Effects of Density and Impurity on Edge Localized Modes in Tokamaks

Science.gov (United States)

Zhu, Ping

2017-10-01

Plasma density and impurity concentration are believed to be two of the key elements governing the edge tokamak plasma conditions. Optimal levels of plasma density and impurity concentration in the edge region have been searched for in order to achieve the desired fusion gain and divertor heat/particle load mitigation. However, how plasma density or impurity would affect the edge pedestal stability may have not been well known. Our recent MHD theory modeling and simulations using the NIMROD code have found novel effects of density and impurity on the dynamics of edge-localized modes (ELMs) in tokamaks. First, previous MHD analyses often predict merely a weak stabilizing effect of toroidal flow on ELMs in experimentally relevant regimes. We find that the stabilizing effects on the high- n ELMs from toroidal flow can be significantly enhanced with the increased edge plasma density. Here n denotes the toroidal mode number. Second, the stabilizing effects of the enhanced edge resistivity due to lithium-conditioning on the low- n ELMs in the high confinement (H-mode) discharges in NSTX have been identified. Linear stability analysis of the experimentally constrained equilibrium suggests that the change in the equilibrium plasma density and pressure profiles alone due to lithium-conditioning may not be sufficient for a complete suppression of the low- n ELMs. The enhanced resistivity due to the increased effective electric charge number Zeff after lithium-conditioning provides additional stabilization of the low- n ELMs. These new effects revealed in our theory analyses may help further understand recent ELM experiments and suggest new control schemes for ELM suppression and mitigation in future experiments. They may also pose additional constraints on the optimal levels of plasma density and impurity concentration in the edge region for H-mode tokamak operation. Supported by National Magnetic Confinement Fusion Science Program of China Grants 2014GB124002 and 2015GB

Impurity seeding in ASDEX upgrade tokamak modeled by COREDIV code

Energy Technology Data Exchange (ETDEWEB)

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

2016-08-15

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

Impurity screening of scrape-off plasma in a tokamak

International Nuclear Information System (INIS)

Kishimoto, Hiroshi; Tani, Keiji; Nakamura, Hiroo

1981-11-01

Impurity screening effect of a scrape-off layer has been studied in a tokamak, based on a simple model of wall-released impurity behavior. Wall-sputtered impurities are stopped effectively by the scrape-off plasma for a medium-Z or high-Z wall system while major part of impurities enters the main plasma in a low-Z wall system. The screening becomes inefficient with increase of scrape-off plasma temperature. Successive multiplication of recycling impurities in the scrape-off layer is large for a high-Z wall and is enhanced by a rise of scrape-off plasma temperature. The stability of plasma-wall interaction is determined by a multiplication factor of recycling impurities. (author)

Poloidal density variation of impurities in a rotating tokamak plasma - flux surface coordinates and effect on transport coefficients

International Nuclear Information System (INIS)

Romanelli, M.

1999-09-01

The poloidal variation of impurity densities over magnetic surfaces brings about an enhancement of neoclassical transport coefficients, as shown by Romanelli and Ottaviani for impurities in the Pfirsch Schlueter regime and by Helander for particles in the banana-plateau regime, both in a large aspect ratio tokamak. The same effect will occur in a finite aspect ratio tokamak and therefore it is considered to be relevant for inclusion in transport codes for comparison with the experimental measurements of impurity transport. Here an expression for the impurity-density poloidal-variation generated by the fast toroidal rotation of the plasma column is presented in general coordinates. (author)

The impurity transport in HT-6M tokamak

International Nuclear Information System (INIS)

Xu Wei; Wan Baonian; Xie Jikang

2003-01-01

The space-time profile of impurities has been measured with a multichannel visible spectroscopic detect system and UV rotation-mirror system in the HT-6M tokamak. An ideal impurity transport code has been used to simulate impurities (carbon and oxygen) behaviour during the OHM discharge. The profiles of impurities diffusion and convection coefficient, impurities ion densities in different ionized state, loss power density and effective charge number have been derived. The impurity behaviour during low-hybrid current drive has also been analyzed, the results show that the confinement of particles, impurities and energy has been improved, and emission power and effective charge number have been reduced

Texas Experimental Tokamak

International Nuclear Information System (INIS)

Wootton, A.J.

1990-04-01

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

Electron and impurity transport studies in the TCV Tokamak

Energy Technology Data Exchange (ETDEWEB)

Wagner, D.

2013-05-15

In this thesis electron and impurity transport are studied in the Tokamak à Configuration Variable (TCV) located at CRPP-EPFL in Lausanne. Understanding particle transport is primordial for future nuclear fusion power plants. Modeling of experiments in many specific plasma scenarios can help to understand the common elements of the physics at play and to interpret apparently contradictory experiments on the same machine and across different machines. The first part of this thesis deals with electron transport in TCV high confinement mode plasmas. It was observed that the electron density profile in these plasmas flatten when intense electron heating is applied, in contrast to observations on other machines where the increase of the profile peakedness was reported. It is shown with quasi-linear gyrokinetic simulations that this effect, usually interpreted as collisionality dependence, stems from the combined effect of many plasma parameters. The influence of the collisionality, electron to ion temperature ratio, the ratio of temperature gradients, and the Ware-pinch are studied with detailed parameter scans. It is shown that the complex interdependence of the various plasma parameters is greatly simplified when the simulation results are interpreted as a function of the average frequency of the main modes contributing to radial transport. In this way the model is able to explain the experimental results. It was also shown that the same basic understanding is at play in L-modes, H-modes and electron internal transport barriers. The second part of the thesis is devoted to impurity transport. A multi-purpose gas injection system is developed, commissioned and calibrated. It is shown that the system is capable of massive gas injections to provoke disruptions and delivering small puffs of gaseous impurities for perturbative transport experiments. This flexible tool is exploited in a series of impurity transport measurements with argon and neon injections. The impurities

Electron and impurity transport studies in the TCV Tokamak

International Nuclear Information System (INIS)

Wagner, D.

2013-05-01

In this thesis electron and impurity transport are studied in the Tokamak à Configuration Variable (TCV) located at CRPP-EPFL in Lausanne. Understanding particle transport is primordial for future nuclear fusion power plants. Modeling of experiments in many specific plasma scenarios can help to understand the common elements of the physics at play and to interpret apparently contradictory experiments on the same machine and across different machines. The first part of this thesis deals with electron transport in TCV high confinement mode plasmas. It was observed that the electron density profile in these plasmas flatten when intense electron heating is applied, in contrast to observations on other machines where the increase of the profile peakedness was reported. It is shown with quasi-linear gyrokinetic simulations that this effect, usually interpreted as collisionality dependence, stems from the combined effect of many plasma parameters. The influence of the collisionality, electron to ion temperature ratio, the ratio of temperature gradients, and the Ware-pinch are studied with detailed parameter scans. It is shown that the complex interdependence of the various plasma parameters is greatly simplified when the simulation results are interpreted as a function of the average frequency of the main modes contributing to radial transport. In this way the model is able to explain the experimental results. It was also shown that the same basic understanding is at play in L-modes, H-modes and electron internal transport barriers. The second part of the thesis is devoted to impurity transport. A multi-purpose gas injection system is developed, commissioned and calibrated. It is shown that the system is capable of massive gas injections to provoke disruptions and delivering small puffs of gaseous impurities for perturbative transport experiments. This flexible tool is exploited in a series of impurity transport measurements with argon and neon injections. The impurities

A theoretical study of impurity production at limiters in tokamaks

International Nuclear Information System (INIS)

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

1988-01-01

The spatial distribution of neutral impurity emissions around graphite limiters in the DITE tokamak is investigated using a Monte Carlo neutral transport code based on physical sputtering. The Monte Carlo code results are compared with experiments for the CI distributions observed toroidally around the probe limiter and radially around the fixed limiter. The comparison between code and experiment demonstrates the ability of camera observations and Langmuir probe measurements to provide detailed information on the production of impurities at limiters. From the Monte Carlo simulation it is shown that the toroidal distribution of impurity emission is determined mainly by the geometry of the limiter and the radial variation of the sputtering yield, while the radial distribution is sensitive mainly to the sputtered atom velocity distribution and the rate coefficients for ionization and excitation

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.)

On the role of impurity radiation on edge turbulence in the TJ-1 Tokamak

International Nuclear Information System (INIS)

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

1994-01-01

The correlation between edge radiation and electron temperature and density fluctuations has been studied in the vicinity of the upper poloidal limiter of the TJ-I tokamak. When edge impurity radiation is strongly raked in the proximity of the limiter radius, electron temperature fluctuations are notably higher than density fluctuations. Results provide experimental evidence of edge turbulence driven by thermal instabilities

Steady-state tokamak reactor with non-divertor impurity control: STARFIRE

International Nuclear Information System (INIS)

Baker, C.C.

1980-01-01

STARFIRE is a conceptual design study of a commercial tokamak fusion electric power plant. Particular emphasis has been placed on simplifying the reactor concept by developing design concepts to produce a steady-state tokamak with non-divertor impurity control and helium ash removal. The concepts of plasma current drive using lower hybrid rf waves and a limiter/vacuum system for reactor applications are described

Electromagnetic effects on trace impurity transport in tokamak plasmas

Science.gov (United States)

Hein, T.; Angioni, C.

2010-01-01

The impact of electromagnetic effects on the transport of light and heavy impurities in tokamak plasmas is investigated by means of an extensive set of linear gyrokinetic numerical calculations with the code GYRO [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)] and of analytical derivations with a fluid model. The impurity transport is studied by appropriately separating diffusive and convective contributions, and conditions of background microturbulence dominated by both ion temperature gradient (ITG) and trapped electron modes (TEMs) are analyzed. The dominant contribution from magnetic flutter transport turns out to be of pure convective type. However it remains small, below 10% with respect to the E ×B transport. A significant impact on the impurity transport due to an increase in the plasma normalized pressure parameter β is observed in the case of ITG modes, while for TEM the overall effect remains weak. In realistic conditions of high β plasmas in the high confinement (H-) mode with dominant ITG turbulence, the impurity diffusivity is found to decrease with increasing β in qualitative agreement with recent observations in tokamaks. In contrast, in these conditions, the ratio of the total off-diagonal convective velocity to the diagonal diffusivity is not strongly affected by an increase in β, particularly at low impurity charge, due to a compensation between the different off-diagonal contributions.

Coefficients of viscosity for heavy impurity element in tokamak

Energy Technology Data Exchange (ETDEWEB)

El-Sharif, R N; Bekhit, A M [Plasma Physics dept., NRC, Atomic energy Authority, Cairo, (Egypt)

1997-12-31

The transport of heavy impurity element in to tokamak was studied theoretically. The viscosity coefficients of chromium impurities has been calculated in 13 and 21 moment approximation, in the limit of strong fields where is the gyrofrequency of species it was found that the off diagonal coefficient approximately tends to zero. This means that the friction force in the off-diagonal direction is very small, for the perpendicular viscosity coefficient the two approximation coincide to each other. 3 figs.

On the Role of Impurity Radiation on Edge Turbulence in the TJ-I Tokamak

International Nuclear Information System (INIS)

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

1994-01-01

The correlation between edge radiation and electron temperature and density fluctuations has been studied in the vicinity of the upper poloidal limiter of the TJ-I tokamak. When edge impurity radiation is strongly peaked in the proximity of the limiter radius, electron temperature fluctuations are notably higher than density fluctuations. Results provide experimental evidence of edge turbulence driven by thermal instabilities. (Author) 16 refs

On the Role of Impurity Radiation on Edge Turbulence in the TJ-I Tokamak

Energy Technology Data Exchange (ETDEWEB)

Ochando, M A; Pedrosa, M A; Balbin, R; Garcia-Cortes, I; Hidalgo, C

1994-07-01

The correlation between edge radiation and electron temperature and density fluctuations has been studied in the vicinity of the upper poloidal limiter of the TJ-I tokamak. When edge impurity radiation is strongly peaked in the proximity of the limiter radius, electron temperature fluctuations are notably higher than density fluctuations. Results provide experimental evidence of edge turbulence driven by thermal instabilities. (Author) 16 refs.

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

Energy Technology Data Exchange (ETDEWEB)

Navas, G; Zurro, B

1982-07-01

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

Effects of impurities and magnetic divertors on high-temperature tokamaks

International Nuclear Information System (INIS)

Meade, D.M.; Furth, H.P.; Rutherford, P.H.; Seidl, F.G.P.; Duechs, D.F.

1974-10-01

A one-dimensional tokamak plasma transport code has been adapted to include impurity influx, stripping, radiation, and diffusion, as well as the usual processes of hydrogen plasma and heat transport, recycling at the boundary, and multigeneration charge-exchange. Neutral-beam heating, adiabatic compression, and divertor boundary conditions are included as optional features. Illustrative computations are given for present-day and next-generation tokamaks. The problems of impurity control are discussed, and two technical approaches are examined in greater detail: the transient cold-plasma shield, and the poloidal divertor. (auth)

Impurity control in near-term tokamak reactors

International Nuclear Information System (INIS)

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

1976-10-01

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

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

Parametric dependences of impurity transport in the Tore Supra tokamak

International Nuclear Information System (INIS)

Parisot, Th.

2007-09-01

During this Ph.D. work, a full setup of tools for an experimental investigation of impurity transport has been developed on the Tore Supra tokamak. It includes a laser blow-off system for metallic impurity injections and developments for ITC (Impurity Transport Code), a transport code which allows the extraction of the experimental impurity transport coefficients (diffusion and convection velocity). This tool has been used to perform and analyse several experiments, to evidence parametric dependences of impurity transport. In a first experiment, a confinement time law for nickel in Tore Supra has been obtained as a function of collisionality ν * and normalized Larmor radius ρ * . Then the impurity charge Z role has been investigated in various conditions: ohmic regime with or without sawteeth, and sawtooth less L-mode with LH power. No Z effect is observed, consistently with theoretical predictions, whether neoclassical (NCLASS) or for turbulent transport with both non linear gyro-fluid (TRB) and quasilinear gyrokinetic (QuaLiKiz) simulations. An exception is found for LH heated plasmas where the confinement time seems to decrease for the heaviest impurities. This is not explained by any model available. The observed transport is close to neoclassical between sawtooth relaxations, in the centre (r q-1 ) of ohmic plasmas, turbulent outside. Without sawteeth, it is turbulent in the whole plasma, for ohmic or L mode discharges. The profile shape of the diffusion coefficient is here qualitatively different, with a stronger and deeper transition between the low diffusion central region and a more turbulent peripheral region for LH heated plasmas. (author)

TIBER (Tokamak Ignition/Burn Experimental Reactor) II as a precursor to an international thermonuclear experimental reactor

International Nuclear Information System (INIS)

Henning, C.D.; Gilleland, J.R.

1988-01-01

The Tokamak Ignition/Burn Experimental Reactor (TIBER) was pursued in the US as one option for an International Thermonuclear Experimental Reactor (ITER). This concept evolved from earlier work on the Tokamak Fusion Core Experiment (TFCX) to develop a small, ignited tokamak. While the copper-coil versions of TFCX became the short-pulsed, 1.23-m radius, Compact Ignition Tokamak (CIT), the superconducting TIBER with long pulse or steady state and a 2.6-m radius was considered for international collaboration. Recently the design was updated to TIBER II, to accommodate more conservative confinement scaling, double-poloidal divertors for impurity control, steady-state current drive, and nuclear testing. 18 refs., 1 fig

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

Experimental study of impurity production in the Tokapole II tokamak

International Nuclear Information System (INIS)

Brickhouse, N.S.

1984-01-01

The release mechanism for low-Z impurities in Tokapole II has been characterized through impurity doping and isotopic exchange experiments. The desorption mechanism responsible for the low-Z impurity concentrations during the rise phase of the plasma current depends on the mass of the plasma ions. Doping with small amounts of any gas studied (H 2 , D 2 , He, N 2 , O 2 , Ne, Ar, Kr, and Xe) increases the early-time radiation of O, C, and N. For exotic gas doping this increase is linear with the dopant concentration, and proportional to the mass of the dopant, as expected for a momentum transfer process. Isotopic exchange experiments confirm the mass-dependence of oxygen production. A time-dependent coronal model is compared with the vacuum ultraviolet spectroscopic signals of the ionizing oxygen. The quantity sigma/tau (desorption cross section divided by particle confinement time) is determined to be 4 x 10 13 cm 2 /msec. The oxygen influx has a large peak early in the start-up

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.

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.

HESTER: a hot-electron superconducting tokamak experimental reactor at M.I.T

International Nuclear Information System (INIS)

Schultz, J.H.; Montgomery, D.B.

1983-04-01

HESTER is an experimental tokamak, designed to resolve many of the central questions in the tokamak development program in the 1980's. It combines several unique features with new perspectives on the other major tokamak experiments scheduled for the next decade. The overall objectives of HESTER, in rough order of their presently perceived importance, are the achievement of reactor-like wall-loadings and plasma parameters for long pulse periods, determination of a good, reactor-relevant method of steady-state or very long pulse tokamak current drive, duplication of the planned very high temperature neutral injection experiments using only radio frequency heating, a demonstration of true steady-state tokamak operation, integration of a high-performance superconducting magnet system into a tokamak experiment, determination of the best methods of long term impurity control, and studies of transport and pressure limits in high field, high aspect ratio tokamak plasmas. These objectives are described

Tokamak Engineering Technology Facility scoping study

Energy Technology Data Exchange (ETDEWEB)

Stacey, W.M. Jr.; Abdou, M.A.; Bolta, C.C.

1976-03-01

A scoping study for a Tokamak Engineering Technology Facility (TETF) is presented. The TETF is a tokamak with R = 3 m and I/sub p/ = 1.4 MA based on the counterstreaming-ion torus mode of operation. The primary purpose of TETF is to demonstrate fusion technologies for the Experimental Power Reactor (EPR), but it will also serve as an engineering and radiation test facility. TETF has several technological systems (e.g., superconducting toroidal-field coil, tritium fuel cycle, impurity control, first wall) that are prototypical of EPR.

Tokamak Engineering Technology Facility scoping study

International Nuclear Information System (INIS)

Stacey, W.M. Jr.; Abdou, M.A.; Bolta, C.C.

1976-03-01

A scoping study for a Tokamak Engineering Technology Facility (TETF) is presented. The TETF is a tokamak with R = 3 m and I/sub p/ = 1.4 MA based on the counterstreaming-ion torus mode of operation. The primary purpose of TETF is to demonstrate fusion technologies for the Experimental Power Reactor (EPR), but it will also serve as an engineering and radiation test facility. TETF has several technological systems (e.g., superconducting toroidal-field coil, tritium fuel cycle, impurity control, first wall) that are prototypical of EPR

Effects of density asymmetries on heavy-impurity transport in a rotating tokamak-plasma

International Nuclear Information System (INIS)

Romanelli, M.; Ottaviani, M.

1997-12-01

The transport equations of heavy trace-impurities in a Tokamak plasma with strong toroidal rotation have been studied analytically in the collisional regime. It is found that the poloidal asymmetry of the impurity-density, which occurs because of the rotation, brings about a large enhancement of the diffusivity and indeed of the pinch velocity above the conventional Pfirsh-Schlueter values. (author)

Characterization of intermittency of impurity turbulent transport in tokamak edge plasmas

International Nuclear Information System (INIS)

Futatani, S.; Benkadda, S.; Nakamura, Y.; Kondo, K.

2008-01-01

The statistical properties of impurity transport of a tokamak edge plasma embedded in a dissipative drift-wave turbulence are investigated using structure function analysis. The impurities are considered as a passive scalar advected by the plasma flow. Two cases of impurity advection are studied and compared: A decaying impurities case (given by a diffusion-advection equation) and a driven case (forced by a mean scalar gradient). The use of extended self-similarity enables us to show that the relative scaling exponent of structure functions of impurity density and vorticity exhibit similar multifractal scaling in the decaying case and follows the She-Leveque model. However, this property is invalidated for the impurity driven advection case. For both cases, potential fluctuations are self-similar and exhibit a monofractal scaling in agreement with Kolmogorov-Kraichnan theory for two-dimensional turbulence. These results obtained with a passive scalar model agree also with test-particle simulations.

VUV study of impurity generation during ICRF heating experiments on the Alcator C tokamak

International Nuclear Information System (INIS)

Manning, H.L.

1986-06-01

A 2.2 meter grazing incidence VUV monochromator has been converted into a time-resolving spectrograph by the addition of a new detector system, based on a microchannel plate image intensifier linked to a 1024-element linear photodiode array. The system covers the wavelength range 15 to 1200 A (typically 40 A at a time) with resolution of up to .3 A FWHM. Time resolution is selectable down to 0.5 msec. The system sensitivity was absolutely calibrated below 150 A by a soft x-ray calibration facility. The spectrograph was installed on the Alcator C tokamak at MIT to monitor plasma impurity emission. There, cross-calibration with a calibrated EUV monochromator was performed above 400 A. Calibration results, system performance characteristics, and data from Alcator C are presented. Observations of impurity behavior are presented from a series of ICRF heating experiments (180 MHz, 50 to 400 kW) performed on the Alcator C tokamak, using graphite limiters and stainless steel antenna Faraday shields

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

Application of secondary ion emission to impurity control in tokamaks

International Nuclear Information System (INIS)

Krauss, A.R.; Gruen, D.M.

1979-01-01

The extent to which high Z impurities enter the plasma of a magnetic confinement fusion device depends on the kinetic energy, angle of emission, and very importantly, the charge state of the ejected material. We have been studying both the fundamental process of secondary ion emission and possible techniques for producing surfaces which give rise to high ion fractions during sputtering, with a view to assessing the potential of this approach to impurity control in tokamaks. By carefully choosing materials exposed to fusion plasmas and by properly modifying the surface it may be possible to insure that nearly all the impurities are ejected as ions. As long as certain gas blanket configurations are avoided and especially if a divertor is used, it should then be possible to remove the impurities before they reach the plasma. The relative merits of a variety of materials are considered with regard to this application

Impurity flux collection at the plasma edge of the tokamak MT-1

International Nuclear Information System (INIS)

Hildebrandt, D.; Bakos, J.S.; Petravich, G.

1989-09-01

Fluxes of intrinsic and injected impurities and background plasma ions were collected using a bidirectional probe at the plasma edge of the tokamak MT-1. The directional and radial dependences of injected impurities and plasma ions were very similar indicating a strong coupling of the impurity transport to the dynamics of the background plasma. The measured intrinsic concentration of about 10 -4 for Mo at the plasma edge is derived. (author) 17 refs.; 5 figs

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)

Effect of magnetic islands on the impurity transport in a tokamak

International Nuclear Information System (INIS)

Ivanov, N.V.; Khvostenko, P.P.; Chudnovskij, A.N.

1986-01-01

Effect of magnetic islands m=2, created in plasma with the help of special quadrupole winding, on the behaviour of impurity in the T-7 tokamak is studied. The use of quadrupole winding permitted to exlude the magnetic island nonstationarity typical to spontaneous development of the Tiaring instability. The results obtained confirm the point of view that splitting of rational magnetic surface results in the change of impurity ion density gradient in its vicinity. This change occurs under the action of ambipolar electric field, that is excited due to the increase of radial electron transport in magnetic islands

Effect of impurities and ripple upon power regulation in self-sustained tokamaks

International Nuclear Information System (INIS)

Bromberg, L.; Cohn, D.R.

1981-01-01

Tokamak power reactors will likely operate in a self sustained heating mode where additional power losses are introduced to permit higher levels of alpha particle heating (and thus higher levels of total fusion power) at thermal equilibrium. Illustrative 0-dimensional calculations are made to assess requirements for power regulation of self sustained tokamak plasmas by the use of impurity radiation. Effects of impurities upon allowable fuel density and thermal stability are determined. Requirements are calculated for passive thermal stability control by temperature driven radial motion in the presence of ripple transport losses; it appears that stability might be attained over a relatively wide temperature range with a small amount of ripple transport loss. Requirements for power regulation by the use of ripple transport are also determined

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

Direction of Impurity Pinch and Auxiliary Heating in Tokamak Plasmas

International Nuclear Information System (INIS)

Angioni, C.; Peeters, A.G.

2006-01-01

A mechanism of particle pinch for trace impurities in tokamak plasmas, arising from the effect of parallel velocity fluctuations in the presence of a turbulent electrostatic potential, is identified analytically by means of a reduced fluid model and verified numerically with a gyrokinetic code for the first time. The direction of such a pinch reverses as a function of the direction of rotation of the turbulence in agreement with the impurity pinch reversal observed in some experiments when moving from dominant auxiliary ion heating to dominant auxiliary electron heating

Experimental investigations at the Soviet tokamaks

International Nuclear Information System (INIS)

Bobrovskij, G.A.; Golant, V.E.; AN SSSR, Leningrad. Fiziko-Tekhnicheskij Inst.)

1978-01-01

The review is devoted to the basic results obtained on the Soviet tokamaks during 1976-1977. Behaviour of impurities, tearing instability, additional methods of plasma heating, energy distribution function were investigated. A brief description of new T-7, TM-4, ''Tuman-3'' tokamaks is given. It is shown that despite inflow of impurities to the pinch periphery, no their appreciable accumulation is observed at least during the discharge time. It is shown that the helical perturbations with m=2 and 1 present the greatest danger. The suppression of the tearing instability is related with suppression of the mode with m=2. The helical perturbation prevents formation of skin configuration at the initial stage of the discharge. As a rule, the transition of an appreciable fraction of electrons to continuous acceleration does not take place, although a significant deformation of electron distribution function under the action of electric field occurs. Plasma compression by increasing magnetic field induces oscillations and improves thermal plasma isolation. It is shown experimentally that the considerable efficiency of energy contribution to the ion component at the central part of plasma may be obtained by means of HF heating under conditions of low-hybrid resonance. It is shown that the recombination has a considerable effect on concentration of neutral particles in the central region

Extended numerical modeling of impurity neoclassical transport in tokamak edge plasmas

International Nuclear Information System (INIS)

Inoue, H.; Yamoto, S.; Hatayama, A.; Homma, Y.

2016-01-01

Understanding of impurity transport in tokamaks is an important issue in order to reduce the impurity contamination in fusion core plasmas. Recently, a new kinetic numerical scheme of impurity classical/neoclassical transport has been developed. This numerical scheme makes it possible to include classical self-diffusion (CL SD), classical inward pinch (CL IWP), and classical temperature screening effect (CL TSE) of impurity ions. However, impurity neoclassical transport has been modeled only in the case where background plasmas are in the Pfirsch-Schluter (PS) regime. The purpose of this study is to extend our previous model to wider range of collisionality regimes, i.e., not only the PS regime, but also the plateau regime. As in the previous study, a kinetic model with Binary Collision Monte-Carlo Model (BMC) has been adopted. We focus on the modeling of the neoclassical self-diffusion (NC SD) and the neoclassical inward pinch (NC IWP). In order to simulate the neoclassical transport with the BCM, velocity distribution of background plasma ions has been modeled as a deformed Maxwell distribution which includes plasma density gradient. Some test simulations have been done. As for NC SD of impurity ions, our scheme reproduces the dependence on the collisionality parameter in wide range of collisionality regime. As for NC IWP, in cases where test impurity ions and background ions are in the PS and plateau regimes, parameter dependences have been reproduced. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Extended numerical modeling of impurity neoclassical transport in tokamak edge plasmas

Energy Technology Data Exchange (ETDEWEB)

Inoue, H.; Yamoto, S.; Hatayama, A. [Graduate School of Science and Technology, Keio University, Hiyoshi, Yokohama (Japan); Homma, Y. [Graduate School of Science and Technology, Keio University, Hiyoshi, Yokohama (Japan); Research Fellow of Japan Society for the Promotion of Science, Tokyo (Japan)

2016-08-15

Understanding of impurity transport in tokamaks is an important issue in order to reduce the impurity contamination in fusion core plasmas. Recently, a new kinetic numerical scheme of impurity classical/neoclassical transport has been developed. This numerical scheme makes it possible to include classical self-diffusion (CL SD), classical inward pinch (CL IWP), and classical temperature screening effect (CL TSE) of impurity ions. However, impurity neoclassical transport has been modeled only in the case where background plasmas are in the Pfirsch-Schluter (PS) regime. The purpose of this study is to extend our previous model to wider range of collisionality regimes, i.e., not only the PS regime, but also the plateau regime. As in the previous study, a kinetic model with Binary Collision Monte-Carlo Model (BMC) has been adopted. We focus on the modeling of the neoclassical self-diffusion (NC SD) and the neoclassical inward pinch (NC IWP). In order to simulate the neoclassical transport with the BCM, velocity distribution of background plasma ions has been modeled as a deformed Maxwell distribution which includes plasma density gradient. Some test simulations have been done. As for NC SD of impurity ions, our scheme reproduces the dependence on the collisionality parameter in wide range of collisionality regime. As for NC IWP, in cases where test impurity ions and background ions are in the PS and plateau regimes, parameter dependences have been reproduced. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Expanded-boundary approach to impurity control in tokamaks

International Nuclear Information System (INIS)

Ohyabu, N.

1981-01-01

It is proposed to expand the outermost flux surfaces in tokamaks to divert the heat flux emerging from the plasma core. The expanded flux surfaces provide a large volume for radiative cooling. The radiative power at the boundary is enhanced by the effects of plasma flow as well as by a volumetric factor, and the resultant edge cooling and reduced heat load on the limiter may significantly retard impurity generation. Furthermore, it seems to be compatible with reactor engineering requirements. (author)

The ARIES-I tokamak reactor study

International Nuclear Information System (INIS)

1991-01-01

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

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

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)

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)

Measurements on injected impurity transport in TEXT [Texas Experimental Tokamak] using multiply filtered soft x-ray detectors

International Nuclear Information System (INIS)

Wenzel, K.W.

1990-01-01

Aluminum was injected into TEXT to study trace, non-recycling impurity transport. A 92-channel, three array x-ray imaging system was constructed and installed to measure temporally-resolved density profiles of the three highest charge states. A novel krypton filter in one array discriminated between the He-like and H-like resonance lines, and a hard filter responded mostly to the fully stripped charge state. The impurity confinement time scaled approximately as τ c ∼ bar n e Z eff √m i /Z i /Ip (i denotes the background gas). Aluminum density profiles averaged over a sawtooth crashes were also measured for a few discharges. Sawteeth strongly enhanced the inward impurity flow immediately following injection, when the density was still peaked near the plasma edge. Those discharges with the longest sawtooth period obtained the most peaked aluminum density profiles; thus sawteeth were also important in ameliorating impurity accumulation on the tokamak axis. The charge state balance of the aluminum ions obtained from the measured profiles was compared to predictions of coronal equilibrium. Somewhat surprisingly the aluminum ions were close to coronal, except in those discharges with very short sawtooth periods or very large inversion radii. Preliminary evidence of up-down asymmetric density profiles was also found. Numerical simulations of aluminum transport were performed. The effect of sawtooth oscillations was taken into account with a simple flattening model. The data disagreed with a constant D anomalous model except in the plasma center; enhanced outward transport was required. The experiments did not agree with neoclassical simulations, because the theory had outward convection that was too large. 237 refs., 86 figs., 8 tabs

Investigation of oxygen impurity transport using the O4+ visible spectral line in the Aditya tokamak

International Nuclear Information System (INIS)

Chowdhuri, M.B.; Ghosh, J.; Banerjee, S.; Dey, Ritu; Manchanda, R.; Kumar, Vinay; Vasu, P.; Patel, K.M.; Atrey, P.K.; Shankara Joisa, Y.; Rao, C.V.S.; Tanna, R.L.; Raju, D.; Chattopadhyay, P.K.; Jha, R.; Gupta, C.N.; Bhatt, S.B.; Saxena, Y.C.

2013-01-01

Intense visible lines from Be-like oxygen impurity are routinely observed in the Aditya tokamak. The spatial profile of brightness of a Be-like oxygen spectral line (2p3p 3 D 3 –2p3d 3 F 4 ) at 650.024 nm is used to investigate oxygen impurity transport in typical discharges of the Aditya tokamak. A 1.0 m multi-track spectrometer (Czerny–Turner) capable of simultaneous measurements from eight lines of sight is used to obtain the radial profile of brightness of O 4+ spectral emission. The emissivity profile of O 4+ spectral emission is obtained from the spatial profile of brightness using an Abel-like matrix inversion. The oxygen transport coefficients are determined by reproducing the experimentally measured emissivity profiles of O 4+ , using a one-dimensional empirical impurity transport code, STRAHL. Much higher values of the diffusion coefficient compared with the neo-classical values are observed in both the high magnetic field edge region (D inboard max ∼30 m 2 s -1 ) and the low magnetic field edge region (D outboard max ∼45 m 2 s -1 ) of typical Aditya ohmic plasmas, which seems to be due to fluctuation-induced transport. The diffusion coefficient at the limiter radius in the low-field (outboard) region is typically ∼ twice as high as that at the limiter radius in the high-field (inboard) region. (paper)

Impurity transport and plasma rotation in the ISX-B tokamak

International Nuclear Information System (INIS)

Isler, R.C.; Murray, L.E.; Crume, E.C.

1983-01-01

Recent calculations have shown that when external momentum sources and plasma rotation are included in the neoclassical theory, the standard results for impurity transport can be strongly altered. Under appropriate conditions, inward convection is reduced by co-injection and enhanced by counter-injection. In order to examine the theoretical predictions, several observations of impurity transport have been made in the ISX-B tokamak during neutral-beam injection for comparison with the transport seen with Ohmic heating alone. Both intrinsic contaminants and deliberately introduced test impurities display a behaviour that is in qualitative agreement with the predicted beam-driven effects. These correlations are particularly noticeable when the comparisons are made for deuterium when the impurity transport in the Ohmically heated discharges exhibits neoclassical-like characteristics, i.e. accumulation and long confinement times. Similar but smaller effects are observed in beam-heated hydrogen discharges; neoclassical-like behaviour is not seen in Ohmically heated hydrogen sequences. Emphasis has been placed on measuring toroidal plasma rotation, and semiquantitative comparisons with the theories of beam-induced impurity transport have been made. It is possible that radial electric fields other than those associated with momentum transfer and increased anomalous processes during injection could also play a role. (author)

Impurity and neutral effects on the dissipative drift wave in tokamak edge plasmas

International Nuclear Information System (INIS)

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

1991-05-01

Possible destabilizing mechanisms for the liner electrostatic dissipative drift waves (in tokamak edge plasmas) are investigated in slab geometry. The effects of processes such as ionization, charge exchange, radiation, and rippling are examined. In particular, the impurity condensation associated with radiation cooling is evaluated appropriately for the drift wave ordering, which is found to be an important driving mechanism in contrast to the results of earlier studies. It also shown that the role of ionization is quite complicated, and depends strongly on the manner in which the equilibrium is achieved. The linear eigenmode equation is studied both analytically and numerically. For the range of parameters relevant to TEXT tokamak, both the charge exchange of the rippling effect are found to be unimportant for instability. 25 refs., 6 figs

Effect of density control and impurity transport on internal transport barrier formation in tokamak plasma

International Nuclear Information System (INIS)

Yamakami, Tomoyuki; Fujita, Takaaki; Arimoto, Hideki; Yamazaki, Kozo

2014-01-01

In future fusion reactors, density control, such as fueling by pellet injection, is an effective method to control the formation of the internal transport barrier (ITB) in reversed magnetic shear plasma, which can improve plasma performance. On the other hand, an operation with ITB can cause accumulation of impurities inside the core ITB region. We studied the relation between pellet injection and ITB formation and the effect of impurity transport on the core of ITB for tokamak plasmas by using the toroidal transport analysis linkage. For ITB formation, we showed that the pellet has to be injected beyond the position where the safety factor q takes the minimum value. We confirmed that the accumulation of impurities causes the attenuation of ITB owing to radiation loss inside the ITB region. Moreover, in terms of the divertor heat flux reduction by impurity gas, the line radiation loss is high for high-Z noble gas impurities, such as Kr, whereas factor Q decreases slightly. (author)

Spectroscopic studies of carbon impurities in PISCES-A

International Nuclear Information System (INIS)

Ra, Y.; Hirooka, Y.; Leung, W.K.; Conn, R.W.; Pospieszczyk, A.

1989-08-01

The graphite used for the limiter of the tokamak reactor produces carbon-containing molecular impurities as a result of the interactions with the edge plasma. The behavior of these molecular impurities has been studied using emission spectroscopy. The present study includes: finding molecular bands and atomic lines in the visible spectral range which can be used for the study of the molecular impurities, studying the breakup processes of the molecular impurities on their way from the source into the plasma, developing a spectroscopic diagnostic method for the absolute measurement of the molecular impurity flux resulting from graphite erosion. For these studies, carbon-containing molecules such as CH 4 , C 2 H 2 , C 2 H 4 , and CO 2 were injected into the tokamak-boundary,like plasma generated by PISCES-A. The spectrograms of these gases were taken. Many useful bands and lines were determined from the spectrograms. The breakup processes of these gases were studied by observing the spatial profiles of the emission of the molecules and their radicals for different plasma conditions. For the absolute measurement of the eroded molecular impurity flux, the photon efficiency of the lines and bands were found by measuring the absolute number of the emitted photons and injected gas molecules. The chemical sputtering yield of graphite by hydrogen plasma was spectroscopically measured using the previously obtained photon efficiencies. It showed good agreement with results obtained by weight loss measurements. 16 refs., 7 figs., 1 tab

Amount of impurity and its behavior in the STP-2 screw pinch tokamak

International Nuclear Information System (INIS)

Yamaguchi, S.

1981-05-01

Temporal and spatial evolution of oxygen spectral line intensities have been measured in the STP-2 screw pinch tokamak. The electron density and temperature as measured by Thomson scattering are of the order of 10 14 cm -3 and 10 eV, respectively. On the basis of these measurements, quasi-steady-state rate equations have been solved to give the OII and OIII ion densities. It is found that the density of oxygen impurity is about several percent of the electron density, and the impurity moves with the bulk plasma. It is confirmed that the impurity originates from the wall of the discharge tube during the initial phase of the discharge. (author)

Radio frequency induced and neoclassical asymmetries and their effects on turbulent impurity transport in a tokamak

Energy Technology Data Exchange (ETDEWEB)

Pusztai, I. [Applied Physics, Chalmers University of Technology and Euratom-VR Association, SE-41296 Goeteborg (Sweden); Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Landreman, M. [University of Maryland, College Park, MD 20742 (United States); Mollen, A.; Fueloep, T. [Applied Physics, Chalmers University of Technology and Euratom-VR Association, SE-41296 Goeteborg (Sweden); Kazakov, Ye.O. [Laboratory for Plasma Physics, ERM/KMS, Association ' EURATOM-Belgian State' , TEC Partner, BE-1000 Brussels (Belgium)

2014-06-15

Poloidal asymmetries in the impurity density can be generated by radio frequency heating in the core and by neoclassical effects in the edge of tokamak plasmas. In a pedestal case study, using global neoclassical simulations we find that finite orbit width effects can generate significant poloidal variation in the electrostatic potential, which varies on a small radial scale. Gyrokinetic modeling shows that these poloidal asymmetries can be strong enough to significantly modify turbulent impurity peaking. In the pedestal the E x B drift in the radial electric field can give a larger contribution to the poloidal motion of impurities than that of their parallel streaming. Under such circumstances we find that up-down asymmetries can also affect impurity peaking. (copyright 2014 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

Calculation of impurity poloidal rotation from measured poloidal asymmetries in the toroidal rotation of a tokamak plasma

Energy Technology Data Exchange (ETDEWEB)

Chrystal, C. [University of California-San Diego, La Jolla, California 92186-5608 (United States); Burrell, K. H.; Groebner, R. J.; Kaplan, D. H. [General Atomics, San Diego, California 92186-5608 (United States); Grierson, B. A. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States)

2012-10-15

To improve poloidal rotation measurement capabilities on the DIII-D tokamak, new chords for the charge exchange recombination spectroscopy (CER) diagnostic have been installed. CER is a common method for measuring impurity rotation in tokamak plasmas. These new chords make measurements on the high-field side of the plasma. They are designed so that they can measure toroidal rotation without the need for the calculation of atomic physics corrections. Asymmetry between toroidal rotation on the high- and low-field sides of the plasma is used to calculate poloidal rotation. Results for the main impurity in the plasma are shown and compared with a neoclassical calculation of poloidal rotation.

Calculation of impurity poloidal rotation from measured poloidal asymmetries in the toroidal rotation of a tokamak plasma.

Science.gov (United States)

Chrystal, C; Burrell, K H; Grierson, B A; Groebner, R J; Kaplan, D H

2012-10-01

To improve poloidal rotation measurement capabilities on the DIII-D tokamak, new chords for the charge exchange recombination spectroscopy (CER) diagnostic have been installed. CER is a common method for measuring impurity rotation in tokamak plasmas. These new chords make measurements on the high-field side of the plasma. They are designed so that they can measure toroidal rotation without the need for the calculation of atomic physics corrections. Asymmetry between toroidal rotation on the high- and low-field sides of the plasma is used to calculate poloidal rotation. Results for the main impurity in the plasma are shown and compared with a neoclassical calculation of poloidal rotation.

Studies on fundamental technologies for producing tokamak-plasma

International Nuclear Information System (INIS)

Matsuzaki, Yoshimi

1987-10-01

The report describes studies on fundamental technologies to produce tokamak-plasma of the JFT-2 and JFT-2M tokamaks. (1) In order to measure the particle number of residual gases, calibration methods of vacuum gauges have been developed. (2) Devices for a Taylor-type discharge cleaning (TDC), a glow discharge cleaning (GDC) and ECR discharge cleaning (ECR-DC) have been made and the cleaning effects have been investigated. In TDC the most effective plasma for cleaning is obtained in the plasma with 5 eV of electron temperature. GDC is effective in removing carbon impurities, but is less effective for removing oxygen impurities. ECR-DC has nearly the similar effect as TDC. The cleaning effect of these three types were studied by comparing the properties of resulting tokamak plasmas in the JFT-2M tokamak. (3) Experimental studies of pre-ionization showed as following results; A simple pre-ionization equipment as a hot-electron-gun and a J x B gun was effective in reducing breakdown voltage. An ordinary mode wave of the electron cyclotron frequency was very effective for pre-ionization. The RF power whose density is 3.6 x 10 -2 W/cm 3 produced plasma of an electron density of 5 x 10 11 cm -3 . In this case, it is possible to start up with negligible consumption of the magnetic flux caused by the plasma resistance. (4) Concerning to studies on plasma control, the following results were obtained; In order to obtain constant plasma current, a pulse forming network was constructed and sufficient constant plasma current was achieved. In applying an iso-flux method for measuring the plasma position, it is no problem practically to use only one loop-coil and one magnetic probe. (author)

Collimator type monochromator as a possible impurities monitor for fusion plasmas. Preliminary tests on the Tokamak TM-1-MH

International Nuclear Information System (INIS)

Musa, G.; Lungu, C.P.; Badalec, J.; Jakubka, K.; Kopecky, V.; Stoeckel, J.; Zacek, F.

1984-09-01

A collimator type monochromator has been tested for the first time as the impurity monitor on Tokamak. The possibility to use this type of monochromator in fusion devices is analyzed and a monoslit device is proposed as a convenient monitor for impurities. (authors)

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

Energy Technology Data Exchange (ETDEWEB)

Valencia A, R

2006-07-01

As a main objective of this work, we present some experimental results obtained from studying the process of extracting those impurities created by the interaction plasma with its vessel wall in the case of Novillo tokamak. Likewise, we describe the main cleaning and conditioning techniques applied to it, fundamentally that of glow discharge cleaning at a low electron temperature (<10 eV), both in noble and reactive gases, as well as the conditioning by thin film deposits of hydrogen rich amorphous carbon (carbonization) leading to a reduction in the plasma resistivity from 8.99 x 10{sup -6} to 4.5 x 10{sup -6} {omega}-m, thus taking the Z{sub ef} value from 3.46 to 2.07 which considerably improved the operational parameters of the machine. With a view to justifying the fact that controlled nuclear fusion is a feasible alternative for the energy demand that humanity will face in the future, we review in Chapter 1 some fundamentals of the energy production by nuclear fusion reactions while, in Chapter 2, we examine two relevant plasma wall interaction processes. Our experimental array used to produce both cleaning and intense plasma discharges is described in Chapter 3 along with the associated diagnostics equipment. Chapter 4 contains a description of the vessel conditioning techniques followed in the process. Finally, we report our results in Chapter 5 while, in Chapter 6, some conclusions and remarks are presented. It is widely known that tokamak impurities are generated mainly by the plasma-wall interaction, particularly in the presence of high potentials between the plasma sheath and the limiter or wall. Given that impurities affect most adversely the plasma behaviour, understanding and controlling the impurity extraction mechanisms is crucial for optimizing the cleaning and wall conditioning discharge processes. Our study of one impurity extraction mechanism for both low and high Z in Novillo tokamak was carried out though mass spectrometry, optical emission

Particle and momentum confinement in tokamak plasmas with unbalanced neutral beam injection and strong rotation

International Nuclear Information System (INIS)

Malik, M.A.

1988-01-01

There is a self-consistent theory of the effects of neutral beam injection on impurity transport in tokamak plasmas. The theory predicts that co-injection drives impurities outward and that counter-injection enhances the normally inward flow of impurities. The theory was applied to carry out a detailed analysis of the large experimental database from the PLT and the ISX-B tokamaks. The theory was found to generally model the experimental data quite well. It is, therefore, concluded that neutral beam co-injection can drive impurities outward to achieve clean central plasmas and a cool radiating edge. Theoretical predictions for future thermonuclear reactors such as INTOR, TIBER II, and ITER indicated that neutral beam driven flow reversal might be an effective impurity control method if the rate of beam momentum deposited per plasma ion is adequate. The external momentum drag, which is a pivotal concept in impurity flow reversal theory, is correctly predicted by the gyroviscous theory of momentum confinement. The theory was applied to analyze experimental data from the PLT and the PDX tokamaks with exact experimental conditions. The theory was found to be in excellent agreement with experiment over a wide range of parameters. It is, therefore, possible to formulate the impurity transport theory from first principles, without resort to empiricism

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.)

Design of charge exchange recombination spectroscopy for the joint Texas experimental tokamak

Energy Technology Data Exchange (ETDEWEB)

Chi, Y.; Zhuang, G., E-mail: ge-zhuang@hust.edu.cn; Cheng, Z. F.; Hou, S. Y.; Cheng, C.; Li, Z.; Wang, J. R.; Wang, Z. J. [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan 430074 (China)

2014-11-15

The old diagnostic neutral beam injector first operated at the University of Texas at Austin is ready for rejoining the joint Texas experimental tokamak (J-TEXT). A new set of high voltage power supplies has been equipped and there is no limitation for beam modulation or beam pulse duration henceforth. Based on the spectra of fully striped impurity ions induced by the diagnostic beam the design work for toroidal charge exchange recombination spectroscopy (CXRS) system is presented. The 529 nm carbon VI (n = 8 − 7 transition) line seems to be the best choice for ion temperature and plasma rotation measurements and the considered hardware is listed. The design work of the toroidal CXRS system is guided by essential simulation of expected spectral results under the J-TEXT tokamak operation conditions.

Rotation characteristics of main ions and impurity ions in H-mode tokamak plasma

International Nuclear Information System (INIS)

Kim, J.; Burrell, K.H.; Gohil, P.; Groebner, R.J.; Kim, Y.; St. John, H.E.; Seraydarian, R.P.; Wade, M.R.

1994-01-01

Poloidal and toroidal rotation of the main ions (He 2+ ) and the impurity ions (C 6+ and B 5+ ) in H-mode helium plasmas have been measured via charge exchange recombination spectroscopy in the DIII-D tokamak. It was discovered that the main ion poloidal rotation is in the ion diamagnetic drift direction while the impurity ion rotation is in the electron diamagnetic drift direction, in qualitative agreement with the neoclassical theory. The deduced radial electric field in the edge is of the same negative-well shape regardless of which ion species is used, validating the fundamental nature of the electric field in L-H transition phenomenology

Long-time impurity confinement as a precursor to disruptions in ohmically heated tokamaks

International Nuclear Information System (INIS)

Isler, R.C.; Rowan, W.L.

1988-01-01

It has been observed in several tokamaks that the confinement of test impurities increases dramatically when operating near density limits. The characteristics of the working gas transport coefficients also change character under these conditions. These changes appear to be caused by a suppression of the anomalous transport mechanisms. This series of vugraphs investigates the role of these changes in initiating disruptions

Deposit of thin films for Tokamaks conditioning

International Nuclear Information System (INIS)

Valencia A, R.

2006-01-01

As a main objective of this work, we present some experimental results obtained from studying the process of extracting those impurities created by the interaction plasma with its vessel wall in the case of Novillo tokamak. Likewise, we describe the main cleaning and conditioning techniques applied to it, fundamentally that of glow discharge cleaning at a low electron temperature ( -6 to 4.5 x 10 -6 Ω-m, thus taking the Z ef value from 3.46 to 2.07 which considerably improved the operational parameters of the machine. With a view to justifying the fact that controlled nuclear fusion is a feasible alternative for the energy demand that humanity will face in the future, we review in Chapter 1 some fundamentals of the energy production by nuclear fusion reactions while, in Chapter 2, we examine two relevant plasma wall interaction processes. Our experimental array used to produce both cleaning and intense plasma discharges is described in Chapter 3 along with the associated diagnostics equipment. Chapter 4 contains a description of the vessel conditioning techniques followed in the process. Finally, we report our results in Chapter 5 while, in Chapter 6, some conclusions and remarks are presented. It is widely known that tokamak impurities are generated mainly by the plasma-wall interaction, particularly in the presence of high potentials between the plasma sheath and the limiter or wall. Given that impurities affect most adversely the plasma behaviour, understanding and controlling the impurity extraction mechanisms is crucial for optimizing the cleaning and wall conditioning discharge processes. Our study of one impurity extraction mechanism for both low and high Z in Novillo tokamak was carried out though mass spectrometry, optical emission spectroscopy and plasma resistivity measurement. Such mechanism depends fundamentally on the mass of the ions that interact with the wall during the plasma current formation phase. The reaction products generated by the glow

Studies of the impurity pellet ablation in the high-temperature plasma of magnetic confinement devices

International Nuclear Information System (INIS)

Sergeev, V. Yu.; Bakhareva, O. A.; Kuteev, B. V.; Tendler, M.

2006-01-01

The ablation of impurity pellets in tokamak and stellarator plasmas is investigated. Different mechanisms for shielding the heat fluxes from the surrounding plasma to the pellet surface are discussed. A model for impurity pellet ablation is developed that can account for both neutral and electrostatic shielding. It is shown that the experimental values of the impurity pellet ablation rate are well described by the neutral gas shielding model over a wide range of plasma temperatures and densities. Taking into account the electrostatic shielding leads to worse agreement between the predictions of the model and the experimental data; this result still remains unclear. Scaling laws are obtained that allow one to estimate the local ablation rate of impurity pellets made of various materials over a wide range of plasma parameters in the neutral gas shielding model

Tokamak power systems studies, FY 1985

International Nuclear Information System (INIS)

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

1985-12-01

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

Tokamak power systems studies, FY 1985

Energy Technology Data Exchange (ETDEWEB)

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

1985-12-01

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

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.)

Tokamak experimental power reactor

International Nuclear Information System (INIS)

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

1978-01-01

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

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

Impurity transport studies on the FTU tokamak

International Nuclear Information System (INIS)

Pacella, D.; Romanelli, F.; Gregory, B.

1999-01-01

In this work, the radial profile of the diffusion coefficient D and the convective velocity V in the plasma core (0 2 /s and V ∼ 100 m/s. A model for the anomalous transport induced by electrostatic turbulence is developed. With a typical fluctuation spectrum (ω = 10 5 -2x10 5 Hz), calculations can reproduce very well the experimental results. To investigate the impurity behavior in a non-stationary phase, Kr gas was injected into the plasma. It is found that the total flux of Kr gas flowing into the core is also driven by diffusion but the magnitude is much lower than the single ion fluxes derived for Mo ions. The effect of the turbulence on the single ion is very strong but it is reduced when averaged over many charge states. (author)

Experimental methods to study tokamak plasma stability

International Nuclear Information System (INIS)

Perez-Navarro, A.

1978-01-01

Experimental devices to measure external instability modes with small pick-up coils to detect poloidal magnetic field fluctuations, and internal modes with soft-X-ray detectors are discussed. The characteristics of these devices are calculated for a small tokamak (R 0 = 30 cm, a = 10 cm, I 0 50 KA). (author)

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

Energy Technology Data Exchange (ETDEWEB)

Zhang, Ling; Xu, Zong; Wu, Zhenwei; Zhang, Pengfei; Wu, Chengrui; Gao, Wei; Shen, Junsong; Chen, Yingjie; Liu, Xiang; Wang, Yumin; Gong, Xianzu; Hu, Liqun; Chen, Junlin; Zhang, Xiaodong; Wan, Baonian; Li, Jiangang [Institute of Plasma Physics Chinese Academy of Sciences, Hefei 230026, Anhui (China); Morita, Shigeru; Ohishi, Tetsutarou; Goto, Motoshi [National Institute for Fusion Science, Toki 509-5292, Gifu (Japan); Department of Fusion Science, Graduate University for Advanced Studies, Toki 509-5292, Gifu (Japan); Dong, Chunfeng [Southwestern Institute of Physics, Chengdu 610041, Sichuan (China); and others

2015-12-15

A flat-field extreme ultraviolet (EUV) spectrometer working in the 20-500 Å wavelength range with fast time response has been newly developed to measure line emissions from highly ionized tungsten in the Experimental Advanced Superconducting Tokamak (EAST) with a tungsten divertor, while the monitoring of light and medium impurities is also an aim in the present development. A flat-field focal plane for spectral image detection is made by a laminar-type varied-line-spacing concave holographic grating with an angle of incidence of 87°. A back-illuminated charge-coupled device (CCD) with a total size of 26.6 × 6.6 mm{sup 2} and pixel numbers of 1024 × 255 (26 × 26 μm{sup 2}/pixel) is used for recording the focal image of spectral lines. An excellent spectral resolution of Δλ{sub 0} = 3-4 pixels, where Δλ{sub 0} is defined as full width at the foot position of a spectral line, is obtained at the 80-400 Å wavelength range after careful adjustment of the grating and CCD positions. The high signal readout rate of the CCD can improve the temporal resolution of time-resolved spectra when the CCD is operated in the full vertical binning mode. It is usually operated at 5 ms per frame. If the vertical size of the CCD is reduced with a narrow slit, the time response becomes faster. The high-time response in the spectral measurement therefore makes possible a variety of spectroscopic studies, e.g., impurity behavior in long pulse discharges with edge-localized mode bursts. An absolute intensity calibration of the EUV spectrometer is also carried out with a technique using the EUV bremsstrahlung continuum at 20-150 Å for quantitative data analysis. Thus, the high-time resolution tungsten spectra have been successfully observed with good spectral resolution using the present EUV spectrometer system. Typical tungsten spectra in the EUV wavelength range observed from EAST discharges are presented with absolute intensity and spectral identification.

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

Energy Technology Data Exchange (ETDEWEB)

Wootton, A.J.

1994-07-01

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

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

International Nuclear Information System (INIS)

Wootton, A.J.

1994-07-01

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

Experimental and theoretical study of particle transport in the TCV Tokamak

International Nuclear Information System (INIS)

Fable, E.

2009-06-01

The main scope of this thesis work is to compare theoretical models with experimental observations on particle transport in particular regimes of plasma operation from the Tokamak à Configuration Variable (TCV) located at CRPP–EPFL in Lausanne. We introduce the main topics in Tokamak fusion research and the challenging problems in the first Chapter. A particular attention is devoted to the modelling of heat and particle transport. In the second Chapter the experimental part is presented, including an overview of TCV capabilities, a brief review of the relevant diagnostic systems, and a discussion of the numerical tools used to analyze the experimental data. In addition, the numerical codes that are used to interpret the experimental data and to compare them with theoretical predictions are introduced. The third Chapter deals with the problem of understanding the mechanisms that regulate the transport of energy in TCV plasmas, in particular in the electron Internal Transport Barrier (eITB) scenario. A radial transport code, integrated with an external module for the calculation of the turbulence-induced transport coefficients, is employed to reproduce the experimental scenario and to understand the physics at play. It is shown how the sustainment of an improved confinement regime is linked to the presence of a reversed safety factor profile. The improvement of confinement in the eITB regime is visible in the energy channel and in the particle channel as well. The density profile shows strong correlation with the temperature profile and has a large local logarithmic gradient. This is an important result obtained from the TCV eITB scenario analysis and is presented in the fourth Chapter. In the same chapter we present the estimate of the particle diffusion and convection coefficients obtained from density transient experiments performed in the eITB scenario. The theoretical understanding of the strong correlation between density and temperature observed in the e

Atomic spectroscopy on fusion relevant ions and studies of light impurities in the JET tokamak

International Nuclear Information System (INIS)

Tunklev, M.

1999-03-01

The spectrum and energy levels of C IV and the 3l-4l system of the Mg-like ions in the iron group elements have been investigated. This has led to several hundred identified transitions, many of them previously unknown. Using the Charge Exchange Diagnostic system at JET, ion temperatures, rotation velocities and densities have been derived from visible spectroscopic measurements on fully ionised light impurities, such as He, C, N and Ne. The existence of plume contribution from beam produced hydrogen-like ions has been proven beyond any doubt to affect the deduction of the active charge exchange signal of He II. In the case of C VI the plume signal was estimated to be at least a factor of five lower than the active charge exchange signal. Line integrated passive charge exchange emission between neutral background atoms and fully stripped impurity ions has been investigated and modelled. When the synthetic spectrum is fitted into the experimentally detected spectra the neutral background density can be deduced. The importance of including background atoms (H, D and T) as charge exchange donors, not only in state 2s, but also in state 1s, has shown to be crucial in high temperature shots. Transport of light impurities has been studied with gas puff injections into steady state H-mode plasmas. The results suggest that light impurities are transported as described by the neo-classical Pfirsch-Schlueter regime at the edge, whilst in the centre, sawtoothing, preferably to Banana transport, is mixing the plasma and increases the measured values on the diffusion. For the peaking of impurities in a steady state plasma an anomalous treatment was more in agreement with the experimental data. Certain confinement information, previously predicted theoretically as a part of the peaking equation, has been experimentally verified

Control of edge localized modes by pedestal deposited impurity in the HL-2A tokamak

Science.gov (United States)

Zhang, Y. P.; Mazon, D.; Zou, X. L.; Zhong, W. L.; Gao, J. M.; Zhang, K.; Sun, P.; Dong, C. F.; Cui, Z. Y.; Liu, Yi; Shi, Z. B.; Yu, D. L.; Cheng, J.; Jiang, M.; Xu, J. Q.; Isobe, M.; Xiao, G. L.; Chen, W.; Song, S. D.; Bai, X. Y.; Zhang, P. F.; Yuan, G. L.; Ji, X. Q.; Li, Y. G.; Zhou, Y.; Delpech, L.; Ekedahl, A.; Giruzzi, G.; Hoang, T.; Peysson, Y.; Song, X. M.; Song, X. Y.; Li, X.; Ding, X. T.; Dong, J. Q.; Yang, Q. W.; Xu, M.; Duan, X. R.; Liu, Y.; the HL-2A Team

2018-04-01

Effect of the pedestal deposited impurity on the edge-localized mode (ELM) behaviour has been observed and intensively investigated in the HL-2A tokamak. Impurities have been externally seeded by a newly developed laser blow-off (LBO) system. Both mitigation and suppression of ELMs have been realized by LBO-seeded impurity. Measurements have shown that the LBO-seeded impurity particles are mainly deposited in the pedestal region. During the ELM mitigation phase, the pedestal density fluctuation is significantly increased, indicating that the ELM mitigation may be achieved by the enhancement of the pedestal transport. The transition from ELM mitigation to ELM suppression was triggered when the number of the LBO-seeded impurity exceeds a threshold value. During the ELM suppression phase, a harmonic coherent mode (HCM) is excited by the LBO-seeded impurity, and the pedestal density fluctuation is significantly decreased, the electron density is continuously increased, implying that HCM may reduce the pedestal turbulence, suppress ELMs, increase the pedestal pressure, thus extending the Peeling-Ballooning instability limit. It has been found that the occurance of the ELM mitigation and ELM suppression closely depends on the LBO laser spot diameter.

Impurity study experiment proposal

International Nuclear Information System (INIS)

1975-05-01

ISX is a modest tokamak which emphasizes the production of a predictable test plasma, experimental flexibility, ease of assembly and disassembly, and good diagnostic access. Its plasma models the outer cooler layers in EPR like plasmas. In addition, provisions will be made for long discharge times which may be necessary to observe some impurity effects. These machine characteristics will enable one to study the collisional transport of impurities in the plasma, perform systematic studies of wall and limiter materials and geometries, study methods of cleaning the walls, and develop and test new diagnostic techniques. ISX will employ water-cooled copper coils to produce a maximum toroidal magnetic field of 20 kG at the plasma axis, which is 77 cm from the major axis. The plasma minor radius will be about 15 cm, and the maximum plasma current will be 100 kA which will be induced by an iron core transformer with a capability of up to 0.9 volt-sec for long discharges. An aspect ratio of five and the modest magnetic field permit a design with ample space for thick wall structures such as honeycomb walls. The ''picture frame'' toroidal field coil provides additional space, while removable coil top sections allow easy replacement of the vacuum chamber. The 72-turn toroidal field coil is grouped into 24 sections for increased access. Absence of a conducting shell and placement of the vertical field and transformer primary coils away from the plasma allow easy viewing of the plasma and good diagnostic access. (U.S.)

Density and impurity profile behaviours in HL-2A tokamak with different gas fuelling methods

International Nuclear Information System (INIS)

Zheng-Ying, Cui; Yan, Zhou; Wei, Li; Bei-Bin, Feng; Ping, Sun; Chun-Feng, Dong; Yi, Liu; Wen-Yu, Hong; Qing-Wei, Yang; Xuan-Tong, Ding; Xu-Ru, Duan

2009-01-01

The electron density profile peaking and the impurity accumulation in the HL-2A tokamak plasma are observed when three kinds of fuelling methods are separately used at different fuelling particle locations. The density profile becomes more peaked when the line-averaged electron density approaches the Greenwald density limit n G and, consequently, impurity accumulation is often observed. A linear increase regime in the density range n e G and a saturation regime in n e > 0.6n G are obtained. There is no significant difference in achieved density peaking factor f ne between the supersonic molecular beam injection (SMBI) and gas puffing into the plasma main chamber. However, the achieved f ne is relatively low, in particular, in the case of density below 0.7n G , when the working gas is puffed into the divertor chamber. A discharge with a density as high as 1.2n G , i.e. n e = 1.2n G , can be achieved by SMBI just after siliconization as a wall conditioning. The metallic impurities, such as iron and chromium, also increase remarkably when the impurity accumulation happens. The mechanism behind the density peaking and impurity accumulation is studied by investigating both the density peaking factor versus the effective collisionality and the radiation peaking versus density peaking. (fluids, plasmas and electric discharges)

Measurement of impurity ion densities and energies in the divertor and edge regions of Alcator C-Mod tokamak

International Nuclear Information System (INIS)

Griem, H.R.; Moreno, J.; Welch, B.L.

1992-01-01

A study to investigate impurity production and transport in the divertor and edge regions of the Alcator C-Mod tokamak through spectroscopic techniques is described. A 0.75-meter Czerny-Turner spectrometer with a 1200-g/mm grating and a 35-meter quartz optic bundle transmission line were tested. A high-resolution 2-meter spectrometer will be ordered. Data acquisition considerations are being addressed

Magnetic field structure of experimental high beta tokamak equilibria

International Nuclear Information System (INIS)

Deniz, A.V.

1986-01-01

The magnetic field structure of several low and high β tokamaks in the Columbia High Beta Tokamak (HBT) was determined by high-impedance internal magnetic probes. From the measurement of the magnetic field, the poloidal flux, toroidal flux, toroidal current, and safety factor are calculated. In addition, the plasma position and cross-sectional shape are determined. The extent of the perturbation of the plasma by the probe was investigated and was found to be acceptably small. The tokamaks have major radii of approx.0.24 m, minor radii of approx.0.05 m, toroidal plasma current densities of approx.10 6 A/m 2 , and line-integrated electron densities of approx.10 20 m -2 . The major difference between the low and high β tokamaks is that the high β tokamak was observed to have an outward shift in major radius of both the magnetic center and peak of the toroidal current density. The magnetic center moves inward in major radius after 20 to 30 μsec, presumably because the plasma maintains major radial equilibrium as its pressure decreases from radiation due to impurity atoms. Both the equilibrium and the production of these tokamaks from a toroidal field stabilized z-pinch are modeled computationally. One tokamak evolves from a state with low β features, through a possibly unstable state, to a state with high β features

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)

Influence of an ergodic magnetic limiter on the impurity content in a tokamak

International Nuclear Information System (INIS)

Engelhardt, W.; Feneberg, W.

1978-01-01

This work deals with the properties of an ergodic magnetic limiter and presents calculations concerning the reduction of the impurity rate in a tokamak by a boundary sheath with decreased confinement time. The layer is produced by resonant helical windings superposed on an equilibrium magnetic field with closed magnetic surfaces. The transport coefficients in the boundary layer, which yield the temperature and density distribution, are obtained from the movement of particles along a stochastic magnetic field. The resulting line density can be made a factor of ten higher than is expected for a poloidal divertor experiment. From this it is concluded that all impurities coming from the wall will be ionized in the boundary layer. The concentration of the impurities in the plasma center is calculated according to a model which uses an anomalous diffusion coefficient being consistent with the ergodization in the boundary layer. The resulting concentration can be reduced proportional to the factor (nsub(e)Dsub(e)) -1 where nsub(e) and Dsub(e) are electron density and diffusion coefficient in the boundary layer. (Auth.)

A study on temperature effects on hydrogen recycling and molybdenum impurity emission from a movable limiter in TRIAM-1M Tokamak

International Nuclear Information System (INIS)

Bhattacharyay, R.; Zushi, H.; Nakashima, K.; Shikama, T.; Sakamoto, M.; Yoshida, N.; Kado, S.; Sawada, K.; Hirooka, Y.; Nakamura, K.; Hanada, K.; Idei, H.; Hasegawa, M.; Sato, K.N.; Ogawa, M.; Takaki, O.; Sasaki, K.; Xu, H.; Kawasaki, S.; Nakashima, H.; Higashijima, A.

2007-01-01

In order to investigate the surface temperature effects on plasma fuel recycling and impurity release from the plasma facing components, plasma discharges have been performed under selected plasma-wall interaction (PWI) conditions in the high-field superconducting tokamak, TRIAM-1M. By moving a water-cooled molybdenum movable limiter (ML) beyond the last closed flux surface, as defined by poloidal limiters, the surface temperature profile on it is varied. Hot spots have been observed on the ML surface in such conditions. The release behaviour of fuel as well as impurity particles from the ML surface has been studied as a function of hot spot temperature (T hot ) by means of wide range spectroscopy (200-1600 nm). A critical T hot is found to be ∼2100 K above which the emission of both hydrogen and impurity particles enhances significantly. This is indicative of some thermally activated process playing an important role in PWIs between the limiter and the edge plasma. With the rise in hot spot temperature localized PWI at the ML is found to dominate the global recycling even when external fuelling is stopped

The origin of metal impurities in DIVA

International Nuclear Information System (INIS)

Ohasa, Kazumi; Sengoku, Seio; Maeda, Hikosuke; Ohtsuka, Hideo; Yamamoto, Shin

1978-10-01

The origin of metal impurities in DIVA (JFT-2a Tokamak) has been studied experimentally. Three processes of metal impurity release from the first wall were identified; i.e. ion sputtering, evaporation, and arcing. Among of these, ion sputtering is the predominant process in the quiet phase of the discharge, which is characterized by no spikes in the loop voltage and no localized heat flux concentrations on the first wall. ''Cones'' formation due to the sputtering is observed on the gold protection plate (guard limiter) exposed to about 10,000 discharges by scanning electron micrograph. In the SEM photographs, the spacial distribution of cones on the shell surface due to the ion sputtering coincides with the spacial distribution of intensity of Au-I line radiation. Gold is the dominant metal impurity in DIVA. The honeycomb structure can decrease release of the metal impurity. (author)

Simplified models for radiational losses calculating a tokamak plasma

International Nuclear Information System (INIS)

Arutiunov, A.B.; Krasheninnikov, S.I.; Prokhorov, D.Yu.

1990-01-01

To determine the magnitudes and profiles of radiational losses in a Tokamak plasma, particularly for high plasma densities, when formation of MARFE or detached-plasma takes place, it is necessary to know impurity distribution over the ionization states. Equations describing time evolution of this distribution are rather cumbersome, besides that, transport coefficients as well as rate constants of the processes involving complex ions are known nowadays with high degree of uncertainty, thus it is believed necessary to develop simplified, half-analytical models describing time evolution of the impurities analysis of physical processes taking place in a Tokamak plasma on the base of the experimental data. (author) 6 refs., 2 figs

Impurity behavior during ion-Bernstein wave heating in PBX-M

Science.gov (United States)

Isler, R. C.; Post-Zwicker, A. P.; Paul, S. F.; Tighe, W.; Ono, M.; Leblanc, B. P.; Bell, R.; Kugel, H. W.; Kaita, R.

1994-07-01

Ion-Bernstein-wave heating (IBWH) has been tested in several tokamaks. In some cases the results have been quite positive, producing temperature increases and also improving both energy and particle confinement times, whereas in others, no distinctive changes were observed. Most recently, IBWH has been utilized in the Princeton Beta Experiment-Modified (PBX-M) where the long-range goal is the achievement of operation in the second stable region by current and pressure profile control. Investigations have been performed in this machine using IBWH as the sole source of auxiliary power or using IBWH in conjunction with neutral-beam injection (NBI) or with lower-hybrid current drive (LHCD). Impurity studies seem particularly important for IBWH since not only have influxes often been observed to increase, but the global impurity confinement time has also been shown to lengthen as the confinement of the working gas improved. The authors present here a set of characteristic experimental results regarding the impurity behavior in PBX-M; in general, these are consonant with previous observations in other tokamaks.

Experimental result of poloidal limiter baking of Aditya tokamak

International Nuclear Information System (INIS)

Jadeja, K.A.; Arambhadiya, B.G.; Bhatt, S.B.; Bora, D.

2005-01-01

In tokamak Aditya, Poloidal limiter function as the operational limiter and are subjected to very high particles load and heat flux during plasma discharge. In addition, Poloidal limiter is the first material surface to come in contact with the hot plasma. In plasma discharge, the impurity generations from limiter are mostly by adsorbed particles. The baking of limiter provides high degassing rate and thermal desorption of adsorbed particles of limiter to reduce impurities from the limiter tiles. The series of experiments are done with different conditions like, Baking of limiter SS ring by heating element with and without limiter tiles in atmosphere and vacuum. Than Poloidal limiter is structured with 14 numbers of graphite tiles and electrical isolated to the vessel and support structure. As a heating element and for electrical isolation, Nychrome wire and ceramic block with ceramic tubes are used. In addition, Thermo couple and two DC power supply (0-10 Ampere) are used for limiter baking. Mass analyzer gives partial pressures of different species to observe effect of limiter baking. For the period of Poloidal limiter baking in Aditya, the partial pressures of different species like hydrogen, water vapor, and oxygen are extremely increased with time duration. This paper presents series of experimental results of poloidal limiter baking. (author)

Neutronic scoping studies for the tokamak experimental power reactor

International Nuclear Information System (INIS)

Santoro, R.T.; Bettis, E.S.; McAlees, D.G.; Watts, H.L.; Williams, M.L.

1976-02-01

One-dimensional neutron and photon radiation transport methods have been used to investigate candidate blanket configurations and compositions for use in the Tokamak Experimental Power Reactor. Seven blanket designs are compared in terms of energy recovery, radiation attenuation, potential radiation damage, and, where applicable, tritium breeding

Varennes Tokamak

International Nuclear Information System (INIS)

Cumyn, P.B.

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

Studies on impurity control and hydrogen pumping with chromium gettering in ISX-B

International Nuclear Information System (INIS)

Mioduszewski, P.; Simpkins, J.E.; Edmonds, P.H.; Isler, R.C.; Lazarus, E.A.; Ma, C.H.; Murakami, M.; Wootton, A.J.

1984-01-01

Chromium gettering has been proven to be a trouble-free and efficient method of surface pumping in tokamaks. The impurity control capabilities are excellent and comparable to that of titanium. The hydrogen uptake is reduced to monolayer quantities on the surface. The expansion of the operating space is similar to that seen with titanium without the disadvantage of strongly increased hydrogen fluxes. Possible applications of chromium gettering are: impurity control in contemporary tokamaks; surface pumping in short pulse DT-burning devices to minimize tritium inventory, and wall conditioning of future large machines prior to operation

Impurity transport in the Wendelstein VII-A stellarator

International Nuclear Information System (INIS)

1985-01-01

Impurity radiation losses in net-current-free neutral-beam-heated plasmas in the Wendelstein W VII-A stellarator are the combined effect of particularly strong impurity sources and improved particle confinement as compared with ohmically heated tokamak-like plasma discharges. Experiments are described and conclusions are drawn about the impurity species, their origin and their transport behaviour. The impurity transport is modelled by a 1-D impurity transport and radiation code. The evolution of the total radiation in time and space deduced from soft-X-ray and bolometer measurements can be fairly well simulated by the code. Experimentally, oxygen was found to make the main contribution to the radiation losses. In the calculations, an influx of cold oxygen desorbed from the walls of the order of 10 13 -10 14 cm -2 .s -1 and a rate of fast injected oxygen corresponding to a 1% impurity content of the neutral beams in combination with neoclassical impurity transport leads to quantitative agreement between the simulation and the observed radiation. The transport of A1 trace impurities injected by the laser blow-off technique was experimentally studied by soft-X-ray measurements using a differential method allowing extraction of the time evolution of A1 XII, XIII radial profiles. These are compared with code predictions, together with additional spectroscopic measurements. The main features of the impurity transport are consistent with neoclassical predictions, which explain particularly the central impurity accumulation. Some details, however, seem to require additional 'anomalous' transport. Such an enhancement is correlated with distortions of the magnetic configuration around resonant magnetic surfaces. (author)

Impurity studies in fusion devices using laser-fluorescence-spectroscopy

International Nuclear Information System (INIS)

Husinsky, W.R.

1980-08-01

Resonance fluorescence excitation of neutral atoms using tunable radiation from dye lasers offers a number of unique advantages for impurity studies in fusion devices. Using this technique, it is possible to perform local, time-resolved measurements of the densities and velocity distributions of metallic impurities in fusion devices without disturbing the plasma. Velocities are measured by monitoring the fluorescence intensity while tuning narrow bandwidth laser radiation through the Doppler - broadened absorbtion spectrum of the transition. The knowledge of the velocity distribution of neutral impurities is particularly useful for the determination of impurity introduction mechanisms. The laser fluorescence technique will be described in terms of its application to metallic impurities in fusion devices and related laboratory experiments. Particular attention will be given to recent results from the ISX-B tokamak using pulsed dye lasers where detection sensitivities for neutral Fe of 10 6 atoms/cm 3 with a velocity resolution of 600 m/sec (0.1 eV) have been achieved. Techniques for exciting plasma particles (H,D) will also be discussed

TFR, the tokamak of Fontenay-aux-Roses

International Nuclear Information System (INIS)

1985-06-01

Twelve years of the TFR tokamak operation are briefly reviewed. In the historical introduction, justifications of the experiment and main development stages are described. The importance of the choice of materials in contact with the plasma (impurities production) appears in the various paragraphs. The power lost by impurity radiation is compared with other losses in ohmic conditions (paragraph 2). Additional heating experiments: neutral beam heating paragraph 3 and ion cyclotron heating paragraph 4 are reported; their efficiency as well as deleterious induced effects are described. Important diagnostics development to measure impurities were made, giving experimental results used also in astrophysics and atomic physics. In the last paragraph magnetohydrodynamic phenomena are reported

Multiscaling Dynamics of Impurity Transport in Drift-Wave Turbulence

International Nuclear Information System (INIS)

Futatani, S.; Benkadda, S.; Nakamura, Y.; Kondo, K.

2008-01-01

Intermittency effects and the associated multiscaling spectrum of exponents are investigated for impurities advection in tokamak edge plasmas. The two-dimensional Hasagawa-Wakatani model of resistive drift-wave turbulence is used as a paradigm to describe edge tokamak turbulence. Impurities are considered as a passive scalar advected by the plasma turbulent flow. The use of the extended self-similarity technique shows that the structure function relative scaling exponent of impurity density and vorticity follows the She-Leveque model. This confirms the intermittent character of the impurities advection in the turbulent plasma flow and suggests that impurities are advected by vorticity filaments

Space-resolved vacuum ultra-violet spectroscopy on T.F.R. Tokamak plasmas

International Nuclear Information System (INIS)

1978-01-01

Results are reported of space-resolved vacuum-ultraviolet spectroscopy (between 100 A and 2000A) on T.F.R. Tokamak plasmas and examples are given of profiles for both heavy and light impurity ions. The experimental method and the associated uncertainties and problems are stressed. The great importance of numerical calculations in the interpretation of the impurity profiles is pointed out. (author)

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)

On impurities transport in a tokamak

International Nuclear Information System (INIS)

Rozhanskij, V.A.

1980-01-01

Transport of impurity ions is analitically analized in the case when main plasma is in plateau or banana regimes but impurity ions - in the Pfirsch-Schlutter mode. It is shown that in the large region of parameters the impUrity transport represents a drift in a p oloidal electric field, averaged from magnetic surface with provision for disturbance of concentration on it. Therefore, transport velocity does not depend on Z value and impurity type, as well as collision frequency both in the plateau and banana regimes. A value of flows is determined by the value of poloidal rotation velocity. At the rotation velocity corresponding to the electric field directed from the centre to periphery impurities are thrown out of a discharge, in the reverse case the flow is directed inside. Refusal from the assumption that Zsub(eff) > approximately 2, does not considerably change the results of work. The approach developed in the process of work can be applied to the case when impurity ions are in the plateau or banana modes

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.)

Demonstration tokamak power plant

International Nuclear Information System (INIS)

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

1983-01-01

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

ICRF heating experiments in JFT-2 tokamak

International Nuclear Information System (INIS)

Matsumoto, Hiroshi

1986-01-01

This is an experimental study of ICRF heating on JFT-2 Tokamak in Japan Atomic Energy Research Institute. In this study, we first clarified physical and engineering problems of ICRF heating of tokamak plasma. Next, we optimized the design of the ICRF heating system, and the plasma parameters for the heating. Finally, we could demonstrate a high efficiency of this additional heating method by launching RF power which is two or three times as large as an ohmic input power to a plasma. And we achieved following things. (1) We optimized a design of an antenna, and we improved a durability of the system for high voltage. With the result that we achieved the maximum power density on an antenna. (2) We demonstrated that electron heating regime and ion heating regime can be easily accessed by controlling plasma parameters. Also we found the optimum heating conditions in each heating regime. (3) We experimentally clarified the production mechanism of impurities during ICRF heating. We could reduce the influx of metal impurity ions to a plasma by employing low z materials for limiters and antenna shields. Consequently, we improved a heating efficiency of electrons. Next, we studied a power balance of plasma during ICRF heating, and we could compare heating characteristics of ICRF with other additional heatings on JFT-2. (author)

Effects of classical and neo-classical cross-field transport of tungsten impurity in realistic tokamak geometry

Energy Technology Data Exchange (ETDEWEB)

Yamoto, S.; Inoue, H.; Sawada, Y.; Hatayama, A. [Faculty of Science and Technology, Keio University, Yokohama (Japan); Homma, Y.; Hoshino, K. [Japan Atomic Energy Agency, Rokkasho, Aomori (Japan); Bonnin, X. [ITER Organization, St. Paul Lez Durance (France); Coster, D. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Schneider, R. [Ernst-Moritz-Arndt University Greifswald (Germany)

2016-08-15

The initial simulation study of the neoclassical perpendicular self-diffusion transport in the SOL/Divertor regions for a realistic tokamak geometry with the IMPGYRO code has been performed in this paper. One of the most unique features of the IMPGYRO code is calculating exact Larmor orbit of the test particle instead of assuming guiding center approximation. Therefore, effects of the magnetic drifts in realistic tokamaks are naturally taken into account in the IMPGYRO code. This feature makes it possible to calculate neoclassical transport processes, which possibly become large in the SOL/divertor plasma. Indeed, neoclassical self-diffusion process, the resultant effect of the combination of magnetic drift and Coulomb collisions with background ions, has already been included in the IMPGYRO model. In the present paper, prior to implementing the detailed model of neoclassical transport process into IMPGYRO, we have investigated the effect of neoclassical selfdiffusion in a realistic tokamak geometry with lower single null X-point. We also use a model with guiding center approximation in order to compare with the IMPGYRO full orbit model. The preliminary calculation results of each model have shown differences in the perpendicular average velocity of impurity ions at the top region of the SOL. The mechanism which leads to the difference has been discussed. (copyright 2015 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

Light impurity production in tokamaks

International Nuclear Information System (INIS)

Philipps, V.; Vietzke, E.; Erdweg, M.

1989-01-01

A review is given of the different erosion processes of carbon materials with special emphasis on conditions relevant to plasma surface interactions. New results on the chemical erosion and radiation enhanced sublimation of boron-carbon layers are presented. The chemical hydrocarbon formation produced by the interaction of the TEXTOR scrape-off plasma with a carbon target has been investigated up to temperatures of 1500K using a Sniffer probe. The chemical interaction of the plasma with the carbon walls in TEXTOR is also analysed by measuring the hydrocarbon and CO and CO 2 partial pressures built up on the surrounding walls during the discharges. The recycling of oxygen impurities in an all carbon surrounding occurs predominantly in the form of CO and Co 2 molecules and the analysis of both neutral pressures during the discharges has been used as an additional diagnostic for the oxygen impurity situation in TEXTOR. These data are discussed in view of spectroscopic measurements on the influx of carbon and oxygen atoms from the walls and impurity line radiation. CD-band spectroscopy in addition is employed to identify the hydrocarbon chemical carbon erosion. Our present understanding of the oxygen impurity recycling and the oxygen sources are described. Particle induced release of CO molecules from the entire first wall is believed to be the dominant influx process of oxygen in the SOL of plasmas with carbon facing materials. The influence of coating the TEXTOR first wall with a boron-carbon film (B/C ≅1) on the light impurity behaviour is shown. (author)

Surface effects and impurity production in tokamak machines

International Nuclear Information System (INIS)

Staib, P.; Staudenmaier, G.

1978-01-01

Plasma-wall interactions are presently investigated in two ways: a) The a priori assumption of a mechanism responsible for impurity release. Relevant experimental data can be used in a model and calculations made in order to understand the observed impurity behaviour in plasma. b) The comprehensive investigation of samples exposed to a plasma. Recent investigations have confirmed the earlier assumption that the interactions occur only in the topmost atomic layers of the wall, and so emphasize the major role of surface physics on this field. These investigations have further shown that besides atomic processes, such a desorption or sputtering, other processes occur, extending on a microscopic rather than an atomic scale. These are for example evaporation, arcing, and mechanical stress. Both aspects are discussed as far as possible in a quantitative way. The contribution of most probable processes is estimated using data available on flux and energy of particles and yields of single processes. The conclusion is reached that no process can be disregarded. Several processes seem to contribute to the impurity release and are different at different phases of the discharge. An interdependence between these processes is likely. (Auth.)

On the origin, properties, and implications of asymmetries in the tungsten impurity density in tokamak plasmas

Energy Technology Data Exchange (ETDEWEB)

Odstrcil, Tomas

2017-07-03

In this thesis, the transport of tungsten ions is studied in the plasma of ASDEX Upgrade tokamak. The plasma facing components of the fusion reactors are expected to be built from high-Z materials such as W, Mo or Fe. These materials provide advantages like a high melting point, small erosion rates, and low tritium retention. However, due to the interaction of the plasma with the wall, ions of this material will be inevitably present also in the main plasma. These ions are not entirely stripped even at fusion plasma temperatures, and therefore emit strong line radiation, which can significantly degrade the performance of the fusion plasma. Thus the understanding and control of impurity transport are of critical importance to the success of fusion. The high mass and charge of the heavy impurities make them susceptible to some of the forces acting upon the plasma, resulting in a poloidal variation of their density. The most prominent are the centrifugal force arising from the plasma rotation and the electric force caused by magnetically trapped non-thermal ions. Furthermore, the poloidal asymmetries should have a significant impact on the radial transport of heavy ions, which was widely ignored up to date. In the present work, the poloidal asymmetries in the heavy impurity density were inferred from the soft X-ray radiation using a newly developed tomographic method. The high accuracy of the tomography and of the model for the centrifugal force allowed to identify for the first time in an experiment the effect of the fast ion distribution produced by neutral beam injection on the poloidal asymmetry of the tungsten density. The measured asymmetry was compared to several fast ion models, and the best match was found with the Monte Carlo code in the TRANSP code suite that includes finite orbits effects of the fast ions. Similarly, fast ions accelerated by ion cyclotron heating and localized mainly in the outboard side of the plasma due to a magnetic trapping and produce

Tokamak power systems studies, FY 1986: A second stability power reactor

International Nuclear Information System (INIS)

Ehst, D.; Baker, C.; Billone, M.

1987-03-01

This report presents the results of the work at Argonne National Laboratory (ANL) during FY-1986 on the Tokamak Power Systems Study (TPSS). The purpose of the TPSS is to explore and develop ideas that would lead to improvements in the tokamak as a power reactor concept. The work at ANL concentrated on plasma engineering, impurity control, and the blanket/first wall/shield system. The work in FY-1986 extended these studies and focused them on a reference design point. The key features of the design point include: second stability regime with higher β and larger aspect ratio, steady-state operation with fast wave current drive, impurity control via a self-pumped slot limiter, a self-cooled liquid lithium, vanadium alloy blanket with simplified poloidal flow, and reduced reactor building volume with vertical lift maintenance. Sufficient work was carried out to report a preliminary cost estimate. In addition, reactor implications of steady-state operation in the first stability regime were also studied. 174 refs., 124 figs., 65 tabs

Experimental study of external kink instabilities in the Columbia High Beta Tokamak

International Nuclear Information System (INIS)

Ivers, T.H.

1991-01-01

The generation of power through controlled thermonuclear fusion reactions in a magnetically confined plasma holds promise as a means of supplying mankind's future energy needs. The device most technologically advanced in pursuit of this goal is the tokamak, a machine in which a current-carrying toroidal plasma is thermally isolated from its surroundings by a strong magnetic field. To be viable, the tokamak reactor must produce a sufficiently large amount of power relative to that needed to sustain the fusion reactions. Plasma instabilities may severely limit this possibility. In this work, I describe experimental measurements of the magnetic structure of large-scale, rapidly-growing instabilities that occur in a tokamak when the current or pressure of the plasma exceeds a critical value relative to the magnetic field, and I compare these measurements with theoretical predictions

Experimental and calculating study on the stressed state of superconducting coils of toroidal field in the T-15 tokamak

International Nuclear Information System (INIS)

Vaulina, I.G.; Gusev, S.V.; Sivkova, G.N.

1987-01-01

Results of calculational and experimental atudy of stress-deformed state of superconducting coils of the T-15 tokamak toroidal field are presented. The calculations are made using the method of finite elements and refined theory of cores. Experimental studies were carried out using elastic tensometric model of polymer materials. Test results are compared with the calculational results. Divergence between calculational and experimental values of displacement of characteristic points in the unit does not exceed 20 %. Results of model studies confirm the expediency of the calculational model used for designing SOTP unit for the T-15 tokamak

Experimental study of electron temperature gradient influence on impurity turbulent transport in fusion plasmas

International Nuclear Information System (INIS)

Villegas, D.

2010-01-01

Understanding impurity transport is a key to an optimal regime for a future fusion device. In this thesis, the theoretical and experimental influence of the electron temperature gradient R/L Te on heavy impurity transport is analyzed both in Tore Supra and ASDEX Upgrade. The electron temperature profile is modified locally by heating the plasma with little ECRH power deposited at two different radii. Experimental results have been obtained with the impurity transport code (ITC) which has been completed with a genetic algorithm allowing to determine the transport coefficient profiles with more accuracy. Transport coefficient profiles obtained by a quasilinear gyrokinetic code named QuaLiKiz are consistent with the experimental ones despite experimental uncertainties on gradients. In the core dominated by electron modes, the lower R/L Te the lower the nickel diffusion coefficient. The latter tends linearly to the neoclassical level when the instability threshold is approached. The experimental threshold is in agreement with the one computed by QuaLiKiz. Further out, where the plasma is dominated by ITG, which are independent of R/L Te , both experimental and simulated results show no modification in the diffusion coefficient profile. Furthermore, the convection velocity profile is not modified. This is attributed to a very small contribution of the thermodiffusion (1/Z dependence) in the total convection. On ASDEX, the preliminary results, very different from the Tore Supra ones, show a internal transport barrier for impurities located at the same radius as the strong ECRH power deposit. (author) [fr

An enhanced tokamak startup model

Science.gov (United States)

Goswami, Rajiv; Artaud, Jean-François

2017-01-01

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

Soft x-ray measurements in the TRIAM-1 tokamak

Energy Technology Data Exchange (ETDEWEB)

Satoh, T; Toi, K; Nakamura, K; Nakamura, Y; Hiraki, N [Kyushu Univ., Fukuoka (Japan). Research Inst. for Applied Mechanics

1981-07-01

Soft X-ray pulse height analysis system has been designed and constructed for measurements of electron distribution function and impurity with high spatial resolution (0.5 cm) and temporal resolution (2 msec) in the TRIAM-1 tokamak. The experimental results about electron temperature, enhancement factor, Z sub(eff) and runaway electrons are presented and discussed.

Effect of impurities on kinetic transport processes in fusion plasmas

Energy Technology Data Exchange (ETDEWEB)

Braun, Stefanie

2010-12-10

Within the framework of this thesis, different problems arising in connection with impurities have been investigated. Collisional damping of zonal flows in tokamaks: Since the Coulomb collision frequency increases with increasing ion charge, heavy, highly charged impurities play an important role in this process. The effect of such impurities on the linear response of the plasma to an external potential perturbation, as caused by zonal flows, is calculated with analytical methods. In comparison with a pure plasma, the damping of the flows occurs, as expected, considerably faster; for experimentally relevant parameters, the enhancement exceeds the effective charge Z{sub eff} of the plasma. Impurity transport driven by microturbulence in tokamaks: With regard to impurities, it is especially important whether the resulting flows are directed inwards or outwards, since they are deleterious for core energy confinement on the one hand, but on the other hand help protecting plasma-facing components from too high energy fluxes in the edge region. A semi-analytical model is presented describing the resulting impurity fluxes and the stability boundary of the underlying mode. The main goal is to bridge the gap between, on the one hand, costly numerical simulations, which are applicable to a broad range of problems but yield scarcely traceable results, and, on the other hand, analytical theory, which might ease the interpretation of the results but is so far rather rudimentary. The model is based on analytical formulae whenever possible but resorts to a numerical treatment when the approximations necessary for an analytical solution would lead to a substantial distortion of the results. Both the direction of the impurity flux and the stability boundary are found to depend sensitively on the plasma parameters such as the impurity density and the temperature gradient. Pfirsch-Schlueter transport in stellarators: Due to geometry effects, collisional transport plays a much more

The ARIES tokamak fusion reactor study

International Nuclear Information System (INIS)

Bartlit, J.R.; Bathke, C.G.; Krakowski, R.A.; Miller, R.L.; Beecraft, W.R.; Hogan, J.T.; Peng, Y.K.M.; Reid, R.L.; Strickler, D.J.; Whitson, J.C.; Blanchard, J.P.; Emmert, G.A.; Santarius, J.F.; Sviatoslavsky, I.N.; Wittenberg, L.J.

1989-01-01

The ARIES study is a community effort to develop several visions of the tokamak as fusion power reactors. The aims are to determine their potential economics, safety, and environmental features 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 in 2nd stability regime and employs both potential advances in the physics and expected advances in technology and engineering; and ARIES-III is a conceptual D 3 He reactor. This paper focuses on the ARIES-I design. Parametric systems studies show that the optimum 1st stability tokamak has relatively low plasma current (∼ 12 MA), high plasma aspect ratio (∼ 4-6), and high magnetic field (∼ 24 T at the coil). ARIES-I is 1,000 MWe (net) reactor with a plasma major radius of 6.5 m, a minor radius of 1.4 m, a neutron wall loading of about 2.8 MW/m 2 , and a mass power density of about 90 kWe/ton. The ARIES-I reactor operates at steady state using ICRF fast waves to drive current in the plasma core and lower-hybrid waves for edge-plasma current drive. The current-drive system supplements a significant (∼ 57%) bootstrap current contribution. The impurity control system is based on high-recycling poloidal divertors. Because of the high field and large Lorentz forces in the toroidal-field magnets, innovative approaches with high-strength materials and support structures are used. 24 refs., 4 figs., 1 tab

Tokamak experimental section

International Nuclear Information System (INIS)

Berry, L.A.; Dunlap, J.L.; Arakawa, E.T.

1977-01-01

Descriptions of research during this period are given for the following topics: (1) ion and electron heating, (2) high-beta and gas puff experiments, (3) beam trapping by impurities, (4) counterinjection studies, (5) impurity measurements, (6) Balmer alpha line profiles, (7) internal mode structure, (8) sawtooth oscillations and plasma transport, (9) Ormak plasma modeling, (10) charge exchange measurements, (11) wall power measurements, (12) neutron time behavior due to deuterium neutral beam injection into a hydrogen plasma, (13) wall impurities in Ormak, (14) relativistic electron studies, (15) fast x-ray energy analyzer for the 1 to 10 keV range, and (16) CTR related atomic physics

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.)

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

Grazing incidence EUV study of the Alcator tokamaks

International Nuclear Information System (INIS)

Castracane, J.

1982-01-01

The use of impurity radiation to examine plasma conditions is a well known technique. To gain access, however, to the hot, central portion of the plasma created in the present confinement machines it is necessary to be able to observe radiation from medium and heavy elements such as molybdenum and iron. These impurities radiate primarily in the extreme ultra violet region of the spectrum and can play a role in the power balance of the tokamak. Radiation from highly ionized molybdenum was examined on the Alcator A and C tokamaks using a photometrically calibrated one meter grazing incidence monochromator. On Alcator A, a pseudo-continuum of Mo emissions in the 60 to 100 A ranges were seen to comprise 17% of the radiative losses from the plasma. This value closely matched measurements by a broad band bolometer array. Following these preliminary measurements, the monochromator was transferred to Alcator C for a more thorough examination of EUV emissions. Deviations from predicted scaling laws for energy confinement time vs density were observed on this machine

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.

Nonequilibrium effects and structure of X-ray lines in tokamak plasma

Science.gov (United States)

Gontis, V. G.; Lisitsa, V. S.

1986-02-01

The sensitivity of X-ray spectra to a number of typical non-equilibrium effects occurring in modern tokamaks is examined. Experimental data from the T-10 and ST Tokamaks are cited to illustrate the degree of deviation from coronal equilibrium. The analysis exploits recent atomic data for radiation and autoionization line widths; standard semiempirical formulas are used to calculate the rates of collision processes. Ion diffusion and impurity distribution by degrees of ionization are investigated. The sensitivity of K radiation to electron nonequilibrium and ion charge exchange is examined.

Sawtooth driven particle transport in tokamak plasmas

International Nuclear Information System (INIS)

Nicolas, T.

2013-01-01

The radial transport of particles in tokamaks is one of the most stringent issues faced by the magnetic confinement fusion community, because the fusion power is proportional to the square of the pressure, and also because accumulation of heavy impurities in the core leads to important power losses which can lead to a 'radiative collapse'. Sawteeth and the associated periodic redistribution of the core quantities can significantly impact the radial transport of electrons and impurities. In this thesis, we perform numerical simulations of sawteeth using a nonlinear tridimensional magnetohydrodynamic code called XTOR-2F to study the particle transport induced by sawtooth crashes. We show that the code recovers, after the crash, the fine structures of electron density that are observed with fast-sweeping reflectometry on the JET and TS tokamaks. The presence of these structure may indicate a low efficiency of the sawtooth in expelling the impurities from the core. However, applying the same code to impurity profiles, we show that the redistribution is quantitatively similar to that predicted by Kadomtsev's model, which could not be predicted a priori. Hence finally the sawtooth flushing is efficient in expelling impurities from the core. (author) [fr

Collisional-radiative models for hydrogen-like and helium-like carbon and oxygen ions and applications to experimental data from the TS Tokamak and the reversed field pinch RFX

International Nuclear Information System (INIS)

Carraro, L.; Sattin, F.; Puiatty, M.E.; Scarin, P.; Valisa, M.; Mattioli, M.; Demichelis, C.; Mandl, W.

1996-07-01

Collisional radiative models (CRM) are needed to simulate experimental line brightnesses and emissivities from fusion devices. CRM are built for H-like and He-like carbon and oxygen ions. The impurity ion radial distribution is obtained using a transport code with two radius dependent transport parameters: a diffusion coefficient D and an inward convection velocity V. Examples are given of the quantitative interpretation of experimental spectroscopic data from two fusion devices: the Tore Supra Tokamak and the Reversed Field Pinch RFX. (K.A.)

Economic trends of tokamak power plants independent of physics scaling models

International Nuclear Information System (INIS)

Reid, R.L.; Steiner, D.

1978-01-01

This study examines the effects of plasma radius, field on axis, plasma impurity level, and aspect ratio on power level and unit capital cost, $/kW/sub e/, of tokamak power plants sized independent of plasma physics scaling models. It is noted that tokamaks sized in this manner are thermally unstable based on trapped particle scaling relationships. It is observed that there is an economic advantage for larger power level tokamaks achieved by physics independent sizing; however, the incentive for increased power levels is less than that for fission reactors. It is further observed that the economic advantage of these larger power level tokamaks is decreased when plasma thermal stability measures are incorporated, such as by increasing the plasma impurity concentration. This trend of economy with size obtained by physics independent sizing is opposite to that observed when the tokamak designs are constrained to obey the trapped particle and empirical scaling relationships

Pyroelectric detector study and realization measuring the plasma radiated power in a tokamak

International Nuclear Information System (INIS)

Simonet, F.

1981-10-01

The study of a additional heating method and the perfection of impurities rate control and reduction means are presently actively investigated. Petula experiment must demonstrate heating efficiency by high frequency oscillating electromagnetic fields. Impurities will probably dissipate an important part of the ohmic power and electromagnetic power left in plasma. In this report, experimental device is described, which has been realized, and introduced in the tokamak, to measure precisely the energy losses by radiation in the ionized medium. In a first part, tokomak Petula is presented and it is shown how different chemical species can introduce numerously in the discharge gas. In a second part, plasma cooling by photon and fast neutron strong emission is stressed on. In a third part, the measuring device is explained; the detector part is a pyroelectric crystal. In a fourth and last part, results are discussed, insisting on the signal temporal evolution and on the value of the following ratio: power lost by plasma towards the walls/ohmic power left in plasma [fr

Silicon drift detector based X-ray spectroscopy diagnostic system for the study of non-thermal electrons at Aditya tokamak.

Science.gov (United States)

Purohit, S; Joisa, Y S; Raval, J V; Ghosh, J; Tanna, R; Shukla, B K; Bhatt, S B

2014-11-01

Silicon drift detector based X-ray spectrometer diagnostic was developed to study the non-thermal electron for Aditya tokamak plasma. The diagnostic was mounted on a radial mid plane port at the Aditya. The objective of diagnostic includes the estimation of the non-thermal electron temperature for the ohmically heated plasma. Bi-Maxwellian plasma model was adopted for the temperature estimation. Along with that the study of high Z impurity line radiation from the ECR pre-ionization experiments was also aimed. The performance and first experimental results from the new X-ray spectrometer system are presented.

L-H power threshold studies with tungsten/carbon divertor on the EAST tokamak

DEFF Research Database (Denmark)

Chen, L.; Xu, G. S.; Gao, W.

2016-01-01

The power threshold for low (L) to high (H) confinement mode transition achieved by radio-frequency heating and molybdenum first wall with lithium coating has been experimentally investigated on the EAST tokamak for two sets of divertor geometries and materials: tungsten/carbon divertor and full...... carbon divertor. For both sets of divertors, the power threshold was found to decrease with gradual accumulation of the lithium wall coating, suggesting the important role played by the low Z impurities and/or the edge neutral density on the L-H power threshold. When operating in the upper single null...

Transport simulations of a density limit in radiation-dominated tokamak discharges: II

International Nuclear Information System (INIS)

Stotler, D.P.

1991-05-01

The procedures developed previously to simulate the radiatively induced tokamak density limit are used to examine in more detail the scaling of the density limit. It is found that the maximum allowable density increases with auxiliary power and decreases with impurity concentration. However, it is demonstrated that there is little dependence of the density limit on plasma elongation. These trends are consistent with experimental results. Our previous work used coronal equilibrium impurities; the primary result of that paper was that the maximum density increases with current when peaked profiles are assumed. Here, this behavior is shown to occur with a coronal nonequilibrium impurity as well. 26 refs., 4 figs

Surface changes on probes inserted into the limiter shadow of the T-10 tokamak

International Nuclear Information System (INIS)

Chicherov, V.M.; Hildebrandt, D.; Laux, M.; Lingertat, J.; Lukianov, S.U.; Pech, P.; Reiner, H.D.; Stepanchikov, V.A.; Wolff, H.

1980-01-01

Time-resolved measurements of impurity fluxes which are oriented parallel and antiparallel to the toroidal and poloidal magnetic field have been performed in the limiter shadow of the T-10 tokamak. A qualitative model is proposed which explains the main features of the experimental results. (orig.)

Collisional-radiative models for hydrogen-like and helium-like carbon and oxygen ions and applications to experimental data from the TS Tokamak and the reversed field pinch RFX

Energy Technology Data Exchange (ETDEWEB)

Carraro, L.; Sattin, F.; Puiatty, M.E.; Scarin, P.; Valisa, M. [Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy); Mattioli, M.; Demichelis, C.; Mandl, W. [Association Euratom-CEA, Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Hogan, J.T. [Oak Ridge National Lab., TN (United States)

1996-07-01

Collisional radiative models (CRM) are needed to simulate experimental line brightnesses and emissivities from fusion devices. CRM are built for H-like and He-like carbon and oxygen ions. The impurity ion radial distribution is obtained using a transport code with two radius dependent transport parameters: a diffusion coefficient D and an inward convection velocity V. Examples are given of the quantitative interpretation of experimental spectroscopic data from two fusion devices: the Tore Supra Tokamak and the Reversed Field Pinch RFX. (K.A.). 60 refs.

Spectroscopic Measurements of Impurity Spectra on the EAST Tokamak

International Nuclear Information System (INIS)

Fu Jia; Li Yingying; Shi Yuejiang; Wang Fudi; Zhang Wei; Lv Bo; Huang Juan; Wan Baonian; Zhou Qian

2012-01-01

Ultraviolet (UV) and visible impurity spectra (200∼750 nm) are commonly used to study plasma and wall interactions in magnetic fusion plasmas. Two optical multi-channel analysis (OMA) systems have been installed for the UV-visible spectrum measurement on EAST. These two OMA systems are both equipped with the Czerny-Turner (C-T) type spectrometer. The upper vacuum vessel and inner divertor baffle can be viewed simultaneously through two optical lenses. The OMA1 system is mainly used for multi-impurity lines radiation measurement. A 280 nm wavelength range can be covered by a 300 mm focal length spectrometer equipped with a 300 grooves/mm grating. The Dα/Hα line shapes can be resolved by the OMA2 system. The focal length is 750 mm. The spectral resolution can be up to 0.01 nm using a 1800 grooves/mm grating. The impurity behaviour and hydrogen ratio evolution after boroniztion, lithium coating, and siliconization are compared. Lithium coating has shown beneficial effects on the reduction of edge recycling and low Z impurity (C, O) influx. The impurity expelling effect of the divertor configuration is also briefly discussed through multi-channels observation of OMA1 system. (magnetically confined plasma)

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

Neoclassical poloidal and toroidal rotation in tokamaks

International Nuclear Information System (INIS)

Kim, Y.B.; Diamond, P.H.; Groebner, R.J.

1991-01-01

Explicit expressions for the neoclassical poloidal and toroidal rotation speeds of primary ion and impurity species are derived via the Hirshman and Sigmar moment approach. The rotation speeds of the primary ion can be significantly different from those of impurities in various interesting cases. The rapid increase of impurity poloidal rotation in the edge region of H-mode discharges in tokamaks can be explained by a rapid steepening of the primary ion pressure gradient. Depending on ion collisionality, the poloidal rotation speed of the primary ions at the edge can be quite small and the flow direction may be opposite to that of the impurities. This may cast considerable doubts on current L to H bifurcation models based on primary ion poloidal rotation only. Also, the difference between the toroidal rotation velocities of primary ions and impurities is not negligible in various cases. In Ohmic plasmas, the parallel electric field induces a large impurity toroidal rotation close to the magnetic axis, which seems to agree with experimental observations. In the ion banana and plateau regime, there can be non-negligible disparities between primary ion and impurity toroidal rotation velocities due to the ion density and temperature gradients. Detailed analytic expressions for the primary ion and impurity rotation speeds are presented, and the methodology for generalization to the case of several impurity species is also presented for future numerical evaluation

Characteristics of edge-localized modes in the experimental advanced superconducting tokamak (EAST)

DEFF Research Database (Denmark)

Jiang, M.; Xu, G.S.; Xiao, C.

2012-01-01

Edge-localized modes (ELMs) are the focus of tokamak edge physics studies because the large heat loads associated with ELMs have great impact on the divertor design of future reactor-grade tokamaks such as ITER. In the experimental advanced superconducting tokamak (EAST), the first ELMy high...... confinement modes (H-modes) were obtained with 1 MW lower hybrid wave power in conjunction with wall conditioning by lithium (Li) evaporation and real-time Li powder injection. The ELMs in EAST at this heating power are mostly type-III ELMs. They were observed close to the H-mode threshold power and produced...

Plasma jet source parameter optimisation and experiments on injection into Globus-M spherical tokamak

International Nuclear Information System (INIS)

Gusev, V.K.; Petrov, Yu.V.; Sakharov, N.V.; Semenov, A.A.; Voronin, A.V.

2005-01-01

Results of theoretical and experimental research on the plasma sources and injection of plasma and gas jet produced by the modified source into tokamak Globus-M are presented. An experimental test stand was developed for investigation of intense plasma jet generation. Optimisation of pulsed coaxial accelerator parameters by means of analytical calculations is performed with the aim of achieving the highest flow velocity at limited coaxial electrode length and discharge current. The optimal parameters of power supply to generate a plasma jet with minimal impurity contamination and maximum flow velocity were determined. A comparison of experimental and calculation results is made. Plasma jet parameters are measured, such as: impurity species content, pressure distribution across the jet, flow velocity, plasma density, etc. Experiments on the interaction of a higher kinetic energy plasma jet with the magnetic field and plasma of the Globus-M tokamak were performed. Experimental results on plasma and gas jet injection into different Globus-M discharge phases are presented and discussed. Results are presented on the investigation of plasma jet injection as the source for discharge breakdown, plasma current startup and initial density rise. (author)

Impurity injection into tokamak plasmas by erosion probes

International Nuclear Information System (INIS)

Hildebrandt, D.; Bakos, J.S.; Buerger, G.; Paszti, F.; Petravich, G.

1987-08-01

Exposing special erosion probes into the edge plasma of MT-1 the impurities Li and Ti were released and contaminated the plasma. By the use of collector probes the torodial transport of these impurities were investigated. The results indicate a preferential impurity flow into codirection of the plasma current. However, the asymmetric component of this flow is much larger than expected from the toroidal drift correlated to the plasma current. (author)

Impurity studies and discharge cleaning in Doublet III

Energy Technology Data Exchange (ETDEWEB)

Marcus, F.B.

1979-10-01

The goal of present and next generation tokamak experiments is to produce high-density, high-purity plasmas during high-power, extended-duration discharges. Plasma discharges with Z/sub eff/ values near unity and low concentrations of medium and high-Z metallic impurities have been obtained in Doublet III using a combination of low-power hydrogen discharge cleaning, gas puffing, precise plasma shape and position control, and high-Z limiters. Analysis of the first wall surface and residual gas impurities confirmed that clean conditions have been achieved. The high-Z limiters showed very limited amounts of melting or arcing. The progress of the wall cleaning process was monitored by three diagnostic techniques: Auger electron spectroscopy of metallic samples at the vessel wall, residual gas analysis, and the resistivity of full power discharges.

Impurity studies and discharge cleaning in Doublet III

International Nuclear Information System (INIS)

Marcus, F.B.

1979-10-01

The goal of present and next generation tokamak experiments is to produce high-density, high-purity plasmas during high-power, extended-duration discharges. Plasma discharges with Z/sub eff/ values near unity and low concentrations of medium and high-Z metallic impurities have been obtained in Doublet III using a combination of low-power hydrogen discharge cleaning, gas puffing, precise plasma shape and position control, and high-Z limiters. Analysis of the first wall surface and residual gas impurities confirmed that clean conditions have been achieved. The high-Z limiters showed very limited amounts of melting or arcing. The progress of the wall cleaning process was monitored by three diagnostic techniques: Auger electron spectroscopy of metallic samples at the vessel wall, residual gas analysis, and the resistivity of full power discharges

An empirical method for determination of elemental components of radiated powers and impurity concentrations from VUV and XUV spectral features in tokamak plasmas

International Nuclear Information System (INIS)

Lawson, K.; Peacock, N.; Gianella, R.

1998-12-01

The derivation of elemental components of radiated powers and impurity concentrations in bulk tokamak plasmas is complex, often requiring a full description of the impurity transport. A novel, empirical method, the Line Intensity Normalization Technique (LINT) has been developed on the JET (Joint European Torus) tokamak to provide routine information about the impurity content of the plasma and elemental components of radiated power (P rad ). The technique employs a few VUV and XUV resonance line intensities to represent the intrinsic impurity elements in the plasma. From a data base comprising these spectral features, the total bolometric measurement of the radiated power and the Z eff measured by visible spectroscopy, separate elemental components of P rad and Z eff are derived. The method, which converts local spectroscopic signals into global plasma parameters, has the advantage of simplicity, allowing large numbers of pulses to be processed, and, in many operational modes of JET, is found to be both reliable and accurate. It relies on normalizing the line intensities to the absolute calibration of the bolometers and visible spectrometers, using coefficients independent of density and temperature. Accuracies of the order of ± 15% can be achieved for the elemental P rad components of the most significant impurities and the impurity concentrations can be determined to within ±30%. Trace elements can be monitored, although with reduced accuracy. The present paper deals with limiter discharges, which have been the main application to date. As a check on the technique and to demonstrate the value of the LINT results, they have been applied to the transport modelling of intrinsic impurities carried out with the SANCO transport code, which uses atomic data from ADAS. The simulations provide independent confirmation of the concentrations empirically derived using the LINT technique. For this analysis, the simple case of the L-mode regime is considered, the chosen

A cryogenic system for TIBER II [Tokamak Ignition/Burn Experimental Reactor

International Nuclear Information System (INIS)

Slack, D.S.; Kerns, J.A.

1987-01-01

Phase II of the Tokamak Ignition/Burn Experimental Reactor (TIBER II) study describes one option for a small, economical, next-generation tokamak [1,2]. Because of its small size, minimum shielding is used between the plasma and the toroidal-field (TF) coils. Consequently, a large cryogenic system (approximately 70 kW at 4.5 K) capable of delivering forced-flow helium is required. This paper describes a cryogenic system that meets this requirement and includes TIBER-II requirements. 3 refs

Shielding and maintainability in an experimental tokamak

International Nuclear Information System (INIS)

Abdou, M.A.; Fuller, G.; Hager, E.R.; Vogelsang, W.F.

1979-01-01

This paper presents the results of an attempt to develop an understanding of the various factors involved. This work was performed as a part of the task assigned to one of the expert groups on the International Tokamak Reactor (INTOR). However, the results of this investigation are believed to be generally applicable to the broad class of the next generation of experimental tokamak facilities such as ETF. The shielding penalties for requiring personnel access are quantified. This is followed by a quantitative estimate of the benefits associated with personnel access. The penalties are compared to the benefits and conclusions and recommendations are developed on resolving the issue

Beam heating requirements for a tokamak experimental power reactor

International Nuclear Information System (INIS)

Bertoncini, P.J.; Brooks, J.N.; Fasolo, J.A.; Stacey, W.M. Jr.

1976-01-01

Typical beam heating requirements for effective tokamak experimental power reactor (TEPR) operation have been studied in connection with the Argonne preliminary conceptual TEPR design. For an ignition level plasma (approximately 100 MWt fusion power) for the nominal case envisioned, the neutral beam is only used to heat the plasma to ignition. This typically requires a beam power output of 40 MW at 180 keV for about 3 sec with a total energy of 114 MJ supplied to the plasma. The beam requirements for an ignition device are not very sensitive to changes in wall-sputtered impurity levels or plasma resistivity. For a plasma that must be driven due to poor confinement, the beam must remain on for most of the burn cycle. For representative cases, beam powers of approximately 23 MW are required for a total on-time of 20 to 50 sec. Reqirements on power level, beam energy, on-time, and beam-generation efficiency all represent considerable advances over present technology. For the Argonne TEPR design, a total of 16 to 32 beam injectors is envisioned. For a 40-MW, 180-keV, one-component beam, each injector supplies about 7 to 14 A of neutrals to the plasma. For positive ion sources, about 50 to 100 A of ions are required per injector and some form of particle and/or energy recycling appears to be essential in order to meet the power and efficiency requirements

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)

Orbit effects on impurity transport in a rotating plasma

International Nuclear Information System (INIS)

Wong, K.L.; Cheng, C.Z.

1988-01-01

In 1985, very high ion temperature plasmas were first produced in TFTR with co-injecting neutral beams in low current, low density plasmas. This mode of operation is called the energetic ion mode in which the plasma rotates at very high speed. It was found that heavy impurities injected into these plasmas diffused out very quickly. In this paper, the authors calculate the impurity ion orbits in a rotating tokamak plasma based on the equation of motion in the frame that rotates with the plasma. It is shown that heavy particles in a rotating plasma can drift away from magnetic surfaces significantly faster. Particle orbits near the surface of a rotating tokamak are also analyzed. During impurity injection experiments, freshly ionized impurities near the plasma surface are essentially stationary in the laboratory frame and they are counter-rotating in the plasma frame with co-beam injection. The results are substantiated by numeral particle simulation. The computer code follows the impurity guiding center positions by integrating the equation of motion with the second order predictor-corrector method

Impurity production and transport at limiters

International Nuclear Information System (INIS)

Matthews, G.F.

1989-01-01

This paper concentrates on the description and evaluation of experiments on the DITE tokamak. These are designed to characterise the processes involved in the production and transport of neutral and ionised impurities near carbon limiters. The need for good diagnostics in the scrape-off layer is highlighted. Langmuir probes are used to provide input data for models of impurity production at limiters. Observations of the radial profiles of carbon and oxygen impurities are compared with the code predictions. Changeover experiments involving hydrogen and helium plasmas are used as a means for investigating the role of the atomic physics and chemistry. The impurity control limiter (ICL) experiment is described which shows how geometry plays an important role in determining the spatial distributions of the neutral and ionised carbon. New diagnostics are required to study the flux and charge state distribution of impurities in the boundary. Preliminary results from an in-situ plasma ion mass-spectrometer are presented. The role of oxygen and the importance of evaluating the wall sources of impurity are emphasised. (orig.)

New directions in tokamak reactors

International Nuclear Information System (INIS)

Baker, C.C.

1985-01-01

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

Modeling of impurity transport in the core plasma

International Nuclear Information System (INIS)

Hulse, R.A.

1992-01-01

This paper presents a brief overview of computer modeling of impurity transport in the core region of controlled thermonuclear fusion plasmas. The atomic processes of importance in these high temperature plasmas and the numerical formulation of the model are described. Selected modeling examples are then used to highlight some features of the physics of impurity behavior in large tokamak fusion devices, with an emphasis on demonstrating the sensitivity of such modeling to uncertainties in the rate coefficients used for the atomic processes. This leads to a discussion of current requirements and opportunities for generating the improved sets of comprehensive atomic data needed to support present and future fusion impurity modeling studies

Impurity and trace tritium transport in tokamak edge turbulence

DEFF Research Database (Denmark)

Naulin, V.

2005-01-01

The turbulent transport of impurity or minority species, as for example tritium, is investigated in drift-Alfven edge turbulence. The full effects of perpendicular and parallel convection are kept for the impurity species. The impurity density develops a granular structure with steep gradients...... and locally exceeds its initial values due to the compressibility of the flow. An approximate decomposition of the impurity flux into a diffusive part and an effective convective part (characterized by a pinch velocity) is performed and a net inward pinch effect is recovered. The pinch velocity is explained...

System studies for quasi-steady-state advanced physics tokamak

International Nuclear Information System (INIS)

Reid, R.L.; Peng, Y.K.M.

1983-11-01

Parametric studies were conducted using the Fusion Engineering Design Center (FEDC) Tokamak Systems Code to investigate the impact of veriation in physics parameters and technology limits on the performance and cost of a low q/sub psi/, high beta, quasi-steady-state tokamak for the purpose of fusion engineering experimentation. The features and characteristics chosen from each study were embodied into a single Advanced Physics Tokamak design for which a self-consistent set of parameters was generated and a value of capital cost was estimated

Demonstration tokamak-power-plant study (DEMO)

International Nuclear Information System (INIS)

1982-09-01

A study of a Demonstration Tokamak Power Plant (DEMO) has been completed. The study's objective was to develop a conceptual design of a prototype reactor which would precede commercial units. Emphasis has been placed on defining and analyzing key design issues and R and D needs in five areas: noninductive current drivers, impurity control systems, tritium breeding blankets, radiation shielding, and reactor configuration and maintenance features. The noninductive current drive analysis surveyed a wide range of candidates and selected relativistic electron beams for the reference reactor. The impurity control analysis considered both a single-null poloidal divertor and a pumped limiter. A pumped limiter located at the outer midplane was selected for the reference design because of greater engineering simplicity. The blanket design activity focused on two concepts: a Li 2 O solid breeder with high pressure water cooling and a lead-rich Li-Pb eutectic liquid metal breeder (17Li-83Pb). The reference blanket concept is the Li 2 O option with a PCA structural material. The first wall concept is a beryllium-clad corrugated panel design. The radiation shielding effort concentrated on reducing the cost of bulk and penetration shielding; the relatively low-cost outborad shield is composed of concrete, B 4 C, lead, and FE 1422 structural material

Mathematical modeling plasma transport in tokamaks

Energy Technology Data Exchange (ETDEWEB)

Quiang, Ji [Univ. of Illinois, Urbana-Champaign, IL (United States)

1997-01-01

In this work, the author applied a systematic calibration, validation and application procedure based on the methodology of mathematical modeling to international thermonuclear experimental reactor (ITER) ignition studies. The multi-mode plasma transport model used here includes a linear combination of drift wave branch and ballooning branch instabilities with two a priori uncertain constants to account for anomalous plasma transport in tokamaks. A Bayesian parameter estimation method is used including experimental calibration error/model offsets and error bar rescaling factors to determine the two uncertain constants in the transport model with quantitative confidence level estimates for the calibrated parameters, which gives two saturation levels of instabilities. This method is first tested using a gyroBohm multi-mode transport model with a pair of DIII-D discharge experimental data, and then applied to calibrating a nominal multi-mode transport model against a broad database using twelve discharges from seven different tokamaks. The calibrated transport model is then validated on five discharges from JT-60 with no adjustable constants. The results are in a good agreement with experimental data. Finally, the resulting class of multi-mode tokamak plasma transport models is applied to the transport analysis of the ignition probability in a next generation machine, ITER. A reference simulation of basic ITER engineering design activity (EDA) parameters shows that a self-sustained thermonuclear burn with 1.5 GW output power can be achieved provided that impurity control makes radiative losses sufficiently small at an average plasma density of 1.2 X 10²⁰/m³ with 50 MW auxiliary heating. The ignition probability of ITER for the EDA parameters, can be formally as high as 99.9% in the present context. The same probability for concept design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%.

Mathematical modeling plasma transport in tokamaks

International Nuclear Information System (INIS)

Quiang, Ji

1995-01-01

In this work, the author applied a systematic calibration, validation and application procedure based on the methodology of mathematical modeling to international thermonuclear experimental reactor (ITER) ignition studies. The multi-mode plasma transport model used here includes a linear combination of drift wave branch and ballooning branch instabilities with two a priori uncertain constants to account for anomalous plasma transport in tokamaks. A Bayesian parameter estimation method is used including experimental calibration error/model offsets and error bar rescaling factors to determine the two uncertain constants in the transport model with quantitative confidence level estimates for the calibrated parameters, which gives two saturation levels of instabilities. This method is first tested using a gyroBohm multi-mode transport model with a pair of DIII-D discharge experimental data, and then applied to calibrating a nominal multi-mode transport model against a broad database using twelve discharges from seven different tokamaks. The calibrated transport model is then validated on five discharges from JT-60 with no adjustable constants. The results are in a good agreement with experimental data. Finally, the resulting class of multi-mode tokamak plasma transport models is applied to the transport analysis of the ignition probability in a next generation machine, ITER. A reference simulation of basic ITER engineering design activity (EDA) parameters shows that a self-sustained thermonuclear burn with 1.5 GW output power can be achieved provided that impurity control makes radiative losses sufficiently small at an average plasma density of 1.2 X 10 20 /m 3 with 50 MW auxiliary heating. The ignition probability of ITER for the EDA parameters, can be formally as high as 99.9% in the present context. The same probability for concept design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%

Experimental and theoretical basis for advanced tokamaks

International Nuclear Information System (INIS)

Chan, V.S.

1994-09-01

In this paper, arguments will be presented to support the attractiveness of advanced tokamaks as fusion reactors. The premise that all improved confinement regimes obtained to date were limited by magnetohydrodynamic stability will be established from experimental results. Accessing the advanced tokamak regime, therefore, requires means to overcome and enhance the beta limit. We will describe a number of ideas involving control of the plasma internal profiles, e.g. to achieve this. These approaches will have to be compatible with the underlying mechanisms for confinement improvement, such as shear rotation suppression of turbulence. For steady-state, there is a trade-off between full bootstrap current operation and the ability to control current profiles. The coupling between current drive and stability dictates the choice of sources and suggests an optimum for the bootstrap fraction. We summarize by presenting the future plans of the US confinement devices, DIII-D, PBX-M, C-Mod, to address the advanced tokamak physics issues and provide a database for the design of next-generation experiments

Experimental study of the topological aspect of the ergodic divertor in Tore-supra tokamak; Etude experimentale des aspects topologiques du divertor ergodique de Tore Supra

Energy Technology Data Exchange (ETDEWEB)

Costanzo, L

2001-10-01

The control of power deposition onto plasma facing components in tokamaks is a determining factor for future thermonuclear fusion reactors. Plasma surface interaction can be performed using limiters or divertors. The ergodic divertor installed on Tore Supra is an atypical example of a magnetic divertor. It consists in applying a magnetic perturbation which establishes a particular topology of the plasma in contact with the wall (edge plasma). We carried out dedicated experiments in order to study parallel heat flux which strike the divertor neutralizers. This quantitative and qualitative analysis of heat flux as a function of experimental conditions allows to determine the profiles of power deposition along the neutralizers. The influence of plasma electron density, additional heating, impurities and injected gas was established. An experimental study of the sheath heat transmission factor {gamma} was carried out by correlating measurements made with Langmuir probes and infrared imaging. This study gave rise to a major conclusion: for ohmic discharges with deuterium injection and most of the time with helium, it was experimentally confirmed that {gamma}=7 in agreement with classical sheath theory. However, an increase of this factor with additional power has been shown. Detached plasma, which is an attractive regime in order to reduce the power deposition, requires an optimized control. A new measurement of the detachment onset has been developed. It is based on the variation of heat flux onto the plates derived from infrared measurements. A detachment cartography with the determination of a new 2D 'IR' Degree of Detachment was carried out allowing to locate the zone where the detachment starts. We can apply this concept both to other tokamaks such as JET and ITER. A comparison between the axisymmetric divertor and the ergodic divertor is also presented concerning the power deposition in the two configurations. Low heat flux with the ergodic divertor is a

Conceptual studies of toroidal field magnets for the tokamak experimental power reactor. Final report

International Nuclear Information System (INIS)

Buncher, B.R.; Chi, J.W.H.; Fernandez, R.

1976-01-01

This report documents the principal results of a Conceptual Design Study for the Superconducting Toroidal Field System for a Tokamak Experimental Power Reactor. Two concepts are described for peak operating fields at the windings of 8 tesla, and 12 tesla, respectively. The design and manufacturing considerations are treated in sufficient detail that cost and schedule estimates could be developed. Major uncertainties in the design are identified and their potential impact discussed, along with recommendations for the necessary research and development programs to minimize these uncertainties. The minimum dimensions of a sub-size test coil for experimental qualification of the full size design are developed and a test program is recommended

Tokamak physics studies using x-ray diagnostic methods

International Nuclear Information System (INIS)

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

1987-03-01

X-ray diagnostic measurements have been used in a number of experiments to improve our understanding of important tokamak physics issues. The impurity content in TFTR plasmas, its sources and control have been clarified through soft x-ray pulse-height analysis (PHA) measurements. The dependence of intrinsic impurity concentrations and Z/sub eff/ on electron density, plasma current, limiter material and conditioning, and neutral-beam power have shown that the limiter is an important source of metal impurities. Neoclassical-like impurity peaking following hydrogen pellet injection into Alcator C and a strong effect of impurities on sawtooth behavior were demonstrated by x-ray imaging (XIS) measurements. Rapid inward motion of impurities and continuation of m = 1 activity following an internal disruption were demonstrated with XIS measurements on PLT using injected aluminum to enhance the signals. Ion temperatures up to 12 keV and a toroidal plasma rotation velocity up to 6 x 10 5 m/s have been measured by an x-ray crystal spectrometer (XCS) with up to 13 MW of 85-keV neutral-beam injection in TFTR. Precise wavelengths and relative intensities of x-ray lines in several helium-like ions and neon-like ions of silver have been measured in TFTR and PLT by the XCS. The data help to identify the important excitation processes predicted in atomic physics. Wavelengths of n = 3 to 2 silver lines of interest for x-ray lasers were measured, and precise instrument calibration techniques were developed. Electron thermal conductivity and sawtooth dynamics have been studied through XIS measurements on TFTR of heat-pulse propagation and compound sawteeth. A non-Maxwellian electron distribution function has been measured, and evidence of the Parail-Pogutse instability identified by hard x-ray PHA measurements on PLT during lower-hybrid current-drive experiments

Experimental observations related to the thermodynamic properties of tokamak plasmas

International Nuclear Information System (INIS)

Sozzi, C.; Minardi, E.; Lazzaro, E.; Cirant, S.; Mantica, P.; Esposito, B.; Marinucci, M.; Romanelli, M.; Imbeaux, F.

2005-01-01

The coarse-grained tokamak plasma description derived from the magnetic entropy concept presents appealing features as it involves a simple mathematics and it identifies a limited set of characteristic parameters of the macroscopic equilibrium. In this paper a comprehensive review of the work done in order to check the reliability of the Stationary Magnetic Entropy predictions against experimental data collected from different tokamaks, plasma regimes and heating methods is reported. (author)

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.

Experimental and gyrokinetic investigation of core impurity transport in Alcator C-mod

Science.gov (United States)

Howard, N.; Greenwald, M.; Podpaly, Y.; Reinke, M. L.; Rice, J. E.; White, A. E.; Mikkelsen, D. R.; Puetterich, T.

2010-11-01

A new multiple pulse laser blow-off system coupled with an upgraded high resolution x-ray spectrometer with spatial resolution allow for the most detailed studies of impurity transport on Alcator C-mod to date. Trace impurity injections created by the laser blow-off technique were introduced into plasmas with a wide range of parameters and time evolving profiles of He-like calcium were measured. The unique measurement of a single charge state profile and line integrated emission measurements from spectroscopic diagnostics were compared with the simulated emission from the impurity transport code STRAHL. A nonlinear least squares fitting routine was coupled with STRAHL, allowing for core impurity transport coefficients with errors to be determined. With this method, experimental data from trace calcium injections were analyzed and radially dependent, core values (< r/a ˜.6) of the diffusive and convective components of the impurity flux were obtained. The STRAHL results are compared with linear and global, nonlinear simulations from the gyrokinetic code GYRO. Results of this comparison and an investigation of the underlying physics associated with turbulent impurity transport will be presented.

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

International Nuclear Information System (INIS)

Henning, C.D.; Logan, B.G.; Barr, W.L.

1985-01-01

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

Deposition probe measurements of impurities injected into a tokamak plasma

International Nuclear Information System (INIS)

Hildebrandt, D.; Grote, H.; Herrmann, A.

1985-01-01

Impurity confinement behaviour has been studied by using a deposition probe in conjunction with pellet injection. Generally, an exponential decay of the impurity efflux and nearly symmetric ion/electron side toroidal flows have been observed. During phases of strong plasma disturbances, asymmetric flow is seen, indicative of edge transport and prompt recycling from local sources. The application of ECRH may cause such disturbances. (author)

Isotope effects of trapped electron modes in the presence of impurities in tokamak plasmas

Science.gov (United States)

Shen, Yong; Dong, J. Q.; Sun, A. P.; Qu, H. P.; Lu, G. M.; He, Z. X.; He, H. D.; Wang, L. F.

2016-04-01

The trapped electron modes (TEMs) are numerically investigated in toroidal magnetized hydrogen, deuterium and tritium plasmas, taking into account the effects of impurity ions such as carbon, oxygen, helium, tungsten and others with positive and negative density gradients with the rigorous integral eigenmode equation. The effects of impurity ions on TEMs are investigated in detail. It is shown that impurity ions have substantially-destabilizing (stabilizing) effects on TEMs in isotope plasmas for {{L}ez}\\equiv {{L}ne}/{{L}nz}>0 (TEM turbulences in hydrogenic isotope plasmas with and without impurities are performed. The relations between the maximum growth rate of the TEMs with respect to the poloidal wave number and the ion mass number are given in the presence of the impurity ions. The results demonstrate that the maximum growth rates scale as {γ\\max}\\propto Mi-0.5 in pure hydrogenic plasmas. The scale depends on the sign of its density gradient and charge number when there is a second species of (impurity) ions. When impurity ions have density profiles peaking inwardly (i.e. {{L}ez}\\equiv {{L}ne}/{{L}nz}>0 ), the scaling also depends on ITG parameter {ηi} . The maximum growth rates scale as {γ\\max}\\propto M\\text{eff}-0.5 for the case without ITG ({ηi}=0 ) or the ITG parameter is positive ({ηi}>0 ) but the impurity ion charge number is low (Z≤slant 5.0 ). However, when {ηi}>0 and the impurity ion charge number is moderate (Z=6.0-8.0 ), the scaling law is found as {γ\\max}\\propto M\\text{eff}-1.0 . Here, Z is impurity ion charge number, and the effective mass number, {{M}\\text{eff}}=≤ft(1-{{f}z}\\right){{M}i}+{{f}z}{{M}z} , with {{M}i} and {{M}Z} being the mass numbers of the hydrogenic and impurity ions, respectively, and {{f}z}=Z{{n}0z}/{{n}0e} being the charge concentration of impurity ions. In addition, with regard to the case of {{L}ez}<0 , the maximum growth rate scaling is {γ\\max}\\propto Mi-0.5 . The possible relations of the results

Study on the impurity transport in the experiments on macroparticle injection into the FT-1 Tokamak

International Nuclear Information System (INIS)

Zhilinskij, A.P.; Kuteev, B.V.; Larionov, M.M.; Lebedev, A.D.; Mikhajkin, S.S.; Nikiforov, V.A.; Rozhanskij, V.A.; Tsendin, L.D.

1980-01-01

Studied is decomposition of the impurity disturbance (Li, C, O, Si, Ti, stainless steel, Cu, Mo, W) injected by a macroparticle into the central plasma region. It is shown that the impurity are carried out from the discharge at a tau characteristic time weakly depending on their sort. The tau values are in good agreement with the theoretical predictions accounting for the poloidal plasma rotation

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-07-01

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

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

Tokamak experimental power reactor studies

International Nuclear Information System (INIS)

Stacey, W.M. Jr.; Baker, C.C.; Roberts, M.

1976-01-01

Three designs of an EPR are briefly described. The plasma physics involved in the EPR design is discussed. Brief discussions on the following topics are included: (1) MHD equilibrium, (2) impurity control, (3) transport scaling, (4) burn cycle dynamics, (5) first wall system, (6) blanket/shield systems, (7) plasma heating, (8) toroidal field coils, (9) poloidal field coils, (10) energy storage and transfer, (11) tritium fuel cycle, (12) radiation damage, and (13) costs

Abstracts of the International seminar 'Experimental possibilities of KTM tokamak and research programme'

International Nuclear Information System (INIS)

2005-01-01

The International seminar 'Experimental possibilities of KTM tokamak and research programme' was held in 10-12 October 2005 in Astana city (Kazakhstan). The seminar was dedicated to problems of KTM tokamak commissioning. The Collection of abstracts comprises 45 papers

The measurement of the intrinsic impurities of molybdenum and carbon in the Alcator C-Mod tokamak plasma using low resolution spectroscopy

Science.gov (United States)

May, M. J.; Finkenthal, M.; Regan, S. P.; Moos, H. W.; Terry, J. L.; Goetz, J. A.; Graf, M. A.; Rice, J. E.; Marmar, E. S.; Fournier, K. B.; Goldstein, W. H.

1997-06-01

The intrinsic impurity content of molybdenum and carbon was measured in the Alcator C-Mod tokamak using low resolution, multilayer mirror (MLM) spectroscopy ( Delta lambda ~1-10 AA). Molybdenum was the dominant high-Z impurity and originated from the molybdenum armour tiles covering all of the plasma facing surfaces (including the inner column, the poloidal divertor plates and the ion cyclotron resonant frequency (ICRF) limiter) at Alcator C-Mod. Despite the all metal first wall, a carbon concentration of 1 to 2% existed in the plasma and was the major low-Z impurity in Alcator C-Mod. Thus, the behaviour of intrinsic molybdenum and carbon penetrating into the main plasma and the effect on the plasma must be measured and characterized during various modes of Alcator C-Mod operation. To this end, soft X-ray extreme ultraviolet (XUV) emission lines of charge states, ranging from hydrogen-like to helium-like lines of carbon (radius/minor radius, r/a~1) at the plasma edge to potassium to chlorine-like (0.4Data Nucl. Data Tables 33 (1985) 149), which were incorporated into the collisional radiative model. The intrinsic i

Plasma-wall impurity experiments in ISX-A

International Nuclear Information System (INIS)

Colchin, R.J.; Bush, C.E.; Edmonds, P.H.; England, A.; Hill, K.W.; Isler, R.C.; Jernigan, T.C.; King, P.W.; Langley, R.A.; McNeill, D.H.; Murakami, M.; Neidigh, R.V.; Neilson, C.H.; Simpkins, J.E.; Wilgen, J.; DeBoo, J.C.; Burrell, K.H.; Ensberg, E.S.

1978-01-01

The ISX-A was a tokamak designed for studying plasma-wall interactions and plasma impurities. It fulfilled this role quite well, producing reliable and reproducible plasmas which had currents up to 175 kA and energy containment times up to 30 ms. With discharge precleaning, Zsub(eff) was as low as 1.6; with titanium evaporation. Zsub(eff) approached 1.0. Values of Zsub(eff) > approximately 2.0 were found to be proportional to residual impurity gases in the vacuum system immidiately following a discharge. However, there was no clear dependence of Zsub(eff) on base pressure. Stainless steel limiters were used in most of the ISX-A experiments. Upon introducing carbon limiters into the vacuum system, Zsub(eff) increased to 5.6. After twelve days of clean-up with tokamak discharges, during which time Zsub(eff) steadily decreased, the carbon limiters tended to give slightly higher values of Zsub(eff) than stainless steel limiters. Injection of 16 atoms of tungsten into discharges caused the power incident on the wall to double and the electron temperature profile to become hollow. (Auth.)

Theoretical and experimental studies of runaway electrons in the TEXTOR tokamak

International Nuclear Information System (INIS)

Abdullaev, S.S.; Finken, K.H.; Wongrach, K.; Willi, O.

2016-01-01

Theoretical and experimental studies of runaway electrons in tokamaks and their mitigations, particularly the recent studies performed by a group of the Heinrich-Heine University Duesseldorf in collaboration with the Institute of Energy and Climate Research of the Research Centre (Forschungszentrum) of Juelich are reviewed. The main topics focus on (i) runaway generation mechanisms, (ii) runaway orbits in equilibrium plasma, (iii) transport in stochastic magnetic fields, (iv) diagnostics and investigations of transport of runaway electron and their losses in low density discharges (v) runaway electrons during plasma disruptions, and (vi) runaway mitigation methods. The development of runaway diagnostics enables the measurement of runaway electrons in both the centre and edge of the plasma. The diagnostics provide an absolute runaway energy resolved measurement, the radial decay length of runaway electrons and, the structure and dynamics of runaway electron beams. The new mechanism of runaway electron formation during plasma disruptions is discussed.

Theoretical and experimental studies of runaway electrons in the TEXTOR tokamak

Energy Technology Data Exchange (ETDEWEB)

Abdullaev, S.S.; Finken, K.H.; Wongrach, K.; Willi, O.

2016-07-01

Theoretical and experimental studies of runaway electrons in tokamaks and their mitigations, particularly the recent studies performed by a group of the Heinrich-Heine University Duesseldorf in collaboration with the Institute of Energy and Climate Research of the Research Centre (Forschungszentrum) of Juelich are reviewed. The main topics focus on (i) runaway generation mechanisms, (ii) runaway orbits in equilibrium plasma, (iii) transport in stochastic magnetic fields, (iv) diagnostics and investigations of transport of runaway electron and their losses in low density discharges (v) runaway electrons during plasma disruptions, and (vi) runaway mitigation methods. The development of runaway diagnostics enables the measurement of runaway electrons in both the centre and edge of the plasma. The diagnostics provide an absolute runaway energy resolved measurement, the radial decay length of runaway electrons and, the structure and dynamics of runaway electron beams. The new mechanism of runaway electron formation during plasma disruptions is discussed.

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

Design study of toroidal magnets for tokamak experimental power reactors

International Nuclear Information System (INIS)

Stekly, Z.J.J.; Lucas, E.J.

1976-12-01

This report contains the results of a six-month study of superconducting toroidal field coils for a Tokamak Experimental Power Reactor to be built in the late 1980s. The designs are for 8 T and 12 T maximum magnetic field at the superconducting winding. At each field level two main concepts were generated; one in which each of the 16 coils comprising the system has an individual vacuum vessel and the other in which all the coils are contained in a single vacuum vessel. The coils have a D shape and have openings of 11.25 m x 7.5 m for the 8 T coils and 10.2 m x 6.8 m for the 12 T coils. All the designs utilize rectangular cabled conductor made from copper stabilized Niobium Titanium composite which operates at 4.2 K for the 8 T design and at 2.5 K for the 12 T design. Manufacturing procedures, processes and schedule estimates are also discussed

Two dimensional neutral transport analysis in tokamak plasma

International Nuclear Information System (INIS)

Shimizu, Katsuhiro; Azumi, Masafumi

1987-02-01

Neutral particle influences the particle and energy balance, and play an important role on sputtering impurity and the charge exchange loss of neutral beam injection. In order to study neutral particle behaviour including the effects of asymmetric source and divertor configuration, the two dimensional neutral transport code has been developed using the Monte-Carlo techniques. This code includes the calculation of the H α radiation intensity based on the collisional-radiation model. The particle confinement time of the joule heated plasma in JT-60 tokamak is evaluated by comparing the calculated H α radiation intensity with the experimental data. The effect of the equilibrium on the neutral density profile in high-β plasma is also investigated. (author)

Overview of recent experimental results from the DIII-D advanced tokamak program

International Nuclear Information System (INIS)

Allen, S.L.

2001-01-01

The goals of DIII-D Advanced Tokamak (AT) experiments are to investigate and optimize the upper limits of energy confinement and MHD stability in a tokamak plasma, and to simultaneously maximize the fraction of non-inductive current drive. Significant overall progress has been made in the past 2 years, as the performance figure of merit β N H 89P of 9 has been achieved in ELMing H-mode for over 16 τ E without sawteeth. We also operated at β N ∼7 for over 35 τ E or 3 τ R , with the duration limited by hardware. Real-time feedback control of β (at 95% of the stability boundary), optimizing the plasma shape (e.g., δ, divertor strike- and X-point, double/single null balance), and particle control (n e /n GW ∼0.3, Z eff N H 89P of 7. The QDB regime has been obtained to date only with counter neutral beam injection. Further modification and control of internal transport barriers (ITBs) has also been demonstrated with impurity injection (broader barrier), pellets, and ECH (strong electron barrier). The new Divertor-2000, a key ingredient in all these discharges, provides effective density, impurity and heat flux control in the high-triangularity plasma shapes. Discharges at n e /n GW ∼1.4 have been obtained with gas puffing by maintaining the edge pedestal pressure; this operation is easier with Divertor-2000. We are developing several other tools required for AT operation, including real-time feedback control of resistive wall modes (RWMs) with external coils, and control of neoclassical tearing modes (NTMs) with electron cyclotron current drive (ECCD). (author)

Integrated modeling of temperature profiles in L-mode tokamak discharges

Energy Technology Data Exchange (ETDEWEB)

Rafiq, T.; Kritz, A. H.; Tangri, V. [Department of Physics, Lehigh University, Bethlehem, Pennsylvania 18015 (United States); Pankin, A. Y. [Tech-X Corporation, Boulder, Colorado 80303 (United States); Voitsekhovitch, I. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Budny, R. V. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

2014-12-15

Simulations of doublet III-D, the joint European tokamak, and the tokamak fusion test reactor L-mode tokamak plasmas are carried out using the PTRANSP predictive integrated modeling code. The simulation and experimental temperature profiles are compared. The time evolved temperature profiles are computed utilizing the Multi-Mode anomalous transport model version 7.1 (MMM7.1) which includes transport associated with drift-resistive-inertial ballooning modes (the DRIBM model [T. Rafiq et al., Phys. Plasmas 17, 082511 (2010)]). The tokamak discharges considered involved a broad range of conditions including scans over gyroradius, ITER like current ramp-up, with and without neon impurity injection, collisionality, and low and high plasma current. The comparison of simulation and experimental temperature profiles for the discharges considered is shown for the radial range from the magnetic axis to the last closed flux surface. The regions where various modes in the Multi-Mode model contribute to transport are illustrated. In the simulations carried out using the MMM7.1 model it is found that: The drift-resistive-inertial ballooning modes contribute to the anomalous transport primarily near the edge of the plasma; transport associated with the ion temperature gradient and trapped electron modes contribute in the core region but decrease in the region of the plasma boundary; and neoclassical ion thermal transport contributes mainly near the center of the discharge.

Tokamak plasma interaction with limiters

International Nuclear Information System (INIS)

Pitcher, C.S.

1987-11-01

The importance of plasma purity is first discussed in terms of the general requirements of controlled thermonuclear fusion. The tokamak approach to fusion and its inherent problem of plasma contamination are introduced. A main source of impurities is due to the bombardment of the limiter by energetic particles and thus the three main aspects of the plasma-limiter interaction are reviewed, boundary plasma conditions, fuelling/recycling and impurity production. The experiments, carried out on the DITE tokamak at Culham Laboratory, UK, investigated these three topics and the results are compared with predicted behaviour; new physical phenomena are presented in all three areas. Simple one-dimensional fluid equations are found to adequately describe the SOL plasma, except in regard to the pre-sheath electric field and ambipolarity; that is, the electric field adjacent to the limiter surface appears to be weak and the associated plasma flow can be non-ambipolar. Recycling of fuel particles from the limiter is observed to be near unity at all times. The break-up behaviour of recycled and gas puffed D 2 molecules is dependent on the electron temperature, as expected. Impurity production at the limiter is chemical erosion of graphite being negligible. Deposition of limiter and wall-produced impurities is found on the limiter. The spatial distributions of impurities released from the limiter are observed and are in good agreement with a sputtered atom transport code. Finally, preliminary experiments on the transport of impurity ions along field lines away from the limiter have been performed and compared with simple analytic theory. The results suggest that the pre-sheath electric field in the SOL is much weaker than the simple fluid model would predict

Data processing system for spectroscopy at Novillo Tokamak

International Nuclear Information System (INIS)

Ortega C, G.; Gaytan G, E.

1998-01-01

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

Impurity investigations in the boundary layer of the DITE tokamak

International Nuclear Information System (INIS)

McCracken, G.M.; Partridge, J.W.; Erents, S.K.; Sofield, C.J.; Ferguson, S.M.

1982-01-01

The results obtained in the present investigation show large fluctuations both during discharges and from one discharge to the next. The radial density gradient of impurities in the boundary is not large. It is clear that the density and in particular dn/dt can have a strong effect on the impurity level. However there are apparently a number of other factors causing changes in impurity level which have not been well controlled in the present experiments. Possibilities include flaking from the walls, and changes in the level of the light impurities, oxygen and carbon, in the discharges. (orig./RW)

Runaway electrons during tokamak startup

International Nuclear Information System (INIS)

Sharma, A.S.; Jayakumar, R.

1988-01-01

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

[High beta tokamak research and plasma theory

International Nuclear Information System (INIS)

1990-01-01

Our activities on High Beta Tokamak Research during the past 12 months of the present budget period can be divided into four areas: completion of kink mode studies in HBT; completion of carbon impurity transport studies in HBT; design of HBT-EP; and construction of HBT-EP. Each of these is described briefly in the sections of this progress report

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)

Vertical poloidal asymmetries of low-Z element radiation in the PDX tokamak

Energy Technology Data Exchange (ETDEWEB)

Brau, K.; Suckewer, S.; Wong, S.K.

1983-06-01

Vertical poloidal asymmetries of hydrogen isotopes and low-Z impurity radiation in the PDX tokamak may be caused by poloidally asymmetric sources of these elements at gas inlet valves, limiters or vacuum vessel walls, asymmetric magnetic field geometry in the region beyond the plasma boundary, or by ion curvature drifts. Low ionization states of carbon (C II- C IV) are more easily influenced by edge conditions than is CV. Vertical poloidal asymmetries of CV are correlated with the direction of the toroidal field. The magnitude of the asymmetry agrees with the predictions of a quasifluid neoclassical model. Experimental data and numerical simulations are presented to investigate different models of impurity poloidal asymmetries.

Vertical poloidal asymmetries of low-Z element radiation in the PDX tokamak

International Nuclear Information System (INIS)

Brau, K.; Suckewer, S.; Wong, S.K.

1983-06-01

Vertical poloidal asymmetries of hydrogen isotopes and low-Z impurity radiation in the PDX tokamak may be caused by poloidally asymmetric sources of these elements at gas inlet valves, limiters or vacuum vessel walls, asymmetric magnetic field geometry in the region beyond the plasma boundary, or by ion curvature drifts. Low ionization states of carbon (C II- C IV) are more easily influenced by edge conditions than is CV. Vertical poloidal asymmetries of CV are correlated with the direction of the toroidal field. The magnitude of the asymmetry agrees with the predictions of a quasifluid neoclassical model. Experimental data and numerical simulations are presented to investigate different models of impurity poloidal asymmetries

Overview of the Tokamak de Varennes program

International Nuclear Information System (INIS)

Pacher, H.D.

1986-01-01

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

Investigation by the Rutherford backscattering method of impurity deposited on the T-3M tokamak diaphragm

International Nuclear Information System (INIS)

Danelyan, L.S.; Egorova, I.M.; Kulikauskas, V.S.; Baratov, D.G.; Belykh, T.A.

1994-01-01

The Rutherford backscattering of helium-4 ions was used for investigation of impurity deposited on the annular graphite diaphragm as a result of the interaction between hydrogen plasma and liquid-metal spray limiter. The experimental RBS spectra distributions of the impurity elements surface densities along the direction from plasma to the chamber wall are presented as depth of the elements. The erosion coefficient of the main liquid-metal limiter element has been estimated

Estimation of Zeff in Novillo Tokamak

International Nuclear Information System (INIS)

Valencia, R.; Olayo, G.; Cruz, G.; Lopez, R.; Chavez, E.; Melendez, L.; Flores, A.; Gaytan, E.

1996-01-01

We estimated the Z eff in the Novillo Tokamak after having applied a HeGDC process through two different methods: anomaly factor and mass spectrometry. The first one gave a Z eff value of 2.07 for a tokamak discharge of 4350 A plasma current and 3 V of loop voltage. By mass spectrometry 30 s after the discharge had finished a Z eff of 4.19 was obtained for the same discharge. By mass spectrometry we observed that the Z eff value is a time function. Furthermore this method is helpful for evaluating the level of impurities after many discharges in Novillo Tokamak. (orig.)

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.)

Role of codeposited impurities during growth. I. Explaining distinctive experimental morphology on Cu(0 0 1)

Science.gov (United States)

Hamouda, Ajmi Bh.; Sathiyanarayanan, Rajesh; Pimpinelli, Alberto; Einstein, T. L.

2011-01-01

A unified explanation of the physics underlying all the distinctive features of the growth instabilities observed on Cu vicinals has long eluded theorists. Recently, kinetic Monte Carlo studies showed that codeposition of impurities during growth could account for the key distinctive experimental observations [Hamouda , Phys. Rev. BPLRBAQ0556-280510.1103/PhysRevB.77.245430 77, 245430 (2008)]. To identify the responsible impurity atom, we compute the nearest-neighbor binding energies (ENN) and terrace diffusion barriers (Ed) for several candidate impurity atoms on Cu(0 0 1) using DFT-based VASP. Our calculations show that codeposition (with Cu) of midtransition elements, such as Fe, Mn, and W, could—in conjunction with substantial Ehrlich-Schwoebel barriers—cause the observed instabilities; when the experimental setup is considered, W emerges to be the most likely candidate. We discuss the role of impurities in nanostructuring of surfaces.

Experimental Study of Thermal Crisis in Connection with Tokamak Reactor High Heat Flux Components

International Nuclear Information System (INIS)

Gallo, D.; Giardina, M.; Castiglia, F.; Celata, G.P.; Mariani, A.; Zummo, G.; Cumo, M.

2000-01-01

The results of an experimental research on high heat flux thermal crisis in forced convective subcooled water flow, under operative conditions of interest to the thermal-hydraulic design of TOKAMAK fusion reactors, are here reported. These experiments, carried out in the framework of a collaboration between the Nuclear Engineering Department of Palermo University and the National Institute of Thermal - Fluid Dynamics of the ENEA - Casaccia (Rome), were performed on the STAF (Scambio Termico Alti Flussi) water loop and consisted, essentially, in a high speed photographic study which enabled focusing several information on bubble characteristics and flow patterns taking place during the burnout phenomenology

Revised design for the Tokamak experimental power reactor

International Nuclear Information System (INIS)

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

1977-03-01

A new, preliminary design has been identified for the tokamak experimental power reactor (EPR). The revised EPR design is simpler, more compact, less expensive and has somewhat better performance characteristics than the previous design, yet retains many of the previously developed design concepts. This report summarizes the principle features of the new EPR design, including performance and cost

Oak Ridge Tokamak experimental power reactor study scoping report

International Nuclear Information System (INIS)

Roberts, M.

1977-03-01

This report presents the scoping studies performed as the initial part of the program to produce a conceptual design for a Tokamak Experimental Power Reactor (EPR). The EPR as considered in this study is to employ all systems necessary for significant electric power production at continuous high duty cycle operation; it is presently scheduled to be the final technological step before a Demonstration Reactor Plant (Demo). The scoping study tasks begin with an exploration and identification of principal problem areas and then concentrate on consideration and evaluation of alternate design choices for each of the following major systems: Plasma Engineering and Physics, Nuclear, Electromagnetics, Neutral Beam Injection, and Tritium Handling. In addition, consideration has been given to the integration of these systems and requirements arising out of their incorporation into an EPR. One intent of this study is to document the paths explored in search of the appropriate EPR characteristics. To satisfy this intent, the explorations are presented in chart form outlining possible options in key areas with extensive supporting footnotes. An important result of the scoping study has been the development and definition of an EPR reference design to serve as (1) a common focus for the continuing design study and (2) a guide for associated development programs. In addition, the study has identified research and development requirements essential to facilitate the successful conceptual design, construction, and operation of an EPR

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

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

Plasma shut-down with fast impurity puff on ASDEX Upgrade

International Nuclear Information System (INIS)

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

2007-01-01

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

Plasma-wall impurity experiments in ISX-A

International Nuclear Information System (INIS)

Colchin, R.J.; Bush, C.E.; Edmonds, P.H.

1978-08-01

The ISX-A was a tokamak designed for studying plasma-wall interactions and plasma impurities. It fulfilled this role quite well, producing reliable and reproducible plasmas which had currents up to 175 kA and energy containment times up to 30 msec. With discharge precleaning, Z/sub eff/ was as low as 1.6; with titanium evaporation, Z/sub eff/ approached 1.0. Values of Z/sub eff/ greater than or equal to 2.0 were found to be proportional to residual impurity gases in the vacuum system immediately following a discharge. However, there was no clear dependence of Z/sub eff/ on base pressure. Stainless steel limiters were used in most of the ISX-A experiments. When carbon limiters were introduced into the vacuum system, Z/sub eff/ increased to 5.6. After twelve days of cleanup with tokamak discharges, during which time Z/sub eff/ steadily decreased, the carbon limiters tended to give slightly higher values of Z/sub eff/ than stainless steel limiters. Injection of less than 10 16 atoms of tungsten into discharges caused the power incident on the wall to double and the electron temperature profile to become hollow

Coupling of ion temperature gradient and trapped electron modes in the presence of impurities in tokamak plasmas

Science.gov (United States)

Du, Huarong; Wang, Zheng-Xiong; Dong, J. Q.; Liu, S. F.

2014-05-01

The coupling of ion temperature gradient (ITG or ηi) mode and trapped electron mode (TEM) in the presence of impurity ions is numerically investigated in toroidal collisionless plasmas, using the gyrokinetic integral eigenmode equation. A framework for excitations of the ITG modes and TEMs with respect to their driving sources is formulated first, and then the roles of impurity ions played in are analyzed comprehensively. In particular, the characteristics of the ITG and TEM instabilities in the presence of impurity ions are emphasized for both strong and weak coupling (hybrid and coexistent) cases. It is found that the impurity ions with inwardly (outwardly) peaked density profiles have stabilizing (destabilizing) effects on the hybrid (namely the TE-ITG) modes in consistence with previous works. A new finding of this work is that the impurity ions have stabilizing effects on TEMs in small ηi (ηi≤1) regime regardless of peaking directions of their density profiles whereas the impurity ions with density gradient Lez=Lne/Lnz>1 (LezTEMs in large ηi (ηi≥1) regime. In addition, the dependences of the growth rate, real frequency, eigenmode structure, and wave spectrum on charge concentration, charge number, and mass of impurity ions are analyzed in detail. The necessity for taking impurity ion effects on the features of turbulence into account in future transport experimental data analyses is also discussed.

Recent Activities on the Experimental Research Programme Using Small Tokamaks

International Nuclear Information System (INIS)

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

2006-01-01

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 (CRP) is discussed in this paper. Besides the presentation of the recent activities on the experimental research programme using small tokamaks and scientific results achieved at the participating laboratories, information is provided about the organisation of the co-ordinated research project. Future plans of the co-ordinated activities within the CRP are discussed

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

Effect of neutral particles on density limits in tokamaks

International Nuclear Information System (INIS)

Abramov, V.A.; Morozov, D.Kh.; Bachmann, P.; Suender, D.

1993-01-01

The global stability and confinement of a tokamak plasma are significantly influenced by the boundary plasma parameters. The onset of density disruptions, which limit the maximum plasma density, is triggered by impurity radiation in the edge plasma and can be connected with the radiative thermal instability. At the density n c the total radiative power P rad is equal to the total input power P in into the plasma (S:=P rad /P in =1). Above n c (S>1) no steady state of the plasma column exists. Contrary to predictions made elsewhere, where neutral particle kinetics is not taken into consideration, experimental results show that disruptions can occur for S R as a function of the plasma temperature T, ξ N :=N/n and ξ i :=n i /n, where N, n i , n are the densities of hydrogen atoms, impurity ions and the plasma, respectively. We investigate the influence of the neutral particles on the critical densities and the stability of the system, taking into account ionization, charge exchange and impurity cooling. (author) 6 refs., 3 figs

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.)

Tokamak startup with electron cyclotron heating

International Nuclear Information System (INIS)

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

1980-04-01

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

Tokamak startup with electron cyclotron heating

Energy Technology Data Exchange (ETDEWEB)

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

1980-04-01

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

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)

Plasma boundary experiments on DIII-D tokamak

International Nuclear Information System (INIS)

Mahdavi, M.A.; Brooks, N.; Jackson, G.L.; Langhorn, A.; Leikind, B.; Lippmann, S.; Luxon, J.; Petersen, P.; Petrie, P.; Stambaugh, R.D.; Simonen, T.C.; Staebler, G.; Buchenauer, D.; Futch, A.; Hill, D.N.; Rensink, M.; Hogan, J.; Menon, M.; Mioduszewski, P.K.; Owen, L.; Matthews, G.

1990-01-01

A survey of the boundary physics research on the DIII-D tokamak and an outline of the DIII-D Advanced Divertor Program (ADP) is presented. We will present results of experiments on impurity control, impurity transport, neutral particle transport, and particle effects on core confinement over a wide range of plasma parameters, I p T < or approx.10.7%, P(auxiliary)< or approx.20 MW. Based on the understanding gained in these studies, we in collaboration with a number of other laboratories have devised a series of experiments (ADP) to modify the core plasma conditions through changes in the edge electric field, neutral recycling, and plasma-surface interactions. (orig.)

Test particle calculations for the Texas experimental tokamak with resonant magnetic fields

International Nuclear Information System (INIS)

Wootton, A.J.; McCool, S.C.; Zheng, S.

1991-01-01

This paper presents a simple test particle model that attempts to describe particle motion in the presence of intrinsic electrostatic fluctuations in a prescribed tokamak magnetic field. In particular, magnetic field configurations that include externally produced magnetic islands and stochastic regions are considered. The resulting test particle transport is compared with the predictions of analytic models and with the experimentally measured electron heat and particle transport on the Texas Experimental Tokamak (TEXT). Agreement between the test particle results and applicable analytic theories is found. However, there is only partial agreement with the experimental results, and possible reasons for the discrepancies are explored. Good agreement is found between predicted and measured spatially asymmetric particle distributions. The particle collection efficiency of an apertured limiter inside a magnetic island (an intra-island pump limiter) is discussed

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Plasma diagnostics on large tokamaks

International Nuclear Information System (INIS)

Orlinskij, D.V.; Magyar, G.

1988-01-01

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

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

Experimental and theoretical investigation of Fe-catalysis phenomenon in hydrogen thermal desorption form hydrocarbon plasma-discharge films from T-10 tokamak

International Nuclear Information System (INIS)

Stankevich, V.G.; Svechnikov, N.Y.; Lebedev, A.M.; Menshikov, K.A.; Kolbasov, B.N.; Sukhanov, L.P.

2017-01-01

A comprehensive study of hydrocarbon films obtained in the plasma discharge of large fusion facilities will allow the minimization of parasitic capture. The investigation of the effect of Fe impurities on D 2 thermal desorption (TD) from homogeneous CD x films (x ∼ 0.5) formed in the D-plasma discharge of the T-10 tokamak were carried out. The experimental TD spectra of the films showed 2 groups of peaks at 650-850 K and 900-1000 K for 2 adsorption states. The main result of the iron catalysis effect consists in the shift of the high-temperature peak by -24 K and in the increase in the fraction of the weakly bonded adsorption states. To describe the effect of iron impurities on TD of hydrogen isotopes, a structural cluster model based on the interaction of the Fe + ion with the 1,3-C 6 H 8 molecule was proposed. The potential energy surfaces of chemical reactions with the H 2 elimination were calculated using ab initio methods of quantum chemistry. It was established that the activation barrier of hydrogen TD is reduced by about 1 eV due to the interaction of the Fe + ion with the π-subsystem of the 1,3-C 6 H 8 molecule leading to a redistribution of the double bonds along the carbon system

Tokamaks - Third Edition

International Nuclear Information System (INIS)

Rogister, A L

2004-01-01

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

Tokamak instrumentation and controls

International Nuclear Information System (INIS)

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

1979-02-01

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

The study of heat flux for disruption on experimental advanced superconducting tokamak

Science.gov (United States)

Yang, Zhendong; Fang, Jianan; Gong, Xianzu; Gan, Kaifu; Luo, Jiarong; Zhao, Hailin; Cui, Zhixue; Zhang, Bin; Chen, Meiwen

2016-05-01

Disruption of the plasma is one of the most dangerous instabilities in tokamak. During the disruption, most of the plasma thermal energy is lost, which causes damages to the plasma facing components. Infrared (IR) camera is an effective tool to detect the temperature distribution on the first wall, and the energy deposited on the first wall can be calculated from the surface temperature profile measured by the IR camera. This paper concentrates on the characteristics of heat flux distribution onto the first wall under different disruptions, including the minor disruption and the vertical displacement events (VDE) disruption. Several minor disruptions have been observed before the major disruption under the high plasma density in experimental advanced superconducting tokamak. During the minor disruption, the heat fluxes are mainly deposited on the upper/lower divertors. The magnetic configuration prior to the minor disruption is a lower single null with the radial distance between the two separatrices in the outer midplane dRsep = -2 cm, while it changes to upper single null (dRsep = 1.4 cm) during the minor disruption. As for the VDE disruption, the spatial distribution of heat flux exhibits strong toroidal and radial nonuniformity, and the maximum heat flux received on the dome plate can be up to 11 MW/m2.

Heavy impurity collection at the plasma edge of the stellarator W VII A

International Nuclear Information System (INIS)

Schou, J.

1981-12-01

The presence of impurities at the plasma edge of the Wendelstein VII-A stellarator was studied by means of carbon probes that were exposed to up to 200 plasma discharges in helium. The probes were subsequently analysed with 1 MeV 4 He + Rutherford Backscattering. The average impurity deposition for Ti, Mo and wall components (Fe, Cr, Ni) was 2-4 x 10 12 atoms/cm 2 , 6 x 10 10 atoms/cm 2 and 1 x 10 11 atoms/cm 2 per discharge, respectively. With the exception of Ti this impurity deposition is more than one order of magnitude smaller than the corresponding results from comparable tokamaks. (orig.)

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

Energy Technology Data Exchange (ETDEWEB)

Casali, Livia

2015-11-24

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

Understanding of impurity poloidal distribution in the edge pedestal by modelling

Science.gov (United States)

Rozhansky, V.; Kaveeva, E.; Molchanov, P.; Veselova, I.; Voskoboynikov, S.; Coster, D.; Fable, E.; Puetterich, T.; Viezzer, E.; Kukushkin, A. S.; Kirk, A.; the ASDEX Upgrade Team

2015-07-01

Simulation of an H-mode ASDEX Upgrade shot with boron impurity was done with the B2SOLPS5.2 transport code. Simulation results were compared with the unique experimental data available for the chosen shot: radial density, electron and ion temperature profiles in the equatorial midplanes, radial electric field profile, radial profiles of the parallel velocity of impurities at the low-field side (LFS) and high-field side (HFS), radial density profiles of impurity ions at LHS and HFS. Simulation results reproduce all available experimental data simultaneously. In particular strong poloidal HFS-LFS asymmetry of B5+ ions was predicted in accordance with the experiment. The simulated HFS B5+ density inside the edge transport barrier is twice larger than that at LFS. This is consistent with the experimental observations where even larger impurity density asymmetry was observed. A similar effect was predicted in the simulation done for the MAST H-mode. Here the HFS density of He2+ is predicted to be 4 times larger than that at LHS. Such a large predicted asymmetry is connected with a larger ratio of HFS and LFS magnetic fields which is typical for spherical tokamaks. The HFS/LFS asymmetry was not measured in the experiment, however modelling qualitatively reproduces the observed change of sign of He+parallel velocity to the counter-current direction at LFS. The understanding of the asymmetry is based on neoclassical effects in plasma with strong gradients. It is demonstrated that simulation results obtained with account of sources of ionization, realistic geometry and turbulent transport are consistent with the simplified analytical approach. Difference from the standard neoclassical theory is emphasized.

Plasma rotation measurement in small tokamaks using an optical spectrometer and a single photomultiplier as detector.

Science.gov (United States)

Severo, J H F; Nascimento, I C; Kuznetov, Yu K; Tsypin, V S; Galvão, R M O; Tendler, M

2007-04-01

The method for plasma rotation measurement in the tokamak TCABR is reported in this article. During a discharge, an optical spectrometer is used to scan sequentially spectral lines of plasma impurities and spectral lines of a calibration lamp. Knowing the scanning velocity of the diffraction grating of the spectrometer with adequate precision, the Doppler shifts of impurity lines are determined. The photomultiplier output voltage signals are recorded with adequate sampling rate. With this method the residual poloidal and toroidal plasma rotation velocities were determined, assuming that they are the same as those of the impurity ions. The results show reasonable agreement with the neoclassical theory and with results from similar tokamaks.

Experimental study of the β-limit in ohmic H-mode in the TUMAN-3M tokamak

International Nuclear Information System (INIS)

Lebedev, S.V.; Andreiko, M.V.; Askinazi, L.G.; Golant, V.E.; Kornev, V.A.; Krikunov, S.V.; Levin, L.S.; Rozhdestvensky, V.V.; Tukachinsky, A.S.; Yaroshevich, S.P.

1998-01-01

Because of its high confinement properties, the H-mode provides good opportunities to achieve high beta values in a tokamak. In this paper the results of an experimental study of β T and β N limits in the H-mode, obtained in a circular cross section tokamak without auxiliary heating are presented. The experiments were performed in the TUMAN-3M tokamak. The device has the following parameters: R 0 =0.53m, a s =0.22m (limiter configuration), B T ≤1.2T, I p ≤175kA, n-bar e ≤6.2x10 19 m -3 . The stored energy was measured using diamagnetic loops and compared with W calculated from kinetic data obtained by Thomson scattering and microwave interferometry. Measurements of the stored energy and of the β were performed in the ohmic H-mode before and after boronization and in the scenario with fast current ramp-down in ohmic H-mode. A maximum value of β T of 2.0% and β N of 2.0 were achieved. The β N limit achieved reveals itself as a 'soft' (non-disruptive) limit. The stored energy slowly decays after the current ramp-down. No correlation was found between beta restriction and MHD phenomena. Internal transport barrier (ITB) formation was observed in ohmic H-mode. An enhancement factor of 2.0 over ITER93H(ELM-free) was found in the ohmic H-mode with ITB. (author)

Bragg rotor spectrometer for tokamak diagnostics

International Nuclear Information System (INIS)

Barnsley, R.; Evans, K.D.; Peacock, N.J.

1986-01-01

This paper discusses high thru'put broad band (1-24 angstrom) x-ray spectrometer having absolute calibration for wavelength and intensity and demonstrated on the DITE tokamak. This instrument has a self-contained vacuum system which allows full spatial scans of the DITE plasma. Data acquisition and drive mechanism for the rotor and filter selection are operated remotely from a SADA. Results are presented of fast spectral surveys and λ-lock time-evolution of impurity emission during neutral beam injection. Spatial scans of the absolute impurity concentrations are derived

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.)

Annual report of the Division of Thermonuclear Fusion Research and the Division of Large Tokamak Development for the period of April 1, 1977 to March 31, 1978

International Nuclear Information System (INIS)

1979-02-01

Research and development works in fiscal year 1977 of the Division of Thermonuclear Fusion Research and the Division of Large Tokamak Development are described. 1) Theoretical studies on tokamak confinement have continued with more emphasis on computations. A task was started of developing a computer code system for mhd behavior of tokamak plasmas. 2) Experimental studies of lower hybrid heating up to 140 kW were made in JFT-2. The ion temperature was increased by 50% -- 60% near the plasma center. Plasma-wall interactions (particle and thermal fluxes to the wall, and titanium gettering) were studied. In JFT-2a (DIVA) ion sputtering, arcing and evaporation were identified, and the impurity ion sputtering was found to be a dominant origin of metal impurities in the present tokamaks. High temperature and high-density plasma divertor actions were demonstrated; i.e. the divertor decreases the radiation power loss by a factor of 3 and increases the energy confinement time by a factor of 2.5. Various diagnostic instruments operated sufficiently to provide useful information for the research with JFT-2 and JFT-2a(DIVA). 3) JFT-2 and JFT-2a(DIVA) operated as scheduled. Technological improvements were made such as titanium coating of the chamber wall, discharge cleaning and pre-ionization. 4) Detailed design of the prototype JT-60 neutral beam injector was made. A 200 kW, 650 MHz radiofrequency heating system for JFT-2 was completed; a lower hybrid heating experiment in JFT-2 was successful 5) In particle-surface interactions, the sputtering and surface erosion were studied. 6) Improvement designs of a superconducting cluster test facility and a test module coil were made in the toroidal coil development. 7) Second preliminary design of the tokamak experimental fusion reactor JXFR started in April 1977. Safety analyses were made of the main components and system of JXFR on the basis of the first preliminary design. (J.P.N.)

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

STARFIRE: a commercial tokamak fusion power plant study

Energy Technology Data Exchange (ETDEWEB)

1980-09-01

STARFIRE is a 1200 MWe central station fusion electric power plant that utilizes a deuterium-tritium fueled tokamak reactor as a heat source. Emphasis has been placed on developing design features which will provide for simpler assembly and maintenance, and improved safety and environmental characteristics. The major features of STARFIRE include a steady-state operating mode based on continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum system for impurity control and exhaust, high tritium burnup and low vulnerable tritium inventories, superconducting EF coils outside the superconducting TF coils, fully remote maintenance, and a low-activation shield. A comprehensive conceptual design has been developed including reactor features, support facilities and a complete balance of plant. A construction schedule and cost estimate are presented, as well as study conclusions and recommendations.

STARFIRE: a commercial tokamak fusion power plant study

International Nuclear Information System (INIS)

1980-09-01

STARFIRE is a 1200 MWe central station fusion electric power plant that utilizes a deuterium-tritium fueled tokamak reactor as a heat source. Emphasis has been placed on developing design features which will provide for simpler assembly and maintenance, and improved safety and environmental characteristics. The major features of STARFIRE include a steady-state operating mode based on continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum system for impurity control and exhaust, high tritium burnup and low vulnerable tritium inventories, superconducting EF coils outside the superconducting TF coils, fully remote maintenance, and a low-activation shield. A comprehensive conceptual design has been developed including reactor features, support facilities and a complete balance of plant. A construction schedule and cost estimate are presented, as well as study conclusions and recommendations

Plasma boundary experiments on DIII-D tokamak

International Nuclear Information System (INIS)

Mahdavi, M.A.; Brooks, N.; Jackson, G.L.; Langhorn, A.; Leikind, B.; Lippmann, S.; Luxon, J.; Petersen, P.; Petrie, T.; Stambaugh, R.D.; Simonen, T.C.; Staebler, G.; Buchenauer, D.; Futch, A.; Hill, D.N.; Rensink, M.; Hogan, J.; Menon, M.; Mioduszewski, P.; Owen, L.; Matthews, G.

1990-06-01

A survey of the boundary physics research on the DIII-D tokamak and an outline of the DIII-D Advanced Divertor Program (ADP) is presented. We will present results of experiments on impurity control, impurity transport, neutral particle transport, and particle effects on core confinement over a wide range of plasma parameters, I p approx-lt 3 MA, β T approx-lt 10.7%, P(auxiliary) approx-lt 20 MW. Based on the understanding gained in these studies, we in collaboration with a number of other laboratories have devised a series of experiments (ADP) to modify the core plasma conditions through changes in the edge electric field, neutral recycling, and plasma surface interactions. 41 refs., 8 figs., 1 tab

Behaviour of impurities during the H-mode in JET

International Nuclear Information System (INIS)

Gianella, R.; Behringer, K.; Denne, B.; Gottardi, N.; Hellermann, M. von; Morgan, P.D.; Pasini, D.; Stamp, M.F.

1989-01-01

In additionally-heated tokamak discharges, the H-mode phases are reported to display, together with a better energy confinement, a longer global containment time for particles. In particular, steep gradients of electron density and temperature are sustained in the outer region of the plasma column. This enhanced performance is observed especially in discharges in which the activity of edge localized modes (ELMs) is low or absent. High confinement and accumulation of metallic impurities, which quickly give raise to terminal disruptions have been described under similar conditions. In JET H-modes very long impurity confinement times are also observed. However the experimental condition is somewhat more favourable since quiescent H-modes are obtained lasting much longer than the energy confinement times and the radiation from metals is generally negligible. The dominant impurities are normally carbon and oxygen, the latter generally accounting for half or more of the power radiated from the bulk plasma. During the X-point operation the effective influx of carbon into the discharge, which is normally in close correlation with that of deuterium, is substantially reduced while the influx of oxygen, whose production mechanisms is believed to be of a chemical nature, does not show significant variations. (author) 5 refs., 4 figs

Migration of a radioactive beryllium in the ISX-B tokamak

International Nuclear Information System (INIS)

England, A.C.; Hillis, D.L.; Edmonds, P.H.

1985-10-01

One of 12 beryllium tiles on a top rail limiter in the Impurity Study Experiment (ISX-B) tokamak was intentionally made radioactive. The migration of the radioactivity due to melting, ablation, and general excoriation of the radioactive tile was studied. Several hundred milligrams of material from the radioactive tile were spread to the other tiles over the course of the experiment, which consisted of more than 3000 tokamak shots. A small amount of activity was found on the bottom of the tokamak vacuum vessel. The distribution of activity on the other limiter tiles showed a marked radially inward directivity, although all of the tiles received some activity. The possibility that some of the activity was the result of photo- and electro-excitation by runaway electrons cannot be ruled out but probably cannot account for the bulk of the effect

Tokamak start-up with electron-cyclotron heating

International Nuclear Information System (INIS)

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

1981-01-01

Experiments are described in which the start-up voltage in a tokamak is reduced by about a factor of two by the use of a modest amount of electron cyclotron resonance heating power for pre-ionization. The solution of the zero-dimensional start-up equations indicates that the effect is due to the high initial density which increases the rate at which the conductivity increases in the neutral-dominated initial plasma. The effect extrapolates favourably to larger tokamaks. A 50% reduction in the start-up volt-second requirement and impurity reflux is also observed. (author)

Tokamak start-up with electron-cyclotron heating

Energy Technology Data Exchange (ETDEWEB)

Holly, D J; Prager, S C; Shepard, D A; Sprott, J C [Wisconsin Univ., Madison (USA)

1981-11-01

Experiments are described in which the start-up voltage in a tokamak is reduced by about a factor of two by the use of a modest amount of electron cyclotron resonance heating power for pre-ionization. The solution of the zero-dimensional start-up equations indicates that the effect is due to the high initial density which increases the rate at which the conductivity increases in the neutral-dominated initial plasma. The effect extrapolates favourably to larger tokamaks. A 50% reduction in the start-up volt-second requirement and impurity reflux is also observed.

Plasma response on impurity injection in W7-AS

International Nuclear Information System (INIS)

Hildebrandt, D.; Brakel, R.; Elsner, A.; Grigull, P.; Hacker, H.; Burhenn, R.; Fiedler, S.; Giannone, L.; Goerner, C.; Hartfuss, H.J.; Herre, G.; Herrmann, A.; Hofmann, J.V.; Kuehner, G.; Naujoks, D.; Sardei, F.; Weller, A.; Wolf, R.

1997-01-01

In order to study impurity transport and radiation behavior nitrogen has been injected into the scrape-off plasma of limiter-dominated discharges by gas puffing or into natural magnetic islands at the plasma edge of separatrix-dominated discharges by a reciprocating erosion probe. Strong radiative plasma edge cooling with a reduction of the power flux to the limiter could be obtained. The accompanied degradation of the energy confinement and the observed decrease of the central electron temperature can partly be explained by the reduction of the effective heating power. At high plasma densities and sufficiently strong impurity injection phenomena well known from tokamaks, like plasma shrinking, plasma detachment from the limiters and MARFE's were transiently observed. (orig.)

Plasma response on impurity injection in W7-AS

Energy Technology Data Exchange (ETDEWEB)

Hildebrandt, D [Euratom Assoc., Berlin (Germany). Max-Planck-Inst. of Plasma Phys.; Brakel, R; Elsner, A; Grigull, P; Hacker, H; Burhenn, R; Fiedler, S; Giannone, L.; Goerner, C; Hartfuss, H J; Herre, G; Herrmann, A; Hofmann, J V; Kuehner, G; Naujoks, D; Sardei, F; Weller, A; Wolf, R

1997-02-01

In order to study impurity transport and radiation behavior nitrogen has been injected into the scrape-off plasma of limiter-dominated discharges by gas puffing or into natural magnetic islands at the plasma edge of separatrix-dominated discharges by a reciprocating erosion probe. Strong radiative plasma edge cooling with a reduction of the power flux to the limiter could be obtained. The accompanied degradation of the energy confinement and the observed decrease of the central electron temperature can partly be explained by the reduction of the effective heating power. At high plasma densities and sufficiently strong impurity injection phenomena well known from tokamaks, like plasma shrinking, plasma detachment from the limiters and MARFE`s were transiently observed. (orig.).

The study of heat flux for disruption on experimental advanced superconducting tokamak

International Nuclear Information System (INIS)

Yang, Zhendong; Fang, Jianan; Luo, Jiarong; Cui, Zhixue; Gong, Xianzu; Gan, Kaifu; Zhao, Hailin; Zhang, Bin; Chen, Meiwen

2016-01-01

Disruption of the plasma is one of the most dangerous instabilities in tokamak. During the disruption, most of the plasma thermal energy is lost, which causes damages to the plasma facing components. Infrared (IR) camera is an effective tool to detect the temperature distribution on the first wall, and the energy deposited on the first wall can be calculated from the surface temperature profile measured by the IR camera. This paper concentrates on the characteristics of heat flux distribution onto the first wall under different disruptions, including the minor disruption and the vertical displacement events (VDE) disruption. Several minor disruptions have been observed before the major disruption under the high plasma density in experimental advanced superconducting tokamak. During the minor disruption, the heat fluxes are mainly deposited on the upper/lower divertors. The magnetic configuration prior to the minor disruption is a lower single null with the radial distance between the two separatrices in the outer midplane dR_s_e_p = −2 cm, while it changes to upper single null (dR_s_e_p = 1.4 cm) during the minor disruption. As for the VDE disruption, the spatial distribution of heat flux exhibits strong toroidal and radial nonuniformity, and the maximum heat flux received on the dome plate can be up to 11 MW/m"2.

Achieving improved ohmic confinement via impurity injection

International Nuclear Information System (INIS)

Bessenrodt-Weberpals, M.; Soeldner, F.X.

1991-01-01

Improved Ohmic Confinement (IOC) was obtained in ASDEX after a modification of the divertors that allowed a larger (deuterium and impurity) backflow from the divertor chamber. The quality of IOC depended crucially on the wall conditions, i.e. IOC was best for uncovered stainless steels walls and vanished with boronization. Furthermore, IOC was found only in deuterium discharges. These circumstances led to the idea that IOC correlates with the content of light impurities in the plasma. To substantiate this working hypothesis, we present observations in deuterium discharges with boronized wall conditions into which various impurities have been injected with the aim to induce IOC conditions. Firstly, the plasma behaviour in typical IOC discharges is characterized. Secondly, injection experiments with the low-Z impurities nitrogen and neon as well as with the high-Z impurities argon and krypton are discussed. Then, we concentrate on optimized neon puffing that yields the best confinement results which are similar to IOC conditions. Finally, these results are compared with eperiments in other tokamaks and some conclusions are drawn about the effects of the impurity puffing on both, the central and the edge plasma behaviour. (orig.)

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

Material Surface Characteristics and Plasma Performance in the Lithium Tokamak Experiment

Science.gov (United States)

Lucia, Matthew James

The performance of a tokamak plasma and the characteristics of the surrounding plasma facing component (PFC) material surfaces strongly influence each other. Despite this relationship, tokamak plasma physics has historically been studied more thoroughly than PFC surface physics. The disparity is particularly evident in lithium PFC research: decades of experiments have examined the effect of lithium PFCs on plasma performance, but the understanding of the lithium surface itself is much less complete. This latter information is critical to identifying the mechanisms by which lithium PFCs affect plasma performance. This research focused on such plasma-surface interactions in the Lithium Tokamak Experiment (LTX), a spherical torus designed to accommodate solid or liquid lithium as the primary PFC. Surface analysis was accomplished via the novel Materials Analysis and Particle Probe (MAPP) diagnostic system. In a series of experiments on LTX, the MAPP x-ray photoelectron spectroscopy (XPS) and thermal desorption spectroscopy (TDS) capabilities were used for in vacuo interrogation of PFC samples. This represented the first application of XPS and TDS for in situ surface analysis of tokamak PFCs. Surface analysis indicated that the thin (dLi ˜ 100nm) evaporative lithium PFC coatings in LTX were converted to Li2O due to oxidizing agents in both the residual vacuum and the PFC substrate. Conversion was rapid and nearly independent of PFC temperature, forming a majority Li2O surface within minutes and an entirely Li2O surface within hours. However, Li2O PFCs were still capable of retaining hydrogen and sequestering impurities until the Li2 O was further oxidized to LiOH, a process that took weeks. For hydrogen retention, Li2O PFCs retained H+ from LTX plasma discharges, but no LiH formation was observed. Instead, results implied that H+ was only weakly-bound, such that it almost completely outgassed as H 2 within minutes. For impurity sequestration, LTX plasma performance

A comparison of tokamak operation with metallic getters (Ti, Cr, Be) and boronization

International Nuclear Information System (INIS)

Winter, J.

1990-07-01

In addition to discharge cleaning techniques, gettering of tokamaks has been used since 1975 as a powerful tool for controlling the impurity influx into fusion plasmas. High-Z metals like Ti and Cr, evaporated onto the walls of the fusion devices, have first been used. After the introduction of carbon as low Z plasma facing material for the large tokamaks new scenarios were developed, optimizing the low-Z aspect of wall materials. These are the boronization technique and the evaporation of Be in conjunction with the use of Be limiters. A review of the different getter techniques and of the observed results will be given, focussing on the comparison of the tokamak performance achieved with boronization and the use of beryllium. It is shown that in all cases of gettering the most important mechanism for the improved machine performance is the control of the oxygen impurity influx. Very similar results are found for the impurity control potential. The added benefit of boronization and Be gettering arises from the low Z of the materials. Both scenarios essentially lead to the same machine performance. Both render themselves as an option for future devices. (orig.)

Deposition of deuterium and metals on divertor tiles in the DIII-D tokamak

International Nuclear Information System (INIS)

Walsh, D.S.; Doyle, B.L.; Jackson, G.L.

1991-01-01

Hydrogen recycling and impurity influx are important issues in obtaining high confinement discharges in the D3-D tokamak. To reduce metallic impurities in D3-D, 40% of the wall area, including the highest heat flux zones, have been covered with graphite tiles. However erosion, redeposition and hydrogen retention in the tiles, as well as metal transport from the remaining Inconel walls can lead to enhanced recycling and impurity influx. Hydrogen and metal retention in divertor floor tiles have been measured using external ion beam analysis techniques following four campaigns where tiles were exposed to several thousand tokamak discharges. The areal density of deuterium retained following exposure to tokamak plasmas was measured with external nuclear reaction analysis. External proton-induced x-ray emission analysis was used to measure the areal densities of metallic impurities deposited upon the divertor tiles either by sputtering of metallic components during discharges or as contamination during tile fabrication. Measurements for both deuterium and metallic impurities were taken on both the tile surfaces which face the operating plasma and the surfaces on the side of the tiles which form the small gaps separating each of the tiles in the divertor. The highest areal densities of both deuterium and metals were found on the plasma-facing surface near the inner strike point region of each set of divertor tiles. Significant deposits, extending as fast a 1 cm from the plasma-facing and containing up to forty percent of the total divertor deposition, were also observed on the gap-forming surfaces of the tiles

Impurity transport in a collision-dominated rotating tokamak plasma

International Nuclear Information System (INIS)

Eriksson, G.; Liljegren, A.

1981-04-01

The flux of heavy impurities is an axisymmetric, toroidal plasma with all particles in the collision-dominated regime is considered. Plasma rotation and charge-exchange with neutrals are taken into account. A hydrodynamic model employing Braginskii's transport equations is used. The theorry is extended to higher collision freqencies as compared to previous treatments. It is found that the Pfirsch-Schlueter flux is significantly reduced as compared to the value given by Rutherford and that it is of the same order of magnitude, or less, than the classical flux in all regimes considered. It is also shown that the impurity flux can be influenced by charge-exchange with neutrals. (author)

Scaling for scrape-off layer plasma in tokamak

International Nuclear Information System (INIS)

Shimomura, Yasuo; Maeda, Hikosuke; Kimura, Haruyuki; Azumi, Masashi; Odajima, Kazuo

1977-12-01

Scaling for a scrape-off layer plasma in a tokamak is obtained by using DIVA (JFT-2a). The scaling gives the average electron temperature, the width and the mean electron density of the scrape-off layer. The temperature at the edge will be high in a future large tokamak with a small energy-loss by charge-exchange and radiation. The scrape-off layer plasma can easily shield the impurity influx from the wall. The fuel, however, can easily penetrate into the main plasma. (auth.)

Accessibility and replacement as prime constraints in the design of large experimental tokamaks

International Nuclear Information System (INIS)

Challender, R.S.; Reynolds, P.

1976-01-01

An attempt is made to bring together those design features of large, experimental Tokamaks, which would lead to better accessibility during non-active operation and, in particular, permit replacement and repair after activation, thereby making possible an extended period of experimental operation into the ignition phase

Experimental operation of the KT-5C tokamak in USTC

International Nuclear Information System (INIS)

Wen Yizhi; Wan Shude; Zhai Kan; Liu Wandong

1995-01-01

Experimental operation of the KT-5C tokamak was started in early 1991. More than 3 x 10 4 shots of discharges have been performed so far. The authors deals with the major features of operation control of the KT-5C device. As the machine is usually controlled by an experimenter himself without duty operator, an audible indicator makes it much easier to control and operate. Accident and interference prevention systems prove to be reliable, and the operation system works successfully

3D simulation studies of tokamak plasmas using MHD and extended-MHD models

International Nuclear Information System (INIS)

Park, W.; Chang, Z.; Fredrickson, E.; Fu, G.Y.

1996-01-01

The M3D (Multi-level 3D) tokamak simulation project aims at the simulation of tokamak plasmas using a multi-level tokamak code package. Several current applications using MHD and Extended-MHD models are presented; high-β disruption studies in reversed shear plasmas using the MHD level MH3D code, ω *i stabilization and nonlinear island saturation of TAE mode using the hybrid particle/MHD level MH3D-K code, and unstructured mesh MH3D ++ code studies. In particular, three internal mode disruption mechanisms are identified from simulation results which agree which agree well with experimental data

Alpha particle diagnostics using impurity pellet injection (invited)

International Nuclear Information System (INIS)

Fisher, R.K.; McChesney, J.M.; Howald, A.W.; Parks, P.B.; Snipes, J.A.; Terry, J.L.; Marmar, E.S.; Zweben, S.J.; Medley, S.S.

1992-01-01

We have proposed using impurity pellet injection to measure the energy distribution of the fast confined alpha particles in a reacting plasma [R. K. Fisher et al., Fusion Technol. 13, 536 (1988)]. The ablation cloud surrounding the injected pellet is thick enough that an equilibrium fraction F ∞ 0 (E) of the incident alphas should be neutralized as they pass through the cloud. By observing neutrals created in the large spatial region of the cloud which is expected to be dominated by the heliumlike ionization state, e.g., Li + ions, we can determine the incident alpha distribution dn He 2+ /dE from the measured energy distribution of neutral helium atoms dn He 0 /dE using dn He 0 /dE = dn He 2+ /dE·F ∞ 0 (E,Li + ). Initial experiments were performed on the Texas Experimental Tokamak (TEXT) in which we compared pellet penetration with our impurity pellet ablation model [P. B. Parks et al., Nucl. Fusion 28, 477 (1988)], and measured the spatial distribution of various ionization states in carbon pellet clouds [R. K. Fisher et al., Rev. Sci. Instrum. 61, 3196 (1990)]. Experiments have recently begun on the Tokamak Fusion Test Reactor (TFTR) with the goal of measuring the alpha particle energy distribution during D--T operation in 1993--94. A series of preliminary experiments are planned to test the diagnostic concept. The first experiments will observe neutrals from beam-injected deuterium ions and the high energy 3 He tail produced during ion cyclotron (ICH) minority heating on TFTR interacting with the cloud. We will also monitor by line radiation the charge state distributions in lithium, boron, and carbon clouds

Role of impurities in magnetically confined high temperature plasmas

International Nuclear Information System (INIS)

Barnett, C.F.

1976-01-01

A summary is given of the atomic physics concerned with plasma cooling by impurities and the limiting effect that impurities may have on heating of plasmas by neutral injection. A general description is given of the tokamak concept and the present and next generation experiments are described. The time and spatial behavior of O and Mo multicharged ions in present hydrogen plasmas is presented. This is followed by a discussion of the power loss from a plasma containing one percent Fe. Finally, the limitation of plasma heating by energetic H or D injection is summarized

Poloidal asymmetries of the heavy ions in the ASDEX Upgrade tokamak

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-01

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

Experimental study on practicability of self-created spherical tokamak in coilless STPC-EX machine

International Nuclear Information System (INIS)

Sinman, S.

2002-01-01

The aim of this study is to recognize the physical basis of the alternative self organization mechanism occurred STPC-EX machine. The conventional diagnostic tools are used in this study and for photographic recording, open shutter integrated post-fogging method is preferred. The annular coaxial two plasma current sheets one within other at the same direction are created and flowed on the surface of floating conductive central rod. Consequently, spherical tokamak configurated by new creation mechanism of Dual Axial Z-Pinch. (DAZP) yields fairly high beta of 0.4-0.6 at self created spherical tokamak plasma. Sustainment time of DAZP is 5.6-6.3 mili second. (author)

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

Analyzing the Radiation Properties of High-Z Impurities in High-Temperature Plasmas

International Nuclear Information System (INIS)

Reinke, M. L.; Ince-Cushman, A.; Podpaly, Y.; Rice, J. E.; Bitter, M.; Hill, K. W.; Fournier, K. B.; Gu, M. F.

2009-01-01

Most tokamak-based reactor concepts require the use of noble gases to form either a radiative mantle or divertor to reduce conductive heat exhaust to tolerable levels for plasma facing components. Predicting the power loss necessary from impurity radiation is done using electron temperature-dependent 'cooling-curves' derived from ab initio atomic physics models. We present here a technique to verify such modeling using highly radiative, argon infused discharges on Alcator C-Mod. A novel x-ray crystal imaging spectrometer is used to measure spatially resolved profiles of line-emissivity, constraining impurity transport simulations. Experimental data from soft x-ray diodes, bare AXUV diodes and foil bolometers are used to determine the local emissivity in three overlapping spectral bands, which are quantitatively compared to models. Comparison of broadband measurements show agreement between experiment and modeling in the core, but not over the entire profile, with the differences likely due to errors in the assumed radial impurity transport outside of the core. Comparison of Ar 16+ x-ray line emission modeling to measurements suggests an additional problem with the collisional-radiative modeling of that charge state.

Investigation of the energy transport mechanism in the TCA tokamak by studying the plasma dynamical response

International Nuclear Information System (INIS)

Dudok de Wit, Th.; Duval, B.P.; Joye, B.; Lister, J.B.; Moret, J.M.

1989-01-01

The energy transport mechanisms that govern the electron temperature behaviour of a tokamak remain very badly understood and up to now no proper model has been proposed that can explain experimental observations such as profile consistency or the influence of the density profile. One approach to this problem, extensively used on TCA, is to study the dynamical response of the plasma due to externally imposed modifications of parameters which have an influence on the plasma energy content. The temporal evolution of the electron temperature will closely depend on the type and the characteristics of the implied mechanisms. Thus a detailed measurement of the dynamical response would reveal experimentally the dominant properties that would have to be taken into account in the elaboration of a model of the transport processes. Most of the results presented here were obtained by analysing the electron temperature response inferred from soft X-ray emissivity during modification of the plasma density due to either gas puffing, laser impurity ablation or alfven wave heating on TCA (a = 0.18 m, R = 0.61 m, B Φ = 1.52 T). 4 refs., 3 figs

3D simulation studies of tokamak plasmas using MHD and extended-MHD models

International Nuclear Information System (INIS)

Park, W.; Chang, Z.; Fredrickson, E.; Fu, G.Y.; Pomphrey, N.; Sugiyama, L.E.

1997-01-01

The M3D (Multi-level 3D) tokamak simulation project aims at the simulation of tokamak plasmas using a multi-level tokamak code package. Several current applications using MHD and Extended-MHD models are presented; high-β disruption studies in reversed shear plasmas using the MHD level MH3D code, ω *i stabilization and nonlinear island rotation studies using the two-fluid level MH3D-T code, studies of nonlinear saturation of TAE modes using the hybrid particle/MHD level MH3D-K code, and unstructured mesh MH3D ++ code studies. In particular, three internal mode disruption mechanisms are identified from simulation results which agree well with experimental data

Impurity confinement and transport in high confinement regimes without edge localized modes on DIII-D

Energy Technology Data Exchange (ETDEWEB)

Grierson, B. A., E-mail: bgriers@pppl.gov; Nazikian, R. M.; Solomon, W. M. [Princeton Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (United States); Burrell, K. H.; Garofalo, A. M.; Belli, E. A.; Staebler, G. M.; Evans, T. E.; Smith, S. P.; Chrobak, C. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); Fenstermacher, M. E. [Lawerence Livermore National Laboratory, P.O. Box 808, Livermore, California 94550 (United States); McKee, G. R. [Department of Engineering Physics, University of Wisconsin-Madison, Madison, Wisconsin 53796 (United States); Orlov, D. M. [Center for Energy Research, University of California San Diego, 9500 Gilman Dr., La Jolla, California 92093-0417 (United States); Chrystal, C. [University of California San Diego, 9500 Gilman Dr., La Jolla, California 92093-0417 (United States)

2015-05-15

Impurity transport in the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] is investigated in stationary high confinement (H-mode) regimes without edge localized modes (ELMs). In plasmas maintained by resonant magnetic perturbation (RMP), ELM-suppression, and QH-mode, the confinement time of fluorine (Z = 9) is equivalent to that in ELMing discharges with 40 Hz ELMs. For selected discharges with impurity injection, the impurity particle confinement time compared to the energy confinement time is in the range of τ{sub p}/τ{sub e}≈2−3. In QH-mode operation, the impurity confinement time is shown to be smaller for intense, coherent magnetic, and density fluctuations of the edge harmonic oscillation than weaker fluctuations. Transport coefficients are derived from the time evolution of the impurity density profile and compared to neoclassical and turbulent transport models NEO and TGLF. Neoclassical transport of fluorine is found to be small compared to the experimental values. In the ELMing and RMP ELM-suppressed plasma, the impurity transport is affected by the presence of tearing modes. For radii larger than the mode radius, the TGLF diffusion coefficient is smaller than the experimental value by a factor of 2–3, while the convective velocity is within error estimates. Low levels of diffusion are observed for radii smaller than the tearing mode radius. In the QH-mode plasma investigated, the TGLF diffusion coefficient is higher inside of ρ=0.4 and lower outside of 0.4 than the experiment, and the TGLF convective velocity is more negative by a factor of approximately 1.7.

Tokamak power systems studies at ANL

International Nuclear Information System (INIS)

Baker, C.C.; Ehst, D.A.; Brooks, J.N.; Evans, K. Jr.

1986-01-01

A number of advances in plasma physics and engineering promise to greatly improve the reactor prospects of tokamaks. The following features, in particular, are examined: (a) large aspect ratio (A ≅ 6), which may ease maintenance; (b) high beta (β ≥ 0.20) without indentation, which brings the maximum toroidal field down to about 7 T; (c) low toroidal current (I ≅ 5MA), which reduces the cost of the current drive and equilibrium field system; and (d) steady state operation with current density control via fast and slow wave current drive. The key to high beta operation with low toroidal current lies in utilizing second stability regime equilibria with the required current distributions produced by an appropriate selection of wave driver frequencies and power spectra. The ray tracing and current drive calculation is self-consistent with the actual magnetic fields produced in the plasma. In addition to matching desirable high-beta equilibria, this method is capable of producing a large variety of new equilibria, many of which look attractive. The impurity control activities in TPSS have emphasized the self-pumping concept as applied to using the entire first wall or ''slot'' limiters. The blanket design effort has emphasized liquid metal and Flibe concepts. The reference concept is a liquid lithium/vanadium, self-cooled configuration. Overall, there exists a number of major design improvements which will substantially improve the attractiveness of tokamak reactors

Study of a compact reversed shear Tokamak reactor

International Nuclear Information System (INIS)

Okano, K.; Asaoka, Y.; Tomabechi, K.; Yoshida, T.; Hiwatari, R.; Ogawa, Y.; Tokimatsu, K.; Yamamoto, T.; Inoue, N.; Murakami, Y.

1998-01-01

A reversed shear configuration, which was observed recently in some tokamak experiments, might have a possibility to realize compact and cost-competitive tokamak reactors. In this study, a compact (low cost) commercial reactor based on the shear reversed high beta equilibrium with β N =5.5, is considered, namely the compact reversed shear tokamak, CREST-1. The CREST-1 is designed with a moderate aspect ratio (R/a=3.4), which will allow us to experimentally develop this CREST concept by ITER. This will be very advantageous with regard to the fusion development strategy. The current profile for the reversed shear operation is sustained and controlled in steady state by bootstrap (88%), beam and r driven currents, which are calculated by a neo-classical model code in 3D geometry. The MHD stability has been checked by an ideal MHD stability analysis code (ERATO) and it has been confirmed that the ideal low n kink, ballooning and Mercier modes are stable while a closed conductive shell is required for stability. Such a compact tokamak can be cost-competitive as an electric power source in the 21st century and it is one possible scenario in realizing a commercial fusion reactor beyond the ITER project. (orig.)

Study of the electron heat transport in Tore-Supra tokamak

International Nuclear Information System (INIS)

Harauchamps, E.

2004-01-01

This work presents analytical solutions to the electron heat transport equation involving a damping term and a convection term in a cylindrical geometry. These solutions, processed by Matlab, allow the determination of the evolution of the radial profile of electron temperature in tokamaks during heating. The modulated injection of waves around the electron cyclotron frequency is an efficient tool to study heat transport experimentally in tokamaks. The comparison of these analytical solutions with experimental results from Tore-Supra during 2 discharges (30550 and 31165) shows the presence of a sudden change for the diffusion and damping coefficients. The hypothesis of the presence of a pinch spread all along the plasma might explain the shape of the experimental temperature profiles. These analytical solutions could be used to determine the time evolution of plasma density as well or of any parameter whose evolution is governed by a diffusion-convection equation. (A.C.)

New steady-state quiescent high-confinement plasma in an experimental advanced superconducting tokamak.

Science.gov (United States)

Hu, J S; Sun, Z; Guo, H Y; Li, J G; Wan, B N; Wang, H Q; Ding, S Y; Xu, G S; Liang, Y F; Mansfield, D K; Maingi, R; Zou, X L; Wang, L; Ren, J; Zuo, G Z; Zhang, L; Duan, Y M; Shi, T H; Hu, L Q

2015-02-06

A critical challenge facing the basic long-pulse high-confinement operation scenario (H mode) for ITER is to control a magnetohydrodynamic (MHD) instability, known as the edge localized mode (ELM), which leads to cyclical high peak heat and particle fluxes at the plasma facing components. A breakthrough is made in the Experimental Advanced Superconducting Tokamak in achieving a new steady-state H mode without the presence of ELMs for a duration exceeding hundreds of energy confinement times, by using a novel technique of continuous real-time injection of a lithium (Li) aerosol into the edge plasma. The steady-state ELM-free H mode is accompanied by a strong edge coherent MHD mode (ECM) at a frequency of 35-40 kHz with a poloidal wavelength of 10.2 cm in the ion diamagnetic drift direction, providing continuous heat and particle exhaust, thus preventing the transient heat deposition on plasma facing components and impurity accumulation in the confined plasma. It is truly remarkable that Li injection appears to promote the growth of the ECM, owing to the increase in Li concentration and hence collisionality at the edge, as predicted by GYRO simulations. This new steady-state ELM-free H-mode regime, enabled by real-time Li injection, may open a new avenue for next-step fusion development.

Upgrade of MHD data acquisition system from ISX-B [Impurity Study Experiment] to ATF [Advanced Toroidal Facility

International Nuclear Information System (INIS)

Bell, J.D.; Pare, V.L.

1987-01-01

The data acquisition system assembled to study magnetohydrodynamic (MHD) activity on the Impurity Study Experiment (ISX-B) tokamak at Oak Ridge National Laboratory (ORNL) is being revised for use on the Advanced Toroidal Facility (ATF). The new hardware and software architectures are based on ISX-B experience and will feature different modes of operation for storing various subsets of available data, a user interface that requires less routine activity than the earlier system, and continued support of calibration and testing measurement used on ISX-B. The new hardware organization and software components are described in detail. 2 refs., 5 figs., 1 tab

Experimental study of tungsten transport properties in T-10 plasma

Science.gov (United States)

Krupin, V. A.; Nurgaliev, M. R.; Klyuchnikov, L. A.; Nemets, A. R.; Zemtsov, I. A.; Dnestrovskij, A. Yu.; Sarychev, D. V.; Lisitsa, V. S.; Shurygin, V. A.; Leontiev, D. S.; Borschegovskij, A. A.; Grashin, S. A.; Ryjakov, D. V.; Sergeev, D. S.; Mustafin, N. A.; Trukhin, V. M.; Solomatin, R. Yu.; Tugarinov, S. N.; Naumenko, N. N.

2017-06-01

First experimental results of tungsten transport investigation in OH and ECRH plasmas in the T-10 tokamak with W-limiter and movable Li-limiter are presented. It is shown that tungsten tends to accumulate (a joint process of cumulation and peaking) near the plasma axis in ohmic regimes. The cumulation of W is enhanced in discharges with high values of the parameter γ ={{\\bar{n}}\\text{e}}\\centerdot {{\\bar{Z}}\\text{eff}}\\centerdot I\\text{pl}-1.5 that coincides with accumulation conditions of light and medium impurities in T-10 plasmas. Experiments with Li-limiter show the immeasurable level of Li3+ (0.3-0.5% of n e) of T-10 CXRS diagnostics because of the low inflow of Li with respect to other light impurities. Nevertheless, the strong influence of lithium on inflow of light and tungsten impurities is observed. In discharges with lithized walls, vanishing of light impurities occurs and values of {{Z}\\text{eff}}≈ 1 are obtained. It is also shown that the tungsten density in the plasma center decreases by 15 to 20 times while the W inflow reduces only by 2 to 4 times. In lithized discharges with high γ, the flattening of the tungsten density profile occurs and its central concentration decreases up to 10 times during the on-axis ECRH. This effect is observed together with the increase of the W inflow by 3 to 4 times at the ECRH stage.

Tokamak first-wall coating program development

International Nuclear Information System (INIS)

Davis, M.J.; Langley, R.A.; Prevender, T.S.

1977-08-01

The development of a research program to study coatings for control of impurities originating from the first wall of a Tokamak reactor is extensively discussed. The first wall environment and sputtering, temperature, surface chemical, and bulk radiation damage effects are reviewed. Candidate materials and application techniques are discussed. The philosophy and flow chart of a recommended coating development plan are presented and discussed. Projected impacts of the proposed plan include benefits to other aspects of confinement experiments. A list of 45 references is appended

Quasilinear Carbon Transport In An Impurity Hole Plasma In LHD

Energy Technology Data Exchange (ETDEWEB)

Mikkelsen, David R. [PPPL; Tanaka, K. [NIFS; Nunami, M. [NIFS; Watanabe, T-H. [Nagoya University; Sugama, H. [NIFS; Yoshinuma, M. [NIFS; Suzuki, Y. [NIFS; Goto, M. [NIFS; Morita, S. [NIFS; Wieland, B. [NIFS; Yamada, I. [NIFS; Yashura, R. [NIFS; Akiyama, T. [NIFS; Pablant, Novimir A. [PPPL

2014-04-01

Comprehensive electrostatic gyrokinetic linear stability calculations for ion-scale microinstabilities in an LHD plasma with an ion-ITB and carbon "impurity hole" are used to make quasilinear estimates of particle flux to explore whether microturbulence can explain the observed outward carbon fluxes that flow "up" the impurity density gradient. The ion temperature is not stationary in the ion-ITB phase of the simulated discharge, during which the core carbon density decreases continuously. To fully sample these varying conditions the calculations are carried out at three radial locations and four times. The plasma parameter inputs are based on experimentally measured profiles of electron and ion temperature, as well as electron and carbon density. The spectroscopic line-average ratio of hydrogen and helium densities is used to set the density of these species. Three ion species (H,He,C) and the electrons are treated kinetically, including collisions. Electron instability drive does enhance the growth rate significantly, but the most unstable modes have characteristics of ion temperature gradient (ITG) modes in all cases. As the carbon density gradient is scanned between the measured value and zero, the quasilinear carbon flux is invariably inward when the carbon density profile is hollow, so turbulent transport due to the instabilities considered here does not explain the observed outward flux of impurities in impurity hole plasmas. The stiffness of the quasilinear ion heat flux is found to be 1.7-2.3, which is lower than several estimates in tokamaks.

ZORNOC: a 1 1/2-D tokamak data analysis code for studying noncircular high beta plasmas

International Nuclear Information System (INIS)

Zurro, B.; Wieland, R.M.; Murakami, M.; Swain, D.W.

1980-03-01

A new tokamak data analysis code, ZORNOC, was developed to study noncircular, high beta plasmas in the Impurity Study Experiment (ISX-B). These plasmas exhibit significant flux surface shifts and elongation in both ohmically heated and beam-heated discharges. The MHD equilibrium flux surface geometry is determined by solving the Grad-Shafranov equation based on: (1) the shape of the outermost flux surface, deduced from the magnetic loop probes; (2) a pressure profile, deduced by means of Thomson scattering data (electrons), charge exchange data (ions), and a Fokker-Planck model (fast ions); and (3) a safety factor profile, determined from the experimental data using a simple model (Z/sub eff/ = const) that is self-consistently altered while the plasma equilibrium is iterated. For beam-heated discharches the beam deposition profile is determined by means of a Monte Carlo scheme and the slowing down of the fast ions by means of an analytical solution of the Fokker-Planck equation. The code also carries out an electron power balance and calculates various confinement parameters. The code is described and examples of its operation are given

Deposition of deuterium and metals on divertor tiles in the DIII--D tokamak

International Nuclear Information System (INIS)

Walsh, D.S.; Doyle, B.L.; Jackson, G.L.

1992-01-01

Hydrogen recycling and impurity influx are important issues in obtaining high confinement discharges in the DIII--D tokamak. To reduce metallic impurities in DIII--D, 40% of the wall area, including the highest heat flux zones, have been covered with graphite tiles. However, erosion, redeposition, and hydrogen retention in the tiles, as well as metal transport from the remaining Inconel walls, can lead to enhanced recycling and impurity influx. Hydrogen and metal retention in divertor floor tiles have been measured using external ion beam analysis techniques following four campaigns where tiles were exposed to several thousand tokamak discharges. The areal density of deuterium retained following exposure to tokamak plasmas was measured with external nuclear reaction analysis. External proton-induced x-ray emission analysis was used to measure the areal densities of metallic impurities deposited upon the divertor tiles either by sputtering of metallic components during discharges or as contamination during tile fabrication. Measurements for both deuterium and metallic impurities were taken on both the tile surfaces which face the operating plasma and the surfaces on the sides of the tiles which form the small gaps separating each of the tiles in the divertor. The highest areal densities of both deuterium (from 2 to 8 x 10 18 atoms/cm 2 ) and metals (from 0.2 to 1 x 10 18 atoms/cm 2 ) were found on the plasma-facing surface near the inner strike point region of each set of divertor tiles. Significant deposits, extending as far as 1 cm from the plasma-facing surface and containing up to 40% of the total divertor deposition, were also observed on the gap-forming surfaces of the tiles

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

Overview on the progress of tokamak experimental research in China

International Nuclear Information System (INIS)

Xie Jikang . E-mail; Liu Yong; Wen Yizhi; Wang Long

2001-01-01

Tokamak experimental research in China has made important progress. The main efforts were related to quasi-steady-state operation, LHCD, plasma heating with ICRF, IBW, NBI and ECRH, fuelling with pellets and supersonic molecular beams, and first wall conditioning techniques. Plasma parameters in the experiments were much improved, for example n e =8x10 19 m -3 and a plasma pulse length of >10 s were achieved. ICRF boronization and conditioning resulted in Z eff close to unity. Steady state full LH wave current drive has been achieved for more than 3 s. LHCD ramp-up and recharge have also been demonstrated. The best η CD exp ∼0.5(1+0.085exp(4.8(B T -1.45)))n e I CD R p /P LH =10 19 m -2 A W -1 . Quasi-steady-state H-mode-like plasmas with a density close to the Greenwald limit were obtained by LHCD, where the energy confinement time was nearly five times longer than in the ohmic case. The synergy between IBW, pellet and LHCD was tested. Research on the mechanism of macroturbulence has been extensively carried out experimentally. AC tokamak operation has been successfully demonstrated. (author)

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)

Simulation of MHD instability effects on burning plasma transport with ITB in tokamak and helical reactors

International Nuclear Information System (INIS)

Yamazaki, K.; Yamada, I.; Taniguchi, S.; Oishi, T.

2009-01-01

Full text: The high performance plasma behavior is required to realize economic and environmental-friendly fusion reactors compatible with conventional power plant systems. To improve plasma confinement, the formation of internal transport barrier (ITB) is anticipated, and its behavior is analyzed by the simulation code TOTAL (Toroidal Transport Linkage Analysis). This TOTAL code comprises a 2- or 3-dimensional equilibrium and 1-dimensional predictive transport code for both tokamak and helical systems. In the tokamak code TOTAL-T, the external current drive, bootstrap current, sawtooth oscillation, ballooning mode and neoclassical tearing mode (NTM) analyses are included. The steady-state burning plasma operation is achieved by the feedback control of pellet injection fuelling and external heating power control. The impurity dynamics of iron and tungsten is also included in this code. The NTM effects are evaluated using the modified Rutherford Model with the stabilization of the ECCD current drive. The excitation of m=2/n=1 NTM leads to the 20 % reduction in the central temperature in ITER-like reactors. Recently, the external non-resonant helical field application is analyzed and its stabilization properties are evaluated. The pellet injection effects on ITB formation is also clarified in tokamak and helical plasmas. Relationship between sawtooth oscillation and impurity ejection is recently simulated in comparison with experimental data. In this conference, we will show above-stated new results on MHD instability effects on burning plasma transport. (author)

Experimental device for the X-ray energetic distribution measurement in a tokamak plasma

International Nuclear Information System (INIS)

Perez-Navarro, A.

1977-01-01

An experimental system to measure the X-ray spectrum in a tokamak plasma is described, emphasizing its characteristics: resolution, dead time and the pulse pile-up distortion effects on the X-ray spectra. (author) [es

Investigation of impurity confinement in lower hybrid wave heated plasma on EAST tokamak

Science.gov (United States)

Xu, Z.; Wu, Z. W.; Zhang, L.; Gao, W.; Ye, Y.; Chen, K. Y.; Yuan, Y.; Zhang, W.; Yang, X. D.; Chen, Y. J.; Zhang, P. F.; Huang, J.; Wu, C. R.; Morita, S.; Oishi, T.; Zhang, J. Z.; Duan, Y. M.; Zang, Q.; Ding, S. Y.; Liu, H. Q.; Chen, J. L.; Hu, L. Q.; Xu, G. S.; Guo, H. Y.; the EAST Team

2018-01-01

The transient perturbation method with metallic impurities such as iron (Fe, Z  =  26) and copper (Cu, Z  =  29) induced in plasma-material interaction (PMI) procedure is used to investigate the impurity confinement characters in lower hybrid wave (LHW) heated EAST sawtooth-free plasma. The dependence of metallic impurities confinement time on plasma parameters (e.g. plasma current, toroidal magnetic field, electron density and heating power) are investigated in ohmic and LHW heated plasma. It is shown that LHW heating plays an important role in the reduction of the impurity confinement time in L-mode discharges on EAST. The impurity confinement time scaling is given as 42IP0.32Bt0.2\\overline{n}e0.43Ptotal-0.4~ on EAST, which is close to the observed scaling on Tore Supra and JET. Furthermore, the LHW heated high-enhanced-recycling (HER) H-mode discharges with ~25 kHz edge coherent modes (ECM), which have lower impurity confinement time and higher energy confinement time, provide promising candidates for high performance and steady state operation on EAST.

Effect of impurities and post-experimental purification in SAD phasing with serial femtosecond crystallography data.

Science.gov (United States)

Zhang, Tao; Gu, Yuanxin; Fan, Haifu

2016-06-01

In serial crystallography (SX) with either an X-ray free-electron laser (XFEL) or synchrotron radiation as the light source, huge numbers of micrometre-sized crystals are used in diffraction data collection. For a SAD experiment using a derivative with introduced heavy atoms, it is difficult to completely exclude crystals of the native protein from the sample. In this paper, simulations were performed to study how the inclusion of native crystals in the derivative sample could affect the result of SAD phasing and how the post-experimental purification proposed by Zhang et al. [(2015), Acta Cryst. D71, 2513-2518] could be used to remove the impurities. A gadolinium derivative of lysozyme and the corresponding native protein were used in the test. Serial femtosecond crystallography (SFX) diffraction snapshots were generated by CrystFEL. SHELXC/D, Phaser, DM, ARP/wARP and REFMAC were used for automatic structure solution. It is shown that a small amount of impurities (snapshots from native crystals) in the set of derivative snapshots can strongly affect the SAD phasing results. On the other hand, post-experimental purification can efficiently remove the impurities, leading to results similar to those from a pure sample.

Commissioning and preliminary operation of HL-2A tokamak

International Nuclear Information System (INIS)

Liu Dequan; Liu Yong; Yan Jianchen; Cao Zeng; Yang Qingwei; Zhou Caipin; Li Xiaodong

2005-01-01

HL-2A is a divertor tokamak located at new site of Southwestern Institute of Physics (SWIP), Chengdu, China. It can be operated in double-null and single-null divertor with closed configurations. The effects of the divertor on impurity behaviors, MHD instabilities, transport, wall conditioning and divertor physics are key issues to be studied during the first step operation on HL-2A. Preliminary experiments with limiter and single null divertor configurations were carried out in 2003. Main parameters achieved are the plasma current 168 KA, duration time of 920 ms and plasma linear-averaged density of 1.7 x 10 19 m -3 . The impurities, especially of low Z, are clearly decreased during the divertor experiments

Engineering aspects of a D-D commercial tokamak reactor

International Nuclear Information System (INIS)

Evans, K. Jr.; Baker, C.C.; Brooks, J.N.

1981-01-01

This paper presents some of the engineering aspects of WILDCAT, a conceptual design of a D-D tokamak, fusion reactor. This conceptual design has evolved from initial studies of D-D tokamak reactors, and is intended to be a study of a later-model, commerical fusion reactor in the same sense that STARFIRE was such a study for D-T fuel cycle. The major guidelines of the study have been to utilize as fully as possible the advantages of the D-D fuel cycle but to avoid unnecessary extrapolations of parameters from existing D-T designs, in particular STARFIRE. The paper consists of an overview of the reference design, a description of each of the major engineering systems (rf current drive, burn cycle, impurity control, first wall, blanket/shield, TF magnets, and tritium system, and a summary of conclusions)

An electrostatic detector for dust measurement on HT-7 tokamak

International Nuclear Information System (INIS)

Ling, B.L.; Zhang, X.D.; Ti, A.; Gao, X.

2007-01-01

An electrostatic dust detector has been successfully developed to measure dust event in situ and in real time on the HT-7 tokamak. For measuring dust near the edge plasmas and preventing interference of electrons and ions, the shielding plates were designed and installed around the dust detector. The electric signal of dust has been successfully measured during LHCD discharges on HT-7 tokamak. The measured dust signal was in good agreement with bursts appeared on multi-channel H α radiation and on multi-channel ECE diagnostics. Diagnostics of the spectrum and the measurement of impurity emission during dust bursts were studied in detail. It is interesting that there is a delay between dust bursts and CIII line emission. It is observed that the delay time between dust signal and measured CIII line emission is about 0.3 ms in the HT-7 tokamak

A Fast Shutdown Technique for Large Tokamaks

International Nuclear Information System (INIS)

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

1999-01-01

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

Divertor impurity monitor for the International Thermonuclear Experimental Reactor

Science.gov (United States)

Sugie, T.; Ogawa, H.; Nishitani, T.; Kasai, S.; Katsunuma, J.; Maruo, M.; Ebisawa, K.; Ando, T.; Kita, Y.

1999-01-01

The divertor impurity monitoring system of the International Thermonuclear Experimental Reactor has been designed. The main functions of this system are to identify impurity species and to measure the two-dimensional distributions of the particle influxes in the divertor plasmas. The wavelength range is 200-1000 nm. The viewing fans are realized by molybdenum mirrors located in the divertor cassette. With additional viewing fans seeing through the gap between the divertor cassettes, the region approximately from the divertor leg to the x point will be observed. The light from the divertor region passes through the quartz windows on the divertor port plug and the cryostat, and goes through the dog-leg optics in the biological shield. Three different type of spectrometers: (i) survey spectrometers for impurity species monitoring, (ii) filter spectrometers for the particle influx measurement with the spatial resolution of 10 mm and the time resolution of 1 ms, and (iii) high dispersion spectrometers for high resolution wavelength measurements are designed. These spectrometers are installed just behind the biological shield (for λthe transmission loss in fiber and in the diagnostic room (for λ⩾450 nm) from the point of view of accessibility and flexibility. The optics have been optimized by a ray trace analysis. As a result, 10-15 mm spatial resolution will be achieved in all regions of the divertor.

Impurity diagnosis of a KSTAR graphite divertor tile using laser induced breakdown spectroscopy technique

Energy Technology Data Exchange (ETDEWEB)

Kim, Minju; Cho, Min Sang; Cho, Byoung Ick, E-mail: bicho@gist.ac.kr

2017-04-15

Laser induced breakdown spectroscopy (LIBS) has been tested to diagnose impurity elements on a Korea Superconducting Tokamak Advanced Research (KSTAR) divertor tile. Spectral lines of various impurity elements such as iron, chromium, and nickel were detected from the divertor surface. The variation of spectra with consecutive laser pulses demonstrates the potential for depth profiling analysis for the deposited impurity layer. The LIBS plasma parameters have been qualitatively determined from analysis of the relative line intensities and linewidths for each element. The validity of this analysis has been checked with atomic spectral simulations.

Impurity transport during neutral beam injection in the ISX-B tokamak

International Nuclear Information System (INIS)

Isler, R.C.; Crume, E.C.; Arnurius, D.E.; Murray, L.E.

1980-10-01

In ohmically heated ISX-B discharges, both the intrinsic iron impurity ions and small amounts of argon introduced as a test gas accumulate at the center of the plasma. But during certain beam-heated discharges, it appears that this accumulation does not take place. These results may reflect the conclusion of Stacey and Sigmar that momentum transferred from the beams to the plasma can inhibit inward impurity transport

Theoretical Study of Radiation from a Broad Range of Impurity Ions for Magnetic Fusion Diagnostics

Energy Technology Data Exchange (ETDEWEB)

Safronova, Alla [Univ. of Nevada, Reno, NV (United States)

2014-03-14

Spectroscopy of radiation emitted by impurities plays an important role in the study of magnetically confined fusion plasmas. The measurements of these impurities are crucial for the control of the general machine conditions, for the monitoring of the impurity levels, and for the detection of various possible fault conditions. Low-Z impurities, typically present in concentrations of 1%, are lithium, beryllium, boron, carbon, and oxygen. Some of the common medium-Z impurities are metals such as iron, nickel, and copper, and high-Z impurities, such as tungsten, are present in smaller concentrations of 0.1% or less. Despite the relatively small concentration numbers, the aforementioned impurities might make a substantial contribution to radiated power, and also influence both plasma conditions and instruments. A detailed theoretical study of line radiation from impurities that covers a very broad spectral range from less than 1 Å to more than 1000 Å has been accomplished and the results were applied to the LLNL Electron Beam Ion Trap (EBIT) and the Sustained Spheromak Physics Experiment (SSPX) and to the National Spherical Torus Experiment (NSTX) at Princeton. Though low- and medium-Z impurities were also studied, the main emphasis was made on the comprehensive theoretical study of radiation from tungsten using different state-of-the-art atomic structure codes such as Relativistic Many-Body Perturbation Theory (RMBPT). The important component of this research was a comparison of the results from the RMBPT code with other codes such as the Multiconfigurational Hartree–Fock developed by Cowan (COWAN code) and the Multiconfiguration Relativistic Hebrew University Lawrence Atomic Code (HULLAC code), and estimation of accuracy of calculations. We also have studied dielectronic recombination, an important recombination process for fusion plasma, for variety of highly and low charged tungsten ions using COWAN and HULLAC codes. Accurate DR rate coefficients are needed for

Compact toroid fueling of the TdeV tokamak

International Nuclear Information System (INIS)

Martin, F.; Raman, R.; Xiao, C.; Thomas, J.

1993-01-01

Compact toroids have been proposed as a means of centrally fueling tokamak reactors because of the high velocity to which they can be accelerated. These are cold (T e ∼ 10 eV), high density (n e > 10 20 m -3 ) spheromak plasmoids that are accelerated in a magnetized Marshall gun. As a proof of principle experiment, a compact toroid fueler (CTF) has been developed for injection into the TdeV tokamak. The engineering goals of the experiment are to measure and minimize the impurity content of the CT plasma and the neutral gas remaining after CT formation. Also of importance is the effect of CT central fueling on the tokamak density profile and bootstrap current, and the relaxation rate of the density profile providing information on the confinement time of the CT fuel

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)

Studies of runaway electrons via Cherenkov effect in tokamaks

Science.gov (United States)

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

2018-01-01

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

Wavelength calibration of x-ray imaging crystal spectrometer on Joint Texas Experimental Tokamak

International Nuclear Information System (INIS)

Yan, W.; Chen, Z. Y.; Jin, W.; Huang, D. W.; Ding, Y. H.; Li, J. C.; Zhang, X. Q.; Zhuang, G.; Lee, S. G.; Shi, Y. J.

2014-01-01

The wavelength calibration of x-ray imaging crystal spectrometer is a key issue for the measurements of plasma rotation. For the lack of available standard radiation source near 3.95 Å and there is no other diagnostics to measure the core rotation for inter-calibration, an indirect method by using tokamak plasma itself has been applied on joint Texas experimental tokamak. It is found that the core toroidal rotation velocity is not zero during locked mode phase. This is consistent with the observation of small oscillations on soft x-ray signals and electron cyclotron emission during locked-mode phase

The ICRH tokamak fusion test reactor

International Nuclear Information System (INIS)

Perkins, F.W.

1976-01-01

A Tokamak Fusion Test Reactor where the ion are maintained at Tsub(i) approximately 20keV>Tsub(e) approximately 7keV by ion-cyclotron resonance heating is shown to produce an energy amplification of Q>2 provided the principal ion energy loss channel is via collisional transfer to the electrons. Such a reactor produces 19MW of fusion power to the electrons. Such a reactor produces 19MW of fusion power and requires a 50MHz radio-frequency generator capable of 50MW peak power; it is otherwise compatible with the conceptual design for the Princeton TFTR. The required n tausub(E) values for electrons and ions are respectively ntausub(Ee)>1.5.10 13 cm -3 -sec and ntausub(Ei)>4.10 13 cm -3 -sec. The principal areas where research is needed to establish this concept are: tokamak transport calculations, ICRH physics, trapped-particle instability energy losses, tokamak equilibria with high values of βsub(theta), and, of course, impurities

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)

Experimental investigation of solid hydrogen pellet ablation in high-temperature plasmas using holographic interferometry and other diagnostics

International Nuclear Information System (INIS)

Thomas, C.E. Jr.

1981-03-01

The technology currently most favored for the refueling of fusion reactors is the high-velocity injection of solid hydrogen pellets. Design details are presented for a holographic interferometer/shadowgraph used to study the microscopic characteristics of a solid hydrogen pellet ablating in an approx. 1-keV plasma. Experimental data are presented for two sets of experiments in which the interferometer/shadowgraph was used to study approx. 1-mm-diam solid hydrogen pellets injected into the Impurity Study Experiment (ISX-B) tokamak at Oak Ridge National Laboratory (ORNL) at velocities of 1000 m/s. In addition to the use of the holographic interferometer, the pellet ablation process is diagnosed by studying the emission of Balmer-alpha photons and by using the available tokamak diagnostics

Impurity dynamics in stellarator W7-AS plasmas

International Nuclear Information System (INIS)

Igitkhanov, Yuri; Beidler, Craig D.; Burhenn, Reiner; Polunovsky, Eduard; Yamazaki, Kozo

2006-01-01

Numerical efforts to understand the neoclassical transport of impurities in stellarator plasmas have been undertaken. The new code solves the radial continuity equations for each ionization stage of the impurity ions for given background plasma profiles and magnetic configuration. An analytic description of the neoclassical transport coefficients based on numerical results from the DKES (Drift Kinetic Equation Solver) code and monoenergetic Monte-Carlo calculation (C.D. Beidler et al., EPS 1994), is here applied for impurity transport coefficients. The transition between the different charge states due to the ionization and recombination in balance equation is described by using the ADAS (Atomic Data and Analysis Structure) database. The impurity behavior in some typical discharges from W7-AS with moderate (NC) and improved energy confinement (HDH) has been considered. It is shown that the spatial distribution results from the competition between the radial electric field and the thermal force (which together produce a convective flux), and the diffusive term, which flattens the radial impurity distribution. The impurity ions are localized at the radial position where the convective flux goes through zero. It is also shown that for typical stellarator discharges there is no pronounced temperature screening effect as in tokamak plasmas. (author)

Surface impurity removal from DIII-D graphite tiles by boron carbide grit blasting

International Nuclear Information System (INIS)

Lee, R.L.; Hollerbach, M.A.; Holtrop, K.L.; Kellman, A.G.; Taylor, P.L.; West, W.P.

1993-11-01

During the latter half of 1992, the DIII-D tokamak at General Atomics (GA) underwent several modifications of its interior. One of the major tasks involved the removal of accumulated metallic impurities from the surface of the graphite tiles used to line the plasma facing surfaces inside of the tokamak. Approximately 1500 graphite tiles and 100 boron nitride tiles from the tokamak were cleaned to remove the metallic impurities. The cleaning process consisted of several steps: the removed graphite tiles were permanently marked, surface blasted using boron carbide (B 4 C) grit media (approximately 37 μm. diam.), ultrasonically cleaned in ethanol to remove loose dust, and outgassed at 1000 degrees C. Tests were done using, graphite samples and different grit blaster settings to determine the optimum propellant and abrasive media pressures to remove a graphite layer approximately 40-50 μm deep and yet produce a reasonably smooth finish. EDX measurements revealed that the blasting technique reduced the surface Ni, Cr, and Fe impurity levels to those of virgin graphite. In addition to the surface impurity removal, tritium monitoring was performed throughout the cleaning process. A bubbler system was set up to monitor the tritium level in the exhaust gas from the grit blaster unit. Surface wipes were also performed on over 10% of the tiles. Typical surface tritium concentrations of the tiles were reduced from about 500 dpm/100 cm 2 to less than 80 dpm/100 cm 2 following the cleaning. This tile conditioning, and the installation of additional graphite tiles to cover a high fraction of the metallic plasma facing surfaces, has substantially reduced metallic impurities in the plasma discharges which has allowed rapid recovery from a seven-month machine opening and regimes of enhanced plasma energy confinement to be more readily obtained. Safety issues concerning blaster operator exposure to carcinogenic metals and radioactive tritium will also be addressed

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

DEFF Research Database (Denmark)

Ondac, Peter; Horacek, Jan; Seidl, Jakub

2015-01-01

In this article we have used the 2D fluid turbulence numerical model, ESEL, to simulate turbulent transport in edge tokamak plasma. Basic plasma parameters from the ASDEX Upgrade and COMPASS tokamaks are used as input for the model, and the output is compared with experimental observations obtain...... for an extension of the ESEL model from 2D to 3D to fully resolve the parallel dynamics, and the coupling from the plasma to the sheath....

The measurement of the intrinsic impurities of molybdenum and carbon in the Alcator C-Mod tokamak plasma using low resolution spectroscopy

International Nuclear Information System (INIS)

May, M.J.; Finkenthal, M.; Regan, S.P.

1997-01-01

The intrinsic impurity content of molybdenum and carbon was measured in the Alcator C-Mod tokamak using low resolution, multilayer mirror (MLM) spectroscopy (Δλ ∼ 1-10 A). Molybdenum was the dominant high-Z impurity and originated from the molybdenum armour tiles covering all the plasma facing surfaces (including the inner column, the poloidal divertor plates and the ion cyclotron resonant frequency (ICRF) limiter) at Alcator C-Mod. Soft X ray extreme ultraviolet (XUV) emission, lines of charge states, ranging from hydrogen-like to helium-like lines of carbon (radius/minor radius, r/a ∼ 1) at the plasma edge to potassium- to chlorine-like (0.4 eff value, and the power losses through line radiation were estimated. For the diverted ohmically heated plasma examined, the intrinsic molybdenum and carbon concentrations in the core plasma were found to be ∼ 1.2 x 10 10 and ∼ 1.7 x 10 12 cm -3 , respectively. These measurements were obtained before the plasma facing components were boronized. The calculated radiated power from molybdenum was 170 kW; for carbon it was 45 kW. The contribution to the measured Z eff - 1 value of ∼ 0.8 was ∼ 0.11 for molybdenum and ∼ 0.5 for carbon. (author). 36 refs, 11 figs, 3 tabs

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

Source effects on impurity and heat transport in a tokamak

International Nuclear Information System (INIS)

Bennett, R.B.

1980-12-01

A recently developed generalization of neoclassical theory is extended here to study heat flux contributions to impurity transport, as well as the heat fluxes themselves. The theory accounts for the first four source moments, with external drags, which has been studied previously with either fewer moments or restricted to a collisional plasma. Conditions are established for which a momentum source may be used to modify the particle and heat transport. In the course of this work, the particle and heat transport is evaluated for a two species plasma with arbitrary plasma geometry, beta, and collisionality

Study of the electron heat transport in Tore-Supra tokamak; Etude du transport de la chaleur electronique dans le Tokamak Tore Supra

Energy Technology Data Exchange (ETDEWEB)

Harauchamps, E

2004-07-01

This work presents analytical solutions to the electron heat transport equation involving a damping term and a convection term in a cylindrical geometry. These solutions, processed by Matlab, allow the determination of the evolution of the radial profile of electron temperature in tokamaks during heating. The modulated injection of waves around the electron cyclotron frequency is an efficient tool to study heat transport experimentally in tokamaks. The comparison of these analytical solutions with experimental results from Tore-Supra during 2 discharges (30550 and 31165) shows the presence of a sudden change for the diffusion and damping coefficients. The hypothesis of the presence of a pinch spread all along the plasma might explain the shape of the experimental temperature profiles. These analytical solutions could be used to determine the time evolution of plasma density as well or of any parameter whose evolution is governed by a diffusion-convection equation. (A.C.)

Radiation losses from oxygen and iron impurities in a high temperature plasma

International Nuclear Information System (INIS)

Breton, C.; Michelis, C. de; Mattioli, M.

1976-06-01

Radiation and ionization losses due to impurities present in a high temperature plasma have been calculated for a light element (oxygen), which is completely stripped in the core of existing Tokamak discharges, and a heavy one (iron), which is only partially stripped. Two extreme cases have been treated: in the first one coronal equilibrium is reached; the radiated power is then equal to the product of the electron density, the impurity density, and a function of the electron temperature; in the second one impurities recycle with a constant radial velocity v 0 in a background plasma; radiation and ionization losses are proportional to the impurity flux and are a decreasing function of the diffusion velocity. The results presented can be used to evaluate losses in a practical case [fr

Impurity ion transport studies on the PLT tokamak during neutral-beam injection

International Nuclear Information System (INIS)

Suckewer, S.; Cavallo, A.; Cohen, S.

1984-01-01

Radial transport of medium- and high-Z ions during co- and counter-neutral-beam heating in the PLT tokamak is studied, using molybdenum and scandium ions as tracer elements. The time evolution of the radial profiles of several ionization stages of both elements, injected by laser blowoff during the neutral-beam heating, is measured under three significantly different beam-plasma combinations. No noticeable differences in the radial profiles attributable to the beam direction are observed. However, a given injected amount resulted in considerably larger interior concentrations of the tracer element in the counter-beam heating cases, suggesting larger penetration of the plasma periphery. Computer simulation with the MIST code suggests a net inward drift of the order 10 3 cm.s -1 superposed to a diffusion coefficient of the order 10 4 cm 2 .s -1 for both scandium and molybdenum ions. Injection of larger amounts of the tracer element, sufficient to cause measurable central electron temperature changes, resulted in dramatic changes in ion-state distributions, making some appear peaked in the centre while others disappeared. This effect could be produced with both co- and counter-beam heating, but with lesser amounts in the latter case. It is interpreted as rearrangement of the ionization balance, rather than any preferential accumulation of the injected element. (author)

Tokamak experimental power reactor

International Nuclear Information System (INIS)

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

1976-01-01

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

Edge and Core Impurity Transport Study with Spectroscopic Instruments in LHD

International Nuclear Information System (INIS)

Morita, Shigeru; Goto, Motoshi; Kobayashi, Masahiro; Muto, Sadatsugu; Chowdhuri, Malay Bikas; Chunfeng, Dong; Hangyu, Zhou; Zhengying, Cui; Fujii, Keisuke; Hasuo, Masahiro; Iwamae, Atsushi; Furuzawa, Akihiro; Sakurai, Ikuya; Tawara, Yuzuru; Yinxian, Jie; Baonian, Wan; Zhenwei, Wu; Koubiti, Mohammed; Yamaguchi, Naohiro

2009-01-01

Impurity transport was investigated at both edge and core regions in large helical device (LHD) with developed spectroscopic instruments which can measure one- and two-dimensional distributions of impurities. The edge impurity behavior was studied recently using four carbon resonant transitions in different ionization stages of CIII (977A), CIV (1548A), CV (40.3A) and CVI (33.7A). When the line-averaged electron density, n e , is increased from 1 to 6 x 10 13 cm -3 , the ratio of (CIII+CIV)/n e increases while the ratio of (CV+CVI)/n e decreases. Here, CIII+CIV (CV+CVI) expresses the sum of CIII (CV) and CIV (CVI) intensities. The CIII+CIV indicates the carbon influx and the CV+CVI indicates the emissions through the transport in the ergodic layer. The result thus gives experimental evidence on the impurity screening by the ergodic layer in LHD, which is also supported by a three-dimensional edge particle simulation. The core impurity behavior is also studied in high-density discharges (n e ≤ 1x 10 15 cm -3 ) with multi H 2 -pellets injection. It is found that the ratio of V/D (V: convection velocity, D: diffusion coefficient) decreases after pellet injection and Z eff profile shows a flat one at values of 1.1-1.2. These results confirm no impurity accumulation occurs in high-density discharges. As a result, the iron density, n Fe , is analyzed to be 6 x 10 -7 ( = n Fe /n e ) of which the amount can be negligible as radiation source even in such high-density discharges. One- and two-dimensional impurity distributions from space-resolved VUV and EUV spectrometers newly developed for further impurity transport study are also presented with their preliminary results. (magnetically confined plasma)

Moessbauer Studies of Implanted Impurities in Solids

CERN Multimedia

2002-01-01

Moessbauer studies were performed on implanted radioactive impurities in semiconductors and metals. Radioactive isotopes (from the ISOLDE facility) decaying to a Moessbauer isotope were utilized to investigate electronic and vibrational properties of impurities and impurity-defect structures. This information is inferred from the measured impurity hyperfine interactions and Debye-Waller factor. In semiconductors isoelectronic, shallow and deep level impurities have been implanted. Complex impurity defects have been produced by the implantation process (correlated damage) or by recoil effects from the nuclear decay in both semiconductors and metals. Annealing mechanisms of the defects have been studied. \\\\ \\\\ In silicon amorphised implanted layers have been recrystallized epitaxially by rapid-thermal-annealing techniques yielding highly supersaturated, electrically-active donor concentrations. Their dissolution and migration mechanisms have been investigated in detail. The electronic configuration of Sb donors...

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)

Neutron activation of microspheres containing 165Ho: theoretical and experimental radionuclidic impurities study

International Nuclear Information System (INIS)

Squair, Peterson L.; Pozzo, Lorena; Ivanov, Evandro; Osso Junior, Joao A.

2011-01-01

The 166 Ho microspheres are potentially interesting for medical applications for treatment of many tumors. The internal radionuclide therapy can use polymer or glass device that provides structural support for the radionuclide. After activation, beta minus emission of 166 Ho (T 1/2 =26.8h, β - E max =1.84 MeV, γ E p =80.6 keV) can be used for therapeutic purposes. The aim of this work is study the influence of radionuclide impurities between End of Bombardment (EOB) and the medical application. The appropriate specific activities and purity along decay should be adequate for their safe and efficient medical applications. The good practices on neutron activation techniques are choice a high purity target to avoid production of undesirable radionuclides and when possible with enriched targets to obtain higher specific activity. In this work the target used was Ho 2 O 3 and polymeric microspheres containing holmium acetylacetonate (HoAcAc) manufactured at the Biotechnology Center-IPEN/CNEN-SP. Three conditions were evaluated: preliminary test with 1.0x10 13 n.cm -2 s -1 for 1.0 hour; nowadays maximum capability of IEA-R1 reactor with 5.0x10 13 n.cm -2 s -1 for 64.0 hours and the ideal IEA-R1 operation with 5.0x10 13 n.cm -2 s -1 for 120.0 hours. Considering the sample with 99.9% 165 Ho purity and 0.1% for each impurities elements with its natural abundance, the highest radionuclidic impurity is the Lutetium followed by Ytterbium, Lanthanum and Cerium. The intrinsic radionuclidic impurity of 166 mHo is less relevant. This review is important to identify the radionuclidic purity characteristics of the preliminary studies with different time and flux irradiation. The data produced in this paper will help to define strategies for the production of 166 Ho radioisotope at IEA-R1 IPEN/CNEN-SP reactor. (author)

Probing neutral density at the plasma edge of Tore Supra with CX excited impurity ions

International Nuclear Information System (INIS)

Hess, W.R.; Mattioli, M.; Guirlet, R.

1993-01-01

In Tokamak plasma physics renewed interest in visible spectroscopy has grown for two reasons. The use of fiber optics allows observation of local sources of both impurities and of hydrogen by observing radiation of low ionization states. Moreover, charge exchange spectroscopy (CXS) with either auxiliary or heating neutral beams is a standard technique to determine the ion temperature and impurity density profiles. After a short description of the experimental setup and the ergodic divertor of Tore Supra (TS), two discharges in which space-resolved observations of the CVI (8-7) line clearly show the presence of CX-related effects. A well isolated spectral line at 5304.6 A is discussed. Tentative identification as CIII (1s 2 2s, 7-5) is suggested. The conclusion shows the usefulness of the reported results for probing neutral density at the plasma edge by detecting CX excited impurity ions and that highly ionized C 6+ ions exist in the MARFE regions. To the best of our knowledge, only very low ionization C and O ions (such as CIII or OIV) have been previously reported in these regions

High temperature outgassing tests on materials used in the DIII-D tokamak

International Nuclear Information System (INIS)

Holtrop, K.L.; Hansink, M.J.

2006-01-01

This article is a continuation of previous work on determining the outgassing characteristics of materials used in the DIII-D magnetic fusion tokamak [K. L. Holtrop, J. Vac. Sci. Technol. A 17, 2064 (1999)]. Achievement of high performance plasma discharges in the DIII-D tokamak requires careful control of impurity levels. Among the techniques used to control impurities are routine bakes of the vacuum vessel to an average temperature of 350 deg. C. Materials used in DIII-D must release only very small amounts of impurities (below 2x10 -6 mole) at this temperature that could be transferred to the first wall materials and later contaminate plasma discharges. To better study the behavior of materials proposed for use in DIII-D at elevated temperatures, the initial outgassing test chamber was improved to include an independent heating control of the sample and a simple load lock chamber. The goal was to determine not only the total degassing rate of the material during baking, but to also determine the gas species composition and to obtain a quantitative estimate of the degassing rate of each species by the use of a residual gas analyzer. Initial results for aluminum anodized using three different processes, stainless steel plated with black oxide and black chrome, and a commercially available fiber optic feedthrough will be presented

Oak Ridge Tokamak experimental power reactor study reference design

International Nuclear Information System (INIS)

Roberts, M.; Bettis, E.S.

1975-11-01

A Tokamak EPR Reference Design is presented as a basis for further design study leading to a Conceptual Design. The set of basic plasma parameters selected--minor radius of 2.25 m, major radius of 6.75 m, magnetic field on axis of 4.8 T and plasma current of 7.2 MA--should produce a reactor-grade plasma with a significant neutron flux, even with the great uncertainty in plasma physics scaling from present experience to large sizes. Neutronics and heat transfer calculations coupled with mechanical design and materials considerations were used to develop a blanket and shield capable of operating at high temperature, protecting the surrounding coils, being maintained remotely and, in a few experimental modules, breeding tritium. Nb 3 Sn and NbTi superconductors are used in the toroidal field coil design. The coil system was developed for a maximum field of 11 T at the winding (to give a field on axis of 4.8 T), and combines multifilamentary superconducting cable with forced flow of supercritical helium enclosed in a steel conduit. The structural system uses a stainless steel center bucking ring and intercoil box beam bracing to provide rigid support for coils against the centering force, overturning moments from poloidal fields and faults, other external forces, and thermal stresses. The poloidal magnetics system is specially designed both to reduce the total volt-second energy requirements and to reduce the magnitude of the rate of field change at the toroidal field coils. The rate of field change imposed upon the toroidal field coils is reduced by at least a factor of 3.3 compared to that due to the plasma alone. Tritium processing, tritium containment and vacuum systems employ double containment and atmospheric cleanup to minimize releases. The document also contains discussions of systems integration and assembly, key research and development needs, and schedule considerations

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

ARIES tokamak reactor study

International Nuclear Information System (INIS)

Steiner, D.; Embrechts, M.

1990-07-01

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

Tokamak and RFP ignition requirements

International Nuclear Information System (INIS)

Werley, K.A.

1991-01-01

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

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

Soft x-ray measurements on the PLT Tokamak

International Nuclear Information System (INIS)

Von Goeler, S.; Sauthoff, N.; Bitter, M.

1977-10-01

Four experiments are described that currently run on the PLT tokamak and which utilize the soft x-ray emission of the plasma as a diagnostic: the pulse height analysis system for temperature and impurity measurements; the curved crystal Bragg spectrometer for the determination of ionization states of impurities; ''windowless'' surface barrier detectors for the investigation of the ultra soft x-ray radiation in the energy range 0.1 keV < hν < 1 keV and a silicon diode array for x-ray fluctuation measurements. For each diagnostic a short technical description and some recent results obtained with it on PLT are given in order to demonstrate its use

Experimental investigation of turbulent transport at the edge of a tokamak plasma

International Nuclear Information System (INIS)

Fedorczak, N.

2010-01-01

This manuscript is devoted to the experimental investigation of particle transport in the edge region of the tokamak Tore Supra. The first part introduces the motivations linked to energy production, the principle of a magnetic confinement and the elements of physics essential to describe the dynamic of the plasma at the edge region. From data collected by a set of Langmuir probes and a fast visible imaging camera, we demonstrate that the particle transport is dominated by the convection of plasma filaments, structures elongated along magnetic field lines. They present a finite wave number, responsible for the high enhancement of the particle flux at the low field side of the tokamak. This leads to the generation of strong parallel flows, and the strong constraint of filament geometry by the magnetic shear. (author)

Plasma density remote control system of experimental advanced superconductive tokamak

International Nuclear Information System (INIS)

Zhang Mingxin; Luo Jiarong; Li Guiming; Wang Hua; Zhao Dazheng; Xu Congdong

2007-01-01

In Tokamak experiments, experimental data and information on the density control are stored in the local computer system. Therefore, the researchers have to be in the control room for getting the data. Plasma Density Remote Control System (DRCS), which is implemented by encapsulating the business logic on the client in the B/S module, conducts the complicated science computation and realizes the synchronization with the experimental process on the client. At the same time, Web Services and Data File Services are deployed for the data exchange. It is proved in the experiments that DRCS not only meets the requirements for the remote control, but also shows an enhanced capability on the data transmission. (authors)

Demixing of impurities and hydrogen as deduced from Zeff profiles in the boronized ASDEX

International Nuclear Information System (INIS)

Steuer, K.H.; Roehr, H.; Engelhardt, W.; Fussmann, G.; Kallenbach, A.; Kurzan, B.; Murmann, H.D.

1990-01-01

Substantial progress towards fusion has been made in the confinement, stability and heating of tokamak plasmas. The transport behaviour of magnetically confined plasmas, however, is still an unsolved problem. The transport mechanisms of hydrogen and of impurities are known to be different, leading to phenomena such as impurity accumulation on axis, especially in good confinement regimes, and more generally to demixing of the various species. Besides energy losses from impurity radiation, one has to be concerned about dilution of the fuel-ion density, and about effects that impurities may have on the main ion and electron transport. To understand the transport behaviour of the different plasma species, one needs their spatial density profiles. It is convenient to represent the various impurities by a characteristic impurity ion with a density n z and a fictive charge Z (Z = Σn i Z i 2 / Σn i Z i ; i ≥ 2). A typical value for Z is 7, indicating that light impurities are dominating. Comparing the different profiles, we find characteristic differences in the electron, proton and impurity transport behaviour. (orig./AH)

Stability of high-beta tokamak equilibria and transport in Belt-Pinch IIa

Energy Technology Data Exchange (ETDEWEB)

Becker, G; Gruber, O; Krause, H; Mast, F; Wilhelm, R [Association Euratom-Max-Planck-Institut fuer Plasmaphysik, Garching (Germany, F.R.)

1978-01-01

In Belt-Pinch IIa, highly elongated equilibria with poloidal beta values up to the aspect ratio have been achieved. In these tokamak-like configurations, no fast-growing MHD instabilities such as external kink and ballooning modes have been observed. Rigid displacement instabilities have been stabilized by an appropriate poloidal magnetic field configuration and by a conducting shell. By comparing simulation experiments using the Garching high-beta transport code with measurements, it has been found that in the collision-dominated plasma no anomalously enhanced transport occurs. Transport theory in the Pfirsch-Schlueter regime, which includes elongation and high-beta effects, has been confirmed by the experiment. In particular, it has been shown that the perpendicular electrical conductivity is also classical. Detailed investigations of oxygen and carbon impurity losses demonstrated that the impurity subprograms commonly used for tokamaks underestimate the radiation losses in the range Tsub(e)=10 to 30 eV.

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

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

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.)

Study on assembly techniques and procedures for ITER tokamak device

International Nuclear Information System (INIS)

Obara, Kenjiro; Kakudate, Satoshi; Shibanuma, Kiyoshi; Sago, Hiromi; Ue, Koichi; Shimizu, Katsusuke; Onozuka, Masanori

2006-06-01

The International Thermonuclear Experimental Reactor (ITER) tokamak is mainly composed of a doughnut-shaped vacuum vessel (VV), four types of superconducting coils such as toroidal field coils (TF coils) arranged around the VV, and in-vessel components, such as blanket and divertor. The dimensions and weight of the respective components are around a few ten-meters and several hundred-tons. In addition, the whole tokamak assembly, which are composed of these components, are roughly estimated, 26 m in diameter, 18 m in height and over 16,500 tons in total weight. On the other hand, as for positioning and assembly tolerances of the VV and the TF coil are required to be a high accuracy of ±3 mm in spite of large size and heavy weight. The assembly procedures and techniques of the ITER tokamak are therefore studied, taking account of the tolerance requirements as well as the configuration of the tokamak with large size and heavy weight. Based on the above backgrounds, the assembly procedures and techniques, which are able to assemble the tokamak with high accuracy, are described in the present report. The tokamak assembly operations are categorized into six work break down structures (WBS), i.e., (1) preparation for assembly operations, (2) sub-assembly of the 40deg sector composed of 40deg VV sector, two TF coils and thermal shield between VV and TF coil at the assembly hall, (3) completion of the doughnut-shaped tokamak assembly composed of nine 40deg sectors in the cryostat at the tokamak pit, (4) measurement of positioning and accuracy after the completion of the tokamak assembly, (5) installation of the ex-vessel components, and (6) installation of in-vessel components. In the present report, two assembly operations of (2) and (3) in the above six WBS, which are the most critical in the tokamak assembly, are mainly described. The report describes the following newly developed tokamak assembly procedures and techniques, jigs and tools for assembly and metrology

Fundamental studies of fusion plasmas

International Nuclear Information System (INIS)

Aamodt, R.E.; Catto, P.J.; D'Ippolito, D.A.; Myra, J.R.; Russell, D.A.

1990-03-01

This paper discusses tokamak transport, auxiliary heating physics; ICRF impurity study; ponderomotive stabilization studies; ICRF induced fluxes in the edge plasma; runaway electron confinement in TEXT; rf sheath modelling for ICRF antenna Faraday screens; and isotropic energetic in fluxes in tokamaks

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

Turbulent transport of impurities in a magnetized plasma; Transport turbulent d'impuretes dans un plasma magnetise

Energy Technology Data Exchange (ETDEWEB)

Dubuit, N

2006-10-15

This work deals with the transport of impurities in magnetically confined thermonuclear plasmas. The accumulation of impurities in the core of the plasma would imply dramatic losses of energy that may lead to the extinction of the plasma. On the opposite, the injection of impurities in the plasma edge is considered as an efficient means to extract heat without damaging the first wall. The balance between these 2 contradictory constraints requires an accurate knowledge of the impurity transport inside the plasma. The effect of turbulence, the main transport mechanism for impurities is therefore a major issue. In this work, the complete formula of a turbulent flow of impurities for a given fluctuation spectrum has been inferred. The origin and features of the main accumulation processes have been identified. The main effect comes from the compressibility of the electrical shift speed in a plane perpendicular to the magnetic field. This compressibility appears to be linked to the curvature of the magnetic field. A less important effect is a thermal-diffusion process that is inversely proportional to the number of charges and then disappears for most type of impurities except the lightest. This effect implies an impurity flux proportional to the temperature gradient and its direction can change according to the average speed of fluctuations. A new version of the turbulence code TRB has been developed. This new version allows the constraints of the turbulence not by the gradients but by the flux which is more realistic. The importance of the processes described above has been confirmed by a comparison between calculation and experimental data from Tore-supra and the Jet tokamak. The prevailing role of the curvature of the magnetic field in the transport impurity is highlighted. (A.C.)

Internal disruptions in tokamaks

International Nuclear Information System (INIS)

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

1990-01-01

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

Annual report of Division of Thermonuclear Fusion Research and Division of Large Tokamak Development for the period of April 1, 1976 to March 31, 1977

International Nuclear Information System (INIS)

1978-02-01

Research and development activities in the two divisions are closely related. 1) Theoretical and computational studies continued on tokamak confinement and heating related to experimental problems. Studies on NBI heating in JT-60 were completed. 2) Experimental studies on impurities, density control and effects of density fluctuations were made in JFT-2. Neutral beams up to 30 keV and 8 A were injected into JFT-2 plasma perpendicularly. The ion temperature was increased by 10% - 15%, which is in agreement with the prediction by classical Fokker-Planck theory. In JFT-2a(DIVA), plasma-wall interaction (behavior of heavy and light impurities) was studies. The divertor of DIVA reduced the plasma-wall interaction and hence the radiation loss due to heavy impurities by a factor of 3. A grazing-incidence vacuum monochromator was first used in impurity studies in JFT-2 and JFT-2a. 3) Technological improvements were made raising efficiencies of operation, maintenance and plasma research. 4) Neutral beam injector test stand ITS-2 of 100 keV was completed. Construction of a 200 kW, 650 MHz radiofrequency heating system for JFT-2 was started. 5) Sputterings of molybdenum and pyrolytic graphite by low-energy protons and chemical reaction rates of pyrolytic graphite with protons were measured. Honeycomb structure greatly reduced the sputtered particles. 6) The superconducting magnet development group made the design of cluster test apparatus and the development of large current superconductor. 7) Phase-I preliminary design of experimental fusion reactor JXFR was completed and preliminary safety evaluation of JXFR was made. 8) Detailed design of JT-60 was completed in November 1976. Engineering development contracts were all completed by March 1977. 9) Engineering studies and tests on critical components of JT-4 with non-circular plasma cross section and divertors were made, after the preliminary design in fiscal year 1975. (auth.)

Fusion Studies Program. Progress report

International Nuclear Information System (INIS)

Stacey, W.M. Jr.

1984-01-01

Continuation of work in two areas, impurity control and transient electromagnetics, is proposed. In the tokamak impurity control area, an innovative supplemental mechanism, NB-driven impurity flow reversal, has been developed partly under this contract; and the proposed effort is aimed at completing this development, verifying the methodology by comparison with experiment and evaluating its potential in future tokamak experiments. In the tokamak transient electromagnetics area, the proposed effort is aimed at developing a new and more efficient methodology for calculating the currents and resulting magnetic fields in the torus structure and coil systems, which will allow a detailed representation of the latter that can be coupled to the distributed-current plasma model that was implemented for vertical stability and disruption control studies in previous work under this contract; and the application of this methodology to study the control of vertical instabilities and disruptions

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.

LHCD experiments on tokamak CASTOR

International Nuclear Information System (INIS)

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

1990-10-01

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

Conceptual studies of toroidal field magnets for the tokamak (fusion) experimental power reactor. Final report

International Nuclear Information System (INIS)

1976-01-01

This report presents the results of ''Conceptual Studies of Toroidal Field Magnets for the Tokamak Experimental Power Reactor'' performed for the Energy Research and Development Administration, Oak Ridge Operations. Two conceptual coil designs are developed. One design approach to produce a specified 8 Tesla maximum field uses a novel NbTi superconductor design cooled by pool-boiling liquid helium. For a highest practicable field design, a unique NbSn 3 conductor is used with forced-flow, single-phase liquid helium cooling to achieve a 12 Tesla peak field. Fabrication requirements are also developed for these approximately 7 meter horizontal bore by 11 meter vertical bore coils. Cryostat design approaches are analyzed and a hybrid cryostat approach selected. Structural analyses are performed for approaches to support in-plane and out-of-plane loads and a structural approach selected. In addition to the conceptual design studies, cost estimates and schedules are prepared for each of the design approaches, major uncertainties and recommendations for research and development identified, and test coil size for demonstration recommended

The impact of the biasing radial electric field on the SOL in a divertor tokamak

International Nuclear Information System (INIS)

Rozhansky, V.; Tendler, M.

1993-01-01

Strong radial electric field can be induced within the SOL in a divertor tokamak by applying a voltage to divertor plates with respect to the first wall. This biasing scheme results in the strong radial electric field which is much larger than the natural electric field, usually of the order T e /e. Experiments employing this biasing scheme were carried out on the tokamak TdeV. Many interesting effects such as - modifications of the density profile and radial transport of impurities as a function of the polarity and the magnitude of the biasing voltage, the generation of the flux surface average toroidal rotation proportional to the applied voltage, redistribution of the plasma outflow onto divertor plates and so on - were demonstrated to result from the biasing. Furthermore, in contrast to studies carried out employing a different biasing scheme which primarily results in a poloidal electric field, the strong radial electric field impacts more significantly within SOL than the poloidal electric field. Here, we aim to show that the main effects observed experimentally follow from the analysis, provided continuity and momentum balances are employed invoking anomalous viscosity and inertia. (author) 4 refs

Improved charge-coupled device detectors for high-speed, charge exchange spectroscopy studies on the DIII-D tokamak

International Nuclear Information System (INIS)

Burrell, K.H.; Gohil, P.; Groebner, R.J.; Kaplan, D.H.; Robinson, J.I.; Solomon, W.M.

2004-01-01

Charge exchange spectroscopy is one of the key ion diagnostics on the DIII-D tokamak. It allows determination of ion temperature, poloidal and toroidal velocity, impurity density, and radial electric field E r throughout the plasma. For the 2003 experimental campaign, we replaced the intensified photodiode array detectors on the central portion of the DIII-D charge exchange spectroscopy system with advanced charge-coupled device (CCD) detectors mounted on faster (f/4.7) Czerny-Turner spectrometers equipped with toroidal mirrors. The CCD detectors are improved versions of the ones installed on our edge system in 1999. The combination improved the photoelectron signal level by about a factor of 20 and the signal to noise by a factor of 2-8, depending on the absolute signal level. The new cameras also allow shorter minimum integration times while archiving to PC memory: 0.552 ms for the slower, lower-read noise (15 e) readout mode and 0.274 ms in the faster, higher-read noise (30 e) mode

Molybdenum emission from impurity-induced m= 1 snake-modes on the Alcator C-Mod tokamak

Energy Technology Data Exchange (ETDEWEB)

Delgado-Aparicio, L. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); MIT - Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Bitter, M.; Gates, D.; Hill, K.; Pablant, N. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Granetz, R.; Reinke, M.; Podpaly, Y.; Rice, J. [MIT - Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Beiersdorfer, P. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Sugiyama, L. [MIT - Laboratory for Nuclear Science, Cambridge, Massachusetts 02139 (United States)

2012-10-15

A suite of novel high-resolution spectroscopic imaging diagnostics has facilitated the identification and localization of molybdenum impurities as the main species during the formation and lifetime of m= 1 impurity-induced snake-modes on Alcator C-Mod. Such measurements made it possible to infer, for the first time, the perturbed radiated power density profiles from which the impurity density can be deduced.

Self-limitation of impurity production by radiation cooling at the edge of a fusion plasma

International Nuclear Information System (INIS)

Neuhauser, J.; Lackner, K.; Wunderlich, R.

1982-04-01

The influence of radiation cooling at the edge of a fusion plasma on the plasma-wall interaction is numerically studied for parameters typical of the ZEPHYR ignition experiment. Various transport and impurity influx models and different external heating methods are studied using the 1D tokamak transport code BALDUR developed at Princeton. The results demonstrate the self-consistent formation of a radiating boundary layer (photosphere) for a wide range of parameters, limiting the impurity concentration in the plasma to a tolerable value. While the plasma behaviour is rather insensitive to model assumptions, the sputtering rate and the corresponding wall erosion depend on various parameters. Methods for external control of the photosphere and - more important - of the wall erosion are also discussed. (orig.)

HTMR: an experimental tokamak reactor with hybrid copper/superconductor toroidal field magnet

International Nuclear Information System (INIS)

Avanzini, P.G.; Raia, G.; Rosatelli, F.; Zampaglione, V.

1985-01-01

The feasibility of a hybrid configuration superconducting coils/copper coils for a next generation tokamak TF magnet has been investigated. On the basis of this hybrid solution, the conceptual design has been developed for a medium-high toroidal field tokamak reactor (HTMR). The results of this study show the possibility of designing a tokamak reactor with reduced size in comparison with other INTOR like devices, still gaining some margins in front of the uncertainties in the scaling laws for plasma physics parameters and retaining the presence of a blanket with a tritium breeding ratio of about 1

First experimental results with the Current Limit Avoidance System at the JET tokamak

Energy Technology Data Exchange (ETDEWEB)

De Tommasi, G. [Associazione EURATOM-ENEA-CREATE, Università di Napoli Federico II, Via Claudio 21, 80125 Napoli (Italy); Galeani, S. [Dipartimento di Informatica, Sistemi e Produzione, Università di Roma, Tor Vergata, Rome (Italy); Jachmich, S. [Association EURATOM-Belgian State, Koninklijke Militaire School - Ecole Royale Militaire, B-1000 Brussels (Belgium); Joffrin, E. [IRFM-CEA, Centre de Cadarache, 13108 Saint-paul-lez-Durance (France); Lennholm, M. [EFDA Close Support Unit, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); European Commission, B-1049 Brussels (Belgium); Lomas, P.J. [Euratom-CCFE, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); Neto, A.C. [Associazione EURATOM-IST, Instituto de Plasmas e Fusao Nuclear, IST, 1049-001 Lisboa (Portugal); Maviglia, F. [Associazione EURATOM-ENEA-CREATE, Via Claudio 21, 80125 Napoli (Italy); McCullen, P. [Euratom-CCFE, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); Pironti, A. [Associazione EURATOM-ENEA-CREATE, Università di Napoli Federico II, Via Claudio 21, 80125 Napoli (Italy); Rimini, F.G. [Euratom-CCFE, Culham Science Centre, OX14 3DB Abingdon (United Kingdom); Sips, A.C.C. [European Commission, B-1049 Brussels (Belgium); Varano, G.; Vitelli, R. [Dipartimento di Informatica, Sistemi e Produzione, Università di Roma, Tor Vergata, Rome (Italy); Zaccarian, L. [CNRS, LAAS, 7 Avenue du Colonel Roche, F-31400 Toulouse (France); Universitè de Toulouse, LAAS, F-31400 Toulouse (France)

2013-06-15

The Current Limit Avoidance System (CLA) has been recently deployed at the JET tokamak to avoid current saturations in the poloidal field (PF) coils when the eXtreme Shape Controller is used to control the plasma shape. In order to cope with the current saturation limits, the CLA exploits the redundancy of the PF coils system to automatically obtain almost the same plasma shape using a different combination of currents in the PF coils. In the presence of disturbances it tries to avoid the current saturations by relaxing the constraints on the plasma shape control. The CLA system has been successfully implemented on the JET tokamak and fully commissioned in 2011. This paper presents the first experimental results achieved in 2011–2012 during the restart and the ITER-like wall campaigns at JET.

Tore-Supra: a Tokamak with superconducting toroidal field coils

International Nuclear Information System (INIS)

Turck, B.

1987-07-01

Tore Supra is a tokamak under construction on the site of Cen Cadarache by the Euratom-CEA Association. The machine technology integrates all problems related to the fabrication and the operation of large superconducting coils and of the associated cryogenic system. Tore Supra will provide a significant experience to prepare the next generation of machines for plasma physics and controlled fusion. Tore Supra is specially designed to implement a large physics program. The superconducting coils make possible the study of plasma confinement in long pulses (more than 60s), the impurities and the stability, and the efficiency of additional heating sources (neutral particle beams and radio frequency heating). The opportunity is taken to recall the particular features and requirements of the superconducting coils of the large future tokamaks in order to point out the problems that have to be faced by any new material (superconducting or not)

Heating of plasmas in tokamaks by current-driven turbulence

International Nuclear Information System (INIS)

Kluiver, H. de.

1985-10-01

Investigations of current-driven turbulence have shown the potential to heat plasmas to elevated temperatures in relatively small cross-section devices. The fundamental processes are rather well understood theoretically. Even as it is shown to be possible to relax the technical requirements on the necessary electric field and the pulse length to acceptable values, the effect of energy generation near the plasma edge, the energy transport, the impurity influx and the variation of the current profile are still unknown for present-day large-radius tokamaks. Heating of plasmas by quasi-stationary weakly turbulent states caused by moderate increases of the resistivity due to higher loop voltages could be envisaged. Power supplies able to furnish power levels 5-10 times higher than the usual values could be used for a demonstration of those regimes. At several institutes and university laboratories the study of turbulent heating in larger tokamaks and stellarators is pursued

Measurement of electron density profiles by soft X-ray tomography on the RTP tokamak

Energy Technology Data Exchange (ETDEWEB)

Cruz, D.F. da; Donne, A.J.H.; Lyadina, E.S.; Rutteman, R.H.; Tanzi, C.P. [FOM-Instituut voor Plasmafysica, Rijnhuizen (Netherlands)

1993-12-31

Tomographic diagnosis of the soft x-ray emissivity profile is a powerful method for studying several plasma parameters. The x-ray emissivity is a complicated function of plasma quantities like the electron density and temperature, and the impurity content in the plasma. These quantities can be studied separately provided that information is available on the remaining parameters. Soft x-ray emissivity profiles have already been used successfully in other machines to determine local values of impurity densities and the effective charge Z{sub eff}. In the RTP tokamak the electron density profile has been inferred from a modelling of the x-ray emissivity in situations where information is available on the electron temperature profile, the value of Z{sub eff}, and the relative proportion of the impurities. The method can be useful for the study of hollow density profiles that cannot be properly reconstructed by Abel inversion of interferometer or reflectometer data. (author) 7 refs., 2 figs.

Measurement of electron density profiles by soft X-ray tomography on the RTP tokamak

International Nuclear Information System (INIS)

Cruz, D.F. da; Donne, A.J.H.; Lyadina, E.S.; Rutteman, R.H.; Tanzi, C.P.

1993-01-01

Tomographic diagnosis of the soft x-ray emissivity profile is a powerful method for studying several plasma parameters. The x-ray emissivity is a complicated function of plasma quantities like the electron density and temperature, and the impurity content in the plasma. These quantities can be studied separately provided that information is available on the remaining parameters. Soft x-ray emissivity profiles have already been used successfully in other machines to determine local values of impurity densities and the effective charge Z eff . In the RTP tokamak the electron density profile has been inferred from a modelling of the x-ray emissivity in situations where information is available on the electron temperature profile, the value of Z eff , and the relative proportion of the impurities. The method can be useful for the study of hollow density profiles that cannot be properly reconstructed by Abel inversion of interferometer or reflectometer data. (author) 7 refs., 2 figs

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

A system to deposit boron films (boronization) in the DIII-D tokamak

International Nuclear Information System (INIS)

Hodapp, T.R.; Jackson, G.L.; Phillips, J.; Holtrop, K.L.; Peterson, P.L.; Winters, J.

1992-01-01

A system has been added to the DIII-D tokamak to coat its plasma facing surfaces with a film of boron using diborane gas. The system includes special health and safety equipment for handling the diborane gas which is toxic and inflammable. The purpose f the boron film is to reduce the levels of impurity atoms in the DIII-D plasmas. Experiments following the application of the boron film in DIII-D have led to significant reductions in plasma impurity levels and the observation of a new, very high confinement regime

Theory of ion-temperature-gradient-driven turbulence in tokamaks

International Nuclear Information System (INIS)

Lee, G.S.; Diamond, P.H.

1986-01-01

An analytic theory of ion-temperature-gradient-driven turbulence in tokamaks is presented. Energy-conserving, renormalized spectrum equations are derived and solved in order to obtain the spectra of stationary ion-temperature-gradient-driven turbulence. Corrections to mixing-length estimates are calculated explicitly. The resulting anomalous ion thermal diffusivity chi/sub i/ = 0.4[(π/2)ln(1 + eta/sub i/)] 2 [(1 + eta/sub i/)/tau] 2 rho/sub s/ 2 c/sub s//L/sub s/ is derived and is found to be consistent with experimentally-deduced thermal diffusivities. The associated electron thermal diffusivity and particle and heat-pinch velocities are also calculated. The effect of impurity gradients on saturated ion-temperature-gradient-driven turbulence is discussed and a related explanation of density profile steepening during Z-mode operation is proposed. 35 refs., 4 figs

Tokamak power system studies at ANL

International Nuclear Information System (INIS)

Baker, C.C.; Ehst, D.A.; Brooks, J.N.; Evans, K. Jr.

1986-06-01

The following features, in particular, have been examined: (a) large aspect ratio (A ≅ 6), which may ease maintenance; (b) high beta (β ≥ 0.20) without indentation, which brings the maximum toroidal field down to about 6 to 7 T; (c) low toroidal current (I ≅ 4MA), which reduces the cost of the current drive and equilibrium field system; and (d) steady state operation with current density control via fast and slow wave current drive. The key to high beta operation with low toroidal current lies in utilizing second stability regime equilibria with the required current distributions produced by an appropriate selection of wave driver frequencies and power spectra. The ray tracing and current drive calculation is self-consistent with the actual magnetic fields they produce in the plasma. The impurity control activities in TPSS have emphasized the self-pumping concept as applied to using the entire first wall or ''slot'' limiters. The blanket design effort has emphasized liquid metal and Flibe concepts. The reference concept is a liquid lithium/vanadium, self-cooled configuration. Overall, there exists a number of major design improvements which will substantially improve the attractiveness of tokamak reactors

Study on plasma visible radiation spatial distribution at the TO-1 tokamak

International Nuclear Information System (INIS)

Molotkov, L.I.; Shvindt, N.N.

1977-01-01

The results of spatial distribution measurements of radiation intensities of spectral lines of hydrogen and light plasma impurities in the visible-light spectrum TO-1 tokamak are described. The method of electrochemical scanning with the help of rotating disk with notches was used. The experiments were carried out in the stable regime of discharge and in the regime with breakdown instability. In the stable regime absolute intensities of the spectral lines were measured. The concentration of radiating atoms and ions were calculated using the measured absolute intensities. Besides in order to determine the region in which the discharge appears in the cross section of the TO-1 chamber the spatial distributions of Hsub(β) neutral hydrogen spectral line in the initial stage of the discharge (0 9 cm -3 for the C(3) ion and anti nsub(0)=1.1x10 10 cm -3 for hydrogen atoms. The investigation into spatial distributions of the Hsub(β) line radiation in the initial stage of the filament formation showed that in the overwhelming majority of cases the discharge appears near the internal wall of the tokamak chamber

Spectroscopic study of ohmically heated Tokamak discharges

International Nuclear Information System (INIS)

Breton, C.; Michelis, C. de; Mattioli, M.

1980-07-01

Tokamak discharges interact strongly with the wall and/or the current aperture limiter producing recycling particles, which penetrate into the discharge and which can be studied spectroscopically. Working gas (hydrogen or deuterium) is usually studied observing visible Balmer lines at several toroidal locations. Absolute measurements allow to obtain both the recycling flux and the global particle confinement time. With sufficiently high resolution the isotopic plasma composition can be obtained. The impurity elements can be divided into desorbed elements (mainly oxygen) and eroded elements (metals from both walls and limiter) according to the plasma-wall interaction processes originating them. Space-and time-resolved emission in the VUV region down to about 20 A will be reviewed for ohmically-heated discharges. The time evolution can be divided into four phases, not always clearly separated in a particular discharge: a) the initial phase, lasting less than 10 ms (the so-called burn-out phase), b) the period of increasing plasma current and electron temperature, lasting typically 10 - 100 ms, c) an eventual steady state (plateau of the plasma current with almost constant density and temperature), d) the increase of the electron density up to or just below the maximum value attainable in a given device. For all these phases the results reported from different devices will be described and compared

Confinement and diffusion in tokamaks

International Nuclear Information System (INIS)

McWilliams, R.

1988-01-01

The effect of electric field fluctuations on confinement and diffusion in tokamak is discussed. Based on the experimentally determined cross-field turbolent diffusion coefficient, D∼3.7*cT e /eB(δn i /n i ) rms which is also derived by a simple theory, the cross-field diffusion time, tp=a 2 /D, is calculated and compared to experimental results from 51 tokamak for standard Ohmic operation

Monte Carlo impurity transport modeling in the DIII-D transport

International Nuclear Information System (INIS)

Evans, T.E.; Finkenthal, D.F.

1998-04-01

A description of the carbon transport and sputtering physics contained in the Monte Carlo Impurity (MCI) transport code is given. Examples of statistically significant carbon transport pathways are examined using MCI's unique tracking visualizer and a mechanism for enhanced carbon accumulation on the high field side of the divertor chamber is discussed. Comparisons between carbon emissions calculated with MCI and those measured in the DIII-D tokamak are described. Good qualitative agreement is found between 2D carbon emission patterns calculated with MCI and experimentally measured carbon patterns. While uncertainties in the sputtering physics, atomic data, and transport models have made quantitative comparisons with experiments more difficult, recent results using a physics based model for physical and chemical sputtering has yielded simulations with about 50% of the total carbon radiation measured in the divertor. These results and plans for future improvement in the physics models and atomic data are discussed

Feasibility study of a fission-suppressed tokamak fusion breeder

International Nuclear Information System (INIS)

Moir, R.W.; Lee, J.D.; Neef, W.S.

1984-12-01

The preliminary conceptual design of a tokamak fissile fuel producer is described. The blanket technology is based on the fission suppressed breeding concept where neutron multiplication occurs in a bed of 2 cm diameter beryllium pebbles which are cooled by helium at 50 atmospheres pressure. Uranium-233 is bred in thorium metal fuel elements which are in the form of snap rings attached to each beryllium pebble. Tritium is bred in lithium bearing material contained in tubes immersed in the pebble bed and is recovered by a purge flow of helium. The neutron wall load is 3 MW/m 2 and the blanket material is ferritic steel. The net fissile breeding ratio is 0.54 +- 30% per fusion reaction. This results in the production of 4900 kg of 233 U per year from 3000 MW of fusion power. This quantity of fuel will provide makeup fuel for about 12 LWRs of equal thermal power or about 18 1 GW/sub e/ LWRs. The calculated cost of the produced uranium-233 is between $23/g and $53/g or equivalent to $10/kg to $90/kg of U 3 O 8 depending on government financing or utility financing assumptions. Additional topics discussed in the report include the tokamak operating mode (both steady state and long pulse considered), the design and breeding implications of using a poloidal divertor for impurity control, reactor safety, the choice of a tritium breeder, and fuel management

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

Disruption mitigation with high-pressure helium gas injection on EAST tokamak

Science.gov (United States)

Chen, D. L.; Shen, B.; Granetz, R. S.; Qian, J. P.; Zhuang, H. D.; Zeng, L.; Duan, Y.; Shi, T.; Wang, H.; Sun, Y.; Xiao, B. J.

2018-03-01

High pressure noble gas injection is a promising technique to mitigate the effect of disruptions in tokamaks. In this paper, results of mitigation experiments with low-Z massive gas injection (helium) on the EAST tokamak are reported. A fast valve has been developed and successfully implemented on EAST, with valve response time  ⩽150 μs, capable of injecting up to 7 × 1022 particles, corresponding to 300 times the plasma inventory. Different amounts of helium gas were injected into stable plasmas in the preliminary experiments. It is seen that a small amount of helium gas (N_He≃ N_plasma ) can not terminate a discharge, but can trigger MHD activity. Injection of 40 times the plasma inventory impurity (N_He≃ 40× N_plasma ) can effectively radiate away part of the thermal energy and make the electron density increase rapidly. The mitigation result is that the current quench time and vertical displacement can both be reduced significantly, without resulting in significantly higher loop voltage. This also reduces the risk of runaway electron generation. As the amount of injected impurity gas increases, the gas penetration time decreases slowly and asymptotes to (˜7 ms). In addition, the impurity gas jet has also been injected into VDEs, which are more challenging to mitigate that stable plasmas.

Influence of large dust particles on plasma performance in the HL-2A tokamak

Energy Technology Data Exchange (ETDEWEB)

Huang, Z.H., E-mail: huangzh@swip.ac.cn; Yan, L.W.; Feng, Z.; Cheng, J.; Tomita, Y.; Liu, L.; Gao, J.M.; Zhong, W.L.; Jiang, M.; Yang, Q.W.; Xu, Y.; Duan, X.R.

2015-08-15

Visible dust particles generated from plasma-facing components (PFCs) and the impact of the dusts on plasma performance as a source of impurities have been studied in the HL-2A tokamak by means of a fast framing camera together with other diagnostics. The camera images display that during a steady state discharge the dusts are accelerated toriodally by the ion drag force and radially by the centrifugal force. The first experimental evidence shows that dust particles originating from the high field side (HFS) lead to a significant reduction of central electron temperature and divertor heat flux, a considerable rise of total radiated power and effective charge, and a slight growth of local electron density. The results reveal that the dusts at the HFS have much stronger effects on plasma performance than those at the low field side (LFS)

Characteristics of post-disruption runaway electrons with impurity pellet injection

International Nuclear Information System (INIS)

Kawano, Yasunori; Nakano, Tomohide; Isayama, Akihiko; Asakura, Nobuyuki; Tamai, Hiroshi; Kubo, Hirotaka; Takenaga, Hidenobu; Bakhtiari, Mohammad; Ide, Shunsuke; Kondoh, Takashi; Hatae, Takaki

2005-01-01

Characteristics of post-disruption runaway electrons with impurity pellet injection were investigated for the first time using the JT-60U tokamak device. A clear deposition of impurity neon ice pellets was observed in a post-disruption runaway plasma. The pellet ablation was attributed to the energy deposition of relativistic runaway electrons in the pellet. A high normalized electron density was stably obtained with n e bar /n GW ∼2.2. Effects of prompt exhaust of runaway electrons and reduction of runaway plasma current without large amplitude MHD activities were found. One possible explanation for the basic behavior of runaway plasma current is that it follows the balance of avalanche generation of runaway electrons and slowing down predicted by the Andersson-Helander model, including the combined effect of collisional pitch angle scattering and synchrotron radiation. Our results suggested that the impurity pellet injection reduced the energy of runaway electrons in a stepwise manner. (author)

Electronic properties of iron impurity in hcp metals from Moessbauer studies

International Nuclear Information System (INIS)

Janot, C.; Delcroix, P.

1975-01-01

Moessbauer spectroscopy was used in quantitative investigating the electronic properties of iron impurities in hexagonal close-packed metals. Beryllium of the highest commercially obtainable purity containing about 300 ppm residual impurities was used as a host element. Experimental evidence is given for the existence of localized electronic states which have non-spherical distribution and obviously contribute especially to the electric field gradient. Iron impurity seems to retain the same electronic behaviour as long as the host hcp metal is a normal one (Mg, Cd, Zn), but the localized electronic states seem to disappear when the host is a transition hcp metal (Co, Ti, Sc, Zr, etc.). (Z.S.)

Optimization design study of an innovative divertor concept for future experimental tokamak-type fusion reactors

International Nuclear Information System (INIS)

Willem Janssens, Ir.; Crutzen, Y.; Farfaletti-Casali, F.; Matera, R.

1991-01-01

The design optimization study of an innovative divertor concept for future experimental tokamak-type fusion devices is both an answer to the actual problems encountered in the multilayer divertor proposals and an illustration of a rational modelling philosophy and optimization strategy for the development of a new divertor structure. Instead of using mechanical attachment or metallurgical bonding of the protective material to the heat sink as in most actual divertor concepts, the so-called brush divertor in this study uses an array of unidirectional fibers penetrating in both the protective armor and the underling composite heat sink. Although the approach is fully concentrated on the divertor performance, including both a description of its function from the theoretical point of view and an overview of the problems related to the materials choice and evaluation, both the approach followed in the numerical modelling and the judgment of the results are thought to be valid also for other applications. Therefore the spin-off of the study must be situated in both the technological progress towards a feasible divertor solution, which introduces no additional physical uncertainties, and in the general area of the thermo-mechanical finite-element modelling on both macro-and microscale. The brush divertor itself embodies the use, and thus the modelling, of advanced materials such as tailor-made metal matrix composites and dispersion strengthened metals, and is shown to offer large potential advantages, demanding however and experimental validation under working conditions. It is clearly indicated where the need originates for an integrated experimental program which must allow to verify the basic modelling assumptions in order to arrive at the use of numerical computation as a powerful and realistic tool of structural testing and life-time prediction

User's manual for DSTAR MOD1: A comprehensive tokamak disruption code

International Nuclear Information System (INIS)

Merrill, B.J.; Jardin, S.J.

1986-01-01

A computer code, DSTAR, has recently been developed to quantify the surface erosion and induced forces that can occur during major tokamak plasma disruptions. The DSTAR code development effort has been accomplished by coupling a recently developed free boundary tokamak plasma transport computational model with other models developed to predict impurity transport and radiation, and the electromagnetic and thermal dynamic response of vacuum vessel components. The combined model, DSTAR, is a unique tool for predicting the consequences of tokamak disruptions. This informal report discusses the sequence of events of a resistive disruption, models developed to predict plasma transport and electromagnetic field evolution, the growth of the stochastic region of the plasma, the transport and nonequilibrium ionization/emitted radiation of the ablated vacuum vessel material, the vacuum vessel thermal and magnetic response, and user input and code output

Experimental study of parametric dependence of electron-scale turbulence in a spherical tokamak

Energy Technology Data Exchange (ETDEWEB)

Ren, Y.; Guttenfelder, W.; Kaye, S. M.; Mazzucato, E.; Bell, R. E.; Diallo, A.; LeBlanc, B. P. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Domier, C. W.; Lee, K. C. [University of California at Davis, Davis, California 95616 (United States); Smith, D. R. [University of Wisconsin-Madison, Madison, Wisconsin 53706 (United States); Yuh, H. [Nova Photonics, Inc., Princeton, New Jersey 08540 (United States)

2012-05-15

Electron-scale turbulence is predicted to drive anomalous electron thermal transport. However, experimental study of its relation with transport is still in its early stage. On the National Spherical Tokamak Experiment (NSTX), electron-scale density fluctuations are studied with a novel tangential microwave scattering system with high radial resolution of {+-}2 cm. Here, we report a study of parametric dependence of electron-scale turbulence in NSTX H-mode plasmas. The dependence on density gradient is studied through the observation of a large density gradient variation in the core induced by an edge localized mode (ELM) event, where we found the first clear experimental evidence of density gradient stabilization of electron-gyro scale turbulence in a fusion plasma. This observation, coupled with linear gyro-kinetic calculations, leads to the identification of the observed instability as toroidal electron temperature gradient (ETG) modes. It is observed that longer wavelength ETG modes, k{sub Up-Tack }{rho}{sub s} Less-Than-Or-Equivalent-To 10 ({rho}{sub s} is the ion gyroradius at electron temperature and k{sub Up-Tack} is the wavenumber perpendicular to local equilibrium magnetic field), are most stabilized by density gradient, and the stabilization is accompanied by about a factor of two decrease in electron thermal diffusivity. Comparisons with nonlinear ETG gyrokinetic simulations show ETG turbulence may be able to explain the experimental electron heat flux observed before the ELM event. The collisionality dependence of electron-scale turbulence is also studied by systematically varying plasma current and toroidal field, so that electron gyroradius ({rho}{sub e}), electron beta ({beta}{sub e}), and safety factor (q{sub 95}) are kept approximately constant. More than a factor of two change in electron collisionality, {nu}{sub e}{sup *}, was achieved, and we found that the spectral power of electron-scale turbulence appears to increase as {nu}{sub e}{sup *} is

Preionization and start-up in the ISX-B tokamak using electron cyclotron heating at 28 GHz

International Nuclear Information System (INIS)

Kulchar, A.G.; Eldridge, O.C.; England, A.C.

1983-10-01

A 28-GHz gyrotron was used to produce a plasma at the electron cyclotron resonance in the Impurity Study Experiment (ISX-B) tokamak. The influence of the toroidal magnetic field magnitude, error fields, gas pressure, microwave power, microwave pulse length, and microwave timing was studied for experiments with magnetic field and gas only. Also, experiments with preionization followed by capacitor discharges were carried out in which these quantities were varied, as were the capacitor bank voltages. Optimum conditions of preionization for some of the parameters were determined. A theoretical model that adequately reproduces the data is given. Calculations based on this model show the temporal evolution of the electron temperature and density, the neutral density, and the plasma current. The model adequately accounts for present and previous experimental results and can be used to make predictions for future experiments

Numerical studies of edge localized instabilities in tokamaks

International Nuclear Information System (INIS)

Wilson, H.R.; Snyder, P.B.; Huysmans, G.T.A.; Miller, R.L.

2002-01-01

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

Measurement of the central ion and electron temperature of tokamak plasmas from the x-ray line radiation of high-Z impurity ions

International Nuclear Information System (INIS)

Bitter, M.; von Goeler, S.; Goldman, M.; Hill, K.W.; Horton, R.; Roney, W.; Sauthoff, N.; Stodiek, W.

1982-04-01

This paper describes measurements of the central ion and electron temperature of tokamak plasmas from the observation of the 1s - 2p resonance lines, and the associated dielectronic (1s 2 nl - 1s2pnl, with n greater than or equal to 2) satellites, of helium-like iron (Fe XXV) and titanium (Ti XXI). The satellite to resonance line ratios are very sensitive to the electron temperature and are used as an electron temperature diagnostic. The ion temperature is deduced from the Doppler width of the 1s - 2p resonance lines. The measurements have been performed with high resolution Bragg crystal spectrometers on the PLT (Princeton Large Torus) and PDX (Poloidal Divertor Experiment) tokamaks. The details of the experimental arrangement and line evaluation are described, and the ion and electron temperature results are compared with those obtained from independent diagnostic techniques, such as the analysis of charge-exchange neutrals and measurements of the electron cyclotron radiation. The obtained experimental results permit a detailed comparison with theoretical predictions

Impurity study of TMX using ultraviolet spectroscopy

International Nuclear Information System (INIS)

Allen, S.L.; Strand, O.T.; Moos, H.W.; Fortner, R.J.; Nash, T.J.; Dietrich, D.D.

1981-01-01

An extreme ultraviolet (EUV) study of the emissions from intrinsic and injected impurities in TMX is presented. Two survey spectrographs were used to determine that the major impurities present were oxygen, nitrogen, carbon, and titanium. Three absolutely-calibrated monochromators were used to measure the time histories and radial profiles of these impurity emissions in the central cell and each plug. Two of these instruments were capable of obtaining radial profiles as a function of time in a single shot

Experimental results from the TUMAN 3 tokamak

International Nuclear Information System (INIS)

Golant, V.E.; Andrejko, M.V.; Askinazi, L.G.; Korneev, V.A.; Krikunov, S.V.; Lipin, B.M.; Lebedev, S.V.; Levin, L.S.; Podushnikova, K.A.; Razdobarin, G.T.; Rozhansky, V.A.; Rozhdestvensky, V.V.; Tendler, M.; Tukachinsky, A.S.; Jaroshevich, S.P.

1995-01-01

The open-quote open-quote TUMAN-3 close-quote close-quote Tokamak programme concentrates on issues of improved confinement. In 1989 the transition from an ordinary Ohmic regime into an improved confinement mode was achieved. The signatures of the H-mode in auxiliary heated tokamaks have been observed in this regime. The crucial role of the boundary radial electric field was found in the experiments with internal bias probe. Other techniques were demonstrated to disturb the boundary plasma which led to H-mode triggering: short increase of working gas puffing, minor radius magnetic compression and pellet injection. The role scaling of the energy confinement time in the Ohmic H-mode was obtained, which differs dramatically from the scaling for the ordinary Ohmic regime. There were found a strong dependence of τ E on plasma current and a weak dependence on density. The maximum value of τ E was 10 times longer than in the ordinary Ohmic region. The τ E scaling for the Ohmic H-mode is consistent with the scaling proposed for devices with powerful auxiliary heating. The results shows that H-mode physics is universal in tokamaks with different geometries and heating methods. (AIP) copyright 1995 American Institute of Physics

ADVANCES IN COMPREHENSIVE GYROKINETIC SIMULATIONS OF TRANSPORT IN TOKAMAKS

International Nuclear Information System (INIS)

WALTZ, RE; CANDY, J; HINTON, FL; ESTRADA-MILA, C; KINSEY, JE.

2004-01-01

A continuum global gyrokinetic code GYRO has been developed to comprehensively simulate core turbulent transport in actual experimental profiles and enable direct quantitative comparisons to the experimental transport flows. GYRO not only treats the now standard ion temperature gradient (ITG) mode turbulence, but also treats trapped and passing electrons with collisions and finite β, equilibrium ExB shear stabilization, and all in real tokamak geometry. Most importantly the code operates at finite relative gyroradius (ρ * ) so as to treat the profile shear stabilization and nonlocal effects which can break gyroBohm scaling. The code operates in either a cyclic flux-tube limit (which allows only gyroBohm scaling) or a globally with physical profile variation. Rohm scaling of DIII-D L-mode has been simulated with power flows matching experiment within error bars on the ion temperature gradient. Mechanisms for broken gyroBohm scaling, neoclassical ion flows embedded in turbulence, turbulent dynamos and profile corrugations, plasma pinches and impurity flow, and simulations at fixed flow rather than fixed gradient are illustrated and discussed

Influence of impurities on the crystallization of dextrose monohydrate

Science.gov (United States)

Markande, Abhay; Nezzal, Amale; Fitzpatrick, John; Aerts, Luc; Redl, Andreas

2012-08-01

The effects of impurities on dextrose monohydrate crystallization were investigated. Crystal nucleation and growth kinetics in the presence of impurities were studied using an in-line focused beam reflectance monitoring (FBRM) technique and an in-line process refractometer. Experimental data were obtained from runs carried out at different impurity levels between 4 and 11 wt% in the high dextrose equivalent (DE) syrup. It was found that impurities have no significant influence on the solubility of dextrose in water. However, impurities have a clear influence on the nucleation and growth kinetics of dextrose monohydrate crystallization. Nucleation and growth rate were favored by low levels of impurities in the syrup.

Numerical evaluation of experimental models to investigate the dynamic behavior of the ITER tokamak assembly

International Nuclear Information System (INIS)

Onozuka, M.; Takeda, N.; Nakahira, M.; Shimizu, K.; Nakamura, T.

2003-01-01

The most recent assessment method to evaluate the dynamic behavior of the International Thermonuclear Experimental Reactor (ITER) tokamak assembly is outlined. Three experimental models, including a 1/5.8-scale tokamak model, have been considered to validate the numerical analysis methods for dynamic events, particularly seismic ones. The experimental model has been evaluated by numerical calculations and the results are presented. In the calculations, equivalent linearization has been applied for the non-linear characteristics of the support flange connection, caused by the effects of the bolt-fastening and the friction between the flanges. The detailed connecting conditions for the support flanges have been developed and validated for the analysis. Using the conditions, the eigen-mode analysis has shown that the first and second eigen-mode are horizontal vibration modes with the natural frequency of 39 Hz, while the vertical vibration mode is the fourth mode with the natural frequency of 86 Hz. Dynamic analysis for seismic events has shown the maximum acceleration of approximately twofold larger than that of the applied acceleration, and the maximum stress of 104 MPa found in the flange connecting bolt. These values will be examined comparing with experimental results in order to validate the analysis methods

Numerical evaluation of experimental models to investigate the dynamic behavior of the ITER tokamak assembly

Energy Technology Data Exchange (ETDEWEB)

Onozuka, M. E-mail: masanori_onozuka@mhi.co.jp; Takeda, N.; Nakahira, M.; Shimizu, K.; Nakamura, T

2003-09-01

The most recent assessment method to evaluate the dynamic behavior of the International Thermonuclear Experimental Reactor (ITER) tokamak assembly is outlined. Three experimental models, including a 1/5.8-scale tokamak model, have been considered to validate the numerical analysis methods for dynamic events, particularly seismic ones. The experimental model has been evaluated by numerical calculations and the results are presented. In the calculations, equivalent linearization has been applied for the non-linear characteristics of the support flange connection, caused by the effects of the bolt-fastening and the friction between the flanges. The detailed connecting conditions for the support flanges have been developed and validated for the analysis. Using the conditions, the eigen-mode analysis has shown that the first and second eigen-mode are horizontal vibration modes with the natural frequency of 39 Hz, while the vertical vibration mode is the fourth mode with the natural frequency of 86 Hz. Dynamic analysis for seismic events has shown the maximum acceleration of approximately twofold larger than that of the applied acceleration, and the maximum stress of 104 MPa found in the flange connecting bolt. These values will be examined comparing with experimental results in order to validate the analysis methods.

Experimental observation of current generation by asymmetrical heating of ions in a tokamak plasma

International Nuclear Information System (INIS)

Gahl, J.; Ishihara, O.; Wong, K.L.; Kristiansen, M.; Hagler, M.

1986-01-01

The first experimental observation of current generation by asymmetrical heating of ions is reported. Ions were asymmetrically heated by a unidirectional fast Alfven wave launched by a slow wave antenna inside a tokamak. Current generation was detected by measuring the asymmetry of the toroidal plasma current with probes at the top and bottom of the toroidal plasma column

Experimental test of far-infrared polarimetry for Faraday rotation measurements on the TFR 600 Tokamak

Energy Technology Data Exchange (ETDEWEB)

Soltwisch, H [Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.). Inst. fuer Plasmaphysik; Association Euratom-Kernforschungsanlage Juelich G.m.b.H. (Germany, F.R.)); Equipe, T F.R. [Association Euratom-CEA sur la Fusion, Centre d' Etudes Nucleaires de Fontenay-aux-Roses, 92 (France). Dept. de Recherches sur la Fusion Controlee

1981-09-01

The results are reported on the feasibility of using far-infrared polarimetry for Faraday rotation diagnostic measurements on the TRF Tokamak. Precise quantitative results were not obtained but a satisfactory agreement with a simple theoretical model leads to a good understanding of the experimental limitations of the method.

Overview of the STARFIRE reference commercial tokamak fusion power reactor design

International Nuclear Information System (INIS)

Baker, C.C.; Abdou, M.A.; DeFreece, D.A.; Trachsel, C.A.; Graumann, D.; Barry, K.

1980-01-01

The purpose of the STARFIRE study is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The major features for STARFIRE include a steady-state operating mode based on a continuous rf lower-hybrid current drive and auxiliary heating, solid tritium breeder material, pressurized water cooling, limiter/vacuum system for impurity control and exhaust, high tritium burnup, superconducting EF coils outside the TF superconducting coils, fully remote maintenance, and a low-activation shield

A system to deposit boron films (boronization) in the DIII-D tokamak

International Nuclear Information System (INIS)

Hodapp, T.R.; Jackson, G.L.; Phillips, J.; Holtrop, K.L.; Petersen, P.I.; Winter, J.

1991-09-01

A system has been added to the D3-D tokamak to coat its plasma facing surfaces with a film of boron using diborane gas. The system includes special health and safety equipment for handling the diborane gas which is toxic and inflammable. The purpose of the boron film is to reduce the levels of impurity atoms in the D3-D plasmas. Experiments following the application of the boron film in D3-D have led to significant reductions in plasma impurity levels and the observation of a new, very high confinement regime. 9 refs., 1 fig

Phonon-impurity relaxation and acoustic wave absorption in yttrium-aluminium garnet crystals with impurities

International Nuclear Information System (INIS)

Ivanov, S.N.; Kotelyanskij, I.M.; Medved', V.V.

1983-01-01

The experimental results of investigations of the influence of substitution impurities in the yttrium-aluminium garnet lattice on absorption of high-frequency acoustic waves are presented. It is shown that the phonon-impurity relaxation processses affect at most the wave absorption and have resonance character when the acoustic wave interacts with the thermal phonon group in the vicinity of the perturbed part of the phonon spectrum caused by the impurity. The differences of time values between inelastic and elastic thermal phonons relaxations determined from the data on longitudinal and shear waves in pure and impurity garnet crystals are discussed

0-d modeling of fast radiative shutdown of Tokamak discharges following massive gas injection

International Nuclear Information System (INIS)

Hollmann, E.M.; Parks, P.B.; Scott, H.A.

2008-01-01

0-D modeling of fast radiative shutdowns of tokamak discharges following massive gas injection is presented. Realistic neutral deposition rates are used together with a 1-D diffusive model to estimate impurity deposition into the plasma. Non-coronal radiation rates including opacity are used, as are induced wall currents, wall impurity radiation, and neutral and neoclassical corrections to plasma resistivity. The 0-D modeling is found to reproduce the shutdown timescale and free electron density rise seen in DIII-D argon injection experiments well. Opacity, wall currents, and wall impurities can all have a significant (>10%) impact on simulated timescales. (copyright 2008 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

Three novel tokamak plasma regimes in TFTR

International Nuclear Information System (INIS)

Furth, H.P.

1985-10-01

Aside from extending ''standard'' ohmic and neutral beam heating studies to advanced plasma parameters, TFTR has encountered a number of special plasma regimes that have the potential to shed new light on the physics of tokamak confinement and the optimal design of future D-T facilities: (1) High-powered, neutral beam heating at low plasma densities can maintain a highly reactive hot-ion population (with quasi-steady-state beam fueling and current drive) in a tokamak configuration of modest bulk-plasma confinement requirements. (2) Plasma displacement away from limiter contact lends itself to clarification of the role of edge-plasma recycling and radiation cooling within the overall pattern of tokamak heat flow. (3) Noncentral auxiliary heating (with a ''hollow'' power-deposition profile) should serve to raise the central tokamak plasma temperature without deterioration of central region confinement, thus facilitating the study of alpha-heating effects in TFTR. The experimental results of regime (3) support the theory that tokamak profile consistency is related to resistive kink stability and that the global energy confinement time is determined by transport properties of the plasma edge region

Discharge initiation experiments in the Tokapole II tokamak

International Nuclear Information System (INIS)

Shepard, D.A.

1984-01-01

Experiments in the Tokapole II tokamak demonstrate the benefits of high density (n/sub e//n/sub o/ greater than or equal to 0.01) preionization by reducing four quantities at startup: necessary toroidal loop voltage (V 1 ) (50%), volt-second consumption (40-50%), impurity radiation (25-50%), and runaway electron production (approx. 80-100%). A zero-dimensional code models the loop voltage reduction dependence on preionization density and predicts a similar result for reactor scale devices. The code shows low initial resistivity and a high resistivity time derivative contribute to loop voltage reduction. Microwaves at the electron cyclotron resonance (ECR) frequency and plasma gun injection produce high density preionization, which reduces the initial V 1 , volt-second consumption, and runaways. The ECR preionization also reduces impurity radiation by shortening the time from voltage application to current channel formation. This, evidently, reduces the total plasma-wall interaction at startup. The power balance of the ECR plasma in a toroidal-field-only case was studied using Langmuir probes and impurity doping. The vertical electric field and current, which result from curvature drift, were measured as approx. 10 V/cm and 50 amps, respectively, and exceeded expected values for the bulk electron temperature (approx. 10 eV)

Power exhaust by impurity seeding in fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Bernert, Matthias; Kallenbach, Arne; Dux, Ralph; Wischmeier, Marco [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Reimold, Felix [Forschungszentrum Juelich GmbH, IEK, Juelich (Germany); Lipschultz, Bruce [University of York, York Plasma Institute, Heslington, York (United Kingdom); Collaboration: the ASDEX Upgrade team; the EUROfusion MST1 Team

2016-07-01

Power exhaust is one of the big challenges for future fusion reactors. The power load at the divertor targets, the primary plasma-wall interaction zone, would exceed material limits and, thus, must be reduced. Therefore, 90% of the exhaust power needs to be dissipated and the divertor is anticipated to be in the detached regime, where the interaction of the plasma with the wall is significantly reduced. Radiation is the dominant dissipation process and is increased by impurity seeding. The radiation distribution can be tailored by using different seed impurities (N for radiation outside, Ne and Ar for radiation at the edge of and Kr for radiation inside the confined region). The tailoring of the radiation profile is required in order to maximize the radiated power and at the same time minimize the impact on the energy confinement. Recent experiments with intense impurity seeding at the ASDEX Upgrade tokamak demonstrate operation at highest heat fluxes and detached divertor targets at radiated power fractions of up to 90%. In these scenarios the radiation originates predominantly from the confined region and leads to an unexpectedly small confinement reduction.

Numerical calculation of impurity charge state distributions

International Nuclear Information System (INIS)

Crume, E.C.; Arnurius, D.E.

1977-09-01

The numerical calculation of impurity charge state distributions using the computer program IMPDYN is discussed. The time-dependent corona atomic physics model used in the calculations is reviewed, and general and specific treatments of electron impact ionization and recombination are referenced. The complete program and two examples relating to tokamak plasmas are given on a microfiche so that a user may verify that his version of the program is working properly. In the discussion of the examples, the corona steady-state approximation is shown to have significant defects when the plasma environment, particularly the electron temperature, is changing rapidly

On a mechanism of antenna phasing effect on impurity production during ICRF plasma heating

International Nuclear Information System (INIS)

Chechkin, V.V.; Grigor'eva, L.I.

1990-01-01

An appreciable reduction of the metal impurity in flux and a decrease in SOL plasma parameter disturbance occure during ICRP heating in some tokamaks when toroidally adjacent antennae are driven in anti-phase. Also cancelled are low-frequency electric field fluctuations arising in the sheaths and the associated charged particle flux fluctuations. 24 refs.; 7 figs

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

The use of internal transport barriers in tokamak plasmas

Energy Technology Data Exchange (ETDEWEB)

Challis, C D [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxon, OX14 3DB (United Kingdom)

2004-12-01

Internal transport barriers (ITBs) can provide high tokamak confinement at modest plasma current. This is desirable for operation with most of the current driven non-inductively by the bootstrap mechanism, as currently envisaged for steady-state power plants. Maintaining such plasmas in steady conditions with high plasma purity is challenging, however, due to MHD instabilities and impurity transport effects. Significant progress has been made in the control of ITB plasmas: the pressure profile has been varied using the barrier location; q-profile modification has been achieved with non-inductive current drive, and means have been found to affect density peaking and impurity accumulation. All these features are, to some extent, interdependent and must be integrated self-consistently to demonstrate a sound basis for extrapolation to future devices.

Zeff measurements and low-Z impurity transport for NBI and ICRF heated plasma in JIPP T-IIU tokamak

International Nuclear Information System (INIS)

Ida, K.; Amano, T.; Kawahata, K.; Kaneko, O.

1988-12-01

A visible bremsstrahlung detector array system for Z eff measurements and a charge exchange recombination spectroscopy (CXRS) system for fully ionized impurity profile measurements were installed on JIPP TII-U to study impurity transport for NBI and ICRF heated plasma. More impurities are sputtered by ICRF heating than by NBI and/or ohmic heatings. The carbon contribution to Z eff is 80-90 % for NBI heated plasmas, and 60 % for NBI + ICRF heated plasmas. With a carbon coating of vacuum vessel, the Z eff value decreases 2.4 to 1.7 and the carbon contribution to Z eff increases up to 80-90 %. We obtain the diffusion coefficient D a = 1.0 m 2 /s and the convective velocity V a (a) = 13 m/s at the plasma edge for carbon impurity from the radial profile and time evolution of fully ionized carbon after the ICRF pulse is turned on. (author)

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)

Studies of nanostructures formed in T-10 tokamak

Energy Technology Data Exchange (ETDEWEB)

Kolbasov, B N; Stankevich, V G; Svechnikov, N Yu; Lebedev, A M; Menshikov, K A; Somenkov, V A; Trunova, V A; Veligzhanin, A A; Zubavichus, Y V [Kurchatov Institute, Kurchatov square 1, Moscow 123182 (Russian Federation); Rajarathnam, D, E-mail: kolbasov@nfi.kiae.ru [CERAR, University of South Australia, 5095 (Australia)

2011-06-23

According to the X-ray diffraction (XRD) studies, hydrocarbon films and flakes formed under deuterium plasma discharges in T-10 tokamak are amorphous with graphene-like sheets. They have atomic ratio (D + H)/C about 1 and higher. The XRD peak positions revealed the presence of structural defects with interplane distances of 0.12, 0.24 and 0.66 nm. The peak widths gave the in-plane sizes of the scattering structures equal to about 1 nm. The properties of such films were studied with application of small-angle and wide-angle X-ray scattering measurements, neutron diffraction and other techniques. These experiments have shown that the films contain about 63% of sp{sup 3} and {approx}37% of sp{sup 2} states. X-ray fluorescence spectroscopy employing synchrotron radiation revealed that the films contain at least 12 impurities of Fe, Mo, Cr, Ni, Nb and other transition metals. Difference between film properties on its opposite sides was revealed using Fourier-transform infrared spectroscopy and analysis of current-voltage characteristics (CVC). On the wall facing side of the film, graphite-like Csp{sup 2} structures dominate. On the plasma facing side, diamond-like Csp{sup 3} structures prevail. Deuterium retention can be monitored by two groups of vibrational sp{sup 3} modes with different oscillator strengths, depending on the amount of deuterium in films.

Thermal hydraulics of the impurity control system for FED/INTOR

International Nuclear Information System (INIS)

Cha, Y.S.; Mattas, R.F.; Abdou, M.A.; Haines, J.R.

1983-01-01

This paper addresses two important aspects of thermal hydraulics related to the design of the impurity control system (limiter and divertor) of the Fusion Engineering Device (FED) and the International Tokamak Reactor (INTOR). The first part of the paper is devoted to the determination of temperature distributions in various combinations of the coating/structural materials proposed for the limiter/divertor of FED and INTOR. The second part of the paper describes the analysis of the tangential motion of the melt layer under the influence of magnetic force during plasma disruption. The results of both analysis provide inputs to the determination of the life time of the limiter (or divertor) which is the most critical problem for the impurity control system as far as engineering and materials consideration is concerned

Operation and control of high density tokamak reactors

International Nuclear Information System (INIS)

Attenberger, S.E.; McAlees, D.G.

1976-01-01

The incentive for high density operation of a tokamak reactor was discussed. It is found that high density permits ignition in a relatively small, moderately elongated plasma with a moderate magnetic field strength. Under these conditions, neutron wall loadings approximately 4 MW/m 2 must be tolerated. The sensitivity analysis with respect to impurity effects shows that impurity control will most likely be necessary to achieve the desired plasma conditions. The charge exchange sputtered impurities are found to have an important effect so that maintaining a low neutral density in the plasma is critical. If it is assumed that neutral beams will be used to heat the plasma to ignition, high energy injection is required (approximately 250 keV) when heating is accompished at full density. A scenario is outlined where the ignition temperature is established at low density and then the fueling rate is increased to attain ignition. This approach may permit beams with energies being developed for use in TFTR to be successfully used to heat a high density device of the type described here to ignition

Recent advances in the HL-2A tokamak experiments

International Nuclear Information System (INIS)

Liu, Y.; Ding, X.T.; Yang, Q.W.; Yan, L.W.; Liu, D.Q.; Xuan, W.M.; Chen, L.Y.; Song, X.M.; Cao, Z.; Zhang, J.H.; Mao, W.C.; Zhou, C.P.; Li, X.D.; Wang, S.J.; Yan, J.C.; Bu, M.N.; Chen, Y.H.; Cui, C.H.; Cui, Z.Y.; Deng, Z.C.; Hong, W.Y.; Hu, H.T.; Huang, Y.; Kang, Z.H.; Li, B.; Li, W.; Li, F.Z.; Li, G.S.; Li, H.J.; Li, Q.; Li, Y.G.; Li, Z.J.; Liu, Yi; Liu, Z.T.; Luo, C.W.; Mao, X.H.; Pan, Y.D.; Rao, J.; Shao, K.; Song, X.Y.; Wang, M.; Wang, M.X.; Wang, Q.M.; Xiao, Z.G.; Xie, Y.F.; Yao, L.H.; Yao, L.Y.; Zheng, Y.J.; Zhong, G.W.; Zhou, Y.; Pan, C.H.

2005-01-01

Two experiment campaigns were conducted on the HL-2A tokamak in 2003 and 2004 after the first plasma was obtained at the end of 2002. Progresses in many aspects have been made, especially in the divertor discharge and feedback control of plasma configuration. Up to now, the following operation parameters have been achieved: I p = 320 kA, B t = 2.2 T and discharge duration T d = 1580 ms. With the feedback control of plasma current and horizontal position, an excellent repeatability of the discharge has been achieved. The tokamak has been operated at both limiter configuration and single null (SN) divertor configuration. The HL-2A SN divertor configuration is simulated with the MHD equilibrium code SWEQU. When the divertor configuration is formed, the impurity radiation in the main plasma decreases remarkably

Edge plasma physical investigations of tokamak plasmas in CRIP

International Nuclear Information System (INIS)

Bakos, J.; Ignacz, P.; Koltai, L.; Paszti, F.; Petravich, G.; Szigeti, J.; Zoletnik, S.

1988-01-01

The results of the measurements performed in the field of thermonuclear high temperature plasma physics in CRIP (Hungary) are summarized. In the field of the edge plasma physics solid probes were used to test the external zone of plasma edges, and atom beams and balls were used to investigate both the external and internal zones. The plasma density distribution was measured by laser blow-off technics, using Na atoms, which are evaporated by laser pulses. The excitation of Na atom ball by tokamak plasma gives information on the status of the plasma edge. The toroidal asymmetry of particle transport in tokamak plasma was measured by erosion probes. The evaporated and transported impurities were collected on an other part of the plasma edge and were analyzed by SIMS and Rutherford backscattering. The interactions in plasma near the limiter were investigated by a special limiter with implemented probes. Recycling and charge exchange processes were measured. Disruption phenomena of tokamak plasma were analyzed and a special kind of disruptions, 'soft disruptions' and the related preliminary perturbations were discovered. (D.Gy.) 10 figs

Lithium capillary porous system behavior as PFM in FTU Tokamak experiments

International Nuclear Information System (INIS)

Apichela, M.L.; Mazzitelli, G.; Lyublinski, I.E.; Lazarev, V.; Mirnov, S.; Vertkov, A.

2007-01-01

Full text of publication follows: Liquid lithium use on the base of capillary porous systems (CPS) application as plasma facing material (PFM) of tokamaks is advanced way to solve the problems of plasma contamination with high Z impurity, PFM degradation and tritium retention. In frame of joint program between ENEA (Italy) and FSUE 'Red Star' and TRINITI (RF) started at the end of 2005 die test of liquid lithium limiter (LLL) with CPS in a high field, medium size, carbon free tokamak FTU have been performed successfully. The LLL has been inserted in ohmic plasma discharges and at additional heating with LH and ECR at power levels in the MW range without any particular problem (BT = 6 T, Ip = 0.5- 0.9 MA, n e = 0.2 -2.6x10 20 m -3 , t = 1.5 s, P∼ 2-5 MW/m 2 at a normal discharge). The behavior of lithium CPS based on stainless steel wire mesh and its surface modification in normal discharges and at disruptions has been studied. Results of microscopic analyses of CPS structure after experimental campaigns are presented. The possibility to withstand heat load exceeding 5 MW/m 2 without damage, lithium surface renewal, mechanical stabilization of liquid lithium against MHD forces have been confirmed. Application of W, Mo as the base material and possible structure types of CPS have been considered for operating parameters improvement of long-living plasma facing components. (authors)

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)

Graphene plasmons: Impurities and nonlocal effects

Science.gov (United States)

Viola, Giovanni; Wenger, Tobias; Kinaret, Jari; Fogelström, Mikael

2018-02-01

This work analyzes how impurities and vacancies on the surface of a graphene sample affect its optical conductivity and plasmon excitations. The disorder is analyzed in the self-consistent Green's function formulation and nonlocal effects are fully taken into account. It is shown that impurities modify the linear spectrum and give rise to an impurity band whose position and width depend on the two parameters of our model, the density and the strength of impurities. The presence of the impurity band strongly influences the electromagnetic response and the plasmon losses. Furthermore, we discuss how the impurity-band position can be obtained experimentally from the plasmon dispersion relation and discuss this in the context of sensing.

Conditioning of SST-1 Tokamak Vacuum Vessel by Baking and Glow Discharge Cleaning

International Nuclear Information System (INIS)

Khan, Ziauddin; George, Siju; Semwal, Pratibha; Dhanani, Kalpeshkumar R.; Pathan, Firozkhan S.; Paravastu, Yuvakiran; Raval, Dilip C.; Babu, Gattu Ramesh; Khan, Mohammed Shoaib; Pradhan, Subrata

2016-01-01

Highlights: • SST-1 Tokamak was successfully commissioned. • Vacuum vessel was pumped down to 4.5 × 10"–"8 mbar after baking and continuous GDC. • GDC reduced the water vapour by additional 57% while oxygen was reduced by 50%. • Under this condition, an initial plasma breakdown with current of 40 kA for 75 ms was achieved. - Abstract: Steady-state Superconducting Tokamak (SST-1) vacuum vessel (VV) adopts moderate baking at 110 ± 10 °C and the limiters baking at 250 ± 10 °C for ∼ 200 h followed by glow discharge cleaning in hydrogen (GDC-H) with 0.15 A/m"2 current density towards its conditioning prior to plasma discharge experiment. The baking in SST-1 reduces the water (H_2O) vapor by 95% and oxygen (O_2) by 60% whereas the GDC reduces the water vapor by an additional 57% and oxygen by another 50% as measured with residual gas analyzer. The minimum breakdown voltage for H-GDC in SST-1 tokamak was experimentally observed to 300 V at 8 mbar cm. As a result of these adherences, SST-1 VV achieves an ultimate of 4.5 × 10"−"8 mbar with two turbo-molecular pumps with effective pumping speed of 3250 l/s. In the last campaign, SST-1 has achieved successful plasma breakdown, impurity burn through and a plasma current of ∼ 40 kA for 75 ms.

Conditioning of SST-1 Tokamak Vacuum Vessel by Baking and Glow Discharge Cleaning

Energy Technology Data Exchange (ETDEWEB)

Khan, Ziauddin, E-mail: ziauddin@ipr.res.in; George, Siju; Semwal, Pratibha; Dhanani, Kalpeshkumar R.; Pathan, Firozkhan S.; Paravastu, Yuvakiran; Raval, Dilip C.; Babu, Gattu Ramesh; Khan, Mohammed Shoaib; Pradhan, Subrata

2016-02-15

Highlights: • SST-1 Tokamak was successfully commissioned. • Vacuum vessel was pumped down to 4.5 × 10{sup –8} mbar after baking and continuous GDC. • GDC reduced the water vapour by additional 57% while oxygen was reduced by 50%. • Under this condition, an initial plasma breakdown with current of 40 kA for 75 ms was achieved. - Abstract: Steady-state Superconducting Tokamak (SST-1) vacuum vessel (VV) adopts moderate baking at 110 ± 10 °C and the limiters baking at 250 ± 10 °C for ∼ 200 h followed by glow discharge cleaning in hydrogen (GDC-H) with 0.15 A/m{sup 2} current density towards its conditioning prior to plasma discharge experiment. The baking in SST-1 reduces the water (H{sub 2}O) vapor by 95% and oxygen (O{sub 2}) by 60% whereas the GDC reduces the water vapor by an additional 57% and oxygen by another 50% as measured with residual gas analyzer. The minimum breakdown voltage for H-GDC in SST-1 tokamak was experimentally observed to 300 V at 8 mbar cm. As a result of these adherences, SST-1 VV achieves an ultimate of 4.5 × 10{sup −8} mbar with two turbo-molecular pumps with effective pumping speed of 3250 l/s. In the last campaign, SST-1 has achieved successful plasma breakdown, impurity burn through and a plasma current of ∼ 40 kA for 75 ms.

Digital control of plasma position in Damavand tokamak

Energy Technology Data Exchange (ETDEWEB)

Emami, M.; Babazadeh, A.R.; Roshan, M.V.; Memarzadeh, M.; Habibi, H. [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of). Nuclear Fusion Research Center. Plasma Physics Lab.

2002-03-01

Plasma position control is one of the important issues in the design and operation of tokamak fusion research device. Since a tokamak is basically an electrical system consisting of power supplies, coils, plasma and eddy currents, a model in which these components are treated as an electrical circuits is used in designing Damavand plasma position control system. This model is used for the simulation of the digital control system and its parameters have been verified experimentally. In this paper, the performance of a high-speed digital controller as well as a simulation study and its application to the Damavand tokamak is discussed. (author)

Advanced commercial Tokamak optimization studies

International Nuclear Information System (INIS)

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

1985-01-01

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

Design studies of Tokamak power reactor in JAERI

International Nuclear Information System (INIS)

Tone, T.; Nishikawa, M.; Tanaka, Y.

1985-01-01

Recent design studies of tokamak power reactor and related activities conducted in JAERI are presented. A design study of the SPTR (Swimming-Pool Type Reactor) concept was carried out in FY81 and FY82. The reactor design studies in the last two years focus on nuclear components, heat transport and energy conversion systems. In parallel of design studies, tokamak systems analysis code is under development to evaluate reactor performances, cost and net energy balance

The influence of optical parameters on impurity determinations by IR spectroscopy

International Nuclear Information System (INIS)

Lombard, O.J.

1985-01-01

The important role of impurities in semiconductor materials is the subject of continuous research. The concentration of interstitial oxygen impurities in silicon are determined with the aid of infrared spectroscopy. The maximum absorption coefficient of the oxygen absorption peak, centered at 9,06 μm, is determined and the impurity concentration is then calculated using a calibration factor. This procedure was evaluated, paying particular attention to those optical parameters which may influence these impurity determinations. A thorough discussion of the theoretical and experimental aspects of infrared spectroscopy in general is followed by an overview of previous experimental work. This lead to some theoretical analysis regarding the influence of the index of refraction, the index of absorption and multiple reflections in the silicon wafer on impurity determinations. This lead to specific experimental investigations. The influence of the surface morphology of samples on impurity determinations was studied by determining the reflectance of silicon surfaces. It was established that the surface reflectance plays a role and that it must be taken into consideration for accurate impurity concentration determinations. The most accurate values for the absorption coefficient due to oxygen in silicon are calculated. This requires that the surface of the silicon wafers must be highly polished for the formula to be valid. Acceptable values for the absorption coefficient of damaged surfaces are obtained if the uncorrected formula is used. Experimental results may deviate as much as 32% from the real impurity concentration if the wrong formula is used to calculate the absorption coefficient of oxygen in silicon at 9,06 μm

Physics analysis of the Apollo D-3He tokamak reactor

International Nuclear Information System (INIS)

Santarius, J.F.; Emmert, G.A.

1990-01-01

Recent developments in the analysis and conceptual design of Apollo, a D- 3 He Tokamak Reactor are presented. Encouraging experimental results on TEXT motivated a key change in the Apollo concept utilization of an ergodic magnetic limiter for impurity control instead of a divertor. Parameters for the updated Apollo design and an analysis of the ergoidc magnetic limiter are given. The Apollo reference case uses direct conversion of synchrotron radiation to electricity by rectifying antennas (rectennas) for its power conversion system. Previous analyses of this concept are expanded, including further details of the rectennas and of the loss of synchrotron power to the waveguides and walls. Although Apollo will burn D- 3 He fuel, a significant amount of unburned tritium will be generated by D4D reactions. The possibility of operating a short, dedicated, T+ 3 He burn phase to eliminate this tritium will be examined

Study of ion cyclotron fluctuations. Application to the measurement of the ion temperature

International Nuclear Information System (INIS)

Lehner, T.

1982-02-01

A diagnostic technique for measuring the ion temperature of tokamak-type plasmas was developed. A theoretical study was made of the form factor associated with the ion cyclotron waves; the influence of Te/Ti on the frequency of the extrema of the dispersion relations was demonstrated. The different effects able to modify the spectral density (in particular the drift velocity and the impurities) were investigated. The mechanisms of suprathermal excitation of cylotron waves in tokamaks were reviewed together with the various effects stabilizing the spectrum: collisions, shear of the magnetic field lines. The experimental realization of the diagnostic technique is based on Thomson scattering by the electron density fluctuations [fr

Self-consistent modeling of plasma response to impurity spreading from intense localized source

International Nuclear Information System (INIS)

Koltunov, Mikhail

2012-07-01

Non-hydrogen impurities unavoidably exist in hot plasmas of present fusion devices. They enter it intrinsically, due to plasma interaction with the wall of vacuum vessel, as well as are seeded for various purposes deliberately. Normally, the spots where injected particles enter the plasma are much smaller than its total surface. Under such conditions one has to expect a significant modification of local plasma parameters through various physical mechanisms, which, in turn, affect the impurity spreading. Self-consistent modeling of interaction between impurity and plasma is, therefore, not possible with linear approaches. A model based on the fluid description of electrons, main and impurity ions, and taking into account the plasma quasi-neutrality, Coulomb collisions of background and impurity charged particles, radiation losses, particle transport to bounding surfaces, is elaborated in this work. To describe the impurity spreading and the plasma response self-consistently, fluid equations for the particle, momentum and energy balances of various plasma components are solved by reducing them to ordinary differential equations for the time evolution of several parameters characterizing the solution in principal details: the magnitudes of plasma density and plasma temperatures in the regions of impurity localization and the spatial scales of these regions. The results of calculations for plasma conditions typical in tokamak experiments with impurity injection are presented. A new mechanism for the condensation phenomenon and formation of cold dense plasma structures is proposed.

Relativistic runaway electrons in tokamak plasmas

International Nuclear Information System (INIS)

Jaspers, R.E.

1995-01-01

Runaway electrons are inherently present in a tokamak, in which an electric field is applied to drive a toroidal current. The experimental work is performed in the tokamak TEXTOR. Here runaway electrons can acquire energies of up to 30 MeV. The runaway electrons are studied by measuring their synchrotron radiation, which is emitted in the infrared wavelength range. The studies presented are unique in the sense that they are the first ones in tokamak research to employ this radiation. Hitherto, studies of runaway electrons revealed information about their loss in the edge of the discharge. The behaviour of confined runaways was still a terra incognita. The measurement of the synchrotron radiation allows a direct observation of the behaviour of runaway electrons in the hot core of the plasma. Information on the energy, the number and the momentum distribution of the runaway electrons is obtained. The production rate of the runaway electrons, their transport and the runaway interaction with plasma waves are studied. (orig./HP)

Multilayer mirror based monitors for impurity controls in large fusion reactor type devices

International Nuclear Information System (INIS)

Regan, S.P.; May, M.J.; Soukhanovskii, V.; Finkenthal, M.; Moos, H.W.

1995-01-01

Multilayer Mirror (MLM) based monitors are compact, high throughput diagnostics capable of extracting XUV emissions (the wavelength range including the soft-x-ray and the extreme ultraviolet, 10 angstrom to 304 angstrom) of impurities from the harsh environment of large fusion reactor type devices. For several years the Plasma Spectroscopy Group at Johns Hopkins University has investigated the application of MLM based XUV spectroscopic diagnostics for magnetically confined fusion plasmas. MLM based monitors have been constructed for and extensively used on DIII-D, Alcator C-mod, TEXT, Phaedrus-T, and CDX-U tokamaks to study the impurity behavior of elements ranging from He to Mo. On ITER MLM based devices would be used to monitor the spectral line emissions from Li I-like to F I-like charge states of Fe, Cr, and Ni, as well as extractors for the bands of emissions from high Z elements such as Mo or W for impurity controls of the fusion plasma. In addition to monitoring the impurity emissions from the main plasma, MLM based devices can also be adapted for radiation measurements of low Z elements in the divertor. The concepts and designs of these MLM based monitors for impurity controls in ITER will be presented. The results of neutron irradiation experiments of the MLMs performed in the Los Alamos Spallation Radiation Effects Facility (LASREF) at the Los Alamos National Laboratory will also be discussed. These preliminary neutron exposure studies show that the dispersive and reflective qualities of the MLMs were not affected in a significant manner

Accessibility and replacement as prime constraints in the design of large experimental Tokamaks

International Nuclear Information System (INIS)

Challender, R.S.; Reynolds, P.

1976-01-01

Many of the designs being developed for large Tokamak experiments are based on the classical geometry of the small machines in laboratories throughout the world: a circular array of coils interlinked by an inner toroidal vessel. An attempt is made to bring together those design features which would lead to better accessibility during non-active operation and in particular permit replacement and repair after activation, thereby making possible an extended period of experimental operation into the ignition phase

Quantitative spectroscopy of X-ray lines and continua in tokamaks

International Nuclear Information System (INIS)

Peacock, N.J.; Lawson, K.D.; Patel, A.; Barnsley, R.; Melnick, I.M.; O'Mullane, M.G.; Singleton, M.A.

1996-10-01

Crystal and synthetic multilayer diffractors, deployed either as flat Bragg reflectors, or curved, as in the Johann configuration, are used to study the spectrum of COMPASS-D and other tokamaks in the wavelength region 1-100 A. In this paper we concentrate on the measurement of absolute photon fluxes and the derivation of volume emissivities of the lines and continua in the X-ray region. The sensitivities of these instruments to absolute photon flux have been constructed ab initio from the individual component efficiencies, including published values of the diffractor reflectivities, which have been checked or supplemented by measurements using a double-axis goniometer or from line branching ratios. For these tokamak plasmas where the elemental abundances and effective ion charge are documented, the X-ray continuum intensity itself has been used as a calibration source to derive absolute instrument sensitivity, in reasonable agreement with the ab initio method. In the COMPASS-D tokamak, changes in the effective ion charge state, Z eff , have been derived for different operating conditions, from the absolute intensity of the continuum at ∼4 A. From the irradiances of the line emission, changes in the absolute level of impurities following ''boronisation'' of the vacuum vessel have also been documented. (UK)

The effect of resonant magnetic perturbations on the impurity transport in TEXTOR-DED plasmas

International Nuclear Information System (INIS)

Greiche, Albert Josef

2009-01-01

Thermonuclear fusion provides a new mechanism for the generation of electrical power which has the perspective to serve humanity for several millions of years. One possibility to implement fusion on earth is to magnetically confine hot deuterium tritium plasmas in so called tokamaks. The fusion reactions take place in the hot plasma core. Each of the fusion reactions between deuterium and tritium yields 17.6 MeV which can be used in the process of generating electrical power. Impurities contaminate the plasma which then is cooled down and diluted. This leads to a reduction of the fusion reactions and in consequence the energy yield. The transport behaviour of the impurities in the plasma is not fully understood up to now. Nevertheless, experiments have shown that the application of resonant magnetic perturbations (RMP) can control the impurity content in the plasma. The dynamic ergodic divertor (DED) on the tokamak Textor is able to induce static and dynamic RMPs. During the application of RMPs transient impurity transport experiments with argon have been performed and the time evolution of the impurity concentrations have been monitored. The line emission intensity of the impurities in the plasma is measured in the vacuum ultraviolet (VUV) and in the soft X-ray (SXR) with the absolutely calibrated VUV spectrometer Hexos and SXR PIN diodes, respectively. The analysis of the transient impurity transport experiments is performed with the help of the transport code Strahl. The impurity flows in Strahl are described by a combination of a diffusive and a convective flow. In the computing process the code solves the coupled set of continuity equations of each of the ionization stages of an impurity. With this method the time evolution of the impurity ion densities and the line emission intensities of the ionization stages can be computed. The adaption to the experimental measurements is performed with the help of the diffusion coefficient and the drift velocity which

Comparative studies of stellarator and tokamak transport

Energy Technology Data Exchange (ETDEWEB)

Stroth, U; Burhenn, R; Geiger, J; Giannone, L.; Hartfuss, H J; Kuehner, G; Ledl, L; Simmet, E E; Walter, H [Max-Planck-Inst. fuer Plasmaphysik, IPP-Euratom Association, Garching (Germany); ECRH Team; W7-AS Team

1997-09-01

Transport properties in the W7-AS stellarator and in tokamaks are compared. The parameter dependences and the absolute values of the energy confinement time are similar. Indications are found that the density dependence, which is usually observed in stellarator confinement, can vanish above a critical density. The density dependence in stellarators seems to be similar to that in the linear ohmic confinement regime, which, in small tokamaks, extends to high density values, too. Because of the similarity in the gross confinement properties, transport in stellarators and tokamaks should not be dominated by the parameters which are very different in the two concepts, i.e. magnetic shear, major rational values of the rotational transform and plasma current. A difference in confinement is that there exists evidence for pinches in the particle and, possibly, energy transport channels in tokamaks whereas in stellarators no pinches have been observed, so far. In order to study the effect of plasma current and toroidal electric fields, stellarator discharges were carried out with an increasing amount of plasma current. From these experiments, no clear evidence of a connection of pinches with these parameters is found. The transient response in W7-AS plasmas can be described in terms of a non-local model. As in tokamaks, also cold pulse experiments in W7-AS indicate the importance of non-local transport. (author). 8 refs, 5 figs.

Active beam scattering apparatus and its application to JFT-2 tokamak

International Nuclear Information System (INIS)

Takeuchi, Hiroshi; Matsuda, Toshiaki; Nishitani, Takeo; Shiho, Makoto; Maeda, Hikosuke; Konagai, Chikara; Kimura, Hironobu.

1983-09-01

The capability to assess the ion temperatures using a neutral beam scattering system is investigated on the JFT-2 tokamak. The neutral beam scattering system consists of a 15 KeV neutral hydrogen atom beam and a momentum analyser with silicon surface barrier detectors. The energy analysis of scattered particles on the scattering angle of 4 0 gives the estimation of ion temperatures, which agree well with the one deduced from passive charge-exchange neutral measurements. The influence of impurity ions to the scattering spectrum is not observed and the results of gas scattering experiments suggests that this phenomenon occurs because of the ionization of neutral beam due to the collisions with impurity ions. (author)

Particle fueling and impurity control in PDX

International Nuclear Information System (INIS)

Fonck, R.J.; Bell, M.; Bol, K.

1984-12-01

Fueling requirements and impurity levels in neutral-beam-heated discharges in the PDX tokamak have been compared for plasmas formed with conventional graphite rail limiters, a particle scoop limiter, and an open or closed poloidal divertor. Gas flows necessary to obtain a given density are highest for diverted discharges and lowest for the scoop limiter. Hydrogen pellet injection provides an efficient alternate fueling technique, and a multiple pellet injector has produced high density discharges for an absorbed neutral beam power of up to 600 kW, above which higher speeds or more massive pellets are required for penetration to the plasma core. Power balance studies indicate that 30 to 40% of the total input power is radiated while approx. 15% is absorbed by the limiting surface, except in the open divertor case, where 60% flows to the neutralizer plate. In all operating configurations, Z/sub eff/ usually rises at the onset of neutral beam injection. Both open divertor plasmas and those formed on a well conditioned water-cooled limiter have Z/sub eff/ less than or equal to 2 at the end of neutral injection. A definitive comparison of divertors and limiters for impurity control purposes requires longer beam pulses or higher power levels than available on present machines

Experimental validation of a method for performance monitoring of the impurity processing stage in the TEP system of ITER

International Nuclear Information System (INIS)

Bornschein, B.; Corneli, D.; Glugla, M.; Guenther, K.; Le, T.L.; Simon, K.H.

2007-01-01

The Tokamak Exhaust Processing (TEP) system within the Tritium Plant of ITER needs to be designed such that tritium is recovered from all exhaust gases produced during different modes and operational conditions of the vacuum vessel. The reference process for the TEP system of ITER is called CAPER and comprises three different, consecutive steps to recover hydrogen isotopes at highest purity for direct transfer to the cryogenic Isotope Separation system. The second step ('impurity processing', IP) is carried out in a closed loop involving heterogeneously catalyzed cracking or conversion reactions to liberate tritium from tritiated hydrocarbons or tritiated water combined with permeation of hydrogen isotopes through a Pd/Ag permeator. This combination shifts chemical equilibria towards dehydrogenation and, therefore, enables detritiation factors higher than 1000 in the IP stage. Such a high decontamination factor requires the optimal performance of the permeator, which on the other hand is operated under conditions which provoke coking of the permeator membrane by hydrocarbon cracking. For this reason the permeator in the impurity processing loop needs to be repeatedly regenerated in order to sustain decontamination factors higher/in the order of 1000. At the Tritium Laboratory Karlsruhe (TLK) a method to measure the actual performance of the second stage of the CAPER process has been developed. This method has been successfully tested with the CAPER facility and appears feasible for the TEP system of ITER

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)

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

M.H.D. activity associated with the q=1 surface in the Tore-Supra tokamak; Activite M.H.D. associee a la surface q=1 dans le tokamak Tore-Supra

Energy Technology Data Exchange (ETDEWEB)

Cristofani, P.

1996-02-12

In order to increase the temperature, density and confinement time of the plasma energy inside tokamak devices, several heating and fuel injection techniques have been used. However, the increase of the energy content of the central part of the plasma leads to instabilities in the confinement magnetic structure which can degrade the confinement properties and the temperature performances. Inside the plasma, the ``q=1`` surface plays an important role in the confinement process. The aim of this thesis is to study the experimental physics related to this surface with the analysis of the ``saw-tooth`` periodical internal relaxations and of the ``snake`` structure. The first chapter gives a general introduction about thermonuclear fusion and a description of the plasma and of its equilibrium. Chapter 2 is devoted to the description of the soft X-ray tomography, the diagnostic technique used in this work. In chapter 3, a theoretical presentation of plasma stability and a comparison with experimental results obtained in the Tore-Supra tokamak are given. The observations of saw-tooth instabilities are presented with the principal theoretical models which are used to explain this phenomenon. The snake density instability localized in the central part of the plasma is described in chapter 4 with an attempt of interpretation. The equation of the size evolution of a magnetic island was modified to test different models which can explain the snake stability. One model is based on the modification of the bootstrap current induced by the presence of the snake, and on the local modification of the current induced by the accumulation of impurities inside the snake. (J.S.). 107 refs.

Ion temperature increase during MHD events on the TST-2 spherical tokamak

International Nuclear Information System (INIS)

Ejiri, A.; Shiraiwa, S.; Takase, Y.; Yamada, T.; Nagashima, Y.; Kasahara, H.; Iijima, D.; Kobori, Y.; Nishi, T.; Taniguchi, T.; Aramasu, M.; Ohara, S.; Ushigome, M.; Yamagishi, K.

2003-01-01

Various types of MHD events including internal reconnection events are studied on the TST-2 spherical tokamak. In weak MHD events no positive current spike was observed, but in strong MHD events with positive current spikes, a rapid and significant impurity ion temperature increase was observed. The decrease in the poloidal magnetic energy is the most probable energy source for ion heating. The plasma current shows a stepwise change. The magnitude of this step correlates with the temperature increase and is found to be a good indicator of the strength of each event. (author)

UWMAK-II: a conceptual tokamak reactor design

International Nuclear Information System (INIS)

1975-10-01

This report describes the conceptual design of a Tokamak fusion power reactor, UWMAK-II. The aim of this study is to perform a self consistent and thorough analysis of a probable future fusion power reactor in order to assess the technological problems posed by such a system and to examine feasible solutions. UWMAK-II is a conceptual Tokamak fusion reactor designed to deliver 1716 MWe continuously and to generate 5000 MW(th) during the plasma burn. The structural material is 316 stainless steel and the primary coolant is helium. UWMAK-II is a low aspect ratio, low field design and includes a double null, axisymmetric poloidal field divertor for impurity control. In addition, a carbon curtain, made of two dimensional woven carbon fiber, is mounted on the first vacuum chamber wall to protect the plasma from high Z impurities and to protect the first wall from erosion by charged particle bombardment. The blanket is designed to minimize the inventory of both tritium and lithium while achieving a breeding ratio greater than one. This has led to a blanket design based on the use of a solid breeding material (LiAlO 2 ) with beryllium as a neutron multiplier. The lithium is enriched to 90 percent 6 Li and the blanket coolant is helium at a maximum pressure of 750 psia (5.2 x 10 6 N/m 2 ). A cell of the UWMAK-II blanket design is shown. The breeding ratio is between 1.11 and 1.19 based on one-dimensional discrete ordinates transport calculations, depending on the method of homogenization. Detailed Monte Carlo calculations, which take into account the more complicated geometry, give a breeding ratio of 1.06. The total energy per fusion is 21.56 MeV, which is fairly high

Inward particle transport at high collisionality in the Experimental Advanced Superconducting Tokamak

International Nuclear Information System (INIS)

Wang, G. Q.; Ma, J.; Weiland, J.; Zang, Q.

2013-01-01

We have made the first drift wave study of particle transport in the Experimental Advanced Superconducting Tokamak (Wan et al., Nucl. Fusion 49, 104011 (2009)). The results reveal that collisions make the particle flux more inward in the high collisionality regime. This can be traced back to effects that are quadratic in the collision frequency. The particle pinch is due to electron trapping which is not very efficient in the high collisionality regime so the approach to equilibrium is slow. We have included also the electron temperature gradient (ETG) mode to give the right electron temperature gradient, since the Trapped Electron Mode (TE mode) is weak in this regime. However, at the ETG mode number ions are Boltzmann distributed so the ETG mode does not give particle transport

Experimental studies of fast deuterons, impurity- and admixture-ions emitted from a plasma focus

International Nuclear Information System (INIS)

Mozer, A.; Sadowski, M.; Herold, H.; Schmidt, H.

1982-01-01

The energy and mass analysis of ions emitted from a 50-kJ, 18-kV, plasma focus machine was performed with a Thomson analyzer. Energy distribution functions of fast deuterons (E> or =350 keV) and those of impurity ions have been determined. The energy distributions of the O, N, and C impurity ions in different ionization states have similar character. They usually increase exponentially and after reaching the maximum at E/Zroughly-equal1.0 MeV they decrease exponentially to E/Zroughly-equal1.8 MeV. For deuterons at lower operating pressures (p 0 + -Ar 7+ ions of energy from 0.5 to 14 MeV are produced

Engineering solutions for components facing the plasma in experimental power reactors

International Nuclear Information System (INIS)

Casini, G.; Farfaletti-Casali, F.

1985-01-01

A review of the engineering problems related to the structures in front of the plasma of experimental Tokamak-type reactors is made. Attention is focused on the so-named ''first wall'', i.e. the wall side of the blanket segments facing the plasma, and on the collector plates of the impurity control system, in particular for the case of the single-null poloidal divertor. Even if the uncertainties related to the plasma-wall interaction are stil relevant, some engineering solutions which look manageable are identified and described. (orig.)

The diagnostic neutral beam injector with arc-discharge plasma source on the TCV Tokamak

Energy Technology Data Exchange (ETDEWEB)

Karpushov, Alexander N. [Ecole Polytechnique Federale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, Association Euratom-Confederation Suisse, CH-1015 Lausanne (Switzerland)], E-mail: alexander.karpushov@epfl.ch; Andrebe, Yanis; Duval, Basil P.; Bortolon, Alessandro [Ecole Polytechnique Federale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas, Association Euratom-Confederation Suisse, CH-1015 Lausanne (Switzerland)

2009-06-15

The diagnostic neutral beam injector (DNBI) together with a charge exchange recombination spectroscopy (CXRS) system has been used on the TCV Tokamak as a diagnostic tool for local measurements of plasma ion temperature, velocity and carbon impurity density based on analysis of the beam induced impurity radiation emission since 2000. To improve the performance of the CXRS diagnostic, several upgrades of both the optical system and the neutral beam were performed. An increase of the plasma source size together with beam optimization in 2003 resulted in a twofold increase the beam current. The RF plasma generator was replaced by an arc-discharge plasma source together with a new ion optical system (IOS) in 2006 and subsequent beam optimization is presented herein. This was designed to increase the line brightness of the beam in the CXRS observation region without increasing of the injected power (to avoid plasma perturbation by the beam). The beam characteristics are measured by a multi-chord scanning of Doppler-shifted H{sub {alpha}} emission, thermal measurements on a movable calorimeter and visible optical measurements inside the Tokamak vessel.

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

Physics aspects of the Compact Ignition Tokamak

International Nuclear Information System (INIS)

Post, D.; Bateman, G.; Houlberg, W.

1986-11-01

The Compact Ignition Tokamak (CIT) is a proposed modest-size ignition experiment designed to study the physics of alpha-particle heating. The basic concept is to achieve ignition in a modest-size minimum cost experiment by using a high plasma density to achieve the condition of ntau/sub E/ ∼ 2 x 10 20 sec m -3 required for ignition. The high density requires a high toroidal field (10 T). The high toroidal field allows a large plasma current (10 MA) which improves the energy confinement, and provides a high level of ohmic heating. The present CIT design also has a gigh degree of elongation (k ∼ 1.8) to aid in producing the large plasma current. A double null poloidal divertor and a pellet injector are part of the design to provide impurity and particle control, improve the confinement, and provide flexibility for impurity and particle control, improve the confinement, and provide flexibility for improving the plasma profiles. Since auxiliary heating is expected to be necessary to achieve ignition, 10 to 20 MW of Ion Cyclotron Radio Frequency (ICRF) is to be provided

Impurity-induced moments in underdoped cuprates

International Nuclear Information System (INIS)

Khaliullin, G.; Kilian, R.; Krivenko, S.; Fulde, P.

1997-01-01

We examine the effect of a nonmagnetic impurity in a two-dimensional spin liquid in the spin-gap phase, employing a drone-fermion representation of spin-1/2 operators. The properties of the local moment induced in the vicinity of the impurity are investigated and an expression for the nuclear-magnetic-resonance Knight shift is derived, which we compare with experimental results. Introducing a second impurity into the spin liquid an antiferromagnetic interaction between the moments is found when the two impurities are located on different sublattices. The presence of many impurities leads to a screening of this interaction as is shown by means of a coherent-potential approximation. Further, the Kondo screening of an impurity-induced local spin by charge carriers is discussed. copyright 1997 The American Physical Society

Edge and coupled core/edge transport modelling in tokamaks

International Nuclear Information System (INIS)

Lodestro, L.L.; Casper, T.A.; Cohen, R.H.

1999-01-01

Recent advances in the theory and modelling of tokamak edge, scrape-off-layer (SOL) and divertor plasmas are described. The effects of the poloidal E x B drift on inner/outer divertor-plate asymmetries within a 1D analysis are shown to be in good agreement with experimental trends; above a critical v ExB , the model predicts transitions to supersonic flow at the inboard midplane. 2D simulations show the importance of E x B flow in the private-flux region and of ∇ B-drifts. A theory of rough plasma-facing surfaces is given, predicting modifications to the SOL plasma. The parametric dependence of detached-plasma states in slab geometry has been explored; with sufficient pumping, the location of the ionization front can be controlled; otherwise only fronts near the plate or the X-point are stable. Studies with a more accurate Monte-Carlo neutrals model and a detailed non-LTE radiation-transport code indicate various effects are important for quantitative modelling. Detailed simulations of the DIII-D core and edge are presented; impurity and plasma flow are discussed and shown to be well modelled with UEDGE. (author)

Edge and coupled core-edge transport modelling in tokamaks

International Nuclear Information System (INIS)

Lodestro, L.L.; Casper, T.A.; Cohen, R.H.

2001-01-01

Recent advances in the theory and modelling of tokamak edge, scrape-off-layer (SOL) and divertor plasmas are described. The effects of the poloidal ExB drift on inner/outer divertor-plate asymmetries within a 1D analysis are shown to be in good agreement with experimental trends; above a critical v ExB, the model predicts transitions to supersonic SOL flow at the inboard midplane. 2D simulations show the importance of ExB flow in the private-flux region and of ∇ B-drifts. A theory of rough plasma-facing surfaces is given, predicting modifications to the SOL plasma. The parametric dependence of detached-plasma states in slab geometry has been explored; with sufficient pumping, the location of the ionization front can be controlled; otherwise only fronts near the plate or the X-point are stable. Studies with a more accurate Monte-Carlo neutrals model and a detailed non-LTE radiation-transport code indicate various effects are important for quantitative modelling. Detailed simulations of the DIII-D core and edge are presented; impurity and plasma flow are discussed and shown to be well modelled with UEDGE. (author)

Integral torque balance in tokamaks

International Nuclear Information System (INIS)

Pustovitov, V.D.

2011-01-01

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

Coherent structures in tokamak plasmas workshop: Proceedings

International Nuclear Information System (INIS)

Koniges, A.E.; Craddock, G.G.

1992-08-01

Coherent structures have the potential to impact a variety of theoretical and experimental aspects of tokamak plasma confinement. This includes the basic processes controlling plasma transport, propagation and efficiency of external mechanisms such as wave heating and the accuracy of plasma diagnostics. While the role of coherent structures in fluid dynamics is better understood, this is a new topic for consideration by plasma physicists. This informal workshop arose out of the need to identify the magnitude of structures in tokamaks and in doing so, to bring together for the first time the surprisingly large number of plasma researchers currently involved in work relating to coherent structures. The primary purpose of the workshop, in addition to the dissemination of information, was to develop formal and informal collaborations, set the stage for future formation of a coherent structures working group or focus area under the heading of the Tokamak Transport Task Force, and to evaluate the need for future workshops on coherent structures. The workshop was concentrated in four basic areas with a keynote talk in each area as well as 10 additional presentations. The issues of discussion in each of these areas was as follows: Theory - Develop a definition of structures and coherent as it applies to plasmas. Experiment - Review current experiments looking for structures in tokamaks, discuss experimental procedures for finding structures, discuss new experiments and techniques. Fluids - Determine how best to utilize the resource of information available from the fluids community both on the theoretical and experimental issues pertaining to coherent structures in plasmas. Computation - Discuss computational aspects of studying coherent structures in plasmas as they relate to both experimental detection and theoretical modeling

Some problems of a tokamak with fast-neutral injection

International Nuclear Information System (INIS)

Pistunovich, V.I.

1976-01-01

An attempt has been made to formulate the fundamental physical problems which should be solved for designing a reactor on the basis of a tokamak with injection. The problems enumerated should be considered particularly for the suggested system, though qualitative solutions of some of them have been already known. They are the following. 1) Ionization of the atomic beam and distribution of the generated ion density. 2) Capture of fast ions by the tokamak magnetic field. 3) Anisotropy of the velocity distribution function. 4) The effect of the neutral gas density on the shape of the one-dimensional distribution function. 5)Stability of the ion distribution function on deceleration of the ion beam in plasma. The paper permits evaluating the advantages and shortcomings of the reactor on the basis of a tokamak with injection from viewpoints of the requirements on confinement and heating of plasma, and of the effect of impurities and neutral working gas on the output parameters of the reactor. The author believes that the regime of a two-component tokamak with small values of quality Q (<=) 3 shows greatest advantages over the ignition reactor, though in this case a reactor may be economically profitable (Q approximately 10) only with an increase of the total quality of the reactor by additional energy release in a blanket on using fission reactions