Dissipative processes in interchange driven scrape-off layer turbulence

DEFF Research Database (Denmark)

Fundamenski, W.; Garcia, Odd Erik; Naulin, Volker

2007-01-01

First principles expressions are given for the parameters governing collisional diffusion and parallel losses of mass, momentum and energy in tokamak scrape-off layer (SOL) plasmas. These dissipative, or damping, coefficients are based on neoclassical perpendicular transport (Pfirsch-Schluter dif......First principles expressions are given for the parameters governing collisional diffusion and parallel losses of mass, momentum and energy in tokamak scrape-off layer (SOL) plasmas. These dissipative, or damping, coefficients are based on neoclassical perpendicular transport (Pfirsch......-Schluter diffusion) and classical parallel transport (sub-sonic advection and Spitzer-Harm diffusion). When numerical values derived from these expressions are used to compute damping coefficients for the edge-SOL electrostatic (ESEL) turbulence code, simulations correctly reproduce the radial profiles of particle...... density, n, and electron temperature, T-e, as well as statistical distributions and temporal correlations of particle density and flux density measured in Ohmic and L-mode plasmas on the TCV tokamak. Similarly, preliminary calculations agree reasonably well with radial profiles of T-e measured in Ohmic...

Poloidal asymmetries in the scrape-off layer plasma of the Alcator C tokamak

International Nuclear Information System (INIS)

LaBombard, B.; Lipschultz, B.

1987-01-01

Large poloidal asymmetries in density, electron temperature, radial density e-folding length and floating potential have been measured in the plasma existing between the limiter radius and the wall of the Alcator C tokamak. Typically, variations in density by factors of about 4-20 and variations in radial density e-folding length by factors of about 3-8 are recorded in discharges which are bounded by poloidally symmetric ring limiters. These poloidal asymmetries show that pressure is a function of poloidal angle on open magnetic flux surfaces in this region of the plasma. Observations of toroidally symmetric MARFE (multifaceted asymmetric radiation from the edge) phenomena further imply that density and perhaps pressure are also a function of poloidal angle on closed flux surfaces existing just inside the limiter radius. The magnitude of these poloidal asymmetries and their dependence on poloidal angle persists independent of machine parameters (central plasma density, plasma current, toroidal field, MARFE versus non-MARFE discharges). Analysis of the data indicates that these asymmetries are caused by poloidal variations in perpendicular particle and heat transport in both the main plasma and the scrape-off layer. A number of possible asymmetric perpendicular transport processes in the scrape-off layer plasma are examined, including diffusion and E-vectorxB-vector plasma convection. (author)

Properties of the plasma of the scrape-off layer including the effects of polarization drift

International Nuclear Information System (INIS)

Petrov, V.G.

1987-01-01

The plasma of the scrape-off layer of a tokamak is analyzed. The toroidal electric drift and the polarization drift of the charged particles are taken into account. The buildup of electric charge in the shadow of the poloidal limiter which results from toroidal drift is offset by a current to the limiter. The radial electric current associated with the polarization drift of ions is important near the inner boundary of the scrape-off layer. The distributions of the electric potential and the plasma density in the scrape-off layer are derived

Modelling of impurity production and transport in the scrape-off layer of a high density limiter tokamak

International Nuclear Information System (INIS)

Zagorski, R.; Romanelli, F.

1996-01-01

A simple analytical model is presented that describes impurity ion production and transport in the tokamak scrape-off layer (SOL). The equations of the model are solved analytically in the test particle approximation. The solution, as a function of different plasma parameters and target materials, is discussed in the case in which the background plasma is described by the simple SOL model and a comparison between the model and the numerical results of a 2-D multifluid code is presented. (author). 18 refs, 8 figs, 2 tabs

Vlasov modelling of parallel transport in a tokamak scrape-off layer

International Nuclear Information System (INIS)

Manfredi, G; Hirstoaga, S; Devaux, S

2011-01-01

A one-dimensional Vlasov-Poisson model is used to describe the parallel transport in a tokamak scrape-off layer. Thanks to a recently developed 'asymptotic-preserving' numerical scheme, it is possible to lift numerical constraints on the time step and grid spacing, which are no longer limited by, respectively, the electron plasma period and Debye length. The Vlasov approach provides a good velocity-space resolution even in regions of low density. The model is applied to the study of parallel transport during edge-localized modes, with particular emphasis on the particles and energy fluxes on the divertor plates. The numerical results are compared with analytical estimates based on a free-streaming model, with good general agreement. An interesting feature is the observation of an early electron energy flux, due to suprathermal electrons escaping the ions' attraction. In contrast, the long-time evolution is essentially quasi-neutral and dominated by the ion dynamics.

Vlasov modelling of parallel transport in a tokamak scrape-off layer

Energy Technology Data Exchange (ETDEWEB)

Manfredi, G [Institut de Physique et Chimie des Materiaux, CNRS and Universite de Strasbourg, BP 43, F-67034 Strasbourg (France); Hirstoaga, S [INRIA Nancy Grand-Est and Institut de Recherche en Mathematiques Avancees, 7 rue Rene Descartes, F-67084 Strasbourg (France); Devaux, S, E-mail: Giovanni.Manfredi@ipcms.u-strasbg.f, E-mail: hirstoaga@math.unistra.f, E-mail: Stephane.Devaux@ccfe.ac.u [JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom)

2011-01-15

A one-dimensional Vlasov-Poisson model is used to describe the parallel transport in a tokamak scrape-off layer. Thanks to a recently developed 'asymptotic-preserving' numerical scheme, it is possible to lift numerical constraints on the time step and grid spacing, which are no longer limited by, respectively, the electron plasma period and Debye length. The Vlasov approach provides a good velocity-space resolution even in regions of low density. The model is applied to the study of parallel transport during edge-localized modes, with particular emphasis on the particles and energy fluxes on the divertor plates. The numerical results are compared with analytical estimates based on a free-streaming model, with good general agreement. An interesting feature is the observation of an early electron energy flux, due to suprathermal electrons escaping the ions' attraction. In contrast, the long-time evolution is essentially quasi-neutral and dominated by the ion dynamics.

Scrape-off layer plasma modeling for the DIII-D tokamak

International Nuclear Information System (INIS)

Porter, G.D.; Rognlien, T.D.; Allen, S.L.

1994-09-01

The behavior of the scrape-off layer (SOL) region in tokamaks is believed to play an important role determining the overall device performance. In addition, control of the exhaust power has become one of the most important issues in the design of future devices such as ITER and TPX. This paper presents the results of application of 2-D fluid models to the DII-D tokamak, and research into the importance of processes which are inadequately treated in the fluid models. Comparison of measured and simulated profiles of SOL plasma parameters suggest the physics model contained in the UEDGE code is sufficient to simulate plasmas which are attached to the divertor plates. Experimental evidence suggests the presence of enhanced plasma recombination and momentum removal leading to the existence of detached plasma states. UEDGE simulation of these plasmas obtains a bifurcation to a low temperature plasma at the divertor, but the plasma remains attached. Understanding the physics of this detachment is important for the design of future devices. Analytic studies of the behavior of SOL plasmas enhance our understanding beyond that achieved with fluid modeling. Analysis of the effect of drifts on sheath structure suggest these drifts may play a role in the detachment process. Analysis of the turbulent-transport equations indicate a bifurcation which is qualitatively similar to the experimentally different behavior of the L- and H-mode SOL. Electrostatic simulations of conducting wall modes suggest possible control of the SOL width by biasing

A two-dimensional kinetic model of the scrape-off layer

International Nuclear Information System (INIS)

Catto, P.J.; Hazeltine, R.D.

1993-09-01

A two-dimensional (radius and poloidal angle), analytically tractable kinetic model of the ion (or energetic electron) behavior in the scrape-off layer of a limiter or divertor plasma in a tokamak is presented. The model determines the boundary conditions on the core ion density and ion temperature gradients, the power load on the limiter or divertor plates, the energy carried per particle to the walls, and the effective flux limit. The self-consistent electrostatic potential in the quasi-neutral scrape-off layer is determined by using the ion kinetic model of the layer along with a Maxwell-Boltzmann electron response that occurs because most electrons are reflected by the Debye sheaths (assumed to be infinitely thin) at the limiter or divertor plates

Influence of the Alfven wave spectrum on the scrape-off layer of the TCA tokamak

International Nuclear Information System (INIS)

Martin, Y.; Hollenstein, Ch.

1988-01-01

The study of the Scrape-Off Layer (SOL) during Alfven wave heating may lead to a better understanding of the antenna-plasma interaction. The SOL of the TCA tokamak has been widely investigated by means of Langmuir probes. The aim of the present work is to present in detail the influence of the Alfven wave spectrum on the SOL. The experiments have shown that the plasma boundary layer is strongly affected by the RF, in particular the ion density, the electron temperature and the floating potential. In TCA, as the spectrum evolves due to a density rise, the passage of the Alfven continua and their associated eigenmodes (DAW) induces a strong depletion in the edge density of up to 70% during the continuum part and a density increase during the crossing of an eigenmode. The floating potential becomes negative during the continua and even more negative crossing the eigenmodes. This behaviour changes as a function of the power transmitted to the plasma through the antennae, especially we have found with MHD modes a change around 100 kW. The profiles of the basic parameters are modified, depending on the wave spectrum. MHD mode activity which can occur during the RF phase considerably alters the behaviour mentioned above. Finally, the modulation of the RF power allows us to characterize the difference in coupling, for the continua and the eigenmodes, between the Alfven wave field and the scrape-off layer. (author) 5 figs., 6 refs

Blobs and front propagation in the scrape-off layer of magnetic confinement devices

International Nuclear Information System (INIS)

Bian, N.; Benkadda, S.; Paulsen, J.-V.; Garcia, O.E.

2003-01-01

In this paper we show the self-consistent evolution of an isolated density perturbation in models of tokamak scrape-off layer turbulence. Our purpose is to explain the possible mechanisms responsible for radial propagation of density perturbations observed in the scrape-off layer. Results of both two-dimensional numerical simulations and one-dimensional quasilinear modeling of the propagative events are presented, and shown to be consistent with many experimental observations. The role of sheath dissipation for front propagation and turbulent mixing is also addressed

The effect of ion drifts on the properties of the tokamak scrape-off plasma

International Nuclear Information System (INIS)

Petravic, M.; Kuo-Petravic, G.

1988-09-01

A plasma fluid model which takes into account ion drifts has been constructed and applied to the scrape-off layer of a tokamak with a poloidal divertor. This model predicts near-sonic toroidal velocities and large poloidal flows in most of the scrapeoff together with steep gradients in the pressure and electrostatic potential along the magnetic field near the X-point, contrary to the predictions of the standard model. The potential step at X-point should reduce parallel heat transport and could act as an H-mode trigger. 12 refs., 4 figs

Particle and heat balance analysis in scrape-off and divertor regions of the JFT-2M tokamak

International Nuclear Information System (INIS)

Nagashima, K.; Shoji, T.; Tamai, H.; Miura, Y.; Takenaga, H.; Maeda, H.

1995-01-01

Particle and heat balance in the scrape-off layer and the divertor region were studied in the JFT-2M tokamak. Using particle and energy conservation laws, particle and heat diffusivities perpendicular to the flux surface were evaluated just outside the magnetic separatrix. It was found that the particle diffusivity decreases with increasing electron density in the scrape-off layer and decreases by a factor of 2-3 in the H-mode phase as compared with that in L-mode. The heat diffusivity has almost the same dependence on the electron density. The ratio of the heat diffusivity to the particle diffusivity is about 2. ((orig.))

Plasma motion in the scrape-off layer of a nonequipotential tokamak limiter

International Nuclear Information System (INIS)

Fidel'man, G.N.

1987-01-01

A theory is derived for the laminar convection of the plasma in the scrape-off layer of a poloidal limiter with a distributed potential. The plasma potential and density distributions are derived for the cases of poloidal and slightly nonuniform radial distributions of the potential on the limiter. The stability of these solutions against flute perturbations is analyzed. Criteria for the control of the density profile are derived in the models of laminar and turbulent scrape-off layers. The energy expended in sustaining the limiter potential distribution is studied. If a suitable potential distribution is selected, it is possible to obtain ∼10% of the total convective heat flux from the plasma column in the form of purely electrical energy

Scrape-off measurements during Alfven wave heating in the TCA tokamak

International Nuclear Information System (INIS)

Hofmann, F.; Hollenstein, C.; Joye, B.; Lietti, A.; Lister, J.B.; Pochelon, A.; Gimzewski, J.K.; Veprek, S.

1984-01-01

Plasma parameters and impurity fluxes in the scrape-off layer of the TCA tokamak have been measured during Alfven wave heating. Langmuir probes are used to measure electron density, electron temperature and plasma potential. Collection probes, in conjunction with XPS surface analysis, are used to determine impurity fluxes and ion impact energies. During RF heating, the electron edge temperature rises, the plasma potential drops and impurity fluxes are enhanced. Probe erosion due to impurity sputtering is clearly observed. The measurements are correlated with other diagnostics on TCA. (orig.)

Experimental and theoretical investigation of density and potential fluctuations in the scrape-off layer of ASDEX

Energy Technology Data Exchange (ETDEWEB)

Endler, M; Giannone, L.; Niedermeyer, H; Rudyj, A; Theimer, G [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

1994-12-31

Electrostatic fluctuations (i.e. the magnetic field is assumed constant) are candidates for the explanation of the anomalous transport of particles and energy in both tokamaks and stellarators. While most theoretical effort has been directed to an explanation of the anomalous transport in the bulk plasma, it is now widely being realized that the anomalous radial transport in the scrape-off layer, determining the width of the power flow channel at limiter or divertor plates, may be equally important to a future reactor experiment. In the divertor tokamak ASDEX density and potential fluctuations in the scrape-off layer were investigated with high temporal and spatial resolution by Langmuir probes and an H{sub {alpha}} diagnostic. Many results of these measurements were reported and are summarized below. Several of these properties of the fluctuations have also been reported from other experiments. (author) 3 refs., 4 figs.

Interpretive modelling of scrape-off plasmas on the MAST tokamak

Energy Technology Data Exchange (ETDEWEB)

Harrison, J. [Euratom/UKAEA Fusion Association, Culham Science Centre, D2/2.01 Fusion Association, Abingdon, Oxfordshire OX14 3DB (United Kingdom); University of York, Heslington, York (United Kingdom)], E-mail: james.harrison@ukaea.org.uk; Lisgo, S. [Euratom/UKAEA Fusion Association, Culham Science Centre, D2/2.01 Fusion Association, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Counsell, G.F. [Fusion for Energy, Barcelona (Spain); Gibson, K. [University of York, Heslington, York (United Kingdom); Dowling, J. [Euratom/UKAEA Fusion Association, Culham Science Centre, D2/2.01 Fusion Association, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Trojan, L. [University of Manchester, Oxford Road, Manchester (United Kingdom); Reiter, D. [IPP, Forschungszentrum Juelich GmbH, EURATOM Association, D-52425 Juelich (Germany)

2009-06-15

Electrical currents in the scrape-off layer (SOL) of MAST are modelled using an interpretive Onion-Skin Model (OSM) constrained with experimental data from MAST diagnostics. The model was extended to include the effects of the magnetic mirror force, which has a strong influence on the particle and momentum balance in spherical tokamaks, such as MAST . These modifications serve to more accurately model the parallel electric fields present in the MAST SOL, which can alter plasma dynamics via the E x B drift. Simulations show that the electrical current at the divertor targets is predominantly thermoelectric, whereas Pfirsch-Schlueter currents have a greater contribution to the total current in the bulk of the SOL plasma.

Characteristics of steady-state plasma flow in the tokamak limiter scrape-off layer

International Nuclear Information System (INIS)

Petrov, V.G.

1984-01-01

Steady state plasma flow in the scrape-off layer of a toroidal limiter is discussed. The force balance along the torus minor radius is taken into account, from which follows that the plasma pressure gradient is balanced by the ponderomotive force (1/c) j-vectorxB-vector, which arises in the presence of a current density component perpendicular to the magnetic field. The limiter has an important effect on the electric current flow in the scrape-off layer. It is shown that the electric potential and plasma density values differ from one side of the limiter to the other; this leads to plasma drift along the minor radius. The characteristic length of change in the plasma density is found to be of the order of the ion cyclotron radius calculated for a poloidal magnetic field. (author)

Tokamak fusion reactor exhaust

International Nuclear Information System (INIS)

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

1981-08-01

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

Simulation of the scrape-off layer plasma during a disruption

International Nuclear Information System (INIS)

Rognlien, T.D.; Crotinger, J.A.; Porter, G.D.; Smith, G.R.; Kellman, A.G.; Taylor, P.L.

1996-01-01

The evolution of the scrape-off layer (SOL) during a disruption in the DIII-D tokamak is modeled using the 2-D UEDGE transport code. The focus is on the thermal quench phase when most of the energy content of the discharge is rapidly transported across the magnetic separatrix where it then flows to material surfaces or is radiated. Comparisons between the simulation and an experiment on the DIII-D tokamak are made with the heat flux to the divertor plate, and temperature and density profiles at the SOL midplane. The temporal response of the separate electron and ion heat-flux components to the divertor plate is calculated. The sensitivity of the solution to assumptions of electron heat-flux models and impurity radiation is investigated

Transport of wall released impurities in the limiter scrape-off layer of a tokamak

International Nuclear Information System (INIS)

Claassen, H.A.; Repp, H.

1978-01-01

A collisional theory for the transport of heavy wall released impurities in the plasma scrape-off layer is developed, which to zero order approximation considers electron impact ionization and Coulomb collisions with the plasma ions. Impurity ion convection parallel to the magnetic field and radial drift motion are treated as first order correction terms. The theory, which under certain restrictions to the integral coefficients of the Fokker-Planck collision operator is independent of the special form of the plasma ion distribution, is applied to the calculation of the impurity ion fluxes in the scrape-off layer. Preliminary numerical results are presented for a model plasma ion distribution of the loss ellipse type and a half-maxwellian distribution of the wall released impurity atoms. (Auth.)

Evidence for a poloidally localized enhancement of radial transport in the scrape-off layer of the Tore Supra tokamak

International Nuclear Information System (INIS)

Gunn, J.P.; Boucher, C.; Dionne, M.; Duran, I.; Fuchs, V.; Loarer, T.; Nanobashvili, I.; Panek, R.; Pascal, J.-Y.; Saint-Laurent, F.; Stoeckel, J.; Rompuy, T. van; Zagorski, R.; Adamek, J.; Bucalossi, J.; Dejarnac, R.; Devynck, P.; Hertout, P.; Hron, M.; Lebrun, G.; Moreau, P.; Rimini, F.; Sarkissian, A.; Oost, G. van

2007-01-01

Near-sonic parallel flows are systematically observed in the far scrape-off layer (SOL) of the limiter tokamak Tore Supra, as in many L-mode X-point divertor tokamak plasmas. The poloidal variation of the parallel flow has been measured by moving the contact point of a small circular plasma onto limiters at different poloidal angles. The resulting variations of flow are consistent with the existence of a poloidally localized enhancement of radial transport concentrated in a 30 deg. sector near the outboard midplane. If the plasma contact point is placed on the inboard limiters, then the SOL expands to fill all the space between the plasma and the outboard limiters, with density decay lengths between 10 and 20 cm. On the other hand, if the contact point lies on the outboard limiters, the localized plasma outflux is scraped off and the SOL is very thin with decay lengths around 2-3 cm. The outboard radial transport would have to be about two orders of magnitude stronger than inboard to explain these results

Discontinuous Galerkin methods for plasma physics in the scrape-off layer of tokamaks

International Nuclear Information System (INIS)

Michoski, C.; Meyerson, D.; Isaac, T.; Waelbroeck, F.

2014-01-01

A new parallel discontinuous Galerkin solver, called ArcOn, is developed to describe the intermittent turbulent transport of filamentary blobs in the scrape-off layer (SOL) of fusion plasma. The model is comprised of an elliptic subsystem coupled to two convection-dominated reaction–diffusion–convection equations. Upwinding is used for a class of numerical fluxes developed to accommodate cross product driven convection, and the elliptic solver uses SIPG, NIPG, IIPG, Brezzi, and Bassi–Rebay fluxes to formulate the stiffness matrix. A novel entropy sensor is developed for this system, designed for a space–time varying artificial diffusion/viscosity regularization algorithm. Some numerical experiments are performed to show convergence order on manufactured solutions, regularization of blob/streamer dynamics in the SOL given unstable parameterizations, long-time stability of modon (or dipole drift vortex) solutions arising in simulations of drift-wave turbulence, and finally the formation of edge mode turbulence in the scrape-off layer under turbulent saturation conditions

Scrape-off model and pumped-limiter design for reversed-field pinches (RFP)

International Nuclear Information System (INIS)

Embrechts, M.J.; Bathke, C.G.; Krakowski, R.A.

1983-01-01

In order to develop a better understanding of the plasma/first-wall interaction in an RFP configuration, the models being developed to describe edgeplasma and scrapeoff regions for the tokamak have been adopted. Specifically, a scrapeoff model similar to the one developed and used for the tokamak FED design is used to model the parameter range where pumped limiters may be applied to the compact RFP. The one-dimensional, steady-state heat and particle equations are solved in the scrapeoff layer for an RFP geometry, assuming equal electron and ion temperature and density, convective transport along field lines, and Bohm-like diffusion. All charge-exchange and radiation effects are assumed to take place in the region between plasma and scrapeoff layer, and only a specific fraction of the total plasma energy will enter the scrapeoff layer in the form of conduction and convection. A simplified recycling model based on an effective recycling coefficient is used. For a given particle and energy flux entering the scrapeoff and for specified relationships between the particle flux and the energy flux near the first wall, the temperature and density profiles in the scrapeoff layer region are determined. The shape of the limiter surface is determined iteratively for a specified number of poloidal limiters by specifying a constant (design) heat flux for the limiter surfaces

Scrape-off layer turbulence in TCV: evidence in support of stochastic modelling

Czech Academy of Sciences Publication Activity Database

Theodorsen, A.; Garcia, O.E.; Horáček, Jan; Kube, R.; Pitts, R.A.

2016-01-01

Roč. 58, č. 4 (2016), č. článku 044006. ISSN 0741-3335 R&D Projects: GA ČR(CZ) GAP205/12/2327 Institutional support: RVO:61389021 Keywords : turbulence * intermittency * transport * scrape-off layer * tcv * plasma * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.392, year: 2016 http://iopscience.iop.org/article/10.1088/0741-3335/58/4/044006/meta

Ion Temperature Measurements in the Tore Supra Scrape-Off Layer Using a Retarding Field Analyzer

International Nuclear Information System (INIS)

Kocan, M.; Gunn, J.P.; Pascal, J.Y.; Gauthier, E.

2010-01-01

The retarding field analyzer (RFA) is one of the only widely accepted diagnostics for measuring the ion temperature T i )in the tokamak scrape-off layer. An overview of the outstanding RFA performance over ten years of operation in Tore Supra tokamak is given and the validation of T i measurements is addressed. The RFA measurements in Tore Supra are found to be well reproducible. The ion-to-electron temperature ratio is higher than one at low-to-moderate ion-electron collisionality regime and converges to unity at high collisionality regime. (authors)

Experimental and theoretical investigation of density and potential fluctuations in the scrape-off layer of ASDEX

International Nuclear Information System (INIS)

Endler, M.; Giannone, L.; Niedermeyer, H.; Rudyj, A.; Theimer, G.

1993-01-01

In the divertor tokamak ASDEX density and potential fluctuations in the scrape-off layer were investigated with high temporal and spatial resolution by Langmuir probes and an H α diagnostic. Many results of these measurements were reported and are summarized below. Several of these properties of the fluctuations have also been reported from other experiments. (orig.)

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

Czech Academy of Sciences Publication Activity Database

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

2013-01-01

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

Diagnostics of mobile dust in scrape-off layer plasmas

International Nuclear Information System (INIS)

Ratynskaia, S; Castaldo, C; Bergsaaker, H; Rudakov, D

2011-01-01

Dust production and accumulation pose serious safety and operational implications for the next generation fusion devices. Mobile dust particles can result in core plasma contamination with impurities, and those with high velocities can significantly contribute to the wall erosion. Diagnostics for monitoring dust in tokamaks during plasma discharges are hence important as they can provide information on dust velocity and size, and-in some cases-on dust composition. Such measurements are also valuable as an input for theoretical models of dust dynamics in scrape-off layer plasmas. Existing in situ dust diagnostics, focusing on the range of dust parameters they can detect, are reviewed. Particular attention is paid to the diagnostics which allow us to detect tails of the dust velocity and size distributions, e.g. small and very fast particles. Some of the techniques discussed have been adopted from space-related research and have been shown to be feasible and useful for tokamak applications as well.

2D scrape-off layer turbulence measurement using Deuterium beam emission spectroscopy on KSTAR

Science.gov (United States)

Lampert, M.; Zoletnik, S.; Bak, J. G.; Nam, Y. U.; Kstar Team

2018-04-01

Intermittent events in the scrape-off layer (SOL) of magnetically confined plasmas, often called blobs and holes, contribute significantly to the particle and heat loss across the magnetic field lines. In this article, the results of the scrape-off layer and edge turbulence measurements are presented with the two-dimensional Deuterium Beam Emission Spectroscopy system (DBES) at KSTAR (Korea Superconducting Tokamak Advanced Research). The properties of blobs and holes are determined in an L-mode and an H-mode shot with statistical tools and conditional averaging. These results show the capabilities and limitations of the SOL turbulence measurement of a 2D BES system. The results from the BES study were compared with the analysis of probe measurements. It was found that while probes offer a better signal-to-noise ratio and can measure blobs down to 3 mm size, BES can monitor the two-dimensional dynamics of larger events continuously during full discharges, and the measurement is not limited to the SOL on KSTAR.

Ion temperature measurements in the scrape-off layer of the Tore Supra Tokamak

International Nuclear Information System (INIS)

Kocan, M.

2009-10-01

The thesis describes measurements of the scrape-off layer (SOL) ion temperature T i with a retarding field analyzer (RFA) in the limiter tokamak Tore Supra. Considerable emphasis is placed on study of the instrumental effects of RFAs and their influence on T i measurements. In general, the influence of instrumental effects on T i measurements is found to be relatively small. The instrumental study is followed by systematic measurements of T i (as well as other parameters) in the Tore Supra SOL. This includes the scaling of SOL temperatures and electron density with the main plasma parameters (such as the plasma density, toroidal magnetic field, working gas, and the radiated power fraction). Except at very high densities or in detached plasmas, SOL T i is found to be higher than T e by up to a factor of 7. While SOL T i is found to vary by almost two orders of magnitude, following the variation of the core temperatures, SOL T e changes only little and seems to be decoupled from the core plasma. The first continuous T i /T e profile from the edge of the confined plasma into the SOL is constructed using data from different tokamaks. It is shown that T i /T e > 1 in the SOL but also in the confined plasma, and increases with radius. The first evidence of poloidal asymmetry of the radial ion and electron energy transport in the SOL is reported. Implications for ITER start-up phase are discussed. Correlation of the asymmetries of SOL T i and T e measured from both directions along the magnetic field lines with changes of the parallel Mach number is studied. SOL T i was measured for the first time in Tore Supra by charge exchange recombination spectroscopy (CXRS) and compared to RFA data. A factor of 4 higher T i measured by CXRS is a subject of further analysis. (A.C.)

Poloidal Asymmetry in the Narrow Heat Flux Feature in the TCV Scrape-Off Layer.

Czech Academy of Sciences Publication Activity Database

Tsui, C.K.; Boedo, J. A.; Halpern, F.D.; Loizu, J.; Nespoli, F.; Horáček, Jan; Labit, B.; Morales, J.; Reimerdes, H.; Ricci, P.; Theiler, C.; Coda, S.; Duval, B. P.; Furno, I.

2017-01-01

Roč. 24, č. 6 (2017), č. článku 062508. ISSN 1070-664X R&D Projects: GA ČR(CZ) GA15-10723S EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : Scrape-Off Layer * TCV * tokamak * plasma Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.115, year: 2016 http://dx.doi.org/10.1063/1.4985075

Understanding and suppressing the near Scrape-Off Layer in inboard-limited plasmas in TCV.

Czech Academy of Sciences Publication Activity Database

Nespoli, F.; Labit, B.; Furno, I.; Horáček, Jan; Tsui, C.K.; Boedo, J. A.; Maurizio, R.; Reimerdes, H.; Theiler, C.; Ricci, P.; Halpern, F.D.; Sheikh, U.; Verhaegh, K.; Pitts, R.A.; Militello, F.

2017-01-01

Roč. 57, č. 12 (2017), č. článku 126029. ISSN 0029-5515 R&D Projects: GA ČR(CZ) GA15-10723S EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : tokamak * TCV * scrape-off layer * heat flux * limiter * infrared thermography * Langmuir probes Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 3.307, year: 2016 http://iopscience.iop.org/article/10.1088/1741-4326/aa84e0

Detection of lower hybrid waves in the scrape-off layer of tokamak plasmas with microwave backscattering

International Nuclear Information System (INIS)

Baek, S. G.; Shiraiwa, S.; Parker, R. R.; Bonoli, P. T.; Marmar, E. S.; Wallace, G. M.; Lau, C.; Dominguez, A.; Kramer, G. J.

2014-01-01

Microwave backscattering experiments have been performed on the Alcator C-Mod tokamak in order to investigate the propagation of lower hybrid (LH) waves in reactor-relevant, high-density plasmas. When the line-averaged density is raised above 1 × 10 20 m –3 , lower hybrid current drive efficiency is found to be lower than expected [Wallace et al., Phys. Plasmas 19, 062505 (2012)] and LH power is thought to be dissipated at the plasma edge. Using a single channel (60 GHz) ordinary-mode (O-mode) reflectometer system, we demonstrate radially localized LH wave measurements in the scrape-off layer of high density plasmas (n ¯ e  ≳ 0.9×10 20  m −3 ). Measured backscattered O-mode power varies depending on the magnetic field line mapping, suggesting the resonance cone propagation of LH waves. Backscattered power is also sensitive to variations in plasma density and the launched parallel refractive index of the LH waves. LH ray-tracing simulations have been carried out to interpret the observed variations. To understand the measured LH waves in regions not magnetically connected to the launcher, two hypotheses are examined. One is the weak single pass absorption and the other is scattering of LH waves by non-linear effects

Dual cascade and minimum enstrophy state in the tokamak scrape-off layer

International Nuclear Information System (INIS)

Mattor, N.; Cohen, R.H.; Xu, X.Q.

1993-01-01

In the Tokamak Scrape-off layer (SOL), there is experimental, theoretical, and computational evidence of an inverse energy cascade, wherein fluctuation energy transfers nonlinearly to large scale lengths. If the inverse cascade proceeds to the largest scales, it gives transport which is inherently nonlocal, precluding standard descriptions with local transport coefficients. This includes DIA based renormalization theories, γ/k 2 open-quotes mixing lengthclose quotes theories, and spectral or pseudo-spectral codes, all of which tend to involve a two-scale assumption, that turbulence acts on very short time and length scales relative to the equilibrium. The two-scale assumption is violated by turbulence undergoing a significant inverse cascade, and a different approach is called for. The authors postulate that the net effect of such turbulence is not local transport, but rather to supply the equilibrium with a steady source of energy at the minimum enstrophy. The form of the supplied energy is assessed through a variational calculation, which gives an equation for the equilibrium velocity profile, ∇ 2 V = λ 2 V, where λ 2 is an undetermined Lagrange multiplier. For a slab model, the solution in the SOL is V = V a exp[-λ(r-a)]y, where V a is the poloidal velocity at the SOL/edge interface. This velocity (from E x B in the simple model), leads to the potential profile, φ = -(V a B/λc)exp[-λ(r-a)]. For field lines connected to an endplate eφ = ΛT e , (where Λ ∼ 4 is nearly constant) giving also the T e profile. Thus, the profiles are given and the transport problem is solved, up to the two unknown constants λ and V a . One relation comes from heat balance. There are several candidates for the second constant, and the authors present numerical simulations which evaluate these

Measurements of the parallel wavenumber of lower hybrid waves in the scrape-off layer of a high-density tokamak

International Nuclear Information System (INIS)

Baek, S. G.; Wallace, G. M.; Parker, R. R.; Shiraiwa, S.; Bonoli, P. T.; Brunner, D.; Faust, I.; LaBombard, B. L.; Wukitch, S.; Shinya, T.; Takase, Y.

2016-01-01

In lower hybrid current drive (LHCD) experiments on tokamaks, the parallel wavenumber of lower hybrid waves is an important physics parameter that governs the wave propagation and absorption physics. However, this parameter has not been experimentally well-characterized in the present-day high density tokamaks, despite the advances in the wave physics modeling. In this paper, we present the first measurement of the dominant parallel wavenumber of lower hybrid waves in the scrape-off layer (SOL) of the Alcator C-Mod tokamak with an array of magnetic loop probes. The electric field strength measured with the probe in typical C-Mod plasmas is about one-fifth of that of the electric field at the mouth of the grill antenna. The amplitude and phase responses of the measured signals on the applied power spectrum are consistent with the expected wave energy propagation. At higher density, the observed k || increases for the fixed launched k || , and the wave amplitude decreases rapidly. This decrease is correlated with the loss of LHCD efficiency at high density, suggesting the presence of loss mechanisms. Evidence of the spectral broadening mechanisms is observed in the frequency spectra. However, no clear modifications in the dominant k || are observed in the spectrally broadened wave components, as compared to the measured k || at the applied frequency. It could be due to (1) the probe being in the SOL and (2) the limited k || resolution of the diagnostic. Future experiments are planned to investigate the roles of the observed spectral broadening mechanisms on the LH density limit problem in the strong single pass damping regime.

Characterization of Scrape-Off layer turbulence changes induced by a non-axisymmetric magnetic perturbation in an ASDEX upgrade low density L-mode

International Nuclear Information System (INIS)

Mueller, H.W.; Carralero, D.; Birkenmeier, G.; Conway, G.D.; Fischer, R.; Happel, T.; Manz, P.; Suttrop, W.; Wolfrum, E.

2014-01-01

In the tokamak ASDEX Upgrade the influence of a non-axisymmetric n = 2 error field on the turbulence in the far scrape-off layer of a low density L-mode discharge has been studied. There is no density pump-out with the non-axisymmetric perturbation but an increase of the scrape-off layer density at the outer midplane. While the relative ion saturation current fluctuation level in the far scrape-off layer is decreasing, the skewness rises and especially the excess kurtosis grows by a factor of 1.5-3. The frequency of intermittent events (blobs) is increasing by 50 %. Also the poloidal velocity grows with the magnetic perturbation while the typical turbulent structure size becomes smaller by a factor 5-10 about 20-25 mm outside the separatrix. The local spectral density has been calculated from a two-point measurement of the ion saturation current. It is used to derive a dispersion relation. Two poloidal propagation velocities depending on the wave number have been found. One is an upper limit for the bulk E x B velocity and the second one the lower limit of the phase velocity. There is a significant contribution of the phase velocity to the propagation speed in the far scrape-off layer. (copyright 2014 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

Theory of the scrape-off layer width in inner-wall limited tokamak plasmas

International Nuclear Information System (INIS)

Halpern, F.D.; Ricci, P.; Jolliet, S.; Loizu, J.; Mosetto, A.

2014-01-01

We develop a predictive theory applicable to the scrape-off layer (SOL) of inner-wall limited plasmas. Using the non-linear flattening of the pressure profile as a saturation mechanism for resistive ballooning modes, we are able to demonstrate and quantify the increase of the SOL width with plasma size, connection length, plasma β, and collisionality. Individual aspects of the theory, such as saturation physics, parallel dynamics, and system size scaling, are tested and verified using non-linear, 3D flux-driven SOL turbulence simulations. Altogether, very good agreement between theory and simulation is found. (paper)

Probe measurements for impurity transport in the scrape-off layer of JIPP T-II

International Nuclear Information System (INIS)

Mohri, M.; Satake, T.; Hashiba, H.; Yamashina, T.; Amemiya, S.

1982-05-01

Impurity transport processes in the scrape-off layer of the JIPP T-II device have been studied by a probe method. A cubical silicon probe was inserted and exposed to 20 identical tokamak discharges in the scrape-off region. Deposited impurities were analyzed with use of AES, RBS and PIXE equipments. The main metallic impurities were molybdenum and iron whose deposition behavior was almost the same on any side of the probe, and their fluxes were observed to be 1.2 x 10 13 /cm 2 .discharge on the electron drift side and 5.2 x 10 13 /cm 2 .discharge on the ion drift side, respectively at the distance of 18.3 cm from the center line of the plasma. The mean transport energy of the impurities striking the probe surface was estimated from the depth concentration profile applying the LSS theory for iron as 90 eV on the electron drift side and 250 eV on the ion drift side, respectively. The e-folding length of the scrape-off plasma density was measured by the radial distribution of a deposited tantalum amount to be 0.64 cm on the electron drift side and 1.73 cm on the ion drift side, respectively. (author)

Retention of neon in graphite after ion beam implantation or exposures to the scrape-off layer plasma in the TEXTOR tokamak

International Nuclear Information System (INIS)

Kim, Y.M.; Philipps, V.; Rubel, M.; Vietzke, E.; Pospieszczyk, A.; Unterberg, B.; Jaspers, R.

2002-01-01

The interaction of neon ions with graphite was investigated for targets either irradiated with ion beams (2-10 keV range) or exposed to the scrape-off layer plasma in the TEXTOR tokamak during discharges with neon edge cooling. The emphasis was on the influence of the target temperature (300-1200 K) and the implantation dose on the neon retention and reemission. The influence of deuterium impact on the retention of neon implanted into graphite has also been addressed. In ion beam experiments saturation is observed above a certain ion dose with a saturation level, which decreases with increasing target temperature. The temperature dependence of the thermal desorption corresponds to an apparent binding energy of about 2.06 eV. The retention of neon (C Ne /C C ) decreases with increasing ion energy with values from 0.55 to 0.15 following irradiation with 2 and 10 keV ions, respectively. The reemission yield during the irradiation increases with target temperature and above 1200 K all impinging ions are reemitted instantaneously. The retention densities measured using the sniffer probe at the TEXTOR tokamak are less than 1% of the total neon fluence and are over one order of magnitude smaller than those observed in ion beam experiments. The results are discussed in terms of different process decisive for ion deposition and release under the two experimental conditions

Heuristic drift-based model of the power scrape-off width in low-gas-puff H-mode tokamaks

International Nuclear Information System (INIS)

Goldston, R.J.

2012-01-01

A heuristic model for the plasma scrape-off width in low-gas-puff tokamak H-mode plasmas is introduced. Grad B and curv B drifts into the scrape-off layer (SOL) are balanced against near-sonic parallel flows out of the SOL, to the divertor plates. The overall particle flow pattern posited is a modification for open field lines of Pfirsch–Schlüter flows to include order-unity sinks to the divertors. These assumptions result in an estimated SOL width of ∼2aρ p /R. They also result in a first-principles calculation of the particle confinement time of H-mode plasmas, qualitatively consistent with experimental observations. It is next assumed that anomalous perpendicular electron thermal diffusivity is the dominant source of heat flux across the separatrix, investing the SOL width, derived above, with heat from the main plasma. The separatrix temperature is calculated based on a two-point model balancing power input to the SOL with Spitzer–Härm parallel thermal conduction losses to the divertor. This results in a heuristic closed-form prediction for the power scrape-off width that is in reasonable quantitative agreement both in absolute magnitude and in scaling with recent experimental data. Further work should include full numerical calculations, including all magnetic and electric drifts, as well as more thorough comparison with experimental data.

Catastrophe in the stochastic layer due to dipole perturbation for a single-null divertor Tokamak

International Nuclear Information System (INIS)

Ali, H.; Watson, M.; Punjabi, A.; Boozer, A.

1996-01-01

We use the method of maps developed by Punjabi and Boozer to investigate the motion of magnetic field lines in stochastic scrape-off layer in the presence of dipole perturbation of a single-null divertor Tokamak. This method is based on the idea that the magnetic field line trajectories in a divertor tokamak are mathematically equivalent to a single degree of freedom, time dependent Hamiltonian System, and that the basic features of motion near a separatrix broadened by asymmetric perturbations are generic for such Hamiltonian and near-Hamiltonian systems. The magnetic topology of a single-null divertor tokamak with the effects on dipole perturbations is represented by the Symmetric Simple Map followed by Dipole Map. We have found that as the amplitude of the dipole perturbation increases, the width of the stochastic layer also increases. At some critical value of the amplitude is reached, there is a catastrophic increase in the width of stochastic layer. This may have significant implications for tokamak divertor physics

Electrostatic and magnetic fluctuations in the proximity of the velocity shear layer in the TJ-I Tokamak

International Nuclear Information System (INIS)

Garcia-Cortes, I.; Pedrosa, M.A.; Hidalgo, C.

1992-01-01

The structure of the electrostatic and magnetic turbulence changes in the proximity of the naturally velocity shear layer in the TJ-I tokamak. A decorrelation in the broad-band magnetic fluctuations and a decreasing in the density fluctuation levels have been observed in the proximity (scrape-off layer side) of the shear layer. The results are interpreted in terms of turbulence characteristics modified by sheared poloidal flows or/and magnetic configuration. (author) 8 fig. 16 ref

Gyrofluid computations of filament dynamics in tokamak scrape-off layers

International Nuclear Information System (INIS)

Wiesenberger, M.

2014-01-01

This work contributes to the theoretical modelling, the simulation, and the analysis of plasma filaments in the scrape-off layer of magnetic fusion devices. A gyrofluid model that explicitly incorporates high fluctuation amplitudes and finite Larmor radius effects is used. First, the derivation of gyrofluid equations is thoroughly reviewed. Lie transform perturbation methods are able to remove the fast space-time scales associated with the fast gyration of particles in a strong background magnetic field. This form of perturbation theory conserves the Hamiltonian nature of the system while not being restricted to canonical coordinates. Variational methods lead to the gyrofluid equations. We explicitly derive an energy theorem and consistency equations that are necessary for the theorem to hold. The next step is to carefully choose, describe, and implement numerical methods necessary for computational studies. The discontinuous Galerkin methods have proven very flexible and easy to implement in our newly developed formulation. Furthermore, we show how field-aligned derivatives can be numerically treated in a cylindrical coordinate system. The numerical code is able to harness the computational power of today's parallel hardware architecture. A GPU, an OpenMP, and an MPI version of the source code for two- and three-dimensional simulations have been carefully planned, implemented, and tested. In the course of this work, we develop a novel conservative space discretization of the two-dimensional Poisson bracket extending the well-known Arakawa scheme to a high order discontinuous Galerkin method. Together with an existing discretization for the Laplacian we are able to discretize the two-dimensional incompressible Navier-Stokes and Euler equations. Simulations confirm the high order and the conservative properties of our method. Finally, we show that we can numerically solve the nonlinear polarization equation in the context of a mass and energy conserving, two

Adaptive grids and numerical fluid simulations for scrape-off layer plasmas

International Nuclear Information System (INIS)

Klingshirn, Hans-Joachim

2010-01-01

Magnetic confinement nuclear fusion experiments create plasmas with local temperatures in excess of 100 million Kelvin. In these experiments the scrape-off layer, which is the plasma region in direct contact with the device wall, is of central importance both for the quality of the energy confinement and the wall material lifetime. To study the behaviour of the scrape-off layer, in addition to experiments, numerical simulations are used. This work investigates the use of adaptive discretizations of space and compatible numerical methods for scrape-off layer simulations. The resulting algorithms allow dynamic adaptation of computational grids aligned to the magnetic fields to precisely capture the strongly anisotropic energy and particle transport in the plasma. The methods are applied to the multi-fluid plasma code B2, with the goal of reducing the runtime of simulations and extending the applicability of the code.

Poloidal electric field and variation of radial transport during ICRF heating in the JET scrape-off layer

International Nuclear Information System (INIS)

Clement, S.; Tagle, J.A.; Bures, M.; Vince, J.; Kock, L. de; Stangeby, P.C.

1989-01-01

The highly anomalous perpendicular transport in the plasma edge of a tokamak is generally attributed to plasma turbulence, primarily to density and electrostatic potential fluctuations. The edge transport could be modified by changing the geometry of objects in contact with the plasma (limiters, radio frequency antennae ...) and during additional heating experiments. Poloidal asymmetries in the scrape-off layer (SOL) in tokamaks using poloidal limiters (eg. ALCATOR-C) have been recently reported, indicating a poloidal asymmetry in cross-field transport. A poloidal ring limiter obstructs communications between different flux tubes in the SOL, thus permitting poloidal asymmetries in n e and T e to develop if D perpendicular is θ-dependent. When JET was operated with discrete limiters, equivalent to a single toroidal limiter at the outside mid-plane, little poloidal variation in the SOL plasma properties was observed. Currently JET is operated with two complete toroidal belt limiters located approximately one meter above and below the outside mid-plane. This configuration breaks the SOL into two regions: the low field side SOL (LFS), between the limiters, and the rest of the SOL on the high field side (HFS). Differences on the scrape-off lengths in the two SOLs are reported here, indicating that cross-field transport is faster on the LFS-SOL, in agreement with observations made on ASDEX and T-10. (author) 8 refs., 6 figs

Computations of intermittent transport in scrape-off layer plasmas

DEFF Research Database (Denmark)

Garcia, O.E.; Naulin, V.; Nielsen, A.H.

2004-01-01

in the form of blobs. These structures propagate far into the scrape-off layer where they are dissipated due to transport along open magnetic field lines. From single-point recordings it is shown that the blobs have asymmetric conditional wave forms and lead to positively skewed and flattened probability......Two-dimensional fluid simulations of interchange turbulence for geometry and parameters relevant for the scrape-off layer of magnetized plasmas are presented. The computations, which have distinct plasma production and loss regions, reveal bursty ejection of particles and heat from the bulk plasma...... distribution functions. The radial propagation velocity may reach one-tenth of the sound speed. These results are in excellent agreement with recent experimental measurements....

Studies of mobile dust in scrape-off layer plasmas using silica aerogel collectors

Energy Technology Data Exchange (ETDEWEB)

Bergsaker, H., E-mail: henricb@kth.se [Division of Space and Plasma Physics, Association EURATOM-VR, School of Electrical Engineering, Royal Institute of Technology KTH, SE-10044 Stockholm (Sweden); Ratynskaia, S. [Division of Space and Plasma Physics, Association EURATOM-VR, School of Electrical Engineering, Royal Institute of Technology KTH, SE-10044 Stockholm (Sweden); Litnovsky, A. [Institut fur Energieforschung - Plasmaphysik, Forschungszentrum Julich, Trilateral Euregio Cluster, Association EURATOM-FZ Julich, D-52425 Julich (Germany); Ogata, D. [Division of Space and Plasma Physics, Association EURATOM-VR, School of Electrical Engineering, Royal Institute of Technology KTH, SE-10044 Stockholm (Sweden); Sahle, W. [Functional Materials Division, KTH-Electrum 229, Isafjordsgatan 22, SE-16440 Stockholm (Sweden)

2011-08-01

Dust capture with ultralow density silica aerogel collectors is a new method, which allows time resolved in situ capture of dust particles in the scrape-off layers of fusion devices, without substantially damaging the particles. Particle composition and morphology, particle flux densities and particle velocity distributions can be determined through appropriate analysis of the aerogel surfaces after exposure. The method has been applied in comparative studies of intrinsic dust in the TEXTOR tokamak and in the Extrap T2R reversed field pinch. The analysis methods have been mainly optical microscopy and SEM. The method is shown to be applicable in both devices and the results are tentatively compared between the two plasma devices, which are very different in terms of edge plasma conditions, time scale, geometry and wall materials.

Ion and electron parameters in the alcator C tokamak scrape-off region

International Nuclear Information System (INIS)

Wan, A.S.H.

1986-05-01

Janus is a bi-directional, multi-functional edge probe used to diagnose the ion and electron parameters in the Alcator C tokamak scrape-off region. Two mirror image sets of diagnostics are aligned to face the electron and ion sides along magnetic field lines. Each set of diagnostics consists of a retarding-field energy analyzer (RFEA), a Langmuir probe, and a calorimeter. The RFEA can alternatively sample both the ion and electron parallel energy distribution functions during a tokamak discharge. From the Langmuir probe, one can infer electron temperature, density, and the plasma floating potential. Simple Langmuir probe theory is found to yield the best agreement between the measured Langmuir probe characteristics and the RFEA-inferred T/sub e/. The calorimeter independently detects the total parallel heat flux incident to an electrically floating plate. The measured sheath transmission coefficient, however, is typically lower than the theoretically predicted value by a factor of approx.3. Together these diagnostics enable detailed, localized edge plasma characterization on Alcator C

Turbulent transport regimes and the scrape-off layer heat flux width

Science.gov (United States)

Myra, J. R.; D'Ippolito, D. A.; Russell, D. A.

2015-04-01

Understanding the responsible mechanisms and resulting scaling of the scrape-off layer (SOL) heat flux width is important for predicting viable operating regimes in future tokamaks and for seeking possible mitigation schemes. In this paper, we present a qualitative and conceptual framework for understanding various regimes of edge/SOL turbulence and the role of turbulent transport as the mechanism for establishing the SOL heat flux width. Relevant considerations include the type and spectral characteristics of underlying instabilities, the location of the gradient drive relative to the SOL, the nonlinear saturation mechanism, and the parallel heat transport regime. We find a heat flux width scaling with major radius R that is generally positive, consistent with the previous findings [Connor et al., Nucl. Fusion 39, 169 (1999)]. The possible relationship of turbulence mechanisms to the neoclassical orbit width or heuristic drift mechanism in core energy confinement regimes known as low (L) mode and high (H) mode is considered, together with implications for the future experiments.

Turbulent transport regimes and the scrape-off layer heat flux width

International Nuclear Information System (INIS)

Myra, J. R.; D'Ippolito, D. A.; Russell, D. A.

2015-01-01

Understanding the responsible mechanisms and resulting scaling of the scrape-off layer (SOL) heat flux width is important for predicting viable operating regimes in future tokamaks and for seeking possible mitigation schemes. In this paper, we present a qualitative and conceptual framework for understanding various regimes of edge/SOL turbulence and the role of turbulent transport as the mechanism for establishing the SOL heat flux width. Relevant considerations include the type and spectral characteristics of underlying instabilities, the location of the gradient drive relative to the SOL, the nonlinear saturation mechanism, and the parallel heat transport regime. We find a heat flux width scaling with major radius R that is generally positive, consistent with the previous findings [Connor et al., Nucl. Fusion 39, 169 (1999)]. The possible relationship of turbulence mechanisms to the neoclassical orbit width or heuristic drift mechanism in core energy confinement regimes known as low (L) mode and high (H) mode is considered, together with implications for the future experiments

Measurement of flow in the scrape-off layer of TdeV

International Nuclear Information System (INIS)

MacLatchy, C.S.; Gunn, J.P.; Boucher, C.; Poirier, D.A.; Stansfield, B.L.; Zuzak, W.W.

1992-01-01

Two techniques are used to monitor the flow in the scrape-off layer of Tokamak de Varennes (TdeV); one is based on a new multipin Langmuir/Mach probe called Gundestrup while the other depends on the measurement of the upstream/downstream asymmetry of the power absorbed by a test limiter inserted into the plasma edge. Gundestrup has been used to measure the components of velocity parallel and perpendicular to the magnetic field as a function of the radial electric field. Both components vary linearly with the radial field and inversely as the magnetic field (U parallel ∝E r /B θ and U perpendicular to ∝E r /B). The pattern of power deposition on the test limiter implies that the flow is in the same direction as that measured by Gundestrup and the e-folding length for the power deposition is in agreement with Gundestrup measurements of temperature and density. The test limiter observations indicate that the flow reverses just inside the separatrix. (orig.)

Intermittent fluctuations in the Alcator C-Mod scrape-off layer for ohmic and high confinement mode plasmas

Science.gov (United States)

Garcia, O. E.; Kube, R.; Theodorsen, A.; LaBombard, B.; Terry, J. L.

2018-05-01

Plasma fluctuations in the scrape-off layer of the Alcator C-Mod tokamak in ohmic and high confinement modes have been analyzed using gas puff imaging data. In all cases investigated, the time series of emission from a single spatially resolved view into the gas puff are dominated by large-amplitude bursts, attributed to blob-like filament structures moving radially outwards and poloidally. There is a remarkable similarity of the fluctuation statistics in ohmic plasmas and in edge localized mode-free and enhanced D-alpha high confinement mode plasmas. Conditionally averaged waveforms have a two-sided exponential shape with comparable temporal scales and asymmetry, while the burst amplitudes and the waiting times between them are exponentially distributed. The probability density functions and the frequency power spectral densities are similar for all these confinement modes. These results provide strong evidence in support of a stochastic model describing the plasma fluctuations in the scrape-off layer as a super-position of uncorrelated exponential pulses. Predictions of this model are in excellent agreement with experimental measurements in both ohmic and high confinement mode plasmas. The stochastic model thus provides a valuable tool for predicting fluctuation-induced plasma-wall interactions in magnetically confined fusion plasmas.

Measurements and simulations of scrape-off layer flows in the DIII-D Tokamak

Energy Technology Data Exchange (ETDEWEB)

Groth, M. [Lawrence Livermore National Laboratory, General Atomics, P.O. Box 85608, Livermore, San Diego, CA 92186-5608 (United States); General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States)], E-mail: groth@fusion.gat.com; Porter, G.D. [Lawrence Livermore National Laboratory, General Atomics, P.O. Box 85608, Livermore, San Diego, CA 92186-5608 (United States); Boedo, J.A. [University of California San Diego, La Jolla, CA 92093 (United States); Brooks, N.H. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Isler, R.C. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); West, W.P.; Bray, B.D. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Fenstermacher, M.E. [Lawrence Livermore National Laboratory, General Atomics, P.O. Box 85608, Livermore, San Diego, CA 92186-5608 (United States); Groebner, R.J.; Leonard, A.W. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Moyer, R.A. [University of California San Diego, La Jolla, CA 92093 (United States); Rognlien, T.D. [Lawrence Livermore National Laboratory, General Atomics, P.O. Box 85608, Livermore, San Diego, CA 92186-5608 (United States); Watkins, J.G. [Sandia National Laboratory, P.O. Box 5800, Albuquerque, NM 87185 (United States); Yu, J.H. [University of California San Diego, La Jolla, CA 92093 (United States)

2009-06-15

Flow velocities of the order 10-20 km/s in the direction of the high-field side divertor have been measured for deuterons and low charge-state carbon ions in the scrape-off layer at the crown of low-density L-mode plasmas, suggesting that these carbon ions at the crown move with the background plasma flow. Simulations with the multi-fluid edge code UEDGE including cross-field drifts due to E x B and B x {nabla}B yield calculated divertor conditions which are more consistent with the measurements, but flows at the crown that are stagnant or in the opposite direction than observed. The simulations indicate that both the ion temperature gradient force and deuteron frictional drag play a role in determining the flow direction and magnitude of low charge-state carbon ions. The effect of the assumed radial transport model, toroidal core rotation, and neutral pumping at the divertor plates on the flow at the crown is investigated.

Turbulent fluctuations and radial transport in the scrape-off layer of the ASDEX tokamak

Energy Technology Data Exchange (ETDEWEB)

Endler, M [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85740 Garching (Germany); Giannone, L. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85740 Garching (Germany); McCormick, K [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85740 Garching (Germany); Niedermeyer, H [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85740 Garching (Germany); Rudyj, A [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85740 Garching (Germany); Theimer, G [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85740 Garching (Germany); Tsois, N [NCSR ` Demokritos` , Athens (Greece); ASDEX Team

1995-04-01

Electrostatic fluctuations have been measured in the scrape-off layer of ASDEX by Langmuir probes and by observation of H{sub {alpha}} light with high poloidal and temporal resolution. It was demonstrated that these fluctuations contribute a significant, if not dominant, fraction of the ``anomalous`` radial particle transport. A model for an instability mechanism specific to the SOL is presented including density, temperature and electric potential fluctuations. From this model mixing length estimates for the radial transport and resulting density and pressure gradients in the SOL are derived and compared to measurements in the mid-plane and in the divertor of ASDEX. In spite of several simplifications in the model a quantitative agreement up to factors of 1-3 and a qualitative agreement for variations of discharge parameters is achieved between the model predictions and the measurements. ((orig.)).

Turbulent fluctuations and radial transport in the scrape-off layer of the ASDEX tokamak

International Nuclear Information System (INIS)

Endler, M.; Giannone, L.; McCormick, K.; Niedermeyer, H.; Rudyj, A.; Theimer, G.; Tsois, N.

1995-01-01

Electrostatic fluctuations have been measured in the scrape-off layer of ASDEX by Langmuir probes and by observation of H α light with high poloidal and temporal resolution. It was demonstrated that these fluctuations contribute a significant, if not dominant, fraction of the ''anomalous'' radial particle transport. A model for an instability mechanism specific to the SOL is presented including density, temperature and electric potential fluctuations. From this model mixing length estimates for the radial transport and resulting density and pressure gradients in the SOL are derived and compared to measurements in the mid-plane and in the divertor of ASDEX. In spite of several simplifications in the model a quantitative agreement up to factors of 1-3 and a qualitative agreement for variations of discharge parameters is achieved between the model predictions and the measurements. ((orig.))

Plasma boundary phenomena in tokamaks

International Nuclear Information System (INIS)

Stangeby, P.C.

1989-06-01

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

A theoretical interpretation of the main scrape-off layer heat-flux width scaling for tokamak inner-wall limited plasmas

Czech Academy of Sciences Publication Activity Database

Halpern, F.D.; Horáček, Jan; Pitts, R. A.; Ricci, P.

2016-01-01

Roč. 58, č. 8 (2016), č. článku 084003. ISSN 0741-3335 R&D Projects: GA ČR(CZ) GAP205/12/2327 Institutional support: RVO:61389021 Keywords : edge plasma * heat-flux width * scrape-off layer Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.392, year: 2016 http://iopscience.iop.org/article/10.1088/0741-3335/58/8/084003/meta

Gyrokinetic particle simulation of neoclassical transport in the pedestal/scrape-off region of a tokamak plasma

Energy Technology Data Exchange (ETDEWEB)

Ku, S [Courant Institute of Mathematical Sciences, New York University (United States); Chang, C-S [Courant Institute of Mathematical Sciences, New York University (United States); Adams, M [Columbia University (United States); Cummings, J [California Institute of Technology (United States); Hinton, F [Hinton Associates (United States); Keyes, D [Columbia University (United States); Klasky, S [Oak Ridge National Laboratory (United States); Lee, W [Princeton Plasma Physics Laboratory (United States); Lin, Z [University of California at Irvine (United States); Parker, S [University of Colorado at Boulder (United States)

2006-09-15

A gyrokinetic neoclassical solution for a diverted tokamak edge plasma has been obtained for the first time using the massively parallel Jaguar XT3 computer at Oak Ridge National Laboratory. The solutions show similar characteristics to the experimental observations: electric potential is positive in the scrape-off layer and negative in the H-mode layer, and the parallel rotation is positive in the scrape-off layer and at the inside boundary of the H-mode layer. However, the solution also makes a new physical discovery that there is a strong ExB convective flow in the scrape-off plasma. A general introduction to the edge simulation problem is also presented.

Plasma confinement modification and convective transport suppression in the scrape-off layer using additional gas puffing in the STOR-M tokamak

International Nuclear Information System (INIS)

Dreval, M; Hubeny, M; Ding, Y; Onchi, T; Liu, Y; Hthu, K; Elgriw, S; Xiao, C; Hirose, A

2013-01-01

The influence of short gas puffing (GP) pulses on the scrape-off layer (SOL) transport is studied. Similar responses of ion saturation current and floating potential measured near the GP injection valve and in the 90° toroidally separated cross-section suggest that the GP influence on the SOL region should be global. A drop in plasma temperature and a decrease in the rotational velocity of the plasma are observed in the SOL region immediately after the GP pulse; however, an unexpected increase in electron and ion temperatures is observed in the second stage of the plasma response. The decrease in floating potential fluctuations indicates that the turbulent transport is dumped immediately after the GP pulse. The GP-induced modification of turbulence properties in the SOL points to a convective transport suppression in the STOR-M tokamak. A substantial decrease in the skewness and kurtosis of ion saturation current fluctuations is observed in the SOL region resulting in the probability distribution function (PDF) getting closer to the Gaussian distribution. The plasma potential reduction, the change in plasma rotation and the suppression of turbulent transport in the SOL region indicate that the plasma confinement is modified after the GP injection. Some features of the H-mode-like confinement in the plasma bulk also accompany the SOL observations after application of the additional sharp GP pulse. (paper)

Measurements and Simulations of Scrape-off Layer Flows in the DIII-D Tokamak

International Nuclear Information System (INIS)

Groth, M.; Porter, G.D.; Boedo, J.A.; Brooks, N.H.; Isler, Ralph C.; West, W.P.; Bray, B.D.; Fenstermacher, M.E.; Groebner, R.J.; Leonard, A.W.; Moyer, R.A.; Rognlien, T.D.; Watkins, J.G.; Yu, J.H.

2009-01-01

Flow velocities of the order 10-20 km/s in the direction of the high-field side divertor have been measured for deuterons and low charge-state carbon ions in the scrape-off layer at the crown of low-density L-mode plasmas, suggesting that these carbon ions at the crown move with the background plasma flow. Simulations with the multi-fluid edge code UEDGE including cross-field drifts due to E x B and B x del B yield calculated divertor conditions which are more consistent with the measurements, but flows at the crown that are stagnant or in the opposite direction than observed. The simulations indicate that both the ion temperature gradient force and deuteron frictional drag play a role in determining the flow direction and magnitude of low charge-state carbon ions. The effect of the assumed radial transport model, toroidal core rotation, and neutral pumping at the divertor plates on the flow at the crown is investigated. (C) 2009 Elsevier B.V. All rights reserved.

Measurements and modelling of electrostatic fluctuations in the scrape-off layer of ASDEX

Energy Technology Data Exchange (ETDEWEB)

Endler, M; Niedermeyer, H; Giannone, L.; Holzhauer, E; Rudyj, A; Theimer, G; Tsois, N [Association Euratom-Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); ASDEX Team

1995-11-01

In the edge plasma of the ASDEX tokamak, electrostatic fluctuations were observed with Langmuir probes and in H{sub {alpha}} light with high poloidal and temporal resolution. These fluctuations contribute a significant fraction to the `anomalous` radial particle transport in the scrape-off layer (SOL). The basic properties and the dependence of the fluctuations parameters on the discharge conditions are documented. A model for an instability mechanism specific to the SOL is introduced and the experimentally observed fluctuation parameters are compared with the predictions of the linearized version of this model. For plasma temperatures above {approx} 10eV in the SOL the observed parameter dependences of the fluctuations are well reproduced by the model. By mixing length arguments the radial transport and the resulting density and pressure gradients in the SOL are estimated from the model. Their dependence on plasma temperature and density qualitatively agrees with the behaviour observed in ohmic discharges on ASDEX. (author). 54 refs, 25 figs.

Measurements and modelling of electrostatic fluctuations in the scrape-off layer of ASDEX

International Nuclear Information System (INIS)

Endler, M.; Niedermeyer, H.; Giannone, L.; Holzhauer, E.; Rudyj, A.; Theimer, G.; Tsois, N.

1995-01-01

In the edge plasma of the ASDEX tokamak, electrostatic fluctuations were observed with Langmuir probes and in H α light with high poloidal and temporal resolution. These fluctuations contribute a significant fraction to the 'anomalous' radial particle transport in the scrape-off layer (SOL). The basic properties and the dependence of the fluctuations parameters on the discharge conditions are documented. A model for an instability mechanism specific to the SOL is introduced and the experimentally observed fluctuation parameters are compared with the predictions of the linearized version of this model. For plasma temperatures above ∼ 10eV in the SOL the observed parameter dependences of the fluctuations are well reproduced by the model. By mixing length arguments the radial transport and the resulting density and pressure gradients in the SOL are estimated from the model. Their dependence on plasma temperature and density qualitatively agrees with the behaviour observed in ohmic discharges on ASDEX. (author). 54 refs, 25 figs

Analysis of drift effects on the tokamak power scrape-off width using SOLPS-ITER

Science.gov (United States)

Meier, E. T.; Goldston, R. J.; Kaveeva, E. G.; Makowski, M. A.; Mordijck, S.; Rozhansky, V. A.; Senichenkov, I. Yu; Voskoboynikov, S. P.

2016-12-01

SOLPS-ITER, a comprehensive 2D scrape-off layer modeling package, is used to examine the physical mechanisms that set the scrape-off width ({λq} ) for inter-ELM power exhaust. Guided by Goldston’s heuristic drift (HD) model, which shows remarkable quantitative agreement with experimental data, this research examines drift effects on {λq} in a DIII-D H-mode magnetic equilibrium. As a numerical expedient, a low target recycling coefficient of 0.9 is used in the simulations, resulting in outer target plasma that is sheath limited instead of conduction limited as in the experiment. Scrape-off layer (SOL) particle diffusivity (D SOL) is scanned from 1 to 0.1 m2 s-1. Across this diffusivity range, outer divertor heat flux is dominated by a narrow (˜3-4 mm when mapped to the outer midplane) electron convection channel associated with thermoelectric current through the SOL from outer to inner divertor. An order-unity up-down ion pressure asymmetry allows net ion drift flux across the separatrix, facilitated by an artificial mechanism that mimics the anomalous electron transport required for overall ambipolarity in the HD model. At {{D}\\text{SOL}}=0.1 m2 s-1, the density fall-off length is similar to the electron temperature fall-off length, as predicted by the HD model and as seen experimentally. This research represents a step toward a deeper understanding of the power scrape-off width, and serves as a basis for extending fluid modeling to more experimentally relevant, high-collisionality regimes.

Finite mean-free-path effects in tokamak scrape-off layers

International Nuclear Information System (INIS)

Cohen, R.H.; Rognlien, T.D.; Xu, X.Q.; Bernstein, I.B.; Chen, Q.

1993-01-01

When the electron mean free path (mfp) becomes bigger than about 1/10 of the parallel electron-temperature gradient scale length, it is well-known that departures from the Spitzer thermal conductivity become important. These departures are commonly modeled by limiting the parallel heat flux q parallel to an empirically determined fraction of nT e v te where v te is the electron thermal speed. The use of flux limit expressions in 2-D scrape-off layer (SOL) modeling codes leads to the qualitatively correct result that the electron temperature drops along a field line as heat is leaked by radial transport, but perhaps for the wrong reasons. In particular the flux-limiting form is demonstratably incorrect in the long-mfp limit. Here the authors re-examine this issue. Recognizing that the heat flux is carried by superthermal electrons, they formulate a linearized 3-D Fokker-Planck problem. They depart from previous treatments by noting that, for typical SOL parameters, the superthermal particles classically carrying the bulk of the heat flux have long mean free paths and are in the loss cone, and so are absent from the distribution function. They argue that this is a key feature which will reduce the heat flux below that calculated previously. They outline several strategies for reducing the Fokker-Planck equation to analytically tractable and/or computationally more tractable forms. In particular, they discuss a diffusion model for the isotropic part of the distribution function, its numerical implementation, and limits in which approximate analytic solutions can be obtained. They also present a heuristic model for the heat flux that accounts for the physical effects discussed above and which has the correct asymptotic limits for small and large mean free path. They compare this model and preliminary analytic and numerical results from the diffusion model with Monte Carlo simulations

Langmuir probe measurements of the scrape-off plasma in ISX-A

International Nuclear Information System (INIS)

Namkung, W.; England, A.C.; Eldridge, O.C.

1978-11-01

A fixed double Langmuir probe was used to investigate the temporal behavior of the scrape-off plasma in the ISX-A tokamak. During gas puffing, the ion saturation current dropped rapidly to a very low level while the line average density showed a steady increase. This sudden transition was due mainly to a density change of more than a factor of five while the electron temperature remained relatively constant at approximately 10 eV. This behavior was easily observed at points away from the limiter with mild and moderate gas puffing rates, and near the inner edge of the limiter with strong gas puffing. In order to explain the phenomenon, it is suggested that there may be two distinct layers in the scrape-off plasma and that the boundary between the layers moves inward toward the limiter. The existence of the boundary has been confirmed indirectly by sudden shifts of the plasma during feedback control experiments

Two-dimensional analysis of limiter/divertor transition in scrape-off layer plasmas

International Nuclear Information System (INIS)

Ueda, N.; Itoh, K.; Itoh, S.I.

1989-01-01

The structures of scrape-off layer and divertor plasmas have been studied numerically with a neutral code and a two-dimensional fluid code. Doublet-III is taken as an example for an open divertor configuration. A decisive parameter is the distance between the plasma surface (determined by the magnetic separatrix) and the limiter, which is varied in order to assess the interaction of the plasma with the limiter as well as the effect of neutrals on the main plasma. The minimum value of the limiter clearance needed to prevent plasma-limiter interaction is determined. The scaling of the edge temperature and the dependence of the e-folding length of the scrape-off layer plasma on the heating power are obtained. (author). 16 refs, 17 figs

Linear and nonlinear studies of resistive-ballooning modes in a tokamak edge plasma with scrape-off layer

International Nuclear Information System (INIS)

Lau, Y.T.; Novakovskii, S.V.; Drake, J.F.

1996-01-01

We will present 2D linear and 3D nonlinear studies of resistive-ballooning modes in tokamak edge plasmas which include a closed flux region, as well as a limiter scrape-off layer (SOL) region. These studies therefore go beyond most earlier work, where the stability of the edge in the closed flux region and in the SOL have been considered separately. A 2D linear code, 2D-BALLOON, examines the stability of these curvature driven modes and provides the complete 2D eigenfunction spanning the closed flux surface region as well the open field line region. The sheath boundary condition in the SOL introduces an important new parameter λ = (m e /m i ) 1/2 v ei qR/v Te . This parameter plays a significant role in determining the stability of these modes in both the closed flux and SOL regions because of the radial coupling across the last closed flux surface (LCFS). For small λ the spectrum of unstable modes is broad and extends into the low toroidal mode number exclamation point regime where the spatial structure is flute-like. The amplitude for these modes is larger in the SOL compared to the closed flux region. However when A is increased, the low mode numbers are strongly stabilized and the high mode numbers which are strongly ballooning are the dominant modes. In this regime the radial modes straddle the LCFS. In both these cases, the variation in the plasma density is necessary for the radial localization. In the three-dimensional nonlinear simulations, we have solved a set of fluid equations in a toroidal geometry with both the closed flux region and the SOL. The introduction of the SOL to the twisted tube for the closed flux region, has been a major addition to our 3D code. We find that the turbulent transport in the SOL drops significantly as A is increased, which is consistent with our expectations from the 2D linear code results

Comparison of Scrape-off Layer Turbulence in Alcator C-Mod with Three Dimensional Gyrofluid Computations

International Nuclear Information System (INIS)

Zweben, S.J.; Scott, B.D.; Terry, J.L.; LaBombard, B.; Hughes, J.W.; Stotler, D.P.

2009-01-01

This paper describes quantitative comparisons between turbulence measured in the scrape-off layer (SOL) of Alcator C-Mod (S. Scott, A. Bader, M. Bakhtiari et al., Nucl. Fusion 47, S598 (2007)) and three dimensional computations using electromagnetic gyrofluid equations in a two-dimensional tokamak geometry. These comparisons were made for the outer midplane SOL for a set of inner-wall limited, near-circular Ohmic plasmas. The B field and plasma density were varied to assess gyroradius and collisionality scaling. The poloidal and radial correlation lengths in the experiment and computation agreed to within a factor of 2 and did not vary significantly with either B or density. The radial and poloidal propagation speeds and the frequency spectra and poloidal k-spectra also agreed fairly well. However, the autocorrelation times and relative Da fluctuation levels were higher in the experiment by more than a factor of 2. Possible causes for these disagreements are discussed.

Dynamics of impurities in the scrape-off layer

International Nuclear Information System (INIS)

Stangeby, P.C.; Commission of the European Communities, Abingdon

1988-01-01

Impurity modelling of the Scrape-Off Layer, SOL, is reviewed. Simple analytic models are sometimes adequate for relating central impurity levels to edge plasma conditions and for explaining the patterns of net erosion/deposition found on limiters. More sophisticated approaches, which are also necessary, are categorized and reviewed. A plea is made for the acquisition of a more comprehensive data base of edge plasma properties since reliable impurity modelling appears to be dependent on more extensive use of experimental input. (author)

Far scrape-off layer particle and heat fluxes in high density

DEFF Research Database (Denmark)

Müller, H. W.; Bernert, M.; Carralero, D.

2014-01-01

The far scrape-off layer transport is studied in ASDEX Upgrade H-mode discharges with high divertor neutral density N0,div, high power across the separatrix Psep and nitrogen seeding to control the divertor temperature. Such conditions are expected for ITER but usually not investigated in terms...

The scrape-off layer in a finite-aspect-ratio Torus: The influence of limiter position

International Nuclear Information System (INIS)

Harbour, P.J.; Loarte, A.

1995-01-01

The effect on the scrape-off layer (SOL) of changing the position of a tokamak limiter from the low field side (LFS) to the high field side (HFS) of the plasma is considered. Conservation of magnetics flux in the SOL requires that the area, A parallel , for flow of particles and energy parallel to the magnetic field, be smaller on the HFS. The effect that this reduction in A parallel has on the SOL is that, when the limiter is on the HFS, as compared with the LFS, then the characteristic scrape-off thickness is increased and the plasma particle and energy densities are higher. Such increases are described. They were observed in an experiment in JET in which the plasma was limited first on the LFS and then, during the same discharge, on the HFS. The magnitude of the effect was larger than expected, with the line integrated density of the SOL, nλ n , having an average increase by a factor of 5.8 ± 1.1, and an increase by a factor of 7 at the respective limiter. Moreover, there was an increase in the characteristic scrape-off thickness for power flow, λ p , by a factor of 2.5 ± 0.4 on average, and a corresponding increase by a factor of 3.0 at the respective limiter. Such an increase in nλ n can substantially improve the screening of recycled particles and can increase wall pumping, although not strongly in the experiments described, which were in helium to avoid these effects. The large increase in λ p measured in JET is useful in spreading power over a larger area of a limiter on the HFS. Evidence is presented which suggests that either the observed enhancement of the effect on the SOL above expectations is due to a reduction in the Mach number at the limiter sheath when on the HFS or there are fundamental problems with the modelling procedure usually used for the SOL. Related experiments are discussed. (author). 29 refs, 3 figs, 3 tabs

Numerical simulation of feedback stabilization of axisymmetric modes in tokamaks using driven halo currents

International Nuclear Information System (INIS)

Jardin, S.C.; Schmidt, J.A.

1998-01-01

The Tokamak Simulation Code (TSC) has been used to model a new method of feedback stabilization of the axisymmetric instability in tokamaks using driven halo (or scrape-off layer) currents. The method appears to be feasible for a wide range of plasma edge parameters. It may offer advantages over the more conventional method of controlling this instability when applied in a reactor environment. (author)

Scrape-off layer transport and deposition studies in DIII-D

International Nuclear Information System (INIS)

Groth, M.; Allen, S. L.; Fenstermacher, M. E.; Lasnier, C. J.; Porter, G. D.; Rensink, M. E.; Rognlien, T. D.; Boedo, J. A.; Rudakov, D. L.; Brooks, N. H.; Groebner, R. J.; Leonard, A. W.; West, W. P.; Elder, J. D.; McLean, A. G.; Lisgo, S.; Stangeby, P. C.; Wampler, W. R.; Watkins, J. G.; Whyte, D. G.

2007-01-01

Trace 13 CH 4 injection experiments into the main scrape-off layer (SOL) of low density L-mode and high-density H-mode plasmas have been performed in the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] to mimic the transport and deposition of carbon arising from a main chamber sputtering source. These experiments indicated entrainment of the injected carbon in plasma flow in the main SOL, and transport toward the inner divertor. Ex situ surface analysis showed enhanced 13 C surface concentration at the corner formed by the divertor floor and the angled target plate of the inner divertor in L-mode; in H-mode high surface concentration was found both at the corner and along the surface bounding the private flux region inboard of the outer strike point. Interpretative modeling was made consistent with these experimental results by imposing a parallel carbon ion flow in the main SOL toward the inner target, and a radial pinch toward the separatrix. Predictive modeling carried out to better understand the underlying plasma transport processes suggests that the deuterium flow in the main SOL is related to the degree of detachment of the inner divertor leg. These simulations show that carbon ions are entrained with the deuteron flow in the main SOL via frictional coupling, but higher charge-state carbon ions may be suspended upstream of the inner divertor X-point region due to balance of the friction force and the ion temperature gradient force

Blob sizes and velocities in the Alcator C-Mod scrape-off layer

DEFF Research Database (Denmark)

Kube, R.; Garcia, O.E.; LaBombard, B.

A new blob-tracking algorithm for the GPI diagnostic installed in the outboard-midplane of Alcator C-Mod is developed. I t tracks large-amplitude fluctuations propagating through the scrape-off layer and calculates blob sizes and velocities. We compare the results of this method to a blob velocity...

Effect of lower hybrid waves on turbulence and transport of particles and energy in the FTU tokamak scrape-off layer plasma

International Nuclear Information System (INIS)

Ridolfini, V Pericoli

2011-01-01

All the main features of the scrape-off layer turbulence, magnitude, frequency spectrum and perpendicular wave vector, ξ t , are strongly affected by the injection of lower hybrid (LH) power into the FTU tokamak. The governing parameters are the local last closed magnetic surface values of density, n e,LCMS , and temperature, T e,LCMS . n e,LCMS determines the perpendicular wave vector of the LH waves, which is a key parameter for the multiple scattering processes, and together with T e,LCMS the collisionality that exerts a stabilizing effect on the fluctuations. This effect, still to be examined in the light of theoretical models, leads to an asymptotic value for the fluctuation relative amplitude in the ohmic phase close to 25%, and ∼10% in the LH phase, or even less, since the saturation level is not yet attained. The LH waves also can strongly raise ξ t , about 3 times, and double the root mean square frequency. The transfer of momentum and energy in the mutual scattering of LH and turbulence 'waves' drives these changes. An increase also of the cross-correlation between temperature and electric potential fluctuations should occur in order to explain the magnitude of the fluctuation amplitude drop and the large increment of the temperature e-folding decay, by more than a factor of 2.5. Particle transport, however, does not appear to be affected to a large extent-the density e-folding decay length is almost unchanged but the power flow typical length rises by about a factor of 1.5, which is a relevant figure in view of the problem of mitigating the power loads on divertor targets in future reactors. These changes are confined mainly within the flux tube connected with the LH waves launching antenna, but start to spread significantly out of it at high plasma densities.

Effect of chaos on plasma filament dynamics and turbulence in the scrape-off layer

International Nuclear Information System (INIS)

Meyerson, D.; Waelbroeck, F.; Horton, W.; Michoski, C.

2014-01-01

Naturally occurring error fields as well as resonant magnetic perturbations applied for stability control are known to cause magnetic field-line chaos in the scrape-off layer (SOL) region of tokamaks. Here, 2D simulations with the BOUT++ simulation framework are used to investigate the effect of the field-line chaos on the SOL and in particular on its width and peak particle flux. The chaos enters the SOL dynamics only through the connection length, which is evaluated using a Poincaré map. The variation of experimentally relevant quantities, such as the SOL gradient length scale and the intermittency of the particle flux in the SOL, is described as a function of the strength of the magnetic perturbation. It is found that the effect of the chaos is to broaden the profile of the sheath-loss coefficient, which is proportional to the inverse connection length. That is, the SOL transport in a chaotic field is equivalent to that in a model where the sheath-loss coefficient is replaced by its average over the unperturbed flux surfaces. The model does not include the effects of chaotic features other than the parallel connection length

Theory-based scaling of the SOL width in circular limited tokamak plasmas

International Nuclear Information System (INIS)

Halpern, F.D.; Ricci, P.; Labit, B.; Furno, I.; Jolliet, S.; Loizu, J.; Mosetto, A.; Arnoux, G.; Silva, C.; Gunn, J.P.; Horacek, J.; Kočan, M.; LaBombard, B.

2013-01-01

A theory-based scaling for the characteristic length of a circular, limited tokamak scrape-off layer (SOL) is obtained by considering the balance between parallel losses and non-linearly saturated resistive ballooning mode turbulence driving anomalous perpendicular transport. The SOL size increases with plasma size, resistivity, and safety factor q. The scaling is verified against flux-driven non-linear turbulence simulations, which reveal good agreement within a wide range of dimensionless parameters, including parameters closely matching the TCV tokamak. An initial comparison of the theory against experimental data from several tokamaks also yields good agreement. (letter)

Determining advection mechanism of plasma filaments in the scrape-off layer of MAST

International Nuclear Information System (INIS)

Higgins, D; Hnat, B; Kirk, A; Tamain, P; Ben Ayed, N

2012-01-01

The scrape-off layer (SOL) of fusion devices is typically composed of filamentary structures that propagate with a high radial velocity away from the bulk plasma. When radial and parallel transport times are comparable, these coherent structures constitute an intermittent heat and particle flux which can reach the material wall; in time causing wear to plasma facing components. Qualitative models predict that the parallel currents, driven by the divertor sheath, have a direct impact on this radial velocity. In this work, the predictions for radial velocity of plasma filaments in the SOL from models are tested against data from the MAST tokamak and simulation. We apply a statistical method of window averaging to MAST Langmuir probe data in order to examine the scaling of the radial velocity of filaments with the plasma density inside the filaments. Our analysis strongly suggests that the radial dynamics emerge from the competition of multiple mechanisms and not from a single process. At intermediate distances from the bulk plasma, a new model proposed here, in which the parallel current depends on a constant target density appears to be the most relevant for the MAST plasma. This is confirmed using a TOKAM2D simulation with a modified parallel current term.

Three-dimensional simulations of plasma turbulence in the RFX-mod scrape-off layer and comparison with experimental measurements

Science.gov (United States)

Riva, Fabio; Vianello, Nicola; Spolaore, Monica; Ricci, Paolo; Cavazzana, Roberto; Marrelli, Lionello; Spagnolo, Silvia

2018-02-01

The tokamak scrape-off layer (SOL) plasma dynamics is investigated in a circular limiter configuration with a low edge safety factor. Focusing on the experimental parameters of two ohmic tokamak inner-wall limited plasma discharges in RFX-mod [Sonato et al., Fusion Eng. Des. 74, 97 (2005)], nonlinear SOL plasma simulations are performed with the GBS code [Ricci et al., Plasma Phys. Controlled Fusion 54, 124047 (2012)]. The numerical results are compared with the experimental measurements, assessing the reliability of the GBS model in describing the RFX-mod SOL plasma dynamics. It is found that the simulations are able to quantitatively reproduce the RFX-mod experimental measurements of the electron plasma density, electron temperature, and ion saturation current density (jsat) equilibrium profiles. Moreover, there are indications that the turbulent transport is driven by the same instability in the simulations and in the experiment, with coherent structures having similar statistical properties. On the other hand, it is found that the simulation results are not able to correctly reproduce the floating potential equilibrium profile and the jsat fluctuation level. It is likely that these discrepancies are, at least in part, related to simulating only the tokamak SOL region, without including the plasma dynamics inside the last close flux surface, and to the limits of applicability of the drift approximation. The turbulence drive is then identified from the nonlinear simulations and with the linear theory. It results that the inertial drift wave is the instability driving most of the turbulent transport in the considered discharges.

Scrape-off layer power flux measurements in the Tore Supra tokamak

Czech Academy of Sciences Publication Activity Database

Gunn, J. P.; Dejarnac, Renaud; Devynck, P.; Fedorczak, N.; Fuchs, Vladimír; Gil, C.; Kočan, M.; Komm, Michael; Kubič, M.; Lunt, T.; Monier-Garbet, P.; Pascal, J.-Y.; Saint-Laurent, F.

2013-01-01

Roč. 438, suppl (2013), S184-S188 ISSN 0022-3115. [International Conference on Plasma-Surface Interactions in Controlled Fusion Devices/20./. Aachen, 21.05.2012-25.05.2012] Institutional support: RVO:61389021 Keywords : plasma * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.016, year: 2013 http://www.sciencedirect.com/science/article/pii/S0022311513000639#

Impurity transport calculations for the limiter shadow region of a tokamak

International Nuclear Information System (INIS)

Claassen, H.A.; Repp, H.

1981-01-01

Impurity transport calculations are presented for the scrape-off layer of a tokamak with a poloidal ring limiter. The theory is based on the drift-kinetic equations for the impurity ions in their different ionization states. It is developed in the limit of low impurity concentrations under due consideration of electron impact ionization, Coulomb collisions with hydrogen ions streaming onto a neutralizing surface, a convection along the magnetic field, and a radial drift. The background plasma and the impurity sources at the walls enter the theory as input parameters. Numerical results are given for the radial profiles of density, temperature, particle flux, and energy flux of wall-released impurity ions as well as for the screening efficiency of the scrape-off layer neglecting impurity re-emission from the limiter. (author)

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

NARCIS (Netherlands)

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

2017-01-01

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

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

DEFF Research Database (Denmark)

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

2017-01-01

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

Filamentary probe on the COMPASS tokamak

Czech Academy of Sciences Publication Activity Database

Kovařík, Karel; Ďuran, Ivan; Stöckel, Jan; Seidl, Jakub; Adámek, Jiří; Spolaore, M.; Vianello, N.; Háček, Pavel; Hron, Martin; Pánek, Radomír

2017-01-01

Roč. 88, č. 3 (2017), č. článku 035106. ISSN 0034-6748 R&D Projects: GA MŠk(CZ) 8D15001; GA ČR(CZ) GA15-10723S; GA ČR(CZ) GA16-25074S Institutional support: RVO:61389021 Keywords : tokamak * filaments * scrape-off layer Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: 2.11 Other engineering and technologies Impact factor: 1.515, year: 2016 http://aip.scitation.org/doi/10.1063/1.4977591

Characteristics of divertor plasma and scrape-off layer in JT-60U

International Nuclear Information System (INIS)

Itami, K.; Shimada, M.; Hosogane, N.

1992-01-01

Heat flux to the divertor is measured by thermography and the heat transport in the scrape-off layer is studied in beam heated discharges of JT-60U. The heat flux onto the divertor is ∝50% of total beam power at maximum. The in-out asymmetry of the heat flux P HEAT in /P HEAT out is as large as 20-40% when the ion grad-B drift is toward the divertor. Differences in P HEAT in /P HEA T out due to the direction of ion grad-B drift are as large as large as ∝40%. A scaling of the peaking factor Y of heat flux, defined by Y=2πRfq max /P HEAT , is obtained for beam heated discharges in JT-60U with a wide range of plasma parameters. The Y corresponds to the inverse of the thickness of the scrape-off layer. From a statistical analysis, it is found that the peaking factor Y of heat flux scales as P HEAT 0.49±0.18 anti n e -0.45±0.22 q eff -0.67±0.18 . (orig.)

Relevance, Realization and stability of a cold layer at the plasma edge for fusion reactors

International Nuclear Information System (INIS)

1990-09-01

The workshop was dedicated to the realization and stability of a cold layer at the plasma edge for fusion reactors. The subjects of the communications presented were: impurity transport, and control, plasma boundary layers, power balance, radiation control and modifications, limiter discharges, tokamak density limit, Asdex divertor discharges, thermal stability of a radiating diverted plasma, plasma stability, auxiliary heating in Textor, detached plasma in Tore Supra, poloidal divertor tokamak, radiation cooling, neutral-particle transport, plasma scrape-off layer, edge turbulence

Kinetic simulations of scrape-off layer physics in the DIII-D tokamak

Directory of Open Access Journals (Sweden)

R.M. Churchill

2017-08-01

The XGCa simulation of the DIII-D tokamak in a nominally sheath-limited regime show many noteworthy features in the SOL. The density and ion temperature are higher at the low-field side, indicative of ion orbit loss. The SOL ion Mach flows are at experimentally relevant levels (Mi ∼ 0.5, with similar shapes and poloidal variation as observed in various tokamaks. Surprisingly, the ion Mach flows close to the sheath edge remain subsonic, in contrast to the typical fluid Bohm criterion requiring ion flows to be above sonic at the sheath edge. Related to this are the presence of elevated sheath potentials, eΔΦ/Te∼3−4, over most of the SOL, with regions in the near-SOL close to the separatrix having eΔΦ/Te > 4. These two results at the sheath edge are a consequence of non-Maxwellian features in the ions and electrons there.

Effect of lower hybrid waves on turbulence and transport of particles and energy in the FTU tokamak scrape-off layer plasma

Energy Technology Data Exchange (ETDEWEB)

Ridolfini, V Pericoli [ENEA-CR Frascati, Via Enrico Fermi 45-00044 Frascati, Roma (Italy)

2011-11-15

All the main features of the scrape-off layer turbulence, magnitude, frequency spectrum and perpendicular wave vector, {xi}{sub t}, are strongly affected by the injection of lower hybrid (LH) power into the FTU tokamak. The governing parameters are the local last closed magnetic surface values of density, n{sub e,LCMS}, and temperature, T{sub e,LCMS}. n{sub e,LCMS} determines the perpendicular wave vector of the LH waves, which is a key parameter for the multiple scattering processes, and together with T{sub e,LCMS} the collisionality that exerts a stabilizing effect on the fluctuations. This effect, still to be examined in the light of theoretical models, leads to an asymptotic value for the fluctuation relative amplitude in the ohmic phase close to 25%, and {approx}10% in the LH phase, or even less, since the saturation level is not yet attained. The LH waves also can strongly raise {xi}{sub t}, about 3 times, and double the root mean square frequency. The transfer of momentum and energy in the mutual scattering of LH and turbulence 'waves' drives these changes. An increase also of the cross-correlation between temperature and electric potential fluctuations should occur in order to explain the magnitude of the fluctuation amplitude drop and the large increment of the temperature e-folding decay, by more than a factor of 2.5. Particle transport, however, does not appear to be affected to a large extent-the density e-folding decay length is almost unchanged but the power flow typical length rises by about a factor of 1.5, which is a relevant figure in view of the problem of mitigating the power loads on divertor targets in future reactors. These changes are confined mainly within the flux tube connected with the LH waves launching antenna, but start to spread significantly out of it at high plasma densities.

Total hydrogen and oxygen fluxes in the edge plasma of tokamaks

International Nuclear Information System (INIS)

Kastelewicz, H.

1988-01-01

A relativistic model of the edge plasma of tokamaks is described considering the primary neutral fluxes emitted from limiter and wall. The primary neutrals, which determine essentially the particle flux balance in the plasma edge, the scrape-off layer plasma and the particles adsorbed at limiter and wall are treated as separate subsystems which are iteratively coupled through the mutual particle sinks and sources. The model is used for the calculation of total hydrogen and oxygen fluxes in edge plasma of tokamaks. The results for different fractions of and contributions to the total fluxes are illustrated and discussed

Effects of ionizing scrape-off layers on local recycling in Tore Supra pump limiter experiments

International Nuclear Information System (INIS)

Owen, L.W.; Hogan, J.T.; Klepper, C.C.; Mioduszewski, P.K.; Uckan, T.; Chatelier, M.; Loarer, T.

1992-01-01

A series of ohmic discharges with active pumping in the Tore Supra outboard pump limiter has been analyzed with the DEGAS neutrals transport code and an analysis scrape-off layer (SOL) plasma model. Pumping speed and plenum pressure measurements indicated 5--10 torr-L/s throughput with only modest effects on density (dN core /dt + source rate from ionization and dissociation of wall-desorbed molecules is seen to peak very near the radial position of the limiter throat. Consequently, a strong recycling vortex is created in the region of the limiter, with the ion flux amplified by factors of ∼2 at the outer limiter surfaces and >3 within the limiter throat. The calculations indicate that less than 30% of the pump throughput is due to first-generation ions from the core efflux, with the balance from local recycling in the strongly ionizing scrape-off layer

Analysis of plasma flow in a scrape-off layer in a tokamak

International Nuclear Information System (INIS)

Petrov, V.G.

1988-01-01

Plasma drift on the periphery of a tokamak in a magnetic tube, on sides of which coins are placed, is considered. Convection caused by toroidal particle drift is taken into account. Distribution of plasma parameters in such a tube is found. Transition from the total poloidal diaphragm to a sectioned one is traced

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

International Nuclear Information System (INIS)

Ferro, C.; Zanino, R.

1992-01-01

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

Conceptual analysis of a tokamak reactor with lithium dust jet

International Nuclear Information System (INIS)

Kuteev, B.V.; Krylov, S.V.; Sergeev, V.Yu.; Skokov, V.G.; Timokhin, V.M.

2010-01-01

The steady-state operation of tokamak reactors requires radiating a substantial part of the fusion energy dissipated in plasma to make more uniform the heat loads onto the first wall and to reduce the erosion of the divertor plates. One of the approaches to realize this goal uses injection of lithium dust jet into the scrape-off layer (SOL). A quantitative conceptual analysis of the reactor parameters with lithium dust jet injection is presented here. The effects of the lithium on the core and SOL plasma are considered. The first results of developing the lithium jet injection technology and its application to the T-10 tokamak are also presented.

Carbon impurity transport around limiters in the DITE tokamak

International Nuclear Information System (INIS)

Pitcher, C.S.; Stangeby, P.C.; Goodall, D.H.J.; Matthews, G.F.; McCracken, G.M.

1989-01-01

The transport of impurity ions originating at the limiter in a tokamak is critically dependent on the location of the ion in the boundary plasma. In the confined plasma, just inboard of the limiter, impurity ions will disperse freely into the discharge whilst in the scrape-off layer the pre-sheath plasma flow and the associated ambipolar electric field may tend to sweep impurities back to the limiter surface. In this paper we have studied, both by experiment and by theory, the transport of carbon impurity ions in the vicinity of the limiter. By comparing experimental measurements of the spatial distributions of impurities around the limiter with that predicted from a Monte Carlo computer code it appears that the parallel dispersal on closed field lines in the confined plasma is consistent with classical transport processes and that in the scrape-off layer the dispersal is indeed impeded by the pre-sheath plasma flow. (orig.)

Boundary Plasma Turbulence Simulations for Tokamaks

International Nuclear Information System (INIS)

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

2008-05-01

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 e ; T 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

Numerical scalings of the decay lengths in the scrape-off layer

DEFF Research Database (Denmark)

Militello, F.; Naulin, V; Nielsen, Anders Henry

2013-01-01

Numerical simulations of L-mode turbulence in the scrape-off layer (SOL) are used to construct power scaling laws for the characteristic decay lengths of the temperature, density and heat flux at the outer mid-plane. Most of the results obtained are in qualitative agreement with the experimental...... observations despite the known limitation of the model. Quantitative agreement is also obtained for some exponents. In particular, an almost linear inverse dependence of the heat flux decay length with the plasma current is recovered. The relative simplicity of the theoretical model used allows one to gain...

Local gas injection as a scrape-off layer diagnostic on the Alcator C-Mod tokamak

International Nuclear Information System (INIS)

Jablonski, D.F.

1996-05-01

A capillary puffing array has been installed on Alcator C-Mod which allows localized introduction of gaseous species in the scrape-off layer. This system has been utilized in experiments to elucidate both global and local properties of edge transport. Deuterium fueling and recycling impurity screening are observed to be characterized by non-dimensional screening efficiencies which are independent of the location of introduction. In contrast, the behavior of non-recycling impurities is seen to be characterized by a screening time which is dependent on puff location. The work of this thesis has focused on the use of the capillary array with a camera system which can view impurity line emission plumes formed in the region of an injection location. The ionic plumes observed extend along the magnetic field line with a comet-like asymmetry, indicative of background plasma ion flow. The flow is observed to be towards the nearest strike-point, independent of x-point location, magnetic field direction, and other plasma parameters. While the axes of the plumes are generally along the field line, deviations are seen which indicate cross-field ion drifts. A quasi-two dimensional fluid model has been constructed to use the plume shapes of the first charge state impurity ions to extract information about the local background plasma, specifically the temperature, parallel flow velocity, and radial electric field. Through comparisons of model results with those of a three dimensional Monte Carlo code, and comparisons of plume extracted parameters with scanning probe measurements, the efficacy of the model is demonstrated. Plume analysis not only leads to understandings of local edge impurity transport, but also presents a novel diagnostic technique

Local gas injection as a scrape-off layer diagnostic on the Alcator C-Mod tokamak

Energy Technology Data Exchange (ETDEWEB)

Jablonski, David F. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

1996-05-01

A capillary puffing array has been installed on Alcator C-Mod which allows localized introduction of gaseous species in the scrape-off layer. This system has been utilized in experiments to elucidate both global and local properties of edge transport. Deuterium fueling and recycling impurity screening are observed to be characterized by non-dimensional screening efficiencies which are independent of the location of introduction. In contrast, the behavior of non-recycling impurities is seen to be characterized by a screening time which is dependent on puff location. The work of this thesis has focused on the use of the capillary array with a camera system which can view impurity line emission plumes formed in the region of an injection location. The ionic plumes observed extend along the magnetic field line with a comet-like asymmetry, indicative of background plasma ion flow. The flow is observed to be towards the nearest strike-point, independent of x-point location, magnetic field direction, and other plasma parameters. While the axes of the plumes are generally along the field line, deviations are seen which indicate cross-field ion drifts. A quasi-two dimensional fluid model has been constructed to use the plume shapes of the first charge state impurity ions to extract information about the local background plasma, specifically the temperature, parallel flow velocity, and radial electric field. Through comparisons of model results with those of a three dimensional Monte Carlo code, and comparisons of plume extracted parameters with scanning probe measurements, the efficacy of the model is demonstrated. Plume analysis not only leads to understandings of local edge impurity transport, but also presents a novel diagnostic technique.

Assessment of feasibility of helium ash exhaust and heat removal by pumped-limiter in tokamak fusion reactor

International Nuclear Information System (INIS)

Hitoki, Shigehisa; Sugihara, Masayoshi; Saito, Seiji; Fujisawa, Noboru

1985-01-01

A detailed calculation of the behavior of fuel and He particles in tokamak reactor with pumped-limiter is performed by one-dimensional tokamak transport code. Energy of neutral particles flowing back from limiter chamber is calculated by two-dimensional Monte Carlo neutral code. Feasibility of He ash exhaust and heat removal by the pumped-limiter are analyzed. Following features of the pumped-limiter are clarified: (1) Electron temperature decays rapidly in radial direction in scrape-off layer, while density profile is broader than that of temperature. (2) Helium accumulation in main plasma can be kept at desired level by rather short limiter and moderate pumping system. (3) Minimum amount of tritium pumped out little depends on limiter length. (4) Although high temperature plasma in scrape-off layer could be realized by large pumping and ideal pellet injection, it is not sufficiently high to reduce the erosion of the limiter surface and the leading edge. In conclusion, He ash exhaust may be possible by the pumped-limiter, while the heat load and erosion will be so high that the pumped-limiter may not be applicable unless the boundary plasma is cooled by radiation or by some other means. (author)

Scrape-off layer profile modifications by convective cells

International Nuclear Information System (INIS)

Myra, J.R.; DIppolito, D.A.

1996-01-01

Convective cells (CC close-quote s) are important in understanding density profile modifications induced by ion cyclotron range of frequencies (ICRF) antennas. This has motivated the present work in which the effect of CC close-quote s on transport in the scrape-off layer is studied, in the regime where the density gradient scale length L n and the cell size L are comparable. Monte Carlo simulations show that closed cell convection acts to flatten the density profile, and that open cells enhance the particle flow to the wall, depleting the density and yielding profiles similar to those measured near ICRF antennas. A new one-dimensional, two-branch model of CC transport is shown to agree well with the simulations. The model gives rise to two characteristic scale lengths, only one of which is retained in the enhanced diffusion models that are applicable for L n >L. The two-branch model is expected to be useful in analyzing ICRF experiments. copyright 1996 American Institute of Physics

Numerical analysis of the impact of an RF sheath on the Scrape-Off Layer in 2D and 3D turbulence simulations

Directory of Open Access Journals (Sweden)

P. Tamain

2017-08-01

Full Text Available Motivated by Radio Frequency (RF heating studies, the response of the plasma of tokamaks to the presence of a locally polarized limiter is studied. In a first part, we use the TOKAM3X3D global edge turbulence code to analyse the impact of such biasing in a realistic geometry. Key features of experimental observations are qualitatively recovered, especially the extension of a potential and density perturbation on long, but finite, distances along connected field lines. The perturbation is also found to extend in the transverse direction. Both observations demonstrate the influence of perpendicular current loops on the plasma confirming the need for an accurate description in reduced models. In a second part, we use the TOKAM2D slab turbulence code to determine the validity of using a transverse Ohm's law for this purpose. Results indicate that a local Ohm's law with a constant and uniform perpendicular resistivity appears at least as an oversimplified description of perpendicular charge transport in a turbulent Scrape-Off Layer.

Poloidally asymmetric potential increases in tokamak scrape-off layer plasmas by radiofrequency power

International Nuclear Information System (INIS)

Diebold, D.A.; Majeski, R.; Tanaka, T.

1992-01-01

Langmuir probe data are presented which show poloidally asymmetric increases in floating potential, electron temperature and, hence, plasma potential on magnetic field lines which map to the Faraday shield of an ICRF antenna in a medium size tokamak, Phaedrus-T, during radiofrequency power injection. These data are consistent with and suggestive of the existence of radiofrequency generated sheath voltages on those field lines. (author). Letter-to-the-editor. 20 refs, 3 figs

Langmuir probe evaluation of the plasma potential in tokamak edge plasma for non-Maxwellian EEDF

Energy Technology Data Exchange (ETDEWEB)

Popov, Ts.K. [Faculty of Physics, St. Kliment Ohridski University (Bulgaria); Dimitrova, M. [Institute of Plasma Physics, Academy of Sciences of the Czech Republic v.v.i., Prague (Czech Republic); Institute of Electronics, Bulgarian Academy of Sciences, Sofia (Bulgaria); Ivanova, P. [Institute of Electronics, Bulgarian Academy of Sciences, Sofia (Bulgaria); Hasan, E. [Faculty of Physics, St. Kliment Ohridski University (Bulgaria); Institute of Electronics, Bulgarian Academy of Sciences, Sofia (Bulgaria); Horacek, J.; Dejarnac, R.; Stoeckel, J.; Weinzettl, V. [Institute of Plasma Physics, Academy of Sciences of the Czech Republic v.v.i., Prague (Czech Republic); Kovacic, J. [Jozef Stefan Institute, Ljubljana (Slovenia)

2014-04-15

The First derivative probe technique for a correct evaluation of the plasma potential in the case of non-Maxwellian EEDF is presented and used to process experimental data from COMPASS tokamak. Results obtained from classical and first derivative techniques are compared and discussed. The first derivative probe technique provides values for the plasma potential in the scrape-off layer of tokamak plasmas with an accuracy of about ±10%. Classical probe technique can provide values of the plasma potential only, if the electron and ion temperatures are known as well as the coefficient of secondary electron emission. (copyright 2014 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

Peeling-off of the external kink modes at tokamak plasma edge

International Nuclear Information System (INIS)

Zheng, L. J.; Furukawa, M.

2014-01-01

It is pointed out that there is a current jump between the edge plasma inside the last closed flux surface and the scrape-off layer and that the current jump can lead the external kink modes to convert to the tearing modes, due to the current interchange effects [L. J. Zheng and M. Furukawa, Phys. Plasmas 17, 052508 (2010)]. The magnetic reconnection in the presence of tearing modes subsequently causes the tokamak edge plasma to be peeled off to link to the divertors. In particular, the peeling or peeling-ballooning modes can become the “peeling-off” modes in this sense. This phenomenon indicates that the tokamak edge confinement can be worse than the expectation based on the conventional kink mode picture

Peeling-off of the external kink modes at tokamak plasma edge

Energy Technology Data Exchange (ETDEWEB)

Zheng, L. J. [Institute for Fusion Studies, University of Texas at Austin, Austin, Texas 78712 (United States); Furukawa, M. [Graduate School of Engineering, Tottori University, Tottori 680-8552 (Japan)

2014-08-15

It is pointed out that there is a current jump between the edge plasma inside the last closed flux surface and the scrape-off layer and that the current jump can lead the external kink modes to convert to the tearing modes, due to the current interchange effects [L. J. Zheng and M. Furukawa, Phys. Plasmas 17, 052508 (2010)]. The magnetic reconnection in the presence of tearing modes subsequently causes the tokamak edge plasma to be peeled off to link to the divertors. In particular, the peeling or peeling-ballooning modes can become the “peeling-off” modes in this sense. This phenomenon indicates that the tokamak edge confinement can be worse than the expectation based on the conventional kink mode picture.

Stochastic modelling of intermittent fluctuations in the scrape-off layer: Correlations, distributions, level crossings, and moment estimation

Energy Technology Data Exchange (ETDEWEB)

Garcia, O. E., E-mail: odd.erik.garcia@uit.no; Kube, R.; Theodorsen, A. [Department of Physics and Technology, UiT The Arctic University of Norway, N-9037 Tromsø (Norway); Pécseli, H. L. [Physics Department, University of Oslo, PO Box 1048 Blindern, N-0316 Oslo (Norway)

2016-05-15

A stochastic model is presented for intermittent fluctuations in the scrape-off layer of magnetically confined plasmas. The fluctuations in the plasma density are modeled by a super-position of uncorrelated pulses with fixed shape and duration, describing radial motion of blob-like structures. In the case of an exponential pulse shape and exponentially distributed pulse amplitudes, predictions are given for the lowest order moments, probability density function, auto-correlation function, level crossings, and average times for periods spent above and below a given threshold level. Also, the mean squared errors on estimators of sample mean and variance for realizations of the process by finite time series are obtained. These results are discussed in the context of single-point measurements of fluctuations in the scrape-off layer, broad density profiles, and implications for plasma–wall interactions due to the transient transport events in fusion grade plasmas. The results may also have wide applications for modelling fluctuations in other magnetized plasmas such as basic laboratory experiments and ionospheric irregularities.

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

DEFF Research Database (Denmark)

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

2013-01-01

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

Tokamak pump limiters

International Nuclear Information System (INIS)

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

1984-01-01

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

Viscosity in the edge of tokamak plasmas

International Nuclear Information System (INIS)

Stacey, W.M.

1993-05-01

A fluid representation of viscosity has been incorporated into a set of fluid equations that are maximally ordered in the ''short-radial-gradient-scale-length'' (srgsl) ordering that is appropriate for the edge of tokamak plasmas. The srgsl ordering raises viscous drifts and other viscous terms to leading order and fundamentally alters the character of the fluid equations. A leasing order viscous drift is identified. Viscous-driven radial particle and energy fluxes in the scrape-off layer and divertor channel are estimated to have an order unity effect in reducing radial peaking of energy fluxes transported along the field lines to divertor collector plates

Scrape-off layer properties of ITER-like limiter start-up plasmas in JET

Czech Academy of Sciences Publication Activity Database

Arnoux, G.; Farley, T.; Silva, C.; Devaux, S.; Firdaouss, M.; Frigione, D.; Goldston, R.J.; Gunn, J.; Horáček, Jan; Jachmich, S.; Lomas, P. J.; Marsen, S.; Matthews, G. F.; Pitts, R.A.; Stamp, M.; Stangeby, P.C.

2013-01-01

Roč. 53, č. 7 (2013), 073016-073016 ISSN 0029-5515 Institutional support: RVO:61389021 Keywords : Plasma-material interactions * boundary layer effect * power exhaust * divertors * electric and magnetic measurements * tokamaks * spherical tokamaks Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.243, year: 2013 http://iopscience.iop.org/0029-5515/53/7/073016/pdf/0029-5515_53_7_073016.pdf

Scrape-off layer reflectometer for Alcator C-Moda)

Science.gov (United States)

Lau, Cornwall; Hanson, Greg; Wilgen, John; Lin, Yijun; Wukitch, Steve

2010-10-01

A swept-frequency X-mode reflectometer is being built for Alcator C-Mod to measure the scrape-off layer density profiles at the top, middle, and bottom locations in front of both the new lower hybrid launcher and the new ion cyclotron range of frequencies antenna. The system is planned to operate between 100 and 146 GHz at sweep rates from 10 μs to 1 ms, and will cover a density range of approximately 1016-1020 m-3 at B0=5-5.4 T. To minimize the effects of density fluctuations, both differential phase and full phase reflectometry will be employed. Design, test data, and calibration results of this electronics system will be discussed. To reduce attenuation losses, tallguide (TE01) will be used for most of the transmission line system. Simulations of high mode conversion in tallguide components, such as e-plane hyperbolic secant radius of curvature bends, tapers, and horn antennas will be shown. Experimental measurements of the total attenuation losses of these components in the lower hybrid waveguide run will also be presented.

Limiter effects on scrape-off layer fluctuations and transport

International Nuclear Information System (INIS)

Thayer, D.R.; Diamond, P.H.; Wootton, A.J.

1987-01-01

Edge turbulence experiments indicate that radial particle flux increases as a function of radius up to the scrape-off layer (SOL), and that the Boltzman relation is violated. Resistivity gradient driven turbulence (RGDT) theory has been shown to track the radial dependence of the particle flux in the plasma edge closer than dissipative density gradient driven turbulence (DDGDT) theory. Also, the Boltzman relation is not invoked for RGDT while it is usually assumed for DDGDT. Consequently, RGDT is a more likely candidate for an edge turbulence model. However, Langmuir probe experiments indicate that the particle flux is reduced by as much as 50% in the SOL. Thus, since basic turbulence theories do not account for limiter effects, the primary focus of this study is to include such effects in a RGDT theory of the SOL. We present an analysis of SOL fluctuations using a rippling mode or RGDT calculation which incorporates the essential limiter boundary condition.(orig./GG)

Transport code calculations concerning the plasma parameters in the scrape-off layer of poloidal limiter and the possible advantage of high Z wall materials in the cool plasma blanket approach

International Nuclear Information System (INIS)

Nicolai, A.; Fuchs, G.

1978-01-01

The plasma parameters in the scrape-off layer (SoL) of a tokamak are calculated by introducing appropriate particles and energy loss terms into the six-regime version of Duechs code. These terms take secondary electron smission from the limiter surface and a potential sheath in front of it into account. In the SoL Bohn diffusion is assumed. Limiter materials with large secondary emission coefficients (SEC)(e.g. Mo) give lower potential steps (U = 90 V) than low SEC materials (e.g. Be) which cause (U = 250 V). The flux of the sputtered liner material and the resulting radiation losses can be decreased by neutral gas influx. When the same neutral gas influx and the same additional heating are used, it is found that radiation losses due to molybdenum are lower than those due to iron, although Mo is more toxic. (Auth.)

Electromagnetic ELM and inter-ELM filaments detected in the COMPASS Scrape-Off Layer.

Czech Academy of Sciences Publication Activity Database

Spolaore, M.; Kovařík, Karel; Stöckel, Jan; Adámek, Jiří; Dejarnac, Renaud; Ďuran, Ivan; Komm, Michael; Markovič, Tomáš; Martines, E.; Pánek, Radomír; Seidl, Jakub; Vianello, N.

2017-01-01

Roč. 12, August (2017), s. 844-851 ISSN 2352-1791. [PSI 2016 - 22nd International Conference on Plasma Surface Interactions in Controlled Fusion Devices/22./. Roma, 30.05.2016-03.06.2016] R&D Projects: GA MŠk(CZ) 8D15001; GA MŠk(CZ) LM2015045 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : Electromagnetic filaments * ELMs * Scrape-Off Layer * Magnetic fluctuations * Current density Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) http://www.sciencedirect.com/science/article/pii/S2352179116301934

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

OpenAIRE

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

2017-01-01

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

Spatially resolved characterization of electrostatic fluctuations in the scrape-off layer of the CASTOR tokamak

Czech Academy of Sciences Publication Activity Database

Devynck, P.; Bonhomme, G.; Martines, E.; Stöckel, Jan; Van Oost, G.; Voitsekhovitch, I.; Adámek, Jiří; Azeroual, A.; Doveil, F.; Ďuran, Ivan; Gravier, E.; Gunn, J.; Hron, Martin

2005-01-01

Roč. 47, č. 2 (2005), s. 269-280 ISSN 0741-3335 R&D Projects: GA ČR GA202/03/0786 Grant - others:GA - INTAS 2001 2056 Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * plasma * turbulence Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.902, year: 2005

Calculation of poloidal rotation in the edge plasma of limiter tokamaks

International Nuclear Information System (INIS)

Gerhauser, H.; Claassen, H.A.

1987-05-01

The existing 2-d two-fluid code for computing the plasma profiles in the scrape-off layer of limiter tokamaks has been further developed to include the effect of poloidal rotation in the basic equations. This rotation is produced by radial electric fields which arise in the limiter shadow due to radial gradients in the Langmuir sheath potential in front of the limiter. As a consequence slight deviations from ambipolar motion must occur. A strong increase of rotation near the separatrix is connected with an electric current circuit closed via the limiter edge. The 2-d profiles of all relevant quantities are calculated and discussed for TEXTOR-typical parameters including also the effect of limiter recycled neutrals. The results agree well with the known experimental evidence on poloidal rotation and should be transferable to all limiter tokamaks. (orig.)

Broadband magnetic and density fluctuations in the TCA tokamak

International Nuclear Information System (INIS)

Hollenstein, Ch.; Keller, R.; Pochelon, A.; Ryter, F.; Sawley, M.L.; Simm, W.; Weisen, H.

1987-01-01

The results of comparative studies of broadband magnetic and density fluctuations during ohmic discharges in the TCA tokamak are described. Long coherence lengths are observed in poloidal and toroidal directions between magnetic probes in the scrape-off layer. A phase contrast diagnostic provides a newly accessible range of density fluctuations in the bulk plasma with very long wavelengths. Langmuir probes provide similar measurements in the scrape-off layer. Statistical dispersion relations for both density and magnetic fluctuations are deduced and are shown to be substantially different. Low mean poloidal wavenumbers (m ∼ 2 at 100 kHz) are obtained for the magnetic fluctuations, in contrast to the much higher values measured for density fluctuations. The difference between magnetic and density fluctuations is also reflected in different scalings with plasma parameters and with electron confinement time. The helicity of the coherent magnetic structures is analyzed to show that interior regions of the plasma, such as the q = 2 region contribute to the magnetic activity at the edge. This explains why the magnetic fluctuations measured at the edge are likely to reflect the confinement properties of the bulk plasma. The results of detailed probe rotation experiments and coherence measurements give indications of the physical nature and origin of magnetic fluctuations

Scrape-off layer-induced beam density fluctuations and their effect on beam emission spectroscopy

Science.gov (United States)

Moulton, D.; Marandet, Y.; Tamain, P.; Dif-Pradalier, G.

2015-07-01

A statistical model is presented to calculate the magnitude of beam density fluctuations generated by a turbulent scrape-off layer (SOL). It is shown that the SOL can induce neutral beam density fluctuations of a similar magnitude to the plasma density fluctuations in the core, potentially corrupting beam emission spectroscopy measurements. The degree of corruption is quantified by combining simulations of beam and plasma density fluctuations inside a simulated measurement window. A change in pitch angle from the separatrix to the measurement window is found to reduce the effect of beam fluctuations, whose largest effect is to significantly reduce the measured correlation time.

Propagation of Blob in boundary of HT-7 tokamak

International Nuclear Information System (INIS)

Yan Ning; Zhang Wei; Chang Jiafeng; Ming Tingfeng; Ding Siye

2011-01-01

Intermittent characteristics of turbulence induced by coherent structures (Blob) are manifested clearly on the Langmuir probe signal of HT-7 tokamak. With conditional analysis, asymmetric characteristics of the intermittent bursts are demonstrated. The parameter of plasma inside the Blob is larger than the background plasma parameter. Due to the radial propagation of the coherent structures, the particle density and temperature profiles in the scrape-off layer (SOL) are non-exponential and flat away from the last closed flux surface (LCFS). Around the LCFS, large burst fluctuations are responsible for about 50% of the total transport. These experimental finds may imply that the coherent structures are distorted by the developed shear flow in the shear layer. In SOL region, the coherent structures propagate in the direction of ion diamagnetic drift. (authors)

Tokamak plasma boundary layer model

International Nuclear Information System (INIS)

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

1983-01-01

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

Electron temperature gradient driven instability in the tokamak boundary plasma

International Nuclear Information System (INIS)

Xu, X.Q.; Rosenbluth, M.N.; Diamond, P.H.

1992-01-01

A general method is developed for calculating boundary plasma fluctuations across a magnetic separatrix in a tokamak with a divertor or a limiter. The slab model, which assumes a periodic plasma in the edge reaching the divertor or limiter plate in the scrape-off layer(SOL), should provide a good estimate, if the radial extent of the fluctuation quantities across the separatrix to the edge is small compared to that given by finite particle banana orbit. The Laplace transform is used for solving the initial value problem. The electron temperature gradient(ETG) driven instability is found to grow like t -1/2 e γmt

Investigation of radial propagation of electrostatic fluctuations in the IR-T1 tokamak plasma edge

Energy Technology Data Exchange (ETDEWEB)

Shariatzadeh, R; Ghoranneviss, M; Salem, M K [Plasma Physics Research Center, Science and Research Branch, Islamic Azad University (IAU), PO Box 14665-678, Tehran (Iran, Islamic Republic of); Emami, M, E-mail: rezashariatzadeh@gmail.com [Laser and Optics Research School, NSTRI, AEOI, PO Box 14155-1339, Tehran (Iran, Islamic Republic of)

2011-01-15

The radial propagation of electrostatic fluctuation is considered extremely important for understanding cross-field anomalous transport. In this paper, two arrays of Langmuir probes are used to analyze electrostatic fluctuations in the edge of IR-T1 tokamak plasma in both the radial and the poloidal directions. The propagation characteristics of the floating potential fluctuations are analyzed by the two-point correlation technique. The wavenumber spectrum shows that there is a net radially outward propagation of turbulent fluctuations in the edge and scrape-off layer (SOL) regions. Hence, edge turbulence presumably originates from core fluctuations.

Investigation of radial propagation of electrostatic fluctuations in the IR-T1 tokamak plasma edge

International Nuclear Information System (INIS)

Shariatzadeh, R; Ghoranneviss, M; Salem, M K; Emami, M

2011-01-01

The radial propagation of electrostatic fluctuation is considered extremely important for understanding cross-field anomalous transport. In this paper, two arrays of Langmuir probes are used to analyze electrostatic fluctuations in the edge of IR-T1 tokamak plasma in both the radial and the poloidal directions. The propagation characteristics of the floating potential fluctuations are analyzed by the two-point correlation technique. The wavenumber spectrum shows that there is a net radially outward propagation of turbulent fluctuations in the edge and scrape-off layer (SOL) regions. Hence, edge turbulence presumably originates from core fluctuations.

Parametric study of biased scrape-off layer

International Nuclear Information System (INIS)

Parbhakar, K.

1992-01-01

The particle and energy balance equations in the scrape-off layer (SOL) plasma are solved analytically in a rectangular domain. The parallel flow is supposed to be convective, whereas the cross field flow is assumed diffusive, with modifications due to a radial E field. When the energy balance equation is neglected we get a quadratic equation for Λ, the e-folding density scale length. The normalized e-folding scale length Λ/Λ 0 (Λ 0 = Λ for E = 0) is a function of single parameter α (= eEλ 0 /2kT, where T is the plasma temperature, e the elementary charge and k is Boltzmann's constant), and monotonically increases with α. Taking the energy balance equation into account and neglecting R i (the ionization rate), the quadratic equation is further simplified, and Λ/Λ 0 again depends on a single parameter, determined by Q (the input power flux), D (the cross field diffusion coefficient), Λ (the sheath transmission coefficient for energy) and n p (the plate density). Now Λ/Λ 0 decreases with Q and increases linearly with D, γ and n p . The variation of Λ/Λ 0 with E is estimated for TdeV and ITER. We find that, by a proper choice of parameters, biasing may be a very valuable tool to modify the SOL properties at modest E fields, provided the heat load on the divertor plate is not excessively large

Properties of the tokamak edge plasma

International Nuclear Information System (INIS)

Wolff, H.

1988-01-01

A short review of some features of the edge plasma in limiter tokamaks is given. The limits of the simple one-dimensional scrape-off layer (SOL) model and the relation between the core plasma are discussed. Multifaceted asymmetric radiation from the edge (MARFE) phenomena and detached plasma are closely connected with the particle and energy balance of the SOL. Their occurrence is based on the relation of plasma parameters of the edge plasma to those of the core. Important problems of plasma wall interactions are the detection of the impurity sources and sinks and the study of the impurity transport and shielding. The non-uniform character of plasma wall interactions and their dependence on the discharge performance still renders difficult any theoretical forecast of impurity distribution and transport and calls for better diagnostics. (author)

Monte-Carlo Impurity transport simulations in the edge of the DIII-D tokamak using the MCI code

International Nuclear Information System (INIS)

Evans, T.E.; Mahdavi, M.A.; Sager, G.T.; West, W.P.; Fenstermacher, M.E.; Meyer, W.H.; Porter, G.D.

1995-07-01

A Monte-Carlo Impurity (MCI) transport code is used to follow trace impurities through multiple ionization states in realistic 2-D tokamak geometries. The MCI code is used to study impurity transport along the open magnetic field lines of the Scrape-off Layer (SOL) and to understand how impurities get into the core from the SOL. An MCI study concentrating on the entrainment of carbon impurities ions by deuterium background plasma into the DIII-D divertor is discussed. MCI simulation results are compared to experimental DIII-D carbon measurements

Monte-Carlo Impurity transport simulations in the edge of the DIII-D tokamak using the MCI code

International Nuclear Information System (INIS)

Evans, T.E.; Sager, G.T.; Mahdavi, M.A.; Porter, G.D.; Fenstermacher, M.E.; Meyer, W.H.

1995-01-01

A Monte-Carlo Impurity (MCI) transport code is used to follow trace impurities through multiple ionization states in realistic 2-D tokamak geometries. The MCI code is used to study impurity transport along the open magnetic field lines of the Scrape-off Layer (SOL) and to understand how impurities get into the core from the SOL. An MCI study concentrating on the entrainment of carbon impurities ions by deuterium background plasma into the DII-D divertor is discussed. MCI simulation results are compared to experimental DII-D carbon measurements. 2 refs

Scrape-off layer based modelling of the density limit in beryllated JET limiter discharges

International Nuclear Information System (INIS)

Borrass, K.; Campbell, D.J.; Clement, S.; Vlases, G.C.

1993-01-01

The paper gives a scrape-off layer based interpretation of the density limit in beryllated JET limiter discharges. In these discharges, JET edge parameters show a complicated time evolution as the density limit is approached and the limit is manifested as a non-disruptive density maximum which cannot be exceeded by enhanced gas puffing. The occurrence of Marfes, the manner of density control and details of recycling are essential elements of the interpretation. Scalings for the maximum density are given and compared with JET data. The relation to disruptive density limits, previously observed in JET carbon limiter discharges, and to density limits in divertor discharges is discussed. (author). 18 refs, 10 figs, 1 tab

Parametric analysis of the thermal effects on the divertor in tokamaks during plasma disruptions

International Nuclear Information System (INIS)

Bruhn, M.L.

1988-04-01

Plasma disruptions are an ever present danger to the plasma-facing components in today's tokamak fusion reactors. This threat results from our lack of understanding and limited ability to control this complex phenomenon. In particular, severe energy deposition occurs on the divertor component of the double-null configured tokamak reactor during such disruptions. A hybrid computational model developed to estimate and graphically illustrate global thermal effects of disruptions on the divertor plates is described in detail. The quasi-two-dimensional computer code, TADDPAK (Thermal Analysis Divertor during Disruptions PAcKage), is used to conduct parametric analysis for the TIBER II Tokamak Engineering Test Reactor Design. The dependence of these thermal effects on divertor material choice, disruption pulse length, disruption pulse shape, and the characteristic thickness of the plasma scrape-off layer is investigated for this reactor design. Results and conclusions from this analysis are presented. Improvements to this model and issues that require further investigation are discussed. Cursory analysis for ITER (International Thermonuclear Experimental Reactor) is also presented in the appendix. 75 refs., 49 figs., 10 tabs

Far-reaching Impact of Intermittent Transport across the Scrape-off Layer: Latest Results from ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Kocan, M.; Muller, W.; Conway, G.; De Marne, P.; Eich, T.; Fischer, R.; Fuchs, C.; Herrmann, A.; Ionita, C.; Kallenbach, A.; Lunt, T.; Maraschek, M.; Muller, S.; Nold, B.; Ribeiro, T.; Rohde, V.; Scott, B.; Stroth, U.; Suttrop, W.; Wolfrum, E., E-mail: martin.kocan@ipp.mpg.de [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Garching (Germany); Adamek, J.; Horacek, J.; Komm, M. [Association EURATOM-IPP CR, Prague (Czech Republic); Gennrich, F.; Maszl, C.; Mehlmann, F.; Schrittwieser, R. [Institute for Ion Physics and Applied Physics, Association Euratom-OAW (Austria); Huang, Z. [Institut fuer Plasmaforschung, Universitat Stuttgart, Stuttgart (Germany)

2012-09-15

Full text: Latest research of intermittent transport in the scrape-off layer (SOL) of the ASDEX Upgrade tokamak is presented. Near the separatrix the fluctuations of the plasma and the floating potentials, measured by various Langmuir probes (LPs), are found to be anti-correlated due to fluctuations of the electron temperature. This indicates that, in contrast to a widely used experimental practice, a free exchange of both potentials is unjustified and can lead to significant error. Measurements of ion energies in turbulent L-mode and ELM filaments were carried out using a retarding field analyzer. In L-mode plasma, the filament ion temperature measured at 2 cm outside the separatrix is 80 - 110 eV, i.e., 3 - 4x the background ion temperature. Turbulent filaments also convect plasma to the wall with larger density than the background plasma density. Qualitatively similar observations were obtained during inter-ELM periods. Such enhanced particle and energy fluxes can potentially raise the erosion of the first wall in ITER. The ion temperature averaged over an ELM measured 35 - 60 mm outside the separatrix is in the range of 20 - 200 eV (5 - 50% of the pedestal top ion temperature). This demonstrates that ELM filaments carry hot ions over large radial distances in the SOL, which, in turn, can lead to enhanced sputtering from the first wall in future tokamaks. Lowest ion energies are observed during ELMs mitigated by in-vessel magnetic perturbations (MPs). The ELM ion temperature in the far SOL is found to increase with the ELM energy, indicating that on average the filaments in large ELMs propagate faster radially. The filamentary structure of the ion current density measured by LPs at the outboard mid-plane during mitigated ELMs is qualitatively similar to that observed during type I ELMs. The amplitude of the ion current density decreases only slightly when ELMs are mitigated, but, in contrast to type I ELMs, bursts of the ion current are observed throughout the

Modification of boundary fluctuations by LHCD in the HT-7 tokamak

International Nuclear Information System (INIS)

Song Mei; Wan Baonian; Xu Guosheng; Ling Bili

2003-01-01

Measurements of boundary fluctuations and fluctuation driven electron fluxes have been performed in ohmic and lower hybrid current drive enhanced confinement plasma using a graphite Langmuir probe array on HT-7 tokamak. The fluctuations are significantly suppressed and the turbulent fluxes are remarkably depressed in the enhanced plasma. We characterized the statistical properties of fluctuations and the particle flux and found a non-Gaussian character in the whole scrape-off layer with minimum deviations from Gaussian in the proximity of the velocity shear layer in ohmic plasma. In the enhanced plasma the deviations in the boundary region are all reduces obviously. The fluctuations and induced electron fluxes show sporadic bursts asymmetric in time and the asymmetry is remarkably weakened in the lower hybrid current driving (LHCD) phase. The results suggest a coupling between the statistical behaviour of fluctuations and the turbulent flow

Measurements of RF-induced sol modifications in Tore Supra tokamak

International Nuclear Information System (INIS)

Kubic, Martin; Gunn, James P.; Colas, Laurent; Heuraux, Stephane; Faudot, Eric

2012-01-01

Since spring 2011, one of the three ion cyclotron resonance heating (ICRH) antennas in the Tore Supra (TS) tokamak is equipped with a new type of Faraday screen (FS). Results from Radio Frequency (RF) simulations of the new Faraday screen suggest the innovative structure with cantilevered bars and 'shark tooth' openings significantly changes the current flow pattern on the front of the antenna which in turn reduces the RF potential and RF electrical field in particular parallel to the magnetic field lines which contributes to generating RF sheaths. Effects of the new FS operation on RF-induced scrape-off layer (SOL) modifications are studied for different plasma and antenna configurations - scans of strap power ratio imbalance, phasing, injected power and SOL density. (authors)

Effects of radial electrical field on neoclassical transport in tokamaks

International Nuclear Information System (INIS)

Wang Zhongtian; Le Clair, G.

1996-07-01

Neoclassical transport theory for tokamaks in presence of a radial electrical field with shear is developed using Hamiltonian formalism. Diffusion coefficients are derived in both plateau regime including a large electric field and banana regime including the squeezing factor which can greatly affect diffusion at the plasma edge. The scaling on squeezing factor is different from the one given by Shaing and Hazeltine. Rotation speeds are calculated in the scrape-off region. They are in good agreement with measurements on TdeV Tokamak. (2 figs.)

Interplay between edge and outer core fluctuations in the tokamak Tore Supra

International Nuclear Information System (INIS)

Fenzi, C.; Garbet, X.; Capes, H.; Devynck, P.; Laviron, C.; Truc, A.; Gervais, F.; Hennequin, P.; Quemeneur, A.

2000-01-01

In the tokamak Tore Supra, when a poloidally and toroidally localized limiter, called a modular limiter, is introduced into the lower part of the scrape-off layer, density fluctuations located in the vicinity of this limiter present a specific feature with the appearance of a new spectral pattern in the associated frequency spectrum. This leads to a strong up-down asymmetry observed in both the plasma edge and the plasma outer core, with a maximum level of turbulence at the bottom of the plasma. The observed asymmetry characteristics show that magnetic connection lengths play a critical role here and that the limiter configuration has some effect on the outer core turbulence. (author)

Probability distribution functions for intermittent scrape-off layer plasma fluctuations

Science.gov (United States)

Theodorsen, A.; Garcia, O. E.

2018-03-01

A stochastic model for intermittent fluctuations in the scrape-off layer of magnetically confined plasmas has been constructed based on a super-position of uncorrelated pulses arriving according to a Poisson process. In the most common applications of the model, the pulse amplitudes are assumed exponentially distributed, supported by conditional averaging of large-amplitude fluctuations in experimental measurement data. This basic assumption has two potential limitations. First, statistical analysis of measurement data using conditional averaging only reveals the tail of the amplitude distribution to be exponentially distributed. Second, exponentially distributed amplitudes leads to a positive definite signal which cannot capture fluctuations in for example electric potential and radial velocity. Assuming pulse amplitudes which are not positive definite often make finding a closed form for the probability density function (PDF) difficult, even if the characteristic function remains relatively simple. Thus estimating model parameters requires an approach based on the characteristic function, not the PDF. In this contribution, the effect of changing the amplitude distribution on the moments, PDF and characteristic function of the process is investigated and a parameter estimation method using the empirical characteristic function is presented and tested on synthetically generated data. This proves valuable for describing intermittent fluctuations of all plasma parameters in the boundary region of magnetized plasmas.

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

Science.gov (United States)

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

2018-03-01

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

Exhaust, ELM and Halo physics using the MAST tokamak

International Nuclear Information System (INIS)

Counsell, G.F.; Ahn, J-W.; Kirk, A.; Helander, P.; Martin, R.; Tabasso, A.; Wilson, H.R.; Cohen, R.H.; Ryutov, D.D.; Yang, Y.

2003-01-01

The scrape-off layer (Sol) and divertor target plasma of a large spherical tokamak (ST) is characterised in detail for the first time. Scalings for the SOL heat flux width in MAST are developed and compared to conventional tokamaks. Modelling reveals the significance of the mirror force to the ST SOL. Core energy losses, including during ELMs, in MAST arrive predominantly (>80%) to the outboard targets in all geometries. Convective transport dominates energy losses during ELMs and MHD analysis suggests ELMs in MAST are Type III even at auxiliary heating powers well above the L-H threshold. ELMs are associated with a poloidally and/or toroidally localised radial efflux at ∼1 km/s well into the far SOL but not with any broadening of the target heat flux width. Toroidally asymmetric divertor biasing experiments have been conducted in an attempt to broaden the target heat flux width, with promising results. During vertical displacement events, the maximum product of the toroidal peaking factor and halo current fraction remains below 0.3, around half the ITER design limit. Evidence is presented that the resistance of the halo current path may have an impact on the distribution of halo current. (author)

Heuristic Drift-based Model of the Power Scrape-off width in H-mode Tokamaks

International Nuclear Information System (INIS)

Goldston, Robert J.

2011-01-01

An heuristic model for the plasma scrape-off width in H-mode plasmas is introduced. Grad B and curv B drifts into the SOL are balanced against sonic parallel flows out of the SOL, to the divertor plates. The overall particle flow pattern posited is a modification for open field lines of Pfirsch-Shlueter flows to include sinks to the divertors. These assumptions result in an estimated SOL width of ∼ 2αρ p /R. They also result in a first-principles calculation of the particle confinement time of H-mode plasmas, qualitatively consistent with experimental observations. It is next assumed that anomalous perpendicular electron thermal diffusivity is the dominant source of heat flux across the separatrix, investing the SOL width, defined above, with heat from the main plasma. The separatrix temperature is calculated based on a two-point model balancing power input to the SOL with Spitzer-Haerm parallel thermal conduction losses to the divertor. This results in a heuristic closed-form prediction for the power scrape-off width that is in reasonable quantitative agreement both in absolute magnitude and in scaling with recent experimental data from deuterium plasmas. Further work should include full numerical calculations, including all magnetic and electric drifts, as well as more thorough comparison with experimental data.

Magnetic topology changes induced by lower hybrid waves and their profound effect on edge-localized modes in the EAST tokamak.

Science.gov (United States)

Liang, Y; Gong, X Z; Gan, K F; Gauthier, E; Wang, L; Rack, M; Wang, Y M; Zeng, L; Denner, P; Wingen, A; Lv, B; Ding, B J; Chen, R; Hu, L Q; Hu, J S; Liu, F K; Jie, Y X; Pearson, J; Qian, J P; Shan, J F; Shen, B; Shi, T H; Sun, Y; Wang, F D; Wang, H Q; Wang, M; Wu, Z W; Zhang, S B; Zhang, T; Zhang, X J; Yan, N; Xu, G S; Guo, H Y; Wan, B N; Li, J G

2013-06-07

Strong mitigation of edge-localized modes has been observed on Experimental Advanced Superconducting Tokamak, when lower hybrid waves (LHWs) are applied to H-mode plasmas with ion cyclotron resonant heating. This has been demonstrated to be due to the formation of helical current filaments flowing along field lines in the scrape-off layer induced by LHW. This leads to the splitting of the outer divertor strike points during LHWs similar to previous observations with resonant magnetic perturbations. The change in the magnetic topology has been qualitatively modeled by considering helical current filaments in a field-line-tracing code.

Ultrafast two-dimensional lithium beam emission spectroscopy diagnostic on the EAST tokamak

Science.gov (United States)

Zoletnik, S.; Hu, G. H.; Tál, B.; Dunai, D.; Anda, G.; Asztalos, O.; Pokol, G. I.; Kálvin, S.; Németh, J.; Krizsanóczi, T.

2018-06-01

A diagnostic instrument is described for the Experimental Advanced Superconducting Tokamak (EAST) for the measurement of the edge plasma electron density profile and plasma turbulence properties. An accelerated neutral lithium beam is injected into the tokamak and the Doppler shifted 670.8 nm light emission of the Li2p-2s transition is detected. A novel compact setup is used, where the beam injection and observation take place from the same equatorial diagnostic port and radial-poloidal resolution is achieved with microsecond time resolution. The observation direction is optimized in order to achieve a sufficient Doppler shift of the beam light to be able to separate from the strong edge lithium line emission on this lithium coated device. A 250 kHz beam chopping technique is also demonstrated for the removal of background light. First results show the capability of measuring turbulence and its poloidal flow velocity in the scrape-off layer and edge region and the resolution of details of transient phenomena like edge localized modes with few microsecond time resolution.

Influence of plasma background including neutrals on scrape-off layer filaments using 3D simulations

Directory of Open Access Journals (Sweden)

D. Schwörer

2017-08-01

Full Text Available This paper investigates the effect of the plasma background, including neutrals in a self-consistent way, on filaments in the scrape-off layer (SOL of fusion devices. A strong dependency of filament motion on background density and temperature is observed. The radial filament motion shows an increase in velocity with decreasing background density and increasing background temperature. In the simulations presented here, three neutral-filament interaction models have been compared, one with a static neutral background, one with no interaction between filaments and neutrals, and one co-evolving the neutrals self consistently with the filaments. With the background conditions employed here, which do not show detachment, there are no significant effects of neutrals on filaments, as by the time the filament reaches maximum velocity, the neutral density has not changed significantly.

Turbulence and transport with spatial-temporal biasing on the scrape-off layer on CASTOR tokamak

International Nuclear Information System (INIS)

Stoeckel, J.

2002-01-01

Experiments with the poloidal ring of 32 plane electrodes were performed on the CASTOR tokamak (R=0.4 m, a=0.06 m, B=1 T) to measure, for the first time, the complete poloidal structure of the electrostatic edge turbulence. In addition, the possibility of active modification of the edge turbulence was checked. The main results are as follows: Quite regular turbulent structures with the pronounced poloidal periodicity are observed by passive measuring signals of the individual electrodes. The dominant poloidal mode number, m=6-8, is approximately of the same value as the edge safety factor. Propagating waves of potential (f=10-40 kHz) with the wave numbers in the range of m=2-8, applied to the ring of the electrodes, modify the edge turbulence significantly due to their interaction with turbulent structures. (author)

Steady-state and time-dependent modelling of parallel transport in the scrape-off layer

DEFF Research Database (Denmark)

Havlickova, E.; Fundamenski, W.; Naulin, Volker

2011-01-01

The one-dimensional fluid code SOLF1D has been used for modelling of plasma transport in the scrape-off layer (SOL) along magnetic field lines, both in steady state and under transient conditions that arise due to plasma turbulence. The presented work summarizes results of SOLF1D with attention...... given to transient parallel transport which reveals two distinct time scales due to the transport mechanisms of convection and diffusion. Time-dependent modelling combined with the effect of ballooning shows propagation of particles along the magnetic field line with Mach number up to M ≈ 1...... temperature calculated in SOLF1D is compared with the approximative model used in the turbulence code ESEL both for steady-state and turbulent SOL. Dynamics of the parallel transport are investigated for a simple transient event simulating the propagation of particles and energy to the targets from a blob...

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

Science.gov (United States)

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

2007-06-01

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

Computational studies in tokamak equilibrium and transport

International Nuclear Information System (INIS)

Braams, B.J.

1986-01-01

This thesis is concerned with some problems arising in the magnetic confinement approach to controlled thermonuclear fusion. The work address the numerical modelling of equilibrium and transport properties of a confined plasma and the interpretation of experimental data. The thesis is divided in two parts. Part 1 is devoted to some aspects of the MHD equilibrium problem, both in the 'direct' formulation (given an equation for the plasma current, the corresponding equilibrium is to be determined) and in the 'inverse' formulation (the interpretation of measurements at the plasma edge). Part 2 is devoted to numerical studies of the edge plasma. The appropriate Navier-Stokes system of fluid equations is solved in a two-dimensional geometry. The main interest of this work is to develop an understanding of particle and energy transport in the scrape-off layer and onto material boundaries, and also to contribute to the conceptual design of the NET/INTOR tokamak reactor experiment. (Auth.)

Density profiles and particle fluxes of heavy impurities in the limiter shadow region of a tokamak

International Nuclear Information System (INIS)

Claassen, H.A.; Repp, H.

1980-01-01

For the case of low impurity concentration, transport calculations have been performed for heavy impurities, in the scrape-off layer plasma of a tokamak with a poloidal ring limiter. The theory is based on the drift-kinetic equations for the various ionization states of the impurity ions taking due consideration of the convection and collision processes. The background plasma and the impurity sources from the torus wall and the limiter surface enter the theory as input parameters. The theory is developed for the first two orders of the drift approximation. Numerical results are given to zero order drift approximation for the radial profiles of density and particle fluxes parallel to the magnetic field. (orig.)

Analytical studies of multidimensional plasma transport in the scrape-off layer

International Nuclear Information System (INIS)

Tendler, M.; Rozhansky, V.

1992-01-01

2-D effects originating from the lack of geometrical overlapping of equipotential and magnetic surfaces are examined. It is shown that poloidal rotation in the scrape-off layer (SOL), which emerges due to the drift caused by the radial electric field, shifts the maximum of the plasma density in the SOL out of the equatorial midplane provided it is accompanied by a poloidally asymmetric turbulent diffusion. Another effect resulting in asymmetry relates the significant enhancement of the plasma flow to the ion side of a limiter (and the corresponding reduction to the electron side) with the momentum balance in the SOL. A rigorous mathematical theory has been developed in order to assess the 3-D effect quantitatively. In the SOL global ambipolarity has to be maintained. Therefore flux surface average radial current cancels at the separatrix. Employing the toroidal and the parallel components of momentum balance, it is shown that the ambipolarity constraint on open field lines yields the ion sonic flow driven in the same direction as the inductive current. This results in the reduction to almost zero of a flow toward the electron side of a limiter and the almost two-fold increase of a flow toward the ion side. (orig.)

An Heuristic Drift-Based Model of the Power Scrape-Off Width in H-Mode Tokamaks

International Nuclear Information System (INIS)

Goldston, Robert J.

2011-01-01

An heuristic model for the plasma scrape-off width in H-mode plasmas is introduced. Grad B and curv B drifts into the SOL are balanced against sonic parallel flows out of the SOL, to the divertor plates. The overall mass flow pattern posited is a modification for open field lines of Pfirsch-Shlueter flows to include sinks to the divertors. These assumptions result in an estimated SOL width of 2αρ p /R. They also result in a first-principles calculation of the particle confinement time of H-mode plasmas, qualitatively consistent with experimental observations. It is next assumed that anomalous perpendicular electron thermal diffusivity is the dominant source of heat flux across the separatrix, investing the SOL width, defined above, with heat from the main plasma. The separatrix temperature is calculated based on a two-point model balancing power input to the SOL with Spitzer-Haerm parallel thermal conduction losses to the divertor. This results in an heuristic closed-form prediction for the power scrape-off width that is in remarkable quantitative agreement both in absolute magnitude and in scaling with recent experimental data. Further work should include full numerical calculations, including all magnetic and electric drifts, as well as more thorough comparison with experimental data.

Tokamak advanced pump limiter experiments and analysis

International Nuclear Information System (INIS)

Conn, R.W.

1983-06-01

Experiments with pump limiter modules on several operating tokamaks establish such limiters as efficient collectors of particles and has demonstrated the importance of ballistic scattering as predicted theoretically. Plasma interaction with recycling neutral gas appears to become important as the plasma density increases and the effective ionization mean free path within the module decreases. In limiters with particle collection but without active internal pumping, the neutral gas pressure is found to vary nonlinearly with the edge plasma density at the highest densities studies. Both experiments and theory indicate that the energy spectrum of gas atoms in the pump ducting is non-thermal, consistent with the results of Monte Carlo neutral atom transport calculations. The distribution of plasma power over the front surface of such modules has been measured and appears to be consistent with the predictions of simple theory. Initial results from the latest experiment on the ISX-B tokamak with an actively pumped limiter module demonstrates that the core plasma density can be controlled with a pump limiter and that the scrape-off layer plasma can partially screen the core plasma from gas injection. The results from module pump limiter experiments and from the theory and design analysis of advanced pump limiters for reactors are used to suggest the major features of a definitive, axisymmetric, toroidal belt pump limiter experiment

An advanced plasma control system for the DIII-D tokamak

International Nuclear Information System (INIS)

Ferron, J.R.; Kellman, A.; McKee, E.; Osborne, T.; Petrach, P.; Taylor, T.S.; Wight, J.; Lazarus, E.

1991-11-01

An advanced plasma control system is being implemented for the DIII-D tokamak utilizing digital technology. This system will regulate the position and shape of tokamak discharges that range from elongated limiter to single-null divertor and double-null divertor with elongation as high as 2.6. Development of this system is expected to lead to control system technology appropriate for use on future tokamaks such as ITER and BPX. The digital system will allow for increased precision in shape control through real time adjustment of the control algorithm to changes in the shape and discharge parameters such as β p , ell i and scrape-off layer current. The system will be used for research on real time optimization of discharge performance for disruption avoidance, current and pressure profile control, optimization of rf antenna loading, or feedback on heat deposition patterns through divertor strike point position control, for example. Shape control with this system is based on linearization near a target shape of the controlled parameters as a function of the magnetic diagnostic signals. This digital system is unique in that it is designed to have the speed necessary to control the unstable vertical motion of highly elongated tokamak discharges such as those produced in DIII-D and planned for BPX and ITER. a 40 MHz Intel i860 processor is interfaced to up to 112 channels of analog input signals. The commands to the poloidal field coils can be updated at 80 μs intervals for the control of vertical position with a delay between sampling of the analog signal and update of the command of less than 80 μs

LIBS: study of elemental profile of different layers of the optical window of Tokamak

International Nuclear Information System (INIS)

Maurya, Gulab Singh; Jyotsna, Aradhana; Rai, Awadhesh Kumar; Ajai Kumar

2012-01-01

In the Tokamak, during confinement of plasma, impurities are deposited on optical window, mirror, limiters, etc. of the tokamak. Thus a layer of impurity on the surface of the optical window causes less visibility which creates problem in the study of plasma parameters and other diagnostics of the plasma generated in the tokamak. Laser Induced Breakdown Spectroscopy (LIBS) is a useful atomic spectroscopic technique for analysis of materials in any phase (Solid, Liquid, Gas etc). LIBS spectra of optical window have been recorded in the spectral range of 200-500 nm. In present study we have focused laser on the surface of the window, to study the layer-wise elemental profile of optical window, we have recorded the LIBS spectra with increasing number of laser shots on the same point of the window. In first laser shot, spectral signature of Cr, Fe, and Ni etc. are present in the LIBS spectra, which is related with the impurity but after five to six laser shots at the same point of the optical window spectral signature Si, B are observed which is related to the glass material. Thus our study demonstrates the capability of LIBS as an in-situ monitoring tool for detection of elemental profile in different layers of optical window of the Tokamak. (author)

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

International Nuclear Information System (INIS)

Kočan, M.; Müller, H.W.; Lunt, T.; Bernert, M.; Conway, G.D.; De Marné, P.; Eich, T.; Herrmann, A.; Kallenbach, A.; Maraschek, M.; Müller, S.; Nold, B.; Huang, Z.; Adámek, J.; Horacek, J.; Komm, M.; Allan, S.Y.; Elmore, S.; Gennrich, F.P; Mehlmann, F.

2013-01-01

We report the latest results of turbulence and transport studies in the ASDEX Upgrade scrape-off layer (SOL). Dissimilarity between the plasma and the floating potential fluctuations is studied experimentally and by gyrofluid simulations. Measurements by a retarding field analyser reveal that both, edge-localized mode (ELM) and turbulent filaments, convey hot ions over large radial distances in the SOL. The measured far SOL ELM ion temperature increases with the ELM energy, consistent with earlier observations that large ELMs deposit a large fraction of their energy outside the divertor. In the SOL, the ELM suppression by magnetic perturbations (MPs) results in lower ELM ion energy in the far SOL. At the same time, large filaments of ion saturation current are replaced by more continuous bursts. Splitting of the divertor strike zones observed by the infrared imaging in H-mode with MPs agree with predictions from the EMC3-Eirene simulations. This suggests that the ‘lobe’ structures due to perturbation fields observed near the X-point are not significantly affected by plasma screening, and can be described by a vacuum approach, as in the EMC3-Eirene. Finally, some effects of the MPs on the L-mode SOL are addressed. (paper)

A phenomenological model for cross-field plasma transport in non-ambipolar scrape-off layers

International Nuclear Information System (INIS)

LaBombard, B.; Grossman, A.A.; Conn, R.W.

1990-01-01

A simplified two-fluid transport model which includes phenomenological coefficients of particle diffusion, mobility, and thermal diffusivity is used to investigate the effects of nonambipolar particle transport on scrape-off layer (SOL) plasma profiles. A computer code (BSOLRAD3) has been written to iteratively solve for 2-D cross-field density, potential, and electron temperature profiles for arbitrary boundary conditions, including segments of 'limiters' that are electrically conducting or non-conducting. Numerical results are presented for two test cases: (1) a 1-D slab geometry showing the interdependency of the density, potential, and temperature gradient scale lengths on particle diffusion, mobility, and thermal diffusivity coefficients and limiter bias conditions, and (2) a 2-D geometry illustrating ExB plasma flow effects. It is shown that the SOL profiles can be quite sensitive to non-ambipolarity conditions imposed by the limiter and, in particular, whether the limiter surfaces are biased. Such effects, if overlooked in SOL transport analysis, can lead to erroreous conclusions about the magnitude of the local ambipolar diffusion coefficient. (orig.)

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

International Nuclear Information System (INIS)

1982-03-01

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

Investigation into the formation of the scrape-off layer density shoulder in JET ITER-like wall L-mode and H-mode plasmas

Science.gov (United States)

Wynn, A.; Lipschultz, B.; Cziegler, I.; Harrison, J.; Jaervinen, A.; Matthews, G. F.; Schmitz, J.; Tal, B.; Brix, M.; Guillemaut, C.; Frigione, D.; Huber, A.; Joffrin, E.; Kruzei, U.; Militello, F.; Nielsen, A.; Walkden, N. R.; Wiesen, S.; Contributors, JET

2018-05-01

The low temperature boundary layer plasma (scrape-off layer or SOL) between the hot core and the surrounding vessel determines the level of power loading, erosion and implantation of material surfaces, and thus the viability of tokamak-based fusion as an energy source. This study explores mechanisms affecting the formation of flattened density profiles, so-called ‘density shoulders’, in the low-field side (LFS) SOL, which modify ion and neutral fluxes to surfaces—and subsequent erosion. We find that increases in SOL parallel resistivity, Λdiv (=[L || ν eiΩi]/c sΩe), postulated to lead to shoulder growth through changes in SOL turbulence characteristics, correlates with increases in SOL shoulder amplitude, A s, only under a subset of conditions (D2-fuelled L-mode density scans with outer strike point on the horizontal target). Λdiv fails to correlate with A s for cases of N2 seeding or during sweeping of the strike point across the horizontal target. The limited correlation of Λdiv and A s is also found for H-mode discharges. Thus, while it may be necessary for Λdiv to be above a threshold of ~1 for shoulder formation and/or growth, another mechanism is required. More significantly, we find that in contrast to parallel resistivity, outer divertor recycling, as quantified by the total outer divertor Balmer D α emission, I-D α , does scale with A s where Λdiv does and even where Λdiv does not. Divertor recycling could lead to SOL density shoulder formation through: (a) reducing the parallel to the field flow (loss) of ions out of the SOL to the divertor; and (b) changes in radial electric fields which lead to E  ×  B poloidal flows as well as potentially affecting SOL turbulence birth characteristics. Thus, changes in divertor recycling may be the sole process involved in bringing about SOL density shoulders or it may be that it acts in tandem with parallel resistivity.

Research using small tokamaks

International Nuclear Information System (INIS)

1991-05-01

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

Shear flows at the tokamak edge and their interaction with edge-localized modes

International Nuclear Information System (INIS)

Aydemir, A. Y.

2007-01-01

Shear flows in the scrape-off layer (SOL) and the edge pedestal region of tokamaks are shown to arise naturally out of transport processes in a magnetohydrodynamic model. In quasi-steady-state conditions, collisional resistivity coupled with a simple bootstrap current model necessarily leads to poloidal and toroidal flows, mainly localized to the edge and SOL. The role of these flows in the grad-B drift direction dependence of the power threshold for the L (low) to H (high) transition, and their effect on core rotation, are discussed. Theoretical predictions based on symmetries of the underlying equations, coupled with computational results, are found to be in agreement with observations in Alcator C-Mod [Phys. Plasmas 12, 056111 (2005)]. The effects of these self-consistent flows on linear peeling/ballooning modes and their nonlinear consequences are also examined

Impact of E × B flow shear on turbulence and resulting power fall-off width in H-mode plasmas in experimental advanced superconducting tokamak

Energy Technology Data Exchange (ETDEWEB)

Yang, Q. Q., E-mail: yangqq@ipp.ac.cn; Zhong, F. C., E-mail: gsxu@ipp.ac.cn, E-mail: fczhong@dhu.edu.cn; Jia, M. N. [College of Science, Donghua University, Shanghai 201620 (China); Xu, G. S., E-mail: gsxu@ipp.ac.cn, E-mail: fczhong@dhu.edu.cn; Wang, L.; Wang, H. Q.; Chen, R.; Yan, N.; Liu, S. C.; Chen, L.; Li, Y. L.; Liu, J. B. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

2015-06-15

The power fall-off width in the H-mode scrape-off layer (SOL) in tokamaks shows a strong inverse dependence on the plasma current, which was noticed by both previous multi-machine scaling work [T. Eich et al., Nucl. Fusion 53, 093031 (2013)] and more recent work [L. Wang et al., Nucl. Fusion 54, 114002 (2014)] on the Experimental Advanced Superconducting Tokamak. To understand the underlying physics, probe measurements of three H-mode discharges with different plasma currents have been studied in this work. The results suggest that a higher plasma current is accompanied by a stronger E×B shear and a shorter radial correlation length of turbulence in the SOL, thus resulting in a narrower power fall-off width. A simple model has also been applied to demonstrate the suppression effect of E×B shear on turbulence in the SOL and shows relatively good agreement with the experimental observations.

Drifts, currents, and power scrape-off width in SOLPS-ITER modeling of DIII-D

International Nuclear Information System (INIS)

Meier, E. T.; Goldston, R. J.; Kaveeva, E. G.; Makowski, M. A.; Mordijck, S.

2016-01-01

The effects of drifts and associated flows and currents on the width of the parallel heat flux channel (λ q ) in the tokamak scrape-off layer (SOL) are analyzed using the SOLPS-ITER 2D fluid transport code. Motivation is supplied by Goldston’s heuristic drift (HD) model for λ q , which yields the same approximately inverse poloidal magnetic field dependence seen in multi-machine regression. The analysis, focusing on a DIII-D H-mode discharge, reveals HD-like features, including comparable density and temperature fall-off lengths in the SOL, and up-down ion pressure asymmetry that allows net cross-separatrix ion magnetic drift flux to exceed net anomalous ion flux. In experimentally relevant high-recycling cases, scans of both toroidal and poloidal magnetic field (B tor and B pol ) are conducted, showing minimal λ q dependence on either component of the field. Insensitivity to B tor is expected, and suggests that SOLPS-ITER is effectively capturing some aspects of HD physics. Absence of λ q dependence on B pol , however, is inconsistent with both the HD model and experimental results. As a result, the inconsistency is attributed to strong variation in the parallel Mach number, which violates one of the premises of the HD model.

Low plasma edge temperatures for the self-pumped limiter

International Nuclear Information System (INIS)

Terry, W.K.; Brooks, J.N.

1985-03-01

Transport code calculations have been performed to study the operation of an INTOR-like tokamak plasma from which helium is removed by a self-pumped limiter, which traps helium, but not hydrogen, in its surface layers. To prevent saturation by helium, the surface is renewed by continuous injection of the surface material (vanadium in this study) into the scrape-off layer. The presence of the injected vanadium leads to plasma temperatures well below 50 eV in the scrape-off layer, with supplementary rf heating. Operation in this edge temperature regime is essential for the use of medium- and high-Z limiter coatings

Comparison of the effects of an ICRF antenna with insulating side limiters with and without a Faraday screen on the edge parameters of a tokamak plasma

International Nuclear Information System (INIS)

Sorensen, J.; Diebold, D.A.; Majeski, R.; Hershkowitz, N.

1993-01-01

The PHAEDRUS-T tokamak was operated with radiofrequency power near the ion cyclotron frequency at 90 deg. C phasing between two current straps with and without a stainless steel Faraday screen covering the antenna. In both cases, the sides of the antenna were protected by insulating limiters. The plasma parameters in the scrape-off layer were measured and were shown to be essentially the same when radiofrequency power was applied from the Faraday screen covered antenna as compared with the antenna without a Faraday screen. The intensity of Fe(XVI) light dropped an order of magnitude after the screen was removed. (author). 18 refs, 3 figs

Turbulent and neoclassical toroidal momentum transport in tokamak plasmas

International Nuclear Information System (INIS)

Abiteboul, J.

2012-10-01

then analyzed and, although the conventional gyro-Bohm scaling is recovered on average, local processes are found to be clearly non-diffusive. The impact of scrape-off layer flows on core toroidal rotation is also analyzed by modifying the boundary conditions in GYSELA. Finally, the equilibrium magnetic field in tokamaks, which is not rigorously axisymmetric, provides another means of breaking the toroidal symmetry, through purely collisional processes. This effect is found to contribute significantly to toroidal momentum transport and can compete with the turbulence-driven toroidal rotation in tokamaks. (author)

Combined core/boundary layer plasma transport simulations in tokamaks

International Nuclear Information System (INIS)

Prinja, A.K.; Schafer, R.F. Jr.; Conn, R.W.; Howe, H.C.

1987-01-01

Significant new numerical results are presented from self-consistent core and boundary or scrape-off layer plasma simulations with 3-D neutral transport calculations. For a symmetric belt limiter it is shown that, for plasma conditions considered here, the pump limiter collection efficiency increases from 11% to 18% of the core efflux as a result of local reionization of blade deflected neutrals. This hitherto unobserved effect causes a significant amplification of upstream ion flux entering the pump limiter. Results from coupling of an earlier developed two-zone edge plasma model ODESSA to the PROCTR core plasma simulation code indicates that intense recycling divertor operation may not be possible because of stagnation of upstream flow velocity. This results in a self-consistent reduction of density gradient in an intermediate region between the central plasma and separatrix, and a concomitant reduction of core-efflux. There is also evidence of increased recycling at the first wall. (orig.)

Convergence of statistical moments of particle density time series in scrape-off layer plasmas

Energy Technology Data Exchange (ETDEWEB)

Kube, R., E-mail: ralph.kube@uit.no; Garcia, O. E. [Department of Physics and Technology, UiT - The Arctic University of Norway, N-9037 Tromsø (Norway)

2015-01-15

Particle density fluctuations in the scrape-off layer of magnetically confined plasmas, as measured by gas-puff imaging or Langmuir probes, are modeled as the realization of a stochastic process in which a superposition of pulses with a fixed shape, an exponential distribution of waiting times, and amplitudes represents the radial motion of blob-like structures. With an analytic formulation of the process at hand, we derive expressions for the mean squared error on estimators of sample mean and sample variance as a function of sample length, sampling frequency, and the parameters of the stochastic process. Employing that the probability distribution function of a particularly relevant stochastic process is given by the gamma distribution, we derive estimators for sample skewness and kurtosis and expressions for the mean squared error on these estimators. Numerically, generated synthetic time series are used to verify the proposed estimators, the sample length dependency of their mean squared errors, and their performance. We find that estimators for sample skewness and kurtosis based on the gamma distribution are more precise and more accurate than common estimators based on the method of moments.

Convergence of statistical moments of particle density time series in scrape-off layer plasmas

International Nuclear Information System (INIS)

Kube, R.; Garcia, O. E.

2015-01-01

Particle density fluctuations in the scrape-off layer of magnetically confined plasmas, as measured by gas-puff imaging or Langmuir probes, are modeled as the realization of a stochastic process in which a superposition of pulses with a fixed shape, an exponential distribution of waiting times, and amplitudes represents the radial motion of blob-like structures. With an analytic formulation of the process at hand, we derive expressions for the mean squared error on estimators of sample mean and sample variance as a function of sample length, sampling frequency, and the parameters of the stochastic process. Employing that the probability distribution function of a particularly relevant stochastic process is given by the gamma distribution, we derive estimators for sample skewness and kurtosis and expressions for the mean squared error on these estimators. Numerically, generated synthetic time series are used to verify the proposed estimators, the sample length dependency of their mean squared errors, and their performance. We find that estimators for sample skewness and kurtosis based on the gamma distribution are more precise and more accurate than common estimators based on the method of moments

Physics design and experimental study of tokamak divertor

International Nuclear Information System (INIS)

Yan Jiancheng; Gao Qingdi; Yan Longwen; Wang Mingxu; Deng Baiquan; Zhang Fu; Zhang Nianman; Ran Hong; Cheng Fayin; Tang Yiwu; Chen Xiaoping

2007-06-01

The divertor configuration of HL-2A tokamak is optimized, and the plasma performance in divertor is simulated with B2-code. The effects of collisionality on plasma-wall transition in the scrape-off layer of divertor are investigated, high performances of the divertor plasma in HL-2A are simulated, and a quasi- stationary RS operation mode is established with the plasma controlled by LHCD and NBI. HL-2A tokamak has been successfully operated in divertor configuration. The major parameters: plasma current I p =320 kA, toroidal field B t =2.2 T, plasma discharger duration T d =1580 ms ware achieved at the end of 2004. The preliminary experimental researches of advanced diverter have been carried out. Design studies of divertor target plate for high power density fusion reactor have been carried out, especially, the physical processes on the surface of flowing liquid lithium target plate. The exploration research of improving divertor ash removal efficiency and reducing tritium inventory resulting from applying the RF ponderomotive force potential is studied. The optimization structure design studies of FEB-E reactor divertor are performed. High flux thermal shock experiments were carried on tungsten and carbon based materials. Hot Isostatic Press (HIP) method was employed to bond tungsten to copper alloys. Electron beam simulated thermal fatigue tests were also carried out to W/Cu bondings. Thermal desorption and surface modification of He + implanted into tungsten have been studied. (authors)

Impurity penetration through the stochastic layer near the separatrix in tokamaks

International Nuclear Information System (INIS)

Morozov, D.K.; Herrera, J.J.E.; Rantsev-Kartinov, V.A.

1995-01-01

It is shown that a stochastic layer produced by ripple perturbations near the separatrix in tokamaks, leads to anomalous plasma flow out of the bulk plasma along perturbed field lines, which brings out impurities. This suggests that the stochastic layer may play a cleaning role. There is an opposite process of anomalous impurity diffusion into the plasma. The balance of these two processes defines the impurity concentration in the bulk plasma. copyright 1995 American Institute of Physics

Non-axisymmetric SOL-transport study for tokamaks and stellarators

International Nuclear Information System (INIS)

Sardei, F.; Feng, Y.; Kisslinger, J.; Grigull, P.; Kobayashi, M.; Harting, D.; Reiter, D.; Federici, G.; Loarte, A.

2007-01-01

The paper addresses basic features of non-axisymmetric edge transport induced in tokamaks by local limiters or external magnetic perturbations and in low-shear stellarators by the presence of edge magnetic islands. 3D simulations and, if available for comparison, experimental results are presented and discussed for three devices, ITER during start-up operation, TEXTOR-DED and W7-AS, having edge topologies totally different from each other. The modeling is performed with the EMC3/EIRENE code, which treats self-consistently plasma, neutral and impurity transport in a general 3D scrape-off layer (SOL) with arbitrarily complex geometry of magnetic configuration and plasma-facing components. Shown are code predictions of the power load on the ITER start-up limiters as well as modeling results on the transport in the TEXTOR-DED stochastic edge and on the physics of stable detachment in W7-AS. Experimental observations confirming the code simulations are referenced for both TEXTOR-DED and W7-AS, a direct comparison between modeling and experimental results is shown for W7-AS

Edge turbulence imaging in the Alcator C-Mod tokamak

International Nuclear Information System (INIS)

Zweben, S.J.; Stotler, D.P.; Terry, J.L.; La Bombard, B.; Greenwald, M.; Muterspaugh, M.; Pitcher, C.S.; Hallatschek, K.; Maqueda, R.J.; Rogers, B.; Lowrance, J.L.; Mastrocola, V.J.; Renda, G.F.

2002-01-01

The two-dimensional (2D) radial vs poloidal structure of edge turbulence in the Alcator C-Mod tokamak [I. H. Hutchinson, R. Boivin, P. T. Bonoli et al., Nucl. Fusion 41, 1391 (2001)] was measured using fast cameras and compared with three-dimensional numerical simulations of edge plasma turbulence. The main diagnostic is gas puff imaging, in which the visible D α emission from a localized D 2 gas puff is viewed along a local magnetic field line. The observed D α fluctuations have a typical radial and poloidal scale of ≅1 cm, and often have strong local maxima ('blobs') in the scrape-off layer. The motion of this 2D structure motion has also been measured using an ultrafast framing camera with 12 frames taken at 250 000 frames/s. 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

Collisional boundary layer analysis for neoclassical toroidal plasma viscosity in tokamaks

Czech Academy of Sciences Publication Activity Database

Shaing, K.C.; Cahyna, Pavel; Bécoulet, M.; Park, J.-K.; Sabbagh, S.A.; Chu, M.S.

2008-01-01

Roč. 15, č. 8 (2008), 082506-1-7 ISSN 1070-664X Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma boundary layers * plasma toroidal confinement * Tokamak devices Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.427, year: 2008 http://dx.doi.org/10.1063/1.2969434

Thermographic analysis of plasma facing components covered by carbon surface layer in tokamaks

International Nuclear Information System (INIS)

Gardarein, Jean-Laurent

2007-01-01

Tokamaks are reactors based on the thermonuclear fusion energy with magnetic confinement of the plasma. In theses machines, several MW are coupled to the plasma for about 10 s. A large part of this power is directed towards plasma facing components (PFC). For better understanding and control the heat flux transfer from the plasma to the surrounding wall, it is very important to measure the surface temperature of the PFC and to estimate the imposed heat flux. In most of tokamaks using carbon PFC, the eroded carbon is circulating in the plasma and redeposited elsewhere. During the plasma operations, this leads at some locations to the formation of thin or thick carbon layers usually poorly attached to the PFC. These surface layers with unknown thermal properties complicate the calculation of the heat flux from IR surface temperature measurements. To solve this problem, we develop first, inverse method to estimate the heat flux using thermocouple (not sensitive to the carbon surface layers) temperature measurements. Then, we propose a front face pulsed photothermal method allowing an estimation of layers thermal diffusivity, conductivity, effusivity and the thermal contact resistance between the layer and the tile. The principle is to study with an infrared sensor, the cooling of the layer surface after heating by a short laser pulse, this cooling depending on the thermal properties of the successive layers. (author) [fr

The role of drifts in the plasma transport at the tokamak core–SOL interface

Energy Technology Data Exchange (ETDEWEB)

Chankin, A.V., E-mail: Alex.Chankin@ipp.mpg.de [Max-Planck-Institut für Pasmaphysik, EURATOM Association, 85748 Garching (Germany); Coster, D.P. [Max-Planck-Institut für Pasmaphysik, EURATOM Association, 85748 Garching (Germany)

2013-07-15

The interface between the core (inside the magnetic separatrix in X-point configurations) and the scrape-off layer (SOL) of tokamaks is a delicate region of the magnetic topology transition from closed to open field lines where neither the standard neoclassical theory nor the SOL physics fully apply. Sharp gradients of plasma parameters in the outer core, caused by the proximity of divertor sinks in the near SOL, invalidate some ordering assumptions of the neoclassical theory. At the same time, the existence of closed flux surfaces in the core enforces ambipolarity of radial plasma flows, in difference to the situation in the SOL where the current loop may close through the divertor. Detailed analysis of the plasma transport and flows with the emphasis on the outer core region, just inside the separatrix, is carried out in the paper, based on EDGE2D modelling and analytical formulas.

Statistical characterization of turbulence in the boundary plasma of EAST

DEFF Research Database (Denmark)

Yan, Ning; Nielsen, Anders Henry; Xu, G.S.

2013-01-01

In Ohmic heated low confinement mode (L-mode) discharges, the intermittent statistical characteristics of turbulent fluctuations have been investigated in the edge and the scrape-off layer (SOL) plasma on EAST (the experimental advanced superconducting tokamak) by fast reciprocating Langmuir probe...

Characteristics of the scrape-off layer in DIII-D high-performance negative central magnetic shear discharges

Energy Technology Data Exchange (ETDEWEB)

Lasnier, C.J. [General Atomics, San Diego, CA (United States); Maingi, R. [General Atomics, San Diego, CA (United States); Leonard, A.W. [General Atomics, San Diego, CA (United States); Allen, S.L. [General Atomics, San Diego, CA (United States); Buchenauer, D.A. [General Atomics, San Diego, CA (United States); Burrell, K.H. [General Atomics, San Diego, CA (United States); Casper, T.A. [General Atomics, San Diego, CA (United States); Cuthbertson, J.W. [General Atomics, San Diego, CA (United States); Fenstermacher, M.E. [General Atomics, San Diego, CA (United States); Hill, D.N. [General Atomics, San Diego, CA (United States); Jong, R.A. [General Atomics, San Diego, CA (United States); Lao, L.L. [General Atomics, San Diego, CA (United States); Lazarus, E.A. [General Atomics, San Diego, CA (United States); Moyer, R.A. [General Atomics, San Diego, CA (United States); Petrie, T.W. [General Atomics, San Diego, CA (United States); Porter, G.D. [General Atomics, San Diego, CA (United States); Rice, B.W. [General Atomics, San Diego, CA (United States); Stallard, B.W. [General Atomics, San Diego, CA (United States); Taylor, T.S. [General Atomics, San Diego, CA (United States); Watkins, J.G. [General Atomics, San Diego, CA (United States)

1997-02-01

In this paper we present measurements of the global power and particle balance in the high-performance phase of negative central magnetic shear (NCS) discharges and compare with reference VH-mode discharges. The principal differences observed are that NCS has a much lower fraction of the total input power flowing into the boundary, less core radiation, and larger rate of stored energy increase as a fraction of total power. Scrape-off layer (SOL) temperature and divertor heat flux profiles, and radiation profiles at the midplane, are similar to VH-mode. Due to the good core particle confinement and efficient fueling by neutral beam injection (NBI), with little gas puffing, the gas load on the walls and the recycling are very low during the NCS discharges. This results in a rate of density rise relative to beam fueling at the L to H transition time which is 1/3 of the value for VH transitions, which is in turn 1/2 that for L-to-ELMing-H-mode transitions. (orig.).

Reduction of the turbulent blob transport in the scrape-off layer by a resonant magnetic perturbation in TEXTOR

International Nuclear Information System (INIS)

Xu, Y.; Weynants, R.R.; Van Schoor, M.; Vergote, M.; Jachmich, S.; Jakubowski, M.W.; Mitri, M.; Schmitz, O.; Unterberg, B.; Reiser, D.; Finken, K.H.; Lehnen, M.; Beyer, P.

2009-01-01

During the static 6/2 Dynamic Ergodic Divertor experiments in TEXTOR, a significant influence of the edge resonant magnetic perturbation (RMP) on the turbulent blob transport in the scrape-off layer (SOL) has been observed. In ohmic discharges without the RMP, the blobs extend 4-5 cm deep into the SOL with a radially outward moving speed of about 1 km s -1 and hence constitute a strong outflow of mass. With the application of the RMP, the blob amplitudes and their radially moving velocity are both reduced, resulting in a significant reduction of the blob transport in the SOL. The reduction effect of the RMP on blobs is found to be robust to changes in the operational regime and to phasing variations of the RMP as well. The blob dynamics appears to be consistent with the paradigm of the radial motions of the blob structures driven by the interchange instability.

A monte carlo simulation model for the steady-state plasma in the scrape-off layer

International Nuclear Information System (INIS)

Wang, W.X.; Okamoto, M.; Nakajima, N.; Murakami, S.; Ohyabu, N.

1995-12-01

A new Monte Carlo simulation model for the scrape-off layer (SOL) plasma is proposed to investigate a feasibility of so-called 'high temperature divertor operation'. In the model, Coulomb collision effect is accurately described by a nonlinear Monte Carlo collision operator; a conductive heat flux into the SOL is effectively modelled via randomly exchanging the source particles and SOL particles; secondary electrons are included. The steady state of the SOL plasma, which satisfies particle and energy balances and the neutrality constraint, is determined in terms of total particle and heat fluxes across the separatrix, the edge plasma temperature, the secondary electron emission rate, and the SOL size. The model gives gross features of the SOL such as plasma temperatures and densities, the total sheath potential drop, and the sheath energy transmission factor. The simulations are performed for collisional SOL plasma to confirm the validity of the proposed model. It is found that the potential drop and the electron energy transmission factor are in close agreement with theoretical predictions. The present model can provide primarily useful information for collisionless SOL plasma which is difficult to be understood analytically. (author)

Attempt to model the edge turbulence of a tokamak as a random superposition of eddies

Energy Technology Data Exchange (ETDEWEB)

Endler, M; Theimer, G; Weinlich, M; Carlson, A; Giannone, L.; Niedermeyer, H; Rudyj, A [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

1993-12-31

Turbulence is considered to be the most likely origin of the anomalous transport in tokamaks. Although the main interest is focussed on the bulk plasma, transport in the scrape-off layer is very important for reactor design. For this reason extensive experimental investigations of the edge turbulence were performed on the ASDEX divertor tokamak. Langmuir probe arrays were used in the floating potential mode and in the ion saturation mode to measure the poloidal distribution of density and plasma potential fluctuations neglecting temperature fluctuations. Density fluctuations integrated radially over the boundary layer were derived from H{sub {alpha}}-measurements. Data from up to 16 channels were sampled with a frequency of 1 MHz during time windows of 1 s. Often one parameter like the plasma density or the radial probe position were scanned during this interval. It is impossible to derive physical mechanisms directly from these statistical observations. We draw general conclusions about the physics involved from the entity of observations and propose a set of basic effects to include in a theoretical model. Being still unable to solve the complex nonlinear problem of the fully developed turbulence exactly we attempt to describe the turbulence with a simple non-self-consistent statistical model. This allows to derive plausible physical interpretations of several features of the statistical functions and may be used as a guide-line for the development of a manageable theoretical model. (author) 6 refs., 3 figs.

The electron temperature gradient instability in presence of a limiter with tilted plates

International Nuclear Information System (INIS)

Farina, D.; Pozzoli, R.; Ryutov, D.

1994-01-01

The analysis of the electron temperature gradient instability in the scrape-off layer is generalized to the case of non-orthogonal intersections of the magnetic field with the wall surface, a situation which is most typical for a tokamak with a limiter. (orig.)

TEMPEST simulations of the plasma transport in a single-null tokamak geometry

International Nuclear Information System (INIS)

Xu, X.Q.; Cohen, R.H.; Rognlien, T.D.; Bodi, K.; Krasheninnikov, S.

2010-01-01

We present edge kinetic ion transport simulations of tokamak plasmas in magnetic divertor geometry using the fully nonlinear (full-f) continuum code TEMPEST. Besides neoclassical transport, a term for divergence of anomalous kinetic radial flux is added to mock up the effect of turbulent transport. To study the relative roles of neoclassical and anomalous transport, TEMPEST simulations were carried out for plasma transport and flow dynamics in a single-null tokamak geometry, including the pedestal region that extends across the separatrix into the scrape-off layer and private flux region. A series of TEMPEST simulations were conducted to investigate the transition of midplane pedestal heat flux and flow from the neoclassical to the turbulent limit and the transition of divertor heat flux and flow from the kinetic to the fluid regime via an anomalous transport scan and a density scan. The TEMPEST simulation results demonstrate that turbulent transport (as modelled by large diffusion) plays a similar role to collisional decorrelation of particle orbits and that the large turbulent transport (large diffusion) leads to an apparent Maxwellianization of the particle distribution. We also show the transition of parallel heat flux and flow at the entrance to the divertor plates from the fluid to the kinetic regime. For an absorbing divertor plate boundary condition, a non-half-Maxwellian is found due to the balance between upstream radial anomalous transport and energetic ion endloss.

Drift-based Model for Power Scrape-off Width in Low-Gas-Puff H-mode Plasmas: Theory and Implications

Energy Technology Data Exchange (ETDEWEB)

Goldston, R., E-mail: rgoldston@pppl.gov [Princeton Plasma Physics Laboratory, Princeton (United States)

2012-09-15

Full text: A heuristic model for the plasma scrape-off width in low-gas-puff tokamak H-mode plasmas is introduced. {nabla}B and curvature drifts into the scrape-off layer (SOL) are balanced against near-sonic parallel flows out of the SOL, to the divertor plates. These assumptions result in an estimated SOL width of order the poloidal gyroradius. It is next assumed that anomalous perpendicular electron thermal diffusivity is the dominant source of heat flux across the separatrix, investing the SOL width, derived above, with heat from the main plasma. The separatrix temperature is then calculated based on a two-point model balancing power input to the SOL with Spitzer-Hiarm parallel thermal conduction losses to the divertor. This results in a heuristic closed-form prediction for the power scrape-off width that is in quantitative agreement both in absolute magnitude and in scaling with recent experimental data. The applicability of the Spitzer-Harm model to this regime can be questioned at the lowest densities, where the presence of a sheath can raise the divertor target electron temperature. A more general two-point model including a finite ratio of divertor target to upstream electron temperature shows only a 5% effect on the SOL width with target temperature f{sub T} = 75% of upstream, so this effect is likely negligible in experimentally relevant regimes. To achieve the near-sonic flows measured experimentally, and assumed in this model, sets requirements on the ratio of upstream to total SOL particle sources relative to the square-root of the ratio of target to upstream temperature. As a result very high recycling regimes may allow significantly wider power fluxes. The Pfisch-Schluter model for equilibrium flows has been modified to allow near-sonic flows, appropriate for gradient scale lengths of order the poloidal gyroradius. This results in a new quadrupole flow pattern that amplifies the usual P-S flows at the outer midplane, while reducing them at the inner

L-mode SOL width scaling in the MAST spherical tokamak

International Nuclear Information System (INIS)

Ahn, J-W; Counsell, G F; Kirk, A

2006-01-01

A new data-set of outboard mid-plane scrape-off layer (SOL) heat flux widths, Δ h , has been constructed for L-mode plasmas in the MAST spherical tokamak (ST). The scaling with key plasma parameters such as density, toroidal magnetic field, parallel connection length in the SOL and surface heat flux at the separatrix is investigated. An empirical scaling is developed for the Δ h data-set, which exhibits a strong positive dependence on both the connection length (or edge safety factor) and density and weak or moderate inverse dependences on the surface heat flux and magnetic field, respectively. The empirical scaling is compared with earlier results for a range of tokamaks with conventional geometry, which show weaker dependence on the density and edge safety factor. Importantly, however, the weak negative dependence on the surface heat flux (and thus heating power) is common in both conventional and ST geometries. The experimental data are also used to test a number of dimensionally correct Δ h scalings developed from theoretical models for perpendicular transport in the SOL coupled with classical transport parallel to the magnetic field. A scaling based on perpendicular transport driven by resistive MHD interchange provides the best fit, although several models are close. A subset of the better fitting theoretical scalings are used to extrapolate for Δ h in one design for a future burning ST machine and finally to predict the peak heat loading on the outboard divertor target plate

Theory of nonaxisymmetric vertical displacement events in tokamaks

International Nuclear Information System (INIS)

Fitzpatrick, R.

2011-01-01

A semi-analytic sharp-boundary model of a nonaxisymmetric vertical displacement event (VDE) in a large aspect-ratio, high-beta (i.e. β ∼ ε), vertically elongated tokamak plasma is developed. The model is used to simulate nonaxisymmetric VDEs with a wide range of different plasma equilibrium and vacuum vessel parameters. These simulations yield poloidal halo current fractions and toroidal peaking factors whose magnitudes are similar to those seen in experiments, and also reproduce the characteristic inverse scaling between the halo current fraction and the toroidal peaking factor. Moreover, the peak poloidal halo current density in the vacuum vessel is found to correlate strongly with the reciprocal of the minimum edge safety factor attained during the VDE. In addition, under certain circumstances, the ratio of the net sideways force acting on the vacuum vessel to the net vertical force is observed to approach unity. Finally, the peak vertical force per unit area acting on the vessel is found to have a strong correlation with the equilibrium toroidal plasma current at the start of the VDE, but is also found to increase with increasing vacuum vessel resistivity relative to the scrape-off layer plasma.

Kinetic and transport theory near the tokamak edge

International Nuclear Information System (INIS)

Hazeltine, R.D.; Catto, P.J.

1995-12-01

Conventional transport orderings employed in the core of a tokamak plasma allow large divergence-free flows in flux surfaces, but only weak radial flows. However, alternate orderings are required in the edge region where radial diffusion must balance the rapid loss due to free-streaming to divertor plates or limiters. Kinetic equations commonly used to study the plasma core do not allow such a balance and are, therefore, inapplicable in the plasma edge. Similarly, core transport formulae cannot be extended to the edge region without major, qualitative alteration. Here the authors address the necessary changes. By deriving and solving a novel kinetic equation, they construct distinctive collisional transport laws for the plasma edge. They find that their edge ordering naturally retains the radial diffusion and parallel flow of particles, momentum and heat to lowest order in the conservation equations. To higher order they find a surprising form for parallel transport in the scrape-off layer, in which the parallel flow of particles and heat are driven by a combination of the conventional gradients, viscosity, and new terms involving radial derivatives. The new terms are not relatively small, and could affect understanding of limiter and divertor operation

Modeling of thermal effects on TIBER II [Tokamak Ignition/Burn Experimental Reactor] divertor during plasma disruption

International Nuclear Information System (INIS)

Bruhn, M.L.; Perkins, L.J.

1987-01-01

Mapping the disruption power flow from the mid-plane of the TIBER Engineering Test Reactor to its divertor and calculating the resulting thermal effects are accomplished through the modification and coupling of three presently existing computer codes. The resulting computer code TADDPAK (Thermal Analysis Divertor during Disruption PAcKage) provides three-dimensional graphic presentations of time and positional dependent thermal effects on a poloidal cross section of the double-null-divertor configured reactor. These thermal effects include incident heat flux, surface temperature, vaporization rate, total vaporization, and melting depth. The dependence of these thermal effects on material choice, disruption pulse shape, and the characteristic thickness of the plasma scrape-off layer is determined through parametric analysis with TADDPAK. This computer code is designed to be a convenient, rapid, and user-friendly modeling tool which can be easily adapted to most tokamak double-null-divertor reactor designs. 14 refs

The influence of secondary electron emission on the floating potential of tokamak-born dust

International Nuclear Information System (INIS)

Vaverka, J; Richterová, I; Vyšinka, M; Pavlů, J; Šafránková, J; Němeček, Z

2014-01-01

Dust production and its transport into the core plasma is an important issue for magnetic confinement fusion. Dust grains are charged by various processes, such as the collection of plasma particles and electron emissions, and their charge influences the dynamics of the dust. This paper presents the results of calculations of the surface potential of dust grains in a Maxwellian plasma. Our calculations include the charging balance of a secondary electron emission (SEE) from the dust. The numerical model that we have used accounts for the influence of backscattered electrons and takes into account the effects of grain size, material, and it is also able to handle both spherical and non-spherical grains. We discuss the role of the SEE under tokamak conditions and show that the SEE is a leading process for the grains crossing the scrape-off layer from the edge to core plasma. The results of our calculations are relevant for materials related to fusion experiments in ITER. (paper)

Density profiles in the Scrape-Off Layer interpreted through filament dynamics

Science.gov (United States)

Militello, Fulvio

2017-10-01

We developed a new theoretical framework to clarify the relation between radial Scrape-Off Layer density profiles and the fluctuations that generate them. The framework provides an interpretation of the experimental features of the profiles and of the turbulence statistics on the basis of simple properties of the filaments, such as their radial motion and their draining towards the divertor. L-mode and inter-ELM filaments are described as a Poisson process in which each event is independent and modelled with a wave function of amplitude and width statistically distributed according to experimental observations and evolving according to fluid equations. We will rigorously show that radially accelerating filaments, less efficient parallel exhaust and also a statistical distribution of their radial velocity can contribute to induce flatter profiles in the far SOL and therefore enhance plasma-wall interactions. A quite general result of our analysis is the resiliency of this non-exponential nature of the profiles and the increase of the relative fluctuation amplitude towards the wall, as experimentally observed. According to the framework, profile broadening at high fueling rates can be caused by interactions with neutrals (e.g. charge exchange) in the divertor or by a significant radial acceleration of the filaments. The framework assumptions were tested with 3D numerical simulations of seeded SOL filaments based on a two fluid model. In particular, filaments interact through the electrostatic field they generate only when they are in close proximity (separation comparable to their width in the drift plane), thus justifying our independence hypothesis. In addition, we will discuss how isolated filament motion responds to variations in the plasma conditions, and specifically divertor conditions. Finally, using the theoretical framework we will reproduce and interpret experimental results obtained on JET, MAST and HL-2A.

Plasma particle sources due to interactions with neutrals in a turbulent scrape-off layer of a toroidally confined plasma

DEFF Research Database (Denmark)

Thrysøe, Alexander Simon; Løiten, M.; Madsen, J.

2018-01-01

The conditions in the edge and scrape-off layer (SOL) of magnetically confined plasmas determine the overall performance of the device, and it is of great importance to study and understand the mechanics that drive transport in those regions. If a significant amount of neutral molecules and atoms...... is present in the edge and SOL regions, those will influence the plasma parameters and thus the plasma confinement. In this paper, it is displayed how neutrals, described by a fluid model, introduce source terms in a plasma drift-fluid model due to inelastic collisions. The resulting source terms...... are included in a four-field drift fluid model, and it is shown how an increasing neutral particle density in the edge and SOL regions influences the plasma particle transport across the lastclosed-flux-surface. It is found that an appropriate gas puffing rate allows for the edge density in the simulation...

Application of the radiating divertor approach to innovative tokamak divertor concepts

Energy Technology Data Exchange (ETDEWEB)

Petrie, T.W., E-mail: petrie@fusion.gat.com [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Allen, S.L.; Fenstermacher, M.E. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Groebner, R.J. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Holcomb, C.T. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Kolemen, E. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); La Haye, R.J. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Lasnier, C.J. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Leonard, A.W.; Luce, T.C. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Maingi, R. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); Moyer, R.A. [University of California San Diego, 9500 Gilman Dr., La Jolla, CA 92093-0417 (United States); Solomon, W.M. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Turco, F. [Columbia University, 2960 Broadway, New York, NY 10027 (United States); Watkins, J.G. [Sandia National Laboratory, PO Box 5800, Albuquerque, NM 87185 (United States)

2015-08-15

We survey the results of recent DIII-D experiments that tested the effectiveness of three innovative tokamak divertor concepts in reducing divertor heat flux while still maintaining acceptable energy confinement under neon/deuterium-based radiating divertor (RD) conditions: (1) magnetically unbalanced high performance double-null divertor (DND) plasmas, (2) high performance double-null “Snowflake” (SF-DN) plasmas, and (3) single-null H-mode plasmas having different isolation from their divertor targets. In general, all three concepts adapt well to RD conditions, achieving significant reduction in divertor heat flux (q{sub ⊥p}) and maintaining high performance metrics, e.g., 50–70% reduction in peak divertor heat flux for DND and SF-DN plasmas that are characterized by β{sub N} ≅ 3.0 and H{sub 98(y,2)} ≈ 1.35. It is also demonstrated that q{sub ⊥p} could be reduced ≈50% by extending the parallel connection length (L{sub ||-XPT}) in the scrape-off layer between the X-point and divertor targets over a variety of the RD and non-RD environments tested.

Frequency spectral broadening of lower hybrid waves in tokamak plasmas - causes and effects

Energy Technology Data Exchange (ETDEWEB)

Pericoli Ridolfini, V; Giannone, L.; Bartiromo, R [Associazione Euratom-ENEA sulla Fusione, Rome (Italy). Centro Ricerche Energia Frascati

1994-04-01

The frequency spectral broadening of lower hybrid (LH) waves injected into tokamak plasmas is extensively analyzed with reference mostly to experimental data from the ASDEX tokamak. The link between the magnitude of the pump spectral width and the degradation of the LH current drive efficiency (up to a factor of 2), pointed out in previous works, is explained. The experimental behaviour of LH power absorption is also well reproduced, even in situations when the access of the launched LH waves to the core plasma should be largely forbidden. Experiments are described that are aimed at determined whether parametric decay instabilities (PDIs) or scattering of LH waves by density fluctuations in the plasma edge are causes of the broadening of the LH pump frequency spectrum. Fluctuations emerge as the largely dominant process, while no signature of PDI processes is observed. Careful measurements of the density fluctuations in the ASDEX scrape-off layer plasma allow the analytical description given by Andrews and Perkins to be assumed as the appropriate model of LH scattering. Indeed, it supplies the correct magnitude for the frequency spectral width of the LH pump, and explains quantitatively, together with a ray tracing code, why the CD efficiency decreases with the broadening of the pump spectrum. It can also account for the observed LH power absorption coefficient. (author). 48 refs, 13 figs, 2 tabs.

Effects of q and high beta on tokamak stability

International Nuclear Information System (INIS)

Brickhouse, N.S.; Callen, J.D.; Dexter, R.N.

1984-08-01

In the Columbia University Torus II tokamak plasmas have been studied with volume averaged toroidal beta values as high as 15%. Experimental equilibria have been compared with a 2D free boundary MHD equilibrium code PSEC. The stability of these equilibria has been computed using PEST, the predictions of which are compatible with an observed instability in Torus II which may be characterized as a high toroidal mode number ballooning fluctuation. In the University of Wisconsin Tokapole II tokamak disruptive instability behavior is investigated, with plasma able to be confined on closed magnetic surfaces in the scrape-off region, as the cylindrical edge safety factor is varied from q approx. 3 to q approx. 0.5. It is observed that at q/sub a/ approx. 3 major disruption activity occurs without current terminations, at q/sub a/ less than or equal to 2 well-confined plasmas are obtained without major disruption, and at q/sub a/ approx. 0.5 only partial reconnection accompanies minor disruptions

Effect of the scrape-off layer in AORSA full wave simulations of fast wave minority, mid/high harmonic, and helicon heating regimes

Energy Technology Data Exchange (ETDEWEB)

Bertelli, N., E-mail: nbertell@pppl.gov; Gerhardt, S.; Hosea, J. C.; LeBlanc, B.; Perkins, R. J.; Phillips, C. K.; Taylor, G.; Valeo, E. J.; Wilson, J. R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Jaeger, E. F. [XCEL Engineering Inc., Oak Ridge, TN 37830 (United States); Lau, C.; Blazevski, D.; Green, D. L.; Berry, L.; Ryan, P. M. [Oak Ridge National Laboratory, Oak Ridge, TN 37831-6169 (United States); Bonoli, P. T.; Wright, J. C. [MIT Plasma Science and Fusion Center, Cambridge, MA 02139 (United States); Pinsker, R. I.; Prater, R. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Qin, C. M. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); and others

2015-12-10

Several experiments on different machines and in different fast wave (FW) heating regimes, such as hydrogen minority heating and high harmonic fast waves, have found strong interactions between radio-frequency (RF) waves and the scrape-off layer (SOL) region. This paper examines the propagation and the power loss in the SOL by using the full wave code AORSA, in which the edge plasma beyond the last closed flux surface (LCFS) is included in the solution domain and a collisional damping parameter is used as a proxy to represent the real, and most likely nonlinear, damping processes. 3D AORSA results for the National Spherical Torus eXperiment (NSTX), where a full antenna spectrum is reconstructed, are shown, confirming the same behavior found for a single toroidal mode results in Bertelli et al, Nucl. Fusion, 54 083004, 2014, namely, a strong transition to higher SOL power losses (driven by the RF field) when the FW cut-off is moved away from in front of the antenna by increasing the edge density. Additionally, full wave simulations have been extended to “conventional” tokamaks with higher aspect ratios, such as the DIII-D, Alcator C-Mod, and EAST devices. DIII-D results show similar behavior found in NSTX and NSTX-U, consistent with previous DIII-D experimental observations. In contrast, a different behavior has been found for Alcator C-Mod and EAST, which operate in the minority heating regime unlike NSTX/NSTX-U and DIII-D, which operate in the mid/high harmonic regime. A substantial discussion of some of the main aspects, such as (i) the pitch angle of the magnetic field; (ii) minority heating vs. mid/high harmonic regimes is presented showing the different behavior of the RF field in the SOL region for NSTX-U scenarios with different plasma current. Finally, the preliminary results of the impact of the SOL region on the evaluation of the helicon current drive efficiency in DIII-D is presented for the first time and briefly compared with the different regimes

Dual wavelength imaging of a scrape-off layer in an advanced beam-driven field-reversed configuration

Energy Technology Data Exchange (ETDEWEB)

Osin, D.; Schindler, T., E-mail: dosin@trialphaenergy.com [Tri Alpha Energy, Inc., P.O. Box 7010, Rancho Santa Margarita, California 92688-7010 (United States)

2016-11-15

A dual wavelength imaging system has been developed and installed on C-2U to capture 2D images of a He jet in the Scrape-Off Layer (SOL) of an advanced beam-driven Field-Reversed Configuration (FRC) plasma. The system was designed to optically split two identical images and pass them through 1 nm FWHM filters. Dual wavelength images are focused adjacent on a large format CCD chip and recorded simultaneously with a time resolution down to 10 μs using a gated micro-channel plate. The relatively compact optical system images a 10 cm plasma region with a spatial resolution of 0.2 cm and can be used in a harsh environment with high electro-magnetic noise and high magnetic field. The dual wavelength imaging system provides 2D images of either electron density or temperature by observing spectral line pairs emitted by He jet atoms in the SOL. A large field of view, combined with good space and time resolution of the imaging system, allows visualization of macro-flows in the SOL. First 2D images of the electron density and temperature observed in the SOL of the C-2U FRC are presented.

Effects of ionizing scrape-off layers on local recycling in Tore Supra pump limiter experiments

International Nuclear Information System (INIS)

Chatelier, M.; Loarer, T.

1992-01-01

A series of ohmic discharges with active pumping in the Tore Supra outboard pump limiter has been analyzed with the DEGAS neutrals transport code and an analytic scrape-off layer (SOL) plasma model. Pumping speed and plenum pressure measurements indicated 5-10 torr.L/s throughput with only modest effects on density. A model is developed in which large exhaust fluxes, with little attendant effect on core plasma density, are explained in terms of SOL ionization of recycled and wall-desorbed neutrals. Particle balance with active pumping and constant line density requires that the wall return flux exceed the incident flux by approximately the pump throughput in the absence of external fueling. The radial profile of the H + source rate from ionization and dissociation of wall-desorbed molecules is seen to peak very near the radial position of the limiter throat. Consequently, a strong recycling vortex is created in the region of the limiter, with the ion flux amplified by factors of 2 at the outer limiter surfaces and > 3 within the limiter throat. The calculations indicate that less than 30% of the pump throughput is due to first-generation ions from the core efflux, with the balance from local recycling in the strongly ionizing SOL

Scrape-off layer and divertor theory meeting: Proceedings

International Nuclear Information System (INIS)

1994-03-01

This report contains viewgraphs on the following topics: fluid modelling of neutrals in the SOL and divertor; instabilities of gas-fueled divertors: theory and adaptive simulations; stability of ionization fronts of gaseous divertor plasmas; monte carlo calculation of heat transport; reduced charge model for edge impurity flows; thermally collapsed solutions for gaseous/radiative divertors; adaptive grid methods in transport simulation; advanced numerical solution algorithms applied to the multispecies edge plasma equations; two-dimensional edge plasma simulation using the multigrid method; neutral behavior and the effects of neutral-neutral and neutral-ion elastic scattering in the ITER gaseous divertor; particle throughput in the TPX divertor; marfes in tokamaks; a comparative study of the limiter and divertor edge plasmas in TEXT-U; issues of toroidal tokamak-type divertor simulators; ASDEX upgrade; the ITER divertor; the DIII-D divertor program and TPX divertor; DEGAS 2: a transmission/escape probabilities model for neutral particle transport: comparison with DEGAS 2; a collisional radiative model of hydrogen for high recycling divertors; comparison of fluid and non- fluid neutral models in B2.5; DIII-D radiative divertor simulations; 3-D fluid simulations of turbulence from conducting wall mode; turbulence and drifts in SOL plasmas; recent results for 1 1/2-D ITER gas target divertor modelling; evaluation of pumping and fueling in coupled core, SOL, and divertor chamber calculations; and ITER gas target divertors: comparison of volume recombination and large radial transport scenarios using DEGAS

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

Science.gov (United States)

Colas, L.; Jacquet, Ph.; Van Eester, D.; Bobkov, V.; Brix, M.; Meneses, L.; Tamain, P.; Marsen, S.; Silva, C.; Carralero, D.; Kočan, M.; Müller, H.-W.; Crombé, K.; Křivska, A.; Goniche, M.; Lerche, E.; Rimini, F. G.; JET-EFDA Contributors

2015-08-01

Combining Lithium beam emission spectroscopy and edge reflectometry, localized Scrape-Off Layer (SOL) density modifications by Ion Cyclotron Range of Frequencies (ICRF) near fields were characterized in JET L-mode plasmas. When using the ICRF wave launchers connected magnetically to the Li-beam chord, the density decreased more steeply 2-3 cm outside the last closed flux surface (mapped onto the outer mid-plane) and its value at the outer limiter radial position was half the ohmic value. The depletion depends on the ICRF power and on the phasing between adjacent radiating straps. Convection due to ponderomotive effects and/or E × B0 drifts is suspected: during ICRF-heated H-mode discharges in 2013, DC potentials up to 70 V were measured locally in the outer SOL by a floating reciprocating probe, located toroidally several metres from the active antennas. These observations are compared with probe measurements on ASDEX-Upgrade. Their implications for wave coupling, heat loads and impurity production are discussed.

Plasma particle sources due to interactions with neutrals in a turbulent scrape-off layer of a toroidally confined plasma

Science.gov (United States)

Thrysøe, A. S.; Løiten, M.; Madsen, J.; Naulin, V.; Nielsen, A. H.; Rasmussen, J. Juul

2018-03-01

The conditions in the edge and scrape-off layer (SOL) of magnetically confined plasmas determine the overall performance of the device, and it is of great importance to study and understand the mechanics that drive transport in those regions. If a significant amount of neutral molecules and atoms is present in the edge and SOL regions, those will influence the plasma parameters and thus the plasma confinement. In this paper, it is displayed how neutrals, described by a fluid model, introduce source terms in a plasma drift-fluid model due to inelastic collisions. The resulting source terms are included in a four-field drift-fluid model, and it is shown how an increasing neutral particle density in the edge and SOL regions influences the plasma particle transport across the last-closed-flux-surface. It is found that an appropriate gas puffing rate allows for the edge density in the simulation to be self-consistently maintained due to ionization of neutrals in the confined region.

Drift-based scrape-off particle width in X-point geometry

Science.gov (United States)

Reiser, D.; Eich, T.

2017-04-01

The Goldston heuristic estimate of the scrape-off layer width (Goldston 2012 Nucl. Fusion 52 013009) is reconsidered using a fluid description for the plasma dynamics. The basic ingredient is the inclusion of a compressible diamagnetic drift for the particle cross field transport. Instead of testing the heuristic model in a sophisticated numerical simulation including several physical mechanisms working together, the purpose of this work is to point out basic consequences for a drift-dominated cross field transport using a reduced fluid model. To evaluate the model equations and prepare them for subsequent numerical solution a specific analytical model for 2D magnetic field configurations with X-points is employed. In a first step parameter scans in high-resolution grids for isothermal plasmas are done to assess the basic formulas of the heuristic model with respect to the functional dependence of the scrape-off width on the poloidal magnetic field and plasma temperature. Particular features in the 2D-fluid calculations—especially the appearance of supersonic parallel flows and shock wave like bifurcational jumps—are discussed and can be understood partly in the framework of a reduced 1D model. The resulting semi-analytical findings might give hints for experimental proof and implementation in more elaborated fluid simulations.

Fast electron flux driven by lower hybrid wave in the scrape-off layer

International Nuclear Information System (INIS)

Li, Y. L.; Xu, G. S.; Wang, H. Q.; Wan, B. N.; Chen, R.; Wang, L.; Gan, K. F.; Yang, J. H.; Zhang, X. J.; Liu, S. C.; Li, M. H.; Ding, S.; Yan, N.; Zhang, W.; Hu, G. H.; Liu, Y. L.; Shao, L. M.; Li, J.; Chen, L.; Zhao, N.

2015-01-01

The fast electron flux driven by Lower Hybrid Wave (LHW) in the scrape-off layer (SOL) in EAST is analyzed both theoretically and experimentally. The five bright belts flowing along the magnetic field lines in the SOL and hot spots at LHW guard limiters observed by charge coupled device and infrared cameras are attributed to the fast electron flux, which is directly measured by retarding field analyzers (RFA). The current carried by the fast electron flux, ranging from 400 to 6000 A/m 2 and in the direction opposite to the plasma current, is scanned along the radial direction from the limiter surface to the position about 25 mm beyond the limiter. The measured fast electron flux is attributed to the high parallel wave refractive index n || components of LHW. According to the antenna structure and the LHW power absorbed by plasma, a broad parallel electric field spectrum of incident wave from the antennas is estimated. The radial distribution of LHW-driven current density is analyzed in SOL based on Landau damping of the LHW. The analytical results support the RFA measurements, showing a certain level of consistency. In addition, the deposition profile of the LHW power density in SOL is also calculated utilizing this simple model. This study provides some fundamental insight into the heating and current drive effects induced by LHW in SOL, and should also help to interpret the observations and related numerical analyses of the behaviors of bright belts and hot spots induced by LHW

Interaction of stochastic boundary layer with plasma facing components

International Nuclear Information System (INIS)

Nguyen, F.; Ghendrih, P.; Grosman, A.

1997-01-01

To alleviate the plasma-wall interaction problems in magnetic confinement devices, a stochastic layer is used at the edge of the Tore Supra tokamak (ergodic divertor). A very important point is to determine the power deposition on the plasma facing components. Two different kinds of transport can be identified in such a configuration: Stochastic transport surrounding the confined plasma, with a random walk process, and scrape-off layer (SOL) like transport, a laminar transport, near the plasma facing components. The laminar regime is investigated in terms of a simple criterion, namely that the power deposition is proportional to the radial penetration of the laminar zone flux tubes over a finite parallel length. The magnetic connection properties of the first wall components are then determined. The connection lengths are quantified with two characteristic scales. The larger corresponds to one poloidal turn and appears to be the characteristic parallel length for laminar transport. A field line tracing code MASTOC (magnetic stochastic configuration) is used to computer the complex topology and the statistics of the connection in the real tokamak geometry. The numerical simulations are then compared with the experimental heat deposition on the modules and neutralizer plates of the Tore Supra ergodic divertor. Good agreement is found. Further evidence of laminar transport is also provided by the tangential view of such structures revealed from H α structures in detached plasma experiments. (author). 27 refs, 14 figs

Modification of the internal electric field by biasing of the divertor plates in the Tokamak de Varennes (TdeV)

International Nuclear Information System (INIS)

Lafrance, D.; Huang, R.; Stansfield, B.L.; Haddad, E.; Lachambre, J.

1997-01-01

The radial electric field inside the separatrix has been deduced from spectroscopic measurements of impurities on TdeV (Tokamak de Varennes), using the reduced radial momentum balance and two neoclassical models [R. D. Hazeltine, Phys. Fluids 17, 961 (1974) and Y. B. Kim, P. H. Diamond, and R. J. Groebner, Phys. Fluids B 3, 2050 (1991)]. The results from all three models are in fair agreement. Furthermore, the electric field has been deduced using the same models both with and without biasing the divertor plates relative to the machine wall, showing an inward propagation of the effect of the biasing created in the scrape-off layer (SOL). Undeniably, the electric field has been modified well inside the separatrix (0.6 approx-lt r/a approx-lt 0.9), revealing the possibility of modifying the internal electric field by external means. copyright 1997 American Institute of Physics

Observation of Blobs and Holes in the Boundary Plasma of EAST Tokamak

DEFF Research Database (Denmark)

Yan, Ning; Xu, Guosheng; Zhang, Wei

2011-01-01

Intermittent convective transport at the edge and in the scrape-off layer (SOL) of EAST was investigated by using fast reciprocating Langmuir probe. Holes, as part of plasma structures, were detected for the first time inside the shear layer. The amplitude probability distribution function...

Divertor impurity injection using high voltage arcs for impurity transport studies on the Mega Amp Spherical Tokamak

International Nuclear Information System (INIS)

Leggate, H. J.; Turner, M. M.; Lisgo, S. W.; Harrison, J. R.; Elmore, S.; Allan, S. Y.; Gaffka, R. C.; Stephen, R. C.

2014-01-01

The operation of next-generation fusion reactors will be significantly affected by impurity transport in the scrape-off layer (SOL). Current modelling efforts are restricted by a lack of detailed data on impurity transport in the SOL. In order to address this, a carbon injector has been designed and installed on the Mega Amp Spherical Tokamak (MAST). The injector creates short lived carbon plumes originating at the MAST divertor lasting less than 50 μs. High voltage capacitor banks are used to create a discharge across concentric carbon electrodes located in a probe mounted on the Divertor Science Facility in the MAST lower divertor. This results in a very short plume duration allowing observation of the evolution of the plume and precise localisation of the plume relative to the X-point on MAST. The emission from the carbon plume was imaged using fast visible cameras filtered in order to isolate the carbon II and carbon III emission lines centered around 514 nm and 465 nm

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

DEFF Research Database (Denmark)

Militello, F; Fundamenski, W; Naulin, Volker

2012-01-01

The L-mode interchange turbulence in the edge and scrape-off-layer (SOL) of the tight aspect ratio tokamak MAST is investigated numerically. The dynamics of the boundary plasma are studied using the 2D drift-fluid code ESEL, which has previously shown good agreement with large aspect ratio machin...

Combined Langmuir-magnetic probe measurements of type-I ELMy filaments in the EAST tokamak

Science.gov (United States)

Qingquan, YANG; Fangchuan, ZHONG; Guosheng, XU; Ning, YAN; Liang, CHEN; Xiang, LIU; Yong, LIU; Liang, WANG; Zhendong, YANG; Yifeng, WANG; Yang, YE; Heng, ZHANG; Xiaoliang, Li

2018-06-01

Detailed investigations on the filamentary structures associated with the type-I edge-localized modes (ELMs) should be helpful for protecting the materials of a plasma-facing wall on a future large device. Related experiments have been carefully conducted in the Experimental Advanced Superconducting Tokamak (EAST) using combined Langmuir-magnetic probes. The experimental results indicate that the radially outward velocity of type-I ELMy filaments can be up to 1.7 km s‑1 in the far scrape-off layer (SOL) region. It is remarkable that the electron temperature of these filaments is detected to be ∼50 eV, corresponding to a fraction of 1/6 to the temperature near the pedestal top, while the density (∼ 3× {10}19 {{{m}}}-3) of these filaments could be approximate to the line-averaged density. In addition, associated magnetic fluctuations have been clearly observed at the same time, which show good agreement with the density perturbations. A localized current on the order of ∼100 kA could be estimated within the filaments.

Practical aspects of a 2-D edge-plasma model

International Nuclear Information System (INIS)

Rensink, M.E.; Hill, D.N.; Porter, G.D.; Braams, B.J.; Princeton Univ., NJ

1989-07-01

The poloidal divertor configuration is considered the most promising solution to the particle and energy exhaust problem for a tokamak reactor. The scrape-off layer plasma surrounding the core and the high-recycling plasma near the divertor plates can be modelled by fluid equations for particle, momentum and energy transport. A numerical code (B2) based on a two-dimensional multi-fluid model has been developed for the study of edge plasmas in tokamaks. In this report we identify some key features of this model as applied to the DIII-D tokamak. 2 refs., 1 fig

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)

Characterizing electrostatic turbulence in tokamak plasmas with high MHD activity

Energy Technology Data Exchange (ETDEWEB)

Guimaraes-Filho, Z O; Santos Lima, G Z dos; Caldas, I L; Nascimento, I C; Kuznetsov, Yu K [Instituto de Fisica, Universidade de Sao Paulo, Caixa Postal 66316, 05315-970, Sao Paulo, SP (Brazil); Viana, R L, E-mail: viana@fisica.ufpr.b [Departamento de Fisica, Universidade Federal do Parana, Caixa Postal 19044, 81531-990, Curitiba, PR (Brazil)

2010-09-01

One of the challenges in obtaining long lasting magnetic confinement of fusion plasmas in tokamaks is to control electrostatic turbulence near the vessel wall. A necessary step towards achieving this goal is to characterize the turbulence level and so as to quantify its effect on the transport of energy and particles of the plasma. In this paper we present experimental results on the characterization of electrostatic turbulence in Tokamak Chauffage Alfven Bresilien (TCABR), operating in the Institute of Physics of University of Sao Paulo, Brazil. In particular, we investigate the effect of certain magnetic field fluctuations, due to magnetohydrodynamical (MHD) instabilities activity, on the spectral properties of electrostatic turbulence at plasma edge. In some TCABR discharges we observe that this MHD activity may increase spontaneously, following changes in the edge safety factor, or after changes in the radial electric field achieved by electrode biasing. During the high MHD activity, the magnetic oscillations and the plasma edge electrostatic turbulence present several common linear spectral features with a noticeable dominant peak in the same frequency. In this article, dynamical analyses were applied to find other alterations on turbulence characteristics due to the MHD activity and turbulence enhancement. A recurrence quantification analysis shows that the turbulence determinism radial profile is substantially changed, becoming more radially uniform, during the high MHD activity. Moreover, the bicoherence spectra of these two kinds of fluctuations are similar and present high bicoherence levels associated with the MHD frequency. In contrast with the bicoherence spectral changes, that are radially localized at the plasma edge, the turbulence recurrence is broadly altered at the plasma edge and the scrape-off layer.

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.

Challenges in edge modeling

International Nuclear Information System (INIS)

Schneider, R.

2007-01-01

Fluid models like B2, UEDGE or EDGE2D are the working horses for scrape-off layer physics, both for design and experimental support. The concept of a numerical tokamak, aiming at a predictive code for ITER, triggers the need to re-assess the available tools and their necessary extensions. These additional physics issues will be summarized from a personal point-of-view. Depending on the specific problem, several complexity levels of scrape-off layer models will be needed. Therefore, a hierarchy of tools is necessary, which will be discussed. Furthermore, the experience existing in other scientific fields with multi-scale problems and modeling should be used. Here, the coupling of different length and time scales are in particular of interest for fusion problems. (author)

Impurity production and transport in the boundary layer of tokamaks

International Nuclear Information System (INIS)

McCracken, G.M.

1987-01-01

The processes by which impurities are produced and enter the discharge are discussed. Emphasis is placed on sputtering at the limiter and an analytical global model is described which incorporates the self-stabilizing effects whch control the edge temperature. Predictions of the scaling of edge temperature and of total radiated power are compared with experimental data from JET and other tokamaks operating with limiters. Under many conditions the scaling of the edge conditions and of the radiated power is accurately predicted. Impurity transport in the boundary and the question of how to control the boundary layer is then discussed. The example of the Impurity Control Limiter on DITE is described. (author)

Fast reciprocating probe system on the HL-2A tokamak

International Nuclear Information System (INIS)

Yan Longwen; Hong Wenyu; Qian Jun; Luo Cuiwen; Pan Li

2005-01-01

A reciprocating probe system has been installed at the midplane of the HL-2A tokamak. The probe is used to measure plasma edge density, temperature, floating potential, and corresponding fluctuation profiles with 8 cm scan from the scrape-off layer to the plasma boundary. The reciprocating probe can move at a speed of 1 m/s. A digital grating displacement measurement system that can provide a high displacement resolution of 0.04 mm is applied to the reciprocating probe system for the first time. A port located behind the vacuum isolation valve is designed for viewing and the exchange of the probe head. Different probe heads can be used to satisfy different experimental requirements. The first probe head had four graphite measurement tips. For high frequency response, no isolation amplifier is used in the electric circuit of the probe measurement. A personal computer via an analog-to-digital digitizer card acquires probe system data, which are sent to a data server by optical fiber after a discharge. All data are sent to the centralized data management system of the HL-2A. In this article we presented the edge temperature and density profiles for the limiter and divertor configurations of a selected plasma discharge

Energy conversion options for ARIES-III - A conceptual D-3He tokamak reactor

International Nuclear Information System (INIS)

Santarius, J.F.; Blanchard, J.P.; Emmert, G.A.; Sviatoslavsky, I.N.; Wittenberg, L.J.; Ghoneim, N.M.; Hasan, M.Z.; Mau, T.K.; Greenspan, E.; Herring, J.S.; Kernbichler, W.; Klein, A.C.; Miley, G.H.; Miller, R.L.; Peng, Y.K.M.

1989-01-01

The potential for highly efficient conversion of fusion power to electricity provides one motivation for investigating D- 3 He fusion reactors. This stems from: (1) the large fraction of D- 3 He power produced in the forms of charged particles and synchrotron radiation which are amenable to direct conversion, and (2) the low neutron fluence and lack of tritium breeding constraints, which increase design flexibility. The design team for a conceptual D- 3 He tokamak reactor, ARIES-III, has investigated numerous energy conversion options at a scoping level in attempting to realize high efficiency. The energy conversion systems have been studied in the context of their use on one or more of three versions of a D- 3 He tokamak: a first stability regime device, a second stability regime device, and a spherical torus. The set of energy conversion options investigated includes bootstrap current conversion, compression-expansion cycles, direct electrodynamic conversion, electrostatic direct conversion, internal electric generator, liquid metal heat engine blanket, liquid metal MHD, plasma MHD, radiation boiler, scrape-off layer thermoelectric, synchrotron radiation conversion by rectennas, synchrotron radiation conversion by thermal cycles, thermionic/AMTEC/thermal systems, and traveling wave conversion. The original set of options is briefly discussed, and those selected for further study are described in more detail. The four selected are liquid metal MHD, plasma MHD, rectenna conversion, and direct electrodynamic conversion. Thermionic energy conversion is being considered, and some options may require a thermal cycle in parallel or series. 17 refs., 3 figs., 1 tab

Plasma equilibrium profiles with applied resonant fields on TBR-1 tokamak

International Nuclear Information System (INIS)

Castro, R.M. de; Heller, M.V.A.P.; Caldas, I.L.; Silva, R.P. da; Brasilio, Z.A.; Oda, G.A.

1995-01-01

In this work we present the measurements of the plasma potential, in the edge and in the scrape-off layer regions of plasma, with and without the presence of the magnetic field perturbations produced by resonant helical windings. (author). 6 refs., 6 figs

Low Z impurity transport in tokamaks

International Nuclear Information System (INIS)

Hawryluk, R.J.; Suckewer, S.; Hirshman, S.P.

1978-10-01

Low Z impurity transport in tokamaks was simulated with a one-dimensional impurity transport model including both neoclassical and anomalous transport. The neoclassical fluxes are due to collisions between the background plasma and impurity ions as well as collisions between the various ionization states. The evaluation of the neoclassical fluxes takes into account the different collisionality regimes of the background plasma and the impurity ions. A limiter scrapeoff model is used to define the boundary conditions for the impurity ions in the plasma periphery. In order to account for the spectroscopic measurements of power radiated by the lower ionization states, fluxes due to anomalous transport are included. The sensitivity of the results to uncertainties in rate coefficients and plasma parameters in the periphery are investigated. The implications of the transport model for spectroscopic evaluation of impurity concentrations, impurity fluxes, and radiated power from line emission measurements are discussed

A mechanism for large divertor plasma energy loss via lithium radiation in tokamaks

Science.gov (United States)

Rognlien, T. D.; Meier, E. T.; Soukhanovskii, V. A.

2012-10-01

Lithium has been used as a wall-conditioning element in a number of tokamaks over the years, including TFTR, FTU, and NSTX, where core plasma energy confinement and particle control are often found to improve following such conditioning. Here the possible role of Li in providing substantial energy loss for divertor plasmas via line radiation is reported. A multi-charge-state 2D UEDGE fluid model is used where the hydrogenic and Li ions and neutrals are each evolved as separate species and separate equations are solved for the electron and ion temperatures. It is shown that a sufficient level of Li neutrals evolving from the divertor surface via sputtering or evaporation can induce energy detachment of the divertor plasma, yielding a strongly radiating zone near the divertor where ionization and recombination from/to neutral Li can radiate most of the power flowing into the scrape-off layer while maintaining low core contamination. A local peaking of Li emissivity for electron temperatures near 1 eV appears to play an important role in the detachment of the mixed deuterium/Li plasma. Evidence of such behavior from NSTX discharges will be discussed.

Ion transport in turbulent edge plasmas

International Nuclear Information System (INIS)

Helander, P.; Massachusetts Inst. of Tech., Cambridge, MA; Hazeltine, R.D.; Catto, P.J.

1996-02-01

Edge plasmas, such as the tokamak scrape-off layer, exist as a consequence of a balance between cross-field diffusion and parallel losses. The former is usually anomalous, and is widely thought to be driven by strong electrostatic turbulence. It is shown that the anomalous diffusion affects the parallel ion transport by giving rise to a new type of thermal force between different ion species. This force is parallel to the magnetic field, but arises entirely because of perpendicular gradients, and could be important for impurity retention in the tokamak divertor. (author)

Lyapunov spectra of density fluctuations in TBR-1

International Nuclear Information System (INIS)

Oiwa, N.N.; Fidler-Ferrara, N.

1993-01-01

The results for the Lyapunov exponents associated with density fluctuations measured by Langmuir probes placed in the scrape-off layer of the Tokamak TBR-1 are reported. By a judicious use of the Sano-Sawada and Eckmann-Ruelle algorithms conclusive values for the positive Lyapunov exponents for most of the analysed signals are used showing evidences of chaotic behavior. (author)

The design of central column protection tiles for the TCV tokamak

International Nuclear Information System (INIS)

Pitts, R.A.; Chavan, R.; Moret, J.M.

1999-01-01

The large variety of plasma shapes produced in the TCV tokamak places unique demands on the plasma facing surfaces. In particular, the central column graphite armour tiles are solicited during the creation of all TCV plasmas and function as power handling surfaces for both limited and diverted discharges. The higher power flux densities accompanying the addition of electron cyclotron heating systems have necessitated a new, optimized, design for these tiles. The optimization process and the subsequent new tile design are described. A basic 'two point' model of the scrape-off layer plasma in conjunction with TCV equilibrium reconstructions and a simplified representation of the local magnetic field line geometry are used to impose simulated power flux densities onto a parametric toroidal tile contour. The thermo-mechanical response of the tile is then investigated via full 3-D finite element simulations accounting for the non-linear temperature dependence of the graphite thermal diffusivity and radiation from the tile surface. The final design choice is a compromise between the requirements for adequate power handling for a range of magnetic configurations, the need to protect against tile edge misalignment in the presence of grazing field line angles of incidence and the space restrictions imposed by vacuum vessel design. (author)

Metal droplet erosion and shielding plasma layer under plasma flows typical of transient processes in tokamaks

Energy Technology Data Exchange (ETDEWEB)

Martynenko, Yu. V., E-mail: Martynenko-YV@nrcki.ru [National Research Nuclear University “MEPhI” (Russian Federation)

2017-03-15

It is shown that the shielding plasma layer and metal droplet erosion in tokamaks are closely interrelated, because shielding plasma forms from the evaporated metal droplets, while droplet erosion is caused by the shielding plasma flow over the melted metal surface. Analysis of experimental data and theoretical models of these processes is presented.

Measurement of inner wall limiter SOL widths in KSTAR tokamak

Directory of Open Access Journals (Sweden)

J.G. Bak

2017-08-01

Full Text Available Scrape-off layer (SOL widths λq are presented from the KSTAR tokamak using fast reciprocating Langmuir probe assembly (FRLPA measurements at the outboard mid-plane (OMP and the infra-Red (IR thermography at inboard limiter tiles in moderately elongated (κ = 1.45 – 1.55 L-mode inner wall-limited (IWL plasmas under experimental conditions such as BT = 2.0 T, PNBI = 1.4 – 1.5 MW, line averaged densities 2.5 – 5.1 × 1019 m−3 and plasma current Ip = 0.4 − 0.7 MA. There is clear evidence for a double exponential structure in q||(r from the FRLPA such that, for example at Ip = 0.6 MA, a narrow feature, λq,near (=3.5 mm is found close to the LFCS, followed by a broader width, λq,main (=57.0 mm. Double exponential profiles (λq,near = 1.5 – 2.8 mm, λq,main = 17.0 – 35.0 mm can be also observed in the IR heat flux mapped to the OMP throughout the range of Ip investigated. In addition, analysis of SOL turbulence statistics obtained with the FRLPA shows high relative fluctuation levels and positively skewed distributions in electron temperature and ion particle flux across the SOL, with both properties increasing for longer distance from the LCFS, as often previously observed in the tokamaks. Interestingly, the fluctuation character expressed in terms of spectral distributions remains unchanged in passing from the narrow to the broad SOL heat flux channel.

Study of heat flux deposition in the Tore Supra Tokamak

International Nuclear Information System (INIS)

Carpentier, S.

2009-02-01

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 λ 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 λ 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)

Lower hybrid experiments in the PETULA Tokamak

International Nuclear Information System (INIS)

Melin, G.; Bardet, R.; Bernard, M.

1979-01-01

The present paper deals with the application of high RF power near the lower hybrid (LH) frequency to a Tokamak discharge. A double waveguide grill antenna has been used to launch a 1.25 GHz-500kW-6 ms RF pulse in PETULA. One of the objectives of this study is to investigate the technological problems associated with LH heating and caused by the high level of RF power involved. Two aspects have been carefully studied: grill conditioning and grill coupling. The latter is strongly dependent on the scrape-off layer of the PETULA plasma and the observations are possibly not significant. The former aspect though not really understood is probably to be expected in other LH experiments. Plasma experiments have been carried out with a wide range of electron density 2. 13 cm -3 in both hydrogen and deuterium. For hydrogen this allows theoretically linear mode conversion to take place in the discharge. Although some heating has been observed, the experiments are not, as yet, very conclusive. The actual RF pulse duration is hardly larger than the ion energy confinement time, and the radial plasma displacement caused by the RF is not feedback controlled. The main discharge parameters were the following: plasma radius 14 cm, toroidal magnetic field 27 kG, plasma current 75 kA, loop voltage 1.6 V, Zsub(eff) < 2., 800 < Tsub(e)(o)< 1100eV, 200 < Tsub(i)(o)< 350 eV

Fast reciprocating probe system for local scrape-off layer measurements in front of the lower hybrid launcher on JT-60U

International Nuclear Information System (INIS)

Asakura, N.; Tsuji-Iio, S.; Ikeda, Y.; Neyatani, Y.; Seki, M.

1995-01-01

A fast reciprocating probe system with a long drive shaft was incorporated into a multi-junction lower hybrid (LH) wave launcher on JT-60U in order to investigate an improved coupling mechanism of the radio frequency wave to the core plasma. The system has been operated reliably over a horizontal scan of 25 cm in 1.5 s using a compact pneumatic cylinder drive and springs. A double probe measurement provided the scrape-off layer plasma profile between the last closed flux surface and the first wall with the spatial resolution of 1-2 mm measured with a laser displacement gauge. The profiles of the electron density n e and temperature T e were in good agreement with those obtained with a triple probe method. During the LH wave injection with good coupling to the core plasma, an increase in the local T e was observed in front of the LH launcher mouth. The local n e was (7-10)x10 16 m -3 , consistent values needed for the good coupling. copyright 1995 American Institute of Physics

Transport-driven scrape-off layer flows and the x-point dependence of the L-H power threshold in Alcator C-Moda)

Science.gov (United States)

LaBombard, B.; Rice, J. E.; Hubbard, A. E.; Hughes, J. W.; Greenwald, M.; Granetz, R. S.; Irby, J. H.; Lin, Y.; Lipschultz, B.; Marmar, E. S.; Marr, K.; Mossessian, D.; Parker, R.; Rowan, W.; Smick, N.; Snipes, J. A.; Terry, J. L.; Wolfe, S. M.; Wukitch, S. J.

2005-05-01

Factor of ˜2 higher power thresholds for low- to high-confinement mode transitions (L-H) with unfavorable x-point topologies in Alcator C-Mod [Phys. Plasmas 1, 1511 (1994)] are linked to flow boundary conditions imposed by the scrape-off layer (SOL). Ballooning-like transport drives flow along magnetic field lines from low- to high-field regions with toroidal direction dependent on upper/lower x-point balance; the toroidal rotation of the confined plasma responds, exhibiting a strong counter-current rotation when B ×∇B points away from the x point. Increased auxiliary heating power (rf, no momentum input) leads to an L-H transition at approximately twice the edge electron pressure gradient when B ×∇B points away. As gradients rise prior to the transition, toroidal rotation ramps toward the co-current direction; the H mode is seen when the counter-current rotation imposed by the SOL flow becomes compensated. Remarkably, L-H thresholds in lower-limited discharges are identical to lower x-point discharges; SOL flows are also found similar, suggesting a connection.

Edge Plasma Response to Non-Axisymmetric Fields in Tokamaks

Energy Technology Data Exchange (ETDEWEB)

Ferraro, N. M.; Lao, L. L.; Buttery, R. J.; Evans, T. E.; Snyder, P. B.; Wade, M.R., E-mail: ferraro@fusion.gat.com [General Atomics, San Diego (United States); Moyer, R. A.; Orlov, D. M. [University of California San Diego, La Jolla (United States); Lanctot, M. J. [Lawrence Livermore National Laboratory, Livermore (United States)

2012-09-15

Full text: The application of non-axisymmetric fields is found to have significant effects on the transport and stability of H-mode tokamak plasmas. These effects include dramatic changes in rotation and particle transport, and may lead to the partial or complete suppression of edge-localized modes (ELMs) under some circumstances. The physical mechanism underlying these effects is presently not well understood, in large part because the response of the plasma to non- axisymmetric fields is significant and complex. Here, recent advances in modeling the plasma response to non-axisymmetric fields are discussed. Calculations using a resistive two-fluid model in diverted toroidal geometry confirm the special role of the perpendicular electron velocity in suppressing the formation of islands in the plasma. The possibility that islands form near the top of the pedestal, where the zero-crossing of the perpendicular electron velocity may coincide with a mode-rational surface, is explored, and the implications for ELM suppression are discussed. Modeling results are compared with empirical data. It is shown that numerical modeling is successful in reproducing some experimentally observed effects of applied non-axisymmetric fields on the edge temperature and density profiles. The numerical model self-consistently includes the plasma, separatrix, and scrape-off layer. Rotation and diamagnetic effects are also included self-consistently. Solutions are calculated using the M3D-C1 extended-MHD code. (and others)

Two-zone model for the transport of wall released impurities in the edge plasma of a limiter tokamak

International Nuclear Information System (INIS)

Claassen, H.A.; Gerhauser, H.

1987-02-01

The transmission of a Gaussian metal impurity pulse uniformly injected from the torus wall is studied within a two-zone plasma model, which separates the plasma into scrape-off layer and plasma core and allows for discontinuous changes of the plasma parameters at the separatrix. The plasma parameters are supposed to ensure a collision dominated scrape-off plasma, in which case we may restrict the solution of the transport equations to its zero order approximation. (orig./GG)

Models for impurity production and transport in tokamaks

International Nuclear Information System (INIS)

Hogan, J.T.

1976-01-01

Models for the edge conditions which are commonly used in tokamak transport codes have been kept simple partly because of a lack of data. A report is presented on an improved model for the particle and energy balance of e - , H 1 + , H 1 0 , H 2 + , H 2 0 , O 0 , O/sup (1 yields 8) + / in the plasma scrape-off region. Experiments should yield the needed data in the near future, and allow one to test the model. The diffusion of impurities has been studied with a neoclassical model. The role of 'anomalous spreading' of the impurity distribution has been studied for the case of Fe. A model is presented for the expulsion of low-Z (oxygen) impurities for cases where q(0) greater than 1, but in which a large shear-free region is produced in the plasma core

Particle supply and recovery device

International Nuclear Information System (INIS)

Yamada, Kimio; Okazaki, Takashi.

1988-01-01

This invention concerns a particle supply and recovery device suitable to the supply of fuels and exhaustion of reaction products in a tokamak device. The divertor chamber is divided into an inner side and an outer side, in which only the outer side is constituted as a tightly closed structure. Particles are supplied from the inside of main plasmas and exhausted from the outer side of the divertor chamber. In the divertor equilibrium arrangement, particles escaping from the main plasmas are conveyed mainly passing through the outer side scrape-off layer to the divertor chamber. The particle density is higher at the outer side and lower at the inner side of the scrape-off layer. By making the outer side as a tightly closed structure, the pressure is increased and the particle exhaustion efficiency is improved. Since the particle density is low in the inner scrape-layer, the particle supply efficiency to the main plasma is increased by supplying the particles from the inside. Further, particles ionized in the inner scrape-off layer are returned to the main plasma chamber and then supplied again since the inner divertor chamber is not closed. Accordingly, the particle supply efficiency can further be improved. (K.M.)

Investigations of radial electric field and global circulation layer in limiter tokamaks

International Nuclear Information System (INIS)

Zagorski, R.; Gerhauser, H.; Lehnen, M.; Loarer, T.

2002-01-01

An updated version of the 2D multifluid code TECXY is used to study the radial electric field structure and the appearance of a global circulation layer (GCL) inside the separatrix of the limiter tokamaks TEXTOR-94 and Tore-Supra-CIEL. The dependence of the driving forces on device geometry, limiter position, magnetic field orientation, impurity content and other parameters is investigated. The centrifugal force in the vicinity of the limiter head always determines the direction of the poloidal velocity in the GCL. There is good agreement with experimentally measured profiles of the poloidal velocity at the TEXTOR low field side. (orig.)

Intermittent electron density and temperature fluctuations and associated fluxes in the Alcator C-Mod scrape-off layer

Science.gov (United States)

Kube, R.; Garcia, O. E.; Theodorsen, A.; Brunner, D.; Kuang, A. Q.; LaBombard, B.; Terry, J. L.

2018-06-01

The Alcator C-Mod mirror Langmuir probe system has been used to sample data time series of fluctuating plasma parameters in the outboard mid-plane far scrape-off layer. We present a statistical analysis of one second long time series of electron density, temperature, radial electric drift velocity and the corresponding particle and electron heat fluxes. These are sampled during stationary plasma conditions in an ohmically heated, lower single null diverted discharge. The electron density and temperature are strongly correlated and feature fluctuation statistics similar to the ion saturation current. Both electron density and temperature time series are dominated by intermittent, large-amplitude burst with an exponential distribution of both burst amplitudes and waiting times between them. The characteristic time scale of the large-amplitude bursts is approximately 15 μ {{s}}. Large-amplitude velocity fluctuations feature a slightly faster characteristic time scale and appear at a faster rate than electron density and temperature fluctuations. Describing these time series as a superposition of uncorrelated exponential pulses, we find that probability distribution functions, power spectral densities as well as auto-correlation functions of the data time series agree well with predictions from the stochastic model. The electron particle and heat fluxes present large-amplitude fluctuations. For this low-density plasma, the radial electron heat flux is dominated by convection, that is, correlations of fluctuations in the electron density and radial velocity. Hot and dense blobs contribute only a minute fraction of the total fluctuation driven heat flux.

A helical magnetic limiter for boundary layer control in large tokamaks

International Nuclear Information System (INIS)

Feneberg, W.; Wolf, G.H.

1981-01-01

In a tokamak configuration, superposition of the magnetic field of resonant helical windings which surround the toroidal plasma current outside the first wall destroys the magnetic surfaces in the boundary layer (ergodization). A transport model is analysed, where convective flow of the plasma from the boundary layer to the first wall permits elevated particle densities in the boundary layer and leads to very high particle and energy transport. The convective flow is driven by the pressure gradient along the field lines which intersect the toroidal wall at an oblique small angle epsilon. The required thickness Δ of the boundary layer is around 10 15 n -1 .cm -2 . As a result, the plasma temperature there can be reduced towards the threshold of critical plasma-wall-interaction processes, the plasma core can be shielded against impurities from the wall and, at the same time, a very short life-time of all particles in the boundary layer can be achieved (use of pumpholes and/or scrape-off-limiters for removing ash). Thus, this model also improves the concepts of edge radiation cooling. An estimate is given of the parameters of INTOR using only a weak helical perturbation field which conserves the magnetic surfaces in the plasma core: one can reach wall temperatures Tsub(w) between 20 and 30 eV in the presence of wall densities nsub(w) approaching 10 14 cm -3 . (author)

Narrow heat flux channels in the COMPASS limiter scrape-off layer

Czech Academy of Sciences Publication Activity Database

Horáček, Jan; Vondráček, Petr; Pánek, Radomír; Dejarnac, Renaud; Komm, Michael; Pitts, R.A.; Kocan, M.; Goldston, R.J.; Stangeby, P.C.; Gauthier, E.; Háček, Pavel; Havlíček, Josef; Hron, Martin; Imríšek, Martin; Janky, Filip; Seidl, Jakub

2015-01-01

Roč. 463, August (2015), s. 385-388 ISSN 0022-3115. [PLASMA-SURFACE INTERACTIONS 21: International Conference on Plasma-Surface Interactions in Controlled Fusion Devices. Kanazawa, 26.05.2014-30.05.2014] R&D Projects: GA ČR(CZ) GAP205/12/2327; GA MŠk(CZ) LM2011021 Institutional support: RVO:61389021 Keywords : plasma * tokamak * COMPASS Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 2.199, year: 2015 http://www.sciencedirect.com/science/article/pii/S0022311514009398

Numerical simulation of edge plasma in tokamak

International Nuclear Information System (INIS)

Chen Yiping; Qiu Lijian

1996-02-01

The transport process and transport property of plasma in edge layer of Tokamak are simulated by solving numerically two-dimensional and multi-fluid plasma transport equations using suitable simulation code. The simulation results can show plasma parameter distribution characteristics in the area of edge layer, especially the characteristics near the first wall and divertor target plate. The simulation results play an important role in the design of divertor and first wall of Tokamak. (2 figs)

The Effects of Radial and Poloidal ExB Drifts in the Tokamak SOL

International Nuclear Information System (INIS)

Ou Jing; Zhu Sizheng

2006-01-01

The effects of radial and poloidal ExB drifts in the scrape-off layer (SOL) of a limiter tokamak are studied with a one-dimensional fluid code. The transport equations are solved in the poloidal direction with the radial influxes as the source terms. The simulation results show that in the high recycling regime, the effect of the radial ExB drift on plasma density tends to be stronger than that of the poloidal ExB drift. In the sheath-limited regime, the effects of the radial ExB drift and poloidal ExB drift on plasma density are almost equally important. Considering the influence on the electron temperature, the poloidal ExB drift tends to be more important than the radial ExB drift in both the high recycling regime and sheath-limited regime. For the normal B φ , the poloidal ExB drift tends to raise the pressure at the low field side while the radial ExB drift favours the high field side. The simulation results also show that the ExB drift influences the asymmetries on the parameter distributions at the high field side and low field side, and the distributions are much more symmetric with the field reversal

On lateral deflection of the SOL plasma in tokamaks during giant ELMs

International Nuclear Information System (INIS)

Landman, I.S.; Wuerz, H.

2000-06-01

In recent H-mode experiments at JET with giant ELMs a lateral deflection of hot tokamak plasma leaving the scrape-off layer and striking the divertor plate has been observed. This deflection can effect the divertor erosion caused by the hot plasma irradiation, because of enlarging the irradiated area. A simplified MHD model of the vapor shield plasma and of the hot plasma initially formed at time t → -∞ is analyzed. At t = -∞ both plasmas are assumed to stay on rest and to be separated by a boundary, which is parallel to the plate surface. The interaction between plasmas is assumed to develop gradually ('adiabatically') as exp(t/t 0 ) with t 0 ∝ 10 2 μs the ELM duration time. Electrical insulation of the core tokamak plasma is assumed everywhere except for the contact with the divertor. Electric currents are flowing only in the toroidal direction. These currents developing in the interaction zone of the hot plasma and the rather cold target plasma are calculated for inclined impact of the magnetized hot plasma. At such conditions the J x B force in the lateral direction accelerates the interacting plasmas. The motion of the cold plasma and the gradual increase of the plasma interaction intensity are shown to be important for the appropriate deflection magnitude. Adiabatically responding against the increase of the interaction intensity the cold plasma motion compensates significantly the currents thus decreasing the deflection compared to motionless approach. The calculated magnitude of the hot plasma deflection is comparable to the observed one. The results of the modeling are discussed in relation to the experiments. It is shown that sudden switching on of the interaction produces Alfven oscillations of large amplitudes causing much larger amplitudes of the magnetic field induced by the currents than in the adiabatic case. (orig.)

The optimisation of limiter geometry to reduce impurity influx in tokamaks

International Nuclear Information System (INIS)

Matthews, G.F.; McCracken, G.M.; Sewell, P.; Goodall, D.H.J.; Stangeby, P.C.; Pitcher, C.S.

1987-01-01

Conventional limiters are designed to withstand large power loadings and hence are constructed with surfaces at grazing angles to the toroidal magnetic field. As a result any impurities released from the limiter surface are projected towards the centre of the plasma and are poorly screened from it. The impurity control limiter (ICL), an alternative concept which has an inverted geometry is discussed. The ICL shape is designed to direct the impurities towards the wall. Results are presented from a two-dimensional neutral particle code which maps the ionisation of carbon physically sputtered by deuterons from a carbon limiter. This ionisation source is coupled to a one-dimensional impurity transport code which calculates the implied central impurity density. The results demonstrate that the ICL achieves impurity control in two ways. Firstly, many of the sputtered impurities directed towards the wall are not ionised and return to the wall as neutrals. Secondly, much of the ionisation which does occur is located in the scrape-off layer. Here there is a strong ion sink which may also be enhanced by the flow of hydrogenic ions entraining impurity ions created close to the limiter surface. We conclude that a reduction in central impurity density of a factor of 10 is possible in a Tokamak such as DITE provided that the limiter is the main source of impurities. (author)

Advances in measurement and modeling of the high-confinement-mode pedestal on the Alcator C-Mod tokamak

International Nuclear Information System (INIS)

Hughes, J.W.; LaBombard, B.; Mossessian, D.A.; Hubbard, A.E.; Terry, J.; Biewer, T.

2006-01-01

Edge transport barrier (ETB) studies on the Alcator C-Mod tokamak [Phys. Plasmas 1, 1511 (1994)] investigate pedestal scalings and the radial transport of plasma and neutrals. Pedestal profiles show trends with plasma operational parameters such as total current I P . A ballooning-like I P 2 dependence is seen in the pressure gradient, despite calculated stability to ideal ballooning modes. A similar scaling is seen in the near scrape-off layer for both low-confinement (L-mode) and H-mode discharges, possibly due to electromagnetic fluid drift turbulence setting transport near the separatrix. Neutral density diagnosis allows an examination of D 0 fueling in H-modes, yielding profiles of effective particle diffusivity in the ETB, which vary as I P is changed. Edge neutral transport is studied using a one-dimensional kinetic treatment. In both experiment and modeling, the C-Mod density pedestal exhibits a weakly increasing pedestal density and a nearly invariant density pedestal width as the D 0 source rate increases. Identical modeling performed on pedestal profiles typical of DIII-D [Nucl. Fusion 42, 614 (2002)] reveal differences in pedestal scalings qualitatively similar to experimental results

Sheared flow layer formation in tokamak plasmas with reversed magnetic shear

International Nuclear Information System (INIS)

Dong, J.Q.; Long, Y.X.; Mou, Z.Z.; Zhang, J.H.; Li, J.Q.

2005-01-01

Sheared flow layer (SFL) formation due to magnetic energy release through tearing-reconnections in tokamak plasmas is investigated. The characteristics of the SFLs created in the development of double tearing mode, mediated by electron viscosity in configurations with non-monotonic safety factor q profiles and, therefore, two rational flux surfaces of same q value, are analyzed in detail as an example. Quasi-linear simulations demonstrate that the sheared flows induced by the mode have desirable characteristics (lying at the boundaries of the magnetic islands), and sufficient levels required for internal transport barrier (ITB) formation. A possible correlation of the SFLs with experimental observations, that double transport barrier structures are preferentially formed in proximity of the two rational surfaces, is also proffered. (author)

Low Z impurity transport in tokamaks. [Neoclassical transport theory

Energy Technology Data Exchange (ETDEWEB)

Hawryluk, R.J.; Suckewer, S.; Hirshman, S.P.

1978-10-01

Low Z impurity transport in tokamaks was simulated with a one-dimensional impurity transport model including both neoclassical and anomalous transport. The neoclassical fluxes are due to collisions between the background plasma and impurity ions as well as collisions between the various ionization states. The evaluation of the neoclassical fluxes takes into account the different collisionality regimes of the background plasma and the impurity ions. A limiter scrapeoff model is used to define the boundary conditions for the impurity ions in the plasma periphery. In order to account for the spectroscopic measurements of power radiated by the lower ionization states, fluxes due to anomalous transport are included. The sensitivity of the results to uncertainties in rate coefficients and plasma parameters in the periphery are investigated. The implications of the transport model for spectroscopic evaluation of impurity concentrations, impurity fluxes, and radiated power from line emission measurements are discussed.

A constant heat flux plasma limiter for TEXTOR

International Nuclear Information System (INIS)

Mioduszewski, P.

1980-10-01

In future large tokamak machines heat removal from the plasma is going to play an important role. In TEXTOR the total plasma power is expected to be in the range of 0.5-2.5 MW. Typical fractions of about 50% of this power have to be removed from the plasma by limiters. The power flux from the limiter scrape-off layer to the limiter surface decays rapidly with distance into the scrape-off layer resulting in a highly space-dependent heat load on the limiter. Therefore, limiters are shaped in a way to smooth of the heat load, and the ideal limiter shape should produce a constant heat flux over the whole limiter surface. The ideally shaped limiter offers a better chance to handle the high heat loads with the preferred materials like stainless steel (or inconel 625 as in the case of TEXTOR). (orig./GG)

The implementation of a toroidal limiter model into the gyrokinetic code ELMFIRE

Energy Technology Data Exchange (ETDEWEB)

Leerink, S.; Janhunen, S.J.; Kiviniemi, T.P.; Nora, M. [Euratom-Tekes Association, Helsinki University of Technology (Finland); Heikkinen, J.A. [Euratom-Tekes Association, VTT, P.O. Box 1000, FI-02044 VTT (Finland); Ogando, F. [Universidad Nacional de Educacion a Distancia, Madrid (Spain)

2008-03-15

The ELMFIRE full nonlinear gyrokinetic simulation code has been developed for calculations of plasma evolution and dynamics of turbulence in tokamak geometry. The code is applicable for calculations of strong perturbations in particle distribution function, rapid transients and steep gradients in plasma. Benchmarking against experimental reflectometry data from the FT2 tokamak is being discussed and in this paper a model for comparison and studying poloidal velocity is presented. To make the ELMFIRE code suitable for scrape-off layer simulations a simplified toroidal limiter model has been implemented. The model is be discussed and first results are presented. (copyright 2008 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

Understanding SOL plasma turbulence by interchange motions

Czech Academy of Sciences Publication Activity Database

Horáček, Jan; Pitts, R. A.; Nielsen, A.H.; Garcia, O.E.

2007-01-01

Roč. 52, č. 16 (2007), s. 192-193 ISSN 0003-0503. [Annual meeting of the division of plasma physics/49th./. Orlando , 12.11.2007-16.11.2007] Grant - others:-(XE) European Training fellowships and Grants (Euratom), EDGETURB Institutional research plan: CEZ:AV0Z20430508 Keywords : tokamak * plasma * scrape-off layer * turbulence * interchange instability Subject RIV: BL - Plasma and Gas Discharge Physics http://meetings.aps.org/Meeting/DPP07/Event/70125

Spectroscopic investigation of ELM phenomena in the ASDEX-Upgrade divertor with high time resolution

International Nuclear Information System (INIS)

Field, A.R.; Buechl, K.; Fuchs, C.J.; Fussmann, G.; Herrmann, A.; Lieder, G.; Napiontek, B.; Radtke, R.; Wenzel, U.; Zohm, H.

1993-01-01

Improved tokamak H-mode confinement is associated with the formation of an insulating zone just within the separatrix. At a critical pressure gradient a sudden burst of MHD activity (an ELM) degrades edge confinement, releasing particles and energy into the scrape-off layer (SOL) which is subsequently transported to the divertor. Here, these phenomena are studied using spectroscopic diagnostics and target plate thermography of high spatial and temporal resolution. (author) 3 refs., 6 figs

Spectroscopic investigation of ELM phenomena in the ASDEX-Upgrade divertor with high time resolution

Energy Technology Data Exchange (ETDEWEB)

Field, A R; Buechl, K; Fuchs, C J; Fussmann, G; Herrmann, A; Lieder, G; Napiontek, B; Radtke, R; Wenzel, U; Zohm, H [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

1994-12-31

Improved tokamak H-mode confinement is associated with the formation of an insulating zone just within the separatrix. At a critical pressure gradient a sudden burst of MHD activity (an ELM) degrades edge confinement, releasing particles and energy into the scrape-off layer (SOL) which is subsequently transported to the divertor. Here, these phenomena are studied using spectroscopic diagnostics and target plate thermography of high spatial and temporal resolution. (author) 3 refs., 6 figs.

Progress in diagnostics of the COMPASS tokamak

Science.gov (United States)

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

2017-12-01

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

Comparing scrape-off layer and divertor physics in JET pure He and D discharges

NARCIS (Netherlands)

Pitts, R.A.; Andrew, P.; Andrew, Y.; Becoulet, M.; Coffey, I.; Coster, D.; McDonald, D. C.; Eich, T.; Erents, S.K.; Fenstermacher, M.E.; Fundamenski, W.; Haas, G.; Hermann, A.; Hidalgo, C.; Hillis, D.; Huber, A.; Ingesson, L. C.; Jachmich, S.; Kallenbach, A.; Korotkov, A.; Lawson, K.; Lomas, P.; Loarer, T.; Loarte, A.; Matthews, G. F.; McCracken, G.; Meigs, A.; Mertens, P.; O' Mullane, M.; Phillipps, V.; Porter, G.; Pospieszczyk, A.; Rapp, J.; Reiter, D.; Riccardo, V.; Saibene, G.; Sartori, R.; Stamp, M. F.; Tsitrone, E.; Wischmeier, W.; Gafert, J.

2003-01-01

Though helium plasmas are one option for the low activation phase of ITER, little effort has thus far been devoted to studying them in a large, diverted tokamak. A recent campaign on JET has therefore sought to address some of the important questions related to helium operation (He concentrations

Magnetic- and material-limiter discharges in Tokapole II

International Nuclear Information System (INIS)

Moyer, R.A.

1988-01-01

Disruptive instabilities were studied in Tokapole II, a small poloidal-divertor tokamak, in magnetic- and material-limiter configurations. In the magnetic limiter configuration, the divertor separatrix defines the tokamak current channel boundary. Limiters or neutralizer plate are not used to remove plasma in the scrape-off region. The relatively hot, dense plasma in the scrape-off region carries 5-20% of the current. In the material-limiter configuration, limiter plates are inserted to the separatrix to remove plasma and current in the scrape-off region. The plates vary the tokamak current-channel boundary condition in a controlled manner, and provide a benchmark for comparison with other tokamaks. Internal and external disruptions have been studied, and several unique features in the magnetic-limiter configuration were identified. The magnetic-limiter configuration enables routine passing of the stability barriers at q(a) = 2 and q(a) = 1, where q(a) is the edge safety factor, without a close-fitting wall, external windings, or detailed profile control techniques. Passing the q(a) = 1 barrier permits operation in the q < 1 regime where total reconnection of the sawtooth does not occur

A universal access layer for the Integrated Tokamak Modelling Task Force

International Nuclear Information System (INIS)

Manduchi, G.; Iannone, F.; Imbeaux, F.; Huysmans, G.; Lister, J.B.; Guillerminet, B.; Strand, P.; Eriksson, L.-G.; Romanelli, M.

2008-01-01

The Integrated Tokamak Modelling (ITM) Task Force aims at providing a suite of codes for preparing and analyzing future ITER discharges. In the framework of the ITM, the universal access layer (UAL) provides the capability of storing and retrieving data involved in simulation. The underlying data structure is hierarchical and the granularity in data access is given by the definition of a set of consistent physical objects (CPOs). To describe the data structure of the overall ITM database, the XML schema description (XSD) has been used. Originally intended to describe the structure of XML documents, XSD is used here to provide an unambiguous way of describing how data are structured, regardless of the actual implementation of the underlying database. The MDSplus-based UAL implementation is currently under test and other prototypes for investigating alternative data storage systems are foreseen

An Extension of the Miller Equilibrium Model into the X-Point Region

Science.gov (United States)

Hill, M. D.; King, R. W.; Stacey, W. M.

2017-10-01

The Miller equilibrium model has been extended to better model the flux surfaces in the outer region of the plasma and scrape-off layer, including the poloidally non-uniform flux surface expansion that occurs in the X-point region(s) of diverted tokamaks. Equations for elongation and triangularity are modified to include a poloidally varying component and grad-r, which is used in the calculation of the poloidal magnetic field, is rederived. Initial results suggest that strong quantitative agreement with experimental flux surface reconstructions and strong qualitative agreement with poloidal magnetic fields can be obtained using this model. Applications are discussed. A major new application is the automatic generation of the computation mesh in the plasma edge, scrape-off layer, plenum and divertor regions for use in the GTNEUT neutral particle transport code, enabling this powerful analysis code to be routinely run in experimental analyses. Work supported by US DOE under DE-FC02-04ER54698.

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

Confinement and exhaust in the Mega Ampere Spherical Tokamak

International Nuclear Information System (INIS)

Counsell, G F; Ahn, J-W; Akers, R; Arends, E; Buttery, R; Field, A R; Gryaznevich, M; Helander, P; Kirk, A; Meyer, H; Valovic, M; Wilson, H R; Yang, Y

2002-01-01

The Mega Ampere Spherical Tokamak (MAST) is now accessing regimes with high normalized confinement relative to international scalings, H H (IPB98(y, 2))>1 at high normalized density, n-bar e >60% of the Greenwald density. Data from MAST H-modes suggest that the aspect ratio dependency of international confinement and L-H threshold scalings may need to be modified to improve predictions for ITER. Access to H-mode on MAST is strongly affected by both the divertor magnetic geometry and fuelling location, with the formation of an edge transport barrier being facilitated by operation near the symmetric, connected double-null configuration and with poloidally localized inboard gas puffing. The ELMs on MAST appear to be Type III in nature, even in the highest performance plasmas and with the maximum available auxiliary heating power. ELM energy losses are less than 4% of stored energy in all regimes so far explored. These Type III ELMs are associated with a reduction in the pedestal density but no significant change in the pedestal temperature or temperature profile, indicating that energy is convected from the pedestal region into the scrape-off layer. Analysis of the energy observed to arrive at the divertor targets indicates that ELM losses are predominantly on the low field side. ELM effluxes are observed up to 20 cm from the plasma edge at the outboard mid-plane and are associated with the radial motion of a feature at an average velocity of 1.2 km s -1

GRILLIX. A 3D turbulence code for magnetic fusion devices based on a field line map

International Nuclear Information System (INIS)

Stegmeir, Andreas Korbinian

2015-01-01

The complex geometry in the scrape-off layer of tokamaks poses problems to existing turbulence codes. The usually employed field aligned coordinates become ill defined at the separatrix. Therefore the parallel code GRILLIX was developed, which is based on a field line map. This allows simulations in additional complex geometries, especially across the separatrix. A new discretisation, based on the support operator method, for the highly anisotropic diffusion was developed and applied to a simple turbulence model (Hasegawa-Wakatani).

Tokamak pump limiters

International Nuclear Information System (INIS)

Conn, R.W.

1984-05-01

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

Concepual design of Langmuir probes for the diagnosis of plasma edge of Aditya-U

International Nuclear Information System (INIS)

Lachhvani, Lavkesh T.; Pandya, Shwetang N.; Iyer, Ramakrishnan B.; Barot, Akash; Patel, Kaushal M.; Jadeja, Kumarpalsinh; Gautam, Pramila; Joshi, Nishita H.; Ghosh, Joydeep; Raj, Harshita

2017-01-01

The role of the Tokamak edge plasma in influencing the fusion energy yield of Tokamaks is now widely recognized and is reflected in the increasing efforts devoted to the experimental and theoretical study of scrape-off layer (SOL) physics. Of particular concern are aspects of the plasma-surface interaction leading to impurity production and the subsequent impurity transport and contamination of the core plasma. The impurity transport depends strongly on the background properties of the SOL plasma, such as the plasma density, potential, electron and ion temperature, ion flows, flow velocity and their fluctuations and transport coefficients. The poster discusses the design considerations and technical details for variety of probes installed on Aditya-U

Kinetic modeling of divertor heat load fluxes in the Alcator C-Mod and DIII-D tokamaks

Energy Technology Data Exchange (ETDEWEB)

Pankin, A. Y. [Tech-X Corporation, Boulder, Colorado 80303 (United States); Rafiq, T.; Kritz, A. H. [Department of Physics, Lehigh University, Bethlehem, Pennsylvania 18015 (United States); Park, G. Y. [National Fusion Research Institute, Daejeon, 305-333 (Korea, Republic of); Chang, C. S.; Ku, S. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Brunner, D.; Hughes, J. W.; LaBombard, B.; Terry, J. L. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Groebner, R. J. [General Atomics, San Diego, California 92121 (United States)

2015-09-15

The guiding-center kinetic neoclassical transport code, XGC0 [Chang et al., Phys. Plasmas 11, 2649 (2004)], is used to compute the heat fluxes and the heat-load width in the outer divertor plates of Alcator C-Mod and DIII-D tokamaks. The dependence of the width of heat-load fluxes on neoclassical effects, neutral collisions, and anomalous transport is investigated using the XGC0 code. The XGC0 code includes realistic X-point geometry, a neutral source model, the effects of collisions, and a diffusion model for anomalous transport. It is observed that the width of the XGC0 neoclassical heat-load is approximately inversely proportional to the total plasma current I{sub p.} The scaling of the width of the divertor heat-load with plasma current is examined for an Alcator C-Mod discharge and four DIII-D discharges. The scaling of the divertor heat-load width with plasma current is found to be weaker in the Alcator C-Mod discharge compared to scaling found in the DIII-D discharges. The effect of neutral collisions on the 1/I{sub p} scaling of heat-load width is shown not to be significant. Although inclusion of poloidally uniform anomalous transport results in a deviation from the 1/I{sub p} scaling, the inclusion of the anomalous transport that is driven by ballooning-type instabilities results in recovering the neoclassical 1/I{sub p} scaling. The Bohm or gyro-Bohm scalings of anomalous transport do not strongly affect the dependence of the heat-load width on plasma current. The inclusion of anomalous transport, in general, results in widening the width of neoclassical divertor heat-load and enhances the neoclassical heat-load fluxes on the divertor plates. Understanding heat transport in the tokamak scrape-off layer plasmas is important for strengthening the basis for predicting divertor conditions in ITER.

Impedance of an intense plasma-cathode electron source for tokamak startup

Science.gov (United States)

Hinson, E. T.; Barr, J. L.; Bongard, M. W.; Burke, M. G.; Fonck, R. J.; Perry, J. M.

2016-05-01

An impedance model is formulated and tested for the ˜1 kV , 1 kA/cm2 , arc-plasma cathode electron source used for local helicity injection tokamak startup. A double layer sheath is established between the high-density arc plasma ( narc≈1021 m-3 ) within the electron source, and the less dense external tokamak edge plasma ( nedge≈1018 m-3 ) into which current is injected at the applied injector voltage, Vinj . Experiments on the Pegasus spherical tokamak show that the injected current, Iinj , increases with Vinj according to the standard double layer scaling Iinj˜Vinj3 /2 at low current and transitions to Iinj˜Vinj1 /2 at high currents. In this high current regime, sheath expansion and/or space charge neutralization impose limits on the beam density nb˜Iinj/Vinj1 /2 . For low tokamak edge density nedge and high Iinj , the inferred beam density nb is consistent with the requirement nb≤nedge imposed by space-charge neutralization of the beam in the tokamak edge plasma. At sufficient edge density, nb˜narc is observed, consistent with a limit to nb imposed by expansion of the double layer sheath. These results suggest that narc is a viable control actuator for the source impedance.

Analysis of divertor asymmetry using a simple five-point model

International Nuclear Information System (INIS)

Hayashi, Nobuhiko; Takizuka, Tomonori; Hatayama, Akiyoshi; Ogasawara, Masatada.

1997-03-01

A simple five-point model of the scrape-off layer (SOL) plasma outside the separatrix of a diverted tokamak has been developed to study the inside/outside divertor asymmetry. The SOL current, gas pumping/puffing in the divertor region, and divertor plate biasing are included in this model. Gas pumping/puffing and biasing are shown to control divertor asymmetry. In addition, the SOL current is found to form asymmetric solutions without external controls of gas pumping/puffing and biasing. (author)

Analytical solution for the mode conversion equations with steep exponential density profiles

International Nuclear Information System (INIS)

Alava, M.J.; Heikkinen, J.A.

1992-01-01

A general analytical solution for the converted power from the fast magnetosonic wave to an ion Bernstein wave in a magnetized plasma with an exponential steeply increasing density profile is given in the closed form. The solution covers both the conversion at the lower-hybrid resonance and the conversion through the density gradient for small parallel wave numbers. As an application, the conversion coefficients at the scrape-off layer plasma are estimated in the context of ion cyclotron heating of a tokamak plasma

Integrated modelling of material migration and target plate power handling at JET

International Nuclear Information System (INIS)

Coster, D.P.; Bonnin, X.; Chankin, A.

2005-01-01

The complexity of the tokamak edge and scrape-off layer (SOL) region is such that extrapolation to ITER requires modelling to be pursued through the integration of a number of edge codes, each of which must be thoroughly tested against results from present day machines. This contribution demonstrates how the edge modelling effort at JET is focused on such an approach by considering two examples, target power loading and material erosion and migration, the understanding of which are crucial issues for ITER. (author)

Propagation velocity analysis of a single blob in the SOL

International Nuclear Information System (INIS)

Sugita, Satoru; Yagi, Masatoshi; Itoh, Sanae-I.; Itoh, Kimitaka

2008-01-01

Nonlinear simulation of plasma blob propagation in the tokamak scrape-off layer is reported. Three types of model equations are introduced and the simulation results are compared. It is found that in the parameter regime where the intercharge instability appears during the propagation process, the theoretical model of propagation velocity determined by the initial blob size provides a good approximation of the simulation results. In the regime where the Kelvin-Helmholtz instability appears, however, the blob velocity saturates at a lower value. (author)

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.

Diffusive heat transport across magnetic islands and stochastic layers in tokamaks

International Nuclear Information System (INIS)

Hoelzl, Matthias

2010-01-01

Heat transport in tokamak plasmas with magnetic islands and ergodic field lines was simulated at realistic plasma parameters in realistic tokamak geometries. This requires the treatment of anisotropic heat diffusion, which is more efficient along magnetic field lines by up to ten orders of magnitude than perpendicular to them. Comparisons with analytical predictions and experimental measurements allow to determine the stability properties of neoclassical tearing modes as well as the experimental heat diffusion anisotropy.

Scrape-off profiles and effects of limiter pumping in Tore Supra

International Nuclear Information System (INIS)

Budny, R.

1986-11-01

A one dimensional plasma scrape-off model was used to simulate Tore Supra discharges which are limited by various combinations of the pumped and inner limiters. Scrape-off profiles of the electron density and temperature, ion temperature, and neutral density are given. For each case, various fractions of the ion flux to the neutralizers were assumed to be pumped. Modifications of the scrap-off profiles caused by pumping are predicted. Pumping efficiencies are calculated including the effects of flux amplification caused by recycling. The pumping efficiency is estimated to be 8% for low-power discharges formed on the outer pumped limiter, 7.5% for intermediate-power discharges formed on the seven-module pumped-limiter system, and 5% for full-power discharges formed on both the inner limiter and the pumped-limiter system. The maximum particle removal rate is estimated to be 150 Tl/s

Low-temperature operating regime of the tokamak evacuating limiter

International Nuclear Information System (INIS)

Tokar', M.Z.

1987-01-01

The conditions for realizing the regime of strong recycling of a cold dense plasma of an evacuating limiter were determined based on a previously proposed model for describing the limiter layer of a tokamak. The scaling for the dependence of the gas pressure in the evacuation system on the average plasma density in the limiter layer was found, and agreed quantitatively with the results of measurements on the Alcator and ISX-B tokamaks. For the tokamak reactor of the INTOR scale the calculations show that the low-temperature operating regime of the evacuating limiter can be realized with a quite low pumping rate. It has the advantages of reduced erosion of the limiter and small fluxes of impurities into the working volume of the reactor. In addition, the relative concentration of the helium ash in the limiter layer does not exceed 2-3%, but the density of the main plasma is comparable to the proposed average density in the reactor. The concept of a stochastic limiter is of interest for lowering the plasma density in the limiter layer and lowering the thermal loads on the limiter

Edge gradient and safety factor effects on electrostatic turbulent transport in tokamaks

International Nuclear Information System (INIS)

Tan, Ing Hwie.

1992-05-01

Electrostatic turbulence and transport measurements are performed on the Tokapole-II tokamak at the University of Wisconsin-Madison, as the safety-factor and the edge equilibrium gradients and varied substantially. Tokapole-II is a poloidal divertor tokamak capable of operating at a wide range of safety factors due to its unique magnetic limiter configuration. It also has retractable material limiters in a large scrape-off region, which permits the study of edge boundary conditions like density and temperature gradients. The turbulence is independent of safety factor, but strongly sensitive to the local density gradient, which itself depends upon the limiter configuration. When a material limiter is inserted in a high discharge, the density gradient is increased locally together with a local increase of the turbulence. On the other hand, limiter insertion in low discharges did not increase the density gradient as much and the turbulence properties are unchanged with respect to the magnetic limiter case. It is conducted then, that electrostatic turbulence is caused by the density gradient. Although the electrostatic fluctuation driven transport is enhanced in the large density gradient case, it is in all cases to small to explain the observed energy confinement times. To explore instabilities with small wavelengths, a 0.5 mm diameter shperical Langmuir probe was constructed, and its power compared with the power measured by larger cylindrical probes

Edge transport studies in the edge and scrape-off layer of the National Spherical Torus Experiment with Langmuir probes

Energy Technology Data Exchange (ETDEWEB)

Boedo, J. A., E-mail: jboedo@ucsd.edu; Rudakov, D. L. [University of California San Diego, 9500 Gilman Dr., La Jolla, California 92093 (United States); Myra, J. R.; D' Ippolito, D. A. [Lodestar Research Corp, 2400 Central Ave., Boulder, Colorado 80301 (United States); Zweben, S.; Maingi, R.; Maqueda, R. J.; Bell, R.; Kugel, H.; Leblanc, B.; Roquemore, L. A. [Princeton University, PO Box 451, Princeton, New Jersey 08543 (United States); Soukhanovskii, V. A. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Ahn, J. W.; Canik, J. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, Tennessee 37830 (United States); Crocker, N. [University of California Los Angeles, PO Box 957099, Los Angeles, California 90095 (United States)

2014-04-15

Transport and turbulence profiles were directly evaluated using probes for the first time in the edge and scrape-off layer (SOL) of NSTX [Ono et al., Nucl. Fusion 40, 557 (2000)] in low (L) and high (H) confinement, low power (P{sub in}∼ 1.3 MW), beam-heated, lower single-null discharges. Radial turbulent particle fluxes peak near the last closed flux surface (LCFS) at ≈4×10{sup 21} s{sup −1} in L-mode and are suppressed to ≈0.2×10{sup 21} s{sup −1} in H mode (80%–90% lower) mostly due to a reduction in density fluctuation amplitude and of the phase between density and radial velocity fluctuations. The radial particle fluxes are consistent with particle inventory based on SOLPS fluid modeling. A strong intermittent component is identified. Hot, dense plasma filaments 4–10 cm in diameter, appear first ∼2 cm inside the LCFS at a rate of ∼1×10{sup 21} s{sup −1} and leave that region with radial speeds of ∼3–5 km/s, decaying as they travel through the SOL, while voids travel inward toward the core. Profiles of normalized fluctuations feature levels of 10% inside LCFS to ∼150% at the LCFS and SOL. Once properly normalized, the intermittency in NSTX falls in similar electrostatic instability regimes as seen in other devices. The L-H transition causes a drop in the intermittent filaments velocity, amplitude and number in the SOL, resulting in reduced outward transport away from the edge and a less dense SOL.

Reactor similarity for plasma–material interactions in scaled-down tokamaks as the basis for the Vulcan conceptual design

International Nuclear Information System (INIS)

Whyte, D.G.; Olynyk, G.M.; Barnard, H.S.; Bonoli, P.T.; Bromberg, L.; Garrett, M.L.; Haakonsen, C.B.; Hartwig, Z.S.; Mumgaard, R.T.; Podpaly, Y.A.

2012-01-01

Highlights: ► Discussion of similarity scalings for reduced-size tokamaks. ► Proposal of a new set of scaling laws for divertor similarity. ► Discussion of how the new scaling provides fidelity to a reactor. ► The new scaling is used as the basis for the Vulcan conceptual design. - Abstract: Dimensionless parameter scaling techniques are a powerful tool in the study of complex physical systems, especially in tokamak fusion experiments where the cost of full-size devices is high. It is proposed that dimensionless similarity be used to study in a small-scale device the coupled issues of the scrape-off layer (SOL) plasma, plasma–material interactions (PMI), and the plasma-facing material (PFM) response expected in a tokamak fusion reactor. Complete similarity is not possible in a reduced-size device. In addition, “hard” technological limits on the achievable magnetic field and peak heat flux, as well as the necessity to produce non-inductive scenarios, must be taken into account. A practical approach is advocated, in which the most important dimensionless parameters are matched to a reactor in the reduced-size device, while relaxing those parameters which are far from a threshold in behavior. “Hard” technological limits are avoided, so that the reduced-size device is technologically feasible. A criticism on these grounds is offered of the “P/R” model, in which the ratio of power crossing the last closed flux surface (LCFS), P, to the device major radius, R, is held constant. A new set of scaling rules, referred to as the “P/S” scaling (where S is the LCFS area) or the “PMI” scaling, is proposed: (i) non-inductive, steady-state operation; (ii) P is scaled with R 2 so that LCFS areal power flux P/S is constant; (iii) magnetic field B constant; (iv) geometry (elongation, safety factor q * , etc.) constant; (v) volume-averaged core density scaled as n≈n ¯ e ∼R −2/7 ; and (vi) ambient wall material temperature T W,0 constant. It is

Investigation of electron parallel pressure balance in the scrape-off layer of deuterium-based radiative divertor discharges IN DIII-D

International Nuclear Information System (INIS)

Petrie, T.W.; Carlstrom, T.N.; Allen, S.L.

1996-10-01

Electron density, temperature, and parallel pressure measurements at several locations along field lines connecting the midplane scrapeoff layer (SOL) with the outer divertor are presented for both attached and partially-detached divertor cases: I p = 1.4 MA, q 95 = 4.2, and P input ∼ 6.7 MW under ELMing H-mode conditions. At the onset of the Partially Detached Divertor (PDD), a high density, low temperature plasma forms in the divertor SOL (divertor MARFE). The electron pressure drops by a factor of ∼ 2 between the midplane separatrix and the X-point, and then an additional ∼3--5 times between the X-point and the outboard separatrix strike point. These results are in contrast to the attached (non-PDD) case, where electron pressure in the SOL is reduced by, at most, a factor of two between the midplane and the divertor target. Divertor MARFEs generally have only marginal adverse impact on important H-mode characteristics, such as confinement time. In fact, PDD discharges at low input power maintains good H-mode characteristics until a high density, low temperature plasma abruptly forms inside the separatrix near the X-point (X-point MARFE). Concurrent with the appearance of this X-point MARFE is a degradation in both energy confinement and the plasma fueling rate, and an increase in the carbon impurity concentration inside the core plasma. The formation of the X-point MARFE is consistent with a thermal instability resulting from the temperature dependence of the carbon radiative cooling rate in the range ∼ 7--30 eV

Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

International Nuclear Information System (INIS)

Riedel, K.S.

1989-03-01

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

Effect of impurity radiation on tokamak equilibrium

International Nuclear Information System (INIS)

Rebut, P.H.; Green, B.J.

1977-01-01

The energy loss from a tokamak plasma due to the radiation from impurities is of great importance in the overall energy balance. Taking the temperature dependence of this loss for two impurities characteristic of those present in existing tokamak plasmas, the condition for radial power balance is derived. For the impurities considered (oxygen and iron) it is found that the radiation losses are concentrated in a thin outer layer of the plasma and the equilibrium condition places an upper limit on the plasma paraticle number density in this region. This limiting density scales with mean current density in the same manner as is experimentally observed for the peak number density of tokamak plasmas. The stability of such equilibria is also discussed. (author)

Response to comment on 'Magnetic topology effects on Alcator C-Mod scrape-off layer flow'

International Nuclear Information System (INIS)

Simakov, Andrei N; Catto, Peter J

2009-01-01

In his comment to our recent work (Simakov et al 2008 Plasma Phys. Control. Fusion 50 105010), Aydemir has asserted that poloidal plasma flow reversal is not a valid response to toroidal magnetic field reversal in an up-down symmetric tokamak, and that the toroidal plasma flow must reverse instead. We show that this assertion is wrong due to his misunderstanding of the corresponding symmetry transformation. (reply)

Stochastic layer scaling in the two-wire model for divertor tokamaks

Science.gov (United States)

Ali, Halima; Punjabi, Alkesh; Boozer, Allen

2009-06-01

The question of magnetic field structure in the vicinity of the separatrix in divertor tokamaks is studied. The authors have investigated this problem earlier in a series of papers, using various mathematical techniques. In the present paper, the two-wire model (TWM) [Reiman, A. 1996 Phys. Plasmas 3, 906] is considered. It is noted that, in the TWM, it is useful to consider an extra equation expressing magnetic flux conservation. This equation does not add any more information to the TWM, since the equation is derived from the TWM. This equation is useful for controlling the step size in the numerical integration of the TWM equations. The TWM with the extra equation is called the flux-preserving TWM. Nevertheless, the technique is apparently still plagued by numerical inaccuracies when the perturbation level is low, resulting in an incorrect scaling of the stochastic layer width. The stochastic broadening of the separatrix in the flux-preserving TWM is compared with that in the low mn (poloidal mode number m and toroidal mode number n) map (LMN) [Ali, H., Punjabi, A., Boozer, A. and Evans, T. 2004 Phys. Plasmas 11, 1908]. The flux-preserving TWM and LMN both give Boozer-Rechester 0.5 power scaling of the stochastic layer width with the amplitude of magnetic perturbation when the perturbation is sufficiently large [Boozer, A. and Rechester, A. 1978, Phys. Fluids 21, 682]. The flux-preserving TWM gives a larger stochastic layer width when the perturbation is low, while the LMN gives correct scaling in the low perturbation region. Area-preserving maps such as the LMN respect the Hamiltonian structure of field line trajectories, and have the added advantage of computational efficiency. Also, for a $1\\frac12$ degree of freedom Hamiltonian system such as field lines, maps do not give Arnold diffusion.

Coupled plasma-neutral transport model for the scrape-off region

International Nuclear Information System (INIS)

Galambos, J.D.; Peng, Y.K.M.; Heifetz, D.

1985-03-01

Analysis of the scrape-off region requires treatment of the plasma transport along and across the field lines and inclusion of the neutral transport effects. A method for modeling the scrape-off region that is presented here uses separate models for each of these aspects that are coupled together through an iteration procedure that requires only minimal numerical effort. The method is applied here to estimate the neutral pumping rates in the pump-limiter and divertor options for a proposed deuterium-tritium (D-T) ignition experiment. High neutral recycling in the vicinity of the neutralizer plate dramatically affects pumping rates for both the pump-limiter and divertor. In both cases, the plasma flow into the channel surrounding the neutralizer plate is greatly reduced by the neutral recycling. The fraction of this flow that is pumped can be large (> 50%), but in general it is dependent on the particular geometry and plasma conditions. It is estimated that pumping speeds approximately greater than 10 5 L/s are adequate for the exhaust requirements in the pump-limiter and the divertor cases. Also, high neutral recycling on the front surface of the limiter tends to increase the neutral pumping rate

On the density limit of Tokamaks

International Nuclear Information System (INIS)

Lehnert, B.

1982-12-01

Under the conditions of so far performed quasi-steady tokamak experiments near the density limit, the plasma pressure gradient in the outer layers of the plasma body becomes mainly determined by the plasma-neutral gas balance. An earlier analysis of ballooning instabilities driven by this gradient in regions of bad curvature has been extended to deduce an explicit stability criterion which determines the density limit. This criterion is closely related to the empirical Murakami limit. At relevant tokamak data, the deduced limit becomes proportional to J(sub)zR(sup)1/2 where J(sub)z is the average current density and R the major plasma radius. It is further found to be independent of the toroidal magnetic field strength and anomalous transport, as well as to be a slow function of the outer layer temperature and the mass number. The deduced stability criterion is consistent with so far performed experiments. Provided that the present analysis can be extrapolated to a wider range of parameter data and be combined with Alcator scaling, conditions near ignition appear to become realizable in small tokamaks by ohmic heating alone. These conditions can be satisfied at relevant magnetic field strengths and plasma currents, by imposing a high plasma current density. (author)

Two-ion ICRF heating in Tokamaks

International Nuclear Information System (INIS)

Tennfors, E.

1985-03-01

The practical consequences for tokamak plasma heating in the ion cyclotron frequency regime of the two-dimensional treatment of the two-ion mode conversion layer are analyzed. The problem of evaluation of the condition for fast wave resonance is analyzed, as well as the limitations imposed by warm plasma effects. Simple ways to find the mode conversion surfaces when they exist are presented. Also for large tokamaks, it is possible to obtain mode conversion conditions for realistic antenna spectra provided species concentration and frequency are chosen such that the surface Epsilon = 0 intersects the plasma midplane just outside of the magnetic axis. (Author)

ITER CTA newsletter. No. 16, January 2003

International Nuclear Information System (INIS)

2003-04-01

This ITER CTA newsletter contains information about some ITER related activities including ITER transitional arrangements (ITA) which will start on 1 January 2003, the USA rejoining ITER and People's Republic of China joining ITER, the visit of Mr. J. Koizumi, Prime Minister of Japan, to Kurchatov Institute, Moscow, Russian Federation on 11 January 2003, and the most recent meeting of the Scrape-Off Layer (SOL) and Divertor Physics Group of the International Tokamak Physics Activity (ITPA), which was held in Lausanne, Switzerland, on October 21-23, 2002 at the CRPP/EFL laboratory

Numerical simulations of blobs with ion dynamics

DEFF Research Database (Denmark)

Nielsen, Anders Henry; Rasmussen, Jens Juul; Madsen, Jens

2017-01-01

The transport of particles and energy into the scrape-off layer (SOL) region at the outboard midplane of medium-sized tokamaks, operating in low confinement mode, is investigated by applying the first-principle HESEL (hot edge-sol-electrostatic) model. HESEL is a four-field drift-fluid model...... of a 'shoulder' in the SOL density profile can be obtained by increasing the collisionality or connection length, thus decreasing the efficiency of the SOL's ability to remove plasma. As the ion pressure has a larger perpendicular but smaller parallel dissipation rate compared to the electron pressure, ion...

Tokamak

International Nuclear Information System (INIS)

Wesson, John.

1996-01-01

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

Coherence imaging of scrape-off-layer and divertor impurity flows in the Mega Amp Spherical Tokamak (invited)

Energy Technology Data Exchange (ETDEWEB)

Silburn, S. A., E-mail: s.a.silburn@durham.ac.uk; Sharples, R. M. [Centre for Advanced Instrumentation, Department of Physics, Durham University, Durham DH1 3LE (United Kingdom); Harrison, J. R.; Meyer, H.; Michael, C. A. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Howard, J. [Plasma Research Laboratory, Australian National University, Canberra, ACT 0200 (Australia); Gibson, K. J. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom)

2014-11-15

A new coherence imaging Doppler spectroscopy diagnostic has been deployed on the UK’s Mega Amp Spherical Tokamak for scrape-off-layer and divertor impurity flow measurements. The system has successfully obtained 2D images of C III, C II, and He II line-of-sight flows, in both the lower divertor and main scrape-off-layer. Flow imaging has been obtained at frame rates up to 1 kHz, with flow resolution of around 1 km/s and spatial resolution better than 1 cm, over a 40° field of view. C III data have been tomographically inverted to obtain poloidal profiles of the parallel impurity flow in the divertor under various conditions. In this paper we present the details of the instrument design, operation, calibration, and data analysis as well as a selection of flow imaging results which demonstrate the diagnostic's capabilities.

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

Science.gov (United States)

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

2017-10-01

Integrating the plasma core performance with an edge and scrape-off layer (SOL) that leads to tolerable heat and particle loads on the wall is a major challenge. The new European medium size tokamak task force (EU-MST) coordinates research on ASDEX Upgrade (AUG), MAST and TCV. This multi-machine approach within EU-MST, covering a wide parameter range, is instrumental to progress in the field, as ITER and DEMO core/pedestal and SOL parameters are not achievable simultaneously in present day devices. A two prong approach is adopted. On the one hand, scenarios with tolerable transient heat and particle loads, including active edge localised mode (ELM) control are developed. On the other hand, divertor solutions including advanced magnetic configurations are studied. Considerable progress has been made on both approaches, in particular in the fields of: ELM control with resonant magnetic perturbations (RMP), small ELM regimes, detachment onset and control, as well as filamentary scrape-off-layer transport. For example full ELM suppression has now been achieved on AUG at low collisionality with n  =  2 RMP maintaining good confinement {{H}\\text{H≤ft(98,\\text{y}2\\right)}}≈ 0.95 . Advances have been made with respect to detachment onset and control. Studies in advanced divertor configurations (Snowflake, Super-X and X-point target divertor) shed new light on SOL physics. Cross field filamentary transport has been characterised in a wide parameter regime on AUG, MAST and TCV progressing the theoretical and experimental understanding crucial for predicting first wall loads in ITER and DEMO. Conditions in the SOL also play a crucial role for ELM stability and access to small ELM regimes. In the future we will refer to the author list of the paper as the EUROfusion MST1 Team.

A model for a scrape-off-layer low-high (L-H) mode transition

International Nuclear Information System (INIS)

Cohen, R.H.; Xu, X.

1995-01-01

Increasing the radial mode number has a stabilizing effect on the conducting-wall and curvature-driven interchange modes in a tokamak scrape-off layer (SOL), arising from the increased polarization response. Such an effect is naturally imposed as the SOL width is decreased, and for a narrow-enough SOL, the stabilizing effect is stronger than the increase in the instability drives. By combining a mixing-length estimate for the thermal diffusivity with energy conservation and heat conduction equations and the condition of continuity of the heat flux at the separatrix, it is found that the resultant turbulence-transport system admits two solutions, one stable and one unstable, at different SOL widths; the inclusion of additional physics can add a second stable root at lower width. These roots are plausibly identified with SOL behavior in low (L) and high (H) modes. Particularly when a model is introduced for finite-β, finite-k parallel effects on the modes, a power threshold for transition to the narrower root is obtained, suggesting a possible L-H transition mechanism. The non-monotonic dependence of the turbulent heat flux vs SOL width and the possibility of multiple solutions for the equilibrium SOL width are verified with nonlinear simulations. copyright 1995 American Institute of Physics

An innovative method for ideal and resistive MHD stability analysis of tokamaks

International Nuclear Information System (INIS)

Tokuda, S.

2001-01-01

An advanced asymptotic matching method of ideal and resistive MHD stability analysis in tokamak is reported. The report explains a solution method of two-dimensional Newcomb equation, dispersion relation for an unstable ideal MHD mode in tokamak, and a new scheme for solving resistive MHD inner layer equations as an initial-value problem. (author)

An innovative method for ideal and resistive MHD stability analysis of tokamaks

International Nuclear Information System (INIS)

Tokuda, S.

2001-01-01

An advanced asymptotic matching method of ideal and resistive MHD stability analysis in tokamaks is reported. A solution method for the two dimensional Newcomb equation, a dispersion relation for an unstable ideal MHD mode in tokamaks and a new scheme for solving resistive MHD inner layer equations as an initial value problem are reported. (author)

Progress towards modeling tokamak boundary plasma turbulence and understanding its role in setting divertor heat flux widths

Science.gov (United States)

Chen, B.; Xu, X. Q.; Xia, T. Y.; Li, N. M.; Porkolab, M.; Edlund, E.; LaBombard, B.; Terry, J.; Hughes, J. W.; Ye, M. Y.; Wan, Y. X.

2018-05-01

The heat flux distributions on divertor targets in H-mode plasmas are serious concerns for future devices. We seek to simulate the tokamak boundary plasma turbulence and heat transport in the edge localized mode-suppressed regimes. The improved BOUT++ model shows that not only Ip but also the radial electric field Er plays an important role on the turbulence behavior and sets the heat flux width. Instead of calculating Er from the pressure gradient term (diamagnetic Er), it is calculated from the plasma transport equations with the sheath potential in the scrape-off layer and the plasma density and temperature profiles inside the separatrix from the experiment. The simulation results with the new Er model have better agreement with the experiment than using the diamagnetic Er model: (1) The electromagnetic turbulence in enhanced Dα H-mode shows the characteristics of quasi-coherent modes (QCMs) and broadband turbulence. The mode spectra are in agreement with the phase contrast imaging data and almost has no change in comparison to the cases which use the diamagnetic Er model; (2) the self-consistent boundary Er is needed for the turbulence simulations to get the consistent heat flux width with the experiment; (3) the frequencies of the QCMs are proportional to Er, while the divertor heat flux widths are inversely proportional to Er; and (4) the BOUT++ turbulence simulations yield a similar heat flux width to the experimental Eich scaling law and the prediction from the Goldston heuristic drift model.

Plasma potential and electron temperature evaluated by ball-pen and Langmuir probes in the COMPASS tokamak

Science.gov (United States)

Dimitrova, M.; Popov, Tsv K.; Adamek, J.; Kovačič, J.; Ivanova, P.; Hasan, E.; López-Bruna, D.; Seidl, J.; Vondráček, P.; Dejarnac, R.; Stöckel, J.; Imríšek, M.; Panek, R.; the COMPASS Team

2017-12-01

The radial distributions of the main plasma parameters in the scrape-off-layer of the COMPASS tokamak are measured during L-mode and H-mode regimes by using both Langmuir and ball-pen probes mounted on a horizontal reciprocating manipulator. The radial profile of the plasma potential derived previously from Langmuir probes data by using the first derivative probe technique is compared with data derived using ball-pen probes. A good agreement can be seen between the data acquired by the two techniques during the L-mode discharge and during the H-mode regime within the inter-ELM periods. In contrast with the first derivative probe technique, the ball-pen probe technique does not require a swept voltage and, therefore, the temporal resolution is only limited by the data acquisition system. In the electron temperature evaluation, in the far scrape-off layer and in the limiter shadow, where the electron energy distribution is Maxwellian, the results from both techniques match well. In the vicinity of the last closed flux surface, where the electron energy distribution function is bi-Maxwellian, the ball-pen probe technique results are in agreement with the high-temperature components of the electron distribution only. We also discuss the application of relatively large Langmuir probes placed in parallel and perpendicularly to the magnetic field lines to studying the main plasma parameters. The results obtained by the two types of the large probes agree well. They are compared with Thomson scattering data for electron temperatures and densities. The results for the electron densities are compared also with the results from ASTRA code calculation of the electron source due to the ionization of the neutrals by fast electrons and the origin of the bi-Maxwellian electron energy distribution function is briefly discussed.

The poloidal distribution of turbulent fluctuations in the Mega-Ampere Spherical Tokamak

International Nuclear Information System (INIS)

Antar, G.Y.; Counsell, G.; Ahn, J.-W.; Yang, Y.; Price, M.; Tabasso, A.; Kirk, A.

2005-01-01

Recently, it was shown that intermittency observed in magnetic fusion devices is caused by large-scales events with high radial velocity reaching about 1/10th of the sound speed (called avaloids or blobs) [G. Antar et al., Phys. Rev. Lett. 87 065001 (2001)]. In the present paper, the poloidal distribution of turbulence is investigated on the Mega-Ampere Spherical Tokamak [A. Sykes et al., Phys. Plasmas 8 2101 (2001)]. To achieve our goal, target probes that span the divertor strike points are used and one reciprocating probe at the midplane. Moreover, a fast imaging camera that can reach 10 μs exposure time looks tangentially at the plasma allowing us to view a poloidal cut of the plasma. The two diagnostics allow us to have a rather accurate description of the particle transport in the poloidal plane for L-mode discharges. Turbulence properties at the low-field midplane scrape-off layer are discussed and compared to other poloidal positions. On the low-field target divertor plates, avaloids bursty signature is not detected but still intermittency is observed far from the strike point. This is a consequence of the field line expansion which transforms a structure localized in the poloidal plane into a structure which expands over several tens of centimeters at the divertor target plates. Around the X point and in the high-field side, however, different phenomena enter into play suppressing the onset of convective transport generation. No signs of intermittency are observed in these regions. Accordingly, like 'normal' turbulence, the onset of convective transport is affected by the local magnetic curvature and shear

Control of long range turbulent transport with biasing in the tokamak scrape-off-layer

International Nuclear Information System (INIS)

Figarella, C.F.; Ghendrih, Ph.; Sarazin, Y.; Attuel, G.; Falchetto, G.; Fleurence, E.; Garbet, X.; Grandgirard, V.

2004-01-01

Cross field transport in the SOL (scrape-off-layer) influences tokamak performance in particular regarding the divertor efficiency. Recent experiment evidence emphasizes non-exponential and/or flat SOL profiles that suggest a large perpendicular transport. A 2-dimensional fluid model based on the interchange instability to simulate the SOL turbulence was found to exhibits intermittent dynamics of the particle flux. We propose a control method that prevents long range transport events from reaching the far SOL: It consists in biasing the far SOL leading to a transport barrier which stops the propagation of these intermittent events. The best trade off is to localize the biased toroidal ring around the baffles. We show that such a control is achievable providing the strength of the barrier is strong enough. The investigation of the minimal biasing power required to achieve the control as well as its experimental estimate is performed. (authors)

Diagnosing transient plasma status: from solar atmosphere to tokamak divertor

International Nuclear Information System (INIS)

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

2016-01-01

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

Investigation of self-organized criticality behavior of edge plasma transport in Torus experiment of technology oriented research

International Nuclear Information System (INIS)

Xu, Y.H.; Jachmich, S.; Weynants, R.R.; Huber, A.; Unterberg, B.; Samm, U.

2004-01-01

The self-organized criticality (SOC) behavior of the edge plasma transport has been studied using fluctuation data measured in the plasma edge and the scrape-off layer of Torus experiment of technology oriented research tokamak [H. Soltwisch et al., Plasma Phys. Controlled Fusion 26, 23 (1984)] before and during the edge biasing experiments. In the 'nonshear' discharge phase before biasing, the fluctuation data clearly show some of the characteristics associated with SOC, including similar frequency spectra to those obtained in 'sandpile' transport and other SOC systems, slowly decaying long tails in the autocorrelation function, values of Hurst parameters larger than 0.5 at all the detected radial locations, and a radial propagation of avalanchelike events in the edge plasma area. During the edge biasing phase, with the generation of an edge radial electric field E r and thus of E r xB flow shear, contrary to theoretical expectation, the Hurst parameters are substantially enhanced in the negative flow shear region and in the scrape-off layer as well. Concomitantly, it is found that the local turbulence is well decorrelated by the E r xB velocity shear, consistent with theoretical predictions

Impurity control in toroidal devices

International Nuclear Information System (INIS)

1990-01-01

This summary report on the Technical Committee Meeting organized by the IAEA and held in Naka-Gun, Japan, 13-15 February 1989, provides an overview of the results presented. Of the twenty-three papers presented, sixteen were devoted to tokamak experiments. These presented data of plasma behavior in the scrape-off layer and divertor regions, as well as effects of impurities on the core plasma; these are summarized here. Other papers summarized deal with plasma-wall interactions, including wall material behavior. Still others deal with theoretical work on physics modelling in the edge region. Refs, figs and tabs

Transport of recycled deuterium to the plasma core in TFTR

International Nuclear Information System (INIS)

Skinner, C.H.; Bell, M.G.; Budny, R.V.; Jassby, D.L.; Park, H.; Ramsey, A.T.; Stotler, D.P.; Strachan, J.D.

1997-10-01

The authors report a study of the fueling of the plasma core by recycling in the Tokamak Fusion Test Reactor (TFTR). They have analyzed discharges fueled by deuterium recycled from the limiter and tritium-only neutral beam injection. In these plasmas, the DT neutron rate provides a measure of the deuterium influx into the core plasma. They find a reduced influx with plasmas using lithium pellet conditioning and with plasmas of reduced major (and minor) radius. Modeling with the DEGAS neutrals code shows that the dependence on radius can be related to the penetration of neutrals through the scrape-off layer

Physics of collisionless scrape-off-layer plasma during normal and off-normal Tokamak operating conditions

International Nuclear Information System (INIS)

Hassanein, A.; Konkashbaev, I.

1999-01-01

The structure of a collisionless scrape-off-layer (SOL) plasma in tokamak reactors is being studied to define the electron distribution function and the corresponding sheath potential between the divertor plate and the edge plasma. The collisionless model is shown to be valid during the thermal phase of a plasma disruption, as well as during the newly desired low-recycling normal phase of operation with low-density, high-temperature, edge plasma conditions. An analytical solution is developed by solving the Fokker-Planck equation for electron distribution and balance in the SOL. The solution is in good agreement with numerical studies using Monte-Carlo methods. The analytical solutions provide an insight to the role of different physical and geometrical processes in a collisionless SOL during disruptions and during the enhanced phase of normal operation over a wide range of parameters

Examination of Deposited Layers Composition on the Discharge Chamber Constructional Elements Tokamak T-11M after Two-Year Operation with Lithium Limiter

International Nuclear Information System (INIS)

Buzhinskij, O.; Barsuk, V.

2006-01-01

In this work the results of the research of internal structural elements state of the T11-M tokamak discharge chamber after two-year operation with lithium limiter are given [V.B. Lazarev, E.A. Azizov et al., Compatibility of the Lithium Capillary Limiter with Plasma in T-11M, 26 th EPS Conf. on Contr. Fusion Plasma Physics, ECA, vol. 231, pp. 845-848, 1999, V.A. Evtikhin, I.E. Lyublinski, A.V. Vertkov et al., Technology Aspects of Lithium Capillary pore Systems Application in Tokamak Device, SOFT-21 (Madrid), A-37, 2000]. The condition of molybdenic wall surface of the discharge chamber and internal steel surface of diagnostic ports has been investigated. X-ray microanalysis of deposited surface of the first wall has shown, that in deposited layer are contained in the main Mo and small amount Cu. In a composition of deposited layer on the ports surface, except the above-named elements, in a small amount is Fe. Because of the instrumental restrictions of this method of analysis, detection opportunity of lithium traces was missing. X-ray diffractometer analysis of deposited layer on the first wall surface has detected a mixture of several phases. The main phase is Li 2 CO 3 , one third from all deposited substance is Li 2 MoO 4 , there is also LiOH-HO phase. The deposited layer on diagnostic ports in the main consists of LiOH-H 2 O phase, there is also Li 2 CO 3 phase. The results of X-ray analysis of a dust probe from the B 4 C coated graphite limiter surface have not detected whatever extra phases, except a crystalline boron carbide phase. (author)

Interpretation of scrape-off layer profile evolution and first-wall ion flux statistics on JET using a stochastic framework based on fillamentary motion

Science.gov (United States)

Walkden, N. R.; Wynn, A.; Militello, F.; Lipschultz, B.; Matthews, G.; Guillemaut, C.; Harrison, J.; Moulton, D.; Contributors, JET

2017-08-01

This paper presents the use of a novel modelling technique based around intermittent transport due to filament motion, to interpret experimental profile and fluctuation data in the scrape-off layer (SOL) of JET during the onset and evolution of a density profile shoulder. A baseline case is established, prior to shoulder formation, and the stochastic model is shown to be capable of simultaneously matching the time averaged profile measurement as well as the PDF shape and autocorrelation function from the ion-saturation current time series at the outer wall. Aspects of the stochastic model are then varied with the aim of producing a profile shoulder with statistical measurements consistent with experiment. This is achieved through a strong localised reduction in the density sink acting on the filaments within the model. The required reduction of the density sink occurs over a highly localised region with the timescale of the density sink increased by a factor of 25. This alone is found to be insufficient to model the expansion and flattening of the shoulder region as the density increases, which requires additional changes within the stochastic model. An example is found which includes both a reduction in the density sink and filament acceleration and provides a consistent match to the experimental data as the shoulder expands, though the uniqueness of this solution can not be guaranteed. Within the context of the stochastic model, this implies that the localised reduction in the density sink can trigger shoulder formation, but additional physics is required to explain the subsequent evolution of the profile.

Magnetic and material limiter discharges in Tokapole II

International Nuclear Information System (INIS)

Moyer, R.A.

1988-01-01

Disruptive instabilities have been studied in Tokapole II, a small poloidal divertor tokamak, in magnetic and material limiter configurations. In the magnetic limiter configuration, the divertor separatrix defines the tokamak current channel boundary. Limiters or neutralizer plates are not used to remove plasma in the scrape-off region. The relatively hot, dense plasma in the scrape-off region carries 5--20% of the current. In the material limiter configuration, limiter plates are inserted to the separatrix to remove plasma and current in the scrape-off region. The plates vary the tokamak current channel boundary condition in a controlled manner, and provide a benchmark for comparison with other tokamaks. Internal and external disruptions have been studied, and several unique features in the magnetic limiter configuration have been identified. The magnitic limiter configuration enables routine passing of the stability barriers at q(a) = 2 and q(a) = 1, where q(a) is the the edge safety factor, without a close fitting wall, external windings, or detailed profile control techniques. Passing the q(a) = 1 barrier permits operation in the q < 1 regime where total reconnection of the sawtooth does not occur. Discharges with q < 1 are also obtained in the material limiter configuration, suggesting that partial reconnection is characteristic of the sawteeth, and not the magnetic limiter configuration. The magnetic limiter configuration suppresses current termination in a major disruption. Current termination occurs in material limiter discharges due to enhanced interaction with the inboard limiter following the post-disruptive shift in major radius

Continuous tokamaks

International Nuclear Information System (INIS)

Peng, Y.K.M.

1978-04-01

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

Tritium system for a tokamak reactor with a self-pumped limiter

International Nuclear Information System (INIS)

Hassanein, A.M.; Sze, D.K.

1986-01-01

The self-pumping concept was proposed as a means of simplifying the impurity control system in a fusion reactor. The idea is to remove helium in-situ by trapping in freshly deposited metal surface layers of a limiter or divertor. Trapping material is added to the plasma scrape-off or edge region where it is transported to the wall. Some of the key issues for this concept are the tritium inventory in the trapping material and the permeation of protium and recycling of tritium. These quantities are shown to be acceptable for the reference design. The tritium issues for a helium-cooled solid breeder reactor design with vanadium alloy as a structural material are also examined. Models are presented for tritium permeation and inventory calculation for structure materials with the effect of a thin layer of coating material

Spatiotemporal Oscillations in Tokamak Edge Layer and their Generation by Various Mechanisms

Energy Technology Data Exchange (ETDEWEB)

Daybelge, U.; Yarim, C., E-mail: daybelge@itu.edu.tr [Istanbul Technical University, Istanbul (Turkey); Nicolai, A. [Forschungszentrum Juelich, Juelich (Germany)

2012-09-15

Full text: Toroidal and poloidal rotations of plasma at the edge region of tokamak devices have long been known to play an important role, such as enhancing the confinement properties by suppressing turbulent behaviour, improving tolerance to error fields and increasing stability to neoclassical tearing modes. Hence, understanding of creation and evolution of rotation is important, since external momentum would not be enough or could not even be realized especially for future large fusion devices. In addition to the externally applied momentum, several mechanisms have been suggested to explain the reasons for spontaneous toroidal rotation of plasmas. For a tokamak edge region as found, for example, within the operational boundaries of the ASDEX upgrade, relevance of the collisional neoclassical theory was recently emphasized. In this regime gyrostresses play a considerable role in modifying the coupled flux surface averaged continuity, energy and momentum equations. Examination of the terms in these equations that are responsible for diffusion or reaction and acting as sources, can show the share of the neoclassical mechanisms to terms like intrinsic rotation, etc. Using similarities of our equations to the nonlinear reaction-diffusion equations with a susceptibility to the Turing instability and applying some robust numerical methods, we present here an approach based on the spatiotemporal simulation of the oscillations in plasma temperature, density, toroidal and poloidal rotation velocities under various perturbative effects. Present study considers a subsonic, collisional plasma in front of the magnetic separatrix. Study indicates a nonlinear, three-time-scales-coupling between the evolutions of the density, temperature and poloidal and toroidal rotation velocities. Numerical solutions of the coupled system for the vector W = [T,N,U{sub {phi}}, U{sub {theta}}] were studied under various given sources such as a periodic pellet injection or loop voltage variation

Tokamak Systems Code

International Nuclear Information System (INIS)

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

1985-03-01

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

Physics of plasma-wall interactions in controlled fusion

International Nuclear Information System (INIS)

Post, D.E.; Behrisch, R.

1984-01-01

In the areas of plasma physics, atomic physics, surface physics, bulk material properties and fusion experiments and theory, the following topics are presented: the plasma sheath; plasma flow in the sheath and presheath of a scrape-off layer; probes for plasma edge diagnostics in magnetic confinement fusion devices; atomic and molecular collisions in the plasma boundary; physical sputtering of solids at ion bombardment; chemical sputtering and radiation enhanced sublimation of carbon; ion backscattering from solid surfaces; implantation, retention and release of hydrogen isotopes; surface erosion by electrical arcs; electron emission from solid surfaces;l properties of materials; plasma transport near material boundaries; plasma models for impurity control experiments; neutral particle transport; particle confinement and control in existing tokamaks; limiters and divertor plates; advanced limiters; divertor tokamak experiments; plasma wall interactions in heated plasmas; plasma-wall interactions in tandem mirror machines; and impurity control systems for reactor experiments

Asymmetric SOL Current in Vertically Displaced Plasma

Science.gov (United States)

Cabrera, J. D.; Navratil, G. A.; Hanson, J. M.

2017-10-01

Experiments at the DIII-D tokamak demonstrate a non-monotonic relationship between measured scrape-off layer (SOL) currents and vertical displacement event (VDE) rates with SOL currents becoming largely n=1 dominant as plasma is displaced by the plasma control system (PCS) at faster rates. The DIII-D PCS is used to displace the magnetic axis 10x slower than the intrinsic growth time of similar instabilities in lower single-null plasmas. Low order (n VDE instabilities observed when vertical control is disabled. Previous inquiry shows VDE asymmetry characterized by SOL current fraction and geometric parameters of tokamak plasmas. We note that, of plasmas displaced by the PCS, short displacement time scales near the limit of the PCS temporal control appear to result in larger n=1/n=2 asymmetries. Work supported under USDOE Cooperative Agreement DE-FC02-04ER54698 and DE-FG02-04ER54761.

Electrical characteristics of an ideal tokamak limiter

International Nuclear Information System (INIS)

Motley, R.W.

1981-01-01

The intrusion of an equipotential poloidal limiter into the edge plasma of a circular tokamak discharge distorts the axisymmetry in several ways: (1) it (partially) shorts out the top-to-bottom Pfirsch-Schlueter potentials, (2) it creates zones of reversed equilibrium current flow into the limiter, and (3) it generates an electrostatic field opposing the loop current. The resulting boundary mismatch between the outer layers and the inner axisymmetric Pfirsch-Schlueter layer increases the free energy available to drive the edge plasma unstable. A number of special limiters are proposed to symmetrize the edge plasma and thereby reduce the electrical and MHD activity in the boundary layer. (author)

Experimental study of the interaction between RF antennas and the edge plasma of a tokamak

International Nuclear Information System (INIS)

Kubic, Martin

2013-01-01

Antennas operating in the ion cyclotron range of frequency (ICRF) provide a useful tool for plasma heating in many tokamaks and are foreseen to play an important role in ITER. However, in addition to the desired heating in the core plasma, spurious interactions with the plasma edge and material boundary are known to occur. Many of these deleterious effects are caused by the formation of radio-frequency (RF) sheaths. The aim of this thesis is to study, mainly experimentally, scrape-off layer (SOL) modifications caused by RF sheaths effects by means of Langmuir probes that are magnetically connected to a powered ICRH antenna. Effects of the two types of Faraday screens' operation on RF-induced SOL modifications are studied for different plasma and antenna configurations - scans of strap power ratio imbalance, injected power and SOL density. In addition to experimental work, the influence of RF sheaths on retarding field analyzer (RFA) measurements of sheath potential is investigated with one-dimensional particle-in-cell code. One-dimensional particle-in-cell simulations show that the RFA is able to measure reliably the sheath potential only for ion plasma frequencies ω π similar to RF cyclotron frequency ω rf , while for the real SOL conditions (ω π ≥ ω rf ), when the RFA is magnetically connected to RF region, it is strongly underestimated. An alternative method to investigate RF sheaths effects is proposed by using broadening of the ion distribution function as an evidence of the RF electric fields in the sheath. RFA measurements in Tore Supra indicate that RF potentials do indeed propagate from the antenna 12 m along magnetic field lines. (author) [fr

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

Science.gov (United States)

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

2005-05-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Tokamak experiments

International Nuclear Information System (INIS)

Robinson, D.C.

1987-01-01

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

Divertor layer during the thermal phase of the disruption in a tokamak

International Nuclear Information System (INIS)

Konkashbaev, I.K.

1993-01-01

Physical processes in the tokamak divertor with a poloidal field during plasma disruption are considered. It is shown that the processes differ qualitatively from the processes in a stationary mode. During the disruption the energy flux is increased by 10 4 times

PPPL tokamak program

International Nuclear Information System (INIS)

Furth, H.P.

1984-10-01

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

Conceptual design of Remote Control System for EAST tokamak

International Nuclear Information System (INIS)

Sun, X.Y.; Wang, F.; Wang, Y.; Li, S.

2014-01-01

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

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.

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.

Radiative edge layers in limiter tokamaks

International Nuclear Information System (INIS)

Monier-Garbet, P.

1997-01-01

The characteristics of the highly radiative edge layers produced in the limiter configuration and with an open ergodic divertor are reviewed, with emphasis on the results obtained in TEXTOR and Tore Supra. In these two experiments an impurity injection technique is used to obtain highly radiating homogeneous peripheral layers. This requires that the peripheral radiation capability be maximized, while at the same time avoiding plasma core contamination; it is also necessary to insure the stability of the radiating layer. These physics issues, governing the success of the highly radiative edge scenario, are discussed. (orig.)

Comparison between stellarator and tokamak divertor transport

International Nuclear Information System (INIS)

Feng, Y.; Lunt, T.; Kobayashi, M.; Reiter, D.

2010-11-01

The paper compares the essential divertor transport features of the poloidal divertor, which is well-developed for tokamaks, and the non-axisymmetric divertors currently investigated on helical devices. It aims at surveying the fundamental similarities and differences in divertor concept and geometry, and their consequences for how the divertor functions. In particular, the importance of various transport terms governing axisymmetric and helical scrape-off-layers (SOLs) is examined, with special attention being paid to energy, momentum and impurity transport. Tokamak and stellarator SOLs are compared by identifying key geometric parameters through which the governing physics can be illustrated by simple models and estimates. More quantitative assessments rely nevertheless on the modeling using EMC3-EIRENE code. Most of the theoretical results are discussed in conjunction with experimental observations. (author)

Anomalous transport in the tokamak edge

International Nuclear Information System (INIS)

Vayakis, G.

1991-04-01

The tokamak edge has been studied with arrays of Langmuir and magnetic probes on the DITE and COMPASS-C devices. Measurements of plasma parameters such as density, temperature and radial magnetic field were taken in order to elucidate the character, effect on transport and origin of edge fluctuations. The tokamak edge is a strongly-turbulent environment, with large electrostatic fluctuation levels and broad spectra. The observations, including direct correlation measurements, are consistent with a picture in which the observed magnetic field fluctuations are driven by the perturbations in electrostatic parameters. The propagation characteristics of the turbulence, investigated using digital spectral techniques, appear to be dominated by the variation of the radial electric field, both in limiter and divertor plasmas. A shear layer is formed, associated in each case with the last closed flux surface. In the shear layer, the electrostatic wavenumber spectra are significantly broader. The predictions of a drift wave model (DDGDT) and of a family of models evolving from the rippling mode (RGDT group), are compared with experimental results. RGDT, augmented by impurity radiation effects, is shown to be the most reasonable candidate to explain the nature of the edge turbulence, only failing in its estimate of the wavenumber range. (Author)

Tokamak engineering mechanics

International Nuclear Information System (INIS)

Song, Yuntao; Wu, Weiyue; Du, Shijun

2014-01-01

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

Tokamak concept innovations

International Nuclear Information System (INIS)

1986-04-01

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

Survey of linear MHD stability in tokamak configurations

International Nuclear Information System (INIS)

Wakatani, M.

1977-01-01

The results found by MHD stability studies for both low-beta and high-beta tokamaks are reviewed. The stability against kink-ballooning modes in equilibria surrounded by vacuum or a layer of force free currents is considered. Internal kink modes and the relation to interchange modes, which should be considered after external kink modes are suppressed, are surveyed

KTM Tokamak operation scenarios software infrastructure

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

DIII-D edge physics database

International Nuclear Information System (INIS)

Jong, R.A.; Porter, G.D.; Hill, D.N.; Buchenauer, D.A.; Bramson, G.

1992-03-01

We have developed an edge-physics database containing data for the plasma in the divertor region and the scrape-off layer (SOL) for the DIII-D tokamak. The database provides many of the parameters necessary to model the power flow to the divertor and other plasma processes in the plasma edge. It will also facilitate the analysis of DIII-D data for comparison with other divertor tokamaks. In addition to the core plasma parameters, edge-specific data are included in this database. Initial results using the database show good agreement between the pressure profiles measured by the Langmuir probes and those determined from the Thomson data for the inner strike point, but not for the outer strike point region. We also find that the ratio of separatrix density to average core density, as well as the in/out asymmetry in the SOL power at the divertor in DIII-D do not agree with values currently assumed in modeling the International Thermonuclear Experimental Reactor (ITER)

Role of the higher harmonics in the launched LH spectrum in the ponderomotive effects at the plasma edge

Energy Technology Data Exchange (ETDEWEB)

Preinhaelter, J [Akademie Ved Ceske Republiky, Prague (Czech Republic). Ustav Fyziky Plazmatu; Vahala, L [Old Dominion Univ., Norfolk, VA (United States). Dept. of Electrical and Computer Engineering; Vahala, G [College of William and Mary, Williamsburg, VA (United States). Dept. of Physics

1995-12-01

The ponderomotive effects of higher harmonic waves in the lower hybrid waveguide grill power spectrum were investigated. Although possessing a few percent of the coupled power, these short-wavelength modes play an important part in the inhomogeneous scrape-off layer in tokamak plasma. Using a barometric representation of the ponderomotive potential for the electron density, the full wave coupled theory was solved self-consistently for 4 modes using a shooting-and-matching method to satisfy the surface impedance boundary conditions determined from infinite grill theory. For a symmetric LH spectrum and below a threshold power, the total reflection coefficient is independent of the coupled power. Isolated plasma density depressions (cavitons) are formed in regions of constructive superposition of short and long wavelength modes. Intense mode conversion also occurs in these regions. For power levels above the threshold, the reflection coefficient increases with power and eventually saturates. Numerical results are presented for the ASDEX tokamak and its 24 waveguide grill. 19 figs., 20 refs.

Physics Basis for a Spherical Torus Power Plant

International Nuclear Information System (INIS)

Kessel, C.E.; Menard, J.; Jardin, S.C.; Mau, T.K.

1999-01-01

The spherical torus, or low-aspect-ratio tokamak, is considered as the basis for a fusion power plant. A special class of wall-stabilized high-beta high-bootstrap fraction low-aspect-ratio tokamak equilibrium are analyzed with respect to MHD stability, bootstrap current and external current drive, poloidal field system requirements, power and particle exhaust and plasma operating regime. Overall systems optimization leads to a choice of aspect ratio A = 1:6, plasma elongation kappa = 3:4, and triangularity delta = 0:64. The design value for the plasma toroidal beta is 50%, corresponding to beta N = 7:4, which is 10% below the ideal stability limit. The bootstrap fraction of 99% greatly alleviates the current drive requirements, which are met by tangential neutral beam injection. The design is such that 45% of the thermal power is radiated in the plasma by Bremsstrahlung and trace Krypton, with Neon in the scrapeoff layer radiating the remainder

Wide-angle ITER-prototype tangential infrared and visible viewing system for DIII-D

Energy Technology Data Exchange (ETDEWEB)

Lasnier, C. J., E-mail: lasnier@LLNL.gov; Allen, S. L.; Ellis, R. E.; Fenstermacher, M. E.; McLean, A. G.; Meyer, W. H.; Morris, K.; Seppala, L. G. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94551-0808 (United States); Crabtree, K. [College of Optics, University of Arizona, Tucson, Arizona 85721 (United States); Van Zeeland, M. A. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States)

2014-11-15

An imaging system with a wide-angle tangential view of the full poloidal cross-section of the tokamak in simultaneous infrared and visible light has been installed on DIII-D. The optical train includes three polished stainless steel mirrors in vacuum, which view the tokamak through an aperture in the first mirror, similar to the design concept proposed for ITER. A dichroic beam splitter outside the vacuum separates visible and infrared (IR) light. Spatial calibration is accomplished by warping a CAD-rendered image to align with landmarks in a data image. The IR camera provides scrape-off layer heat flux profile deposition features in diverted and inner-wall-limited plasmas, such as heat flux reduction in pumped radiative divertor shots. Demonstration of the system to date includes observation of fast-ion losses to the outer wall during neutral beam injection, and shows reduced peak wall heat loading with disruption mitigation by injection of a massive gas puff.

A neutral-particle-based divertor model for Tokamak reactors

International Nuclear Information System (INIS)

Emmert, G.A.

1977-04-01

A zero-dimensional divertor model is derived from one-dimensional 'fluid' equations for the plasma and neutral particles. In this model the important process determining wall bombardment is charge exchange between the background neutral gas and the ions. Simple formulas are presented for the rate of wall bombardment, for the mean temperature and for the shielding efficiency of the plasma in the scrape-off zone. Only the latter parameter is strongly dependent on the parallel transport process in the scrape-off zone. None of the parameters are dependent on the perpendicular transport process. The model shows reasonable agreement with the one-dimensional computer simulation calculations of Mense. (orig.) [de

Joint research using small tokamaks

International Nuclear Information System (INIS)

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

2005-01-01

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

Application of mineral insulated cable (MIC) in Tokamak fusion reactor

International Nuclear Information System (INIS)

Luo Tianyong; Jiang Jiaming; Cen Yishun

2014-01-01

To avoid the instability of plasma and achieve some experimental tasks in Tokamak fusion reactor, many in-vessel coils are designed such as the coils to mitigate the effect of Edge Localized Modes (ELMs coils) and the coils to provide vertical stabilization (VS coils). The in-vessel location presents special challenges in terms of nuclear radiation and temperature, and requires the use of mineral-insulated conductors. The in-vessel coils in ITER are designed to be Mineral-insulated Cable (MIC) with three-layer structures. The inner is hollow-core tube made by OFHC or CuCrZr, the middle is the insulation layer made by Mgo and the outer is the jacket by SS316L or Inconel 718. To control the effect of Edge Localized Modes and vertical instability of plasma, the MIC in-vessel coils shall be used in HL-2M. More details about the application of MIC in Tokamak fusion reactor will be shown in this report. (authors)

Joint research using small tokamaks

International Nuclear Information System (INIS)

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

2005-01-01

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

Tokamaks. 2. ed.

International Nuclear Information System (INIS)

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

1997-01-01

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

Tokamak devices: towards controlled fusion

International Nuclear Information System (INIS)

Trocheris, M.

1975-01-01

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

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

International Nuclear Information System (INIS)

Mueller, E.R.; Steinmetz, K.; Bein, B.K.

1984-06-01

A new, fully computerized and automatic thermographic system has been developed. Its two central components are an AGA THV 780 infrared camera and a PDP-11/34 computer. A combined analytical-numerical method of solving the 1-dimensional heat diffusion equation for a solid of finite thickness bounded by two parallel planes was developed. In high-density (anti nsub(e) = 8 x 10 13 cm -3 ) neutral-beam-heated (L-mode) divertor discharges in ASDEX, the power deposition on the neutralizer plates is reduced to about 10-15% of the total heating power, owing to the inelastic scattering of the divertor plasma from a neutral gas target. Between 30% and 40% of the power is missing in the global balance. The power flow inside the divertor chambers is restricted to an approximately 1-cm-thick plasma scrape-off layer. This width depends only weakly on the density and heating power. During H-phases free of Edge Localized Mode (ELM) activity the energy flow into the divertor is blocked. During H-phases with ELM activity the energy is expelled into the divertor in very short intense pulses (several MW for about one hundred μs). Sawtooth events are able to transport significant amounts of energy from the plasma core to the peripheral zones and the scrape-off layer, and they are frequently correlated with transitions from the L to the H mode. (orig./AH)

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

Indian Academy of Sciences (India)

2016-10-20

Oct 20, 2016 ... ruby laser technology have been extended to deter- mine q. The edge radial electric field is known to ... The control of the shear layer is therefore important to modify transport in tokamaks [26–33]. A ... The calculation of PSD is directly accomplished by the FFT technique presented in the MATLAB software.

Modelling of radial electric fields and currents during divertor plate biasing on TdeV

International Nuclear Information System (INIS)

Lachambre, J.L.; Quirion, B.; Boucher, C.

1994-01-01

A simple model based on non-ambipolar radial transport and planar sheath physics is used to describe the generation of radial electric fields and currents in the scrape-off layer of the Tokamak de Varennes (TdeV) during divertor plate biasing. In general, the calculated predictions compare favourably with TdeV results over a variety of plasma conditions and divertor magnetic configurations. Validated by the experiment, the model is used to study the scaling laws of perpendicular ion mobility and to test existing related theories. Finally, the model is proposed as a useful tool for the design and upgrade of biased divertors through optimization of the plate and throat geometry. (author). 35 refs, 16 figs, 1 tab

Annual technical report

International Nuclear Information System (INIS)

Yang, T.F.; Gierszewski, P.

1982-01-01

This paper describes a Monte Carlo study of the thermal conductivity and impurity transport in a tokamak with an optimized bundle divertor configuration. This configuration has been designed to obtain the best plasma performance. When collisions and electric fields are included, the thermal conductivity was found to be 40% larger than that of an axisymmetric neoclassical value. Without a divertor the oxygen impurity was found to diffuse towards the center of the plasma as predicted theoretically. However, this impurity diffused outwardly when the divertor was turned on. Few impurity particles diffused into the plasma when launched in the scrape-off layer. This optimized bundle divertor meets the engineering constraints of a power producing reactor

Alfven wave studies on a tokamak

International Nuclear Information System (INIS)

Kortbawi, D.

1987-10-01

The continuum modes of the shear Alfven resonance are studied on the Tokapole II device, a small tokamak operated in a four node poloidal divertor configuration. A variety of antenna designs and the efficiency with which they deliver energy to the resonant layer are discussed. The spatial structure of the driven waves is studied by means of magnetic probes inserted into the current channel. In an attempt to optimize the coupling of energy in to the resonant layer, the angle of antenna currents with respect to the equilibrium field, antenna size, and plasma-to-antenna distance are varied. The usefulness of Faraday shields, particle shields, and local limiters are investigated. Antennas should be well shielded, either a dense Faraday shield or particle shield being satisfactory. The antenna should be large and very near to the plasma. The wave magnetic fields measured show a spatial resonance, the position of which varies with the value of the equilibrium field and mass density. They are polarized perpendicular to the equilibrium field. A wave propagates radially in to the resonant surface where it is converted to the shear Alfven wave. The signal has a short risetime and does not propagate far toroidally. These points are all consistent with a strongly damped shear Alfven wave. Comparisons of this work to theoretical predictions and results from other tokamaks are made

Impact of an integrated core/SOL description on the R and B T optimization of tokamak fusion reactors

Science.gov (United States)

Siccinio, M.; Fable, E.; Angioni, C.; Saarelma, S.; Scarabosio, A.; Zohm, H.

2018-01-01

An updated and improved version of the 0D divertor and scrape-off layer (SOL) model published in Siccinio et al (2016 Plasma Phys. Control. Fusion 58 125011) was coupled with the 1.5D transport code ASTRA (Pereverzev 1991 IPP Report 5/42, Pereverzev and Yushmanov 2002 IPP Report 5/98 and Fable et al 2013 Plasma Phys. Control. Fusion 55 124028). The resulting numerical tool was employed for various scans in the major radius R and in the toroidal magnetic field B T—for different safety factors q, allowable loop voltages V loop and H factors—in order to identify the most convenient choices for an electricity producing tokamak. Such a scenario analysis was carried out evaluating self-consistently, and simultaneously, the core profile and transport effects, which significantly impact on the fusion power outcome, and the divertor heat loads, which represent one of the most critical issues in view of the realization of fusion power plants (Zohm et al 2013 Nucl. Fusion 53 073019 and Wenninger et al 2017 Nucl. Fusion 57 046002). The main result is that, when divertor limits are enforced, the curves at constant electrical power output are closed on themselves in the R-BT plane, and a maximum achievable power exists—i.e. no benefits would be obtained from a further increase in R and B T once the optimum is reached. This result appears as an intrinsic physical limit for all those devices where a radiative SOL is needed to deal with the power exhaust, and where a lower limit on the power crossing the separatrix (e.g. because of the L-H transition) is present.

Investigation of scrape-off layer up-down asymmetries in diverted plasmas in TEXT-Upgrade

International Nuclear Information System (INIS)

Bonnin, X.P.G.

1998-05-01

On diverted discharges in TEXT-U, whether it be in single-null or double-null configurations, up-down asymmetries have been noted, the sign and magnitude of which depend on plasma conditions. Geometrical effects play a role in double-null discharges but may not explain the observations. Moreover, they do not play a role in the single-null plasmas, which are topologically up-down symmetric, so the asymmetry must come from physical driving mechanisms. To fully understand these drives a realistic edge plasma must be simulated. In so doing, the authors hope to gain insight as to which physical processes intervene in creating such asymmetries. The B2-Eirene plasma simulation code, upgraded and partially rewritten to handle the various geometries, will be a useful tool in evaluating the respective influences of the driving terms. Several such mechanisms will be investigated, both computationally and experimentally, and compared with data from previous tokamak discharges. In the course of this, theoretical predictions for edge transport phenomena, for example, atomic drives, curvature drives and neoclassical drifts, possibly leading to asymmetries, will be compared to experiment

Temporal evolution of blobs in the scrape-off layer

DEFF Research Database (Denmark)

Nielsen, Anders Henry; Madsen, Jens; Garcia, O.E.

Experimental observations have revealed that the transport in the edge and scrape-off-layer (SOL) of toroidally magnetized plasmas is strongly intermittent and involves large outbreaks of hot plasma. These structures, often referred to as “blobs”, are formed near the last closed flux surface (LCFS......) and propagate far into the SOL. The convective transport mediated by the blob-like structures prevails in virtually all confinement states, including edge-localized modes. They have a profound influence on the pressure profiles in the SOL, the ensuing parallel flows, and the power deposition on plasma facing...... and non-local [3] gyro-fluid equations. The focus of the investigations is the propagation of Gaussian “blob” like density structures. We examine the speed and the associated radial density transport as a function of blob amplitude and width. We observe an increase radial transport if finite ion...

Core-SOL simulations of L-mode tokamak plasma discharges using BALDUR code

Directory of Open Access Journals (Sweden)

Yutthapong Pinanroj

2014-04-01

Full Text Available Core-SOL simulations were carried out of plasma in tokamak reactors operating in a low confinement mode (L-mode, for various conditions that match available experimental data. The simulation results were quantitatively compared against experimental data, showing that the average RMS errors for electron temperature, ion temperature, and electron density were lower than 16% or less for 14 L-mode discharges from two tokamaks named DIII-D and TFTR. In the simulations, the core plasma transport was described using a combination of neoclassical transport calculated by NCLASS module and anomalous transport by Multi-Mode-Model version 2001 (MMM2001. The scrape-off-layer (SOL is the small amount of residual plasma that interacts with the tokamak vessel, and was simulated by integrating the fluid equations, including sources, along open field lines. The SOL solution provided the boundary conditions of core plasma region on low confinement mode (L-mode. The experimental data were for 14 L-mode discharges and from two tokamaks, named DIII-D and TFTR.

Detritiation of tiles from tokamaks by laser cleaning

International Nuclear Information System (INIS)

Coad, J. Paul; Widdowson, Anna; Farcage, Daniel; Semerok, Alexander; Thro, P.-Y.; Likonen, Jari; Renvall, Tommi

2007-01-01

Laser ablation has been used to clean surfaces or to decontaminate hot cells by removing paint, and has been tested on deposited carbon layers from the TEXTOR tokamak. This paper reports on successful trials in the Beryllium Handling Facility of a pulsed laser cleaning system to remove H-isotope containing carbon deposits on tiles from the JET tokamak. The laser beam is rastered over the surface of the tiles to remove the deposit. Two types of JET carbon-fibre composite (CFC) tiles were treated. The first was covered with carbon-based deposits up to 300 μm thick with high H-isotope content, the other was covered with a mixed Be/C film ∼ 50 microns thick. One scan of the laser was sufficient to completely change the appearance and expose the fibre planes. From cross-sectional micrographs, it was found that overall three scans provided the most effective settings for complete film removal. An area 250 cm 2 of the second tile was cleaned in 20 minutes, clearly demonstrating the efficiency of laser cleaning for the removal of tokamak deposits such as likely to occur in ITER. (authors)

The appearance and propagation of filaments in the private flux region in Mega Amp Spherical Tokamak

Energy Technology Data Exchange (ETDEWEB)

Harrison, J. R.; Fishpool, G. M.; Thornton, A. J. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Walkden, N. R. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom)

2015-09-15

The transport of particles via intermittent filamentary structures in the private flux region (PFR) of plasmas in the MAST tokamak has been investigated using a fast framing camera recording visible light emission from the volume of the lower divertor, as well as Langmuir probes and IR thermography monitoring particle and power fluxes to plasma-facing surfaces in the divertor. The visible camera data suggest that, in the divertor volume, fluctuations in light emission above the X-point are strongest in the scrape-off layer (SOL). Conversely, in the region below the X-point, it is found that these fluctuations are strongest in the PFR of the inner divertor leg. Detailed analysis of the appearance of these filaments in the camera data suggests that they are approximately circular, around 1–2 cm in diameter, but appear more elongated near the divertor target. The most probable toroidal quasi-mode number is between 2 and 3. These filaments eject plasma deeper into the private flux region, sometimes by the production of secondary filaments, moving at a speed of 0.5–1.0 km/s. Probe measurements at the inner divertor target suggest that the fluctuations in the particle flux to the inner target are strongest in the private flux region, and that the amplitude and distribution of these fluctuations are insensitive to the electron density of the core plasma, auxiliary heating and whether the plasma is single-null or double-null. It is found that the e-folding width of the time-average particle flux in the PFR decreases with increasing plasma current, but the fluctuations appear to be unaffected. At the outer divertor target, the fluctuations in particle and power fluxes are strongest in the SOL.

Observation of self-organized criticality (SOC) behavior during edge biasing experiment on TEXTOR

Energy Technology Data Exchange (ETDEWEB)

Xu, Y.H.; Jachmich, S.; Weynants, R.R. [Ecole Royale Militaire/Koninklijke Militaire School, Laboratory for Plasma Physics, Euratom-Belgian State Association, Brussels, Belgium, Partner in the Trilateral Euregio Cluster (Belgium)

2004-07-01

The self-organized criticality (SOC) behavior of the edge plasma transport has been investigated using the fluctuation data measured in the plasma edge and the scrape-off layer of TEXTOR tokamak before and during the edge electrode biasing experiments. In the 'non-shear' discharge phase before biasing, both the potential and density fluctuations clearly exhibit some of the characteristics associated with SOC: (1) existence of f{sup -1} power-law dependence in the frequency spectrum, (2) slowly decaying long tails in the autocorrelation function, (3) values of Hurst parameters larger than 0.5 at all the detected radial locations, (4) non-Gaussian probability density function of fluctuations and (5) radial propagation of avalanche-like events in the edge plasma area. During the biasing phase, with the generation of an edge radial electric field E{sub r} and hence a sheared E{sub r} x B flow, the local turbulence is found to be well de-correlated by the E{sub r} x B velocity shear, consistent with theoretical predictions. Nevertheless, it is concomitantly found that the Hurst parameters are substantially enhanced in the negative flow shear region and in the scrape-off layer as well, which is contrary to theoretical expectation. Implication of these observations to our understanding of plasma transport mechanisms is discussed. (authors)

Status of the tokamak program

Science.gov (United States)

Sheffield, J.

1981-08-01

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

Erosion and redeposition at the vessel walls in fusion devices

International Nuclear Information System (INIS)

Naujoks, D.; Behrisch, R.

1995-01-01

The plasma induced erosion and redeposition at the vessel walls in today's fusion devices have been investigated both with the computer simulation code ERO, and in experiments. Well prepared carbon probes with implanted and evaporated markers in the surface layers have been exposed in the scrape-off layer (SOL) of several tokamaks such as JET, TEXTOR and ASDEX-Upgrade. The main plasma parameters (electron density and temperature, impurity concentration in the SOL) are simultaneously determined. After exposure to single plasma discharges, erosion and redeposition of the marker material were measured by surface layer analysis with MeV ion beam techniques. The experimental results were compared with the results from the ERO code. The measured erosion/redeposition could be described with ERO, which takes into account the impurity concentration in the SOL, the dynamical change of the surface composition (causing a modification of the sputtering yield during the exposure) and ExB drift effects. ((orig.))

Conditional analysis of floating potential fluctuations at the edge of the Texas Experimental Tokamak Upgrade (TEXT-U)

International Nuclear Information System (INIS)

Filippas, A.V.; Bengston, R.D.; Li, G.; Meier, M.; Ritz, C.P.; Powers, E.J.

1995-01-01

Fluctuations in floating potential in the scrape-off layer and plasma edge were analyzed using a conditional statistical analysis technique. The floating potential fluctuations had a nearly Gaussian probability density function with the largest deviation from a Gaussian at the shear layer. The conditional averaging technique followed the statistical evolution of selected conditions in the floating potential signal. The decay rate of a conditional feature in time or space showed a small systematic variation with the amplitude of condition chosen. Either long-lived coherent structures are not present in statistically significant numbers, or the fluctuations are dominated by a large number of coherent structures with a nearly Gaussian distribution of fluctuation amplitudes, or conditional analysis using the amplitude of the floating potential as a condition is not a sensitive technique for identifying coherent structures

Present status of Tokamak research

International Nuclear Information System (INIS)

Basu, Jayanta

1991-01-01

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

Macroscopic erosion of divertor and first wall armour in future tokamaks

Science.gov (United States)

Würz, H.; Bazylev, B.; Landman, I.; Pestchanyi, S.; Safronov, V.

2002-12-01

Sputtering, evaporation and macroscopic erosion determine the lifetime of the 'in vessel' armour materials CFC, tungsten and beryllium presently under discussion for future tokamaks. For CFC armour macroscopic erosion means brittle destruction and dust formation whereas for metallic armour melt layer erosion by melt motion and droplet splashing. Available results on macroscopic erosion from hot plasma and e-beam simulation experiments and from tokamaks are critically evaluated and a comprehensive discussion of experimental and numerical macroscopic erosion and its extrapolation to future tokamaks is given. Shielding of divertor armour materials by their own vapor exists during plasma disruptions. The evolving plasma shield protects the armour from high heat loads, absorbs the incoming energy and reradiates it volumetrically thus reducing drastically the deposited energy. As a result, vertical target erosion by vaporization turns out to be of the order of a few microns per disruption event and macroscopic erosion becomes the dominant erosion source.

Macroscopic erosion of divertor and first wall armour in future tokamaks

International Nuclear Information System (INIS)

Wuerz, H.; Bazylev, B.; Landman, I.; Pestchanyi, S.; Safronov, V.

2002-01-01

Sputtering, evaporation and macroscopic erosion determine the lifetime of the 'in vessel' armour materials CFC, tungsten and beryllium presently under discussion for future tokamaks. For CFC armour macroscopic erosion means brittle destruction and dust formation whereas for metallic armour melt layer erosion by melt motion and droplet splashing. Available results on macroscopic erosion from hot plasma and e-beam simulation experiments and from tokamaks are critically evaluated and a comprehensive discussion of experimental and numerical macroscopic erosion and its extrapolation to future tokamaks is given. Shielding of divertor armour materials by their own vapor exists during plasma disruptions. The evolving plasma shield protects the armour from high heat loads, absorbs the incoming energy and reradiates it volumetrically thus reducing drastically the deposited energy. As a result, vertical target erosion by vaporization turns out to be of the order of a few microns per disruption event and macroscopic erosion becomes the dominant erosion source

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

The tokamak as a neutron source

International Nuclear Information System (INIS)

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

1989-11-01

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

Thermal loads on tokamak plasma-facing components during normal operation and disruptions

International Nuclear Information System (INIS)

McGrath, R.T.

1990-01-01

Power loadings experienced by tokamak plasma-facing components during normal operation and during off-normal events are discussed. A model for power and particle flow in the tokamak boundary layer is presented and model predictions are compared to infrared measurements of component heating. The inclusion of the full three-dimensional geometry of the components and of the magnetic flux surface is very important in the modeling. Experimental measurements show that misalignment of component armour tile surfaces by only a millimeter can lead to significant localized heating. An application to the design of plasma-facing components for future machines is presented. Finally, thermal loads expected during tokamak disruptions are discussed. The primary problems are surface melting and vaporization due to localized intense heating during the disruption thermal quench and volumetric heating of the component armour and structure due to localised impact of runaway electrons. (author)

Status of tokamak research

International Nuclear Information System (INIS)

Rawls, J.M.

1979-10-01

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

Divertor experiment for impurity control in DIVA

International Nuclear Information System (INIS)

Nagami, Masayuki

1979-04-01

Divertor actions of controlling the impurities and the transport of impurity ions in the plasma have been investigated in the DIVA device. Following are the results: (1) The radial transport of impurity ions is not described only by neoclassical theory, but it is strongly influenced by anomalous process. Radial diffusion of impurity ions across the whole minor radius is well described by a neoclassical diffusion superposed by the anomalous diffusion for protons. Due to this anomalous process, which spreads the radial density profile of impurity ions, 80 to 90% of the impurity flux in the plasma outer edge is shielded even in a nondiverted discharge. (2) The divertor reduces the impurity flux entering the main plasma by a factor of 2 to 4. The impurity ions shielded by the scrape-off plasma are rapidly guided into the burial chamber with a poloidal excursion time roughly equal to that of the scrape-off plasma. (3) The divertor reduces the impurity ion flux onto the main vacuum chamber by guiding the impurity ions diffusing from the main plasma into the burial chamber, thereby reducing the plasma-wall interaction caused by diffusing impurity ions at the main vacuum chamber. The impurity ions produced in the burial chamber may flow back to the main plasma through the scrape-off layer. However, roughly only 0.3% of the impurity flux into the scrape-off plasma in the burial chamber penetrates into the main plasma due to the impurity backflow. (4) A slight cooling of the scrape-off plasma with light-impurity injection effectively reduces the metal impurity production at the first wall by reducing the potential difference between the plasma and the wall, thereby reducing the accumulation of the metal impurity in the discharge. Radiation cooling by low-Z impurities in the plasma outer edge, which may become an important feature in future large tokamaks both with and without divertor, is numerically evaluated for carbon, oxygen and neon. (author)

Tokamak ARC damage

International Nuclear Information System (INIS)

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

1985-01-01

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

Tokamak ARC damage

Energy Technology Data Exchange (ETDEWEB)

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

1985-01-01

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

Simulation of tokamak runaway-electron events

International Nuclear Information System (INIS)

Bolt, H.; Miyahara, A.; Miyake, M.; Yamamoto, T.

1987-08-01

High energy runaway-electron events which can occur in tokamaks when the plasma hits the first wall are a critical issue for the materials selection of future devices. Runaway-electron events are simulated with an electron linear accelerator to better understand the observed runaway-electron damage to tokamak first wall materials and to consider the runaway-electron issue in further materials development and selection. The electron linear accelerator produces beam energies of 20 to 30 MeV at an integrated power input of up to 1.3 kW. Graphite, SiC + 2 % AlN, stainless steel, molybdenum and tungsten have been tested as bulk materials. To test the reliability of actively cooled systems under runaway-electron impact layer systems of graphite fixed to metal substrates have been tested. The irradiation resulted in damage to the metal compounds but left graphite and SiC + 2 % AlN without damage. Metal substrates of graphite - metal systems for actively cooled structures suffer severe damage unless thick graphite shielding is provided. (author)

Magnetic field structure near the plasma boundary in helical systems and divertor tokamaks

International Nuclear Information System (INIS)

Nagasaki, Kazunobu; Itoh, Kimitaka

1990-02-01

Magnetic field structure of the scrape off layer (SOL) region in both helical systems and divertor tokamaks is studied numerically by using model fields. The connection length of the field line to the wall is calculated. In helical systems, the connection length, L, has a logarithmic dependence on the distance from the outermost magnetic surface or that from the residual magnetic islands. The effect of axisymmetric fields on the field structure is also determined. In divertor tokamaks, the connection length also has logarithmic properties near the separatrix. Even when the perturbations, which resonate to rational surfaces near the plasma boundary, are added, logarithmic properties still remain. We compare the connection length of torsatron/helical-heliotron systems with that of divertor tokamaks. It is found that the former is shorter than the latter by one order magnitude with similar aspect ratio. (author)

Intermittency in the Scrape-off Layer of the National Spherical Torus Experiment During H-mode Confinement

International Nuclear Information System (INIS)

Maqueda, R.J.; Stotler, D.P.; Zweben, S.J.

2010-01-01

A gas puff imaging diagnostic is used in the National Spherical Tokamak Experiment (M. Ono, et al., Nucl. Fusion 40, 557 (2000)) to study the edge turbulence and intermittency present during H-mode discharges. In the case of low power Ohmic H-modes the suppression of turbulence/blobs is maintained through the duration of the (short lived) H-modes. Similar quiescent edges are seen during the early stages of H-modes created with the use of neutral beam injection. Nevertheless, as time progresses following the L-H transition, turbulence and blobs reappear although at a lower level than that typically seen during L-mode confinement. It is also seen that the time-averaged SOL emission profile broadens, as the power loss across the separatrix increases. These broad profiles are characterized by a large level of fluctuations and intermittent events.

Fast measurements of the electron temperature and parallel heat flux in ELMy H-mode on the COMPASS tokamak

Czech Academy of Sciences Publication Activity Database

Adámek, Jiří; Seidl, Jakub; Komm, Michael; Weinzettl, Vladimír; Pánek, Radomír; Stöckel, Jan; Hron, Martin; Háček, Pavel; Imríšek, Martin; Vondráček, Petr; Horáček, Jan; Devitre, A.

2017-01-01

Roč. 57, č. 2 (2017), č. článku 022010. ISSN 0029-5515 R&D Projects: GA ČR(CZ) GA15-10723S; GA MŠk(CZ) LM2015045; GA MŠk(CZ) 8D15001 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : COMPASS * divertor * ELM * scrape-off layer * ball-pen probe * power decay length Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 3.307, year: 2016 http://iopscience.iop.org/article/10.1088/0029-5515/57/2/022010

Tokamak reactor studies

International Nuclear Information System (INIS)

Baker, C.C.

1981-01-01

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

Disassembly of JT-60 tokamak device and ancillary facilities for JT-60 tokamak

International Nuclear Information System (INIS)

Okano, Fuminori; Ichige, Hisashi; Miyo, Yasuhiko; Kaminaga, Atsushi; Sasajima, Tadayuki; Nishiyama, Tomokazu; Yagyu, Jun-ichi; Ishige, Youichi; Suzuki, Hiroaki; Komuro, Kenichi; Sakasai, Akira; Ikeda, Yoshitaka

2014-03-01

The disassembly of JT-60 tokamak device and its peripheral equipments, where the total weight was about 5400 tons, started in 2009 and accomplished in October 2012. This disassembly was required process for JT-60SA project, which is the Satellite Tokamak project under Japan-EU international corroboration to modify the JT-60 to the superconducting tokamak. This work was the first experience of disassembling a large radioactive fusion device based on Radiation Hazard Prevention Act in Japan. The cutting was one of the main problems in this disassembly, such as to cut the welded parts together with toroidal field coils, and to cut the vacuum vessel into two. After solving these problems, the disassembly completed without disaster and accident. This report presents the outline of the JT-60 disassembly, especially tokamak device and ancillary facilities for tokamak device. (author)

Tokamak confinement scaling laws

International Nuclear Information System (INIS)

Connor, J.

1998-01-01

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

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

Directory of Open Access Journals (Sweden)

Ochoukov Roman

2017-01-01

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

Large Aspect Ratio Tokamak Study

International Nuclear Information System (INIS)

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

1980-06-01

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

Physics of the Tokamak Pedestal, and Implications for Magnetic Fusion Energy

Science.gov (United States)

Snyder, Philip

2017-10-01

High performance in tokamaks is achieved via the spontaneous formation of a transport barrier in the outer few percent of the confined plasma. This narrow insulating layer, referred to as a ``pedestal,'' typically results in a >30x increase in pressure across a 0.4-5cm layer. Predicted fusion power scales with the square of the pedestal top pressure (or ``pedestal height''), hence a fusion reactor strongly benefits from a high pedestal, provided this can be attained without large Edge Localized Modes (ELMs), which may erode plasma facing materials. The overlap of drift orbit, turbulence, and equilibrium scales across this narrow layer leads to rich and complex physics, and challenges traditional analytic and computational approaches. We review studies employing gyrokinetic, neoclassical, MHD, and other methods, which have explored how a range of instabilities, influenced by complex geometry, and strong ExB flows and bootstrap current, drive transport across the pedestal and guide its structure and dynamics. Development of high resolution diagnostics, and coordinated experiments on several tokamaks, have validated understanding of important aspects of the physics, while highlighting open issues. A predictive model (EPED) has proven capable of predicting the pedestal height and width to 20-25% accuracy in large statistical studies. This model was used to predict a new, high pedestal ``Super H-Mode'' regime, which was subsequently discovered on DIII-D, and motivated experiments on Alcator C-Mod which achieved world record, reactor relevant pedestal pressure. We review open issues including improved formalism, particle and momentum transport, the role of neutrals and impurities, ELM control, and pedestal formation. Finally we discuss coupling pedestal and core predictive models to enable more comprehensive optimization of the tokamak fusion concept. Supported by the US DOE under DE-FG02-95ER54309, FC02-06ER54873, DE-FC02-04ER54698, DE-FC02-99ER54512.

Plasma flow and transport on the tokamak ISTTOK boundary plasma

International Nuclear Information System (INIS)

Figueiredo, H.; Silva, C.; Goncalves, B.; Duarte, P.; Fernandes, H.

2011-01-01

The ISTTOK boundary plasma velocity near the outer midplane is measured on the parallel and perpendicular directions in four different configurations by reversing independently the toroidal magnetic field and the plasma current directions. The parallel flow is found to not depend significantly on both the toroidal magnetic field and plasma current directions, being always directed towards the nearest limiter in the scrape-off layer. On the contrary, the perpendicular flow is found to follow the E r x B drift direction. The poloidal velocity has also been derived from the correlation of floating potential signals measured on poloidally separated probes and a good agreement with the value derived with the Gundestrup probe is found. Finally, the dynamical interplay between parallel momentum and turbulent particle flux has been investigated and a clear dynamical coupling between these quantities is found in the region inside the limiter.

Thin layer activation techniques in research and industry

International Nuclear Information System (INIS)

Conlon, T.W.

1993-01-01

The following key application of thin layer activation technique (TLA) are discussed: ion-erosion in fusion tokamaks, bio-engineering technology, automobile industry. Future developments of the techniques, such as fission fragment TLA, multi-layer TLA and recoil implantation are discussed as well. 7 refs, 6 figs, 1 tab

Fusion potential for spherical and compact tokamaks

International Nuclear Information System (INIS)

Sandzelius, Mikael

2003-02-01

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

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.

Divertor asymmetry and scrape-off layer flow in various divertor configurations in Experimental Advanced Superconducting Tokamak

DEFF Research Database (Denmark)

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

2012-01-01

moving Mach probe at the outer midplane, which shows similar magnitude to the Pfirsch-Schluter flow. Its contribution to the poloidal particle flux is also assessed and comparison is made with that from the poloidal E x B drift. (C) 2012 American Institute of Physics. [http://dx.doi.org/10.1063/1.4707396]...

Characterized the pattern of the material deposition in the HL-2A tokamak

Energy Technology Data Exchange (ETDEWEB)

Cai, Laizhong, E-mail: cailz@swip.ac.cn [Southwestern Institute of Physics, P.O.Box 432, Chengdu, Sichuan 610041 (China); Wang, Jianbao; Wu, Ting; Zeng, Xiaoxiao [Southwestern Institute of Physics, P.O.Box 432, Chengdu, Sichuan 610041 (China); Hai, Ran; Ding, Hongbin [Dalian University of Technology, Dalian, Liaoning 116024 (China)

2017-03-15

Since the divertor geometry of a tokamak has a strong impact on the material erosion and deposition on the wall and HL-2A has a unique divertor configuration, it is necessary to investigate the material deposition pattern in HL-2A although a few results on other tokamaks have already been published. In this paper, tiles retrieved from the vessel are analyzed ex-situ by SIMS, SEM and laser-induced breakdown spectroscopy (LIBS). And deposition behind the lower divertor is in-situ measured by a quartz crystal microbalance (QMB). The deposition in HL-2A displays a complex pattern and clear localization characteristic. The thickness of the deposition layer varies in the range of 0-4μm. And in-situ diagnostic of QMB indicates that the average thickness of the deposition layer per pulse is over ten nanometers. In addition, the results imply that Si, Fe and D have different behaviors during the material deposition in HL-2A.

Overview of recent results from a Alcator C-Mod tokamak

International Nuclear Information System (INIS)

Marmar, E.; Batishchev, O.; Acedo, P.

1999-01-01

Recent results from the compact, high field, Alcator C-Mod tokamak program are summarized. H-mode threshold studies have demonstrated that the threshold appears to be closely related to local edge plasma parameters: for fixed field and plasma current, T e (ψ 95 ) takes on a density independent value at the transition. The Enhanced D-Alpha H-Mode (EDA) regime has been investigated. EDA is distinct from ELM free H-mode, in that there is no accumulation of impurities, and at the same time EDA does not exhibit large discrete ELMs. The energy confinement is degraded by only about 10%, compared to ELM free. Comparisons for EDA with ELMy H-Mode database scalings indicate τ EDA ∼1.2 τ ITER97H . Strong toroidal rotation is observed in ICRF-only auxiliary heated plasmas; the rotation increases with plasma pressure, and decreases with increasing plasma current. The inferred radial electric field reaches the order of 30 kV/m near the center of the plasma. Through feedback controlled nitrogen impurity puffing, steady state detached EDA H-Modes have been achieved with Z eff E is reduced by about 10% in the detached case, compared to the confinement before the N 2 puff begins. The heat load to the divertor is reduced by a factor of 4. Volume recombination rates are measured in the divertor, using 2-d tomography of Balmer series TV movies. Volume recombination can be a significant contributor to the overall reduction in ion current to the divertor plates which occurs in detachment. Particle balance measurements indicate that the divertor and main chamber plasmas are largely isolated from one another, at least with regard to particle recycling, with most of the main chamber (core plus scrape-off) fueling coming from neutrals in the main chamber volume. With the addition of Lower Hybrid Current Drive, C-Mod would be an ideal vehicle for investigation of advanced tokamak operation with fully relaxed current profiles. Detailed modeling indicates that discharges approaching the �

Overview of recent results from the Alcator C-Mod tokamak

International Nuclear Information System (INIS)

Marmar, E.; Batishchev, O.; Acedo, P.

2001-01-01

Recent results from the compact, high field, Alcator C-Mod tokamak program are summarized. H-mode threshold studies have demonstrated that the threshold appears to be closely related to local edge plasma parameters: for fixed field and plasma current, T e (ψ 95 ) takes on a density independent value at the transition. The Enhanced D-Alpha H-Mode (EDA) regime has been investigated. EDA is distinct from ELM free H mode, in that there is no accumulation of impurities, and at the same time EDA does not exhibit large discrete ELMs. The energy confinement is degraded by only about 10%, compared to ELM free. Comparisons for EDA with ELMy H-Mode database scalings indicate τEDA ∼ 1.2τ ITER97H . Strong toroidal rotation is observed in ICRF-only auxiliary heated plasmas; the rotation increases with plasma pressure, and decreases with increasing plasma current. The inferred radial electric field reaches the order of 30kV/m near the center of the plasma. Through feedback controlled nitrogen impurity puffing, steady state detached EDA H-Modes have been achieved with Z eff E is reduced by about 10% in the detached case, compared to the confinement before the N 2 puff begins. The heat load to the divertor is reduced by a factor of 4. Volume recombination rates are measured in the divertor, using 2-d tomography of Balmer series TV movies. Volume recombination can be a significant contributor to the overall reduction in ion current to the divertor plates which occurs in detachment. Particle balance measurements indicate that the divertor and main chamber plasmas are largely isolated from one another, at least with regard to particle recycling, with most of the main chamber (core plus scrape-off) fueling coming from neutrals in the main chamber volume. With the addition of Lower Hybrid Current Drive, C-Mod would be an ideal vehicle for investigation of advanced tokamak operation with fully relaxed current profiles. Detailed modeling indicates that discharges approaching the �

Microwave Tokamak Experiment

International Nuclear Information System (INIS)

Anon.

1988-01-01

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

Advanced Tokamak Stability Theory

Science.gov (United States)

Zheng, Linjin

2015-03-01

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

Wave trajectory and electron cyclotron heating in tokamak plasmas

International Nuclear Information System (INIS)

Tanaka, S.; Maekawa, T.; Terumichi, Y.; Hamada, Y.

1980-01-01

Wave trajectories in high density tokamak plasmas are studied numerically. Results show that the ordinary wave injected at an appropriate incident angle can propagate into the dense plasmas and is mode-converted to the extraordinary wave at the plasma cutoff, is further converted to the electron Bernstein wave during passing a loop or a folded curve near the upper hybrid resonance layer, and is cyclotron damped away, resulting in local electron heating before arriving at the cyclotron resonance layer. Similar trajectory and damping are obtained when a microwave in a form of extraordinary wave is injected quasi-perpendicularly in the direction of decreasing toroidal field

Burning plasma simulation and environmental assessment of tokamak, spherical tokamak and helical reactors

International Nuclear Information System (INIS)

Yamazaki, K.; Uemura, S.; Oishi, T.; Arimoto, H.; Shoji, T.; Garcia, J.

2009-01-01

Reference 1-GWe DT reactors (tokamak TR-1, spherical tokamak ST-1 and helical HR-1 reactors) are designed using physics, engineering and cost (PEC) code, and their plasma behaviours with internal transport barrier operations are analysed using toroidal transport analysis linkage (TOTAL) code, which clarifies the requirement of deep penetration of pellet fuelling to realize steady-state advanced burning operation. In addition, economical and environmental assessments were performed using extended PEC code, which shows the advantage of high beta tokamak reactors in the cost of electricity (COE) and the advantage of compact spherical tokamak in life-cycle CO 2 emission reduction. Comparing with other electric power generation systems, the COE of the fusion reactor is higher than that of the fission reactor, but on the same level as the oil thermal power system. CO 2 reduction can be achieved in fusion reactors the same as in the fission reactor. The energy payback ratio of the high-beta tokamak reactor TR-1 could be higher than that of other systems including the fission reactor.

The simple map for a single-null divertor tokamak

International Nuclear Information System (INIS)

Punjabi, A.; Verma, A.; Boozer, A.

1996-01-01

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)

Tokamak control simulator

International Nuclear Information System (INIS)

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

1976-01-01

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

Effects of orbit squeezing on neoclassical toroidal plasma viscosity in tokamaks

Czech Academy of Sciences Publication Activity Database

Shaing, K.C.; Sabbagh, S.A.; Chu, M.S.; Bécoulet, M.; Cahyna, Pavel

2008-01-01

Roč. 15, č. 8 (2008), 082505-1-082505-8 ISSN 1070-664X Institutional research plan: CEZ:AV0Z20430508 Keywords : plasma boundary layers * plasma instability * plasma magnetohydrodynamics * plasma toroidal confinement * plasma transport processes * Tokamak devices Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.427, year: 2008 http://dx.doi.org/10.1063/1.2965146

Survey of Tokamak experiments

International Nuclear Information System (INIS)

Bickerton, R.J.

1977-01-01

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

Plasma behavior and plasma-wall interaction in magnetic fusion divices

International Nuclear Information System (INIS)

Ohtsuka, Hideo

1984-10-01

To study the fundamental behavior of plasma in magnetic field is the main subject in the early stage of the magnetic fusion research. At the next stage, it is necessary to overcome some actual problems in order to attain reactor grade plasmas. One of them is to control impurities in the plasma. In these points of view, we carried out several experiments or theoretical analyses. Firstly, anomalous loss mechanisms in magnetic field were investigated in a toroidal multipole device JFT-1 and the role of motions of charged particles in the magnetic field was exhibited. Various measurements of plasma in the scrape-off layer were made in a divertor tokamak JFT-2a and in an ordinary tokamak JFT-2. The former study demonstrated the first successful divertor operation of the tokamak device and the latter one clarified the mechanism of arcing on the tokamak first wall. As to arcing, a new theory which describes the retrograde motion, the well known strange motion of arcs in a magnetic field, was proposed. Good agreement with the experimental results was shown. Finally, by considering a zero-dimensional sputtering model a self-consistent relation between light and metal impurities in tokamak plasmas was obtained. It was shown that the relation well describes some fundamental aspects of the plasma-wall interaction. As a conclusion, the importance of simple behavior of charged particles in magnetic fields was pointed out not only for the plasma confinement but also for the plasma-wall interaction. (author)

Tokamak building-design considerations for a large tokamak device

International Nuclear Information System (INIS)

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

1981-01-01

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

Modelling of Ohmic discharges in ADITYA tokamak using the Tokamak Simulation Code

International Nuclear Information System (INIS)

Bandyopadhyay, I; Ahmed, S M; Atrey, P K; Bhatt, S B; Bhattacharya, R; Chaudhury, M B; Deshpande, S P; Gupta, C N; Jha, R; Joisa, Y Shankar; Kumar, Vinay; Manchanda, R; Raju, D; Rao, C V S; Vasu, P

2004-01-01

Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z eff . Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values

Modelling of Ohmic discharges in ADITYA tokamak using the Tokamak Simulation Code

Energy Technology Data Exchange (ETDEWEB)

Bandyopadhyay, I; Ahmed, S M; Atrey, P K; Bhatt, S B; Bhattacharya, R; Chaudhury, M B; Deshpande, S P; Gupta, C N; Jha, R; Joisa, Y Shankar; Kumar, Vinay; Manchanda, R; Raju, D; Rao, C V S; Vasu, P [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India)

2004-09-01

Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z{sub eff}. Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values.

Advanced commercial tokamak study

International Nuclear Information System (INIS)

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

1985-12-01

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

Improvement of confinement characteristics of tokamak plasma by controlling plasma-wall interactions

International Nuclear Information System (INIS)

Sengoku, Seio

1985-08-01

Relation between plasma-wall interactions and confinement characteristics of a tokamak plasma with respect to both impurity and fuel particle controls is discussed. Following results are obtained from impurity control studies: (1) Ion sputtering is the dominant mechanism of impurity release in a steady state tokamak discharge. (2) By applying carbon coating on entire first wall of DIVA tokamak, dominant radiative region is concentrated more in boundary plasma resulting a hot peripheral plasma with cold boundary plasma. (3) A physical model of divertor functions about impurity control is empilically obtained. By a computer simulation based on above model with respect to divertor functions for JT-60 tokamak, it is found that the allowable electron temperature of the divertor plasma is not restricted by a condition that the impurity release due to ion sputtering does not increase continuously. (4) Dense and cold divertor plasma accompanied with strong remote radiative cooling was diagnosed along the magnetic field line in the simple poloidal divertor of DOUBLET III tokamak. Strong particle recycling region is found to be localized near the divertor plate. by and from particle control studies: (1) The INTOR scaling on energy confinement time is applicable to high density region when a core plasma is fueled directly by solid deuterium pellet injection in DOUBLET III tokamak. (2) As remarkably demonstrated by direct fueling with pellet injection, energy confinement characteristics can be improved at high density range by decreasing particle deposition at peripheral plasma in order to reduce plasma-wall interaction. (3) If the particle deposition at boundary layer is necessarily reduced, the electron temperature at the boundary or divertor region increases due to decrease of the particle recycling and the electron density there. (J.P.N.)

D-D tokamak reactor assessment

International Nuclear Information System (INIS)

Baxter, D.C.; Dabiri, A.E.

1983-01-01

A quantitative comparison of the physics and technology requirements, and the cost and safety performance of a d-d tokamak relative to a d-t tokamak has been performed. The first wall/blanket and energy recovery cycle for the d-d tokamak is simpler, and has a higher efficiency than the d-t tokamak. In most other technology areas (such as magnets, RF, vacuum, etc.) d-d requirements are more severe and the systems are more complex, expensive and may involve higher technical risk than d-t tokamak systems. Tritium technology for processing the plasma exhaust, and tritium refueling technology are required for d-d reactors, but no tritium containment around the blanket or heat transport system is needed. Cost studies show that for high plasma beta and high magnetic field the cost of electricity from d-d and d-t tokamaks is comparable. Safety analysis shows less radioactivity in a d-d reactor but larger amounts of stored energy and thus higher potential for energy release. Consequences of all postulated d-d accidents are significantly smaller than those from d-t reactor tritium releases

Characteristics of ion Bernstein wave heating in JIPPT-II-U tokamak

International Nuclear Information System (INIS)

Okamoto, M.; Ono, M.

1985-11-01

Using a transport code combined with an ion Bernstein wave tokamak ray tracing code, a modelling code for the ion Bernstein wave heating has been developed. Using this code, the ion Bernstein wave heating experiment on the JIPPT-II-U tokamak has been analyzed. It is assumed that the resonance layer is formed by the third harmonic of deuterium-like ions, such as fully ionized carbon, and oxygen ions near the plasma center. For wave absorption mechanisms, electron Landau damping, ion cyclotron harmonic damping, and collisional damping are considered. The characteristics of the ion Bernstein wave heating experiment, such as the ion temperature increase, the strong dependence of the quality factor on the magnetic field strength, and the dependence of the ion temperature increment on the input power, are well reproduced

Edge localized mode physics and operational aspects in tokamaks

Energy Technology Data Exchange (ETDEWEB)

Becoulet, M [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Huysmans, G [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Sarazin, Y [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Garbet, X [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Ghendrih, Ph [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Rimini, F [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Joffrin, E [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Litaudon, X [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Monier-Garbet, P [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Ane, J-M [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Thomas, P [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Grosman, A [Association Euratom-CEA, CEA Cadarache, F-13108 St Paul-lez-Durance (France); Parail, V [Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom); Wilson, H [Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom); Lomas, P [Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom); Vries, P de[Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom); Zastrow, K-D [Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom); Matthews, G F [Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom); Lonnroth, J [Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom); Gerasimov, S [Euratom/UKAEA Association, Fusion Culham Science Centre, Abingdon, OX14 3EA (United Kingdom)] [and others

2003-12-01

Recent progress in experimental and theoretical studies of edge localized mode (ELM) physics is reviewed for the reactor relevant plasma regimes, namely the high confinement regimes, that is, H-modes and advanced scenarios. Theoretical approaches to ELM physics, from a linear ideal magnetohydrodynamic (MHD) stability analysis to non-linear transport models with ELMs are discussed with respect to experimental observations, in particular the fast collapse of pedestal pressure profiles, magnetic measurements and scrape-off layer transport during ELMs. High confinement regimes with different types of ELMs are addressed in this paper in the context of development of operational scenarios for ITER. The key parameters that have been identified at present to reduce the energy losses in Type I ELMs are operation at high density, high edge magnetic shear and high triangularity. However, according to the present experimental scaling for the energy losses in Type I ELMs, the extrapolation of such regimes for ITER leads to unacceptably large heat loads on the divertor target plates exceeding the material limits. High confinement H-mode scenarios at high triangularity and high density with small ELMs (Type II), mixed regimes (Type II and Type I) and combined advanced regimes at high beta{sub p} are discussed for present-day tokamaks. The optimum combination of high confinement and small MHD activity at the edge in Type II ELM scenarios is of interest to ITER. However, to date, these regimes have been achieved in a rather narrow operational window and far from ITER parameters in terms of collisionality, edge safety factor and beta{sub p}. The compatibility of the alternative internal transport barrier (ITB) scenario with edge pedestal formation and ELMs is also addressed. Edge physics issues related to the possible combination of small benign ELMs (Type III, Type II ELMs, quiescent double barrier) and high performance ITBs are discussed for present-day experiments (JET, JT-60U

Edge localized mode physics and operational aspects in tokamaks

International Nuclear Information System (INIS)

Becoulet, M; Huysmans, G; Sarazin, Y; Garbet, X; Ghendrih, Ph; Rimini, F; Joffrin, E; Litaudon, X; Monier-Garbet, P; Ane, J-M; Thomas, P; Grosman, A; Parail, V; Wilson, H; Lomas, P; Vries, P de; Zastrow, K-D; Matthews, G F; Lonnroth, J; Gerasimov, S; Sharapov, S; Gryaznevich, M; Counsell, G; Kirk, A; Valovic, M; Buttery, R; Loarte, A; Saibene, G; Sartori, R; Leonard, A; Snyder, P; Lao, L L; Gohil, P; Evans, T E; Moyer, R A; Kamada, Y; Chankin, A; Oyama, N; Hatae, T; Asakura, N; Tudisco, O; Giovannozzi, E; Crisanti, F; Perez, C P; Koslowski, H R; Eich, T; Sips, A; Horton, L; Hermann, A; Lang, P; Stober, J; Suttrop, W; Beyer, P; Saarelma, S

2003-01-01

Recent progress in experimental and theoretical studies of edge localized mode (ELM) physics is reviewed for the reactor relevant plasma regimes, namely the high confinement regimes, that is, H-modes and advanced scenarios. Theoretical approaches to ELM physics, from a linear ideal magnetohydrodynamic (MHD) stability analysis to non-linear transport models with ELMs are discussed with respect to experimental observations, in particular the fast collapse of pedestal pressure profiles, magnetic measurements and scrape-off layer transport during ELMs. High confinement regimes with different types of ELMs are addressed in this paper in the context of development of operational scenarios for ITER. The key parameters that have been identified at present to reduce the energy losses in Type I ELMs are operation at high density, high edge magnetic shear and high triangularity. However, according to the present experimental scaling for the energy losses in Type I ELMs, the extrapolation of such regimes for ITER leads to unacceptably large heat loads on the divertor target plates exceeding the material limits. High confinement H-mode scenarios at high triangularity and high density with small ELMs (Type II), mixed regimes (Type II and Type I) and combined advanced regimes at high beta p are discussed for present-day tokamaks. The optimum combination of high confinement and small MHD activity at the edge in Type II ELM scenarios is of interest to ITER. However, to date, these regimes have been achieved in a rather narrow operational window and far from ITER parameters in terms of collisionality, edge safety factor and beta p . The compatibility of the alternative internal transport barrier (ITB) scenario with edge pedestal formation and ELMs is also addressed. Edge physics issues related to the possible combination of small benign ELMs (Type III, Type II ELMs, quiescent double barrier) and high performance ITBs are discussed for present-day experiments (JET, JT-60U, DIII-D) in

Transport in the tokamak plasma edge

International Nuclear Information System (INIS)

Vold, E.L.

1989-01-01

Experimental observations characterize the edge plasma or boundary layer in magnetically confined plasmas as a region of great complexity. Evidence suggests the edge physics plays a key role in plasma confinement although the mechanism remains unresolved. This study focuses on issues in two areas: observed poloidal asymmetries in the Scrape Off Layer (SOL) edge plasma and the physical nature of the plasma-neutral recycling. A computational model solves the coupled two dimensional partial differential equations governing the plasma fluid density, parallel and radial velocities, electron and ion temperatures and neutral density under assumptions of toroidal symmetry, ambipolarity, anomalous diffusive radial flux, and neutral-ion thermal equilibrium. Drift flow and plasma potential are calculated as dependent quantities. Computational results are compared to experimental data for the CCT and TEXTOR:ALT-II tokamak limiter cases. Comparisons show drift flux is a major component of the poloidal flow in the SOL along the tangency/separatrix. Plasma-neutral recycling is characterized in several tokamak divertors, including the C-MOD device using magnetic flux surface coordinates. Recycling is characterized by time constant, τ rc , on the order of tens of milliseconds. Heat flux transients from the core into the edge on shorter time scales significantly increase the plasma temperatures at the target and may increase sputtering. Recycling conditions in divertors vary considerably depending on recycled flux to the core. The high density, low temperature solution requires that the neutral mean free path be small compared to the divertor target to x-point distance. The simulations and analysis support H-mode confinement and transition models based on the recycling divertor solution bifurcation

Summary discussion: An integrated advanced tokamak reactor

International Nuclear Information System (INIS)

Sauthoff, N.R.

1994-01-01

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

Facilities for technology testing of ITER divertor concepts, models, and prototypes in a plasma environment

International Nuclear Information System (INIS)

Cohen, S.A.

1991-12-01

The exhaust of power and fusion-reaction products from ITER plasma are critical physics and technology issues from performance, safety, and reliability perspectives. Because of inadequate pulse length, fluence, flux, scrape-off layer plasma temperature and density, and other parameters, the present generation of tokamaks, linear plasma devices, or energetic beam facilities are unable to perform adequate technology testing of divertor components, though they are essential contributors to many physics issues such as edge-plasma transport and disruption effects and control. This Technical Requirements Documents presents a description of the capabilities and parameters divertor test facilities should have to perform accelerated life testing on predominantly technological divertor issues such as basic divertor concepts, heat load limits, thermal fatigue, tritium inventory and erosion/redeposition. The cost effectiveness of such divertor technology testing is also discussed

Meeting of the ITPA Topic Group on SOL and Divertor Physics

International Nuclear Information System (INIS)

Lipschultz, B.; Asakura, N.

2003-01-01

The most recent meeting of the Scrape-Off Layer (SOL) and Divertor Physics Group of the International Tokamak Physics Activity (ITPA) was held in Lausanne, Switzerland, on October 21-23, 2002 at the CRPP/EFL laboratory. The meeting was hosted by Dr. R. Pitts of that laboratory, with the help of other laboratory staff. There were 23 participants. The meeting format was slightly changed from that of the previous meeting (February 25-27, 2002) in so far that more time was allowed for discussions, thus making the meeting more productive. There were 28 detailed presentations over two and a half days falling into three primary areas: the physics of ELMs and their effects; radial transport and wall recycling; materials issues (T codeposition, use of W for tiles, etc.)

Models and applications of the UEDGE code

International Nuclear Information System (INIS)

Rensink, M.E.; Knoll, D.A.; Porter, G.D.; Rognlien, T.D.; Smith, G.R.; Wising, F.

1996-09-01

The transport of particles and energy from the core of a tokamak to nearby material surfaces is an important problem for understanding present experiments and for designing reactor-grade devices. A number of fluid transport codes have been developed to model the plasma in the edge and scrape-off layer (SOL) regions. This report will focus on recent model improvements and illustrative results from the UEDGE code. Some geometric and mesh considerations are introduced, followed by a general description of the plasma and neutral fluid models. A few comments on computational issues are given and then two important applications are illustrated concerning benchmarking and the ITER radiative divertor. Finally, we report on some recent work to improve the models in UEDGE by coupling to a Monte Carlo neutrals code and by utilizing an adaptive grid

A study of X-divertor in NSTX-U with SOLPS simulations

Science.gov (United States)

Chen, Zhong-Ping; Kotschenreuther, Mike; Mahajan, Swadesh; Gerhardt, Stefan

2018-03-01

The X-divertor (XD) geometry in NSTX-U is demonstrated, via SOLPS simulations, to perform better than the standard divertor (SD); in particular, it allows detachment at a lower upstream density and stabilizes the detachment front near the target, away from the main X-point. Consequently a stable detached operation becomes possible—the localization near the plate allows a vast reduction of heat fluxes without degrading the core plasma. Indeed, it is confirmed by our simulation that at similar states of detachment the XD outperforms the SD by reducing the heat fluxes to the target and maintaining higher upstream temperatures, resulting in scrape-off layers that are more favorable for advanced tokamak operation. These advantages are attributed to the unique geometric characteristics of XD—poloidal flaring near the target.

ERO-TEXTOR. 3D-Monte Carlo code for local impurity-modeling in the scrape-off-layer of TEXTOR. Version 2.0

International Nuclear Information System (INIS)

Koegler, U.; Winter, J.

1997-03-01

The ERO-TEXTOR code is described in detail. The code solves the kinetic equations of impurities in the scrape-off layer of a tokamak plasma in the vicinity of material surfaces like limiters or divertors. A relaxation time ansatz in the traced impurity limit is chosen, taking the gyro-motion of the particles into account. Since the background plasma is slightly non-maxwellian at the plasma edge higher order corrections (thermal forces) to the relaxation time ansatz are also considered. Background plasma parameters are calculated from a simple plasma model, i.e. the one dimensional continuity and momentum equations are used to derive the local electron density, the local flow velocity and the pre-sheath and sheath electric fields. Since these calculations are not done in a selfconsistent way, the measured values of electron density and temperature are used as basic input to derive the dependency of these quantities. The regarded magnetic topology is still straight and uniform. Also detailed account is given to the plasma surface interaction and the erosion/deposition processes. A linear differential equation model for multi species impact on a material surface has been developed and is used in a discrete time step approximation. External databases include the ionization rates for atomic species, molecular processes of methane and silane molecules and the sputtering and reflection yields, which are taken from binary collision calculation codes (e.g. TRIM) or from semi-empirical fits (e.g. the Bodhansky and Yamamura fits). (orig.)

Stability of coupled tearing and twisting modes in tokamaks

International Nuclear Information System (INIS)

Fitzpatrick, R.

1994-03-01

A dispersion relation is derived for resistive modes of arbitrary parity in a tokamak plasma. At low mode amplitude, tearing and twisting modes which have nonideal MHD behavior at only one rational surface at a time in the plasma are decoupled via sheared rotation and diamagnetic flows. At higher amplitude, more unstable open-quote compound close-quote modes develop which have nonideal behavior simultaneously at many surfaces. Such modes possess tearing parity layers at some of the nonideal surfaces, and twisting parity layers at others, but mixed parity layers are generally disallowed. At low mode number, open-quote compound close-quote modes are likely to have tearing parity layers at all of the nonideal surfaces in a very low-β plasma, but twisting parity layers become more probable as the plasma β is increased. At high mode number, unstable twisting modes which exceed a critical amplitude drive conventional magnetic island chains on alternate rational surfaces, to form an interlocking structure in which the O-points and X-points of neighboring chains line up

ELECTRIC POTENIAL CELLS AT THE DIVERTED TOKAMAK SEPARATRIX

International Nuclear Information System (INIS)

SCHAFFER, M.J.; PORTER, G.D.; BOEDO, J.A.; BRAY, B.D.; HSIEH, C.L.; MOYER, R.A.; ROGNLIEN, T.D.; STANGEBY, P.C.; WATKINS, J.G.

2000-01-01

OAK-B135 Two-dimensional measurements by probes and Thomson scattering reveal unanticipated electric potential and electron pressure (p e ) maxima near the divertor X-point in L-mode plasmas in the DIII-D tokamak. The potential hill (∼ 50 V) drives E x B circulation (potential cell) of particles, energy and toroidal momentum around the X-point and in and out across the magnetic separatrix. Modeling by the UEDGE two-dimensional edge transport code with plasma drifts shows similar X-point potential and pressure hills. The code predicts additional drift-driven nonuniformity poloidally around the separatrix. Potential cells in UEDGE arise from parallel (to B) viscous stress acting on the Pfirsch-Schlueter ion return flow of the (del)B drift. These experimental and theoretical results demonstrate that the boundary layer just inside the separatrix of low power tokamak plasmas can be far from poloidal uniformity. They speculate that separatrix potential cells might be a major feature of L-mode edge transport and their suppression an important feature of H-mode

Electric potential cells at the diverted tokamak separatrix

Energy Technology Data Exchange (ETDEWEB)

Schaffer, Michael J.; Bray, Bruce D.; Hsieh, Chung-Lih [General Atomics, San Diego, California (United States); Porter, Gary D.; Rognlien, Thomas D. [Lawrence Livermore National Laboratory, Livermore, California (United States); Boedo, Jose A.; Moyer, Richard A. [University of California, San Diego, California (United States); Stangeby, Peter C. [Univ. of Toronto, Toronto (Canada); Watkins, Jonathan G. [Sandia National Laboratories, Albuquerque, New Mexico (United States)

2001-07-01

Two-dimensional measurements by probes and Thomson scattering reveal unanticipated electric potential and electron pressure (p{sub e}) maxima near the divertor X-point in L-mode plasmas in the DIII-D tokamak. The potential hill ({approx}100 V) drives ExB circulation ('potential cell') of particles, energy and toroidal momentum around the X-point and in and out across the magnetic separatrix. Modeling by the UEDGE two-dimensional edge transport code with plasma drifts shows similar X-point potential and pressure hills. The code predicts additional drift-driven nonuniformity poloidally around the separatrix. Potential cells in UEDGE arises from parallel (to B) viscous stress acting on the Pfirsch-Schlueter ion return flow of the {nabla}B drift. These experimental and theoretical results demonstrate that the boundary layer just inside the separatrix of low power tokamak plasmas can be far from poloidal uniformity. We speculate that separatrix potential cells might be a major feature of L-mode edge transport and their suppression an important feature of H-mode. (author)

Electric potential cells at the diverted tokamak separatrix

International Nuclear Information System (INIS)

Schaffer, Michael J.; Bray, Bruce D.; Hsieh, Chung-Lih; Porter, Gary D.; Rognlien, Thomas D.; Boedo, Jose A.; Moyer, Richard A.; Stangeby, Peter C.; Watkins, Jonathan G.

2001-01-01

Two-dimensional measurements by probes and Thomson scattering reveal unanticipated electric potential and electron pressure (p e ) maxima near the divertor X-point in L-mode plasmas in the DIII-D tokamak. The potential hill (∼100 V) drives ExB circulation ('potential cell') of particles, energy and toroidal momentum around the X-point and in and out across the magnetic separatrix. Modeling by the UEDGE two-dimensional edge transport code with plasma drifts shows similar X-point potential and pressure hills. The code predicts additional drift-driven nonuniformity poloidally around the separatrix. Potential cells in UEDGE arises from parallel (to B) viscous stress acting on the Pfirsch-Schlueter ion return flow of the ∇B drift. These experimental and theoretical results demonstrate that the boundary layer just inside the separatrix of low power tokamak plasmas can be far from poloidal uniformity. We speculate that separatrix potential cells might be a major feature of L-mode edge transport and their suppression an important feature of H-mode. (author)

Progress of the ECH·ECCD experiments. Research progress of the ECH·ECCD experiments in tokamaks and spherical tokamaks

International Nuclear Information System (INIS)

Isayama, Akihiko; Tanaka, Hitoshi

2009-01-01

Recent progress in the ECH·ECCD study in tokamak and spherical tokamak devices is described. As for the tokamak study, results on the control of neoclassical tearing modes and sawtooth oscillations, the current profile, the internal transport barrier, the plasma start-up and the discharge cleaning are given. As for the spherical tokamak study, the plasma start-up by ECH·ECCD and the electron-Bernstein-wave heating and the current drive are described. (T.I.)

Preliminary Design of Alborz Tokamak

Science.gov (United States)

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

2012-04-01

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

Linear neoclassical tearing mode in tokamaks

International Nuclear Information System (INIS)

Shaing, K. C.

2007-01-01

The growth rate of linear tearing modes in tokamaks is calculated including the neoclassical dissipation mechanism. It is found that when the growth rate is much smaller than the ion-ion collision frequency, the growth rate is reduced approximately by a factor of (B p /B) 2/5 from its standard value, and when the growth rate is much larger than the ion-ion collision frequency, the growth rate is reduced by a factor [√(ε)/(1.6q 2 )] 1/5 . Here, B p is the poloidal magnetic field strength, B is the magnetic field strength, ε is the inverse aspect ratio, and q is the safety factor. The width of the resistive layer is broadened when compared to that of the standard theory. In both limits, the growth rate and the resistive layer width only depend on B p and are independent of B. The growth rates in the plateau regime and for the inertia dominant modes are also presented

Plasma transport in a compact ignition tokamak

International Nuclear Information System (INIS)

Singer, C.E.; Ku, L.P; Bateman, G.

1987-02-01

Nominal predicted plasma conditions in a compact ignition tokamak are illustrated by transport simulations using experimentally calibrated plasma transport models. The range of uncertainty in these predictions is explored by using various models which have given almost equally good fits to experimental data. Using a transport model which best fits the data, thermonuclear ignition occurs in a Compact Ignition Tokamak design with major radius 1.32 m, plasma half-width 0.43 m, elongation 2.0, and toroidal field and plasma current ramped in six seconds from 1.7 to 10.4 T and 0.7 to 10 MA, respectively. Ignition is facilitated by 20 MW of heating deposited off the magnetic axis near the 3 He minority cyclotron resonance layer. Under these conditions, sawtooth oscillations are small and have little impact on ignition. Tritium inventory is minimized by preconditioning most discharges with deuterium. Tritium is injected, in large frozen pellets, only after minority resonance preheating. Variations of the transport model, impurity influx, heating profile, and pellet ablation rates, have a large effect on ignition and on the maximum beta that can be achieved

Moving Divertor Plates in a Tokamak

International Nuclear Information System (INIS)

Zweben, S.J.; Zhang, H.

2009-01-01

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

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.

The ARIES tokamak reactor study

International Nuclear Information System (INIS)

1989-10-01

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

Summary report on tokamak confinement experiments

International Nuclear Information System (INIS)

1982-03-01

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

Tokamaks - Third Edition

International Nuclear Information System (INIS)

Rogister, A L

2004-01-01

John Wesson's well known book, now re-edited for the third time, provides an excellent introduction to fusion oriented plasma physics in tokamaks. The author's task was a very challenging one, for a confined plasma is a complex system characterised by a variety of dimensionless parameters and its properties change qualitatively when certain threshold values are reached in this multi-parameter space. As a consequence, theoretical description is required at different levels, which are complementary: particle orbits, kinetic and fluid descriptions, but also intuitive and empirical approaches. Theory must be carried out on many fronts: equilibrium, instabilities, heating, transport etc. Since the properties of the confined plasma depend on the boundary conditions, the physics of plasmas along open magnetic field lines and plasma surface interaction processes must also be accounted for. Those subjects (and others) are discussed in depth in chapters 2-9. Chapter 1 mostly deals with ignition requirements and the tokamak concept, while chapter 14 provides a list of useful relations: differential operators, collision times, characteristic lengths and frequencies, expressions for the neoclassical resistivity and heat conduction, the bootstrap current etc. The presentation is sufficiently broad and thorough that specialists within tokamak research can either pick useful and up-to-date information or find an authoritative introduction into other areas of the subject. It is also clear and concise so that it should provide an attractive and accurate initiation for those wishing to enter the field and for outsiders who would like to understand the concepts and be informed about the goals and challenges on the horizon. Validation of theoretical models requires adequately resolved experimental data for the various equilibrium profiles (clearly a challenge in the vicinity of transport barriers) and the fluctuations to which instabilities give rise. Chapter 10 is therefore devoted to

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

Energy Technology Data Exchange (ETDEWEB)

Schmid, Andreas

2008-03-18

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

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

International Nuclear Information System (INIS)

Schmid, Andreas

2008-01-01

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

Research using small tokamaks

International Nuclear Information System (INIS)

1991-01-01

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

Accelerator technology in tokamaks

International Nuclear Information System (INIS)

Kustom, R.L.

1977-01-01

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

Texas Experimental Tokamak

International Nuclear Information System (INIS)

Wootton, A.J.

1993-04-01

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

START: the creation of a spherical tokamak

International Nuclear Information System (INIS)

Sykes, Alan

1992-01-01

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

Tokamaks (Second Edition)

Energy Technology Data Exchange (ETDEWEB)

Stott, Peter [JET, UK (United Kingdom)

1998-10-01

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

Tokamaks (Second Edition)

International Nuclear Information System (INIS)

Stott, Peter

1998-01-01

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

Fuel and helium confinement in fusion reactors

International Nuclear Information System (INIS)

Houlberg, W.A.; Attenberger, S.E.

1993-01-01

An expanded macroscopic model for particle confinement is used to investigate both fuel and helium confinement in reactor plasmas. The authors illustrate the relative effects of external sources of fuel, divertor pumping, and wall and divertory recycle on core, edge and scrape-off layer densities by using separate particle confinement times for open-quote core close-quote fueling (deep pellet or beam penetration, τ c ), open-quote shallow close-quote fueling (shallow pellet penetration or neutral atoms that penetrate the scrape-off layer, τ s ) and fueling in the scrape-off layer (τ sol ). Because τ s is determined by the parallel flow velocity and characteristic distance to the divertor plate, it can be orders of magnitude lower than either τ c or τ sol . A dense scrape-off region, desirable for reduced divertor erosion, leads to a high fraction of the recycled neutrals being ionized in the scrape-off region and poor core fueling efficiency. The overall fueling efficiency can then be dramatically improved with either shallow or deep auxillary fueling. Helium recycle is nearly always coupled to the scrape-off region and does not lead to strong core accumulation unless the helium pumping efficiency is much less than the fuel pumping efficiency, or the plasma preferentially retains helium over hydrogenic ions. Differences between the results of this model, single-τ p macroscopic models, and 1-D and 2-D models are discussed in terms of assumptions and boundary conditions

Modelling dust transport in tokamaks

International Nuclear Information System (INIS)

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

2008-01-01

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

Compact tokamak reactors

International Nuclear Information System (INIS)

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

1997-01-01

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

Numerical modelling of pump limiter biasing on TEXTOR-94 and Tore Supra

International Nuclear Information System (INIS)

Gerhauser, H.; Claassen, H.A.; Mank, G.; Zagorski, R.; Loarer, T.; Gunn, J.; Boucher, C.

2002-01-01

The two-dimensional multifluid code TECXY has been used to model the biasing (with respect to the first wall) of the toroidal belt limiter ALT-II on the tokamak TEXTOR-94 and of the new toroidal pump limiter being installed on Tore Supra tokamak in the framework of the CIEL project. It is well known that the edge flow pattern can be influenced by the poloidal electric drifts from imposing radial electric fields. The modelling with TECXY introduces imprinted bias currents in the scrape-off layer (SOL) for the case of negative (limiter) biasing, and imprinted bias potentials for the case of positive biasing. This allowed us to simulate sufficiently well the experimental I-V characteristics for either biasing of ALT-II and also reproduced the essential features and trends of the observed plasma profiles in the SOL of TEXTOR-94. For negative biasing a moderate improvement of the pumping exhaust efficiency can be achieved in the case of TEXTOR. For Tore Supra, however, only a negligible improvement of the limiter performance with biasing can be predicted, which is explained by the relatively weak drift flows in Tore Supra. (author)

The AMEMIYA probe. Theoretical background

International Nuclear Information System (INIS)

Belitz, Hans Joahim; Althausen, Bernhard; Uehara, Kazuya; Amemiya, Hiroshi

2010-01-01

The present probe was developed in order to measure the temperature T i of positive ions in the scrape-off layer (SOL) of tokamak where T i is usually larger than the electron temperature Ti so that the presheath in front of the probe need not be considered and the ions reach the probe with the thermal velocity. The axis of the cylindrical probe is placed parallel to the magnetic field. The important parameter are L/a, the ratio of the length to the radius of the cylindrical probe and κ, the ratio of the probe radius to (π/4) 1/2 , where is the mean ion Larmor radius. The ion current densities to the side and the end surfaces are expressed by the double integral, which can give an analytical formula with respect to the value of κ. If two electrodes with different lengths are placed parallel to the magnetic field, the difference of current densities can be reduced to κ and hence to Ti. Some examples of the application of the probe to tokamaks, JFT-2M and Textor, are demonstrated. (author)

Study of plasma discharge evolution and edge turbulence with fast visible imaging in the Aditya tokamak

International Nuclear Information System (INIS)

Banerjee, Santanu; Manchanda, R.; Chowdhuri, M.B.

2015-01-01

Study of discharge evolution through the different phases of a tokamak plasma shot viz., the discharge initiation, current ramp-up, current flat-top and discharge termination, is essential to address many inherent issues of the operation of a Tokamak. Fast visible imaging of the tokamak plasma can provide valuable insight in this regard. Further, edge turbulence is considered to be one of the quintessential areas of tokamak research as the edge plasma is at the immediate vicinity of the plasma core and plays vital role in the core plasma confinement. The edge plasma also bridges the core and the scrape off layer (SOL) of the tokamak and hence has a bearing on the particle and heat flux escaping the plasma column. Two fast visible imaging systems are installed on the Aditya tokamak. One of the system is for imaging the plasma evolution with a wide angle lens covering a major portion of the vacuum vessel. The imaging fiber bundle along with the objective lens is installed inside a radial re-entrant viewport, specially designed for the purpose. Another system is intended for tangential imaging of the plasma column. Formation of the plasma column and its evolution are studied with the fast visible imaging in Aditya. Features of the ECRH and LHCD operations on Aditya will be discussed. 3D filaments can, be seen at the plasma edge all along the discharge and they get amplified in intensity at the plasma termination phase. Statistical analysis of these filaments, which are essentially plasma blobs will be presented. (author)

Magnetic confinement experiment -- 1: Tokamaks

International Nuclear Information System (INIS)

Goldston, R.J.

1994-01-01

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

Microwave Tokamak Experiment: Overview and status

International Nuclear Information System (INIS)

1990-05-01

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

Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation

International Nuclear Information System (INIS)

Riedel, K.S.

1989-01-01

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

Magnetic ''islandography'' in tokamaks

International Nuclear Information System (INIS)

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

1978-09-01

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

Double internal transport barrier triggering mechanism in tokamak plasmas

International Nuclear Information System (INIS)

Dong, Jiaqi; Mou, Zongze; Long, Yongxing; Mahajan, Swadesh M.

2004-01-01

Sheared flow layers created by energy released in magnetic reconnection processes are studied with the magneto hydrodynamics (MHD), aimed at internal transport barrier (ITB) dynamics. The double tearing mode induced by electron viscosity is investigated and proposed as a triggering mechanism for double internal transport barrier (DITB) observed in tokamak plasmas with non-monotonic safety factor profiles. The quasi-linear development of the mode is simulated and the emphasis is placed on the structure of sheared poloidal flow layers formed in the vicinity of the magnetic islands. For viscosity double tearing modes, it is shown that the sheared flows induced by the mode may reach the level required by the condition for ITB formation. Especially, the flow layers are found to form just outside the magnetic islands. The scaling of the generated velocity with plasma parameters is given. Possible explanation for the experimental observations that the preferential formation of transport barriers in the proximity of low order rational surface is discussed. (author)

Current drive by spheromak injection into a tokamak

International Nuclear Information System (INIS)

Brown, M.R.; Bellan, P.M.

1990-01-01

The authors report the first observation of current drive by spheromak injection into a tokamak due to the process of helicity injection. Current drive is observed in Caltech's ENCORE tokamak (30% increase, ΔI > 1 kA) only when both the tokamak and injected spheromak have the same sign of helicity (where helicity is defined as positive if current flows parallel to magnetic field lines and negative if anti-parallel). The initial increase (decrease) in current is accompanied by a sharp decrease (increase) in loop voltage and the increase in tokamak helicity is consistent with the helicity content of the injected spheromak. In addition, the injection of the spheromak raises the tokamak central density by a factor of six. The introduction of cold spheromak plasma causes sudden cooling of the tokamak discharge from 12 eV to 4 eV which results in a gradual decline in tokamak plasma current by a factor of three. In a second experiment, the authors inject spheromaks into the magnetized toroidal vacuum vessel (with no tokamak plasma). An m = 1 magnetic structure forms in the vessel after the spheromak undergoes a double tilt; once in the cylindrical entrance between gun and tokamak, then again in the tokamak vessel. A horizontal shift of the spheromak equilibrium is observed in the direction opposite that of the static toroidal field. In the absence of net toroidal flux, the structure develops a helical pitch as predicted by theory. Experiments with a number of refractory metal coatings have shown that tungsten and chrome coatings provide some improvement in spheromak parameters. They have also designed and will soon construct a larger, higher current spheromak gun with a new accelerator section for injection experiments on the Phaedrus-T tokamak

Snowflake Divertor Configuration in NSTX

International Nuclear Information System (INIS)

Soukhanovskii, V.A.; Ahn, Joonwook; Bell, R.E.; Gates, D.A.; Gerhardt, S.; Kaita, R.; Kolemen, E.; Kugel, H.W.; LeBlanc, B.; Maingi, Rajesh; Maqueda, R.J.; McLean, Adam G.; Menard, J.E.; Mueller, D.; Paul, S.F.; Raman, R.; Roquemore, L.; Ryutov, D.D.; Scott, H.A.

2011-01-01

Steady-state handling of divertor heat flux is a critical issue for present and future conventional and spherical tokamaks with compact high power density divertors. A novel 'snowflake' divertor (SFD) configuration that takes advantage of magnetic properties of a second-order poloidal null has been predicted to have a larger plasma-wetted area and a larger divertor volume, in comparison with a standard first-order poloidal X-point divertor configuration. The SFD was obtained in 0.8 MA, 4-6 MW NBI-heated H-mode discharges in NSTX using two divertor magnetic coils. The SFD led to a partial detachment of the outer strike point even in low-collisionality scrape-off layer plasma obtained with lithium coatings in NSTX. Significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with good core confinement and MHD properties.

'Snowflake' divertor configuration in NSTX

International Nuclear Information System (INIS)

Soukhanovskii, V.A.; Ahn, J.-W.; Bell, R.E.; Gates, D.A.; Gerhardt, S.; Kaita, R.; Kolemen, E.; Kugel, H.W.; LeBlanc, B.P.; Maingi, R.; Maqueda, R.; McLean, A.; Menard, J.E.; Mueller, D.M.; Paul, S.F.; Raman, R.; Roquemore, A.L.; Ryutov, D.D.; Scott, H.A.

2011-01-01

Steady-state handling of divertor heat flux is a critical issue for present and future conventional and spherical tokamaks with compact high power density divertors. A novel 'snowflake' divertor (SFD) configuration that takes advantage of magnetic properties of a second-order poloidal null has been predicted to have a larger plasma-wetted area and a larger divertor volume, in comparison with a standard first-order poloidal X-point divertor configuration. The SFD was obtained in 0.8 MA, 4-6 MW NBI-heated H-mode discharges in NSTX using two divertor magnetic coils. The SFD led to a partial detachment of the outer strike point even in low-collisionality scrape-off layer plasma obtained with lithium coatings in NSTX. Significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with good core confinement and MHD properties.

"Snowflake" divertor configuration in NSTX

Science.gov (United States)

Soukhanovskii, V. A.; Ahn, J.-W.; Bell, R. E.; Gates, D. A.; Gerhardt, S.; Kaita, R.; Kolemen, E.; Kugel, H. W.; Leblanc, B. P.; Maingi, R.; Maqueda, R.; McLean, A.; Menard, J. E.; Mueller, D. M.; Paul, S. F.; Raman, R.; Roquemore, A. L.; Ryutov, D. D.; Scott, H. A.

2011-08-01

Steady-state handling of divertor heat flux is a critical issue for present and future conventional and spherical tokamaks with compact high power density divertors. A novel "snowflake" divertor (SFD) configuration that takes advantage of magnetic properties of a second-order poloidal null has been predicted to have a larger plasma-wetted area and a larger divertor volume, in comparison with a standard first-order poloidal X-point divertor configuration. The SFD was obtained in 0.8 MA, 4-6 MW NBI-heated H-mode discharges in NSTX using two divertor magnetic coils. The SFD led to a partial detachment of the outer strike point even in low-collisionality scrape-off layer plasma obtained with lithium coatings in NSTX. Significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with good core confinement and MHD properties.

Recent progress towards a quantitative description of filamentary SOL transport

DEFF Research Database (Denmark)

Carralero, D.; Siccinio, M.; Komm, M.

2017-01-01

A summary of recent results on filamentary transport, mostly obtained with the ASDEX-Upgrade tokamak (AUG), is presented and discussed in an attempt to produce a coherent picture of scrape-off layer (SOL) filamentary transport. A clear correlation is found between L-mode density shoulder formation...... in the outer midplane and a transition between the sheath-limited and the inertial filamentary regimes. Divertor collisionality is found to be the parameter triggering the transition. A clear reduction of the ion temperature takes place in the far SOL after the transition, both for the background......-mode plasmas, although a secondary threshold related to deuterium fueling is observed. EMC3-EIRENE simulations of neutral dynamics show that an ionization front near the main chamber wall is formed after the shoulder formation. Finally, a clear increase of SOL opacity to neutrals is observed, associated...

Edge Zonal Flows and Blob Propagation in Alcator C-Mod

International Nuclear Information System (INIS)

Zweben, S.; Terry, J.L.; Agostini, M.; Davis, B.; Grulke, O.; Hager, R.; Hughes, J.; LaBombard, B.; D'Ippolito, D.A.; Myra, J.R.; Russell, D.A.

2011-01-01

Here we describe recent measurements of the 2-D motion of turbulence in the edge and scrape-off layer (SOL) of the Alcator C-Mod tokamak. This data was taken using the outer midplane gas puff imaging (GPI) camera, which views a 6 cm radial by 6 cm poloidal region near the separatrix just below the outer midplane [1]. The data were taken in Ohmic or RF heated L-mode plasmas at 400,000 frames/sec for ∼50 msec/shot using a Phantom 710 camera in a 64 x 64 pixel format. The resulting 2-D vs. time movies [2] can resolve the structure and motion of the turbulence on a spatial scale covering 0.3-6 cm. The images were analyzed using either a 2-D cross-correlation code (Sec. 2) or a 2-D blob tracking code (Sec. 3).

Conservation Laws for Gyrokinetic Equations for Large Perturbations and Flows

Science.gov (United States)

Dimits, Andris

2017-10-01

Gyrokinetic theory has proved to be very useful for the understanding of magnetized plasmas, both to simplify analytical treatments and as a basis for efficient numerical simulations. Gyrokinetic theories were previously developed in two extended orderings that are applicable to large fluctuations and flows as may arise in the tokamak edge and scrapeoff layer. In the present work, we cast the resulting equations in a field-theoretical variational form, and derive, up to second order in the respective orderings, the associated global and local energy and (linear and toroidal) momentum conservation relations that result from Noether's theorem. The consequences of these for the various possible choices of numerical discretization used in gyrokinetic simulations are considered. Prepared for US DOE by LLNL under Contract DE-AC52-07NA27344 and supported by the U.S. DOE, OFES.

Particle balance in diverted plasmas in TEXT-U

International Nuclear Information System (INIS)

Rowan, W.L.; Bengtson, R.D.; Bonnin, X.; Edmonds, P.H.; Hurwitz, P.D.; Solano, E.R.; Tsui, H.Y.W.; Uglum, J.R.; Wootton, A.J.

1995-01-01

Tokamak plasmas in an open divertor configuration with the X-point at the inner equator were compared with plasmas limited by an inner toroidal belt. This paper describes and compares the particle balance in these two types of discharge. The plasma parameters n e (r), φ(r), T e (r), n e , and φ were measured in the scrape-off layer and in the plasma periphery at one poloidal location and mapped onto the rest of the plasma by assuming constancy on flux surfaces. Emission from neutral hydrogen was measured throughout the plasma. The particle source and then the global particle confinement were inferred from these measurements using a 3-D neutral transport simulation. The SOL profiles are significantly steeper in the diverted discharge. Both the source measurements and estimates from the SOL profiles indicate significantly better confinement in the diverted discharge. ((orig.))

Filament structures at the plasma edge on MAST

International Nuclear Information System (INIS)

Kirk, A; Ayed, N Ben; Counsell, G; Dudson, B; Eich, T; Herrmann, A; Koch, B; Martin, R; Meakins, A; Saarelma, S; Scannell, R; Tallents, S; Walsh, M; Wilson, H R

2006-01-01

The boundary of the tokamak core plasma, or scrape-off layer, is normally characterized in terms of average parameters such as density, temperature and e-folding lengths suggesting diffusive losses. However, as is shown in this paper, localized filamentary structures play an important role in determining the radial efflux in both L mode and during edge localized modes (ELMs) on MAST. Understanding the size, poloidal and toroidal localization and the outward radial extent of these filaments is crucial in order to calculate their effect on power loading both on the first wall and the divertor target plates in future devices. The spatial and temporal evolution of filaments observed on MAST in L-mode and ELMs have been compared and contrasted in order to confront the predictions of various models that have been proposed to predict filament propagation and in particular ELM energy losses

Plasma regimes and research goals of JT-60SA towards ITER and DEMO

International Nuclear Information System (INIS)

Kamada, Y.; Ide, S.; Fujita, T.; Suzuki, T.; Matsunaga, G.; Yoshida, M.; Shinohara, K.; Urano, H.; Nakano, T.; Sakurai, S.; Kawashima, H.; Barabaschi, P.; Lackner, K.; Ishida, S.; Bolzonella, T.

2011-01-01

The JT-60SA device has been designed as a highly shaped large superconducting tokamak with a variety of plasma actuators (heating, current drive, momentum input, stability control coils, resonant magnetic perturbation coils, W-shaped divertor, fuelling, pumping, etc) in order to satisfy the central research needs for ITER and DEMO. In the ITER- and DEMO-relevant plasma parameter regimes and with DEMO-equivalent plasma shapes, JT-60SA quantifies the operation limits, plasma responses and operational margins in terms of MHD stability, plasma transport and confinement, high-energy particle behaviour, pedestal structures, scrape-off layer and divertor characteristics. By integrating advanced studies in these research fields, the project proceeds 'simultaneous and steady-state sustainment of the key performances required for DEMO' with integrated control scenario development applicable to the highly self-regulating burning high-β high bootstrap current fraction plasmas.

Impurity penetration and transport during VH-mode on DIII-D

International Nuclear Information System (INIS)

Lippmann, S.I.; Evans, T.E.; Jackson, G.L.; West, W.P.

1992-05-01

A new modeling effort is made in order to understand the observed relatively low levels of impurity contamination during the VH-mode phase on DIII-D, as compared to those observed during the H-mode phase of selected discharges. The key element is the inclusion of the real 2-D flux surface geometry in the prediction of impurity penetration of sputtered atoms through the scrape-off layer into the core plasma. Of the elements which determine the impurity content in the plasma: sputtering yield, penetration, and core transport, the penetration through the scrape-off layer is found to be the most determinative factor. The low impurity content in VH-mode is attributed to the development of a scrape-off layer with higher density and temperature properties than those normally obtained in H-mode

A need for non-tokamak approaches to magnetic fusion energy

International Nuclear Information System (INIS)

Bathke, C.G.; Krakowski, R.A.; Miller, R.L.

1992-01-01

Focusing exclusively on conventional tokamak physics in the quest for commercial fusion power is premature, and the options for both advanced-tokamak and non-tokamak concepts need continued investigation. The basis for this claim is developed, and promising advanced-tokamak and non-tokamak options are suggested

Submillimeter wave propagation in tokamak plasmas

International Nuclear Information System (INIS)

Ma, C.H.; Hutchinson, D.P.; Staats, P.A.; Vander Sluis, K.L.; Mansfield, D.K.; Park, H.; Johnson, L.C.

1985-01-01

The propagation of submillimeter-waves (smm) in tokamak plasmas has been investigated both theoretically and experimentally to ensure successful measurements of electron density and plasma current distributions in tokamak devices. Theoretical analyses have been carried out to study the polarization of the smm waves in TFTR and ISX-B tokamaks. A multichord smm wave interferometer/polarimeter system has been employed to simultaneously measure the line electron density and poloidal field-induced Faraday rotation in the ISX-B tokamak. The experimental study on TFTR is under way. Computer codes have been developed and have been used to study the wave propagation and to reconstruct the distributions of plasma current and density from the measured data. The results are compared with other measurements

Submillimeter wave propagation in tokamak plasmas

International Nuclear Information System (INIS)

Ma, C.H.; Hutchinson, D.P.; Staats, P.A.; Vander Sluis, K.L.; Mansfield, D.K.; Park, H.; Johnson, L.C.

1986-01-01

Propagation of submillimeter waves (smm) in tokamak plasma was investigated both theoretically and experimentally to ensure successful measurements of electron density and plasma current distributions in tokamak devices. Theoretical analyses were carried out to study the polarization of the smm waves in TFTR and ISX-B tokamaks. A multichord smm wave interferometer/polarimeter system was employed to simultaneously measure the line electron density and poloidal field-induced Faraday rotation in the ISX-B tokamak. The experimental study on TFTR is under way. Computer codes were developed and have been used to study the wave propagation and to reconstruct the distributions of plasma current and density from the measured data. The results are compared with other measurements. 5 references, 2 figures

The role of the spherical tokamak in clarifying tokamak physics

International Nuclear Information System (INIS)

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

1999-01-01

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

X transport and its effect on H-mode and edge pedestal in tokamaks

International Nuclear Information System (INIS)

Chang, C.S.; Darrow, D.; White, R.; Lin, Z.; Lee, W.; Ku, S.H.; Weitzner, H.; Carlstrom, T.N.; Grassie, J.S. de

2001-01-01

A new classical non-ambipolar transport mechanism has been identified which can be a dominant source of strong Er and edge pedestal layer formation immediately inside the separatrix in a diverted tokamak. Due to vanishingly small poloidal B-field and grad-B drift toward x-point, plasma ions with small ν parallel in the X-region do not have confined single particle orbits. This leads to a non-ambipolar convective transport in the X-region (X-transport), either collisional or collisionless, inducing a strong negative Er-shear layer. The X-transport can provide basic understanding of many of the experimental observations. (author)

Steady state technologies for tokamak based fusion neutron sources and hybrids

International Nuclear Information System (INIS)

Azizov, E.A.; Kuteev, B.V.

2015-01-01

injection technology is proposed to use for control of the border plasma radiation and plasma-surface interaction in the scrape-off layer [6]. Concepts of the FNS blankets for pure thermal neutron production and for development of thorium fuel cycle for fission reactors have been proposed and considered. It was shown that thermal neutron fluxes as large as 10 15 n/cm 2 /s are feasible in FNS with Be-central stack. The radial structure, neutronics and thermal hydraulic characteristics as well as the U 233 -production rate and opportunities to self-breed tritium in it have been defined. References: [1] Kuteev, B.V. et al., 2011 Nucl. Fusion 51 073013; [2] Kuteev, B.V. et al., 2012 IAEA FEC-24 FTP/P7-07; [3] Goncharov, P.R. et al. VANT, series Thermonuclear Fusion, 2011, issue 2. p. 43; [4] Golikov, A.A. et al. VANT, series Thermonuclear Fusion, 2011, issue 1. p. 1, and 2012, issue 1, p.43; [5] Dnestrovskij, A.Yu. et al. VANT, series Thermonuclear Fusion, 2011, issue 1. p. 48; [6] Sergeev, V.Yu. et al. Plasma Physics Reports, 2012, issue 5. p. 235. (authors)

Hybrid neural network for density limit disruption prediction and avoidance on J-TEXT tokamak

Science.gov (United States)

Zheng, W.; Hu, F. R.; Zhang, M.; Chen, Z. Y.; Zhao, X. Q.; Wang, X. L.; Shi, P.; Zhang, X. L.; Zhang, X. Q.; Zhou, Y. N.; Wei, Y. N.; Pan, Y.; J-TEXT team

2018-05-01

Increasing the plasma density is one of the key methods in achieving an efficient fusion reaction. High-density operation is one of the hot topics in tokamak plasmas. Density limit disruptions remain an important issue for safe operation. An effective density limit disruption prediction and avoidance system is the key to avoid density limit disruptions for long pulse steady state operations. An artificial neural network has been developed for the prediction of density limit disruptions on the J-TEXT tokamak. The neural network has been improved from a simple multi-layer design to a hybrid two-stage structure. The first stage is a custom network which uses time series diagnostics as inputs to predict plasma density, and the second stage is a three-layer feedforward neural network to predict the probability of density limit disruptions. It is found that hybrid neural network structure, combined with radiation profile information as an input can significantly improve the prediction performance, especially the average warning time ({{T}warn} ). In particular, the {{T}warn} is eight times better than that in previous work (Wang et al 2016 Plasma Phys. Control. Fusion 58 055014) (from 5 ms to 40 ms). The success rate for density limit disruptive shots is above 90%, while, the false alarm rate for other shots is below 10%. Based on the density limit disruption prediction system and the real-time density feedback control system, the on-line density limit disruption avoidance system has been implemented on the J-TEXT tokamak.

Possibility of image converter application to X ray diagnostics in tokamaks

International Nuclear Information System (INIS)

Bryzgunov, V.A.; Chuvatin, S.A.

1980-01-01

To improve the sensitivity of spectral roentgen diagnostics of tokamak plasma possibilities for using a light amplifier having a PMU-1 microchannel plate for photorecording of roentgen spectra have been estimated. Amplification homogeneity over an image field has been improved 5-10 times by means of a screen. In bench tests spectra with an exposure 100 times lesser as compared with the RT-1 roentgen film have been taken. Five-layer protective screen has been developed for experiments in scattered magnetic fields of the tokamak. It is concluded that from the point of view of a number of parameters the PMU-1 electron optical converter quite meets the requirements of roentgen diagnostics. It has a sufficient amplification, small dark noises, small distortion. However, it is very desirable to have an instrument of a better resolution as compared with an obtained one as well as having a more light and comfortable magnetic protection

Advanced tokamak burning plasma experiment

International Nuclear Information System (INIS)

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

2001-01-01

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

Computational studies of tokamak plasmas

International Nuclear Information System (INIS)

Takizuka, Tomonori; Tsunematsu, Toshihide; Tokuda, Shinji

1981-02-01

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

Combined confinement system applied to tokamaks

International Nuclear Information System (INIS)

Ohkawa, Tihiro

1986-01-01

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

Experimental studies of thermal and non-thermal electron cyclotron phenomena in tokamaks

International Nuclear Information System (INIS)

McDermott, F.S.

1984-12-01

A direct measurement of wave absorption in the ISX-B tokamak at the second harmonic of the electron cyclotron frequency is reported. Measurements of the absorption of a wave polarized in the extraordinary mode and propagating perpendicular to the toroidal magnetic field are in agreement with the absorption predicted by the linearized Vlasov equation for a thermal plasma. Agreement is found both for an analytic approximation to the wave absorption and for a numerical simulation of ray propagation in toroidal geometry. Observations are also reported on a non-linear, three-wave interaction process occurring during high power electron cyclotron resonance heating in the Versator II tokamak. The measured spectra and the threshold power are consistent with a model in which the incident power in the extraordinary mode of polarization decays at the upper hybrid resonance layer into a lower hybrid wave and an electron Bernstein wave. Finally, measurements of non-thermal emission at the second harmonic of the electron cyclotron frequency and below the electron plasma frequency are reported from low density, non-Maxwellian plasma in the Versator II tokamak. The emission spectra are in agreement with a model in which waves are driven unstable at the anomalous Doppler resonance, while only weakly damped at the Cerenkov resonance

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.)

Technical aspects of the joint JET-ISX-B beryllium limiter experiment

International Nuclear Information System (INIS)

Edmonds, P.H.; Dietz, K.J.; Mioduszewski, P.K.; Watson, R.D.; Emerson, L.C.; Gabbard, W.A.; Goodall, D.; Simpkins, J.E.; Yarber, J.L.

1984-01-01

An experiment has been performed on the Impurity Study Experiment (ISX-B) tokamak to test beryllium as a limiter material. Beryllium is an attractive candidate for a limiter and has been proposed for use in the Joint European Torus (JET) experiment. A temperature-controlled, segmented, beryllium top-rail limiter was located inside the plasma radius described by the existing titanium limiters. An extended set of diagnostics was added for measurement of scrapeoff and limiter parameters. These included visible and infrared monitoring systems, probes, and surface analysis experiments. Tokamak experiments included parameter surveys of both ohmically heated and neutral-beam-heated plasmas and an extended fluence test of the limiter

Critical condition for current-driven instability excited in turbulent heating of TRIAM-1 tokamak plasma

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Y; Watanabe, T; Nagao, A; Nakamura, K; Kikuchi, M; Aoki, T; Hiraki, N; Itoh, S [Kyushu Univ., Fukuoka (Japan). Research Inst. for Applied Mechanics; Mitarai, O

1982-02-01

Critical condition for current-driven instability excited in turbulently heated TRIAM-1 tokamak plasma is investigated experimentally. Resistive hump in loop voltage, plasma density fluctuation and rapid increase of electron temperature in a skin layer are simultaneously observed at the time when the electron drift velocity amounts to the critical drift velocity for low-frequency ion acoustic instability.

Research into controlled fusion in tokamaks

International Nuclear Information System (INIS)

Zacek, F.

1992-01-01

During the thirty years of tokamak research, physicists have been approaching step by step the reactor breakeven condition defined by the Lawson criterion. JET, the European Community tokamak is probably the first candidate among the world largest tokamaks to reach the ignition threshold and thus to demonstrate the physical feasibility of thermonuclear reaction. The record plasma parameters achieved in JET at H plasma modes due to powerful additional plasma heating and due to substantial reduction of plasma impurities, opened the door to the first experiment with a deuterium-tritium plasma. In the paper, the conditions and results of these tritium experiments are described in detail. The prospects of the world tokamak research and of the participation of Czechoslovak physicists are also discussed. (J.U.) 3 figs., 6 refs

Poloidal inhomogeneity of the particle fluctuation induced fluxes near of the LCFS at lower hybrid heating and improved confinement transition at the FT-2 tokamak

International Nuclear Information System (INIS)

Lashkul, S.I.; Altukhov, A.B.; Gurchenko, A.D.; Gusakov, E.Z.; Dyachenko, V.V.; Esipov, L.A.; Kantor, M.Y.; Kouprienko, D.V.; Stepanov, A.Y.; Sharpeonok, A.P.; Shatalin, S.V.; Vekshina, E.O.

2004-01-01

This paper present our observations and conclusions about development of the transport process at the plasma periphery of the small tokamak FT-2 during additional Lower Hybrid Heating (LHH), when external (ETB) transport barrier followed by Internal (ITB) transport barrier is observed. The peculiarities of the variations of the fluctuation fluxes near periphery are measured by three moveable multi-electrode Langmuir probes (L-probe) located in the same poloidal cross-section of the chamber. So the observed L-H transition and ETB formation after LHH and the associated negative E r rise result mainly from the decrease of the electron temperature (T e ) near inner region of the LCFS (last close flux surface) by greater extent than in SOL (scrape-off layer). This effect is stimulated by decrease of the input power and decrease of the radial correlation coefficient (for r equals 74-77 mm) (and radial particle fluctuation-induced Γ(t)) resulted from ITB formation mechanism during LHH. T e variation in the SOL after LH heating pulse takes place to a lesser extent. Observed non-monotonic radial profile of T e near LCFS with positive δT e /δr rise is kept constant obviously by large longitudinal conductivity and poloidal fluxes from the hotter limiter shadow regions because of the poloidal inhomogeneity of the T e (SOL) and n e (SOL). Such induced negative E r after RF pulse gives fast rise to a quasi-steady-state Γ 0 (t) drift fluxes with reversed direction structure, like 'zonal flows', which may inhibit transport across the flow. Large rise of grad(n e ) after LHH near LCFS with L-H transition is observed after the end of LH pulse for a long time - about 10-15 ms

A study on the heating and diagnostic of a tokamak plasma by electromagnetic waves of the electron cyclotron range of frequencies

International Nuclear Information System (INIS)

Hoshino, Katsumichi

1989-09-01

A study on the heating and diagnosis of tokamak plasma by electromagnetic waves of electron cyclotron range of frequency is summarized. The main results obtained are as follows. On the engineering and technology, the technology of injecting high frequency, large power millimeter waves into tokamak plasma was established by carrying out the design, manufacture and test of a 60 GHz, 400 kW high frequency heating system, and the design, manufacture and test of a heterodyne type electron cyclotron radiation multi-channel mealsuring system were carried out, and the technology of measuring the radiation from tokamak plasma with the time resolution of 10 μs in multi-channel was established. On nuclear fusion reactor core engineering and plasma physics, the high efficiency electron heating of tokamak plasma by the incidence of fundamental irregular and regular waves at electron cyclotron frequency was verified. The discovery and analysis of the heating by electrostatic waves arising due to mode transformation from electromagnetic waves in upper hybrid resonance layer were carried out. By the incidence of second harmonic waves, the high efficiency electron heating of tokamak plasma was verified, and the heating characteristics were clarified. And others. (K.I.) 179 refs

The density limit in Tokamaks

International Nuclear Information System (INIS)

Alladio, F.

1985-01-01

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

Kinetic theory of plasma in the limiter-scrape-off layer

International Nuclear Information System (INIS)

Daybelge, U.; Bein, B.

1977-01-01

An asymptotic solution is given for the ion-drift-kinetic equation with a full Fokker--Planck term for the limiter-scrape-off layer in a tokamak. In this layer, the plasma is assumed to consist of hot, collisionless ions, and cold, collisional electrons. From the solution of the boundary-layer problem, ion and electron particle and energy losses to the limiter are calculated. Limiter load profiles due to ions are explicitly given as functions of the poloidal angle

Turbulence induced radial transport of toroidal momentum in boundary plasma of EAST tokamak

International Nuclear Information System (INIS)

Zhao, N.; Yan, N.; Xu, G. S.; Wang, H. Q.; Wang, L.; Ding, S. Y.; Chen, R.; Chen, L.; Zhang, W.; Hu, G. H.; Shao, L. M.; Wang, Z. X.

2016-01-01

Turbulence induced toroidal momentum transport in boundary plasma is investigated in H-mode discharge using Langmuir-Mach probes on EAST. The Reynolds stress is found to drive an inward toroidal momentum transport, while the outflow of particles convects the toroidal momentum outwards in the edge plasma. The Reynolds stress driven momentum transport dominates over the passive momentum transport carried by particle flux, which potentially provides a momentum source for the edge plasma. The outflow of particles delivers a momentum flux into the scrape-off layer (SOL) region, contributing as a momentum source for the SOL flows. At the L-H transitions, the outward momentum transport suddenly decreases due to the suppression of edge turbulence and associated particle transport. The SOL flows start to decelerate as plasma entering into H-mode. The contributions from turbulent Reynolds stress and particle transport for the toroidal momentum transport are identified. These results shed lights on the understanding of edge plasma accelerating at L-H transitions.

Development of Operation Scenario for Spherical Tokamak at SNU

International Nuclear Information System (INIS)

Sung, C. K.; Park, Y. S.; Lee, H. Y.; Kang, J.; Hwang, Y. S.

2009-01-01

Several concepts for nuclear fusion plant exist. In these concepts, tokamak is the most promising one to realize nuclear fusion plant. Though tokamak has leading concept, and this has world record in fusion heating power, tokamak has the critical drawback: low heating efficiency. That is the reason why we need another alternative concept which compensates tokamak's disadvantage. Spherical Torus(ST) is one of these kinds of concepts. ST is a kind of tokamak which has low aspect ratio. This feature gives ST advantages compared to conventional tokamak: high efficiency, compactness, low cost. However, ST lacks central region for solenoid that is needed to start-up and sustain. Since it is the most efficient that initializing and sustaining by using solenoid, this is ST's intrinsic limitation. To overcome this, a new device which can start-up and sustain ST plasmas by means of continuous tokamak plasma injection has been designed

Economic considerations of commercial tokamak options

International Nuclear Information System (INIS)

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 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 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)

Magnetic propulsion of intense lithium streams in a tokamak magnetic field

International Nuclear Information System (INIS)

Zakharov, Leonid E.

2003-01-01

This paper describes the effect and gives the theory of magnetic propulsion which allows driving free surface plasma facing liquid lithium streams in tokamaks. In the approximation of a thin flowing layer the MHD equations are reduced to one integrodifferential equation which takes into account the propulsion effect, viscosity, and the drag force due to magnetic pumping and other interactions with the magnetic field. A stability criterion is obtained for stabilization of the 'sausage' instability of the streams by centrifugal force

Large aspect ratio tokamak study

International Nuclear Information System (INIS)

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

1979-01-01

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

Nuclear fusion research at Tokamak Energy Ltd

International Nuclear Information System (INIS)

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

2017-01-01

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

Study of the heavy impurity influx into the plasma using laser fluorescence spectroscopy in the TO-2 tokamak with toroidal divertor

International Nuclear Information System (INIS)

Vukolov, K.Yu.; Shvindt, N.N.

1992-01-01

Measurement cycle for determination of iron atom absolute concentrations was carried out in divertor and diaphragm modes of laser fluorescence spectroscopy. The conclusion is made on effective wall shielding by divertor layer as compared to material diaphragm. The basic result of the work consists in creating and testing on the tokamak TO-2 of multichannel diagnostic complex for remote measurement of atom (ion) absolute concentrations of metallic impurities in the near-wall plasma with high spatial and time resolution through laser fluorescence spectroscopy method intended for studies at the Tokamak-15 facility

Tokamak simulation code manual

International Nuclear Information System (INIS)

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

1995-01-01

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

Mercier criterion for high-β tokamaks

International Nuclear Information System (INIS)

Galvao, R.M.O.

1984-01-01

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

Properties of the TEXTOR boundary layer

International Nuclear Information System (INIS)

Bogen, P.; Hartwig, H.; Hintz, E.; Hoethker, K.; Lie, Y.T.; Pospieszczyk, A.; Samm, U.

1984-01-01

First measurements on the TEXTOR boundary layer are reported. The hydrogen recycling in front of the four limiter segments has been studied by means of a CCD-camera, which proved to be a good instrument to center the discharge for symmetric plasma-limiter contact. The composition of the neutral fluxes from the limiter have been measured: oxygen fluxes are about a factor of ten higher than the metal fluxes; within the error limits the composition does not change with varying limiter radius. Electron densities in the scrape-off layer away from the limiter have been determined by injecting an Li-atom beam from a thermal source and by observing its emission as a function of radius. Similar measurements have been made in front of the limiter with sputtered Cr and O atoms. Both methods gave for the magnetic surface of the limiter radius nsub(e) approx.= 1 x 10 12 /cm 23 . Infrared observations of a test limiter with a CCD-camera and a PbSe-detector have been performed to record the thermal loads. About 10% of the input power flows to the limiter. (orig.)

Tokamak Physics Experiment (TPX) design

International Nuclear Information System (INIS)

Schmidt, J.A.

1995-01-01

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

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

International Nuclear Information System (INIS)

Thomassen, K.; Goldston, R.; Nevins, B.; Neilson, H.; Shannon, T.; Montgomery, B.

1992-03-01

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

Concept study of the Steady State Tokamak Reactor (SSTR)

International Nuclear Information System (INIS)

1991-06-01

The Steady State Tokamak Reactor (SSTR) concept has been proposed as a realistic fusion power reactor to be built in the near future. An overall concept of SSTR is introduced which is based on a small extension of the present day physics and technologies. The major feature of SSTR is the maximum utilization of a bootstrap current in order to reduce the power required for the steady state operation. This requirement leads to the choice of moderate current (12 MA), and high βp (2.0) for the device, which are achieved by selecting high aspect ratio (A=4) and high toroidal magnetic field (16.5 T). A negative-ion-based neutral beam injection system is used both for heating and central current drive. Notable engineering features of SSTR are: the use of a uniform vacuum vessel and periodical replacements of the first wall and blanket layers and significant reduction of the electromagnetic force with the use of functionally gradient material. It is shown that a tokamak machine comparable to ITER in size can become a power reactor capable of generating about 1 GW of electricity with a plant efficiency of ∼30%. (author)

An emerging understanding of H-mode discharges in tokamaks

International Nuclear Information System (INIS)

Groebner, R.J.

1992-12-01

A remarkable degree of consistency of experimental results from tokamaks throughout the world has developed with regard to the phenomenology of the transition from L-mode to H-mode confinement in tokamaks. The transition is initiated in a narrow layer at the plasma periphery where density fluctuations are suppressed and steep gradients of temperature and density form in a region with large first and second radial derivatives in the υ E → = (E x B)/B 2 flow velocity. These results are qualitatively consistent with theories which predict suppression of fluctuations by shear or curvature in υE. The required υE flow is generated very rapidly when the magnitude of the heating power or of an externally imposed radial current exceed threshold values and several theoretical models have been developed to explain the observed changes in the υE flow. After the transition occurs, the altered boundary conditions enable the development of improved confinement in the plasma interior on a confinement time scale. The resulting H-mode discharge has typically twice the confinement of L-mode discharges and regimes of further improved confinement have been obtained in some H-mode scenarios

Summary of the 1982 small tokamak users meeting

International Nuclear Information System (INIS)

Sprott, J.C.

1982-11-01

On November 1, 1982, the sixth in a series of approximately annual meetings of the users of small tokamaks was held in conjunction with the APS Division of Plasma Physics meeting at New Orleans. The meeting lasted three hours, with 34 people attending. The interest was on strengthening the ties between the small tokamaks and the large tokamaks. Accordingly, the latest meeting was dedicated to this theme, and in contrast to previous meetings, a few representatives from the large tokamaks were invited to attend and make presentations. Summaries of the various talks are included

Magnetic confinement by Tokamak: physical aspects

International Nuclear Information System (INIS)

Tachon, J.

1980-01-01

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

Tokamak research in the Soviet Union

International Nuclear Information System (INIS)

Strelkov, V.S.

1981-01-01

Important milestones on the way to the tokamak fusion reactor are recapitulated. Soviet tokamak research concentrated at the I.V. Kurchatov Institute in Moscow, the A.F. Ioffe Institute in Leningrad and the Physical-Technical Institute in Sukhumi successfully provides necessary scientific and technological data for reactor design. Achievments include, the successful operation of the first tokamak with superconducting windings (T-7) and the gyrotron set for microwave plasma heating in the T-10 tokamak. The following problems have intensively been studied: Various methods of additional plasma heating, heat and particle transport, and impurity control. The efficiency of electron-cyclotron resonance heating was demonstrated. In the Joule heating regime, both the heat conduction and diffusion rates are anomalously high, but the electron heat conduction rate decreases with increasing plasma density. Progress in impurity control makes it possible to obtain a plasma with effective charge approaching unity. (J.U.)

Neutral beam in ALVAND IIC tokamak

International Nuclear Information System (INIS)

Ghrannevisse, M.; Moradshahi, M.; Avakian, M.

1992-01-01

Neutral beams have a wide application in tokamak experiments. It used to heat; fuel; adjust electric potentials in plasmas and diagnose particles densities and momentum distributions. It may be used to sustain currents in tokamaks to extend the pulse length. A 5 KV; 500 mA ion source has been constructed by plasma physics group, AEOI and it used to produce plasma and study the plasma parameters. Recently this ion source has been neutralized and it adapted to a neutral beam source; and it used to heat a cylindrical DC plasma and the plasma of ALVAND IIC Tokamak which is a small research tokamak with a minor radius of 12.6 cm, and a major radius of 45.5 cm. In this paper we report the neutralization of the ion beam and the results obtained by injection of this neutral beam into plasmas. (author) 2 refs., 4 figs

Model of divertor biasing and control of scrape-off layer and divertor plasmas

International Nuclear Information System (INIS)

Nagasaki, K.; Itoh, K.; Itoh, S.

1991-02-01

Analytic model of the divertor biasing is described. For the given plasma and energy sources from the core plasma, the heat and particle flux densities on the divertor plate as well as scrape-off-layer (SOL)/divertor plasmas are analyzed in a slab model. Using a two-dimensional model, the effects of the divertor biasing and SOL current are studied. The conditions to balance the plasma temperature or sheath potential on different divertor plates are obtained. Effect of the SOL current on the heat channel width is also discussed. (author)

One dimensional simulation on stability of detached plasma in a tokamak divertor

International Nuclear Information System (INIS)

Nakazawa, Shinji; Nakajima, Noriyoshi; Okamoto, Masao; Ohyabu, Nobuyoshi

1999-06-01

The stability of radiation front in the Scrape-Off-Layer (SOL) of a tokamak is studied with a one dimensional fluid code; the time-dependent transport equations are solved in the direction parallel to a magnetic field line. The simulation results show that stable detached solutions exist, where the plasma temperature near the divertor target is ∼2 eV. It is found that whenever such stable detached states are attained, the strong radiation front is contact with or at a small distance from the divertor target. When the energy externally injected into the SOL is decreased below a critical value, the radiation front starts to move towards the X-point, cooling the SOL plasma. In such cases, no stationary solutions such that the radiation front rests in the divertor channel are observed in our parameter space. This qualitatively corresponds to the results of tokamak divertor experiments which show the movement of radiation front. (author)