the sawtooth cycle. Two important parameters, q_{min} and its radial position, are deduced from comparisons of measured density fluctuations with calculations from the ideal MHD code NOVA. These studies provide valuable constraints for further modeling of the sawtooth cycle.

  1. Magneto-hydro-dynamic limits in spherical tokamaks

    Science.gov (United States)

    Hender, T. C.; Allfrey, S. J.; Akers, R.; Appel, L. C.; Bevir, M. K.; Buttery, R. J.; Gryaznevich, M.; Jenkins, I.; Kwon, O. J.; McClements, K. G.; Martin, R.; Medvedev, S.; Nightingale, M. P. S.; Ribeiro, C.; Roach, C. M.; Robinson, D. C.; Sharapov, S. E.; Sykes, A.; Villard, L.; Walsh, M. J.

    1999-05-01

    The operational limits observed in spherical tokamaks, notably the small tight aspect ratio tokamak (START) device [A. Sykes et al., Nucl. Fusion 32, 694 (1992)], are consistent with those found in conventional aspect ratio tokamaks. In particular the highest β achieved (˜40%) is consistent with an ideal magneto-hydro-dynamic (MHD) Troyon type limit, the upper limit on density is well described by the Greenwald density (πa2n¯e/Ip˜1) and the normalized current (Ip/aBt) is limited such that q95≳2. Stability calculations indicate scope for increasing both normalized β and normalized current beyond the values so far achieved, although wall stabilization is generally needed for low-n modes. In double null configurations current terminating disruptions occur at each of the operational boundaries, though the current quench tends to be slow at the density limit and disruptions at high β may be due to the low q. In early limiter START discharges, before the divertor coils were installed, disruptions rarely occurred. Instead internal reconnection events which have all the characteristics of a disruption except the current quench occurred. These various disruptive behaviors are explained in terms of a model in which helicity is conserved during the disruption. Due to the low toroidal field beam ions in START, and α particles in a ST power plant, are super-Alfvénic. This gives the possibility for toroidal Alfvén eigenmodes (TAEs) to occur and such modes are frequently observed in START neutral beam injection (NBI) discharges, but seem to be benign. The features of these observed TAEs are shown to be in agreement with MHD calculations.

  2. Integrated tokamak modeling: when physics informs engineering and research planning

    Science.gov (United States)

    Poli, Francesca

    2017-10-01

    Simulations that integrate virtually all the relevant engineering and physics aspects of a real tokamak experiment are a power tool for experimental interpretation, model validation and planning for both present and future devices. This tutorial will guide through the building blocks of an ``integrated'' tokamak simulation, such as magnetic flux diffusion, thermal, momentum and particle transport, external heating and current drive sources, wall particle sources and sinks. Emphasis is given to the connection and interplay between external actuators and plasma response, between the slow time scales of the current diffusion and the fast time scales of transport, and how reduced and high-fidelity models can contribute to simulate a whole device. To illustrate the potential and limitations of integrated tokamak modeling for discharge prediction, a helium plasma scenario for the ITER pre-nuclear phase is taken as an example. This scenario presents challenges because it requires core-edge integration and advanced models for interaction between waves and fast-ions, which are subject to a limited experimental database for validation and guidance. Starting from a scenario obtained by re-scaling parameters from the demonstration inductive ``ITER baseline'', it is shown how self-consistent simulations that encompass both core and edge plasma regions, as well as high-fidelity heating and current drive source models are needed to set constraints on the density, magnetic field and heating scheme. This tutorial aims at demonstrating how integrated modeling, when used with adequate level of criticism, can not only support design of operational scenarios, but also help to asses the limitations and gaps in the available models, thus indicating where improved modeling tools are required and how present experiments can help their validation and inform research planning. Work supported by DOE under DE-AC02-09CH1146.

  3. 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

  4. On the minimum circulating power of steady state tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Itoh, K.; Itoh, S.; Fukuyama, A.; Yagi, M.

    1995-07-01

    Circulating power for the sustenance and profile control of the steady state tokamak plasmas is discussed. The simultaneous fulfillment of the MHD stability at high beta value, the improved confinement and the stationary equilibrium requires the rotation drive as well as the current drive. In addition to the current drive efficiency, the efficiency for the rotation drive is investigated. The direct rotation drive by the external torque, such as the case of beam injection, is not efficient enough. The mechanism and the magnitude of the spontaneous plasma rotation are studied. (author).

  5. Hall effect on tearing mode instabilities in tokamak

    Science.gov (United States)

    Zhang, W.; Ma, Z. W.; Wang, S.

    2017-10-01

    The tearing mode instability is one of the most important dynamic processes in space and laboratory plasmas. Hall effects, resulting from the decoupling of electron and ion motions, can cause fast development and rotation of the perturbation structure of the tearing mode. A high-accuracy nonlinear magnetohydrodynamics code is developed to study Hall effects on the evolution of tearing modes in the Tokamak geometry. It is found that the linear growth rate increases with the increase in the ion skin depth and the self-consistently generated rotation can greatly alter the dynamic behavior of the double tearing mode.

  6. System design of toroidal field power supply of CDD tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Zheng Zhi

    1996-12-01

    This report deals with system design of Toroidal Field Power Supply of CDD tokamak (CDD-TFPS). The general design philosophy and design variations are introduced. After the outline of CDD-TFPS, the short-circuit calculation, the evaluation of converter parameters, the compatibility of converter and line are carried out. the specifications of major components, semi-conductor devices and accessories are given. High attention is paid to protection system. The design of sub-control and grounding system are described too. Some more general material for power supply design are attached in appendices for reference. (author). 30 tabs., 21 figs.

  7. Dynamic modeling of transport and positional control of tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Jardin, S.C.; Pomphrey, N.; DeLucia, J.

    1985-10-01

    We describe here a numerical model of a free boundary axisymmetric tokamak plasma and its associated control systems. The plasma is modeled with a hybrid method using two-dimensional velocity and flux functions with surface-averaged MHD equations describing the evolution of the adiabatic invariants. Equations are solved for the external circuits and for the effects of eddy currents in nearby conductors. The method is verified by application to several test problems and used to simulate the formation of a bean-shaped plasma in the PBX experiment.

  8. Plasma facing components design of KT-2 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    In, Sang Ryul; Yoon, Byung Joo; Song, Woo Soeb [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Xu, Chao Yin [China Univ. of Science and Technology, Hefei, AH (China)

    1997-04-01

    The vacuum vessel of KT-2 tokamak is protected from high thermal loads by various kinds of plasma facing components (PFC): outer and inner divertors, neutral baffle, inboard limiter, poloidal limiter, movable limiter and passive plate, installed on the inner wall of the vessel. In this report the pre-engineering design of the plasma facing components, including design requirements and function, structures of PFC assemblies, configuration of cooling systems, calculations of some mechanical and hydraulic parameters, is presented. Pumping systems for the movable limiter and the divertor are also discussed briefly. (author). 49 figs.

  9. Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Budny, R.V.; Darrow, D.S.; Medley, S.S.; Nazikian, R.; Zweben, S.J.; et al.

    1998-12-14

    Alpha particle physics experiments were done on the Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium (DT) run from 1993-1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single-particle confinement model in magnetohydrodynamic (MHD) quiescent discharges. Also, the observed alpha particle interactions with sawteeth, toroidal Alfvén eigenmodes (TAE), and ion cyclotron resonant frequency (ICRF) waves were roughly consistent with theoretical modeling. This paper reviews what was learned and identifies what remains to be understood.

  10. Dynamics of nano-dust in tokamak edge plasma

    Energy Technology Data Exchange (ETDEWEB)

    Krasheninnikov, S.I., E-mail: skrash@mae.ucsd.edu [University of California at San Diego, La Jolla, CA 92093 (United States); Soboleva, T.K. [ICN, Universidad Nacional Autonoma de Mexico, Mexico DF (Mexico); Mendis, D.A. [University of California at San Diego, La Jolla, CA 92093 (United States)

    2011-08-01

    The dynamics of nano-scale dust for tokamak edge conditions is reviewed. It is shown that unlike micron-scale grains, where the ion-grain friction is the dominant force acting on the grain, nano-dust dynamics is the subject of both friction and Lorentz forces. Possible impact of nano-dust on MARFE is investigated and it is found that dust, providing plasma particle sink and thus causing plasma flow, can play a dominant role in the localization of impurity in low temperature region. It is also shown that dust can significantly reduce the growth rate of flute instability.

  11. Steady-state current drive in tokamaks workshop summary

    Energy Technology Data Exchange (ETDEWEB)

    1979-02-01

    The purpose of the workshop was to identify the most promising techniques and to outline the expectations for tokamak reactor concepts. The group which included beam and rf specialists were asked to assist in the preparation of specific recommendations for the establishment of a program directed at demonstrating current drive. This report includes the recommendations and conclusions as prepared by A. Bers, D. Jassby, and T. Hsu and the summary papers submitted by the participants describing the various experiments and studies presented at the workshop.

  12. Nonlinear saturation of ballooning modes in tokamaks and stellarators

    Science.gov (United States)

    Bauer, F.; Garabedian, P.; Betancourt, O.

    1988-01-01

    The spectral code BETAS computes plasma equilibrium in a toroidal magnetic field B = [unk]s × [unk]Ψ with remarkable accuracy because the finite difference scheme employed in the radial direction allows for discontinuities of the flux function Ψ across the nested surfaces s = const. Instability of higher modes in stellarators like the Heliotron E can be detected in roughly an hour on the best supercomputers by calculating bifurcated equilibria that are defined over just one field period. The method has been validated by comparing results about nonlinear saturation of ballooning modes in tokamaks with numerical data from the PEST code. PMID:16593984

  13. Protecting Against Damage from Refraction of High Power Microwaves in the DIII-D Tokamak

    Directory of Open Access Journals (Sweden)

    Lohr John

    2017-01-01

    Full Text Available Several new protective systems are being installed on the DIII D tokamak to increase the safety margins for plasma operations with injected ECH power at densities approaching cutoff. Inadvertent overdense operation has previously resulted in reflection of an rf beam back into a launcher causing extensive arcing and melt damage on one waveguide line. Damage to microwave diagnostics, which are located on the same side of the tokamak as the ECH launchers, also has occurred. Developing a reliable microwave based interlock to protect the many vulnerable systems in DIII-D has proved to be difficult. Therefore, multiple protective steps have been taken to reduce the risk of damage in the future. Among these is a density interlock generated by the plasma control system, with setpoint determined by the ECH operators based on rf beam trajectories and plasma parameters. Also installed are enhanced video monitoring of the launchers, and an ambient light monitor on each of the waveguide systems, along with a Langmuir probe at the mouth of each launcher. Versatile rf monitors, measuring forward and reflected power in addition to the mode content of the rf beams, have been installed as the last miter bends in each waveguide line. As these systems are characterized, they are being incorporated in the interlock chains, which enable the ECH injection permits. The diagnostics most susceptible to damage from the ECH waves have also been fitted with a variety of protective devices including stripline filters, thin resonant notch filters tuned to the 110 GHz injected microwave frequency, blazed grating filters and shutters. Calculations of rf beam trajectories in the plasmas are performed using the TORAY ray tracing code with input from kinetic profile diagnostics. Using these calculations, strike points for refracted beams on the vacuum vessel are calculated, which allows evaluation of the risk of damage to sensitive diagnostics and hardware.

  14. Protecting Against Damage from Refraction of High Power Microwaves in the DIII-D Tokamak

    Science.gov (United States)

    Lohr, John; Brambila, Rigo; Cengher, Mirela; Chen, Xi; Gorelov, Yuri; Grosnickle, William; Moeller, Charles; Ponce, Dan; Prater, Ron; Torrezan, Antonio; Austin, Max; Doyle, Edward; Hu, Xing; Dormier, Calvin

    2017-07-01

    Several new protective systems are being installed on the DIII D tokamak to increase the safety margins for plasma operations with injected ECH power at densities approaching cutoff. Inadvertent overdense operation has previously resulted in reflection of an rf beam back into a launcher causing extensive arcing and melt damage on one waveguide line. Damage to microwave diagnostics, which are located on the same side of the tokamak as the ECH launchers, also has occurred. Developing a reliable microwave based interlock to protect the many vulnerable systems in DIII-D has proved to be difficult. Therefore, multiple protective steps have been taken to reduce the risk of damage in the future. Among these is a density interlock generated by the plasma control system, with setpoint determined by the ECH operators based on rf beam trajectories and plasma parameters. Also installed are enhanced video monitoring of the launchers, and an ambient light monitor on each of the waveguide systems, along with a Langmuir probe at the mouth of each launcher. Versatile rf monitors, measuring forward and reflected power in addition to the mode content of the rf beams, have been installed as the last miter bends in each waveguide line. As these systems are characterized, they are being incorporated in the interlock chains, which enable the ECH injection permits. The diagnostics most susceptible to damage from the ECH waves have also been fitted with a variety of protective devices including stripline filters, thin resonant notch filters tuned to the 110 GHz injected microwave frequency, blazed grating filters and shutters. Calculations of rf beam trajectories in the plasmas are performed using the TORAY ray tracing code with input from kinetic profile diagnostics. Using these calculations, strike points for refracted beams on the vacuum vessel are calculated, which allows evaluation of the risk of damage to sensitive diagnostics and hardware.

  15. Transport studies in TJ-I tokamak from steady and perturbative methods

    Energy Technology Data Exchange (ETDEWEB)

    Pardo, C.; Rodriguez-Yunta, A.; Vega, J.; Branas, B.; Estrada, T.; Ochando, M.A.; Tabares, F.L.; Zurro, B. (Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas (CIEMAT), Madrid (Spain))

    1992-01-01

    Transport understanding is an essential task for the development of a future tokamak fusion reactor. In a general formulation, the dependence of particle and electron energy fluxes on density and temperature gradients, may be written as: [Gamma]=-D[nabla]n - D[sub T]n [nabla]T/T - nV, q=Q-5/2[Gamma]T=-[chi][sub n]T[nabla]n-[chi]n[nabla]T-nTU where both fluxes are related to both gradients, and the particle and energy pinches V and U may depend on any other force as could be the parallel electric field. The transport coefficients must be expected to be functions of local plasma parameters such as B, q, n, T, [nabla]n, [nabla]T, ... etc. This means that the fluxes may be non-linear functions of the gradients. Transport analysis in the steady state gives values, from experimental data, for fluxes and gradients. This is not enough to determine the values of the six transport coefficients. A perturbative experiment, such as the simultaneous measurement of density and temperature pulses induced by a sawtooth collapse, give us the incremental transport coefficients or the derivatives of the fluxes with respect to the gradients. By making a coupled analysis of both pulses, we can obtain values for the four derivatives: [partial derivative][Gamma][partial derivative][nabla]n, [partial derivative][Gamma]/[partial derivative][nabla]T, [partial derivative]q/[partial derivative][nabla]n and [partial derivative]q/[partial derivative][nabla]T. The combination of both steady and perturbative studies in discharges with different plasma parameters could give us a better picture of transport processes in a tokamak. (author) 6 refs., 5 figs.

  16. Improved Ohmic Confinement Induced by Turbulent Heating and Electrode Biasing in the Stor-M Tokamak.

    Science.gov (United States)

    Zhang, Wei

    Applying a short current pulse during a nominal Ohmic discharge in the STOR-M tokamak triggers an Ohmic H-mode characterized by reduced H_alpha radiation, increased global energy confinement time, reduced edge density and magnetic fluctuations, and the suppression of sawtooth oscillations. Measurements of the plasma floating potential at the plasma edge and in the scrape-off layer indicate that the Ohmic H-mode is accompanied by negative plasma autobiasing which leads to a steeper radial electric field profile at the edge. Since the duration of the current pulse ( 1) and at the plasma edge (r/a<= 1). It is found that the electromagnetic mode is dominant in the main discharge. The propagation direction of fluctuations is in the electron diamagnetic direction. Around f~eq 250 kHz, the poloidal wavenumber k_theta = 2.0 cm^ {-1} is of the order of the inverse electron skin depth k_theta~eq omega _{pe}/C. In terms of the hybrid ion Larmor radius rho_{s} = c_{s}/Omega_{i}, it corresponds to k_theta rho_{s}~eq 0.1. These observations support the theoretical prediction that a low beta tokamak discharge is susceptible to the skin size electromagnetic instability. In the SOL region, the density fiuctuations propagate in the ion diamagnetic drift direction, but with the local E x B drift velocity which changes sign (along with E,) at r/a~eq 1. However, the magnetic fluctuations continue to propagate in the electron diamagnetic drift direction. The magnetic and density fluctuations in the SOL thus obey different dispersion relations, and they originate from different modes. The magnetic fluctuation intensity decreases towards the SOL, while the density fluctuation level increases. This observation suggests that in the main discharge, the fluctuations are of an electromagnetic nature, while in the SOL, the fluctuations are predominantly electrostatic.

  17. Lower Hybrid Heating and Current Drive on the Alcator C-Mod Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    R. Wilson, R. Parker, M. Bitter, P.T. Bonoli, C. Fiore, R.W. Harvey, K. Hill, A.E. Hubbard, J.W. Hughes, A. Ince-Cushman, C. Kessel, J.S. Ko, O. Meneghini, C.K. Phillips, M. Porkolab, J. Rice, A.E. Schmidt, S. Scott,S. Shiraiwa, E. Valeo, G.Wallace, J.C. Wright and the Alcator C-Mod Team

    2009-11-20

    On the Alcator C-Mod tokamak, lower hybrid current drive (LHCD) is being used to modify the current profile with the aim of obtaining advanced tokamak (AT) performance in plasmas with parameters similar to those that would be required on ITER. To date, power levels in excess of 1 MW at a frequency of 4.6 GHz have been coupled into a variety of plasmas. Experiments have established that LHCD on C-Mod behaves globally as predicted by theory. Bulk current drive efficiencies, n20IlhR/Plh ~ 0.25, inferred from magnetics and MSE are in line with theory. Quantitative comparisons between local measurements, MSE, ECE and hard x-ray bremsstrahlung, and theory/simulation using the GENRAY, TORIC-LH CQL3D and TSC-LSC codes have been performed. These comparisons have demonstrated the off-axis localization of the current drive, its magnitude and location dependence on the launched n|| spectrum, and the use of LHCD during the current ramp to save volt-seconds and delay the peaking of the current profile. Broadening of the x-ray emission profile during ICRF heating indicates that the current drive location can be controlled by the electron temperature, as expected. In addition, an alteration in the plasma toroidal rotation profile during LHCD has been observed with a significant rotation in the counter current direction. Notably, the rotation is accompanied by peaking of the density and temperature profiles on a current diffusion time scale inside of the half radius where the LH absorption is taking place.

  18. Millimeter-wave imaging diagnostics systems on the EAST tokamak (invited)

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, Y. L.; Xie, J. L., E-mail: jlxie@ustc.edu.cn; Yu, C. X.; Zhao, Z. L.; Gao, B. X.; Chen, D. X.; Liu, W. D.; Liao, W.; Qu, C. M.; Luo, C. [School of Physics, University of Science and Technology of China, Anhui 230026 (China); Hu, X.; Spear, A. G.; Luhmann, N. C.; Domier, C. W.; Chen, M.; Ren, X. [University of California, Davis, California 95616 (United States); Tobias, B. J. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

    2016-11-15

    Millimeter-wave imaging diagnostics, with large poloidal span and wide radial range, have been developed on the EAST tokamak for visualization of 2D electron temperature and density fluctuations. A 384 channel (24 poloidal × 16 radial) Electron Cyclotron Emission Imaging (ECEI) system in F-band (90-140 GHz) was installed on the EAST tokamak in 2012 to provide 2D electron temperature fluctuation images with high spatial and temporal resolution. A co-located Microwave Imaging Reflectometry (MIR) will be installed for imaging of density fluctuations by December 2016. This “4th generation” MIR system has eight independent frequency illumination beams in W-band (75-110 GHz) driven by fast tuning synthesizers and active multipliers. Both of these advanced millimeter-wave imaging diagnostic systems have applied the latest techniques. A novel design philosophy “general optics structure” has been employed for the design of the ECEI and MIR receiver optics with large aperture. The extended radial and poloidal coverage of ECEI on EAST is made possible by innovations in the design of front-end optics. The front-end optical structures of the two imaging diagnostics, ECEI and MIR, have been integrated into a compact system, including the ECEI receiver and MIR transmitter and receiver. Two imaging systems share the same mid-plane port for simultaneous, co-located 2D fluctuation measurements of electron density and temperature. An intelligent remote-control is utilized in the MIR electronics systems to maintain focusing at the desired radial region even with density variations by remotely tuning the probe frequencies in about 200 μs. A similar intelligent technique has also been applied on the ECEI IF system, with remote configuration of the attenuations for each channel.

  19. Plasma diagnostics for tokamaks and stellarators

    Energy Technology Data Exchange (ETDEWEB)

    Stott, P. E.; Sanchez, J.

    1994-07-01

    A collection of papers on plasma diagnostics is presented. The papers show the state of the art developments in a series of techniques: Magnetic diagnostics, Edge diagnostics, Langmuir probes, Spectroscopy, Microwave and FIR diagnostics as well as Thomson Scattering. Special interest was focused on those diagnostics oriented to fluctuations measurements in the plasma. (Author) 451 refs.

  20. Magnetic field threshold for runaway generation in tokamak disruptions

    Science.gov (United States)

    F"Ul"Op, T.; Pokol, G.; Smith, H. M.; Helander, P.

    2009-05-01

    Due to a sudden cooling of the plasma in tokamak disruptions a beam of relativistic runaway electrons is sometimes generated, which may cause damage on plasma facing components. Experimental observations on large tokamaks show that the number of runaway electrons produced in disruptions depends on the magnetic field strength. In this work, two possible reasons for this threshold are studied. The first possible explanation for these observations is that the runaway beam excites whistler waves that scatter the electrons in velocity space and prevents the beam from growing. The growth rates of the most unstable whistler waves are inversely proportional to the magnetic field strength and it is possible to derive a magnetic field threshold below which no runaways are expected. The second possible explanation is the magnetic field dependence of the criterion for substantial runaway production determined by the induced electric field available and by the efficiency of the generation mechanisms. It is shown, that even in rapidly cooling plasmas, where hot-tail generation is expected to give rise to substantial runaway population, the whistler waves can stop the runaway formation below a certain magnetic field unless the post-disruption temperature is very low.

  1. Hand-eye coordinative remote maintenance in a tokamak vessel

    Energy Technology Data Exchange (ETDEWEB)

    Qiu, Qiang, E-mail: qiu6401@sjtu.edu.cn; Gu, Kai, E-mail: gukai0707@sjtu.edu.cn; Wang, Pengfei, E-mail: wpf790714@163.com; Bai, Weibang, E-mail: 654253204@qq.com; Cao, Qixin, E-mail: qxcao@sjtu.edu.cn

    2016-03-15

    Highlights: • If there is not rotation between the visual coordinate frame (O{sub e}X{sub e}Y{sub e}) and hand coordinate frame (O{sub h}X{sub h}Y{sub h}), a person can coordinate the movement between hand and eye easily. • We establish an alignment between the movement of the operator's hand and the visual scene of the end-effector as displayed on the monitor. • A potential function is set up in a simplified vacuum vessel model to provide a fast collision checking, and the alignment between repulsive force and Omega 7 feedback force is accomplished. • We carry out an experiment to evaluate its performance in a remote handling task. - Abstract: The reliability is vitally important for the remote maintenance in a tokamak vessel. In order to establish a more accurate and safer remote handling system, a hand-eye coordination method and an artificial potential function based collision avoidance method were proposed in this paper. At the end of this paper, these methods were implemented to a bolts tightening maintenance task, which was carried out in our 1/10 scale tokamak model. Experiment results have verified the value of the hand-eye coordination method and the collision avoidance method.

  2. Active cooling system for Tokamak in-vessel operation manipulator

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Jianjun, E-mail: yuanjj@sjtu.edu.cn; Chen, Tan; Li, Fashe; Zhang, Weijun; Du, Liang

    2015-10-15

    Highlights: • We summarized most of the challenges of fusion devices to robot systems. • Propose an active cooling system to protect all of the necessary components. • Trial design test and theoretical analysis were conducted. • Overall implementation of the active cooling system was demonstrated. - Abstract: In-vessel operation/inspection is an indispensable task for Tokamak experimental reactor, for a robot/manipulator is more capable in doing this than human being with more precise motion and less risk of damaging the ambient equipment. Considering the demanding conditions of Tokamak, the manipulator should be adaptable to rapid response in the extreme conditions such as high temperature, vacuum and so on. In this paper, we propose an active cooling system embedded into such manipulator. Cameras, motors, gearboxes, sensors, and other mechanical/electrical components could then be designed under ordinary conditions. The cooling system cannot only be a thermal shield since the components are also heat sources in dynamics. We carry out a trial test to verify our proposal, and analyze the active cooling system theoretically, which gives a direction on the optimization by varying design parameters, components and distribution. And based on thermal sensors monitoring and water flow adjusting a closed-loop feedback control of temperature is added to the system. With the preliminary results, we believe that the proposal gives a way to robust and inexpensive design in extreme environment. Further work will concentrate on overall implementation and evaluation of this cooling system with the whole inspection manipulator.

  3. ADVANCES IN DUST DETECTION AND REMOVAL FOR TOKAMAKS

    Energy Technology Data Exchange (ETDEWEB)

    Campos, A.; Skinner, C.H.

    2009-01-01

    Dust diagnostics and removal techniques are vital for the safe operation of next step fusion devices such as ITER. In the tokamak environment, large particles or fi bers can fall on the electrostatic detector potentially causing a permanent short. An electrostatic dust detector developed in the laboratory is being applied to the National Spherical Torus Experiment (NSTX). We report on the development of a gas puff system that uses helium to clear such particles from the detector. Experiments at atmospheric pressure with varying nozzle designs, backing pressures, puff durations and exit fl ow orientations have given an optimal confi guration that effectively removes particles from a 25 cm² area. Similar removal effi ciencies were observed under a vacuum base pressure of 1 mTorr. Dust removal from next step tokamaks will be required to meet regulatory dust limits. A tri-polar grid of fi ne interdigitated traces has been designed that generates an electrostatic traveling wave for conveying dust particles to a “drain.” First trials with only two working electrodes have shown particle motion in optical microscope images.

  4. SLPX: superconducting long-pulse tokamak experiment. [NbTi

    Energy Technology Data Exchange (ETDEWEB)

    Jassby, D.L.; File, J.; Bronner, G.

    1978-09-25

    The principal objectives of the SLPX (Superconducting Long-Pulse Experiment) are: (1) to demonstrate quasi-steady operation of 3 to 5 MA hydrogen and deuterium tokamak plasmas at high temperature and high thermal wall loading, and (2) to develop reliable operation of prototypical tokamak reactor magnetics systems featuring a toroidal assembly of high-field niobium-tin coils, and a system of pulsed niobium-titanium superconducting poloidal-field coils. This paper describes the status of the engineering design features of the SLPX, with emphasis on the magnetics systems. The toroidal-field coils have an aperture of 3.1 x 4.8 m and can operate with a maximum field at the conductor of 12 T. The superconducting poloidal field magnetics system consists of a pulsed NbTi central solenoid and a set of dc NbTi equilibrium-field coils. The entire machine is enclosed in an outer vacuum container equipped with re-entrant ports that provide ambient access to the room-temperature plasma vessel.

  5. Equilibrium, confinement and stability of runaway electrons in tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Spong, D A

    1976-03-01

    Some of the ramifications of the runaway population in tokamak experiments are investigated. Consideration is given both to the normal operating regime of tokamaks where only a small fraction of high energy runaways are present and to the strong runaway regime where runaways are thought to carry a significant portion of the toroidal current. In particular, the areas to be examined are the modeling of strong runaway discharges, single particle orbit characteristics of runaways, macroscopic beam-plasma equilibria, and stability against kink modes. A simple one-dimensional, time-dependent model has been constructed in relation to strong runaway discharges. Single particle orbits are analyzed in relation to both the strong runaway regime and the weak regime. The effects of vector E x vector B drifts are first considered in strong runaway discharges and are found to lead to a slow inward shrinkage of the beam. Macroscopic beam-plasma equilibria are treated assuming a pressureless relativistic beam with inertia and using an ideal MHD approximation for the plasma. The stability of a toroidal relativistic beam against kink perturbations is examined using several models. (MOW)

  6. Alternating current plasma operation in the STOR-M tokamak

    Science.gov (United States)

    Mitarai, O.; Xiao, Chijin; Zhang, Liyan; McColl, D.; Zhang, Wei; Conway, G.; Hirose, A.; Skarsgard, H. M.

    1996-10-01

    One cycle alternating current (AC) plasma operation without a dwell time has been achieved in the STOR-M tokamak with good reproducibility using a newly developed ohmic heating circuit. The plasma current of +24 kA is smoothly ramped down in 10 ms with a rampdown rate of around 2.0 kA/ms and then ramped up to between -20 and -24 kA directly without a dwell time. The plasma density of up to (3.7+or-0.6)*1018 m-3 remains at the current reversal as observed in recent soft landing experiments. The key to a successful, reproducible and direct transition in AC tokamak operations on STOR-M is to control both the total vertical field by a feedback control system and the plasma density by careful gas puffing during the current reversal phase. This experiment has demonstrated that the initial loop voltage for the second negative current is minimized when the dwell time approaches zero, and the AC operation without dwelling is possible whenever the plasma current can be softly terminated with a finite residual plasma density

  7. Recent Results from the STOR-M Tokamak

    Science.gov (United States)

    Hirose, A.; Dreval, M.; Elgriw, S.; Mitarai, O.; Pant, A.; Peng, M.; Rohraff, D.; Singh, A. K.; Trembach, D.; Xiao, C.

    2008-04-01

    This paper reports on two recent experiments carried out on the STOR-M tokamak. The first experiment studied the nature of MHD activities based on singular value decomposition algorithm during the improved confinement phase induced by compact torus injection. The typical MHD modes with mode numbers m = 2, 3, and 4 are suppressed during the improved confinement phase. Shortly before the termination of the improved confinement phase, MHD activities reemerge, starting with a gong-mode-like burst followed by oscillations of a rotating m = 2. The second experiment was successful current start-up with a simulated spherical tokamak configuration where the inner Ohmic heating coils surrounding the iron core are deactivated in STOR-M. Current start-up was also achieved with all the vertical equilibrium field coils deactivated. In the latter case, the vertical equilibrium field was provided solely by the image vertical field produced by the magnetization current in the iron core and compensated for by the current through the feedback control vertical field windings. The observed waveforms agree well with numerical simulations.

  8. Tokamak active laser pyrometry for tungsten deposited layer characterisation

    Energy Technology Data Exchange (ETDEWEB)

    Semerok, A., E-mail: alexandre.semerok@cea.fr [CEA Saclay, DEN/DENS/DPC/SCP/LILM, P.C. 56, 91191 Gif-sur-Yvette, Cedex (France); Jaubert, F.; Fomichev, S.V.; Thro, P.-Y. [CEA Saclay, DEN/DENS/DPC/SCP/LILM, P.C. 56, 91191 Gif-sur-Yvette, Cedex (France); Grisolia, C. [CEA Cadarache, IRFM, 13108, Saint Paul-lez-Durance, Cedex (France)

    2012-03-15

    In modern fusion reactors, the erosion of plasma facing surface results in layers deposition on tokamak 'cold' surfaces. To provide efficient operation of tokamaks, it is essential to characterise the deposited layer with high tritium content. In situ rapid surface characterisation without reactor components disassembly is required. Active laser pyrometry together with a repetition rate Nd-YAG laser (1 Hz-1 kHz repetition rate frequency) applied for surface heating can be used to characterise some thermo-physical properties (thermal capacity, thermal contact, and conductivity) of a micrometric layer. The pyrometer system was developed and applied to characterise some properties of a W-layer (140 {mu}m) on a CFC-substrate. The numerical code developed for 3-D simulation of LH of a surface with the deposited layer was applied to simulate the experimental heating temperatures. The experimental and simulation results were compared. W-layer characterisation was performed by fitting the experimental and theoretical heating temperatures.

  9. 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)

  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. Overview of the EUROfusion Medium Size Tokamak program

    Science.gov (United States)

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

    2015-11-01

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

  12. Dynamics and transport of dust particles in tokamak edge plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Krasheninnikov, S I [University of California, San Diego, La Jolla, CA 92093 (United States); Soboleva, T K [UNAM, Mexico D.F., Mexico and Kurchatov Institute, Moscow (Russian Federation)

    2005-05-01

    We discuss the dust particle dynamics in tokamak edge plasmas, with special emphasis on dust particle transport in the sheath and plasma recycling regions. We demonstrate that being dragged by plasma flows in the vicinity of the material surface, dust particles can be accelerated to speeds of {approx}10{sup 3}-10{sup 4} cm s{sup -1}. The opposite direction of plasma recycling flow as well as the frictional forces at the inner and outer divertor legs, propel the dust particles in opposite toroidal directions depending on their location. The interactions of a dust particle with a corrugated surface or plasma turbulence can cause it to exit the recycling region and fly through the scrape-off layer plasma towards the tokamak core. It is conceivable that dust formation in and transport from the divertor region can play an important role in core plasma contamination. However, even then, the dust particle density around the separatrix is {approx}10{sup -2} cm{sup -3}, which makes it difficult to detect.

  13. Observation of MHD phenomenon for SST-1 superconducting tokamak

    Science.gov (United States)

    Bhandarkar, Manisha; Dhongde, Jasraj; Pradhan, Subrata

    2017-04-01

    Steady State Superconducting Tokamak (SST-1) is a medium size Tokamak (major radius = 1.1 m, minor radius = 0.2 m) and is operational at the Institute for Plasma Research (IPR), India. In the last few experimental campaigns SST-1 has successfully achieved plasma current in order of 60-70kA and plasma duration in excess of ∼ 500 ms at a central magnetic field of 1.5T. An attempt has made to study the behavior of the magneto-hydrodynamic (MHD) activity during different phases of plasma pulse which leads to major/minor disruptions, its present modes (poloidal/toroidal mode number i.e. m = 2, n = 1) impact on plasma confinement and signature of lock mode and its frequency in the SST-1 plasma using experimental data from Mirnov signals. Observed MHD phenomenon has also been correlated with other diagnostics (i.e. ECE, Density, Soft X-Ray etc.) and heating system (ECRH) for the recent campaigns of SST-1.

  14. Configuration and operation of detritiation systems for ITER Tokamak Complex

    Energy Technology Data Exchange (ETDEWEB)

    Beloglazov, S., E-mail: sergey.beloglazov@iter.org [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Camp, P. [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Hayashi, T. [Japan Atomic Energy Agency, 2-2-2 Uchisaiwai-cho, Chiyoda, Tokyo 100-0011 (Japan); Lepetit, L.; Perevezentsev, A. [ITER Organization, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Yamanishi, Y. [Japan Atomic Energy Agency, 2-2-2 Uchisaiwai-cho, Chiyoda, Tokyo 100-0011 (Japan)

    2010-12-15

    The ventilation systems design for the ITER nuclear buildings ensures radioactive contamination is confined so that workers, the public and the environment are protected. Nuclear buildings are divided into confinement sectors which connect to the Heating, Ventilation and Air Conditioning (HVAC) system and detritiation system (DS). The Tokamak Complex DS provides centralized air purification for the building confinement sectors. A distributed arrangement of ventilation piping provides networks necessary for two key functions, these being Vent Detritiation (VD), to maintain sub-atmospheric pressure, and Air Detritiation (AD) to collect tritium released into the confinement sector. For the VD function, air extracted from the particular confinement sector is directed to the DS for processing prior to exhaust to the environment. This paper presents the configuration of the DS of the Tokamak Complex and addresses details of the design of the distributed piping network. Dynamic flow and pressure drop modelling has been applied to support the development of the system configuration and provide data for sizing the system and selecting components. Further design development is discussed in view of the safety requirements for operation of the system during design basis events such as earthquake or fire.

  15. Nuclear shielding of openings in ITER Tokamak building

    Energy Technology Data Exchange (ETDEWEB)

    Dammann, A., E-mail: alexis.dammann@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Arumugam, A.P.; Beaudoin, V.; Beltran, D.; Benchikhoune, M.; Berruyer, F.; Cortes, P.; Gandini, F. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ghirelli, N. [ASSYSTEM E.O.S, ZAC Saint Martin, 23, rue Benjamin Franklin, 84120 Pertuis (France); Gray, A.; Hurzlmeier, H.; Le Page, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Lemée, A. [SOGETI High Tech, 180 Rue René Descartes, 13851 Aix en Provence (France); Lentini, G.; Loughlin, M.; Mita, Y.; Patisson, L.; Rigoni, G.; Rathi, D.; Song, I. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: ► Establishment of a methodology to design shielded opening in external wall of the Tokamak building. ► Analysis of the shielding requirement, case by case, depending on the localization and the context. ► Implementation of an integrated solution for shielded opening. -- Abstract: The external walls of the Tokamak building, made of thick concrete, provide the nuclear shielding for operators working in adjacent buildings and for the environment. There are a series of openings to these external walls, devoted to ducts or pipes for ventilation, waveguides and transmission lines for heating systems and diagnostics, cooling pipes, cable trays or busbars. The shielding properties of the wall shall be preserved by adequate design of the openings in order not to affect the radiological zoning in adjacent areas. For some of them, shielding properties of the wall are not affected because the size of the network is quite small or the source is far from the opening. But for most of the openings, specific features shall be considered. Even if the approach is the same and the ways to shield can be standardized, specific analysis is requested in any case because the constraints are different.

  16. Finite pressure effects on the tokamak sawtooth crash

    Energy Technology Data Exchange (ETDEWEB)

    Nishimura, Yasutaro

    1998-07-01

    The sawtooth crash is a hazardous, disruptive phenomenon that is observed in tokamaks whenever the safety factor at the magnetic axis is below unity. Recently, Tokamak Test Fusion Reactor (TFTR) experimental data has revealed interesting features of the dynamical pressure evolution during the crash phase. Motivated by the experimental results, this dissertation focuses on theoretical modeling of the finite pressure effects on the nonlinear stage of the sawtooth crash. The crash phase has been studied numerically employed a toroidal magnetohydrodynamic (MHD) initial value code deduced from the FAR code. For the first time, by starting from a concentric equilibrium, it has been shown that the evolution through an m/n = 1/1 magnetic island induces secondary high-n ballooning instabilities. The magnetic island evolution gives rise to convection of the pressure inside the inversion radius and builds up a steep pressure gradient across the island separatrix, or current sheet, and thereby triggers ballooning instabilities below the threshold for the axisymmetric equilibrium. Due to the onset of secondary ballooning modes, concomitant fine scale vortices and magnetic stochasticity are generated. These effects produce strong flows across the current sheet, and thereby significant modify the m = 1 driven magnetic reconnection process. The resultant interaction of the high-n ballooning modes with the magnetic reconnection process is discussed.

  17. Experimental study of the plasma structure and characterization of the transport behaviour in the laminar zone of a stochastized plasma edge; Experimentelle Untersuchung der Plasmastruktur und Charakterisierung des Transportverhaltens in der laminaren Zone einer stochastisierten Plasmarandschicht

    Energy Technology Data Exchange (ETDEWEB)

    Schmitz, O.

    2006-07-15

    For a detailed study of the plasma structure and the transport characteristics of a stochastized plasma edge at the tokamak TEXTOR the dynamic ergodic divertor (DED) was constructed, by which differently shaped external disturbing fields are statically and dynamically generated. Aim of this thgesis is to study experimentally the radial and poloidal structure of the plasma edge stochastized by the DED disturbing field and to analyze its transport characteristics. For this spatially highly resolved radial profiles of the electron density and temperature were measured by means of radiation-emission spectroscopy on thermal helium at the high- and low-field side of TEXTOR. These experimental results yield a good stating base for the validation and further development of three-dimensional transport codes.

  18. measurements by Thomson scattering system

    Indian Academy of Sciences (India)

    oirity in measuring the electron temperature (Te) and density (ne) in fusion plasma devices like tokamaks. The method is a ... shot to shot basis. This paper discusses the initial measurements of the plasma temperature from ADITYA. Keywords. .... All the in-situ calibrations required for the system are also performed. As the ...

  19. Project and analysis of the toroidal magnetic field production circuits and the plasma formation of the ETE (Spherical Tokamak Experiment) tokamak; Projeto e analise dos circuitos de producao de campo magnetico toroidal e de formacao do plasma do Tokamak ETE (Experimento Tokamak Esferico)

    Energy Technology Data Exchange (ETDEWEB)

    Barbosa, Luis Filipe F.P.W.; Bosco, Edson del

    1994-12-31

    This report presents the project and analysis of the circuit for production of the toroidal magnetic field in the Tokamak ETE (Spherical Tokamak Experiment). The ETE is a Tokamak with a small-aspect-ratio parameter to be used for studying the plasma physics for the research on thermonuclear fusion. This machine is being constructed at the Laboratorio Associado de Plasma (LAP) of the Instituto Nacional de Pesquisas Espaciais (INPE) in Sao Jose dos Campos, SP, Brazil. (author). 20 refs., 39 figs., 4 tabs.

  20. Pellet injection and confinement in the tore supra tokamak; Injection de glacons et confinement dans le tokamak tore supra

    Energy Technology Data Exchange (ETDEWEB)

    Maget, P

    1998-09-23

    Pellet injection in the centre of tokamak plasmas can lead to an improved confinement regime called PEP (Pellet Enhanced Performance). The present work is dedicated to the mechanisms involved in the PEP regimes obtained in the tokamak Tore Supra. A neoclassical approach of transport shows that it is the anomalous transport, due to plasma turbulence, that causes the enhanced confinement. A linear model describing electrostatic instabilities has been developed in order to study the roles of density profile and current profile during the PEP, in the limit of large growth rates. The effect ofradial shear in flows is taken into account by removing the ExB shear flow rate from the linear growth rate, as suggested by non-linear numerical simulations of turbulence. A local transport coefficient is estimated from the knowledge of the linear growth rate and the mode width. We find that the peaked density profile in PEP regime lowers the diffusion coefficient, and that the velocity shear amplifies this effect. The evolution of the current profile is also stabilizing, but this parameter is not known with sufficient accuracy, so that its role in Tore Supra PEP experiments remains uncertain. (author)

  1. Analysis of the plasma turbulence through radar reflectometry in the Tore-Supra tokamak; Analyse de la turbulence de plasma par reflectometrie radar sur le tokamak Tore Supra

    Energy Technology Data Exchange (ETDEWEB)

    Gerbaud, T

    2005-07-01

    The turbulence developing in a tokamak's plasma is liable for a large transport of energy and particles, what slims the plasma magnetic confinement. This turbulence induces electromagnetic fluctuations inside the plasma, which imply local electronic density fluctuations. Using microwave reflectometers 50 - 110 GHz, operating like radars, one can probe the plasma at different depths, and then analyse the wave reflected by the plasma. Probe waves can be polarized ordinarily or extraordinarily, the difference lying in the dispersion relation of the plasma reflection index. The goal of this work is to compare density fluctuations spectrums, obtained in both polarization. Wave numbers spectrums and radials profiles of corresponding RMS values (equivalent to mean quadratic values) allow to conclude on a good agreement between the fluctuations density levels generated by measurement done in ordinary or extraordinary polarization. The comparison of wave numbers spectrums of density fluctuations underlines the growth of turbulence activity in the gradients zone. These results represent the first steps of a advanced analysis of fluctuations profiles and spectrums generated in ordinary polarization. (author)

  2. Fabrication and Characterization of Samples for a Material Migration Experiment on the Experimental Advanced Superconducting Tokamak (EAST).

    Energy Technology Data Exchange (ETDEWEB)

    Wampler, William R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Van Deusen, Stuart B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-12-01

    This report documents work done for the ITER International Fusion Energy Organization (Sponsor) under a Funds-In Agreement FI 011140916 with Sandia National Laboratories. The work consists of preparing and analyzing samples for an experiment to measure material erosion and deposition in the EAST Tokamak. Sample preparation consisted of depositing thin films of carbon and aluminum onto molybdenum tiles. Analysis consists of measuring the thickness of films before and after exposure to helium plasma in EAST. From these measurements the net erosion and deposition of material will be quantified. Film thickness measurements are made at the Sandia Ion Beam Laboratory using Rutherford backscattering spectrometry and nuclear reaction analysis, as described in this report. This report describes the film deposition and pre-exposure analysis. Results from analysis after plasma exposure will be given in a subsequent report.

  3. Development of coupling systems at the hybrid frequency for the non-inductive current generation inside a tokamak; Developpement de coupleurs a la frequence hybride pour la generation non inductive du courant dans un tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Berio, S. [Association Euratom-CEA, Centre d`Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee]|[Aix-Marseille-1 Univ., 13 - Marseille (France)

    1996-12-31

    Used at its first time as an heating method in order to reach the temperature requisite for the fusion of a thermonuclear plasma, the hybrid waves has shown that they were the more efficient method for non-inductive current drive in a tokamak. The size and the objectives of a next machine such as ITER lead to the design of new antennae (in process of realisation on Tore Supra) made of oversized waveguides. This new concept of antenna will be more simple, more robust and will be able to transmit the same if not much power than the present antennae. This thesis contribute to the development of a new code called ALOHA (for `Advanced LOwer Hybrid Antenna`) which, at the end, will be able to give the characteristics and the behaviours of this new oversized antennae in front of a tokamak plasma. This thesis is also a first step in the interpretation of some experimental data concerning the measurement of coupling, absorption and current drive of the actual hybrid wave launched by a grill with rectangular waveguides. Moreover, this thesis lay some foundations of the study of these new antennae in front of a on-parallel confinement magnetic field and/or in front of poloidal inhomogeneities of plasma. (authors) 53 refs.

  4. Development of hybrid frequency couplers for non-inductive current drive in a tokamak; Developpement de coupleurs a la frequence hybride pour la generation non inductive du courant dans un tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Berio, St.

    1996-11-04

    Used at its first time as an heating method in order to reach the temperature requisite for the fusion of a thermonuclear plasma, the hybrid waves has shown that they were the more efficient method for non-inductive current drive in a tokamak. The size and the objectives of a next machine such as ITER lead of the design of new antennae (in process of realisation on Tore Supra) made of oversized waveguides. This new concept of antenna will be more simple, more robust and will be able to transmit the same if not much power than the present antennae. This thesis contribute to the development of a new code called ALOHA (for `Advanced LOwer Hybrid Antenna`) which, at the end, will be able to give the characteristics and the behaviours of this new oversized antennae in front of a tokamak plasma. This thesis is also a first step in the interpretation of some experimental data concerning the measurement of coupling, absorption and current drive of the actual hybrid wave launched by a grill with rectangular waveguides. Moreover, this thesis lay some foundations of the study of these new antennae in front of a non-parallel confinement magnetic field and/or in front of poloidal inhomogeneities of plasma. (author). 53 refs.

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

    CERN Document Server

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

    2015-01-01

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

  6. Geodesic acoustic eigenmode for tokamak equilibrium with maximum of local GAM frequency

    Energy Technology Data Exchange (ETDEWEB)

    Lakhin, V.P. [NRC “Kurchatov Institute”, Moscow (Russian Federation); Sorokina, E.A., E-mail: sorokina.ekaterina@gmail.com [NRC “Kurchatov Institute”, Moscow (Russian Federation); Peoples' Friendship University of Russia, Moscow (Russian Federation)

    2014-01-24

    The geodesic acoustic eigenmode for tokamak equilibrium with the maximum of local GAM frequency is found analytically in the frame of MHD model. The analysis is based on the asymptotic matching technique.

  7. Overview of the ITER Tokamak complex building and integration of plant systems toward construction

    Energy Technology Data Exchange (ETDEWEB)

    Cordier, Jean-Jacques, E-mail: jean-jacques.cordier@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Bak, Joo-Shik [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Baudry, Alain [Engage Consortium, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Benchikhoune, Magali [Fusion For Energy (F4E), c/ Josep Pla, n.2, Torres Diagonal Litoral, E-08019 Barcelona (Spain); Carafa, Leontin; Chiocchio, Stefano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Darbour, Romaric [Fusion For Energy (F4E), c/ Josep Pla, n.2, Torres Diagonal Litoral, E-08019 Barcelona (Spain); Elbez, Joelle; Di Giuseppe, Giovanni; Iwata, Yasuhiro; Jeannoutot, Thomas; Kotamaki, Miikka; Kuehn, Ingo; Lee, Andreas; Levesy, Bruno; Orlandi, Sergio [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Packer, Rachel [Engage Consortium, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Patisson, Laurent; Reich, Jens; Rigoni, Giuliano [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); and others

    2015-10-15

    The ITER Tokamak complex consists of Tokamak, diagnostic and tritium buildings. The Tokamak machine is located in the bioshield pit of the Tokamak building. Plant systems are implemented in the three buildings and are strongly interfacing with the Tokamak. The reference baseline (3D) configuration is a set of over 1000 models that today defines in an exhaustive way the overall layout of Tokamak and plant systems, needed for fixing the interfaces and to complete the construction design of the buildings. During the last two years, one of the main ITER challenges was to improve the maturity of the plant systems layout in order to confirm their integration in the building final design and freeze the interface definitions in-between the systems and to the buildings. The propagation of safety requirements in the design of the nuclear building like confinement, fire zoning and radiation shielding is of first priority. A major effort was placed by ITER Organization together with the European Domestic Agency (F4E) and the Architect Engineer as a joint team to fix the interfaces and the loading conditions to buildings. The most demanding systems in terms of interface definition are water cooling, cryogenic, detritiation, vacuum, cable trays and building services. All penetrations through the walls for piping, cables and other equipment have been defined, as well as all temporary openings needed for the installation phase. Project change requests (PCR) impacting the Tokamak complex buildings have been implemented in a tight allocated time schedule. The most demanding change was to implement a new design of the Tokamak basic machine supporting system. The 18 supporting columns of the cryostat (2001 baseline) were replaced at the end of 2012 by a concrete crown and radial concrete ribs linked to the basemat and to the bioshield surrounding the Tokamak. The change was implemented successfully in the building construction design to allow basemat construction phase being performed

  8. Investigation on the Hard X-ray Radiations of the IR-T1 Tokamak Plasma: Electric and Magnetic Perspectives

    Science.gov (United States)

    Alipour, R.; Ghoranneviss, M.; Salar Elahi, A.

    2017-12-01

    In this experiment, the effect of magnetohydrodynamic (MHD) fluctuations in the hard X-ray radiation from the IR-T1 tokamak plasma is investigated. To reach this goal, the main parameters of plasma such as plasma current and loop voltage are measured. Also, the rake and poloidal Langmuir probes are used to calculate the radial and poloidal electric fields. To detect the hard X-ray radiation, a NaI-scintillator detector is used. To study on the MHD fluctuations, an array of 12 Mirnov coils is used. The obtained data are analyzed by using the singular value decomposition (SVD) algorithm. The wavelet spectrum of the dominant principal components of Mirnov coils is drawn. The results of wavelet and SVD analysis show that the hard X-ray radiation is increased with increasing the fluctuations of the dominant principal components (at the same time). It is also shown that the rate of hard X-ray radiation emitted from the tokamak plasma increased with increasing the MHD fluctuations. The energy of the system is wasted and reduced by these radiations. This an increase in the particle pressure of the plasma.

  9. Outward particle transport by coherent mode in the H-mode pedestal in the Experimental Advanced Superconducting Tokamak (EAST)

    Science.gov (United States)

    Zhang, T.; Han, X.; Gao, X.; Liu, H. Q.; Shi, T. H.; Liu, J. B.; Liu, Y.; Kong, D. F.; Liu, Z. X.; Qu, H.; Xiang, H. M.; Geng, K. N.; Wang, Y. M.; Wen, F.; Zhang, S. B.; Ling, B. L.; the EAST Team

    2017-06-01

    A coherent mode (CM) in the edge pedestal region has been observed on different fluctuation quantities, including density fluctuation, electron temperature fluctuation and magnetic fluctuation in H mode plasma on the Experimental Advanced Superconducting Tokamak (EAST) tokamak. Measurements at different poloidal positions show that the local poloidal wavenumber is smallest at the outboard midplane and will increase with poloidal angle. This poloidal asymmetry is consistent with the flute-like assumption (i.e. k// ˜ 0) from which the toroidal mode number of the mode has been estimated as between 12 and 17. It was further found that the density fluctuation amplitude of the CM also demonstrated poloidal asymmetry. The appearance of a CM can clearly decrease or even stop the increase in the edge density, while the disappearance of a CM will lead to an increase in the pedestal density and density gradient. Statistical analysis showed there was a trend that as the CM mode amplitude increased, the rate of increase of the edge density decreased and the particle flux (Γdiv) onto the divertor plate increased. The CM sometimes showed burst behavior, and these bursts led bursts on Γdiv with a time of about 230 μs, which is close to the time for particle flow from the outer midplane to the divertor targets along the scrape-off layer magnetic field line. This evidence showed that the CM had an effect on the outward transport of particles.

  10. Emission in the 50-80 A region from highly ionized silver in PLT tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Schwob, J.L.; Wouters, A.; Suckewer, S.; Cohen, S.A.; Finkenthal, M.

    1985-09-01

    The spectrum of silver emitted by Princeton Large Torus (PLT) tokamak plasmas has been recorded in the 25 to 150 A region by a multichannel time-resolving grazing-incidence spectrometer. Silver atoms have been introduced in the tokamak plasma using the laser blow-off technique. For the first time, lines emitted within the 3p-3d transitions of Ag XXIX, Ag XXX, and Ag XXXI ions, between 50 and 80 A, have been identified.

  11. Fractional variational problems and particle in cell gyrokinetic simulations with fuzzy logic approach for tokamaks

    Directory of Open Access Journals (Sweden)

    Rastović Danilo

    2009-01-01

    Full Text Available In earlier Rastovic's papers [1] and [2], the effort was given to analyze the stochastic control of tokamaks. In this paper, the deterministic control of tokamak turbulence is investigated via fractional variational calculus, particle in cell simulations, and fuzzy logic methods. Fractional integrals can be considered as approximations of integrals on fractals. The turbulent media could be of the fractal structure and the corresponding equations should be changed to include the fractal features of the media.

  12. Plasma position control in the STOR-M tokamak: A fuzzy logic approach

    Science.gov (United States)

    Morelli, Jordan Edwin

    Adequate control of the position of the plasma column within the STOR-M tokamak is a chief requirement in order for experimental quality discharges to be obtained. Optimal control over tokamak discharge parameters, including the plasma position, is very difficult to achieve. This is due in large part to the difficulty in modelling the tokamak discharge parameters, as they are highly nonlinear and time varying in nature. The difficulty of modelling the tokamak discharge parameters suggests that a control system, such as a fuzzy logic based controller, which does not require a system model may be well suited to the control of fusion plasma. In order to improve the quality of control over the plasma position within the STOR-M tokamak, the existing analog PID controller was modified. These modifications facilitate the application of a digital controller by a personal computer via the Advantech PCL-711B data acquisition card. The performance of the modified plasma position controller and an Arbitrary Signal Generator developed by the author was evaluated. This modified plasma position controller was applied successfully to the STOR-M tokamak during both normal mode and A.C. mode operation. In both cases, the modified controller provided adequate control over the position of the plasma column within the discharge chamber. Furthermore, the modified controller was more convenient to optimize than the original, existing analog PID controller. By taking advantage of the modifications that were made to the plasma position controller, a fuzzy logic controller was developed by the author. The fuzzy logic based plasma position controller was also successfully applied to the STOR-M tokamak during both normal mode and A.C. operation. The fuzzy controller was demonstrated to reliably provide a higher degree of control over the position of the plasma column within the STOR-M tokamak than the modified PID controller.

  13. Transport of fast electrons in lower hybrid current drive plasmas in the HT-7 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Z Y [Institute of Plasma Physics, College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Fang, D; Dai, F; Duan, Z Q; Zhu, J X; Sun, W M [Department of Physics, Yunnan Normal University, Kunming 650092 (China); Wan, B N; Shi, Y J, E-mail: chenzy1003@163.com [Institute of Plasma Physics, Chinese Academy of Sciences, Hefe 230031 (China)

    2011-04-15

    The transport of fast electrons in lower hybrid current drive (LHCD) plasmas in the HT-7 tokamak was investigated in this work. The evolution of fast electron bremsstrahlung emission profiles after switching off the lower hybrid power was analyzed. We found that the dynamics of the fast electrons is governed by the slowing-down process, and the current density profile can be controlled by LHCD in the HT-7 tokamak.

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

    CERN Document Server

    Rampp, Markus; Fischer, Rainer

    2015-01-01

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

  15. Tokamak fusion reactors with less than full tritium breeding

    Energy Technology Data Exchange (ETDEWEB)

    Evans, K. Jr.; Gilligan, J.G.; Jung, J.

    1983-05-01

    A study of commercial, tokamak fusion reactors with tritium concentrations and tritium breeding ratios ranging from full deuterium-tritium operation to operation with no tritium breeding is presented. The design basis for these reactors is similar to those of STARFIRE and WILDCAT. Optimum operating temperatures, sizes, toroidal field strengths, and blanket/shield configurations are determined for a sequence of reactor designs spanning the range of tritium breeding, each having the same values of beta, thermal power, and first-wall heat load. Additional reactor parameters, tritium inventories and throughputs, and detailed costs are calculated for each reactor design. The disadvantages, advantages, implications, and ramifications of tritium-depleted operation are presented and discussed.

  16. WILDCAT: a catalyzed D-D tokamak reactor

    Energy Technology Data Exchange (ETDEWEB)

    Evans, K. Jr.; Baker, C.C.; Brooks, J.N.

    1981-11-01

    WILDCAT is a conceptual design of a catalyzed D-D, tokamak, commercial, fusion reactor. WILDCAT utilizes the beneficial features of no tritium breeding, while not extrapolating unnecessarily from existing D-T designs. The reactor is larger and has higher magnetic fields and plasma pressures than typical D-T devices. It is more costly, but eliminates problems associated with tritium breeding and has tritium inventories and throughputs approximately two orders of magnitude less than typical D-T reactors. There are both a steady-state version with Alfven-wave current drive and a pulsed version. Extensive comparison with D-T devices has been made, and cost and safety analyses have been included. All of the major reactor systems have been worked out to a level of detail appropriate to a complete, conceptual design.

  17. First results from gamma ray diagnostics in EAST Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, R. J.; Hu, L. Q.; Zhong, G. Q., E-mail: gqzhong@ipp.ac.cn; Cao, H. R.; Liu, G. Z.; Li, K.; Zhang, Y.; Lin, S. Y.; Zhang, J. Z. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2016-11-15

    Gamma ray diagnostics has been developed in the EAST tokamak recently. Six BGO scintillator detectors are arranged on the down-half cross-section and pointed at the up-half cross-section of plasma, with space resolution about 15 cm and energy range from 0.3 MeV to 6 MeV. Three main gamma ray peaks in the energy spectra have been observed and are identified as the results of nuclear reactions {sup 207}Pb(n, n′){sup 207m}Pb, H(n, γ) D, and D(p, γ){sup 3}He, respectively. Upgrading of the system is in progress by using LaBr3(Ce) scintillator, fast photo-multiplier tubes, and a fully digital data acquisition system based on high sample frequency digitizers with digital pulse processing algorithms.

  18. Neutral beam optimisation for the spherical tokamak ST40

    Energy Technology Data Exchange (ETDEWEB)

    Salmi, A., E-mail: antti.salmi@vtt.fi [VTT, P.O. Box 1000, FIN-02044 VTT (Finland); Gryaznevich, M.; Buxton, P.; Nightingale, M. [Tokamak Energy Ltd., Culham Science Centre, Abingdon, Oxon OX143DB (United Kingdom); Tala, T. [VTT, P.O. Box 1000, FIN-02044 VTT (Finland)

    2017-04-15

    Orbit following Monte Carlo code (ASCOT) calculations of neutral beam (NB) induced torque, current and heating have been performed for a selected set of plasma scenarios to optimise the neutral beam injection system for the proposed high magnetic field spherical tokamak ST40. It is found that there are strong variations especially in the current drive efficiency and in the toroidal torque depending on the NB alignment and injection energy. The optimal alignments as well as the optimal injection energy depend both on the scenario and whether the current drive or the toroidal torque is to be maximised. Due to the relatively small calculated current drive efficiency (0.05–0.2 MA/MW) we find that for the studied scenarios a relatively central low field side (LFS) deposition below midplane provides the best overall performance still at a tolerable shine through power.

  19. Transport properties of interacting magnetic islands in tokamak plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Gianakon, T.A.; Callen, J.D.; Hegna, C.C.

    1993-10-01

    This paper explores the equilibrium and transient transport properties of a mixed magnetic topology model for tokamak equilibria. The magnetic topology is composed of a discrete set of mostly non-overlapping magnetic islands centered on the low-order rational surfaces. Transport across the island regions is fast due to parallel transport along the stochastic magnetic field lines about the separatrix of each island. Transport between island regions is assumed to be slow due to a low residual cross-field transport. In equilibrium, such a model leads to: a nonlinear dependence of the heat flux on the pressure gradient; a power balance diffusion coefficient which increases from core to edge; and profile resiliency. Transiently, such a model also exhibits a heat pulse diffusion coefficient larger than the power balance diffusion coefficient.

  20. Maintenance features of the Compact Ignition Tokamak fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Spampinato, P.T.; Hager, E.R.

    1987-01-01

    The Compact Ignition Tokamak (CIT) is envisaged to be the next experimental machine in the US Fusion Program. Its use of deuterium/tritium fuel requires the implementation of remote handling technology for maintenance and disassembly operations. The reactor is surrounded by a close-proximity nuclear shield which is designed to permit personnel access within the test cell, one day after shutdown. With the shield in place, certain maintenance activities in the cell may be done hands-on. Maintenance on the reactor is accomplished remotely using a boom-mounted manipulator after disassembling the shield. Maintenance within the plasma chamber is accomplished with two articulated boom manipulators that are capable of operating in a vacuum environment. They are stored in a vacuum enclosure behind movable shield plugs.

  1. STARFIRE: a commercial tokamak fusion power plant study

    Energy Technology Data Exchange (ETDEWEB)

    1980-09-01

    STARFIRE is a 1200 MWe central station fusion electric power plant that utilizes a deuterium-tritium fueled tokamak reactor as a heat source. Emphasis has been placed on developing design features which will provide for simpler assembly and maintenance, and improved safety and environmental characteristics. The major features of STARFIRE include a steady-state operating mode based on continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum system for impurity control and exhaust, high tritium burnup and low vulnerable tritium inventories, superconducting EF coils outside the superconducting TF coils, fully remote maintenance, and a low-activation shield. A comprehensive conceptual design has been developed including reactor features, support facilities and a complete balance of plant. A construction schedule and cost estimate are presented, as well as study conclusions and recommendations.

  2. Anomalous Transport in Tokamaks: Examples of Its Suppression

    Science.gov (United States)

    Hirose, A.; Ding, W.; Elia, M.; Xiao, C.; Yamagiwa, M.

    Studies conducted towards an understanding of improved plasma confinement in small and large tokamaks are presented. In STOR-M, Ohmic H-modes can be induced by various means. The most recent finding is that injection of a high density compact torus triggers a sustained H-mode. For bias induced H-mode, direct evaluation of fluctuation induced particle flux has been made for Ohmic H-modes. The particle flux is intermittent and the average changes its sign from radially outward in L-mode to radially inward in H-mode. The direction and magnitude of the flux are predominantly determined by the phase difference between the density and potential fluctuations. Nonlinear correlation studies have also been made between two radial positions. In L-mode, the inner region behaves as a master and outer region as slave. In H-mode, the relationship is reversed.

  3. On q dependence of thermal transport in tokamaks

    Science.gov (United States)

    Hirose, A.; Livingstone, S.; Singh, A. K.

    2005-12-01

    Analysis based on a gyro-kinetic ballooning stability code predicts that both the ion and electron thermal diffusivities, due to the ion temperature gradient (ITG) and electron temperature gradient (ETG) modes, respectively, increase with the safety factor q almost linearly. In the case of ITG driven ion thermal diffusivity, the q dependence originates from the coupling to the ion acoustic mode, and in the case of the electron thermal diffusivity due to the ETG mode, it emerges from the coupling to the skin size drift mode. In the ETG mode, charge neutrality does not hold for typical tokamak discharges, and mixing length estimates yield a thermal diffusivity large enough to be relevant to experiments.

  4. A Distributed Synchronization and Timing System on the EAST Tokamak

    Science.gov (United States)

    Luo, Jiarong; Wu, Yichun; Shu, Yantai

    2008-08-01

    A key requirement for the EAST distributed control system (EASTDCS) is time synchronization to an accuracy of AVR microcontroller and the Nut/OS embedded Real Time Operating System (RTOS). The DSTS provides the control and the data acquisition systems with reference clocks (0.01 Hz 10 MHz) and delayed trigger times ( 1 mus 4294 s). These are produced by a Core Module Unit (CMU) connected by optical fibres to many Local Synchronized Node Units (LSNU). The fibres provide immunity from electrical noise and are of equal length to match clock and trigger delays between systems. This paper describes the architecture of the DSTS on the EAST tokamak and provides an overview of the characteristics of the main and local units.

  5. Gyrokinetic modelling of stationary electron and impurity profiles in tokamaks

    CERN Document Server

    Skyman, Andreas; Tegnered, Daniel

    2014-01-01

    Particle transport due to Ion Temperature Gradient/Trapped Electron (ITG/TE) mode turbulence is investigated using the gyrokinetic code GENE. Both a reduced quasilinear (QL) treatment and nonlinear (NL) simulations are performed for typical tokamak parameters corresponding to ITG dominated turbulence. A selfconsistent treatment is used, where the stationary local profiles are calculated corresponding to zero particle flux simultaneously for electrons and trace impurities. The scaling of the stationary profiles with magnetic shear, safety factor, electron-to-ion temperature ratio, collisionality, toroidal sheared rotation, triangularity, and elongation is investigated. In addition, the effect of different main ion mass on the zero flux condition is discussed. The electron density gradient can significantly affect the stationary impurity profile scaling. It is therefore expected, that a selfconsistent treatment will yield results more comparable to experimental results for parameter scans where the stationary b...

  6. Plasma recombination in runaway discharges in tokamak TCABR

    Energy Technology Data Exchange (ETDEWEB)

    Soboleva, T.K. [Universidad Nacional Autonoma de Mexico, Mexico City (Mexico). Inst. de Ciencias Nucleares; Galvao, R.M.O.; Kuznetsov, Yu. K.; Nascimento, I.C. [Sao Paulo Univ., SP (Brazil). Inst. de Fisica; Krasheninnikov, S.I. [University of California, San Diego, CA (United States)

    2002-03-01

    A new regime of runaway discharges has been observed in the TCABR tokamak. One of the most distinctive features of this regime is the effect of plasma detachment from the limiter. This experimental fact can only be explained by the volume recombination, which requires a low-temperature plasma. The analysis of the energy and particle balance in the system plasma-relativistic runaway beam in TCABR, which takes into account only the collisional mechanism of the heat transfer from runaways to thermal electrons, predicts electron temperatures T{sub e} = 0.1 - 2 eV; the temperature decreases with the neutral density increase. The recombination process with the rate constant around 10{sup -16} m3/s is required for the explanation of plasma density behavior in the experiment. At present, it is difficult to conclude about the mechanism of recombination. More reliable and detailed experimental data, mainly about the plasma temperature, are necessary. (author)

  7. Kinetic effects and reversal flows in the tokamak edge plasma

    Energy Technology Data Exchange (ETDEWEB)

    Krasheninnikov, S.I.; Soboleva, T.K. (Russian Scientific Center, Kurchatov Inst., Moscow (Russia)); Batischev, O.V.; Zmievskaya, G.I.; Levchenko, V.D.; Ovchenkov, P.A.; Sigov, Yu.S.; Silaev, I.I. (M.V. Keldysh Inst. of Applied Mathematics, Russian Academy of Science, Moscow (Russia))

    1992-12-01

    Preliminary results of the edge tokamak plasma simulation in the frame of 3D kinetic Vlasov-Fokker-Planck equations are presented. Currently two versions of the code are developed: One of them is based on a stochastic modeling method, another uses a finite difference approach via split techniques. In both versions, the self-consistent electric field may be calculated from the quasi-neutrality condition, while the sheath potential is obtained from the ambipolarity of plasma flux towards the neutralizing plate. Simplified equations, describing the dense SOL plasma parameters distributions and allowing the analytical solutions are obtained. It is shown, that for the high particle recycling in SOL, the radial scales of the temperature [Delta][sub T] and densities, [Delta][sub n] are close to each other. The reversal flow is formed within a narrow layer near the separatrix magnetic surface. (orig.).

  8. Divertor research on the DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Hill, D.N.; Allen, S.L. [Lawrence Livermore National Lab., CA (United States); Brooks, N.H. [General Atomics, San Diego, CA (United States)] [and others

    1994-10-01

    In this paper the authors summarize recent progress on DIII-D in developing techniques for divertor power and particle control relevant to next generation tokamaks such as the proposed ITER and TPX devices. Density control and helium removal by divertor pumping have been demonstrated for the first time in high confinement ELMing H-mode discharges ({tau} {approximately} 2 {times} {tau}{sub ITER-89P}) following installation of a divertor cryopumping system. The peak divertor heat flux in similar H-mode discharges has been reduced through production of a radiating mantle with neon or argon puffing (reductions of 3--5). A number of diagnostics have been added to improve the understanding of the physical processes involved. They are now designing modified double-null divertor structures for DIII-D that will provide improved particle control for high-triangularity VH-mode plasmas while at the same time allowing for gas puffing to reduce the divertor heat flux.

  9. Equilibrium system analysis in a tokamak ignition experiment. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Carrera, R.; Weldon, W.F.; Woodson, H.H.

    1989-10-01

    The objective of the IGNITEX Project is to produce and control ignited plasmas for scientific study in the simplest and least expensive way possible. The original concept was proposed by both physics and engineering researchers along the following line of thought. Question: Is there any theoretically simple, compact and reliable way of achieving fusion ignition according to the results of the fusion research program for the last decades? Answer: Yes. An experiment to be carried out in an ohmically heated compact tokamak device with 20 T field on plasma axis. Question: Is there any practical way to carry out that experiment at low cost in the near term? Answer: Yes. Using a single-turn coil magnet system with homopolar power supplies.

  10. Equilibrium system analysis in a tokamak ignition experiment

    Energy Technology Data Exchange (ETDEWEB)

    Carrera, R.; Weldon, W.F.; Woodson, H.H.

    1989-10-01

    The objective of the IGNITEX Project is to produce and control ignited plasmas for scientific study in the simplest and least expensive way possible. The original concept was proposed by both physics and engineering researchers along the following line of thought. Question: Is there any theoretically simple, compact and reliable way of achieving fusion ignition according to the results of the fusion research program for the last decades Answer: Yes. An experiment to be carried out in an ohmically heated compact tokamak device with 20 T field on plasma axis. Question: Is there any practical way to carry out that experiment at low cost in the near term Answer: Yes. Using a single-turn coil magnet system with homopolar power supplies.

  11. Micro-tearing modes in the Mega Ampere Spherical Tokamak

    CERN Document Server

    Applegate, D J; Connor, J W; Cowley, S C; Dorland, W; Hastie, R J; Joiner, N; 10.1088/0741-3335/49/8/001

    2011-01-01

    Recent gyrokinetic stability calculations have revealed that the spherical tokamak is susceptible to tearing parity instabilities with length scales of a few ion Larmor radii perpendicular to the magnetic field lines. Here we investigate this 'micro-tearing' mode in greater detail to uncover its key characteristics, and compare it with existing theoretical models of the phenomenon. This has been accomplished using a full numerical solution of the linear gyrokinetic-Maxwell equations. Importantly, the instability is found to be driven by the free energy in the electron temperature gradient as described in the literature. However, our calculations suggest it is not substantially affected by either of the destabilising mechanisms proposed in previous theoretical models. Instead the instability is destabilised by interactions with magnetic drifts, and the electrostatic potential. Further calculations reveal that the mode is not significantly destabilised by the flux surface shaping or the large trapped particle f...

  12. A general comparison between tokamak and stellarator plasmas

    Directory of Open Access Journals (Sweden)

    Yuhong Xu

    2016-07-01

    Full Text Available This paper generally compares the essential features between tokamaks and stellarators, based on previous review work individually made by authors on several specific topics, such as theories, bulk plasma transport and edge divertor physics, along with some recent results. It aims at summarizing the main results and conclusions with regard to the advantages and disadvantages in these two types of magnetic fusion devices. The comparison includes basic magnetic configurations, magnetohydrodynamic (MHD instabilities, operational limits and disruptions, neoclassical and turbulent transport, confinement scaling and isotopic effects, plasma rotation, and edge and divertor physics. Finally, a concept of quasi-symmetric stellarators is briefly referred along with a comparison of future application for fusion reactors.

  13. First results obtained from the soft x-ray pulse height analyzer on experimental advanced superconducting tokamak.

    Science.gov (United States)

    Xu, P; Lin, S Y; Hu, L Q; Duan, Y M; Zhang, J Z; Chen, K Y; Zhong, G Q

    2010-06-01

    An assembly of soft x-ray pulse height analyzer system, based on silicon drift detector (SDD), has been successfully established on the experimental advanced superconducting tokamak (EAST) to measure the spectrum of soft x-ray emission (E=1-20 keV). The system, including one 15-channel SDD linear array, is installed on EAST horizontal port C. The time-resolved radial profiles of electron temperature and K(alpha) intensities of metallic impurities have been obtained with a spatial resolution of around 7 cm during a single discharge. It was found that the electron temperatures derived from the system are in good agreement with the values from Thomson scattering measurements. The system can also be applied to the measurement of the long pulse discharge for EAST. The diagnostic system is introduced and some typical experimental results obtained from the system are also presented.

  14. Experimental Study of Reversed Shear Alfven Eigenmodes During The Current Ramp In The Alcator C-Mod Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Edlund, E. M.; Porkolab, M.; Kramer, G. J.; Lin, L.; Lin, Y.; Tsuji, N.; Wukitch, S. J.

    2010-08-27

    Experiments conducted in the Alcator C-Mod tokamak at MIT have explored the physics of reversed shear Alfven eigenmodes (RSAEs) during the current ramp. The frequency evolution of the RSAEs throughout the current ramp provides a constraint on the evolution of qmin, a result which is important in transport modeling and for comparison with other diagnostics which directly measure the magnetic field line structure. Additionally, a scaling of the RSAE minimum frequency with the sound speed is used to derive a measure of the adiabatic index, a measure of the plasma compressibility. This scaling bounds the adiabatic index at 1.40 ± 0:15 used in MHD models and supports the kinetic calculation of separate electron and ion compressibilities with an ion adiabatic index close to 7~4.

  15. Development of the saddle loop sensors on the J-TEXT tokamak

    Directory of Open Access Journals (Sweden)

    Daojing Guo

    2017-10-01

    Full Text Available To measure the amplitude and phase of the non-axisymmetric radial magnetic field generated by the locked mode, 12 saddle loop sensors are newly developed on the J-TEXT Tokamak. The saddle loop is made of flexible printed circuit (FPC to adapt the complex installment environment and ensure the installment accuracy. In the experiment, the saddle loop measures the radial magnetic field of locked mode and the axisymmetric equilibrium magnetic fields as well as that of the corresponding eddy current. Precise compensation for the fluxes induced by the horizontal and vertical field coils is realized by utilizing the lumped eddy current circuits based on an analytical model. By using this set of saddle loop sensors, the amplitude and phase of the m/n = 2/1 locked mode are clearly measured for the case of error field locking with slow rotation and the penetration of resonant magnetic perturbations (RMPs respectively. Here, m and n are the poloidal and toroidal mode number.

  16. Near-infrared spectroscopy for divertor plasma diagnosis and control in DIII-D tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V. A., E-mail: vlad@llnl.gov; McLean, A. G.; Allen, S. L. [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States)

    2014-11-15

    New near infrared (NIR) spectroscopic measurements performed in the DIII-D tokamak divertor plasma suggest new viable diagnostic applications: divertor recycling and low-Z impurity flux measurements, a spectral survey for divertor Thomson scattering (DTS) diagnostic, and T{sub e} monitoring for divertor detachment control. A commercial 0.3 m spectrometer coupled to an imaging lens via optical fiber and a InGaAs 1024 pixel array detector enabled deuterium and impurity emission measurements in the range 800–2300 nm. The first full NIR survey identified D, He, B, Li, C, N, O, Ne lines and provided plasma T{sub e}, n{sub e} estimates from deuterium Paschen and Brackett series intensity and Stark line broadening analysis. The range 1.000–1.060 mm was surveyed in high-density and neon seeded divertor plasmas for spectral background emission studies for λ = 1.064 μm laser-based DTS development. The ratio of adjacent deuterium Paschen-α and Brackett Br9 lines in recombining divertor plasmas is studied for divertor T{sub e} monitoring aimed at divertor detachment real-time feedback control.

  17. Near-infrared spectroscopy for divertor plasma diagnosis and control in DIII-D tokamak.

    Science.gov (United States)

    Soukhanovskii, V A; McLean, A G; Allen, S L

    2014-11-01

    New near infrared (NIR) spectroscopic measurements performed in the DIII-D tokamak divertor plasma suggest new viable diagnostic applications: divertor recycling and low-Z impurity flux measurements, a spectral survey for divertor Thomson scattering (DTS) diagnostic, and Te monitoring for divertor detachment control. A commercial 0.3 m spectrometer coupled to an imaging lens via optical fiber and a InGaAs 1024 pixel array detector enabled deuterium and impurity emission measurements in the range 800-2300 nm. The first full NIR survey identified D, He, B, Li, C, N, O, Ne lines and provided plasma Te, ne estimates from deuterium Paschen and Brackett series intensity and Stark line broadening analysis. The range 1.000-1.060 mm was surveyed in high-density and neon seeded divertor plasmas for spectral background emission studies for λ = 1.064 μm laser-based DTS development. The ratio of adjacent deuterium Paschen-α and Brackett Br9 lines in recombining divertor plasmas is studied for divertor Te monitoring aimed at divertor detachment real-time feedback control.

  18. Upgrade of the Mirnov probe arrays on the J-TEXT tokamak.

    Science.gov (United States)

    Guo, Daojing; Hu, Qiming; Li, Da; Shen, Chengshuo; Wang, Nengchao; Huang, Zhuo; Huang, Mingxiang; Ding, Yonghua; Xu, Guo; Yu, Qingquan; Tang, Yuejin; Zhuang, Ge

    2017-12-01

    The magnetic diagnostic of Mirnov probe arrays has been upgraded on the J-TEXT tokamak to measure the magnetohydrodynamic instabilities with higher spatial resolution and better amplitude-frequency characteristics. The upgraded Mirnov probe array contains one poloidal array with 48 probe modules and two toroidal arrays with 25 probe modules. Each probe module contains two probes which measure both the poloidal and the radial magnetic fields (Bp and Br). To ensure that the Mirnov probe possess better amplitude-frequency characteristics, a novel kind of Mirnov probe made of low temperature co-fired ceramics is utilized. The parameters and frequency response of the probe are measured and can meet the experiment requirement. The new Mirnov arrays have been normally applied for a round of experiments, including the observation of tearing modes and their coupling as well as high frequency magnetic perturbation due to the Alfvén eigenmode. In order to extract useful information from raw signals, visualization processing methods based on singular value decomposition and cross-power spectrum are applied to decompose the coupled modes and to determine the mode number.

  19. Reynolds Stress in the Plasma Boundary of the STOR-M Tokamak

    Science.gov (United States)

    Singh, Ajay K.; Liu, Dazhi; Livingstone, Stephen; Xiao, Chijjin; Akira, Hirose

    2004-11-01

    Sheared plasma flows are considered to play an important role in regulating the plasma transport and in the transition of low to high confinement regime in magnetically confined plasmas. There are several mechanisms proposed to explain the generation of sheared poloidal flows. One of the mechanism is through the generation of mean flow from plasma fluctuations via the Reynold's stress. From the examination of momentum balance equation it is apparent that poloidal flow may be nonlinearly accelerated if the turbulent Reynold's stress is finite and has a radial gradient. We present measurements of the radial profile of turbulent Reynolds stress in the boundary of STOR--M tokamak using Langmuir probes. If the plasma flow is dictated by E× B drift, the electrostatic Reynolds stress component is proportional to . The measurements show that the Reynolds stress has a radial gradient close to the velocity shear layer location. The statistical characteristics of fluctuations in the vicinity of this region suggests that gradient in Reynolds stress might drive significant poloidal flow in the plasma edge region. Study is also underway to measure Reynold's stress in the toroidal direction using a Mach probe as well as Langmuir probes.

  20. Stability and properties of electron-driven fishbones in tokamaks

    Science.gov (United States)

    Merle, Antoine

    2013-01-01

    In tokamaks, the stability of magneto-hydrodynamic modes can be modified by populations of energetic particles. In ITER-type fusion reactors, such populations can be generated by fusion reactions or auxiliary heating. The electron-driven fishbone mode results from the resonant interaction of the internal kink mode with the slow toroidal precessional motion of energetic electrons and is frequently observed in present-day tokamaks with Electron Cyclotron Resonance Heating or Lower Hybrid Current Drive. In Tore Supra, electron-driven fishbones are observed during LHCD-powered discharges in which a high-energy tail of the electronic distribution function is created. Although the destabilization of those modes is related to the existence of a fast particle population, the modes are observed at a frequency that is lower than expected. The linear stability analysis of electron-driven fishbone modes is the main focus of this thesis. The fishbone dispersion relation is derived in a form that accounts for the contribution of the parallel motion of passing particles to the resonance condition. The MIKE code is developed to compute and solve the dispersion relation of electron-driven fishbones. The code is successfully benchmarked against theory using simple analytical distributions. Using the code MIKE with parametric distributions and equilibria, we show that both barely trapped and barely passing electrons resonate with the mode and can drive it unstable. More deeply trapped and passing electrons have a non-resonant effect on the mode that is, respectively, stabilizing and destabilizing. MIKE simulations using complete ECRH-like distribution functions show that energetic barely passing electrons can contribute to drive a mode unstable at a relatively low frequency. This observation could provide some insight to the understanding of Tore Supra experiments.

  1. Transport in the plateau regime in a tokamak pedestal

    Science.gov (United States)

    Seol, J.; Shaing, K. C.

    2012-07-01

    In a tokamak H-mode, a strong E × B flow shear is generated during the L-H transition. Turbulence in a pedestal is suppressed significantly by this E × B flow shear. In this case, neoclassical transport may become important. The neoclassical fluxes are calculated in the plateau regime with the parallel plasma flow using their kinetic definitions. In an axisymmetric tokamak, the neoclassical particles fluxes can be decomposed into the banana-plateau flux and the Pfirsch-Schlüter flux. The banana-plateau particle flux is driven by the parallel viscous force and the Pfirsch-Schlüter flux by the poloidal variation of the friction force. The combined quantity of the radial electric field and the parallel flow is determined by the flux surface averaged parallel momentum balance equation rather than requiring the ambipolarity of the total particle fluxes. In this process, the Pfirsch-Schlüter flux does not appear in the flux surface averaged parallel momentum equation. Only the banana-plateau flux is used to determine the parallel flow in the form of the flux surface averaged parallel viscosity. The heat flux, obtained using the solution of the parallel momentum balance equation, decreases exponentially in the presence of sonic Mp without any enhancement over that in the standard neoclassical theory. Here, Mp is a combination of the poloidal E × B flow and the parallel mass flow. The neoclassical bootstrap current in the plateau regime is presented. It indicates that the neoclassical bootstrap current also is related only to the banana-plateau fluxes. Finally, transport fluxes are calculated when Mp is large enough to make the parallel electron viscosity comparable with the parallel ion viscosity. It is found that the bootstrap current has a finite value regardless of the magnitude of Mp.

  2. Neutral particle dynamics in the Alcator C-Mod tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Niemczewski, Artur P. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    1995-08-01

    This thesis presents an experimental study of neutral particle dynamics in the Alcator C-Mod tokamak. The primary diagnostic used is a set of six neutral pressure gauges, including special-purpose gauges built for in situ tokamak operation. While a low main chamber neutral pressure coincides with high plasma confinement regimes, high divertor pressure is required for heat and particle flux dispersion in future devices such as ITER. Thus we examine conditions that optimize divertor compression, defined here as a divertor-to-midplane pressure ratio. We find both pressures depend primarily on the edge plasma regimes defined by the scrape-off-layer heat transport. While the maximum divertor pressure is achieved at high core plasma densities corresponding to the detached divertor state, the maximum compression is achieved in the high-recycling regime. Variations in the divertor geometry have a weaker effect on the neutral pressures. For otherwise similar plasmas the divertor pressure and compression are maximum when the strike point is at the bottom of the vertical target plate. We introduce a simple flux balance model, which allows us to explain the divertor neutral pressure across a wide range of plasma densities. In particular, high pressure sustained in the detached divertor (despite a considerable drop in the recycling source) can be explained by scattering of neutrals off the cold plasma plugging the divertor throat. Because neutrals are confined in the divertor through scattering and ionization processes (provided the mean-free-paths are much shorter than a typical escape distance) tight mechanical baffling is unnecessary. The analysis suggests that two simple structural modifications may increase the divertor compression in Alcator C-Mod by a factor of about 5. Widening the divertor throat would increase the divertor recycling source, while closing leaks in the divertor structure would eliminate a significant neutral loss mechanism.

  3. Conditioning of the vacuum chamber of the Tokamak Novillo; Acondicionamiento de la camara de vacio del Tokamak Novillo

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-03-15

    The obtained experimental results of the implementation of two techniques of present time for the conditioning of the internal wall of the chamber of discharges of the Tokamak Novillo are presented, which has been designed, built and put in operation in the Laboratory of Plasma Physics of the National Institute of Nuclear Research (ININ). These techniques are: the vacuum baking and the low energy pulsed discharges, which were applied after having reached an initial pressure of the order of 10{sup -7} Torr. with a system of turbomolecular pumping previous preparation of surfaces and vacuum seals. The analysis of residual gases was carried out with a mass spectrometer before and after conditioning. The obtained results show that the vacuum baking it was of great effectiveness to reduce the value of the initial pressure in short time, in more of a magnitude order and the low energy discharges reduced the oxygen at worthless levels with regard to the initial values. (Author)

  4. Modification of tokamak edge plasma turbulence and transport by biasing and resonant helical magnetic field

    Science.gov (United States)

    Lafouti, Mansoureh; Ghoranneviss, Mahmood; Meshkani, Sakineh; Salar Elahi, Ahmad

    2013-05-01

    In this paper, both Resonant Helical magnetic Field (RHF) and limiter biasing have been applied to the tokamak. We have investigated their effects on the turbulence and transport of the particles at the edge of the plasma. The biased limiter voltage has been fixed at 200 V and RHF has L = 2 and L = 3. Also, the effects of the time order of the application of RHF and biasing to the tokamak have been explored. The experiment has been performed under three conditions. At first, the biasing and RHF were applied at t = 15 ms and at t = 20 ms. In the next step, RHF and biasing were applied at t = 15 ms and t = 20 ms, respectively. Finally, both of them were turned on at t = 15 ms until the end of the shot. For this purpose, the ion saturation current (Isat) and the floating potential (Vf) have been measured by the Langmuir probe at r/a = 0.9. Moreover, the power spectra of Isat and floating potential gradient (∇Vf), the coherency, the phase between them, and the particle diffusion coefficient have been calculated. The density fluctuations of the particles have been measured by the Rake probe and they have been analyzed with the Probability Distribution Function (PDF) technique. Also the particle diffusion coefficient has been determined by the Fick's law. The results show that, when RHF and biasing were applied at the same time to the plasma (during flatness region of plasma current), the radial particle density gradient, the radial particle flux, and the particle diffusion coefficient decrease about 50%, 60%, and 55%, respectively, compared to the other conditions. For more precision, the average values of the particle flux and the particle density gradient were calculated in the work. When the time is less than 15 ms, the average values of the particle flux and the particle density gradient are identical under all conditions, but in the other time interval they change. They reduce with the simultaneous application of biasing and RHF. The same results obtain from the

  5. Synchrotron emission diagnostic of full-orbit kinetic simulations of runaway electrons in tokamaks plasmas

    Science.gov (United States)

    Carbajal Gomez, Leopoldo; Del-Castillo-Negrete, Diego

    2017-10-01

    Developing avoidance or mitigation strategies of runaway electrons (RE) for the safe operation of ITER is imperative. Synchrotron radiation (SR) of RE is routinely used in current tokamak experiments to diagnose RE. We present the results of a newly developed camera diagnostic of SR for full-orbit kinetic simulations of RE in DIII-D-like plasmas that simultaneously includes: full-orbit effects, information of the spectral and angular distribution of SR of each electron, and basic geometric optics of a camera. We observe a strong dependence of the SR measured by the camera on the pitch angle distribution of RE, namely we find that crescent shapes of the SR on the camera pictures relate to RE distributions with small pitch angles, while ellipse shapes relate to distributions of RE with larger pitch angles. A weak dependence of the SR measured by the camera with the RE energy, value of the q-profile at the edge, and the chosen range of wavelengths is found. Furthermore, we observe that oversimplifying the angular distribution of the SR changes the synchrotron spectra and overestimates its amplitude. Research sponsored by the LDRD Program of ORNL, managed by UT-Battelle, LLC, for the U. S. DoE.

  6. Helium, Iron and Electron Particle Transport and Energy Transport Studies on the TFTR Tokamak

    Science.gov (United States)

    Synakowski, E. J.; Efthimion, P. C.; Rewoldt, G.; Stratton, B. C.; Tang, W. M.; Grek, B.; Hill, K. W.; Hulse, R. A.; Johnson, D .W.; Mansfield, D. K.; McCune, D.; Mikkelsen, D. R.; Park, H. K.; Ramsey, A. T.; Redi, M. H.; Scott, S. D.; Taylor, G.; Timberlake, J.; Zarnstorff, M. C. (Princeton Univ., NJ (United States). Plasma Physics Lab.); Kissick, M. W. (Wisconsin Univ., Madison, WI (United States))

    1993-03-01

    Results from helium, iron, and electron transport on TFTR in L-mode and Supershot deuterium plasmas with the same toroidal field, plasma current, and neutral beam heating power are presented. They are compared to results from thermal transport analysis based on power balance. Particle diffusivities and thermal conductivities are radially hollow and larger than neoclassical values, except possibly near the magnetic axis. The ion channel dominates over the electron channel in both particle and thermal diffusion. A peaked helium profile, supported by inward convection that is stronger than predicted by neoclassical theory, is measured in the Supershot The helium profile shape is consistent with predictions from quasilinear electrostatic drift-wave theory. While the perturbative particle diffusion coefficients of all three species are similar in the Supershot, differences are found in the L-Mode. Quasilinear theory calculations of the ratios of impurity diffusivities are in good accord with measurements. Theory estimates indicate that the ion heat flux should be larger than the electron heat flux, consistent with power balance analysis. However, theoretical values of the ratio of the ion to electron heat flux can be more than a factor of three larger than experimental values. A correlation between helium diffusion and ion thermal transport is observed and has favorable implications for sustained ignition of a tokamak fusion reactor.

  7. Acceleration optimization of real-time equilibrium reconstruction for HL-2A tokamak discharge control

    Science.gov (United States)

    Rui, MA; Fan, XIA; Fei, LING; Jiaxian, LI

    2018-02-01

    Real-time equilibrium reconstruction is crucially important for plasma shape control in the process of tokamak plasma discharge. However, as the reconstruction algorithm is computationally intensive, it is very difficult to improve its accuracy and reduce the computation time, and some optimizations need to be done. This article describes the three most important aspects of this optimization: (1) compiler optimization; (2) some optimization for middle-scale matrix multiplication on the graphic processing unit and an algorithm which can solve the block tri-diagonal linear system efficiently in parallel; (3) a new algorithm to locate the X&O point on the central processing unit. A static test proves the correctness and a dynamic test proves the feasibility of using the new code for real-time reconstruction with 129 × 129 grids; it can complete one iteration around 575 μs for each equilibrium reconstruction. The plasma displacements from real-time equilibrium reconstruction are compared with the experimental measurements, and the calculated results are consistent with the measured ones, which can be used as a reference for the real-time control of HL-2A discharge.

  8. Understanding of Neutral Gas Transport in the Alcator C-Mod Tokamak Divertor

    Energy Technology Data Exchange (ETDEWEB)

    D.P. Stotler; C.S. Pitcher; C.J. Boswell; B. LaBombard; J.L. Terry; J.D. Elder; S. Lisgo

    2002-05-07

    A series of experiments on the effect of divertor baffling on the Alcator C-Mod tokamak provides stringent tests on models of neutral gas transport in and around the divertor region. One attractive feature of these experiments is that a trial description of the background plasma can be constructed from experimental measurements using a simple model, allowing the neutral gas transport to be studied with a stand-alone code. The neutral-ion and neutral-neutral elastic scattering processes recently added to the DEGAS 2 Monte Carlo neutral transport code permit the neutral gas flow rates between the divertor and main chamber to be simulated more realistically than before. Nonetheless, the simulated neutral pressures are too low and the deuterium Balmer-alpha emission profiles differ qualitatively from those measured, indicating an incomplete understanding of the physical processes involved in the experiment. Some potential explanations are examined and opportunities for future exploration a re highlighted. Improvements to atomic and surface physics data and models will play a role in the latter.

  9. Experimental study of reversed shear Alfven eigenmodes during the current ramp in the Alcator C-Mod tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Edlund, E M; Kramer, G J [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Porkolab, M; Lin, Y; Tsujii, N; Wukitch, S J [MIT Plasma Science and Fusion Center, Cambridge, MA 02139 (United States); Lin, L, E-mail: eedlund@pppl.go [University of California Los Angeles, Los Angeles, CA 90095 (United States)

    2010-11-15

    Experiments conducted in the Alcator C-Mod tokamak have explored the physics of reversed shear Alfven eigenmodes (RSAEs) during the current ramp. The frequency evolution of the RSAEs during the current ramp provides a constraint on the evolution of q{sub min}, a result which is important in transport modeling and for comparison with other diagnostics which directly measure the magnetic field line structure. Additionally, a scaling of the RSAE minimum frequency with the sound speed is used to derive bounds on the adiabatic index, a measure of the plasma compressibility. This scaling places the adiabatic index at 1.40 {+-} 0.15 and supports the kinetic calculation of separate electron and ion compressibilities with an ion adiabatic index close to 7/4.

  10. Unraveling the plasma-material interface with real time diagnosis of dynamic boron conditioning in extreme tokamak plasmas

    Science.gov (United States)

    Domínguez-Gutiérrez, F. Javier; Bedoya, Felipe; Krstić, Predrag S.; Allain, Jean P.; Irle, Stephan; Skinner, Charles H.; Kaita, Robert; Koel, Bruce

    2017-08-01

    We present a study of the role of boron and oxygen in the chemistry of deuterium retention in boronized ATJ graphite irradiated by the extreme environment of a tokamak deuterium plasma. The experimental results were obtained by the first XPS measurements inside the plasma chamber of the National Spherical Torus Experiment Upgrade, between the plasma exposures. The subtle interplay of boron, carbon, oxygen and deuterium chemistry is explained by reactive molecular dynamics simulations, verified by quantum-classical molecular dynamics and successfully compared to the measured data. The calculations deciphered the roles of oxygen and boron for the deuterium retention and predict deuterium uptake into a boronized carbon surface close in value to that previously predicted for a lithiated and oxidized carbon surface.

  11. 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

  12. Preliminary project of s Thomson scattering system for the ETE tokamak; Projeto preliminar de um sistema de espalhamento Thomson para o Tokamak ETE

    Energy Technology Data Exchange (ETDEWEB)

    Berni, Luiz Angelo

    1997-12-31

    This report presents the preliminary project of the injection and laser light block system for the Thomson (ET) scattering diagnostic to be implanted at the ETE spheric tokamak of the Instituto Nacional de Pesquisas Espaciais (INPE/LAP). Also, a scanning system for the optics of scattered light 4 refs., 26 figs.

  13. Electronic system for Langmuir probe measurements

    Science.gov (United States)

    Mitov, M.; Bankova, A.; Dimitrova, M.; Ivanova, P.; Tutulkov, K.; Djermanova, N.; Dejarnac, R.; Stöckel, J.; Popov, Tsv K.

    2012-03-01

    A newly developed Langmuir probe system for measurements of current-voltage (IV) characteristics in the tokamak divertor area is presented and discussed. The system is partially controlled by a computer allowing simultaneous and independent feeding and registration of signals. The system is mounted in the COMPASS tokamak, Institute of Plasma Physics, Academy of Sciences of the Czech Republic. The new electronic circuit boards include also active low-pass filters which smooth the signal before recording by the data acquisition system (DAQ). The signal is thus less noisy and the data processing is much easier. We also designed and built a microcontroller-driven waveform generator with resolution of 1 Ms/s. The power supply is linear and uses a transformer. We avoided the use of a switching power supply because of the noise that it could generate. Examples of measurements of the IV characteristics by divertor probes in the COMPASS tokamak and evaluation of the EEDF are presented.

  14. Internal transport barrier in tokamak and helical plasmas

    Science.gov (United States)

    Ida, K.; Fujita, T.

    2018-03-01

    The differences and similarities between the internal transport barriers (ITBs) of tokamak and helical plasmas are reviewed. By comparing the characteristics of the ITBs in tokamak and helical plasmas, the mechanisms of the physics for the formation and dynamics of the ITB are clarified. The ITB is defined as the appearance of discontinuity of temperature, flow velocity, or density gradient in the radius. From the radial profiles of temperature, flow velocity, and density the ITB is characterized by the three parameters of normalized temperature gradient, R/{L}T, the location, {ρ }{ITB}, and the width, W/a, and can be expressed by ‘weak’ ITB (small R/{L}T) or ‘strong’ (large R/{L}T), ‘small’ ITB (small {ρ }{ITB}) or ‘large’ ITB (large {ρ }{ITB}), and ‘narrow’ (small W/a) or ‘wide’ (large W/a). Three key physics elements for the ITB formation, radial electric field shear, magnetic shear, and rational surface (and/or magnetic island) are described. The characteristics of electron and ion heat transport and electron and impurity transport are reviewed. There are significant differences in ion heat transport and electron heat transport. The dynamics of ITB formation and termination is also discussed. The emergence of the location of the ITB is sometimes far inside the ITB foot in the steady-state phase and the ITB region shows radial propagation during the formation of the ITB. The non-diffusive terms in momentum transport and impurity transport become more dominant in the plasma with the ITB. The reversal of the sign of non-diffusive terms in momentum transport and impurity transport associated with the formation of the ITB reported in helical plasma is described. Non-local transport plays an important role in determining the radial profile of temperature and density. The spontaneous change in temperature curvature (second radial derivative of temperature) in the ITB region is described. In addition, the key parameters of the control of the

  15. Magnetic spires for the detection of the position of the plasma column in a Tokamak (linear approximation); Espiras magneticas para la deteccion de la posicion de la columna de plasma en un Tokamak (aproximacion lineal)

    Energy Technology Data Exchange (ETDEWEB)

    Colunga S, S

    1990-07-15

    In this report the simplified analysis of a method to detect the movement of the plasma column of a tokamak in the vertical direction and of the biggest radius is given. The peculiar case of the Tokamak Novillo of the Plasma Physics Laboratory of the ININ is studied. (Author)

  16. A frequency-modulated continuous-wave reflectometer for the Lithium Tokamak Experiment

    Science.gov (United States)

    Kubota, S.; Majeski, R.; Peebles, W. A.; Bell, R. E.; Boyle, D. P.; Kaita, R.; Kozub, T.; Lucia, M.; Merino, E.; Nguyen, X. V.; Rhodes, T. L.; Schmitt, J. C.

    2017-05-01

    The frequency-modulated continuous-wave reflectometer on LTX (Lithium Tokamak Experiment) and the data analysis methods used for determining electron density profiles are described. The diagnostic uses a frequency range of 13.1-33.5 GHz, for covering a density range of 0.21-1.4 ×1013 cm-3 (in O-mode polarization) with a time resolution down to 8 μs. The design of the diagnostic incorporates the concept of an "optimized" source frequency sweep, which minimizes the large variation in the intermediate frequency signal due to a long dispersive transmission line. The quality of the raw data is dictated by the tuning characteristics of the microwave sources, as well as the group delay ripple in the transmission lines, which can generate higher-order nonlinearities in the frequency sweep. Both effects are evaluated for our diagnostic and best practices are presented for minimizing "artifacts" generated in the signals. The quality of the reconstructed profiles is also improved using two additional data analysis methods. First, the reflectometer data are processed as a radar image, where clutter due to echoes from the wall and backscattering from density fluctuations can be easily identified and removed. Second, a weighed least-squares lamination algorithm POLAN (POLynomial ANalysis) is used to reconstruct the electron density profile. Examples of density profiles in LTX are presented, along with comparisons to measurements from the Thomson scattering and the λ = 1 mm interferometer diagnostics.

  17. Development of FEMAG. Calculation code of magnetic field generated by ferritic plates in the tokamak devices

    Energy Technology Data Exchange (ETDEWEB)

    Urata, Kazuhiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2003-03-01

    In design of the future fusion devises in which low activation ferritic steel is planned to use as the plasma facing material and/or the inserts for ripple reduction, the appreciation of the error field effect against the plasma as well as the optimization of ferritic plate arrangement to reduce the toroidal field ripple require calculation of magnetic field generated by ferritic steel. However iterative calculations concerning the non-linearity in B-H curve of ferritic steel disturbs high-speed calculation required as the design tool. In the strong toroidal magnetic field that is characteristic in the tokamak fusion devices, fully magnetic saturation of ferritic steel occurs. Hence a distribution of magnetic charges as magnetic field source is determined straightforward and any iteration calculation are unnecessary. Additionally objective ferritic steel geometry is limited to the thin plate and ferritic plates are installed along the toroidal magnetic field. Taking these special conditions into account, high-speed calculation code ''FEMAG'' has been developed. In this report, the formalization of 'FEMAG' code, how to use 'FEMAG', and the validity check of 'FEMAG' in comparison with a 3D FEM code, with the measurements of the magnetic field in JFT-2M are described. The presented examples are numerical results of design studies for JT-60 modification. (author)

  18. Characteristics of EGAMs in EAST tokamak under ICRF H-mode

    Science.gov (United States)

    di Liu, Ah; Zhou, Chu; Zhang, Xiao Hui; Hu, Jian Qiang; Li, Hong; Lan, Tao; Xie, Jing Lin; Yu, Chang Xuan; Liu, Wan Dong

    2012-10-01

    Doppler reflectometer is common plasma diagnostic used in magnetic confinement devices to measure density fluctuations and poloidal flow velocity. Two set of Doppler reflectometer (Q-band & V-band)were installed on EAST tokamak for the first time. A coherence mode with frequency of 20˜50kHz was observed both on Doppler reflectometer and magnetic coils during ICRF H-mode on EAST. It appeared as zero-symmetric peaks in the spectrum of Doppler backscattering phase signal, implying that the density fluctuation has a standing wave structure with frequency not changing with the plasma rotation. The toroidal mode number is zero according to the magnetic coils. This feather was not observed on ECE and soft-X signals and there isn't obvious relationship between the mode appearance and the neutrons and hard-X signals. Unlike the usual Geodesic Acoustic modes (GAM) in the edge plasma under L-mode, it was found that the mode only appeared in the core regime under H-mode through the ray-tracing code. The mode is suspected to be the energetic ion induced GAM.

  19. Edge multi-energy soft x-ray diagnostic in Experimental Advanced Superconducting Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Li, Y. L.; Xu, G. S.; Wan, B. N.; Lan, H.; Liu, Y. L.; Wei, J.; Zhang, W.; Hu, G. H.; Wang, H. Q.; Duan, Y. M.; Zhao, J. L.; Wang, L.; Liu, S. C.; Ye, Y.; Li, J.; Lin, X.; Li, X. L. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Tritz, K. [Department of Physics and Astronomy, Johns Hopkins University, Baltimore, Maryland 21218 (United States); Zhu, Y. B. [Department of Physics and Astronomy, University of California, Irvine, California 92697-4575 (United States)

    2015-12-15

    A multi-energy soft x-ray (ME-SXR) diagnostic has been built for electron temperature profile in the edge plasma region in Experimental Advanced Superconducting Tokamak (EAST) after two rounds of campaigns. Originally, five preamplifiers were mounted inside the EAST vacuum vessel chamber attached to five vertically stacked compact diode arrays. A custom mechanical structure was designed to protect the detectors and electronics under constraints of the tangential field of view for plasma edge and the allocation of space. In the next experiment, the mechanical structure was redesigned with a barrel structure to absolutely isolate it from the vacuum vessel. Multiple shielding structures were mounted at the pinhole head to protect the metal foils from lithium coating. The pre-amplifiers were moved to the outside of the vacuum chamber to avoid introducing interference. Twisted copper cooling tube was embedded into the back-shell near the diode to limit the temperature of the preamplifiers and diode arrays during vacuum vessel baking when the temperature reached 150 °C. Electron temperature profiles were reconstructed from ME-SXR measurements using neural networks.

  20. Gyrokinetic characterization of the isotope effect in turbulent transport at the FT-2 tokamak

    Science.gov (United States)

    Niskala, P.; Gurchenko, A. D.; Gusakov, E. Z.; Altukhov, A. B.; Esipov, L. A.; Kantor, M. Yu; Kiviniemi, T. P.; Kouprienko, D.; Korpilo, T.; Lashkul, S. I.; Leerink, S.; Perevalov, A. A.; Rochford, R.

    2017-04-01

    Isotope effect allows fusion devices to perform better when heavier hydrogen isotopes are used as fuel, but the reason for this improvement is not yet understood. We present the first direct evidence of the isotope effect on particle confinement in the FT-2 tokamak and investigate it via gyrokinetic simulations. Experimental measurements for comparable hydrogen and deuterium discharges show that the particle confinement time increases by 40% for the heavier isotope species. The isotope effect on particle flux is reproduced in global and local gyrokinetic simulations. Global ELMFIRE simulations demonstrate a systemic reduction in particle fluxes across the radial range, showing a ratio of fluxes {{{Γ }}}{{H}}/{{{Γ }}}{{D}}=1.3 at the edge and {{{Γ }}}{{H}}/{{{Γ }}}{{D}}=1.4 at r/a=0.6. Local GENE simulations agree qualitatively with the result. Besides the fluctuation level, smaller scales and a favorable shift in the cross-phase between the turbulent fluctuations are found to contribute to the isotope effect in the simulations.