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.)

The influence of impurity and particle control on TMX-U [Tandem Mirror Experiment Upgrade] plasma operation

International Nuclear Information System (INIS)

Allen, S.L.; Yu, T.L.; Foote, J.H.; Pickles, W.L.

1985-11-01

A variety of techniques are used in TMX-U to control impurities and reflux: repeated plasma pulses, glow discharge cleaning (GDC), and gettering. A series of experiments under three different plasma-wall conditions was performed: no wall conditioning after a machine maintenance cycle, a glow-discharge-cleaned wall, and a gettered wall. Several plasma diagnostics to determine the effect of these procedures on TMX-U plasma parameters were used. Spectroscopic measurements indicated that GDC reduced impurities and increased the electron temperature, enabling full-duration beam-sustained plasma operation without a large number of repeated plasma pulses. Gettering further reduced the impurities and the neutral pressure, and this improved condition persisted for several shots after gettering was stopped. Measurements from residual gas analyzers and an end-loss ion spectrometer indicated that hydrogen is present in the plasma during the initial deuterium operation after pumpdown; the hydrogen level decreased after plasma operation with gettering, indicating reduced wall recycling

Progress and improvement of KSTAR plasma control using model-based control simulators

Energy Technology Data Exchange (ETDEWEB)

Hahn, Sang-hee, E-mail: hahn76@nfri.re.kr [National Fusion Research Institute, 169-148 Gwahak-ro, yuseong-gu, Daejeon (Korea, Republic of); Welander, A.S. [General Atomics, San Diego, CA (United States); Yoon, S.W.; Bak, J.G. [National Fusion Research Institute, 169-148 Gwahak-ro, yuseong-gu, Daejeon (Korea, Republic of); Eidietis, N.W. [General Atomics, San Diego, CA (United States); Han, H.S. [National Fusion Research Institute, 169-148 Gwahak-ro, yuseong-gu, Daejeon (Korea, Republic of); Humphreys, D.A.; Hyatt, A. [General Atomics, San Diego, CA (United States); Jeon, Y.M. [National Fusion Research Institute, 169-148 Gwahak-ro, yuseong-gu, Daejeon (Korea, Republic of); Johnson, R.D. [General Atomics, San Diego, CA (United States); Kim, H.S.; Kim, J. [National Fusion Research Institute, 169-148 Gwahak-ro, yuseong-gu, Daejeon (Korea, Republic of); Kolemen, E.; Mueller, D. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Penaflor, B.G.; Piglowski, D.A. [General Atomics, San Diego, CA (United States); Shin, G.W. [University of Science and Technology, Daejeon (Korea, Republic of); Walker, M.L. [General Atomics, San Diego, CA (United States); Woo, M.H. [National Fusion Research Institute, 169-148 Gwahak-ro, yuseong-gu, Daejeon (Korea, Republic of)

2014-05-15

Superconducting tokamaks like KSTAR, EAST and ITER need elaborate magnetic controls mainly due to either the demanding experiment schedule or tighter hardware limitations caused by the superconducting coils. In order to reduce the operation runtime requirements, two types of plasma simulators for the KSTAR plasma control system (PCS) have been developed for improving axisymmetric magnetic controls. The first one is an open-loop type, which can reproduce the control done in an old shot by loading the corresponding diagnostics data and PCS setup. The other one, a closed-loop simulator based on a linear nonrigid plasma model, is designed to simulate dynamic responses of the plasma equilibrium and plasma current (I{sub p}) due to changes of the axisymmetric poloidal field (PF) coil currents, poloidal beta, and internal inductance. The closed-loop simulator is the one that actually can test and enable alteration of the feedback control setup for the next shot. The simulators have been used routinely in 2012 plasma campaign, and the experimental performances of the axisymmetric shape control algorithm are enhanced. Quality of the real-time EFIT has been enhanced by utilizations of the open-loop type. Using the closed-loop type, the decoupling scheme of the plasma current control and axisymmetric shape controls are verified through both the simulations and experiments. By combining with the relay feedback tuning algorithm, the improved controls helped to maintain the shape suitable for longer H-mode (10–16 s) with the number of required commissioning shots largely reduced.

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

Control of first-wall surface conditions in the 2XIIB Magnetic Mirror Plasma Confinement experiment

International Nuclear Information System (INIS)

Simonen, T.C.; Bulmer, R.H.; Coensgen, F.H.

1976-01-01

The control of first-wall surface conditions in the 2XIIB Magnetic Mirror Plasma Confinement experiment is described. Before each plasma shot, the first wall is covered with a freshly gettered titanium surface. Up to 5 MW of neutral beam power has been injected into 2XIIB, resulting in first-wall bombardment fluxes of 10 17 atoms . cm -2 . s -1 of 13-keV mean energy deuterium atoms for several ms. The background gas flux is measured with a calibrated, 11-channel, fast-atom detector. Background gas levels are found to depend on surface conditions, injected beam current, and beam pulse duration. For our best operating conditions, an efective reflex coefficient of 0.3 can be inferred from the measurements. Experiments with long-duration and high-current beam injection are limited by charge exchange; however, experiments with shorter beam duration are not limited by first-wall surface conditions. It is concluded that surface effects will be reduced further with smoother walls. (Auth.)

Controlled fusion and plasma physics

International Nuclear Information System (INIS)

1994-07-01

40 papers are presented at this 21. conference on controlled fusion and plasma physics (JET). Titles are: effects of sawtooth crashes on beams ions and fusion product tritons; beta limits in H-modes and VH-modes; impurity induced neutralization of MeV energy protons in JET plasmas; lost α particle diagnostic for high-yield D-T fusion plasmas; 15-MeV proton emission from ICRF-heated plasmas; pulse compression radar reflectometry for density measurements; gamma-ray emission profile measurements during ICRH discharges; the new JET phase ICRH array; simulation of triton burn-up; parametric dependencies of JET electron temperature profiles; detached divertor plasmas; excitation of global Alfven Eigenmodes by RF heating; mechanisms of toroidal rotation; effect of shear in the radial electric field on confinement; plasma transport properties at the L-H transition; numerical study of plasma detachment conditions in JET divertor plasmas; the SOL width and the MHD interchange instability; non linear magnetic reconnection in low collisionality plasmas; topology and slowing down of high energy ion orbits; sawtooth crashes at high beta; fusion performances and alpha heating in future JET D-T plasmas; a stable route to high-beta plasmas with non-monotonic q-profiles; theory of propagation of changes to confinement; spatial distribution of gamma emissivity and fast ions during ICRF heating; multi-camera soft X-ray diagnostic; radiation phenomena and particle fluxes in the X-event; local measurement of transport parameters for laser injected trace impurities; impurity transport of high performance discharges; negative snakes and negative shear; neural-network charge exchange analysis; ion temperature anisotropy in helium neutral beam fuelling; impurity line emission due to thermal charge exchange in edge plasmas; control of convection by fuelling and pumping; VH mode accessibility and global H-mode properties; ion cyclotron emission by spontaneous emission; LHCD/ICRH synergy

Advances in integrated plasma control on DIII-D

International Nuclear Information System (INIS)

Walker, M.L.; Ferron, J.R.; Hahn, S.H.; Humphreys, D.A.; In, Y.; Johnson, R.D.; Kim, J.S.; La Haye, R.J.; Leuer, J.A.; Penaflor, B.G.; Welander, A.S.; Xiao, B.

2007-01-01

The DIII-D advanced tokamak physics program requires extremely high performance from the DIII-D plasma control system, including simultaneous accurate regulation of plasma shape, stored energy, density and divertor characteristics, as well as coordinated suppression of magnetohydrodynamic instabilities. To satisfy these demanding control requirements, we apply the integrated plasma control method, consisting of construction of physics-based plasma and system response models, validation of models against operating experiments, design of integrated controllers that operate in concert with one another, simulation of control action against off-line and actual machine control platforms, and optimization through iteration of the design-test loop. The present work describes progress in development of physics models and development and experimental application of new model-based plasma controllers on DIII-D. We also describe the development of the control software, hardware, and model-based control algorithms for the superconducting EAST and KSTAR tokamaks

Magnum-PSI: A new plasma-wall interaction experiment

International Nuclear Information System (INIS)

Koppers, W.; Eck, H. van; Scholten, J.

2006-01-01

The FOM-Institute for Plasma Physics Rijnhuizen is preparing the construction of Magnum-PSI, a magnetized (3 T), steady-state, large area (diameter 10 cm), high-flux plasma (10 24 ions m -2 s -1 generator. The aim of the linear plasma device Magnum-PSI is to provide a controlled, highly accessible laboratory experiment in which the interaction of a magnetized plasma with different surfaces can be studied in detail. Plasma parameters can be varied over a wide range, in particular covering the high-density, low-temperature conditions expected for the detached divertor plasma of ITER. The target set-up will be extremely flexible allowing the investigation of different materials under a large variety of conditions (temperatures, inclination, biasing, coatings, etc.). A range of target materials will be used, including carbon, tungsten and other metals, and mixed materials. Because of the large plasma beam of 10 cm diameter and spacious vacuum tank, even the test of whole plasma-facing component mock-ups will be possible. Dedicated diagnostics will be installed to allow for detailed studies of the fundamental physics and chemistry of plasma-surface interaction, such as erosion and deposition, hydrogen recycling, retention and removal, dust and layer formation, plasma sheath physics and heat loads (steady-state or transient). Magnum-PSI will be a unique experiment to address the ITER divertor physics which will essentially differ from present day Tokamak and/or linear plasma generator physics. In this contribution, we will present the pre-design of the Magnum-PSI experiment. We will discuss the requirements on the vacuum system, 3T superconducting magnet, plasma source, target manipulator and additional plasma heating. In addition, we will briefly introduce the plasma and surface diagnostics that will be used in the Magnum-PSI experiment. (author)

Development of the 'JFT-2' tokamak plasma position control system

International Nuclear Information System (INIS)

Fujisawa, Noboru; Matsuzaki, Yoshimi; Suzuki, Norio; Murai, Katsuji; Suzuki, Satoshi.

1980-01-01

Digital control technique was applied to control the plasma position in the JFT-2 tokamak experiment device. The detail of the JFT-2 is described elsewhere. The plasma position control system consists of a Hitachi control computer, HIDIC 80, and a Hitachi micro-computer, HIDIC 08E. The plasma position is detected by the position control computer, and compared with a preset value. Then, a reference signal is supplied to the micro-computer controlling power source, and the phase control of the thyristor controlling power source is performed. Since the behavior of plasma is very fast, the fast control is required. The control of the thyristor controlling power source is made by direct digital control (DDC). The main component of the hardware of the present system is the micro-computer HIDIC 08E. The software is the direct task system without the operating system (OS). The results of experiments showed that the feedback control of the system worked well. (Kato, T.)

Plasma wave observations during electron and ion gun experiments

International Nuclear Information System (INIS)

Olsen, R.C.; Lowery, D.R.; Weddle, L.E.

1988-01-01

Plasma wave instruments with high temporal and frequency resolution in the 0-6 kHz frequency range have been used to monitor electron gun-employing charge control experiments with the USAF/NASA p78-2 satellite, in order to determine whether plasma wave signatures consistent with the previous inference of electron heating were present. Strong plasma waves were noted near the electron gyrofrequency; these waves can heat ambient low energy electrons, as previously inferred. Attention is given to the two distinct classes of behavior revealed by the ion gun experiments. 16 references

INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS

International Nuclear Information System (INIS)

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

2004-03-01

OAK-B135 Advanced tokamaks (AT) are distinguished from conventional tokamaks by their high degree of shaping, achievement of profiles optimized for high confinement and stability characteristics, and active stabilization of MHD instabilities to attain high values of normalized beta and confinement. These high performance fusion devices thus require accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating, as well as simultaneous and well-coordinated MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Satisfying the simultaneous demands on control accuracy, reliability, and performance for all of these subsystems requires a high degree of integration in both design and operation of the plasma control system in an advanced tokamak. The present work describes the approach, benefits, and progress made in integrated plasma control with application examples drawn from the DIII-D tokamak. The approach includes construction of plasma and system response models, validation of models against operating experiments, design of integrated controllers which operate in concert with one another as well as with supervisory modules, simulation of control action against off-line and actual machine control platforms, and iteration of the design-test loop to optimize performance

Trends in laser-plasma-instability experiments for laser fusion

International Nuclear Information System (INIS)

Drake, R.P.

1991-01-01

Laser-plasma instability experiments for laser fusion have followed three developments. These are advances in the technology and design of experiments, advances in diagnostics, and evolution of the design of high-gain targets. This paper traces the history of these three topics and discusses their present state. Today one is substantially able to produce controlled plasma conditions and to diagnose specific instabilities within such plasmas. Experiments today address issues that will matter for future laser facilities. Such facilities will irradiate targets with ∼1 MJ of visible or UV light pulses that are tens of nanoseconds in duration, very likely with a high degree of spatial and temporal incoherence. 58 refs., 4 figs

Profile control simulations and experiments on TCV: a controller test environment and results using a model-based predictive controller

Science.gov (United States)

Maljaars, E.; Felici, F.; Blanken, T. C.; Galperti, C.; Sauter, O.; de Baar, M. R.; Carpanese, F.; Goodman, T. P.; Kim, D.; Kim, S. H.; Kong, M.; Mavkov, B.; Merle, A.; Moret, J. M.; Nouailletas, R.; Scheffer, M.; Teplukhina, A. A.; Vu, N. M. T.; The EUROfusion MST1-team; The TCV-team

2017-12-01

The successful performance of a model predictive profile controller is demonstrated in simulations and experiments on the TCV tokamak, employing a profile controller test environment. Stable high-performance tokamak operation in hybrid and advanced plasma scenarios requires control over the safety factor profile (q-profile) and kinetic plasma parameters such as the plasma beta. This demands to establish reliable profile control routines in presently operational tokamaks. We present a model predictive profile controller that controls the q-profile and plasma beta using power requests to two clusters of gyrotrons and the plasma current request. The performance of the controller is analyzed in both simulation and TCV L-mode discharges where successful tracking of the estimated inverse q-profile as well as plasma beta is demonstrated under uncertain plasma conditions and the presence of disturbances. The controller exploits the knowledge of the time-varying actuator limits in the actuator input calculation itself such that fast transitions between targets are achieved without overshoot. A software environment is employed to prepare and test this and three other profile controllers in parallel in simulations and experiments on TCV. This set of tools includes the rapid plasma transport simulator RAPTOR and various algorithms to reconstruct the plasma equilibrium and plasma profiles by merging the available measurements with model-based predictions. In this work the estimated q-profile is merely based on RAPTOR model predictions due to the absence of internal current density measurements in TCV. These results encourage to further exploit model predictive profile control in experiments on TCV and other (future) tokamaks.

Plasma boundary considerations for the national compact stellarator experiment

International Nuclear Information System (INIS)

Mioduszewski, P.; Grossman, A.; Fenstermacher, M.; Koniges, A.; Owen, L.; Rognlien, T.; Umansky, M.

2003-01-01

The national compact stellarator experiment (NCSX) [EPS 2001, Madeira, Portugal, 18-22 June 2001] is a new fusion project located at Princeton Plasma Physics Laboratory, Princeton, NJ. Plasma boundary control in stellarators has been shown to be very effective in improving plasma performance [EPS 2001, Madeira, Portugal, 18-22 June 2001] and, accordingly, will be an important element from the very beginning of the NCSX design. Plasma-facing components will be developed systematically according to our understanding of the NCSX boundary, with the eventual goal to develop a divertor with all the benefits for impurity and neutrals control. Neutrals calculations have been started to investigate the effect of neutrals penetration at various cross-sections

Using the Tritium Plasma Experiment to evaluate ITER PFC safety

International Nuclear Information System (INIS)

Longhurst, G.R.; Anderl, R.A.; Bartlit, J.R.; Causey, R.A.; Haines, J.R.

1993-01-01

The Tritium Plasma Experiment was assembled at Sandia National Laboratories, Livermore to investigate interactions between dense plasmas at low energies and plasma-facing component materials. This apparatus has the unique capability of replicating plasma conditions in a tokamak divertor with particle flux densities of 2 x 10 19 ions/cm 2 · s and a plasma temperature of about 15 eV using a plasma that includes tritium. With the closure of the Tritium Research Laboratory at Livermore, the experiment was moved to the Tritium Systems Test Assembly facility at Los Alamos National Laboratory. An experimental program has been initiated there using the Tritium Plasma Experiment to examine safety issues related to tritium in plasma-facing components, particularly the ITER divertor. Those issues include tritium retention and release characteristics, tritium permeation rates and transient times to coolant streams, surface modification and erosion by the plasma, the effects of thermal loads and cycling, and particulate production. A considerable lack of data exists in these areas for many of the materials, especially beryllium, being considered for use in ITER. Not only will basic material behavior with respect to safety issues in the divertor environment be examined, but innovative techniques for optimizing performance with respect to tritium safety by material modification and process control will be investigated. Supplementary experiments will be carried out at the Idaho National Engineering Laboratory and Sandia National Laboratory to expand and clarify results obtained on the Tritium Plasma Experiment

Plasma Surface interaction in Controlled fusion devices

International Nuclear Information System (INIS)

1990-05-01

The subjects presented in the 9th conference on plasma surface interaction in controlled fusion devices were: the modifications of power scrape-off-length and power deposition during various configurations in Tore Supra plasmas; the effects observed in ergodic divertor experiments in Tore-Supra; the diffuse connexion induced by the ergodic divertor and the topology of the heat load patterns on the plasma facing components in Tore-Supra; the study of the influence of air exposure on graphite implanted by low energy high density deuterium plasma

Review of recent experiments on magnetic reconnection in laboratory plasmas

International Nuclear Information System (INIS)

Yamada, M.

1995-02-01

The present paper reviews recent laboratory experiments on magnetic reconnection. Examples will be drawn from electron current sheet experiments, merging spheromaks, and from high temperature tokamak plasmas with the Lundquist numbers exceeding 10 7 . These recent laboratory experiments create an environment which satisfies the criteria for MHD plasma and in which the global boundary conditions can be controlled externally. Experiments with fully three dimensional reconnection are now possible. In the most recent TFTR tokamak discharges, Motional Stark effect (MSE) data have verified the existence of a partial reconnection. In the experiment of spheromak merging, a new plasma acceleration parallel to the neutral line has been indicated. Together with the relationship of these observations to the analysis of magnetic reconnection in space and in solar flares, important physics issues such as global boundary conditions, local plasma parameters, merging angle of the field lines, and the 3-D aspects of the reconnection are discussed

A large volume uniform plasma generator for the experiments of electromagnetic wave propagation in plasma

International Nuclear Information System (INIS)

Yang Min; Li Xiaoping; Xie Kai; Liu Donglin; Liu Yanming

2013-01-01

A large volume uniform plasma generator is proposed for the experiments of electromagnetic (EM) wave propagation in plasma, to reproduce a “black out” phenomenon with long duration in an environment of the ordinary laboratory. The plasma generator achieves a controllable approximate uniform plasma in volume of 260 mm× 260 mm× 180 mm without the magnetic confinement. The plasma is produced by the glow discharge, and the special discharge structure is built to bring a steady approximate uniform plasma environment in the electromagnetic wave propagation path without any other barriers. In addition, the electron density and luminosity distributions of plasma under different discharge conditions were diagnosed and experimentally investigated. Both the electron density and the plasma uniformity are directly proportional to the input power and in roughly reverse proportion to the gas pressure in the chamber. Furthermore, the experiments of electromagnetic wave propagation in plasma are conducted in this plasma generator. Blackout phenomena at GPS signal are observed under this system and the measured attenuation curve is of reasonable agreement with the theoretical one, which suggests the effectiveness of the proposed method.

Chaos control and taming of turbulence in plasma devices

DEFF Research Database (Denmark)

Klinger, T.; Schröder, C.; Block, D.

2001-01-01

Chaos and turbulence are often considered as troublesome features of plasma devices. In the general framework of nonlinear dynamical systems, a number of strategies have been developed to achieve active control over complex temporal or spatio-temporal behavior. Many of these techniques apply...... to plasma instabilities. In the present paper we discuss recent progress in chaos control and taming of turbulence in three different plasma "model" experiments: (1) Chaotic oscillations in simple plasma diodes, (2) ionization wave turbulence in the positive column of glow discharges, and (3) drift wave...

Control of plasma poloidal shape and position in the DIII-D tokamak

International Nuclear Information System (INIS)

Walker, M.L.; Humphreys, D.A.; Ferron, J.R.

1997-11-01

Historically, tokamak control design has been a combination of theory driving an initial control design and empirical tuning of controllers to achieve satisfactory performance. This approach was in line with the focus of past experiments on simply obtaining sufficient control to study many of the basic physics issues of plasma behavior. However, in recent years existing experimental devices have required increasingly accurate control. New tokamaks such as ITER or the eventual fusion power plant must achieve and confine burning fusion plasmas, placing unprecedented demands on regulation of plasma shape and position, heat flux, and burn characteristics. Control designs for such tokamaks must also function well during initial device operation with minimal empirical optimization required. All of these design requirements imply a heavy reliance on plasma modeling and simulation. Thus, plasma control design has begun to use increasingly modern and sophisticated control design methods. This paper describes some of the history of plasma control for the DIII-D tokamak as well as the recent effort to implement modern controllers. This effort improves the control so that one may obtain better physics experiments and simultaneously develop the technology for designing controllers for next-generation tokamaks

Current status of DIII-D real-time digital plasma control

International Nuclear Information System (INIS)

Penaflor, B.G.; Piglowski, D.A.; Ferron, J.R.; Walker, M.L.

1999-06-01

This paper describes the current status of real-time digital plasma control for the DIII-D tokamak. The digital plasma control system (PCS) has been in place at DIII-D since the early 1990s and continues to expand and improve in its capabilities to monitor and control plasma parameters for DIII-D fusion science experiments. The PCs monitors over 200 tokamak parameters from the DIII-D experiment using a real-time data acquisition system that acquires a new set of samples once every 60 micros. This information is then used in a number of feedback control algorithms to compute and control a variety of parameters including those affecting plasma shape and position. A number of system related improvements has improved the usability and flexibility of the DIII-D PCS. These include more graphical user interfaces to assist in entering and viewing the large and ever growing number of parameters controlled by the PCS, increased interaction and accessibility from other DIII-D applications, and upgrades to the computer hardware and vended software. Future plans for the system include possible upgrades of the real-time computers, further links to other DIII-D diagnostic measurements such as real-time Thomson scattering analysis, and joint collaborations with other tokamak experiments including the NSTX at Princeton

Plasma production for the 50 MeV plasma lens experiment at LBL

International Nuclear Information System (INIS)

Leemans, W.; van der Geer, B.; de Loos, M.; Conde, M.; Govil, R.; Chattopadhyay, S.

1994-06-01

The Center for Beam Physics at LBL has constructed a Beam Test Facility (BTF) housing a 50 MeV electron beam transport line, which uses the linac injector from the Advanced Light Source, and a terawatt Ti:Al 2 O 3 laser system. The linac operates at 50 MeV and generates 15 ps long electron bunches containing a charge of up to 2 nC. The measured unnormalized beam emittance is 0.33 mm-mrad. These parameters allow for a comprehensive study of focusing of relativistic electron beams with plasma columns, in both the overdense and underdense regime (adiabatic and tapered lenses). A study of adiabatic and/or tapered lenses requires careful control of plasma density and scale lengths of the plasma. We present experimental results on the production of plasmas through resonant two-photon ionization, with parameters relevant to an upcoming plasma lens experiment

Plasma Edge Control in Tore Supra

International Nuclear Information System (INIS)

Evans, T.E.; Mioduszewski, P.K.; Foster, C.; Haste, G.; Horton, L.; Grosman, A.; Ghendrih, P.; Chatelier, M.; Capes, H.; Michelis, C. De; Fall, T.; Geraud, A.; Grisolia, C.; Guilhem, D.; Hutter, T.

1990-01-01

TORE SUPRA is a large superconducting tokamak designed for sustaining long inductive pulses (t∼ 30 s). In particular, all the first wall components have been designed for steady-state heat and particle exhaust, particle injection, and additional heating. In addition to these technological assets, a strict control of the plasma-wall interactions is required. This has been done at low power: experiments with ohmic heating have been mainly devoted to the pump limiter, ergodic divertor and pellet injection experiments. Some specific problems arising in large tokamaks are encountered; the pump limiter and the ergodic divertor yield the expected effects on the plasma edge. The effects on the bulk are discussed

Plasma position control device

International Nuclear Information System (INIS)

Takase, Haruhiko.

1987-01-01

Purpose: To conduct position control stably to various plasmas and reduce the burden on the control coil power source. Constitution: Among the proportional, integration and differentiation controls, a proportional-differentiation control section and an integration control section are connected in parallel. Then, a signal switching circuit is disposed to the control signal input section for the proportional-differentiation control section such that either a present position of plasmas or deviation between the present plasma position and an aimed value can be selected as a control signal depending on the control procedures or the state of the plasmas. For instance, if a rapid response is required for the control, the deviation between the present plasma position and the aimed value is selected as the input signal to conduct proportional, integration and differentiation controls. While on the other hand, if it is intended to reduce the burden on the control coil power source, it is adapted such that the control signal inputted to the proportional-differentiation control section itself can select the present plasma position. (Yoshihara, H.)

Boundary plasma control with the ergodic divertor

International Nuclear Information System (INIS)

Ghendrih, Ph.; Becoulet, M.; Beyer, P.

1999-01-01

Ergodic divertor experiments on Tore Supra provide evidence of significant control of plasma-wall interaction. Theoretical investigation of the laminar region (i.e. governed by parallel transport) indicates that control of the plasma state at the target plate can be achieved with plasma states similar to that observed with the axisymmetric divertor. Analysis of the temperature field with a 2-D test particle code allows one to recover the observed spatial modulation and shows that an intrinsic barrier appears to develop at the separatrix. Energy deposition peaking, analysed with a 3-D code, is strongly reduced when moderate transverse transport is considered. Possible control of upstream parameters can thus be achieved in the ergodic region, for instance a lowering of the parallel energy flux by cross field transport. (author)

Boundary plasma control with the ergodic divertor

International Nuclear Information System (INIS)

Ghendrih, Ph.; Becoulet, M.; Beyer, P.

2001-01-01

Ergodic divertor experiments on Tore Supra provide evidence of significant control of plasma-wall interaction. Theoretical investigation of the laminar region (i.e. governed by parallel transport) indicates that control of the plasma state at the target plate can be achieved with plasma states similar to that observed with the axisymmetric divertor. Analysis of the temperature field with a 2-D test particle code allows one to recover the observed spatial modulation and shows that an intrinsic barrier appears to develop at the separatrix. Energy deposition peaking, analysed with a 3-D code, is strongly reduced when moderate transverse transport is considered. Possible control of upstream parameters can thus be achieved in the ergodic region, for instance a lowering of the parallel energy flux by cross field transport. (author)

Chapter 8: Plasma operation and control [Progress in the ITER Physics Basis (PIPB)

International Nuclear Information System (INIS)

Gribov, Y.; Humphreys, D.; Kajiwara, K.; Lazarus, E.A.; Lister, J.B.; Ozeki, T.; Portone, A.; Shimada, M.; Sips, A.C.C.; Wesley, J.C.

2007-01-01

The ITER plasma control system has the same functional scope as the control systems in present tokamaks. These are plasma operation scenario sequencing, plasma basic control (magnetic and kinetic), plasma advanced control (control of RWMs, NTMs, ELMs, error fields, etc) and plasma fast shutdown. This chapter considers only plasma initiation and plasma basic control. This chapter describes the progress achieved in these areas in the tokamak experiments since the ITER Physics Basis (1999 Nucl. Fusion 39 2577) was written and the results of assessment of ITER to provide the plasma initiation and basic control. This assessment was done for the present ITER design (15 MA machine) at a more detailed level than it was done for the ITER design 1998 (21 MA machine) described in the ITER Physics Basis (1999 Nucl. Fusion 39 2577). The experiments on plasma initiation performed in DIII-D and JT-60U, as well as the theoretical studies performed for ITER, have demonstrated that, within specified assumptions on the plasma confinement and the impurity influx, ITER can produce plasma initiation in a low toroidal electric field (0.3 V m -1 ), if it is assisted by about 2 MW of ECRF heating. The plasma basic control includes control of the plasma current, position and shape-the plasma magnetic control, as well as control of other plasma global parameters or their profiles-the plasma performance control. The magnetic control is based on more reliable and simpler models of the control objects than those available at present for the plasma kinetic control. Moreover the real time diagnostics used for the magnetic control in many cases are more precise than those used for the kinetic control. Because of these reasons, the plasma magnetic control was developed for modern tokamaks and assessed for ITER better than the kinetic control. However, significant progress has been achieved in the plasma performance control during the last few years. Although the physics basis of plasma operation

Plasma physics for controlled fusion

International Nuclear Information System (INIS)

Miyamoto, K.

2010-01-01

The primary objective of this lecture note is to present the theories and experiments of plasma physics for recent activities of controlled fusion research for graduate and senior undergraduate students. Chapters 1-6 describe the basic knowledge of plasma and magnetohydrodynamics (MHD). MHD instabilities limit the beta ratio (ratio of plasma pressure to magnetic pressure) of confined plasma. Chapters 7-9 provide the kinetic theory of hot plasma and discuss the wave heating and non-inductive current drive. The dispersion relation derived by the kinetic theory are used to discuss plasma waves and perturbed modes. Landau damping is the essential mechanism of plasma heating and the stabilization of perturbation. Landau inverse damping brings the amplification of waves and the destabilization of perturbed modes. Chapter 10 explains the plasma transport due to turbulence, which is the most important and challenging subject for plasma confinement. Theories and simulations including subject of zonal flow are introduced. Chapters 11, 12 and 13 describe the recent activities of tokamak including ITER as well as spherical tokamak, reversed field pinch (RFP) and stellarator including quasi-symmetric configurations. Emphasis has been given to tokamak research since it made the most remarkable progress and the construction phase of 'International Tokamak Experimental Reactor' called ITER has already started. (author)

Plasma formation and first OH experiments in GLOBUS-M tokamak

International Nuclear Information System (INIS)

Gusev, V.K.; Aleksandrov, S.V.; Burtseva, T.A.

2001-01-01

The paper reports results of experimental campaigns on plasma ohmic heating, performed during 1999-2000 on the spherical tokamak Globus-M. Later experimental results with tokamak fed by thyristor rectifiers are presented in detail. The toroidal magnetic field and plasma pulse duration in these experiments were significantly increased. The method of stray magnetic field compensation is described. The technology of vacuum vessel conditioning, including boronization of the vessel performed at the end of the experiments, is briefly discussed. Also discussed is the influence of ECR preioniziation on the breakdown conditions. Experimental data on plasma column formation and current ramp-up in different regimes of operation with the magnetic flux of the central solenoid (CS) limited to ∼100 mVs are presented. Ramp-up of the plasma current of 0.25 MA for the time interval ∼0.03 s with about 0.02 s flat-top at the toroidal field (TF) strength of 0.35 T allows the conclusion that power supplies, control system and wall conditioning work well. The same conclusion can be drawn from observation of plasma density behavior the density is completely controlled with external gas puff and the influence of the wall is negligible after boronization. The magnetic flux consumption efficiency is discussed. The results of magnetic equilibrium simulations are presented and compared with experiment. (author)

Plasma experiments on staged theta pinch, implosion heating experiment and Scyllac feedback-sector experiment

International Nuclear Information System (INIS)

Bartsch, R.R.; Buchenauer, C.J.; Cantrell, E.L.

1977-01-01

Results of the Los Alamos theta-pinch program in three areas of investigation are summarized: 1) In the Staged Theta Pinch, results are reported on the effects of magnetic field amplitude and time history of plasma formation. 2) In the Implosion Heating Experiment, density, internal-magnetic field and neutron measurements yield a consistent picture of the implosion which agrees with kinetic computations and with a simple dynamic model of the ions and magnetic piston. 3) In the Scyllac Feedback-Sector Experiment, the l=1, 0 equilibrium plasma parameters have been adjusted to accommodate the feedback stabilization system. With a uniform toroidal discharge tube the m=1 instability is feedback-stabilized in the vertical direction, and confinement in the toroidal direction is extended by feedback control. Results with a helical discharge tube are also reported. (author)

Coherent control of plasma dynamics

Science.gov (United States)

He, Zhaohan

2014-10-01

The concept of coherent control - precise measurement or determination of a process through control of the phase of an applied oscillating field - has been applied to numerous systems with great success. Here, we demonstrate the use of coherent control on plasma dynamics in a laser wakefield electron acceleration experiment. A tightly focused femtosecond laser pulse (10 mJ, 35 fs) was used to generate electron beams by plasma wakefield acceleration in the density down ramp. The technique is based on optimization of the electron beam using a deformable mirror adaptive optical system with an iterative evolutionary genetic algorithm. The image of the electrons on a scintillator screen was processed and used in a fitness function as direct feedback for the optimization algorithm. This coherent manipulation of the laser wavefront leads to orders of magnitude improvement to the electron beam properties such as the peak charge and beam divergence. The laser beam optimized to generate the best electron beam was not the one with the ``best'' focal spot. When a particular wavefront of laser light interacts with plasma, it can affect the plasma wave structures and trapping conditions of the electrons in a complex way. For example, Raman forward scattering, envelope self-modulation, relativistic self-focusing, and relativistic self-phase modulation and many other nonlinear interactions modify both the pulse envelope and phase as the pulse propagates, in a way that cannot be easily predicted and that subsequently dictates the formation of plasma waves. The optimal wavefront could be successfully determined via the heuristic search under laser-plasma conditions that were not known a priori. Control and shaping of the electron energy distribution was found to be less effective, but was still possible. Particle-in-cell simulations were performed to show that the mode structure of the laser beam can affect the plasma wave structure and trapping conditions of electrons, which

Real-time control of current and pressure profiles in tokamak plasmas

International Nuclear Information System (INIS)

Laborde, L.

2005-12-01

Recent progress in the field of 'advanced tokamak scenarios' prefigure the operation regime of a future thermonuclear fusion power plant. Compared to the reference regime, these scenarios offer a longer plasma confinement time thanks to increased magnetohydrodynamic stability and to a better particle and energy confinement through a reduction of plasma turbulence. This should give access to comparable fusion performances at reduced plasma current and could lead to a steady state fusion reactor since the plasma current could be entirely generated non-inductively. Access to this kind of regime is provided by the existence of an internal transport barrier, linked to the current profile evolution in the plasma, which leads to steep temperature and pressure profiles. The comparison between heat transport simulations and experiments allowed the nature of the barriers to be better understood as a region of strongly reduced turbulence. Thus, the control of this barrier in a stationary manner would be a remarkable progress, in particular in view of the experimental reactor ITER. The Tore Supra and JET tokamaks, based in France and in the United Kingdom, constitute ideal instruments for such experiments: the first one allows stationary plasmas to be maintained during several minutes whereas the second one provides unique fusion performances. In Tore Supra, real-time control experiments have been accomplished where the current profile width and the pressure profile gradient were controlled in a stationary manner using heating and current drive systems as actuators. In the JET tokamak, the determination of an empirical static model of the plasma allowed the current and pressure profiles to be simultaneously controlled and so an internal transport barrier to be sustained. Finally, the identification of a dynamic model of the plasma led to the definition of a new controller capable, in principle, of a more efficient control. (author)

Bluff Body Flow Control Using Dielectric Barrier Discharge Plasma Actuators

Science.gov (United States)

Thomas, Flint; Kozlov, Alexey

2008-11-01

The results of an experimental investigation involving the use of dielectric barrier discharge plasma actuators to control bluff body flow is presented. The motivation for the work is plasma landing gear noise control for commercial transport aircraft. For these flow control experiments, the cylinder in cross-flow is chosen for study since it represents a generic flow geometry that is similar in all essential aspects to a landing gear strut. The current work is aimed both at extending the plasma flow control concept to Reynolds numbers typical of landing approach and take-off and on the development of optimum plasma actuation strategies. The cylinder wake flow with and without actuation are documented in detail using particle image velocimetry (PIV) and constant temperature hot-wire anemometry. The experiments are performed over a Reynolds number range extending to ReD=10^5. Using either steady or unsteady plasma actuation, it is demonstrated that even at the highest Reynolds number Karman shedding is totally eliminated and turbulence levels in the wake decrease by more than 50%. By minimizing the unsteady flow separation from the cylinder and associated large-scale wake vorticity, the radiated aerodynamic noise is also reduced.

Study of intelligent system for control of the tokamak-ETE plasma positioning

International Nuclear Information System (INIS)

Barbosa, Luis Filipe de Faria Pereira Wiltgen

2003-01-01

The development of an intelligent neural control system of the neural type, capable to perform real time control of the plasma displacement in the experiment tokamak spheric - ETE (spherical tokamak experiment ) is presented. The ETE machine is in operation since Nov 2000, in the LAP - Plasma Associated Laboratory of the Brazilian Institute on Spatial Research (INPE) in Sao Jose dos Campos, S P, Brazil. The experiment is dedicated to study the magnetic confinement of a fusion plasma in a configuration favorable for the construction of future reactors. Nuclear fusion constitutes a renewable energy source with low environmental impact, which uses atomic energy in pacific applications for the sustainable development of humanity. One of the important questions for the attainment of fusion relates to the stability of the plasma and control of its position during the reactor operation. Therefore, the development of systems to control the plasma in tokamaks constitutes a necessary technological advance for the feasibility of nuclear fusion. In particular, the research carried out in this thesis concerns the proposal of a system to control the vertical displacement of the plasma in the ETE tokamak, aiming to obtain steady pulses in this machine. A Magnetic Levitation system (Mag Lev) was developed as part of this work, allowing to study the nonlinear behavior of a device that, from the aspect of position control, is similar (analogous) to the plasma in the ETE tokamak, This magnetic levitation system was designed, mathematically modeled and built in order to test both classical and intelligent type controllers. The results of this comparison are very promising for the use of intelligent controllers in the ETE tokamak as well as other control applications. (author)

New DIII-D tokamak plasma control system

International Nuclear Information System (INIS)

Campbell, G.L.; Ferron, J.R.; McKee, E.; Nerem, A.; Smith, T.; Greenfield, C.M.; Pinsker, R.I.; Lazarus, E.A.

1992-09-01

A state-of-the-art plasma control system has been constructed for use on the DIII-D tokamak to provide high speed real time data acquisition and feedback control of DIII-D plasma parameters. This new system has increased the precision to which discharge shape and position parameters can be maintained and has provided the means to rapidly change from one plasma configuration to another. The capability to control the plasma total energy and the ICRF antenna loading resistance has been demonstrated. The speed and accuracy of this digital system will allow control of the current drive and heating systems in order to regulate the current and pressure profiles and diverter power deposition in the DIII-D machine. Use of this system will allow the machine and power supplies to be better protected from undesirable operating regimes. The advanced control system is also suitable for control algorithm development for future machines in these areas and others such as disruption avoidance. The DIII-D tokamak facility is operated for the US Department of Energy by General Atomics Company (GA) in San Diego, California. The DIII-D experimental program will increase emphasis on rf heating and current drive in the near future and is installing a cryopumped divertor ring during the fall of 1992. To improve the flexibility of this machine for these experiments, the new shape control system was implemented. The new advanced plasma control system has enhanced the capabilities of the DIII-D machine and provides a data acquisition and control platform that promises to be useful far beyond its original charter

Interrelated experiments in laboratory and space plasmas

International Nuclear Information System (INIS)

Koepke, M. E.

2005-01-01

Many advances in understanding space plasma phenomena have been linked to insight derived from theoretical modelling and/or laboratory experiments. Here are discussed advances for which laboratory experiments played an important role. How the interpretation of the space plasma data was influenced by one or more laboratory experiments is described. The space-motivation of laboratory investigations and the scaling of laboratory plasma parameters to space plasma conditions are discussed. Examples demonstrating how laboratory experiments develop physical insight, benchmark theoretical models, discover unexpected behaviour, establish observational signatures, and pioneer diagnostic methods for the space community are presented. The various device configurations found in space-related laboratory investigations are outlined. A primary objective of this review is to articulate the overlapping scientific issues that are addressable in space and lab experiments. A secondary objective is to convey the wide range of laboratory and space plasma experiments involved in this interdisciplinary alliance. The interrelation ship between plasma experiments in the laboratory and in space has a long history, with numerous demonstrations of the benefits afforded the space community by laboratory results. An experiment's suitability and limitations for investigating space processes can be quantitatively established using dimensionless parameters. Even with a partial match of these parameters, aspects of waves, instabilities, nonlinearities, particle transport, reconnection, and hydrodynamics are addressable in a way useful to observers and modelers of space phenomena. Because diagnostic access to space plasmas, laboratory-experimentalists awareness of space phenomena, and efforts by theorists and funding agencies to help scientists bridge the gap between the space and laboratory communities are increasing, the range of laboratory and space plasma experiments with overlapping scientific

Atto-second control of collective electron motion in plasmas

International Nuclear Information System (INIS)

Borot, Antonin; Malvache, Arnaud; Chen, Xiaowei; Jullien, Aurelie; Lopez-Martens, Rodrigo; Geindre, Jean-Paul; Audebert, Patrick; Mourou, Gerard; Quere, Fabien

2012-01-01

Today, light fields of controlled and measured waveform can be used to guide electron motion in atoms and molecules with atto-second precision. Here, we demonstrate atto-second control of collective electron motion in plasmas driven by extreme intensity (approximate to 10 18 W cm -2 ) light fields. Controlled few-cycle near-infrared waves are tightly focused at the interface between vacuum and a solid-density plasma, where they launch and guide sub-cycle motion of electrons from the plasma with characteristic energies in the multi-kilo-electron-volt range-two orders of magnitude more than has been achieved so far in atoms and molecules. The basic spectroscopy of the coherent extreme ultraviolet radiation emerging from the light-plasma interaction allows us to probe this collective motion of charge with sub-200 as resolution. This is an important step towards atto-second control of charge dynamics in laser-driven plasma experiments. (authors)

Plasma position control in SST1 tokamak

Indian Academy of Sciences (India)

also placed inside the vessel, however the controller would ignore fast but insignificant changes in radius arising ... poloidal cross-sectional view of the SST1 plasma along with the stabilizers are shown in figure 1 and ... [1] model which has shown excellent agreement with control experiments in TCV tokamak and also with ...

The plasma position control of ITER EDA plasma

International Nuclear Information System (INIS)

Senda, Ikuo; Nishio, Satoshi; Tsunematsu, Toshihide; Nishino, Toru; Fujieda, Hirobumi.

1994-09-01

The study on the plasma position control of ITER EDA performed by Japan Home Team during the sensitivity study in 1994 is summarized. The controllabilities of plasmas in the Outline Design and elongated version are compared. The model used to describe the motion of the plasma is a rigid model. The PD feedback control is applied with respect to the displacements of the plasma from the equilibrium. Three types of fluctuations, which initiate the motion of the plasma, are examined, namely a finite horizontal fluctuation field, a small horizontal fluctuation field such that the motion of the plasma is governed by the passive structures and an abrupt change of the poloidal beta β p and internal inductance l i . In the simulations of finite horizontal fluctuation fields, controls depend on the strength of the fluctuations, for instance, 3-5V is needed for 5-10G of fluctuation fields in the Outline Design. When the fluctuation field is small and the plasma displacement grows in a characteristic time of the passive structures, a few volt of the control voltage is enough to obtain good controllability. It is shown that the control when (β p , l i ) changes simultaneously is demanding and a large control voltage is required to maintain satisfactory control. Comparing the elongated version with the Outline Design, the control voltage which is larger than the Outline Design by a factor of 2-3 is required to obtain the same controllability in the elongated version. (author)

Plasma physics and controlled nuclear fusion research 1988. V.3

International Nuclear Information System (INIS)

1989-01-01

Volume 3 of the proceedings of the twelfth international conference on plasma physics and controlled nuclear fusion, held in Nice, France, 12-19 October, 1988, contains papers presented on inertial fusion. Direct and indirect laser implosion experiments, programs of laser construction, computer modelling of implosions and resulting plasmas, and light ion beam fusion experiments are discussed. Refs, figs and tabs

Real-time control of Tokamak plasmas: from control of physics to physics-based control

International Nuclear Information System (INIS)

Felici, F. A. A.

2011-11-01

Stable, high-performance operation of a tokamak requires several plasma control problems to be handled simultaneously. Moreover, the complex physics which governs the tokamak plasma evolution must be studied and understood to make correct choices in controller design. In this thesis, the two subjects have been merged, using control solutions as experimental tool for physics studies, and using physics knowledge for developing new advanced control solutions. The TCV tokamak at CRPP-EPFL is ideally placed to explore issues at the interface between plasma physics and plasma control, by combining a digital realtime control system with a flexible and powerful set of actuators, in particular the electron cyclotron heating and current drive system (ECRH/ECCD). This experimental platform has been used to develop and test new control strategies for three plasma physics instabilities: sawtooth, edge localized mode (ELM) and neoclassical tearing mode (NTM). The period of the sawtooth crash, a periodic MHD instability in the core of a tokamak plasma, can be varied by localized deposition of ECRH/ECCD near the q = 1 surface (q: safety factor). A sawtooth pacing controller was developed which is able to control the time of appearance of the next sawtooth crash. Each individual sawtooth period can be controlled in real-time. A similar scheme is applied to H-mode plasmas with type-I ELMs, where it is shown that pacing regularizes the ELM period. The regular, reproducible and therefore predictable sawtooth crashes have been used to study the relationship between sawteeth and NTMs. Postcrash MHD activity can provide the ‘seed’ island for an NTM, which then grows under its neoclassical bootstrap drive. The seeding of 3/2 NTMs by long sawtooth crashes can be avoided by preemptive, crash-synchronized EC power injection pulses at the q = 3/2 rational surface location. NTM stabilization experiments in which the ECRH deposition location is moved in real-time with steerable mirrors have

Architectural concept for the ITER Plasma Control System

Energy Technology Data Exchange (ETDEWEB)

Treutterer, W., E-mail: Wolfgang.Treutterer@ipp.mpg.de [Max-Planck Institute for Plasma Physics, EURATOM Association, Garching (Germany); Humphreys, D., E-mail: humphreys@fusion.gat.com [General Atomics, San Diego, CA (United States); Raupp, G., E-mail: Gerhard.Raupp@ipp.mpg.de [Max-Planck Institute for Plasma Physics, EURATOM Association, Garching (Germany); Schuster, E., E-mail: schuster@lehigh.edu [Lehigh University, Bethlehem, PA (United States); Snipes, J., E-mail: Joseph.Snipes@iter.org [ITER Organization, 13115 St. Paul-lez-Durance (France); De Tommasi, G., E-mail: detommas@unina.it [CREATE/Università di Napoli Federico II, Napoli (Italy); Walker, M., E-mail: walker@fusion.gat.com [General Atomics, San Diego, CA (United States); Winter, A., E-mail: Axel.Winter@iter.org [ITER Organization, 13115 St. Paul-lez-Durance (France)

2014-05-15

The plasma control system is a key instrument for successfully investigating the physics of burning plasma at ITER. It has the task to execute an experimental plan, known as pulse schedule, in the presence of complex relationships between plasma parameters like temperature, pressure, confinement and shape. The biggest challenge in the design of the control system is to find an adequate breakdown of this task in a hierarchy of feedback control functions. But it is also important to foresee structures that allow handling unplanned exceptional situations to protect the machine. Also the management of the limited number of actuator systems for multiple targets is an aspect with a strong impact on system architecture. Finally, the control system must be flexible and reconfigurable to cover the manifold facets of plasma behaviour and investigation goals. In order to prepare the development of a control system for ITER plasma operation, a conceptual design has been proposed by a group of worldwide experts and reviewed by an ITER panel in 2012. In this paper we describe the fundamental principles of the proposed control system architecture and how they were derived from a systematic collection and analysis of use cases and requirements. The experience and best practices from many fusion devices and research laboratories, augmented by the envisaged ITER specific tasks, build the foundation of this collection. In the next step control functions were distilled from this input. An analysis of the relationships between the functions allowed sequential and parallel structures, alternate branches and conflicting requirements to be identified. Finally, a concept of selectable control layers consisting of nested “compact controllers” was synthesised. Each control layer represents a cascaded scheme from high-level to elementary controllers and implements a control hierarchy. The compact controllers are used to resolve conflicts when several control functions would use the same

Architectural concept for the ITER Plasma Control System

International Nuclear Information System (INIS)

Treutterer, W.; Humphreys, D.; Raupp, G.; Schuster, E.; Snipes, J.; De Tommasi, G.; Walker, M.; Winter, A.

2014-01-01

The plasma control system is a key instrument for successfully investigating the physics of burning plasma at ITER. It has the task to execute an experimental plan, known as pulse schedule, in the presence of complex relationships between plasma parameters like temperature, pressure, confinement and shape. The biggest challenge in the design of the control system is to find an adequate breakdown of this task in a hierarchy of feedback control functions. But it is also important to foresee structures that allow handling unplanned exceptional situations to protect the machine. Also the management of the limited number of actuator systems for multiple targets is an aspect with a strong impact on system architecture. Finally, the control system must be flexible and reconfigurable to cover the manifold facets of plasma behaviour and investigation goals. In order to prepare the development of a control system for ITER plasma operation, a conceptual design has been proposed by a group of worldwide experts and reviewed by an ITER panel in 2012. In this paper we describe the fundamental principles of the proposed control system architecture and how they were derived from a systematic collection and analysis of use cases and requirements. The experience and best practices from many fusion devices and research laboratories, augmented by the envisaged ITER specific tasks, build the foundation of this collection. In the next step control functions were distilled from this input. An analysis of the relationships between the functions allowed sequential and parallel structures, alternate branches and conflicting requirements to be identified. Finally, a concept of selectable control layers consisting of nested “compact controllers” was synthesised. Each control layer represents a cascaded scheme from high-level to elementary controllers and implements a control hierarchy. The compact controllers are used to resolve conflicts when several control functions would use the same

Plasma opening switch experiments on the Particle Beam Accelerator II

International Nuclear Information System (INIS)

Sweeney, M.A.; McDaniel, D.H.; Mendel, C.W.; Rochau, G.E.; Moore, W.B.S.; Mowrer, G.R.; Simpson, W.W.; Zagar, D.M.; Grasser, T.; McDougal, C.D.

1989-01-01

Plasma opening switch (POS) experiments have been done since 1986 on the PBFA-II ion beam accelerator to develop a rugged POS that will open rapidly ( 80%) into a high impedance (> 10 ohm) load. In a recent series of experiments on PBFA II, the authors have developed and tested three different switch designs that use magnetic fields to control and confine the injected plasma. All three configurations couple current efficiently to a 5-ohm electron beam diode. In this experimental series, the PBFA-II Delta Series, more extensive diagnostics were used than in previous switch experiments on PBFA II or on the Blackjack 5 accelerator at Maxwell Laboratories. Data from the experiments with these three switch designs is presented

A Physics Exploratory Experiment on Plasma Liner Formation

Science.gov (United States)

Thio, Y. C. Francis; Knapp, Charles E.; Kirkpatrick, Ronald C.; Siemon, Richard E.; Turchi, Peter

2002-01-01

Momentum flux for imploding a target plasma in magnetized target fusion (MTF) may be delivered by an array of plasma guns launching plasma jets that would merge to form an imploding plasma shell (liner). In this paper, we examine what would be a worthwhile experiment to do in order to explore the dynamics of merging plasma jets to form a plasma liner as a first step in establishing an experimental database for plasma-jets driven magnetized target fusion (PJETS-MTF). Using past experience in fusion energy research as a model, we envisage a four-phase program to advance the art of PJETS-MTF to fusion breakeven Q is approximately 1). The experiment (PLX (Plasma Liner Physics Exploratory Experiment)) described in this paper serves as Phase I of this four-phase program. The logic underlying the selection of the experimental parameters is presented. The experiment consists of using twelve plasma guns arranged in a circle, launching plasma jets towards the center of a vacuum chamber. The velocity of the plasma jets chosen is 200 km/s, and each jet is to carry a mass of 0.2 mg - 0.4 mg. A candidate plasma accelerator for launching these jets consists of a coaxial plasma gun of the Marshall type.

Plasma density control in real-time on the COMPASS tokamak

Energy Technology Data Exchange (ETDEWEB)

Janky, F., E-mail: filip.janky.work@gmail.com [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Department of Surface and Plasma Science, Faculty of Mathematics and Physics, Charles University in Prague, V Holešovičkách 2, 180 00 Praha 8 (Czech Republic); Hron, M. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Havlicek, J. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Department of Surface and Plasma Science, Faculty of Mathematics and Physics, Charles University in Prague, V Holešovičkách 2, 180 00 Praha 8 (Czech Republic); Varavin, M.; Zacek, F.; Seidl, J.; Panek, R. [Institute of Plasma Physics AS CR, v.v.i., Association EURATOM/IPP.CR, Za Slovankou 3, 182 00 Praha 8 (Czech Republic)

2015-10-15

Highlights: • We fitted length of the chord of the interferometry crossing plasma in the different plasma scenarios. • We add correction to the actual length of the chord of the interferometry according to plasma shape and position in real-time code. • We used this correction to control plasma density in real-time. - Abstract: The electron density on COMPASS is measured using 2 mm microwave interferometer. Interferometer signal is used as an input for the feedback control loop, running under the MARTe real-time framework. Two different threads are used to calculate (fast 50 μs thread) and to control (slow 500 μs thread) the electron density. The interferometer measures a line averaged density along a measurement chord. This paper describes an approach to control the line-averaged electron density in a real-time loop, using a correction to the real plasma shape, the plasma position, and non-linear effects of the electron density measurement at high densities. Newly developed real-time electron density control give COMPASS the chance to control the electron density more accurately which is essential for parametric scans for diagnosticians, for physics experiments and also for achieving plasma scenarios with H-mode.

Experimental validation of a Lyapunov-based controller for the plasma safety factor and plasma pressure in the TCV tokamak

Science.gov (United States)

Mavkov, B.; Witrant, E.; Prieur, C.; Maljaars, E.; Felici, F.; Sauter, O.; the TCV-Team

2018-05-01

In this paper, model-based closed-loop algorithms are derived for distributed control of the inverse of the safety factor profile and the plasma pressure parameter β of the TCV tokamak. The simultaneous control of the two plasma quantities is performed by combining two different control methods. The control design of the plasma safety factor is based on an infinite-dimensional setting using Lyapunov analysis for partial differential equations, while the control of the plasma pressure parameter is designed using control techniques for single-input and single-output systems. The performance and robustness of the proposed controller is analyzed in simulations using the fast plasma transport simulator RAPTOR. The control is then implemented and tested in experiments in TCV L-mode discharges using the RAPTOR model predicted estimates for the q-profile. The distributed control in TCV is performed using one co-current and one counter-current electron cyclotron heating actuation.

Plasma control system upgrade and increased plasma stability in NSTX

Energy Technology Data Exchange (ETDEWEB)

Mastrovito, D., E-mail: dmastrovito@pppl.go [Princeton Plasma Physics Laboratory, P.O. Box 451 Princeton, NJ 08543 (United States); Gates, D.; Gerhard, S.; Lawson, J.; Ludescher-Furth, C.; Marsala, R. [Princeton Plasma Physics Laboratory, P.O. Box 451 Princeton, NJ 08543 (United States)

2010-07-15

Plasma control on the National Spherical Torus Experiment (NSTX) was previously accomplished using eight 333 MHz G4 processors built by Sky computers. Several planned improvements and additional control algorithms required significant upgrades to our real-time control computers and real-time data acquisition infrastructure. Several in-house modules have been designed and implemented including: the digital time stamp module (DITS) and for digital/analog front panel data port (FPDP) output, the FPDP output module digital/analog (FOMD/A). Standard Linux based Intel computers perform the real-time control tasks and InfiniBand as been employed for communication between a user-accessible 'host' server and the real-time computer. In addition to several independent real-time processes the General Atomics developed PCS (Bell (2006) ) system infrastructure continues to be used on NSTX. While maintaining previous functionality, improvements in the control system software include: an RWM feedback algorithm, beta feedback NBI control, more comprehensive error logging and trapping, more user-friendly interface, more complete archiving and restoring functionality, and better status reporting and diagnostic tools. Once completed, we succeeded in increasing overall plasma stability and decreasing control system latency by several times.

Chapter 8: Plasma operation and control

Science.gov (United States)

ITER Physics Expert Group on Disruptions, Control, Plasma, and MHD; ITER Physics Expert Group on Energetic Particles, Heating, Current and Drive; ITER Physics Expert Group on Diagnostics; ITER Physics Basis Editors

1999-12-01

well as in plasma periphery and divertor. The planned diagnostics (Chapter 7) serve as sensors for kinetic control, while gas and pellet fuelling, auxiliary power and angular momentum input, impurity injection, and non-inductive current drive constitute the control actuators. For example, in an ignited plasma, core density controls fusion power output. Kinetic control algorithms vary according to the plasma state, e.g. H- or L-mode. Generally, present facilities have demonstrated the kinetic control methods required for a reactor scale device. Plasma initiation - breakdown, burnthrough and initial current ramp - in reactor scale tokamaks will not involve physics differing from that found in present day devices. For ITER, the induced electric field in the chamber will be ~0.3V· m-1 - comparable to that required by breakdown theory but somewhat smaller than in present devices. Thus, a start-up 3MW electron cyclotron heating system will be employed to assure burnthrough. Simulations show that plasma current ramp up and termination in a reactor scale device can follow procedures developed to avoid disruption in present devices. In particular, simulations remain in the stable area of the li-q plane. For design purposes, the resistive V·s consumed during initiation is found, by experiments, to follow the Ejima expression, 0.45μ0 RIp. Advanced tokamak control has two distinct goals. First, control of density, auxiliary power, and inductive current ramping to attain reverse shear q profiles and internal transport barriers, which persist until dissipated by magnetic flux diffusion. Such internal transport barriers can lead to transient ignition. Second, combined use poloidal field shape control with non-inductive current drive and NBI angular momentum injection to create and control steady state, high bootstrap fraction, reverse shear discharges. Active n = 1 magnetic feedback and/or driven rotation will be required to suppress resistive wall modes for steady state plasmas

The renewed HT-7 plasma control system based on real-time Linux cluster

Energy Technology Data Exchange (ETDEWEB)

Yuan, Q.P., E-mail: qpyuan@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Xiao, B.J.; Zhang, R.R. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Walker, M.L.; Penaflor, B.G.; Piglowski, D.A.; Johnson, R.D. [General Atomics, DIII-D National Fusion Facility, San Diego, CA (United States)

2012-12-15

Highlights: Black-Right-Pointing-Pointer The hardware and software structure of the new HT-7 plasma control system (HT-7 PCS) is reported. Black-Right-Pointing-Pointer All original systems were integrated in the new HT-7 PCS. And the implementation details of the control algorithms are given in the paper. Black-Right-Pointing-Pointer Different from EAST PCS, the AC operation mode is realized in HT-7 PCS. Black-Right-Pointing-Pointer The experiment results are discussed. Good control performance has been obtained. - Abstract: In order to improve the synchronization, flexibility and expansibility of the plasma control on HT-7, a new plasma control system (HT-7 PCS) was constructed. The HT-7 PCS was based on a real-time Linux cluster with a well-defined, robust and flexible software infrastructure which was adapted from DIII-D PCS. In this paper, the hardware structure and system customization details for HT-7 PCS are reported. The plasma position and current control, plasma density control and off-normal event detection, which were realized in separated systems originally, have been integrated and implemented in such HT-7 PCS. All these control algorithms have been successfully validated in the last several HT-7 experiment campaigns. Good control performance has been achieved and the experiment results are discussed in the paper.

Nonneutral plasma diagnostic commissioning for the ALPHA Antihydrogen experiment

Science.gov (United States)

Konewko, S.; Friesen, T.; Tharp, T. D.; Alpha Collaboration

2017-10-01

The ALPHA experiment at CERN creates antihydrogen by mixing antiproton and positron plasmas. Diagnostic measurements of the precursor plasmas are performed using a diagnostic suite, colloquially known as the ``stick.'' This stick has a variety of sensors and is able to move to various heights to align the desired diagnostic with the beamline. A cylindrical electrode, a faraday cup, an electron gun, and a microchannel-plate detector (MCP) are regularly used to control and diagnose plasmas in ALPHA. We have designed, built, and tested a new, upgraded stick which includes measurement capabilities in both beamline directions.

ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM

International Nuclear Information System (INIS)

HUMPHREYS, DA; FERRON, JR; GAROFALO, AM; HYATT, AW; JERNIGAN, TC; JOHNSON, RD; LAHAYE, RJ; LEUER, JA; OKABAYASHI, M; PENAFLOR, BG; SCOVILLE, JT; STRAIT, EJ; WALKER, ML; WHYTE, DG

2002-01-01

A271 ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM. The principal focus of experimental operations in the DIII-D tokamak is the advanced tokamak (AT) regime to achieve, which requires highly integrated and flexible plasma control. In a high performance advanced tokamak, accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating must be well coordinated with MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Sophisticated monitors of the operational regime must provide detection of off-normal conditions and trigger appropriate safety responses with acceptable levels of reliability. Many of these capabilities are presently implemented in the DIII-D plasma control system (PCS), and are now in frequent or routine operational use. The present work describes recent development, implementation, and operational experience with AT regime control elements for equilibrium control, MHD suppression, and off-normal event detection and response

Real time control of plasmas and ECRH systems on TCV

International Nuclear Information System (INIS)

Paley, J.I.; Berrino, J.; Coda, S.; Duval, B.P.; Felici, F.; Goodman, T.P.; Martin, Y.; Moret, J.M.; Piras, F.; Cruz, N.; Rodriques, A.P.; Santos, B.; Varandas, C.A.F.

2009-01-01

Developments in the real time control hardware on Tokamak a Configuration Variable (TCV) coupled with the flexibility of plasma shaping and electron cyclotron (EC) heating and current drive actuators are opening many opportunities to perform real time experiments and develop algorithms and methods for fusion applications. The ability to control magnetohydrodynamic instabilities is particularly important for achieving high performance fusion plasmas and EC is envisaged as a key actuator in maintaining high performance. We have successfully demonstrated control of the sawtooth instability using the EC launcher injection angle to modify the current profile around the q =1 surface. This paper presents an overview of recent real time control experiments on TCV, developments in the hardware and algorithms together with plans for the future.

Plasma control system for 'Day-One' operation of KSTAR tokamak

International Nuclear Information System (INIS)

Hahn, Sang-hee; Walker, M.L.; Kim, Kukhee; Ahn, H.S.; Penaflor, B.G.; Piglowski, D.A.; Johnson, R.D.; Choi, Jaehoon; Lee, Dong-keun; Kim, Jayhyun; Yoon, S.W.; Seo, Seong-Heon; Kim, H.T.; Kim, K.P.; Lee, T.G.; Park, M.K.; Bak, J.G.; Lee, S.G.; Nam, Y.U.; Eidietis, N.W.

2009-01-01

A complete plasma control system (PCS) has been developed for KSTAR's first plasma campaign as a collaborative project with the DIII-D team. The KSTAR real time plasma control system is based on a conceptual design by Jhang and Choi [Hogun Jhang, I.S. Choi, Fusion Engineering and Design 73 (2005) 35-49] and consists of a fast real-time computer/communication cluster and software derived from the GA-PCS [Penaflor, B.G., et.al., Fusion Engineering and Design, 83 (2) (2008) 176]. The system has been used for simulation testing, poloidal field (PF) coil power supply commissioning and first plasma control. The seven sets of up-down symmetric, superconducting PF coil/power supply systems have been successfully tested. Reflective memory (RFM) is utilized as the primary actuator/PCS real-time communication layer and PCS synchronization with KSTAR timing system and slower control devices is achieved through an EPICS implementation. Consistent feedback loop times of 100 microseconds has been achieved during PF coil power supply testing and first plasma commissioning. Here we present the 'Day-One' plasma control system in its final form for the first plasma experimental campaign of KSTAR and describe how the system has been utilized during magnet commissioning and plasma startup experiments.

ICRF heating on the burning plasma experiment (BPX)

International Nuclear Information System (INIS)

Batchelor, D.B.; Carter, M.D.; Goulding, R.H.; Hoffman, D.J.; Jaeger, E.F.; Ryan, P.M.; Swain, D.W.; Tolliver, J.S.; Yugo, J.J.; Goldston, R.J.; Hosea, J.C.; Kaye, S.M.; Phillips, C.K.; Wilson, J.R.; Mau, T.K.

1991-01-01

RF power in the ion cyclotron range of frequencies (ICRF) has been chosen as the primary heating technique for BPX. This decision is based on the wide success of ICRF heating in existing experiments (JET, TFTR, JT-60), the capability of ion cyclotron waves to penetrate the high-density plasmas of BPX, the ability to concentrate ICRF power deposition near the plasma center, and the ready availability of high-power sources at the appropriate frequency. The primary task of the ICRF system is to heat the plasma to ignition. However, other important roles are envisaged; these include the stabilization of sawteeth, preheating of the plasma during current ramp-up, and possible control of the plasma current profile by means of fast-wave current drive. We give a brief overview of the RF system, describe the operating scenarios planned for BPX, and discuss some of the antenna design issues for BPX. 4 refs., 3 figs

The control of TCV plasmas

International Nuclear Information System (INIS)

Lister, J.B.; Hofmann, F.; Moret, J.M.

1996-07-01

The general control of tokamak plasmas has evolved considerably over the last few years with an increase in the plasma pulse length, an increase in the control of additional heating and fuelling and an increase in the degree to which the shape of the plasma can be varied. The TCV tokamak is specifically designed to explore the operational benefits of plasma shaping over a wide variety of plasma shapes. Consequently, considerable attention has been given to the control of the poloidal field coil currents which impose the desired shape. This paper deals with all aspects of the control of TCV plasmas, from the diagnostic measurements to the power supplies, via control algorithms and overall supervision. (author) 44 figs., tabs., 25 refs

The control of TCV plasmas

Energy Technology Data Exchange (ETDEWEB)

Lister, J B; Hofmann, F; Moret, J M [Ecole Polytechnique Federale, Lausanne (Switzerland). Centre de Recherche en Physique des Plasma (CRPP); and others

1996-07-01

The general control of tokamak plasmas has evolved considerably over the last few years with an increase in the plasma pulse length, an increase in the control of additional heating and fuelling and an increase in the degree to which the shape of the plasma can be varied. The TCV tokamak is specifically designed to explore the operational benefits of plasma shaping over a wide variety of plasma shapes. Consequently, considerable attention has been given to the control of the poloidal field coil currents which impose the desired shape. This paper deals with all aspects of the control of TCV plasmas, from the diagnostic measurements to the power supplies, via control algorithms and overall supervision. (author) 44 figs., tabs., 25 refs.

Magnetic sensorless control of plasma position and shape in a tokamak

International Nuclear Information System (INIS)

Nakamura, K.; Luo, J.R.; Wang, H.Z.

2005-01-01

Magnetic sensorless sensing and control experiments of the plasma horizontal position have been carried out in the superconducting tokamak HT-7. The sensing is made focusing on the ripple frequency component of the power supply with thyristor and directly from them without time integration. There is no drift problem of integrator of magnetic sensors. Two kinds of control experiments were carried out, to keep the position constant and swing the position in a triangular waveform. And magnetic sensorless sensing of plasma shape is discussed. (author)

Flow structure formation in an ion-unmagnetized plasma: The HYPER-II experiments

Science.gov (United States)

Terasaka, K.; Tanaka, M. Y.; Yoshimura, S.; Aramaki, M.; Sakamoto, Y.; Kawazu, F.; Furuta, K.; Takatsuka, N.; Masuda, M.; Nakano, R.

2015-01-01

The HYPER-II device has been constructed in Kyushu University to investigate the flow structure formation in an ion-unmagnetized plasma, which is an intermediate state of plasma and consists of unmagnetized ions and magnetized electrons. High density plasmas are produced by electron cyclotron resonance heating, and the flow field structure in an inhomogeneous magnetic field is investigated with a directional Langmuir probe method and a laser-induced fluorescence method. The experimental setup has been completed and the diagnostic systems have been installed to start the experiments. A set of coaxial electrodes will be introduced to control the azimuthal plasma rotation, and the effect of plasma rotation to generation of rectilinear flow structure will be studied. The HYPER-II experiments will clarify the overall flow structure in the inhomogeneous magnetic field and contribute to understanding characteristic feature of the intermediate state of plasma.

Plasma Equilibrium Control in Nuclear Fusion Devices 2. Plasma Control in Magnetic Confinement Devices 2.1 Plasma Control in Tokamaks

Science.gov (United States)

Fukuda, Takeshi

The plasma control technique for use in large tokamak devices has made great developmental strides in the last decade, concomitantly with progress in the understanding of tokamak physics and in part facilitated by the substantial advancement in the computing environment. Equilibrium control procedures have thereby been established, and it has been pervasively recognized in recent years that the real-time feedback control of physical quantities is indispensable for the improvement and sustainment of plasma performance in a quasi-steady-state. Further development is presently undertaken to realize the “advanced plasma control” concept, where integrated fusion performance is achieved by the simultaneous feedback control of multiple physical quantities, combined with equilibrium control.

Study of Globus-M Tokamak Poloidal System and Plasma Position Control

Science.gov (United States)

Dokuka, V. N.; Korenev, P. S.; Mitrishkin, Yu. V.; Pavlova, E. A.; Patrov, M. I.; Khayrutdinov, R. R.

2017-12-01

In order to provide efficient performance of tokamaks with vertically elongated plasma position, control systems for limited and diverted plasma configuration are required. The accuracy, stability, speed of response, and reliability of plasma position control as well as plasma shape and current control depend on the performance of the control system. Therefore, the problem of the development of such systems is an important and actual task in modern tokamaks. In this study, the measured signals from the magnetic loops and Rogowski coils are used to reconstruct the plasma equilibrium, for which linear models in small deviations are constructed. We apply methods of the H∞-optimization theory to the synthesize control system for vertical and horizontal position of plasma capable to working with structural uncertainty of the models of the plant. These systems are applied to the plasma-physical DINA code which is configured for the tokamak Globus-M plasma. The testing of the developed systems applied to the DINA code with Heaviside step functions have revealed the complex dynamics of plasma magnetic configurations. Being close to the bifurcation point in the parameter space of unstable plasma has made it possible to detect an abrupt change in the X-point position from the top to the bottom and vice versa. Development of the methods for reconstruction of plasma magnetic configurations and experience in designing plasma control systems with feedback for tokamaks provided an opportunity to synthesize new digital controllers for plasma vertical and horizontal position stabilization. It also allowed us to test the synthesized digital controllers in the closed loop of the control system with the DINA code as a nonlinear model of plasma.

Structure of the automatic system for plasma equilibrium position control

International Nuclear Information System (INIS)

Gubarev, V.F.; Krivonos, Yu.G.; Samojlenko, Yu.I.; Snegur, A.A.

1978-01-01

Considered are the principles of construction of the automatic system for plasma filament equilibrium position control inside the discharge chamber for the installation of a tokamak type. The combined current control system in control winding is suggested. The most powerful subsystem creates current in the control winding according to the program calculated beforehand. This system provides plasma rough equilibrium along the ''big radius''. The subsystem performing the current change in small limits according to the principle of feed-back coupling is provided simultaneously. The stabilization of plasma position is achieved in the discharge chamber. The advantage of construction of such system is in decreasing of the automatic requlator power without lowering the requirements to the accuracy of equilibrium preservation. The subsystem of automatic control of plasma position over the vertical is put into the system. Such an approach to the construction of the automatic control system proves to be correct; it is based on the experience of application of similar devices for some existing thermonuclear plants

Plasma control device

International Nuclear Information System (INIS)

Takase, Haruhiko.

1987-01-01

Purpose: To obtain the optimum controllability for the plasmas and the thermonuclear device by selectively executing control operation for proportion, integration and differentiation (PID) by first and second controllers respectively based on selection instruction signals. Constitution: Deviation between a vertical direction equilibrium position: Zp as the plasma status amount measured in a measuring section and an aimed value Zref thereof is inputted to a first PID selection controller. The first controller selectively executes one of the PID control operations in accordance with the first selection signal instruction instructed by a PID control operation instruction circuit. Further, Zp is also inputted to a second PID selection controller, which selectively executes one of the PID control operations in accordance with the second selection instruction signal in the same manner as in the first controller. The deviation amount u between operations signals u1 and u2 from the first and second PID selection controllers is inputted to a power source to thereby supply a predetermined current value to control coils that generate equilibrium magnetic fields for making the vertical direction equilibrium position of plasmas constant. (Kamimura, M.)

Laboratory plasma interactions experiments: Results and implications to future space systems

Science.gov (United States)

Leung, Philip

1986-01-01

The experimental results discussed show the significance of the effects caused by spacecraft plasma interactions, in particular the generation of Electromagnetic Interference. As the experimental results show, the magnitude of the adverse effects induced by Plasma Interactions (PI) will be more significant for spacecraft of the next century. Therefore, research is needed to control possible adverse effects. Several techniques to control the selected PI effects are discussed. Tests, in the form of flight experiments, are needed to validate these proposed ideas.

Plasma control device

International Nuclear Information System (INIS)

Matsutomi, Akiyoshi.

1995-01-01

Plasma position and shape estimation values are outputted based on measured values of coil current. When the measured values of the position and the shape are judged to be abnormal, position and shape estimation values estimated by a plasma position and shape estimation means are outputted as position and shape feed back values to a plasma position and shape control means instead of the position and shape measured values. Since only a portion of the abnormal position and shape measured values may also be replaced with the position and shape estimation values, errors in the plasma position and shape feed back values can be reduced as a whole. In addition, even if the position and shape measured values are abnormal or if they can not be measured, plasma control can be continued by alternative position and shape estimation values, thereby enabling to avoid interruption of plasma control. Since the position and shape estimation values are obtained not only with the measured values of coil current but also with the position and shape estimation values, the accuracy is improved. Further, noises superposed on the position and shape measured values are filtered, and the device is stabilized compared with a prior art device. (N.H.)

Plasma-Jet-Driven Magneto-Inertial Fusion (PJMIF): Physics and Design for a Plasma Liner Formation Experiment

Science.gov (United States)

Hsu, Scott; Cassibry, Jason; Witherspoon, F. Douglas

2014-10-01

Spherically imploding plasma liners are a potential standoff compression driver for magneto-inertial fusion, which is a hybrid of and operates in an intermediate density between those of magnetic and inertial fusion. We propose to use an array of merging supersonic plasma jets to form a spherically imploding plasma liner. The jets are to be formed by pulsed coaxial guns with contoured electrodes that are placed sufficiently far from the location of target compression such that no hardware is repetitively destroyed. As such, the repetition rate can be higher (e.g., 1 Hz) and ultimately the power-plant economics can be more attractive than most other MIF approaches. During the R&D phase, a high experimental shot rate at reasonably low cost (e.g., gun plasma-liner-formation experiment, which will provide experimental data on: (i) scaling of peak liner ram pressure versus initial jet parameters, (ii) liner non-uniformity characterization and control, and (iii) control of liner profiles for eventual gain optimization.

Labotratory Simulation Experiments of Cometary Plasma

OpenAIRE

MINAMI, S.; Baum, P. J.; Kamin, G.; White, R. S.; 南, 繁行

1986-01-01

Laboratory simulation experiment to study the interaction between a cometary plasma and the solar wind has been performed using the UCR-T 1 space simulation facility at the Institute of Geophysics and Planetary Physics, the University of California, Riverside. Light emitting plasma composed of Sr, Ba and/or C simulating cometary coma plasma is produced by a plasma emitter which interacts with intense plasma flow produced by a co-axial plasma gun simulating the solar wind. The purpose of this ...

Positional stability experiment and analysis of elongated plasmas in Doublet III

International Nuclear Information System (INIS)

Yokomizo, Hideaki

1984-04-01

Control systems of the plasma position and shape on Doublet III are explained and experimental results of vertical stability of elongated plasmas are reviewed. Observed results of the vertical instability are qualitatively compared with the predictions from the simplified model and quantitatively compared with the numerical calculations based on a more realistic model. Experiments are in reasonable agreement with the theoretical analyses. (author)

Diagnostics for real-time plasma control in PBX-M

Science.gov (United States)

Kaita, R.; Batha, S.; Bell, R. E.; Bernabei, S.; Hatcher, R.; Kozub, T.; Kugel, H.; Levinton, F.; Okabayashi, M.; Sesnic, S.; von Goeler, S.; Zolfaghari, A.; PBX-M Group

1995-01-01

An important issue for future tokamaks is real-time plasma control for the avoidance of magnetohydrodynamic instabilities and other applications that require detailed plasma profile and fluctuation data. Although measurements from diagnostics providing this information require significantly more processing than magnetic flux data, recent advancements could make them practical for adjusting operational settings for plasma heating and current drive systems as well as field coil currents. On the Princeton Beta Experiment-Modification (PBX-M), the lower hybrid current drive phasing can be varied during a plasma shot using digitally programmable ferrite phase shifters, and neural beam functions can be fully computer controlled. PBX-M diagnostics that may be used for control purposes include motional Stark-effect polarimetry for magnetic field pitch angle profiles, soft x-ray arrays for plasma position control and the separation of βp from li, hard x-ray detectors for energetic electron distributions, a multichannel electron cyclotron emission radiometer for ballooning mode identification, and passive plate eddy current monitors for kink stabilization. We will describe the present status of these systems on PBX-M, and discuss their suitability for feedback applications.

Diagnostics for real-time plasma control in PBX-M

International Nuclear Information System (INIS)

Kaita, R.; Batha, S.; Bell, R.E.; Bernabei, S.; Hatcher, R.; Kozub, T.; Kugel, H.; Levinton, F.; Okabayashi, M.; Sesnic, S.; Goeler, S. von; Zolfaghari, A.

1995-01-01

An important issue for future tokamaks is real-time plasma control for the avoidance of magnetohydrodynamic instabilities and other applications that require detailed plasma profile and fluctuation data. Although measurements from diagnostics providing this information require significantly more processing than magnetic flux data, recent advancements could make them practical for adjusting operational settings for plasma heating and current drive systems as well as field coil currents. On the Princeton Beta Experiment-Modification (PBX-M), the lower hybrid current drive phasing can be varied during a plasma shot using digitally programmable ferrite phase shifters, and neural beam functions can be fully computer controlled. PBX-M diagnostics that may be used for control purposes include motional Stark-effect polarimetry for magnetic field pitch angle profiles, soft x-ray arrays for plasma position control and the separation of β p from l i , hard x-ray detectors for energetic electron distributions, a multichannel electron cyclotron emission radiometer for ballooning mode identification, and passive plate eddy current monitors for kink stabilization. We will describe the present status of these systems on PBX-M, and discuss their suitability for feedback applications

Plasma performance, boundary studies and first experiments with ICRH in TEXTOR

International Nuclear Information System (INIS)

Waidmann, G.; Bay, H.L.; Bertschinger, G.

1985-01-01

The TEXTOR plasma serves as a test bed for plasma/wall interaction studies and ICRH experiments. Reproducible and long-lasting discharges with soft termination were generated in the internal disruptive mode. The operational regime for Ohmic heating is shown in a 1/q versus n-barsub(e)R/Bsub(T) diagram. A comparison of electrical conductivity derived from current density measurements with calculated values favours neoclassical theory. A pump limiter installed on TEXTOR demonstrated a particle removal rate of 6x10 20 particles per second out of the boundary layer. It could decrease the central electron density by 50%. The pump limiter was used to control fuelling and recycling characteristics of stable discharges. First experiments with additional ICRH showed a strong influence on the plasma boundary and scrape-off layer. The interaction of the radiofrequency with the boundary layer at present limits the power input to the plasma. Plasma boundary parameters have been measured by optical methods combined with neutral particle beams. (author)

Spheromak type plasma experiment apparatus

International Nuclear Information System (INIS)

Odagiri, Kiyoyuki; Miyauchi, Yasuyuki; Oomura, Hiroshi

1985-01-01

The fusion power reactor which is expected to be the most promising energy has been developed for several plasma confinement systems. Under these circumstances, Spheromak configuration has recently attracted attention because of its simple structure and efficient plasma confinement. This apparatus was ordered by the Engineering Department of University of Tokyo for basic studies of the Spheromak plasma confinement technologies. This forms Spheromak plasma according to the induction discharge system which injects this plasma with magnetic energy generated by a toroidal current in the plasma and discharges the current through the electrical feed through. Toroidal current is induced by the poloidal coil in the vessel. We worked together with the researchers of University of Tokyo to conduct experiments and confirmed the formation and confinement of Spheromak plasma in the initial test. (author)

Plasma crowbars in cylindrical flux compression experiments

International Nuclear Information System (INIS)

Suter, L.J.

1979-01-01

We have done a series of one- and two-dimensional calculations of hard-core Z-pinch flux compression experiments in order to study the effect of a plasma on these systems. These calculations show that including a plasma can reduce the amount of flux lost during the compression. Flux losses to the outer wall of such experiments can be greatly reduced by a plasma conducting sheath which forms along the wall. This conducting sheath consists of a cold, dense high β, unmagnetized plasma which has enough pressure to balance a large field gradient. Flux which is lost into the center conductor is not effectively stopped by this plasma sheath until late in the implosion, at which time a layer similar to the one formed at the outer wall is created. Two-dimensionl simulations show that flux losses due to arching along the sliding contact of the experiment can be effectively stopped by the formation of a plasma conducting sheath

Post-disruptive plasma loss in the Princeton Beta Experiment (PBX)

International Nuclear Information System (INIS)

Jardin, S.C.; DeLucia, J.; Okabayashi, M.; Pomphrey, N.; Reusch, M.; Kaye, S.; Takahashi, H.

1986-07-01

The free-boundary, axisymmetric tokamak simulation code TSC is used to model the transport time scale evolution and positional stability of PBX. A disruptive thermal quench will cause the plasma column to move inward in major radius. It is shown that the plasma can then lose axisymmetric stability, causing it to displace exponentially off the midplane, terminating the discharge. We verify the accuracy of the code by modeling several controlled experiments shots in PBX

Simultaneous Feedback Control of Plasma Rotation and Stored Energy on the DIII-D Tokamak

International Nuclear Information System (INIS)

Scoville, J.T.; Ferron, J.R.; Humphreys, D.A.; Walker, M.L.

2006-01-01

One of the major modifications made to the DIII-D tokamak during the 2005 Long Torus Opening was the rotation of one of the four two-source neutral beam injection systems. Prior to this modification, all beams injected power with a component in the same direction as the usual plasma current ('' co-injection ''). Starting in early 2006, two of the seven beams inject with a component in the opposite direction ('' counter-injection ''). This new capability allows, for the first time, a partial decoupling of the injected energy and momentum during neutral beam heating experiments. An immediate advantage of mixed co- and counter-injection beams is the capability to control the plasma rotation velocity. High beta plasmas can now be studied over a wide range of the plasma rotation velocity. The stabilizing effect of rotation on the resistive wall mode (RWM), for example, can be directly compared to the stabilization achieved by external feedback coils. This is an advantage over previous techniques to control plasma rotation, such as magnetic braking, which have had only limited success. We describe development and implementation of a model-based control algorithm for simultaneous regulation of plasma rotation and beta. The model includes the two relevant plasma states (plasma rotation and stored energy), and describes the dynamic effects of the relevant actuators on those states. The actuators include the applied beam torque and beam power, which depend on the amount of co and counter-injected beams. Implementation of the model-based control within the plasma control system (PCS) [B.G. Penaflor, et al, '' Current Status of DIII-D Plasma Control System Computer Upgrades,'' Fusion Eng. and Design 71 (2004) 47] requires real-time measurements of the plasma rotation, obtained from the charge exchange recombination (CER) diagnostic, and stored energy calculated by the real-time EFIT equilibrium reconstruction. Details of this model and its development, and a comparison with

Plasma Physics and Controlled Nuclear Fusion Research. Vol. II. Proceedings of a Conference on Plasma Physics and Controlled Physics Research

International Nuclear Information System (INIS)

1966-01-01

Research on controlled nuclear fusion was first disclosed at the Second United Nations Conference on the Peaceful Uses of Atomic Energy, held at Geneva in 1958. From the information given, it was evident that a better understanding of the behaviour of hot dense plasmas was needed before the goal of economic energy release from nuclear fusion could be reached. The fact that research since then has been most complex and costly has enhanced the desirability of international co-operation and exchange of information and experience. Having organized its First Conference on Plasma Physics and Controlled Nuclear Fusion Research at Salzburg in 1961, the International Atomic Energy Agency again provided the means for such cooperation in organizing its Second Conference on this subject on 6-10 September, 1965, at Culham, Abingdon, Berks, England. The meeting was arranged with the generous help of the United Kingdom Atomic Energy Authority at their Culham Laboratory, where the facilities and assistance of the staff were greatly appreciated. At the meeting, which was attended by 268 participants from 26 member states and three international organizations, significant results from many experiments, including those from the new and larger machines, became available. It has now become feasible to intercorrelate data obtained from a number of similar machines; this has led to a more complete understanding of plasma behaviour. No breakthrough was reported nor had been expected towards the economical release of the energy from fusion, but there was increased understanding of the problems of production, control and containment of high-density and high-temperature plasmas

Interprocess communication within the DIII-D plasma control system

International Nuclear Information System (INIS)

Piglowski, D.A.; Penaflor, B.G.; Ferron, J.R.

1999-06-01

The DIII-D tokamak fusion research experiment's real-time digital plasma control system (PCS) is a complex and ever evolving system. During a plasma experiment, it is tasked with some of the most crucial functions at DIII-D. Key responsibilities of the PCS involve sub-system control, data acquisition/storage, and user interface. To accomplish these functions, the PCS is broken down into individual components (both software and hardware), each capable of handling a specific duty set. Constant interaction between these components is necessary prior, during and after a standard plasma cycle. Complicating the matter even more is that some components, mostly those which deal with user interaction, may exist remotely, that is to say they are not part of the immediate hardware which makes up the bulk of the PCS. The four main objectives of this paper are to (1) present a brief outline of the PCS hardware/software and how they relate to each other; (2) present a brief overview of a standard DIII-D plasma cycle (a shot); (3) using three sets of PCS sub-systems, describe in more detail the communication processes; and (4) evaluate the benefits and drawbacks of said systems

Physics of the conceptual design of the ITER plasma control system

Energy Technology Data Exchange (ETDEWEB)

Snipes, J.A., E-mail: Joseph.Snipes@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Bremond, S. [CEA-IRFM, 13108 St Paul-lez-Durance (France); Campbell, D.J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Casper, T. [1166 Bordeaux St, Pleasanton, CA 94566 (United States); Douai, D. [CEA-IRFM, 13108 St Paul-lez-Durance (France); Gribov, Y. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Humphreys, D. [General Atomics, San Diego, CA 92186 (United States); Lister, J. [Association EURATOM-Confédération Suisse, Ecole Polytechnique Fédérale de Lausanne (EPFL), CRPP, Lausanne CH-1015 (Switzerland); Loarte, A.; Pitts, R. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Sugihara, M., E-mail: Sugihara_ma@yahoo.co.jp [Japan (Japan); Winter, A.; Zabeo, L. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France)

2014-05-15

Highlights: • ITER plasma control system conceptual design has been finalized. • ITER's plasma control system will evolve with the ITER research plan. • A sophisticated actuator sharing scheme is being developed to apply multiple coupled control actions simultaneously with a limited set of actuators. - Abstract: The ITER plasma control system (PCS) will play a central role in enabling the experimental program to attempt to sustain DT plasmas with Q = 10 for several hundred seconds and also support research toward the development of steady-state operation in ITER. The PCS is now in the final phase of its conceptual design. The PCS relies on about 45 diagnostic systems to assess real-time plasma conditions and about 20 actuator systems for overall control of ITER plasmas. It will integrate algorithms required for active control of a wide range of plasma parameters with sophisticated event forecasting and handling functions, which will enable appropriate transitions to be implemented, in real-time, in response to plasma evolution or actuator constraints. In specifying the PCS conceptual design, it is essential to define requirements related to all phases of plasma operation, ranging from early (non-active) H/He plasmas through high fusion gain inductive plasmas to fully non-inductive steady-state operation, to ensure that the PCS control functionality and architecture will be capable of satisfying the demands of the ITER research plan. The scope of the control functionality required of the PCS includes plasma equilibrium and density control commonly utilized in existing experiments, control of the plasma heat exhaust, control of a range of MHD instabilities (including mitigation of disruptions), and aspects such as control of the non-inductive current and the current profile required to maintain stable plasmas in steady-state scenarios. Control areas are often strongly coupled and the integrated control of the plasma to reach and sustain high plasma

Physics of the conceptual design of the ITER plasma control system

International Nuclear Information System (INIS)

Snipes, J.A.; Bremond, S.; Campbell, D.J.; Casper, T.; Douai, D.; Gribov, Y.; Humphreys, D.; Lister, J.; Loarte, A.; Pitts, R.; Sugihara, M.; Winter, A.; Zabeo, L.

2014-01-01

Highlights: • ITER plasma control system conceptual design has been finalized. • ITER's plasma control system will evolve with the ITER research plan. • A sophisticated actuator sharing scheme is being developed to apply multiple coupled control actions simultaneously with a limited set of actuators. - Abstract: The ITER plasma control system (PCS) will play a central role in enabling the experimental program to attempt to sustain DT plasmas with Q = 10 for several hundred seconds and also support research toward the development of steady-state operation in ITER. The PCS is now in the final phase of its conceptual design. The PCS relies on about 45 diagnostic systems to assess real-time plasma conditions and about 20 actuator systems for overall control of ITER plasmas. It will integrate algorithms required for active control of a wide range of plasma parameters with sophisticated event forecasting and handling functions, which will enable appropriate transitions to be implemented, in real-time, in response to plasma evolution or actuator constraints. In specifying the PCS conceptual design, it is essential to define requirements related to all phases of plasma operation, ranging from early (non-active) H/He plasmas through high fusion gain inductive plasmas to fully non-inductive steady-state operation, to ensure that the PCS control functionality and architecture will be capable of satisfying the demands of the ITER research plan. The scope of the control functionality required of the PCS includes plasma equilibrium and density control commonly utilized in existing experiments, control of the plasma heat exhaust, control of a range of MHD instabilities (including mitigation of disruptions), and aspects such as control of the non-inductive current and the current profile required to maintain stable plasmas in steady-state scenarios. Control areas are often strongly coupled and the integrated control of the plasma to reach and sustain high plasma

User Control Interface for W7-X Plasma Operation

International Nuclear Information System (INIS)

Spring, A.; Laqua, H.; Schacht, J.

2006-01-01

The WENDELSTEIN 7-X fusion experiment will be a highly complex device operated by a likewise complex control system. The fundamental configuration of the W7-X control system follows two major design principles: It reflects the strict hierarchy of the machine set-up with a set of subordinated components, which in turn can be run autonomously during commissioning and testing. Secondly, it links the basic machine operation (mainly given by the infrastructure status and the components readiness) and the physics program execution (i.e. plasma operation) on each hierarchy level and on different time scales. The complexity of the control system implies great demands on appropriate user interfaces: specialized tools for specific control tasks allowing a dedicated view on the subject to be controlled, hiding complexity wherever possible and reasonable, providing similar operation methods on each hierarchy level and both manual interaction possibilities and a high degree of intelligent automation. The contribution will describe the operation interface for experiment control including the necessary links to the machine operation. The users of ' Xcontrol ' will be both the W7-X session leaders during plasma discharge experiments and the components' or diagnostics' operators during autonomous mode or even laboratory experiments. The main ' Xcontrol ' features, such as program composition and validation, manual and automatic control instruments, resource survey, and process monitoring, will be presented. The implementation principles and the underlying communication will be discussed. (author)

Operation and control of ITER plasmas

International Nuclear Information System (INIS)

2001-01-01

Features incorporated in the design of the International Thermonuclear Experimental Reactor (ITER) tokamak and its ancillary and plasma diagnostic systems that will facilitate operation and control of ignited and/or high-Q DT plasmas are presented. Control methods based upon straight-forward extrapolation of techniques employed in the present generation of tokamaks are found to be adequate and effective for DT plasma control with burn durations of ≥1000 s. Examples of simulations of key plasma control functions including magnetic configuration control and fusion burn (power) control are given. The prospects for the creation and control of steady-state plasmas sustained by non-inductive current drive are also discussed. (author)

Operation and control of ITER plasmas

International Nuclear Information System (INIS)

1999-01-01

Features incorporated in the design of the International Thermonuclear Experimental Reactor (ITER) tokamak and its ancillary and plasma diagnostic systems that will facilitate operation and control of ignited and/or high-Q DT plasmas are presented. Control methods based upon straight-forward extrapolation of techniques employed in the present generation of tokamaks are found to be adequate and effective for DT plasma control with burn durations of ≥1000 s. Examples of simulations of key plasma control functions including magnetic configuration control and fusion burn (power) control are given. The prospects for the creation and control of steady-state plasmas sustained by non-inductive current drive are also discussed. (author)

Gas box control system for Tandem Mirror Experiment-Upgrade

International Nuclear Information System (INIS)

Bell, H.H. Jr.; Hunt, A.L.; Clower, C.A. Jr.

1983-01-01

The Tandem Mirror Experiment-Upgrade (TMX-U) uses several methods to feed gas (usually deuterium) at different energies into the plasma region of the machine. One is an arrangement of eight high-speed piezo-electric valves mounted on special manifolds (gas box) that feed cold gas directly to the plasma. This paper describes the electronic valve control and data acquisition portions of the gas box, which are controlled by a desk-top computer. Various flow profiles have been developed and stored in the control computer for ready access by the operator. The system uses two modes of operation, one that exercises and characterizes the valves and one that operates the valves with the rest of the experiment. Both the valve control signals and the pressure transducers data are recorded on the diagnostics computer so that they are available for experiment analysis

Experiment and simulation on one-dimensional plasma photonic crystals

International Nuclear Information System (INIS)

Zhang, Lin; Ouyang, Ji-Ting

2014-01-01

The transmission characteristics of microwaves passing through one-dimensional plasma photonic crystals (PPCs) have been investigated by experiment and simulation. The PPCs were formed by a series of discharge tubes filled with argon at 5 Torr that the plasma density in tubes can be varied by adjusting the discharge current. The transmittance of X-band microwaves through the crystal structure was measured under different discharge currents and geometrical parameters. The finite-different time-domain method was employed to analyze the detailed properties of the microwaves propagation. The results show that there exist bandgaps when the plasma is turned on. The properties of bandgaps depend on the plasma density and the geometrical parameters of the PPCs structure. The PPCs can perform as dynamical band-stop filter to control the transmission of microwaves within a wide frequency range

Twentyseventh European physical society conference on controlled fusion and plasma physics

International Nuclear Information System (INIS)

Igitkhanov, Y.

2000-01-01

The twentyseventh European physical society conference on controlled fusion and plasma physics was held in Budapest, 12-16 June 2000. About 10 invited papers were presented, covering a wide range of problems in plasma physics, including confinement and transport issues in fusion devices, astrophysics and industrial application of plasmas. More than 100 papers were presented on plasma theory and experiments from tokamaks and stellarators. Some of the ITER-relevant issues covered are described in this newsletter

Developments in remote participation in plasma physics experiments

International Nuclear Information System (INIS)

Blackwell, B.

1999-01-01

Recent growth in the size of plasma experiments and developments in network based software have contributed to a high level of interest in remote participation. Highlights of the recent conferences on this subject, and the ensuing 'white paper' are presented, with demonstrations of various Data Server/Web/Java based remote access techniques. These not only allow AINSE/AFRG users convenient access to H-1NF data from their home laboratory, but are (or soon will be) available to and from many overseas laboratories with similar systems. Many large plasma laboratories predict a large increase in remote access in the next two years. Several demonstrations of remote experiment control have been performed over medium speed networks, and several new experiments are planning on remote access from the beginning. In this paper we consider data access rights and security, access to common documents, and access to processed and raw data. The full version of this document can be viewed on the ANU's H-1NF web page at: http://rsphysse.anu.edu.au/

Streaming-plasma measurements in the Baseball II-T mirror experiment

International Nuclear Information System (INIS)

Damm, C.C.; Foote, J.H.; Futch, A.H.; Goodman, R.K.; Hornady, R.S.; Osher, J.E.; Porter, G.D.

1977-01-01

The warm plasma from a deuterium-loaded titanium washer gun, streaming along magnetic-field lines through the steady-state magnetic well of Baseball II, has been examined for its suitability in this experimental situation as a target plasma for hot-ion buildup experiments and for microinstability control. The gun was positioned near the magnetic axis outside the mirror region. Measurements were made with gridded, end-loss detectors placed outside the opposite mirror, a microwave interferometer, a beam-attenuation detector, and other diagnostics

Automatic plasma control in magnetic traps

International Nuclear Information System (INIS)

Samojlenko, Y.; Chuyanov, V.

1984-01-01

Hot plasma is essentially in thermodynamic non-steady state. Automatic plasma control basically means monitoring deviations from steady state and producing a suitable magnetic or electric field which brings the plasma back to its original state. Briefly described are two systems of automatic plasma control: control with a magnetic field using a negative impedance circuit, and control using an electric field. It appears that systems of automatic plasma stabilization will be an indispensable component of the fusion reactor and its possibilities will in many ways determine the reactor economy. (Ha)

Development of a plasma driven permeation experiment for TPE

Energy Technology Data Exchange (ETDEWEB)

Buchenauer, Dean, E-mail: dabuche@sandia.gov [Sandia National Laboratories, Livermore, CA (United States); Kolasinski, Robert [Sandia National Laboratories, Livermore, CA (United States); Shimada, Masa [Idaho National Laboratory, Idaho Falls, ID (United States); Donovan, David [Sandia National Laboratories, Livermore, CA (United States); Youchison, Dennis [Sandia National Laboratories, Albuquerque, NM (United States); Merrill, Brad [Idaho National Laboratory, Idaho Falls, ID (United States)

2014-10-15

Highlights: • We have designed and fabricated a novel tritium permeation membrane holder for use in the Tritium Plasma Experiment (TPE). • The membrane temperature is controlled by varying the cooling flow rate and proximity of a spiral cooling channel. • Sealing tests have demonstrated adequate helium leak rates up to temperatures of 1000 °C. • Flow modeling indicates a minimal helium pressure drop across the membrane holder (<700 Pa). • Thermal modeling shows good heat removal and minimal membrane temperature variation (±2%) even up to peak TPE ion fluxes. - Abstract: Experiments on retention of hydrogen isotopes (including tritium) at temperatures less than 800 °C have been carried out in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory [1,2]. To provide a direct measurement of plasma driven permeation in plasma facing materials at temperatures reaching 1000 °C, a new TPE membrane holder has been built to hold test specimens (≤1 mm in thickness) at high temperature while measuring tritium permeating through the membrane from the plasma facing side. This measurement is accomplished by employing a carrier gas that transports the permeating tritium from the backside of the membrane to ion chambers giving a direct measurement of the plasma driven tritium permeation rate. Isolation of the membrane cooling and sweep gases from TPE's vacuum chamber has been demonstrated by sealing tests performed up to 1000 °C of a membrane holder design that provides easy change out of membrane specimens between tests. Simulations of the helium carrier gas which transports tritium to the ion chamber indicate a very small pressure drop (∼700 Pa) with good flow uniformity (at 1000 sccm). Thermal transport simulations indicate that temperatures up to 1000 °C are expected at the highest TPE fluxes.

Development of a plasma driven permeation experiment for TPE

International Nuclear Information System (INIS)

Buchenauer, Dean; Kolasinski, Robert; Shimada, Masa; Donovan, David; Youchison, Dennis; Merrill, Brad

2014-01-01

Highlights: • We have designed and fabricated a novel tritium permeation membrane holder for use in the Tritium Plasma Experiment (TPE). • The membrane temperature is controlled by varying the cooling flow rate and proximity of a spiral cooling channel. • Sealing tests have demonstrated adequate helium leak rates up to temperatures of 1000 °C. • Flow modeling indicates a minimal helium pressure drop across the membrane holder (<700 Pa). • Thermal modeling shows good heat removal and minimal membrane temperature variation (±2%) even up to peak TPE ion fluxes. - Abstract: Experiments on retention of hydrogen isotopes (including tritium) at temperatures less than 800 °C have been carried out in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory [1,2]. To provide a direct measurement of plasma driven permeation in plasma facing materials at temperatures reaching 1000 °C, a new TPE membrane holder has been built to hold test specimens (≤1 mm in thickness) at high temperature while measuring tritium permeating through the membrane from the plasma facing side. This measurement is accomplished by employing a carrier gas that transports the permeating tritium from the backside of the membrane to ion chambers giving a direct measurement of the plasma driven tritium permeation rate. Isolation of the membrane cooling and sweep gases from TPE's vacuum chamber has been demonstrated by sealing tests performed up to 1000 °C of a membrane holder design that provides easy change out of membrane specimens between tests. Simulations of the helium carrier gas which transports tritium to the ion chamber indicate a very small pressure drop (∼700 Pa) with good flow uniformity (at 1000 sccm). Thermal transport simulations indicate that temperatures up to 1000 °C are expected at the highest TPE fluxes

Control System Development Plan for the National Spherical Torus Experiment

International Nuclear Information System (INIS)

Neumeyer, C.; Mueller, D.; Gates, D.A.; Ferron, J.R.

1999-01-01

The National Spherical Torus Experiment (NSTX) has as one of its primary goals the demonstration of the attractiveness of the spherical torus concept as a fusion power plant. Central to this goal is the achievement of high plasma β ( = 2 micro 0 /B 2 a measure of the efficiency of a magnetic plasma confinement system). It has been demonstrated both theoretically and experimentally that the maximum achievable β is a strong function of both local and global plasma parameters. It is therefore important to optimize control of the plasma. To this end a phased development plan for digital plasma control on NSTX is presented. The relative level of sophistication of the control system software and hardware will be increased according to the demands of the experimental program in a three phase plan. During Day 0 (first plasma), a simple coil current control algorithm will initiate plasma operations. During the second phase (Day 1) of plasma operations the control system will continue to use the preprogrammed algorithm to initiate plasma breakdown but will then change over to a rudimentary plasma control scheme based on linear combinations of measured plasma fields and fluxes. The third phase of NSTX plasma control system development will utilize the rtEFIT code, first used on DIII-D, to determine, in real-time, the full plasma equilibrium by inverting the Grad-Shafranov equation. The details of the development plan, including a description of the proposed hardware will be presented

A microfluidic chip for blood plasma separation using electro-osmotic flow control

International Nuclear Information System (INIS)

Jiang, Hai; Weng, Xuan; Chon, Chan Hee; Wu, Xudong; Li, Dongqing

2011-01-01

In this paper, a microfluidic-based chip with two straight microchannels and five branch microchannels was designed and tested to separate blood plasma from a small sample of fresh human blood. The electro-osmotic flow method was used to control the separation of blood plasma. Blood cell removal and blood plasma extraction were realized in experiments. The efficiency of extracting blood plasma can be as high as 26%

Real time control of plasmas and ECRH systems on TCV

NARCIS (Netherlands)

Paley, J.I.; Felici, F.; Berrino, J.; Coda, S.; Cruz, N.; Duval, B.P.; Goodman, T.P.; Martin, Y.; Moret, J.-M.; Piras, F.; Rodrigues, A.P.; Santos, B.; Varandas, C.A.F.

2008-01-01

Developments in the real time control hardware on TCV paired with the flexibility of plasma shaping and ECRH actuators are opening many opportunities to perform real time experiments and develop algorithms and methods for fusion applications. The ability to control MHD instabilities is particularly

Edge plasma control using an LID configuration on CHS

Energy Technology Data Exchange (ETDEWEB)

Masuzaki, S.; Komori, A.; Morisaki, T. [National Inst. for Fusion Science, Oroshi, Toki (Japan)] [and others

1997-07-01

A Local Island Divertor (LID) has been proposed to enhance energy confinement through neutral particle control. For the case of the Large Helical Device (LHD), the separatrix of an m/n = 1/1 magnetic island, formed at the edge region, will be utilized as a divertor configuration. The divertor head is inserted in the island, and the island separatrix provides connection between the edge plasma region surrounding the core plasma and the back plate of the divertor head through the field lines. The particle flux and associated heat flux from the core plasma strike the back plate of the divertor head, and thus particle recycling is localized in this region. A pumping duct covers the divertor head to form a closed divertor system for efficient particle exhaust. The advantages of the LID are ease of hydrogen pumping because of the localized particle recycling and avoidance of the high heat load that would be localized on the leading edge of the divertor head. With efficient pumping, the neutral pressure in the edge plasma region will be reduced, and hence the edge plasma temperature will be higher, hopefully leading to a better core confinement region. A LID configuration experiment was done on the Compact Helical System (CHS) to confirm the effect of the LID. The typical effects of the LID configuration on the core plasma are reduction of the line averaged density to a half, and small or no reduction of the stored energy. In this contribution, the experimental results which were obtained in edge plasma control experiments with the LID configuration in the CHS are presented.

An experimental study of icing control using DBD plasma actuator

Science.gov (United States)

Cai, Jinsheng; Tian, Yongqiang; Meng, Xuanshi; Han, Xuzhao; Zhang, Duo; Hu, Haiyang

2017-08-01

Ice accretion on aircraft or wind turbine has been widely recognized as a big safety threat in the past decades. This study aims to develop a new approach for icing control using an AC-DBD plasma actuator. The experiments of icing control (i.e., anti-/de-icing) on a cylinder model were conducted in an icing wind tunnel with controlled wind speed (i.e., 15 m/s) and temperature (i.e., -10°C). A digital camera was used to record the dynamic processes of plasma anti-icing and de-icing whilst an infrared imaging system was utilized to map the surface temperature variations during the anti-/de-icing processes. It was found that the AC-DBD plasma actuator is very effective in both anti-icing and de-icing operations. While no ice formation was observed when the plasma actuator served as an anti-icing device, a complete removal of the ice layer with a thickness of 5 mm was achieved by activating the plasma actuator for ˜150 s. Such information demonstrated the feasibility of plasma anti-/de-icing, which could potentially provide more effective and safer icing mitigation strategies.

Initial operation of NSTX with plasma control

International Nuclear Information System (INIS)

Gates, D.; Bell, M.; Ferron, J.; Kaye, S.; Menard, J.; Mueller, D.; Neumeyer, C.; Sabbagh, S.

2000-01-01

First plasma, with a maximum current of 300kA, was achieved on NSTX in February 1999. These results were obtained using preprogrammed coil currents. The first controlled plasmas on NSTX were made starting in August 1999 with the full 1MA plasma current achieved in December 1999. The controlled quantities were plasma position (R, Z) and current (Ip). Variations in the plasma shape are achieved by adding preprogrammed currents to those determined by the control parameters. The control system is fully digital, with plasma position and current control, data acquisition, and power supply control all occurring in the same four-processor real time computer. The system uses the PCS (Plasma Control Software) system designed at General Atomics. Modular control algorithms, specific to NSTX, were written and incorporated into the PCS. The application algorithms do the actual control calculations, with the PCS handling data passing. The control system, including planned upgrades, will be described, along with results of the initial controlled plasma operations. Analysis of the performance of the control system will also be presented

Edge localized modes control: experiment and theory

International Nuclear Information System (INIS)

Bedoulet, M.; Huysmans, G.; Thomas, P.; Joffrin, E.; Rimini, F.; Monier-Garbet, P.; Grosman, A.; Ghendrih, P.; Parail, V.; Lomas, P.; Matthews, G.; Wilson, H.; Gryaznevich, M.; Gonsell, G.; Loarte, A.; Saibene, G.; Sartori, R.; Leonard, A.; Snyder, P.; Evans, T.; Gohil, P.; Burell, H.; Moyer, R.; Kamada, Y.; Oyama, N.; Hatae, T.; Degeling, A.; Martin, Y.; Lister, J.; Rapp, J.; Perez, C.; Lang, P.; Chankin, A.; Eich, T.; Sips, A.; Stober, J.; Horton, L.; Kallenbach, A.; Suttrop, W.; Saarelma, S.; Cowley, S.; Lonnroth, J.; Kamiya, K.; Shimada, M.; Polevoi, A.; Federici, G.

2004-01-01

The paper reviews recent theoretical and experimental results focusing on the identification of the key factors controlling ELM (energy localized mode) energy and particle losses both in natural ELMs and in the presence of external controlling mechanisms. The theoretical description of the most studied Type-I ELMs is progressing from linear MHD stability analysis for peeling and ballooning modes to the non-linear explosive models and transport codes. Present theories cannot predict the ELM size self-consistently, however they pointed out the benefit of the high plasma shaping, high q 95 and high pedestal density in reducing the ELM affected area. The experimental data also suggest that the conductive energy losses in Type-I ELM can be controlled by working in specific plasma conditions. In particular, the existence of purely convective small Type-I ELMs regimes at high q 95 (>4.5) with ΔW ELM /W ped <5% was demonstrated in high triangularity (δ ∼ 0.5) plasmas in JET. Small benign ELMs regimes in present machines (EDA, HRS, Type-II, grassy, QH, Type-III in impurity seeded discharges at high δ and their relevance for ITER parameters are reviewed briefly. The absence of already developed ITER relevant high confinement scenarios with acceptable ELMs has motivated recent intensive experimental and theoretical studies of active control of ELMs. The possibility of suppression of Type-I ELMs in H-mode scenarios at constant confinement was demonstrated in DIII-D experiments with a stochastic boundary created by external coils. It has been demonstrated in AUG that small pellets can trigger Type-I ELMs with a frequency imposed by the pellet injector. Pellet induced ELMs are similar to the intrinsic Type-I ELMs with the same frequency. At the same time the confinement degradation due to the fuelling can be minimized with pellets small as compared to the gas injection. Recent plasma current ramp experiments (JET, COMPASS-D) and modelling (JETTO) demonstrated that the edge

Edge localized modes control: experiment and theory

Energy Technology Data Exchange (ETDEWEB)

Bedoulet, M.; Huysmans, G.; Thomas, P.; Joffrin, E.; Rimini, F.; Monier-Garbet, P.; Grosman, A.; Ghendrih, P. [Association Euratom-CEA, Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Parail, V.; Lomas, P.; Matthews, G.; Wilson, H.; Gryaznevich, M.; Gonsell, G.; Loarte, A.; Saibene, G.; Sartori, R.; Leonard, A.; Snyder, P.; Evans, T.; Gohil, P.; Burell, H.; Moyer, R.; Kamada, Y.; Oyama, N.; Hatae, T.; Degeling, A.; Martin, Y.; Lister, J.; Rapp, J.; Perez, C.; Lang, P.; Chankin, A.; Eich, T.; Sips, A.; Stober, J.; Horton, L.; Kallenbach, A.; Suttrop, W.; Saarelma, S.; Cowley, S.; Lonnroth, J.; Kamiya, K.; Shimada, M.; Polevoi, A.; Federici, G

2004-07-01

The paper reviews recent theoretical and experimental results focusing on the identification of the key factors controlling ELM (energy localized mode) energy and particle losses both in natural ELMs and in the presence of external controlling mechanisms. The theoretical description of the most studied Type-I ELMs is progressing from linear MHD stability analysis for peeling and ballooning modes to the non-linear explosive models and transport codes. Present theories cannot predict the ELM size self-consistently, however they pointed out the benefit of the high plasma shaping, high q{sub 95} and high pedestal density in reducing the ELM affected area. The experimental data also suggest that the conductive energy losses in Type-I ELM can be controlled by working in specific plasma conditions. In particular, the existence of purely convective small Type-I ELMs regimes at high q{sub 95} (>4.5) with {delta}W{sub ELM}/W{sub ped}<5% was demonstrated in high triangularity ({delta} {approx} 0.5) plasmas in JET. Small benign ELMs regimes in present machines (EDA, HRS, Type-II, grassy, QH, Type-III in impurity seeded discharges at high {delta} and their relevance for ITER parameters are reviewed briefly. The absence of already developed ITER relevant high confinement scenarios with acceptable ELMs has motivated recent intensive experimental and theoretical studies of active control of ELMs. The possibility of suppression of Type-I ELMs in H-mode scenarios at constant confinement was demonstrated in DIII-D experiments with a stochastic boundary created by external coils. It has been demonstrated in AUG that small pellets can trigger Type-I ELMs with a frequency imposed by the pellet injector. Pellet induced ELMs are similar to the intrinsic Type-I ELMs with the same frequency. At the same time the confinement degradation due to the fuelling can be minimized with pellets small as compared to the gas injection. Recent plasma current ramp experiments (JET, COMPASS-D) and

US plans for burning plasma experiments

International Nuclear Information System (INIS)

Nelson, D.

1982-01-01

The first US burning plasma experiment will be the TFTR at Princeton Plasma Physics Laboratory. The initial start-up with hydrogen is expected in December, 1983. The experiment by D-T reaction will begin in 1986. Because of the lack of shielding capability, later experiment is not yet defined. The informal scientific interaction with JET (European project) is kept. The design work on the Fusion Engineering Device (FED) continues, but is delayed. US fusion laboratories collaborated with IPP-Garching on the conceptual design of Zephyr experiment. The US continues to participate in INTOR activities, and will investigate into the critical issues relevant to both INTOR and FED in coming years. (Kato, T.)

2XIIB plasma confinement experiments

International Nuclear Information System (INIS)

Coensgen, F.H.; Clauser, J.F.; Correll, D.L.

1976-01-01

This paper reports results of 2XIIB neutral-beam injection experiments with plasma-stream stabilization. The plasma stream is provided either by a pulsed plasma generator located on the field lines outside the plasma region or by ionization of neutral gas introduced at the mirror throat. In the latter case, the gas is ionized by the normal particle flux through the magnetic mirror. A method of plasma startup and sustenance in a steady-state magnetic field is reported in which the plasma stream from the pulsed plasma generator serves as the initial target for the neutral beams. After an energetic plasma of sufficient density is established, the plasma generator stream is replaced by the gas-fed stream. Lifetimes of the stabilized plasma increase with plasma temperature in agreement with the plasma stabilization of the drift-cyclotron loss-cone mode. The following plasma parameters are attained using the pulsed plasma generator for stabilization: n approximately 5 x 10 13 cm -3 , anti W/sub i/ approximately 13 keV, T/sub e/ = 140 eV, and ntau/sub p/ approximately 7 x 10 10 cm -3 .s. With the gas feed, the mean deuterium ion energy is 9 keV and the peak density n approximately 10 14 cm -3 . In the latter case, the energy confinement parameter reaches ntau/sub E/ = 7 x 10 10 cm -3 .s, and the particle confinement parameter reaches ntau/sub p/ = 1 x 10 11 cm -3 .s

Plasma shape control calculations for BPX divertor design

International Nuclear Information System (INIS)

Strickler, D.J.; Neilson, G.H.; Jardin, S.C.; Pomphrey, N.

1991-01-01

The Burning Plasma Experiment (BPX) divertor is to be capable of withstanding heat loads corresponding to ignited operation and 500 MW of fusion power for a current rise time and flattop lasting several seconds. The poloidal field (PF), diagnostic, and feedback equilibrium control systems must provide precise X-point position control in order to sweep the separatrices across the divertor target surface and optimally distribute the heat loads. A control matrix MHD equilibrium code, BEQ, and the Tokamak Simulation Code (TSC) are used to compute preprogrammed double-null (DN) divertor sweep trajectories that maximize sweep distance while simultaneously satisfying a set of strict constraints: minimum lengths of the field lines between the X-point and strike points, minimum spacing between the inboard plasma edge and the limiter, maximum spacing between the outboard plasma edge and the ICRF antennas, minimum safety factor, and linked poloidal flux. A sequence of DN diverted equilibria and a consistent TSC fiducial discharge simulation are used in evaluating the performance of the BPX divertor shape and possible modifications. 5 refs., 10 figs

Recent plasma control progress on EAST

Energy Technology Data Exchange (ETDEWEB)

Xiao, B.J., E-mail: bjxiao@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Yuan, Q.P. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Humphreys, D.A.; Walker, M.L.; Hyatt, A.W.; Leuer, J.A.; Jackson, G.L. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Mueller, D. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Penaflor, B.G.; Pigrowski, D.A.; Johnson, R.D.; Welander, A.S. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Zhang, R.R.; Luo, Z.P.; Guo, Y.; Xing, Z.; Zhang, Y. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China)

2012-12-15

In recent 2 years, various algorithms to control plasma shape, current and density have been implemented or improved for EAST tokamak. These plasma control performances have been verified by either simulated or actual experimental operation, and thus plasma control basis has been established for the long pulse operation and high performance H-mode plasma operation with low hybrid wave (LHW) and ion cyclotron resonance frequency (ICRF) heating. Startup simulation has been done by using TOKSYS code for the plasma breakdown in either 3.1 Wb or 4.5 Wb initial poloidal flux state and the scenarios proved to be robust and used for routine operation. Various shape configurations have been well feedback controlled by using ISOFLUX limited, double-null or single null algorithms based on RTEFIT equilibrium reconstruction. For the long pulse operation, strike point control and magnetics drift compensation have been implemented in the plasma control system (PCS). To improve the operation safety and efficiency, the verification of magnetic diagnostics before plasma breakdown has been demonstrated adequate to prevent a discharge in case of key sensor failure.

Demonstration of sawtooth period control with EC waves in KSTAR plasma

Directory of Open Access Journals (Sweden)

Jeong J. H.

2015-01-01

Full Text Available The sawtooth period control in tokamak is important issue in recent years because the sawtooth crash can trigger TM/NTM instabilities and drive plasmas unstable. The control of sawtooth period by the modification of local current profile near the q=1 surface using ECCD has been demonstrated in a number of tokamaks [1, 2] including KSTAR. As a result, developing techniques to control the sawtooth period as a way of controlling the onset of NTM has been an important area of research in recent years [3]. In 2012 KSTAR plasma campaign, the sawtooth period control is carried out by the different deposition position of EC waves across the q=1 surface. The sawtooth period is shortened by on-axis co-ECCD (destabilization, and the stabilization of the sawtooth is also observed by off-axis co-ECCD at outside q=1 surface. In 2013 KSTAR plasma campaign, the sawtooth locking experiment with periodic forcing of 170 GHz EC wave is carried out to control the sawtooth period. The optimal target position which lengthens the sawtooth period is investigated by performing a scan of EC beam deposition position nearby q=1 surface at the toroidal magnetic field of 2.9 T and plasma current of 0.7 MA. The sawtooth locking by the modulated EC beam is successfully demonstrated as in [3-5] with the scan of modulation-frequency and duty-ratio at the low beta (βN~0.5 plasma. In this paper, the sawteeth behavior by the location of EC beam and the preliminary result of the sawtooth locking experiments in KSTAR will be presented.

Laboratory simulation of the formation of an ionospheric depletion using Keda Space Plasma EXperiment (KSPEX

Directory of Open Access Journals (Sweden)

Pengcheng Yu

2017-10-01

Full Text Available In the work, the formation of an ionospheric depletion was simulated in a controlled laboratory plasma. The experiment was performed by releasing chemical substance sulfur hexafluoride (SF6 into the pure argon discharge plasma. Results indicate that the plasma parameters change significantly after release of chemicals. The electron density is nearly depleted due to the sulfur hexafluoride-electron attachment reaction; and the electron temperature and space potential experience an increase due to the decrease of the electron density. Compared to the traditional active release experiments, the laboratory scheme can be more efficient, high repetition rate and simpler measurement of the varying plasma parameter after chemical releasing. Therefore, it can effective building the bridge between the theoretical work and real space observation.

Electron energy distribution function control in gas discharge plasmas

International Nuclear Information System (INIS)

Godyak, V. A.

2013-01-01

The formation of the electron energy distribution function (EEDF) and electron temperature in low temperature gas discharge plasmas is analyzed in frames of local and non-local electron kinetics. It is shown, that contrary to the local case, typical for plasma in uniform electric field, there is the possibility for EEDF modification, at the condition of non-local electron kinetics in strongly non-uniform electric fields. Such conditions “naturally” occur in some self-organized steady state dc and rf discharge plasmas, and they suggest the variety of artificial methods for EEDF modification. EEDF modification and electron temperature control in non-equilibrium conditions occurring naturally and those stimulated by different kinds of plasma disturbances are illustrated with numerous experiments. The necessary conditions for EEDF modification in gas discharge plasmas are formulated

RFP plasma experiment at INPE

International Nuclear Information System (INIS)

Ueda, M.; Aso, Y.

1988-01-01

Plasma experiments in CECI, a small Reversed Field Pinch (RFP) apparatus, are described. Preliminary measurements in this device shown the production of a plasma with peak current of 1.3kA and discharge duration of nearly 80μs, when a toroidal DC field of 100G was used. A loop voltage of 40V was measured and a maximum electron temperature of 3eV was estimated for these discharges. Experimental points in the F-θ diagram for CECI indicate that its plasma is approaching the RFP configuration when the discharge is optimize. The probe data also show that the plasma column expands outward. Numerical results indicate that leakage fields have to be reduced below 5G to form appropriate magnetic surfaces. (author) [pt

Experiments on microsecond conduction time plasma opening switch mechanisms

International Nuclear Information System (INIS)

Rix, W.; Coleman, M.; Miller, A.R.; Parks, D.; Robertson, K.; Thompson, J.; Waisman, E.; Wilson, A.

1993-01-01

The authors describe a series of experiments carried out on ACE 2 and ACE 4 to elucidate the mechanisms controlling the conduction and opening phases on plasma opening switches in a radial geometry with conduction times on the order of a microsecond. The results indicate both conduction and opening physics are similar to that observed on lower current systems in a coaxial geometry

The JET PCU project: An international plasma control project

International Nuclear Information System (INIS)

Sartori, F.; Crisanti, F.; Albanese, R.; Ambrosino, G.; Toigo, V.; Hay, J.; Lomas, P.; Rimini, F.; Shaw, S.R.; Luchetta, A.; Sousa, J.; Portone, A.; Bonicelli, T.; Ariola, M.; Artaserse, G.; Bigi, M.; Card, P.; Cavinato, M.; De Tommasi, G.; Gaio, E.

2008-01-01

This paper describes the new JET enhancement project 'Plasma Control Upgrade' (PCU). Initially aimed at an overhaul of JET plasma control capabilities it was eventually focused on improving the vertical stabilisation (VS) system ability to recover from large ELM (edge localised mode) perturbations. The paper describes the results of the first two years where the activity was aimed principally at researching a solution that could be implemented within the timing and budget constraints. A very important task was that of improving the modelling of JET plasma, iron core and passive structures. Using dedicated experiments, the models were progressively refined until it was possible not just to explain the experimental data but predict the VS system behaviour. At the same time the project team studied the best options for power supply (PS) and control system upgrades and evaluated whether a change of turns in the stabilisation coil was desirable and possible. A new fast radial field power supply is now being ordered and the VS control system is being upgraded

Positron Plasma Control Techniques Applied to Studies of Cold Antihydrogen

CERN Document Server

Funakoshi, Ryo

2003-01-01

In the year 2002, two experiments at CERN succeeded in producing cold antihydrogen atoms, first ATHENA and subsequently ATRAP. Following on these results, it is now feasible to use antihydrogen to study the properties of antimatter. In the ATHENA experiment, the cold antihydrogen atoms are produced by mixing large amounts of antiprotons and positrons in a nested Penning trap. The complicated behaviors of the charged particles are controlled and monitored by plasma manipulation techniques. The antihydrogen events are studied using position sensitive detectors and the evidence of production of antihydrogen atoms is separated out with the help of analysis software. This thesis covers the first production of cold antihydrogen in the first section as well as the further studies of cold antihydrogen performed by using the plasma control techniques in the second section.

The plasma focus - numerical experiments leading technology

International Nuclear Information System (INIS)

Saw, S.H.; Lee, S.

2013-01-01

Numerical experiments on the plasma focus are now used routinely to assist design and provide reference points for diagnostics. More importantly guidance has been given regarding the implementation of technology for new generations of plasma focus devices. For example intensive series of experiments have shown that it is of no use to reduce static bank inductance L0 below certain values because of the consistent loading effects of the plasma focus dynamics on the capacitor bank. Thus whilst it was thought that the PF1000 could receive major benefits by reducing its bank inductance L 0 , numerical experiments have shown to the contrary that its present L 0 of 30 nH is already optimum and that reducing L 0 would be a very expensive fruitless exercise. This knowledge gained from numerical experiments now acts as a general valuable guideline to all high performance (ie low inductance) plasma focus devices not to unnecessarily attempt to further lower the static inductance L 0 . The numerical experiments also show that the deterioration of the yield scaling law (e.g. the fusion neutron yield scaling with storage energy) is inevitable again due to the consistent loading effect of the plasma focus, which becomes more and more dominant as capacitor bank impedance reduces with increasing capacitance C 0 as storage energy is increased. This line of thinking has led to the suggestion of using higher voltages (as an alternative to increasing C 0 ) and to seeding of Deuterium with noble gases in order to enhance compression through thermodynamic mechanisms and through radiation cooling effects of strong line radiation. Circuit manipulation e.g. to enhance focus pinch compression by current-stepping is also being numerically experimented upon. Ultimately however systems have to be built, guided by numerical experiments, so that the predicted technology may be proven and realized. (author)

New achievements in the EAST plasma control system

Energy Technology Data Exchange (ETDEWEB)

Yuan, Q.P., E-mail: qpyuan@ipp.ac.c [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Xiao, B.J. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Penaflor, B.G.; Piglowski, D.A. [General Atomics, DIII-D National Fusion Facility, San Diego, CA (United States); Liu, L.Z. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Johnson, R.D.; Walker, M.L.; Humphreys, D.A. [General Atomics, DIII-D National Fusion Facility, San Diego, CA (United States)

2010-07-15

In order to realize the low latency and distortion-free signal transmission between the plasma control system (PCS) and servo systems, the digital output structure configured with reflective memory board (RFM) was adopted in EAST PCS. And the enhanced performances are reported. Another achievement made in the latest EAST PCS was the implementation of density control algorithm, which controlled the line average density in either voltage or width modulation mode. The new integrated algorithm improved the precision of density calculation and control performance greatly. The details and experiment results are presented in this paper.

Validation of ISTTOK Plasma Position Controller

International Nuclear Information System (INIS)

Valcarcel, D. F.; Carvalho, I. S.; Carvalho, B. B.; Fernandes, H.; Silva, C.; Duarte, P.; Duarte, A.; Carvalho, P. J.; Pereira, T.

2008-01-01

Active control of plasma position on the ISTTOK tokamak is of extreme importance due to the inherent instability caused by an unfavourable curvature of the external equilibrium magnetic field. The consequences of this instability can be suppressed by applying a dynamic equilibrium field. A digital real-time plasma position control system for ISTTOK has been developed to perform this task. This system uses magnetic measurements to determine the plasma position and feeds the control signal to power supplies that generate the equilibrium fields. After commissioning, the results obtained have shown some discrepancies between the magnetic plasma position reconstruction and several other diagnostics, such as tomography. This discrepancy at some extent is due to the effect of the external magnetic fields on the poloidal magnetic measurements. This work presents a study that addresses this issue. In a future work it will lead to the development of a corrected plasma position algorithm, aiming at obtaining improved performance of plasma discharges and controlled plasma column displacements

Dielectric barrier discharge plasma actuator for flow control

Science.gov (United States)

Opaits, Dmitry Florievich

Electrohydrodynamic (EHD) and magnetohydrodynamic phenomena are being widely studied for aerodynamic applications. The major effects of these phenomena are heating of the gas, body force generation, and enthalpy addition or extraction, [1, 2, 3]. In particular, asymmetric dielectric barrier discharge (DBD) plasma actuators are known to be effective EHD device in aerodynamic control, [4, 5]. Experiments have demonstrated their effectiveness in separation control, acoustic noise reduction, and other aeronautic applications. In contrast to conventional DBD actuators driven by sinusoidal voltages, we proposed and used a voltage profile consisting of nanosecond pulses superimposed on dc bias voltage. This produces what is essentially a non-self-sustained discharge: the plasma is generated by repetitive short pulses, and the pushing of the gas occurs primarily due to the bias voltage. The advantage of this non-self-sustained discharge is that the parameters of ionizing pulses and the driving bias voltage can be varied independently, which adds flexibility to control and optimization of the actuators performance. Experimental studies were conducted of a flow induced in a quiescent room air by a single DBD actuator. A new approach for non-intrusive diagnostics of plasma actuator induced flows in quiescent gas was proposed, consisting of three elements coupled together: the Schlieren technique, burst mode of plasma actuator operation, and 2-D numerical fluid modeling. During the experiments, it was found that DBD performance is severely limited by surface charge accumulation on the dielectric. Several ways to mitigate the surface charge were found: using a reversing DC bias potential, three-electrode configuration, slightly conductive dielectrics, and semi conductive coatings. Force balance measurements proved the effectiveness of the suggested configurations and advantages of the new voltage profile (pulses+bias) over the traditional sinusoidal one at relatively low

Profile control simulations and experiments on TCV : A controller test environment and results using a model-based predictive controller

NARCIS (Netherlands)

Maljaars, E.; Felici, F.; Blanken, T.C.; Galperti, C.; Sauter, O.; de Baar, M.R.; Carpanese, F.; Goodman, T.P.; Kim, D.; Kim, S.H.; Kong, M.G.; Mavkov, B.; Merle, A.; Moret, J.M.; Nouailletas, R.; Scheffer, M.; Teplukhina, A.A.; Vu, N.M.T.

2017-01-01

The successful performance of a model predictive profile controller is demonstrated in simulations and experiments on the TCV tokamak, employing a profile controller test environment. Stable high-performance tokamak operation in hybrid and advanced plasma scenarios requires control over the safety

Profile control simulations and experiments on TCV: a controller test environment and results using a model-based predictive controller

NARCIS (Netherlands)

Maljaars, B.; Felici, F.; Blanken, T. C.; Galperti, C.; Sauter, O.; de Baar, M. R.; Carpanese, F.; Goodman, T. P.; Kim, D.; Kim, S. H.; Kong, M.; Mavkov, B.; Merle, A.; Moret, J.; Nouailletas, R.; Scheffer, M.; Teplukhina, A.; Vu, T.

2017-01-01

The successful performance of a model predictive profile controller is demonstrated in simulations and experiments on the TCV tokamak, employing a profile controller test environment. Stable high-performance tokamak operation in hybrid and advanced plasma scenarios requires control over the safety

28. Zvenigorod conference on the plasma physics and controlled thermonuclear synthesis. Theses of reports

International Nuclear Information System (INIS)

2001-01-01

Theses of reports, presented at the 28th Conference on the plasma physics and controlled thermonuclear synthesis (Zvenigorod, 19-23 February 2001) are published. 246 reports were heard at the following sections: magnetic confinement, theory and experiments; inertial thermonuclear synthesis; plasma processes and physics of gas-discharge plasma; physical bases of plasma technologies. 17 reports had the summarizing character [ru

QUICK-FIRE: Plasma flow driven implosion experiments

International Nuclear Information System (INIS)

Baker, W.L.; Bigelow, W.S.; Degnan, J.H.

1985-01-01

High speed plasma implosions involving megajoules of energy, and sub-microsecond implosion times are expected to require additional stages of power conditioning between realistic primary energy sources and the implosion system. Plasma flow switches and vacuum inductive stores represent attractive alternates to the high speed fuse and atmospheric store techniques which have been previously reported for powering such plasma experiments. In experiments being conducted at the Air Force Weapons Lab, a washer shaped plasma accelerated to 7-10 cm/microsecond in a coaxial plasma gun configuration, represents the moving element in a vacuum store/power conditioning system of 16.5 nH inductance which stores 1-1.5 MJ at 12-14 MA. At the end of the coaxial gun, the moving element transits the 2cm axial length of the cylindrical implosion gap in 200-400 nS, delivering the magnetic energy to the implosion foil, accelerating the imploding plasma to speeds of 30-40 cm/microsecond in 350-450 nS, and delivering a projected 400 KJ of kinetic energy to the implosion

Plasma position and shape control for ITER

International Nuclear Information System (INIS)

Portone, A.; Gribov, Y.; Huguet, M.

1995-01-01

Key features and main results about the control of the plasma shape in ITER are presented. A control algorithm is designed to control up to 6 gaps between the plasma separatrix and the plasma facing components during the reference burn phase. Nonlinear simulations show the performances of the controller in the presence of plasma vertical position offsets, beta drops and power supply voltage saturation

Conceptual design of plasma position control of SST-1 tokamak using vertical field coil

International Nuclear Information System (INIS)

Gulati, Hitesh Kumar; Patel, Kiritkumar B.; Dhongde, Jasraj

2015-01-01

SST-1 (Steady State Superconducting Tokamak) is a plasma confinement device in Institute for Plasma Research (IPR) India. SST-1 has been commissioned successfully and has been carrying out plasma experiments since the beginning of 2014 achieved a maximum plasma current of 75 kA at a central field of 1.5 T and the plasma duration ∼ 500 ms. SST-1 looks forward to carrying out elongated plasma experiments and stretching plasma pulses beyond 1s. Based on the solution of Grad-Shafranov equation the shift of plasma column center from geometrical centre of vacuum chamber is measured using various magnetic probes and flux loops installed in the machine. The closed feedback loop uses plasma current (Ip), Delta R as feedback signal and manipulate the vertical field current (Ivf). The discharge starts with feed forward loop using initially provided reference then the active feedback starts after discharge by few msec once plasma column is completely formed. The feedback loop time is of the order of 10 msec. The primary objective is to acquire plasma position control related signals, compute plasma position and generate position correction signal for VF coil power supply, communicate correction to VF coil power supply and modify VF power supply output in a deterministic time span. In this we present the methodology used for plasma horizontal displacement control using vertical field and discuss the preliminary results. (author)

Design and Architecture of SST-1 basic plasma control system

Energy Technology Data Exchange (ETDEWEB)

Patel, Kirit, E-mail: kpatel@ipr.res.in; Raju, D.; Dhongde, J.; Mahajan, K.; Chudasama, H.; Gulati, H.; Chauhan, A.; Masand, H.; Bhandarkar, M.; Pradhan, S.

2016-11-15

Highlights: • Reflective Memory network. • FPAG based Timing system for trigger distribution. • IRIG-B network for GPS time synchronization. • PMC based Digital Signal Processors and VME. • Simultaneous sampling ADC. - Abstract: Primary objective of SST-1 Plasma control system is to achieve Plasma position, shape and current profile control. Architecture of control system for SST-1 is distributed in nature. Fastest control loop time requirement of 100 μs is achieved using VME based simultaneous sampling ADCs, PMC based quad core DSP, Reflective Memory [RFM] based real-time network, VME based real-time trigger distribution network and Ethernet network. All the control loops for shape control, position control and current profile control share common signals from Magnetic diagnostic so it is planned to accommodate all the algorithms on the same PMC based quad core DSP module TS C-43. RFM based real-time data network replicate data from one node to next node in a ring network topology at sustained throughput rate of 13.4 MBps. Real-time Timing System network provides guaranteed trigger distribution in 3.8 μs from one node to all node of the network. Monitoring and configuration of different systems participating in the operation of SST-1 is done by Ethernet network. Magnetic sensors data is acquired using Pentek 6802 simultaneously sampling ADC card at the rate of 10KSPS. All the real-time raw data along with the control data will be archived using RFM network and SCSI HDD for the experiment duration of 1000 s. RFM network is also planned for real-time plotting of key parameter of Plasma during long experiment. After experiment this data is transferred to central storage server for archival purpose. This paper discusses the architecture and hardware implementation of the control system by describing all the involved hardware and software along with future plans for up-gradations.

Status of DIII-D plasma control

International Nuclear Information System (INIS)

Walker, M.L.; Ferron, J.R.; Penaflor, B.

1995-10-01

A key component of the DIII-D Advanced Tokamak and Radiative Divertor Programs is the development and implementation of methods to actively control a large number of plasma parameters. These parameters include plasma shape and position, total stored energy, density, rf loading resistance, radiated power and more detailed control of the current profile. To support this research goal, a flexible and easily expanded digital control system has been developed and implemented. We have made parallel progress in modeling of the plasma, poloidal coils, vacuum vessel, and power system dynamics and in ensuring the integrity of diagnostic and command circuits used in control. Recent activity has focused on exploiting the mature digital control platform through the implementation of simple feedback controls of more exotic plasma parameters such as enhanced divertor radiation, neutral pressure and Marfe creation and more sophisticated identification and digital feedback control algorithms for plasma shape, vertical position, and safety factor on axis (q 0 ). A summary of recent progress in each of these areas will be presented

First results of the plasma wakefield acceleration experiment at PITZ

International Nuclear Information System (INIS)

Lishilin, O.; Gross, M.; Brinkmann, R.; Engel, J.; Grüner, F.; Koss, G.; Krasilnikov, M.; Martinez de la Ossa, A.; Mehrling, T.; Osterhoff, J.; Pathak, G.; Philipp, S.; Renier, Y.; Richter, D.; Schroeder, C.; Schütze, R.; Stephan, F.

2016-01-01

The self-modulation instability of long particle beams was proposed as a new mechanism to produce driver beams for proton driven plasma wakefield acceleration (PWFA). The PWFA experiment at the Photo Injector Test facility at DESY, Zeuthen site (PITZ) was launched to experimentally demonstrate and study the self-modulation of long electron beams in plasma. Key aspects for the experiment are the very flexible photocathode laser system, a plasma cell and well-developed beam diagnostics. In this contribution we report about the plasma cell design, preparatory experiments and the results of the first PWFA experiment at PITZ. - Highlights: • A self-modulation mechanism for producing driver beams for PWFA is proposed. • A proof-of-principle experiment is launched at the Photo Injector Test facility at DESY. • The self-modulation instability occurs in long particle beams passing through plasma. • A heat pipe oven and a laser are used to produce plasma.

Plasma position control on Alcator C

International Nuclear Information System (INIS)

Pribyl, P.A.

1981-05-01

The Alcator C MHD equilibrium is investigated from the standpoint of determining the plasma position. A review of equilibrium theory is presented, indicating that the central flux surfaces of the plasma should be displaced about 1 to 2 cm from the outermost. Further, the plasma should have a slightly noncircular cross-section. A comparison is made between the observed and predicted profiles. Flux loops sensitive to plasma position generate the error signal for the feedback control circuit. This measurement agrees with other position-sensitive diagnostics, such as limiter heating, and centroids of density, soft x-ray, and electron cyclotron emission. A linear model is developed for the position control feedback system, including the vertical field SCR supply, plasma, and feedback electronics. Operation of the control system agrees well with that predicted by the model, with acceptable plasma position being maintained for the duration of the discharge. The feedback control system is in daily use for Alcator C runs

Diagnostics for the Plasma Liner Experiment

International Nuclear Information System (INIS)

Lynn, A. G.; Merritt, E.; Gilmore, M.; Hsu, S. C.; Witherspoon, F. D.; Cassibry, J. T.

2010-01-01

The goal of the Plasma Liner Experiment (PLX) is to explore and demonstrate the feasibility of forming imploding spherical ''plasma liners'' via merging high Mach number plasma jets to reach peak liner pressures of ∼0.1 Mbar using ∼1.5 MJ of initial stored energy. Such a system would provide HED plasmas for a variety of fundamental HEDLP, laboratory astrophysics, and materials science studies, as well as a platform for experimental validation of rad-hydro and rad-MHD simulations. It could also prove attractive as a potential standoff driver for magnetoinertial fusion. Predicted parameters from jet formation to liner stagnation cover a large range of plasma density and temperature, varying from n i ∼10 16 cm -3 , T e ≅T i ∼1 eV at the plasma gun mouth to n i >10 19 cm -3 , T e ≅T i ∼0.5 keV at stagnation. This presents a challenging problem for the plasma diagnostics suite which will be discussed.

Active MHD control experiments in RFX-mod

International Nuclear Information System (INIS)

Ortolani, Sergio

2006-01-01

The RFX reversed field pinch experiment has been modified (RFX-mod) to address specific issues of active control of MHD instabilities. A thin shell (τ Bv ∼50 ms) has replaced the old thick one (τ Bv ∼500 ms) and 192 (4 poloidal x 48 toroidal) independently powered saddle coils surround the thin shell forming a cage completely covering the torus. This paper reports the results obtained during the first year of operation. The system has been used with various control scenarios including experiments on local radial field cancellation over the entire torus surface to mimic an ideal wall ('virtual shell') and on single and multiple mode feedback control. Successful virtual shell operation has been achieved leading to: a 3-fold increase in pulse length and well controlled 300 ms pulses(∼6 shell times) up to ∼1 MA plasma current; one order of magnitude reduction of the dominant radial field perturbations at the plasma edge and correspondingly 100% increase in global energy confinement time. Robust feedback stabilization of resistive wall modes has been demonstrated in conditions where rotation does not play a role and multiple unstable modes are present

Implementation of GPU parallel equilibrium reconstruction for plasma control in EAST

Energy Technology Data Exchange (ETDEWEB)

Huang, Yao, E-mail: yaohuang@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Xiao, B.J. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); School of Nuclear Science & Technology, University of Science & Technology of China (China); Luo, Z.P.; Yuan, Q.P.; Pei, X.F. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Yue, X.N. [School of Nuclear Science & Technology, University of Science & Technology of China (China)

2016-11-15

Highlights: • We described parallel equilibrium reconstruction code P-EFIT running on GPU was integrated with EAST plasma control system. • Compared with RT-EFIT used in EAST, P-EFIT has better spatial resolution and full algorithm of EFIT per iteration. • With the data interface through RFM, 65 × 65 spatial grids P-EFIT can satisfy the accuracy and time feasibility requirements for plasma control. • Successful control using ISOFLUX/P-EFIT was established in the dedicated experiment during the EAST 2014 campaign. • This work is a stepping-stone towards versatile ISOFLUX/P-EFIT control, such as real-time equilibrium reconstruction with more diagnostics. - Abstract: Implementation of P-EFIT code for plasma control in EAST is described. P-EFIT is based on the EFIT framework, but built with the CUDA™ architecture to take advantage of massively parallel Graphical Processing Unit (GPU) cores to significantly accelerate the computation. 65 × 65 grid size P-EFIT can complete one reconstruction iteration in 300 μs, with one iteration strategy, it can satisfy the needs of real-time plasma shape control. Data interface between P-EFIT and PCS is realized and developed by transferring data through RFM. First application of P-EFIT to discharge control in EAST is described.

XSC plasma control: Tool development for the session leader

International Nuclear Information System (INIS)

Ambrosino, G.; Albanese, R.; Ariola, M.; Cenedese, A.; Crisanti, F.; Tommasi, G. De; Mattei, M.; Piccolo, F.; Pironti, A.; Sartori, F.; Villone, F.

2005-01-01

A new model-based shape controller (XSC, i.e., eXtreme Shape Controller) able to operate with high elongation and triangularity plasmas has been designed and implemented at JET in 2003. The use of the XSC needs a number of steps, which at present are not automated and therefore imply the involvement of several experts. To help the session leader in preparing an experiment, a number of software tools are needed. The paper describes the SW tools that are currently in the developing phase, and describes the new framework for the preparation of a JET experiment

Progress in the Development of a High Power Helicon Plasma Source for the Materials Plasma Exposure Experiment

Energy Technology Data Exchange (ETDEWEB)

Goulding, Richard Howell [ORNL; Caughman, John B. [ORNL; Rapp, Juergen [ORNL; Biewer, Theodore M. [ORNL; Bigelow, Tim S. [ORNL; Campbell, Ian H. [ORNL; Caneses Marin, Juan F. [ORNL; Donovan, David C. [ORNL; Kafle, Nischal [ORNL; Martin, Elijah H. [ORNL; Ray, Holly B. [ORNL; Shaw, Guinevere C. [ORNL; Showers, Melissa A. [ORNL

2017-09-01

Proto-MPEX is a linear plasma device being used to study a novel RF source concept for the planned Material Plasma Exposure eXperiment (MPEX), which will address plasma-materials interaction (PMI) for nuclear fusion reactors. Plasmas are produced using a large diameter helicon source operating at a frequency of 13.56 MHz at power levels up to 120 kW. In recent experiments the helicon source has produced deuterium plasmas with densities up to ~6 × 1019 m–3 measured at a location 2 m downstream from the antenna and 0.4 m from the target. Previous plasma production experiments on Proto-MPEX have generated lower density plasmas with hollow electron temperature profiles and target power deposition peaked far off axis. The latest experiments have produced flat Te profiles with a large portion of the power deposited on the target near the axis. This and other evidence points to the excitation of a helicon mode in this case.

Modeling and control of plasma rotation for NSTX using neoclassical toroidal viscosity and neutral beam injection

Energy Technology Data Exchange (ETDEWEB)

Goumiri, I. R. [Princeton Univ., NJ (United States). Mechanical and Aerospace Dept.; Rowley, C. W. [Princeton Univ., NJ (United States). Mechanical and Aerospace Dept.; Sabbagh, S. A. [Columbia Univ., New York, NY (United States). Dept. of Applied Physics and Applied Mathematics; Gates, D. A. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Gerhardt, S. P. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Boyer, M. D. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Andre, R. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Kolemen, E. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Taira, K. [Florida State Univ, Dept Mech Engn, Tallahassee, FL USA.

2016-02-19

A model-based feedback system is presented to control plasma rotation in a magnetically confined toroidal fusion device, to maintain plasma stability for long-pulse operation. This research uses experimental measurements from the National Spherical Torus Experiment (NSTX) and is aimed at controlling plasma rotation using two different types of actuation: momentum from injected neutral beams and neoclassical toroidal viscosity generated by three-dimensional applied magnetic fields. Based on the data-driven model obtained, a feedback controller is designed, and predictive simulations using the TRANSP plasma transport code show that the controller is able to attain desired plasma rotation profiles given practical constraints on the actuators and the available measurements of rotation.

Controlling marginally detached divertor plasmas

Science.gov (United States)

Eldon, D.; Kolemen, E.; Barton, J. L.; Briesemeister, A. R.; Humphreys, D. A.; Leonard, A. W.; Maingi, R.; Makowski, M. A.; McLean, A. G.; Moser, A. L.; Stangeby, P. C.

2017-06-01

A new control system at DIII-D has stabilized the inter-ELM detached divertor plasma state for H-mode in close proximity to the threshold for reattachment, thus demonstrating the ability to maintain detachment with minimal gas puffing. When the same control system was instead ordered to hold the plasma at the threshold (here defined as T e  =  5 eV near the divertor target plate), the resulting T e profiles separated into two groups with one group consistent with marginal detachment, and the other with marginal attachment. The plasma dithers between the attached and detached states when the control system attempts to hold at the threshold. The control system is upgraded from the one described in Kolemen et al (2015 J. Nucl. Mater. 463 1186) and it handles ELMing plasmas by using real time D α measurements to remove during-ELM slices from real time T e measurements derived from divertor Thomson scattering. The difference between measured and requested inter-ELM T e is passed to a PID (proportional-integral-derivative) controller to determine gas puff commands. While some degree of detachment is essential for the health of ITER’s divertor, more deeply detached plasmas have greater radiative losses and, at the extreme, confinement degradation, making it desirable to limit detachment to the minimum level needed to protect the target plate (Kolemen et al 2015 J. Nucl. Mater. 463 1186). However, the observed bifurcation in plasma conditions at the outer strike point with the ion B   ×  \

Real time control of plasmas and ECRH systems on TCV

NARCIS (Netherlands)

Paley, J.I.; Berrino, J.; Coda, S.; Cruz, N.; Duval, B.P.; Felici, F.; Goodman, T.P.; Martin, Y.; Moret, J.-M.; Piras, F.; Rodriques, A.P.; Santos, B.; Varandas, C.A.F.

2009-01-01

Developments in the real time control hardware on Tokamak Configuration Variable (TCV) coupled with the flexibility of plasma shaping and electron cyclotron (EC) heating and current drive actuators are opening many opportunities to perform real time experiments and develop algorithms and methods for

The importance of the toroidal magnetic field for the feasibility of a tokamak burning plasma experiment

International Nuclear Information System (INIS)

Mazzucato, E.

2000-01-01

The next step in the demonstration of the scientific feasibility of a tokamak fusion reactor is a DT burning plasma experiment for the study and control of self-heated plasmas. In this paper, the authors examine the role of the toroidal magnetic field on the confinement of a tokamak plasma in the ELMy H-mode regime--the operational regime foreseen for ITER

Quick-fire: Plasma flow driven implosion experiments

International Nuclear Information System (INIS)

Baker, W.L.; Bigelow, W.S.; Degnan, J.H.

1985-01-01

High speed plasma implosions involving megajoules of energy, and sub-microsecond implosion times are expected to require additional stages of power conditioning between realistic primary energy sources and the implosion system. Plasma flow switches and vacuum inductive stores represent attractive alternates to the high speed fuse and atmospheric store techniques which have been previously reported for powering such plasma experiments. In experiments being conducted at the Air Force Weapons Lab, a washer shaped plasma accelerated to 7-10 cm/microsecond in a coaxial plasma gun configuration, represents the moving element in a vacuum store/power conditioning system of 16.5 nH inductance which stores 1-1.5 MJ at 12-14 MA. At the end of the coaxial gun, the moving element transits the 2cm axial length of the cylindrical implosion gap in 200-400 nS, delivering the magnetic energy to the implosion foil, accelerating the imploding plasma to speeds of 30-40 cm/microsecond in 350-450 nS, and delivering a projected 400 KJ of kinetic energy to the implosion. Experiments have been conducted using the SHIVA STAR capacitor bank operating at 6 MJ stored energy in which performance has been monitored by electrical diagnostics, magnetic probes, and axial and radial viewing high speed visible and X-Ray photographs to assess the performance of the coaxial run and coaxial to radial transition. Time and spectrally resolved X-Ray diagnostics are used to assess implosion quality and performance and results are compared to kinematic and MHD models

Plasma automatic control in magnetic traps

International Nuclear Information System (INIS)

Samojlenko, Yu.I.; Chuyanov, V.A.

1983-01-01

Principles of constructing the systems providing a plasma equilibrium and stability in thermonuctear devices are laid down. Operation of the servo system to maintain a plasma equilibrium is described using the tokamak plasma filament as an example. Operation of the system to suppress a flute instability is also described. This system measures electric disturbances on the plasma body surface and controls charge distribution on external electrodes. It is pointed out that systems of automatic control of plasma equilibrium and stability become an essential element of a future thermonuclear reactor and the system potentialities would much determine the reactor economic efficiency

Hot-electron-plasma accumulation in the CIRCE mirror experiment

International Nuclear Information System (INIS)

Bardet, R.; Briand, P.; Dupas, L.; Gormezano, C.; Melin, G.

1975-01-01

In the CIRCE experiment, the plasma is obtained by the trapping of a plasma injected into a magnetic bottle by electron heating at cyclotron resonance. The plasma density lies between 5x10 11 cm -3 and 10 12 cm -3 , the electron temperature is about 100 keV and the ion temperature is in the range of few hundred electronvolts. Gross instabilities are not observed. The ratio of the plasma density to the neutral-gas density inside the plasma is higher than 100. A few kilowatts of r.f. power at 8 GHz are sufficient to obtain these results, a fact which looks encouraging as far as the creation of a more effective fast-neutral-target plasma using the CIRCE-experiment concept is concerned. (author)

Spectroscopic measurements of plasma emission light for plasma-based acceleration experiments

International Nuclear Information System (INIS)

Filippi, F.; Mostacci, A.; Palumbo, L.; Anania, M.P.; Biagioni, A.; Chiadroni, E.; Ferrario, M.; Cianchi, A.; Zigler, A.

2016-01-01

Advanced particle accelerators are based on the excitation of large amplitude plasma waves driven by either electron or laser beams. Future experiments scheduled at the SPARC-LAB test facility aim to demonstrate the acceleration of high brightness electron beams through the so-called resonant Plasma Wakefield Acceleration scheme in which a train of electron bunches (drivers) resonantly excites wakefields into a preformed hydrogen plasma; the last bunch (witness) injected at the proper accelerating phase gains energy from the wake. The quality of the accelerated beam depends strongly on plasma density and its distribution along the acceleration length. The measurements of plasma density of the order of 10 16 –10 17  cm −3 can be performed with spectroscopic measurements of the plasma-emitted light. The measured density distribution for hydrogen filled capillary discharge with both Balmer alpha and Balmer beta lines and shot-to-shot variation are here reported.

Spectroscopic measurements of plasma emission light for plasma-based acceleration experiments

Science.gov (United States)

Filippi, F.; Anania, M. P.; Biagioni, A.; Chiadroni, E.; Cianchi, A.; Ferrario, M.; Mostacci, A.; Palumbo, L.; Zigler, A.

2016-09-01

Advanced particle accelerators are based on the excitation of large amplitude plasma waves driven by either electron or laser beams. Future experiments scheduled at the SPARC_LAB test facility aim to demonstrate the acceleration of high brightness electron beams through the so-called resonant Plasma Wakefield Acceleration scheme in which a train of electron bunches (drivers) resonantly excites wakefields into a preformed hydrogen plasma; the last bunch (witness) injected at the proper accelerating phase gains energy from the wake. The quality of the accelerated beam depends strongly on plasma density and its distribution along the acceleration length. The measurements of plasma density of the order of 1016-1017 cm-3 can be performed with spectroscopic measurements of the plasma-emitted light. The measured density distribution for hydrogen filled capillary discharge with both Balmer alpha and Balmer beta lines and shot-to-shot variation are here reported.

Plasma opening switch experiments on supermite

International Nuclear Information System (INIS)

Mendel, C.W.; Quintenz, J.P.; Rosenthal, S.E.; Savage, M.E.

1988-01-01

Experiments using plasma opening switches with fast field coils and plasmas injected on slow magnetic fields are described. Data showing the measurement of the field penetration into the volume that initially held the plasma fill will be shown. Assuming the plasma is mostly pushed back from the coil, rather than being penetrated by the magnetic field allows the density to be calculated, and gives densities of a few times 10 13 cm -3 for our usual operating range. The data makes it clear that the switch is open well before the initial plasma volume is completely penetrated by the magnetic fields. Additional measurements relating to the magnetic field penetration distance and physical penetration mechanism are presented. Other data presented show a magnetic insulation problem which must be solved before very large voltage multiplication can be accomplished with sufficient switch efficiency

Diagnostics for the plasma liner experiment.

Science.gov (United States)

Lynn, A G; Merritt, E; Gilmore, M; Hsu, S C; Witherspoon, F D; Cassibry, J T

2010-10-01

The goal of the Plasma Liner Experiment (PLX) is to explore and demonstrate the feasibility of forming imploding spherical "plasma liners" via merging high Mach number plasma jets to reach peak liner pressures of ∼0.1 Mbar using ∼1.5 MJ of initial stored energy. Such a system would provide HED plasmas for a variety of fundamental HEDLP, laboratory astrophysics, and materials science studies, as well as a platform for experimental validation of rad-hydro and rad-MHD simulations. It could also prove attractive as a potential standoff driver for magnetoinertial fusion. Predicted parameters from jet formation to liner stagnation cover a large range of plasma density and temperature, varying from n(i)∼10(16) cm(-3), T(e)≈T(i)∼1 eV at the plasma gun mouth to n(i)>10(19) cm(-3), T(e)≈T(i)∼0.5 keV at stagnation. This presents a challenging problem for the plasma diagnostics suite which will be discussed.

A laser plasma beatwave accelerator experiment

International Nuclear Information System (INIS)

Ebrahim, N.A.

1987-03-01

An experiment to test the laser plasma beatware accelerator concept is outlined. A heuristic estimate of the relevant experimental parameters is obtained from fluid theory and considerations of wave-particle interactions. Acceleration of 10 MeV electrons to approximately 70 MeV over a plasma length of 3 cm appears to be feasible. This corresponds to an accelerating gradient of approximately 2.5 GeV/m

Plasma actuators for bluff body flow control

Science.gov (United States)

Kozlov, Alexey V.

The aerodynamic plasma actuators have shown to be efficient flow control devices in various applications. In this study the results of flow control experiments utilizing single dielectric barrier discharge plasma actuators to control flow separation and unsteady vortex shedding from a circular cylinder in cross-flow are reported. This work is motivated by the need to reduce landing gear noise for commercial transport aircraft via an effective streamlining created by the actuators. The experiments are performed at Re D = 20,000...164,000. Circular cylinders in cross-flow are chosen for study since they represent a generic flow geometry that is similar in all essential aspects to a landing gear oleo or strut. The minimization of the unsteady flow separation from the models and associated large-scale wake vorticity by using actuators reduces the radiated aerodynamic noise. Using either steady or unsteady actuation at ReD = 25,000, Karman shedding is totally eliminated, turbulence levels in the wake decrease significantly and near-field sound pressure levels are reduced by 13.3 dB. Unsteady actuation at an excitation frequency of St D = 1 is found to be most effective. The unsteady actuation also has the advantage that total suppression of shedding is achieved for a duty cycle of only 25%. However, since unsteady actuation is associated with an unsteady body force and produces a tone at the actuation frequency, steady actuation is more suitable for noise control applications. Two actuation strategies are used at ReD = 82,000: spanwise and streamwise oriented actuators. Near field microphone measurements in an anechoic wind tunnel and detailed study of the near wake using LDA are presented in the study. Both spanwise and streamwise actuators give nearly the same noise reduction level of 11.2 dB and 14.2 dB, respectively, and similar changes in the wake velocity profiles. The contribution of the actuator induced noise is found to be small compared to the natural shedding

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)

FPGA based Fuzzy Logic Controller for plasma position control in ADITYA Tokamak

International Nuclear Information System (INIS)

Suratia, Pooja; Patel, Jigneshkumar; Rajpal, Rachana; Kotia, Sorum; Govindarajan, J.

2012-01-01

Highlights: ► Evaluation and comparison of the working performance of FLC is done with that of PID Controller. ► FLC is designed using MATLAB Fuzzy Logic Toolbox, and validated on ADITYA RZIP model. ► FLC was implemented on a FPGA. The close-loop testing is done by interfacing FPGA to MATLAB/Simulink. ► Developed FLC controller is able to maintain the plasma column within required range of ±0.05 m and was found to give robust control against various disturbances and faster and smoother response compared to PID Controller. - Abstract: Tokamaks are the most promising devices for obtaining nuclear fusion energy from high-temperature, ionized gas termed as Plasma. The successful operation of tokamak depends on its ability to confine plasma at the geometric center of vacuum vessel with sufficient stability. The quality of plasma discharge in ADITYA Tokamak is strongly related to the radial position of the plasma column in the vacuum vessel. If the plasma column approaches too near to the wall of vacuum vessel, it leads to minor or complete disruption of plasma. Hence the control of plasma position throughout the entire plasma discharge duration is a fundamental requirement. This paper describes Fuzzy Logic Controller (FLC) which is designed for radial plasma position control. This controller is tested and evaluated on the ADITYA RZIP control model. The performance of this FLC was compared with that of Proportional–Integral–Derivative (PID) Controller and the response was found to be faster and smoother. FLC was implemented on a Field Programmable Gate Array (FPGA) chip with the use of a Very High-Speed Integrated-Circuits Hardware Description-Language (VHDL).

Plasma control concepts for ITER

International Nuclear Information System (INIS)

Lister, J.B.; Nieswand, C.

1997-01-01

This overview paper skims over a wide range of issues related to the control of ITER plasmas. Although operation of the ITER project will require extensive developmental work to achieve the degree of control required, there is no indication that any of the identified problems will present overwhelming difficulties compared with the operation of present tokamaks. However, the precision of control required and the degree of automation of the final ITER plasma control system will present a challenge which is somewhat greater than for present tokamaks. In order to operate ITER optimally, integrated use of a large amount of diagnostic information will be necessary, evaluated and interpreted automatically. This will challenge both the diagnostics themselves and their supporting interpretation codes. The intervening years will provide us with the opportunity to implement and evaluate most of the new features required for ITER on existing tokamaks, with the exception of the control of an ignited plasma. (author) 7 figs., 7 refs

Advanced plasma flow simulations of cathodic-arc and ferroelectric plasma sources for neutralized drift compression experiments

Directory of Open Access Journals (Sweden)

Adam B. Sefkow

2008-07-01

Full Text Available Large-space-scale and long-time-scale plasma flow simulations are executed in order to study the spatial and temporal evolution of plasma parameters for two types of plasma sources used in the neutralized drift compression experiment (NDCX. The results help assess the charge neutralization conditions for ion beam compression experiments and can be employed in more sophisticated simulations, which previously neglected the dynamical evolution of the plasma. Three-dimensional simulations of a filtered cathodic-arc plasma source show the coupling efficiency of the plasma flow from the source to the drift region depends on geometrical factors. The nonuniform magnetic topology complicates the well-known general analytical considerations for evaluating guiding-center drifts, and particle-in-cell simulations provide a self-consistent evaluation of the physics in an otherwise challenging scenario. Plasma flow profiles of a ferroelectric plasma source demonstrate that the densities required for longitudinal compression experiments involving ion beams are provided over the drift length, and are in good agreement with measurements. Simulations involving azimuthally asymmetric plasma creation conditions show that symmetric profiles are nevertheless achieved at the time of peak on-axis plasma density. Also, the ferroelectric plasma expands upstream on the thermal expansion time scale, and therefore avoids the possibility of penetration into the acceleration gap and transport sections, where partial neutralization would increase the beam emittance. Future experiments on NDCX will investigate the transverse focusing of an axially compressing intense charge bunch to a sub-mm spot size with coincident focal planes using a strong final-focus solenoid. In order to fill a multi-tesla solenoid with the necessary high-density plasma for beam charge neutralization, the simulations predict that supersonically injected plasma from the low-field region will penetrate and

Plasma experiments with relevance for other branches of science

International Nuclear Information System (INIS)

Sanduloviciu, M.; Lozneanu, E.

2000-01-01

A new scenario of self-organization, suggested by plasma experiments, is presented as an enlightening model able to illustrate, on some examples, the necessity of a paradigm shift in science. Thus, self-organization at criticality in fusion devices, differential negative resistance of semi-conductors, generation of complex space charge configurations under controllable laboratory conditions and in nature are mentioned as phenomena potentially explicable in the frame of a unique framework in which self-organization is the central concept. (authors)

Magnetic Configuration Control of ITER Plasmas

International Nuclear Information System (INIS)

Albanese, R.; Artaserse, G.; Mattei, M.; Ambrosino, G.; Crisanti, F.; Tommasi, G. de; Fresa, R.; Portone, A.; Sartori, F.; Villone, F.

2006-01-01

The aim of this paper is to review the capability of the ITER Poloidal Field (PF) system of controlling the broad range of plasma configurations presently forecasted during ITER operation. The attention is focused on the axi-symmetric aspects of plasma magnetic configuration control since they pose the greatest challenges in terms of control power and they have the largest impact on machine capital cost. The paper is broadly divided in two main sections devoted, respectively, to open loop (feed-forward) and closed loop (feedback) control. In the first part of the study the PF system is assessed with respect to the initiation, ramp-up, sustained burn, ramp-down phases of the main plasma inductive scenario. The limiter-to-divertor configuration transition phase is considered in detail with the aim of assessing the PF capability to form an X-point at the lowest possible current and, therefore, to relax the thermal load on the limiter surfaces. Moreover, during the sustained burn it is important to control plasmas with a broad range of current density profiles. In the second part of the study the plasma vertical feedback control requirements are assessed in details, in particular for the high elongation configurations achievable during the early limiter-to-X point transition phase. Non-rigid plasma displacement models are used to assess the control system voltage and current requirements of different radial field control circuits obtained, for example, by connecting the outermost PF coils, some CS coils, coils sub-sections etc. At last, the main 3D effects of the vessel ports are modeled and their impact of vertical stabilization evaluated. (author)

The COMPASS Tokamak Plasma Control Software Performance

Science.gov (United States)

Valcarcel, Daniel F.; Neto, André; Carvalho, Ivo S.; Carvalho, Bernardo B.; Fernandes, Horácio; Sousa, Jorge; Janky, Filip; Havlicek, Josef; Beno, Radek; Horacek, Jan; Hron, Martin; Panek, Radomir

2011-08-01

The COMPASS tokamak has began operation at the IPP Prague in December 2008. A new control system has been built using an ATCA-based real-time system developed at IST Lisbon. The control software is implemented on top of the MARTe real-time framework attaining control cycles as short as 50 μs, with a jitter of less than 1 μs. The controlled parameters, important for the plasma performance, are the plasma current, position of the plasma current center, boundary shape and horizontal and vertical velocities. These are divided in two control cycles: slow at 500 μs and fast at 50 μs. The project has two phases. First, the software implements a digital controller, similar to the analog one used during the COMPASS-D operation in Culham. In the slow cycle, the plasma current and position are measured and controlled with PID and feedforward controllers, respectively, the shaping magnetic field is preprogrammed. The vertical instability and horizontal equilibrium are controlled with the faster 50-μs cycle PID controllers. The second phase will implement a plasma-shape reconstruction algorithm and controller, aiming at optimized plasma performance. The system was designed to be as modular as possible by breaking the functional requirements of the control system into several independent and specialized modules. This splitting enabled tuning the execution of each system part and to use the modules in a variety of applications with different time constraints. This paper presents the design and overall performance of the COMPASS control software.

Remote automatic control scheme for plasma arc cutting of contaminated waste

International Nuclear Information System (INIS)

Dudar, A.M.; Ward, C.R.; Kriikku, E.M.

1993-01-01

Plasma arc cutting is a popular technique used for size reduction of radioactively contaminated metallic waste such as glove boxes, vessels, and ducts. It is a very aggressive process and is capable of cutting metal objects up to 3 in. thick. The crucial control criteria in plasma cutting is maintaining a open-quotes stand-offclose quotes distance between the plasma torch tip and the material being cut. Manual plasma cutting techniques in radioactive environments require the operator to wear a plastic suit covered by a metallic suit. This is very cumbersome, time-consuming, and also generates additional waste (plastic and metallic suits). Teleoperated remote cutting is preferable to manual cutting, but our experience has shown that remote control of the stand-off distance is particularly difficult because of the brightness of the plasma arc and inadequate viewing angles. Also, the heat generated by the torch causes the sheet metal to deform and warp during plasma cutting, creating a dynamically changing metal surface. The aforementioned factors make it extremely difficult, if not impossible, to perform plasma cuts of waste with a variety of shapes and sizes in a teleoperated fashion with an operator in the loop. Automating the process is clearly desirable

AWAKE, The Advanced Proton Driven Plasma Wakefield Acceleration Experiment at CERN

CERN Document Server

Gschwendtner, E.; Amorim, L.; Apsimon, R.; Assmann, R.; Bachmann, A.M.; Batsch, F.; Bauche, J.; Berglyd Olsen, V.K.; Bernardini, M.; Bingham, R.; Biskup, B.; Bohl, T.; Bracco, C.; Burrows, P.N.; Burt, G.; Buttenschon, B.; Butterworth, A.; Caldwell, A.; Cascella, M.; Chevallay, E.; Cipiccia, S.; Damerau, H.; Deacon, L.; Dirksen, P.; Doebert, S.; Dorda, U.; Farmer, J.; Fedosseev, V.; Feldbaumer, E.; Fiorito, R.; Fonseca, R.; Friebel, F.; Gorn, A.A.; Grulke, O.; Hansen, J.; Hessler, C.; Hofle, W.; Holloway, J.; Huther, M.; Jaroszynski, D.; Jensen, L.; Jolly, S.; Joulaei, A.; Kasim, M.; Keeble, F.; Li, Y.; Liu, S.; Lopes, N.; Lotov, K.V.; Mandry, S.; Martorelli, R.; Martyanov, M.; Mazzoni, S.; Mete, O.; Minakov, V.A.; Mitchell, J.; Moody, J.; Muggli, P.; Najmudin, Z.; Norreys, P.; Oz, E.; Pardons, A.; Pepitone, K.; Petrenko, A.; Plyushchev, G.; Pukhov, A.; Rieger, K.; Ruhl, H.; Salveter, F.; Savard, N.; Schmidt, J.; Seryi, A.; Shaposhnikova, E.; Sheng, Z.M.; Sherwood, P.; Silva, L.; Soby, L.; Sosedkin, A.P.; Spitsyn, R.I.; Trines, R.; Tuev, P.V.; Turner, M.; Verzilov, V.; Vieira, J.; Vincke, H.; Wei, Y.; Welsch, C.P.; Wing, M.; Xia, G.; Zhang, H.

2016-01-01

The Advanced Proton Driven Plasma Wakefield Acceleration Experiment (AWAKE) aims at studying plasma wakefield generation and electron acceleration driven by proton bunches. It is a proof-of-principle R&D experiment at CERN and the world's first proton driven plasma wakefield acceleration experiment. The AWAKE experiment will be installed in the former CNGS facility and uses the 400 GeV/c proton beam bunches from the SPS. The first experiments will focus on the self-modulation instability of the long (rms ~12 cm) proton bunch in the plasma. These experiments are planned for the end of 2016. Later, in 2017/2018, low energy (~15 MeV) electrons will be externally injected to sample the wakefields and be accelerated beyond 1 GeV. The main goals of the experiment will be summarized. A summary of the AWAKE design and construction status will be presented.

Plasma position control in TCABR Tokamak

International Nuclear Information System (INIS)

Galvao, R.M.O.; Kuznetsov, Yu. K.; Nascimento, I.C.; Fonseca, A.M.M.; Silva, R.P. da; Ruchko, L.F.; Tuszel, A.G.; Reis, A.P. dos; Sanada, E.K.

1998-01-01

The plasma control position in the TCABR tokamak is described. The TCA tokamak was transferred from the Centre de Recherches en Physique des Plasmas, Lausanne, to the Institute of Physics of University of Sao Paulo, renamed TCABR (α=0.18 m, R = 0.62 m, B = 1 T,I p = 100 kA). The control system was reconstructed using mainly components obtained from the TCA tokamak. A new method of plasma position determination is used in TCABR to improve its accuracy. A more detailed theoretical analysis of the feed forward and feedback control is performed as compared with. (author)

Theory for beam-plasma millimeter-wave radiation source experiments

International Nuclear Information System (INIS)

Rosenberg, M.; Krall, N.A.

1989-01-01

This paper reports on theoretical studies for millimeter-wave plasma source experiments. In the device, millimeter-wave radiation is generated in a plasma-filled waveguide driven by counter-streaming electron beams. The beams excite electron plasma waves which couple to produce radiation at twice the plasma frequency. Physics topics relevant to the high electron beam current regime are discussed

Numerical simulation and optimal control in plasma physics

International Nuclear Information System (INIS)

Blum, J.

1989-01-01

The topics covered in this book are: A free boundary problem: the axisymmetric equilibrium of the plasma in a Tokamak; Static control of the plasma boundary by external currents; Existence and control of a solution to the equilibrium problem in a simple case; Study of equilibrium solution branches and application to the stability of horizontal displacements; Identification of the plasma boundary and plasma current density from magnetic measurements; Evolution of the equilibrium at the diffusion time scale; Evolution of the equilibrium of a high aspect-ratio circular plasma; Stability and control of the horizontal displacement of the plasma

Worldwide collaborative efforts in plasma control software development

International Nuclear Information System (INIS)

Penaflor, B.G.; Ferron, J.R.; Walker, M.L.; Humphreys, D.A.; Leuer, J.A.; Piglowski, D.A.; Johnson, R.D.; Xiao, B.J.; Hahn, S.H.; Gates, D.A.

2008-01-01

This presentation will describe the DIII-D collaborations with various tokamak experiments throughout the world which have adapted custom versions of the DIII-D plasma control system (PCS) software for their own use. Originally developed by General Atomics for use on the DIII-D tokamak, the PCS has been successfully installed and used for the NSTX experiment in Princeton, the MAST experiment in Culham UK, the EAST experiment in China, and the Pegasus experiment in the University of Wisconsin. In addition to these sites, a version of the PCS is currently being developed for use by the KSTAR tokamak in Korea. A well-defined and robust PCS software infrastructure has been developed to provide a common foundation for implementing the real-time data acquisition and feedback control codes. The PCS infrastructure provides a flexible framework that has allowed the PCS to be easily adapted to fulfill the unique needs of each site. The software has also demonstrated great flexibility in allowing for different computing, data acquisition and real-time networking hardware to be used. A description of the current PCS software architecture will be given along with experiences in developing and supporting the various PCS installations throughout the world

Feedback control of plasma configuration in JT-60

International Nuclear Information System (INIS)

Ninomiya, Hiromasa; Kikuchi, Mitsuru; Yoshino, Ryuji; Hosogane, Nobuyuki; Kimura, Toyoaki; Kurihara, Kenichi; Takahashi, Minoru; Hayashi, Kazuo.

1986-08-01

Plasma current, plasma position (center of the outermost magnetic surface), decay index n index and width of the divertor throat are feedback controlled by using 5 kinds of poloidal field coils in JT-60. 5 control commands are calculated in a feedback control computer in each 1 msec. These feedback control functions are checked in ohmically heated plasma. The control characteristics of the plasma are well understood by the simplified control analysis and are consistent with the precise matrix transfer function analysis in the frequency domain and the simulation analysis which include the effects of eddy currents, delay time elements and mutual interactions between controllers. The usefulness of these analyses is experimentally confirmed. Each controlled variable is well feedback controlled to the command and the experimentally realized equilibrium configuration is checked by the well calibrated magnetic probes. Fast boundary identification code is used for the identification of the equilibrium and results are consistent with the precalculated plasma equilibria. By using this feedback control system of the plasma configuration and the equilibrium identification method, we have obtained the stable limiter and divertor configuration. The maximum parameters obtained during OH(I) experimental period are plasma current I p = 1.8 MA, the effective safety factor q eff e = 5.7 x 10 19 m -3 (Murakami parameter of 4.5) and the pulse length of 5 ∼ 10 sec. (author)

Pre-launch simulation experiment of microwave-ionosphere nonlinear interaction rocket experiment in the space plasma chamber

Energy Technology Data Exchange (ETDEWEB)

Kaya, N. (Kobe University, Kobe, Japan); Tsutsui, M. (Kyoto University, Uji, Japan); Matsumoto, H. (Kyoto University, Kyoto, Japan)

1980-09-01

A pre-flight test experiment of a microwave-ionosphere nonlinear interaction rocket experiment (MINIX) has been carried out in a space plasma simulation chamber. Though the first rocket experiment ended up in failure because of a high voltage trouble, interesting results are observed in the pre-flight experiment. A significant microwave heating of plasma up to 300% temperature increase is observed. Strong excitations of plasma waves by the transmitted microwaves in the VLF and HF range are observed as well. These microwave effects may have to be taken into account in solar power satellite projects in the future.

Using the Tritium Plasma Experiment to evaluate ITER PFC safety

International Nuclear Information System (INIS)

Longhurst, G.R.; Anderl, R.A.; Bartlit, J.R.; Causey, R.A.; Haines, J.R.

1993-01-01

The Tritium Plasma Experiment was assembled at Sandia National Labs., Livermore and is being moved to the Tritium Systems Test Assembly facility at Los Alamos National Lab. to investigate interactions between dense plasmas at low energies and plasma-facing component materials. This apparatus has the unique capability of replicating plasma conditions in a tokamak divertor with particle flux densities of 2 x 10 23 ions/m 2 .s and a plasma temperature of about 15 eV using a plasma that includes tritium. An experimental program has been initiated using the Tritium Plasma Experiment to examine safety issues related to tritium in plasma-facing components, particularly the ITER divertor. Those issues include tritium retention and release characteristics, tritium permeation rates and transient times to coolant streams, surface modification and erosion by the plasma, the effects of thermal loads and cycling, and particulate production. An industrial consortium led by McDonnell Douglas will design and fabricate the test fixtures

Plasma heating by relativistic electron beams: correlations between experiment and theory

International Nuclear Information System (INIS)

Thode, L.E.; Godfrey, B.B.

1975-01-01

The streaming instability is the primary heating mechanism in most, if not all, experiments in which the beam is injected into partially or fully ionized gas. In plasma heating experiments, the relativistic beam must traverse an anode foil before interacting with the plasma. The linear theory for such a scattered beam is discussed, including a criterion for the onset of the kinetic interaction. A nonlinear model of the two-stream instability for a scattered beam is developed. Using this model, data from ten experiments are unfolded to obtain the following correlations: (i) for a fixed anode foil, the dependence of the plasma heating on the beam-to-plasma density ratio is due to anode foil scattering, (ii) for a fixed beam-to-plasma density ratio, the predicted change in the magnitude of plasma heating as a function of the anode foil is in agreement with experiment, and (iii) the plasma heating tentatively appears to be proportional to the beam kinetic energy density and beam pulse length. For a fixed anode foil, theory also predicts that the energy deposition is improved by increasing the beam electron energy γmc 2 . Presently, no experiment has been performed to confirm this aspect of the theory

Plasma control for efficient extreme ultra-violet source

International Nuclear Information System (INIS)

Takahashi, Kensaku; Nakajima, Mitsuo; Kawamura, Tohru; Shiho, Makoto; Hotta, Eiki; Horioka, Kazuhiko

2008-01-01

To generate a high efficiency extreme-ultraviolet (EUV) source, effects of plasma shape for controlling radiative plasmas based on xenon capillary discharge are experimentally investigated. The radiation characteristics observed via tapered capillary discharge are compared with those of straight one. From the comparison, the long emission period and different plasma behaviors of tapered capillary discharge are confirmed. This means that control of the plasma geometry is effective for prolonging the EUV emission period. This result also indicates that the plasma shape control seems to have a potential for enhancing the conversion efficiency. (author)

A Burning Plasma Experiment: the role of international collaboration

Science.gov (United States)

Prager, Stewart

2003-04-01

The world effort to develop fusion energy is at the threshold of a new stage in its research: the investigation of burning plasmas. A burning plasma is self-heated. The 100 million degree temperature of the plasma is maintained by the heat generated by the fusion reactions themselves, as occurs in burning stars. The fusion-generated alpha particles produce new physical phenomena that are strongly coupled together as a nonlinear complex system, posing a major plasma physics challenge. Two attractive options are being considered by the US fusion community as burning plasma facilities: the international ITER experiment and the US-based FIRE experiment. ITER (the International Thermonuclear Experimental Reactor) is a large, power-plant scale facility. It was conceived and designed by a partnership of the European Union, Japan, the Soviet Union, and the United States. At the completion of the first engineering design in 1998, the US discontinued its participation. FIRE (the Fusion Ignition Research Experiment) is a smaller, domestic facility that is at an advanced pre-conceptual design stage. Each facility has different scientific, programmatic and political implications. Selecting the optimal path for burning plasma science is itself a challenge. Recently, the Fusion Energy Sciences Advisory Committee recommended a dual path strategy in which the US seek to rejoin ITER, but be prepared to move forward with FIRE if the ITER negotiations do not reach fruition by July, 2004. Either the ITER or FIRE experiment would reveal the behavior of burning plasmas, generate large amounts of fusion power, and be a huge step in establishing the potential of fusion energy to contribute to the world's energy security.

Status of 2XIIB plasma confinement experiments

International Nuclear Information System (INIS)

Coensgen, F.J.; Clauser, J.F.; Correll, D.L.

1976-01-01

This report describes the status of 2XIIB neutral beam injection experiments with stabilizing plasma. The stream suppresses ion-cyclotron fluctuations and permits density to 5 x 10 13 cm -3 . The ion energy is 13 keV, and electron temperature reaches 140 eV. Plasma confinement increases with ion energy and n tau reaches 7 x 10 10 cm -3 .s at 13 keV. The n tau energy scaling is consistent with electron drag and ion-ion scattering losses. Buildup on a streaming plasma in a steady-state magnetic field is described

LLNL large-area inductively coupled plasma (ICP) source: Experiments

International Nuclear Information System (INIS)

Richardson, R.A.; Egan, P.O.; Benjamin, R.D.

1995-05-01

We describe initial experiments with a large (76-cm diameter) plasma source chamber to explore the problems associated with large-area inductively coupled plasma (ICP) sources to produce high density plasmas useful for processing 400-mm semiconductor wafers. Our experiments typically use a 640-nun diameter planar ICP coil driven at 13.56 MHz. Plasma and system data are taken in Ar and N 2 over the pressure range 3-50 mtorr. RF inductive power was run up to 2000W, but typically data were taken over the range 100-1000W. Diagnostics include optical emission spectroscopy, Langmuir probes, and B probes as well as electrical circuit measurements. The B and E-M measurements are compared with models based on commercial E-M codes. Initial indications are that uniform plasmas suitable for 400-mm processing are attainable

Divertor plasma studies on DIII-D: Experiment and modeling

International Nuclear Information System (INIS)

West, W.P.; Brooks, N.H.; Allen, S.L.

1996-09-01

In a magnetically diverted tokamak, the scrape-off layer (SOL) and divertor plasma provides separation between the first wall and the core plasma, intercepting impurities generated at the wall before they reach the core plasma. The divertor plasma can also serve to spread the heat and particle flux over a large area of divertor structure wall using impurity radiation and neutral charge exchange, thus reducing peak heat and particle fluxes at the divertor strike plate. Such a reduction will be required in the next generation of tokamaks, for without it, the divertor engineering requirements are very demanding. To successfully demonstrate a radiative divertor, a highly radiative condition with significant volume recombination must be achieved in the divertor, while maintaining a low impurity content in the core plasma. Divertor plasma properties are determined by a complex interaction of classical parallel transport, anomalous perpendicular transport, impurity transport and radiation, and plasma wall interaction. In this paper the authors describe a set of experiments on DIII-D designed to provide detailed two dimensional documentation of the divertor and SOL plasma. Measurements have been made in operating modes where the plasma is attached to the divertor strike plate and in highly radiating cases where the plasma is detached from the divertor strike plate. They also discuss the results of experiments designed to influence the distribution of impurities in the plasma using enhanced SOL plasma flow. Extensive modeling efforts will be described which are successfully reproducing attached plasma conditions and are helping to elucidate the important plasma and atomic physics involved in the detachment process

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.)

Rethermalization of a field-reversed configuration plasma in translation experiments

International Nuclear Information System (INIS)

Himura, H.; Okada, S.; Sugimoto, S.; Goto, S.

1995-01-01

A translation experiment of field-reversed configuration (FRC) plasma is performed on the FIX machine [Shiokawa and Goto, Phys. Fluids B 5, 534 (1993)]. The translated FRC bounces between magnetic mirror fields at both ends of a confinement region. The plasma loses some of its axial kinetic energy when it is reflected by the magnetic mirror field, and eventually settles down in the confinement region. In this reflection process, the plasma temperature rises significantly. Such plasma rethermalization has been observed in OCT-L1 experiments [Ito et al., Phys. Fluids 30, 168 (1987)], but rarely in FRX-C/T experiments [Rej et al., Phys. Fluids 29, 852 (1986)]. It is found that the rethermalization depends on the relation between the plasma temperature and the translation velocity. The rethermalization occurs only in the case where the translation velocity exceeds the sound velocity. This result implies the rethermalization is caused by a shock wave induced within the FRC when the plasma is reflected by the magnetic mirror field. copyright 1995 American Institute of Physics

Magnetic sensorless control experiment without drift problem on HT-7

International Nuclear Information System (INIS)

Nakamura, K.; Luo, J.R.; Wang, H.Z.; Ji, Z.S.; Wang, H.; Wang, F.; Qi, N.; Sato, K.N.; Hanada, K.; Sakamoto, M.; Idei, H.; Hasegawa, M.; Iyomasa, A.; Kawasaki, S.; Nakashima, H.; Higashijima, A.

2006-01-01

Magnetic sensorless control experiments of the plasma horizontal position have been carried out in the superconducting tokamak HT-7. Previously the horizontal position was calculated from the vertical field coil current and voltage without using signals of magnetic sensors like magnetic coils and flux loops placed near the plasma. The calculations are made focusing on the ripple frequency component of the power supply with thyristor and directly from them without time integration. There is no drift problem of integrator of magnetic sensors. Two kinds of experiments were carried out, to keep the position constant and swing the position in a triangular waveform

Three Flow Features behind the Flow Control Authority of DBD Plasma Actuator: Result of High-Fidelity Simulations and the Related Experiments

Directory of Open Access Journals (Sweden)

Kozo Fujii

2018-04-01

Full Text Available Both computational and experimental studies are conducted for understanding of the flow separation control mechanism of a DBD (dielectric barrier discharge plasma actuator. Low speed flows over an airfoil are considered. A DBD plasma actuator is attached near the leading edge of an airfoil and the mechanism of flow control of this small device is discussed. The DBD plasma actuator, especially in burst mode, is shown to be very effective for controlling flow separation at Reynolds number of 6.3 × 104, when applied to the flows at an angle of attack higher than the stall. The analysis reveals that the flow structure includes three remarkable features that provide good authority for flow separation control with the appropriate actuator parameters. With proper setting of the actuator parameters to enhance the effective flow features for the application, good flow control can be achieved. Based on the analysis, guidelines for the effective use of DBD plasma actuators are proposed. A DBD plasma actuator is also applied to the flows under cruise conditions. With the DBD plasma actuator attached, a simple airfoil turns out to show higher lift-to-drag ratio than a well-designed airfoil.

First Laser-Plasma Interaction and Hohlraum Experiments on NIF

International Nuclear Information System (INIS)

Dewald, E L; Glenzer, S H; Landen, O L; Suter, L J; Jones, O S; Schein, J; Froula, D; Divol, L; Campbell, K; Schneider, M S; McDonald, J W; Niemann, C; Mackinnon, A J

2005-01-01

Recently the first hohlraum experiments have been performed at the National Ignition Facility (NIF) in support of indirect drive Inertial Confinement Fusion (ICF) designs. The effects of laser beam smoothing by spectral dispersion (SSD) and polarization smoothing (PS) on the beam propagation in long scale gas-filled pipes has been studied at plasma scales as found in indirect drive gas filled ignition hohlraum designs. The long scale gas-filled target experiments have shown propagation over 7 mm of dense plasma without filamentation and beam break up when using full laser smoothing. Vacuum hohlraums have been irradiated with laser powers up to 6 TW, 1-9 ns pulse lengths and energies up to 17 kJ to activate several diagnostics, to study the hohlraum radiation temperature scaling with the laser power and hohlraum size, and to make contact with hohlraum experiments performed at the NOVA and Omega laser facilities. Subsequently, novel long laser pulse hohlraum experiments have tested models of hohlraum plasma filling and long pulse hohlraum radiation production. The validity of the plasma filling assessment in analytical models and in LASNEX calculations has been proven for the first time. The comparison of these results with modeling will be discussed

A control approach for plasma density in tokamak machines

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

Novel aspects of plasma control in ITER

Energy Technology Data Exchange (ETDEWEB)

Humphreys, D.; Jackson, G.; Walker, M.; Welander, A. [General Atomics P.O. Box 85608, San Diego, California 92186-5608 (United States); Ambrosino, G.; Pironti, A. [CREATE/University of Naples Federico II, Napoli (Italy); Vries, P. de; Kim, S. H.; Snipes, J.; Winter, A.; Zabeo, L. [ITER Organization, St. Paul Lez durance Cedex (France); Felici, F. [Eindhoven University of Technology, Eindhoven (Netherlands); Kallenbach, A.; Raupp, G.; Treutterer, W. [Max-Planck Institut für Plasmaphysik, Garching (Germany); Kolemen, E. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543-0451 (United States); Lister, J.; Sauter, O. [Centre de Recherches en Physique des Plasmas, Ecole Polytechnique Federale de Lausanne, Lausanne (Switzerland); Moreau, D. [CEA, IRFM, 13108 St. Paul-lez Durance (France); Schuster, E. [Lehigh University, Bethlehem, Pennsylvania (United States)

2015-02-15

ITER plasma control design solutions and performance requirements are strongly driven by its nuclear mission, aggressive commissioning constraints, and limited number of operational discharges. In addition, high plasma energy content, heat fluxes, neutron fluxes, and very long pulse operation place novel demands on control performance in many areas ranging from plasma boundary and divertor regulation to plasma kinetics and stability control. Both commissioning and experimental operations schedules provide limited time for tuning of control algorithms relative to operating devices. Although many aspects of the control solutions required by ITER have been well-demonstrated in present devices and even designed satisfactorily for ITER application, many elements unique to ITER including various crucial integration issues are presently under development. We describe selected novel aspects of plasma control in ITER, identifying unique parts of the control problem and highlighting some key areas of research remaining. Novel control areas described include control physics understanding (e.g., current profile regulation, tearing mode (TM) suppression), control mathematics (e.g., algorithmic and simulation approaches to high confidence robust performance), and integration solutions (e.g., methods for management of highly subscribed control resources). We identify unique aspects of the ITER TM suppression scheme, which will pulse gyrotrons to drive current within a magnetic island, and turn the drive off following suppression in order to minimize use of auxiliary power and maximize fusion gain. The potential role of active current profile control and approaches to design in ITER are discussed. Issues and approaches to fault handling algorithms are described, along with novel aspects of actuator sharing in ITER.

Diagnostics for Pioneer I imploding plasma experiments

International Nuclear Information System (INIS)

Lee, P.H.Y.; Benjamin, R.F.; Brownell, J.H.

1985-01-01

The Pioneer I series of imploding plasma experiments are aimed at collapsing a thin aluminum foil with a multimegampere, submicrosecond electrical pulse produced by an explosive flux compression generator and fast plasma compression opening switch. Anticipated experimental conditions are bounded by implosion velocities of 2 x 10 7 cm/s and maximum plasma temperatures of 100 eV. A comprehensive array of diagnostics have been deployed to measure implosion symmetry (gated microchannel plate array and other time-resolved imaging), temperature of the imploding plasma (visible/uv spectroscopy), stagnation geometry (x-ray pinhole imaging), radiation emission characteristics at pinch (XRD's, fast bolometry), and electrical drive history (Rogowski loops, Faraday rotation current detectors, and capacitive voltage probes). Diagnostic performance is discussed and preliminary results are presented

Plasma equilibrium control during slow plasma current quench with avoidance of plasma-wall interaction in JT-60U

Science.gov (United States)

Yoshino, R.; Nakamura, Y.; Neyatani, Y.

1997-08-01

In JT-60U a vertical displacement event (VDE) is observed during slow plasma current quench (Ip quench) for a vertically elongated divertor plasma with a single null. The VDE is generated by an error in the feedback control of the vertical position of the plasma current centre (ZJ). It has been perfectly avoided by improving the accuracy of the ZJ measurement in real time. Furthermore, plasma-wall interaction has been avoided successfully during slow Ip quench owing to the good performance of the plasma equilibrium control system

Study of geometrical and operational parameters controlling the low frequency microjet atmospheric pressure plasma characteristics

International Nuclear Information System (INIS)

Kim, Dan Bee; Rhee, J. K.; Moon, S. Y.; Choe, W.

2006-01-01

Controllability of small size atmospheric pressure plasma generated at low frequency in a pin to dielectric plane electrode configuration was studied. It was shown that the plasma characteristics could be controlled by geometrical and operational parameters of the experiment. Under most circumstances, continuous glow discharges were observed, but both the corona and/or the dielectric barrier discharge characteristics were observed depending on the position of the pin electrode. The plasma size and the rotational temperature were also varied by the parameters. The rotational temperature was between 300 and 490 K, being low enough to treat thermally sensitive materials

Plasma boundary shape control and real-time equilibrium reconstruction on NSTX-U

Science.gov (United States)

Boyer, M. D.; Battaglia, D. J.; Mueller, D.; Eidietis, N.; Erickson, K.; Ferron, J.; Gates, D. A.; Gerhardt, S.; Johnson, R.; Kolemen, E.; Menard, J.; Myers, C. E.; Sabbagh, S. A.; Scotti, F.; Vail, P.

2018-03-01

The upgrade to the National Spherical Torus eXperiment (NSTX-U) included two main improvements: a larger center-stack, enabling higher toroidal field and longer pulse duration, and the addition of three new tangentially aimed neutral beam sources, which increase available heating and current drive, and allow for flexibility in shaping power, torque, current, and particle deposition profiles. To best use these new capabilities and meet the high-performance operational goals of NSTX-U, major upgrades to the NSTX-U control system (NCS) hardware and software have been made. Several control algorithms, including those used for real-time equilibrium reconstruction and shape control, have been upgraded to improve and extend plasma control capabilities. As part of the commissioning phase of first plasma operations, the shape control system was tuned to control the boundary in both inner-wall limited and diverted discharges. It has been used to accurately track the requested evolution of the boundary (including the size of the inner gap between the plasma and central solenoid, which is a challenge for the ST configuration), X-point locations, and strike point locations, enabling repeatable discharge evolutions for scenario development and diagnostic commissioning.

Experiments on screw-pinch plasmas with elongated cross section

International Nuclear Information System (INIS)

Lassing, H.W.

1989-01-01

In this thesis experiments are described carried out with SPICA II, a toroidal screw-pinch plasma device. this device is the last one in a series of plasma machines of the toroidal screw-pinch differing from its predecessor in its race-track shaped section. In devices of the type toroidal screw-pinch stable confinement is possible of plasmas with larger β values than in a tokamak discharge. In a pinch the plasma is screwed up, during the formation, in such a way that in a relatively small volume a plasma is formated with a high pressure. During the screwing up the plasma is heated by shock heating as well as adiabatic compression. With the modified snowplow model the density and temperature after the formation can be calculated, starting from the initial conditions. When all ions arrive into the plasma column, the density in the column is determined by the volume compression. First purpose of the experiments was to find a stable discharge. Subsequently discharges have been made with a high as possible β in order to investigate at which maximum β it is possible to confine screw-pinch plasmas stably. When these had been found, the nature and importance could be investigated of the processes following which the screw-pinch plasma looses its energy. (author), 75 res.; 95 figs.; 8 tabs

Review of D-T Experiments Relevant to Burning Plasma Issues

International Nuclear Information System (INIS)

Hawryluk, R.J.

2001-01-01

Progress in the performance of tokamak devices has enabled not only the production of significant bursts of fusion energy from deuterium-tritium (D-T) plasmas in the Tokamak Fusion Test Reactor (TFTR) and the Joint European Torus (JET) but, more importantly, the initial study of the physics of burning magnetically confined plasmas. The TFTR and JET, in conjunction with the worldwide fusion effort, have studied a broad range of topics including magnetohydrodynamic stability, transport, wave-particle interactions, the confinement of energetic particles, and plasma boundary interactions. The D-T experiments differ in three principal ways from previous experiments: isotope effects associated with the use of deuterium-tritium fuel, the presence of fusion-generated alpha particles, and technology issues associated with tritium handling and increased activation. The effect of deuterium-tritium fuel and the presence of alpha particles is reviewed and placed in the perspective of the much large r worldwide database using deuterium fuel and theoretical understanding. Both devices have contributed substantially to addressing the scientific and technical issues associated with burning plasmas. However, future burning plasma experiments will operate with larger ratios of alpha heating power to auxiliary power and will be able to access additional alpha-particle physics issues. The scientific opportunities for extending our understanding of burning plasmas beyond that provided by current experiments is described

Solar array experiments on the SPHINX satellite. [Space Plasma High voltage INteraction eXperiment satellite

Science.gov (United States)

Stevens, N. J.

1974-01-01

The Space Plasma, High Voltage Interaction Experiment (SPHINX) is the name given to an auxiliary payload satellite scheduled to be launched in January 1974. The principal experiments carried on this satellite are specifically designed to obtain the engineering data on the interaction of high voltage systems with the space plasma. The classes of experiments are solar array segments, insulators, insulators with pin holes and conductors. The satellite is also carrying experiments to obtain flight data on three new solar array configurations: the edge illuminated-multijunction cells, the teflon encased cells, and the violet cells.

Control of open end plasma flow utilizing orbital stochasticity

International Nuclear Information System (INIS)

Hojo, Hitoshi

1995-01-01

It has been known that the control of plasma outside the confinement region of diverter plasma and others in a magnetic field confinement device is very important for improveing the confinement of bulk plasma. The control of plasma outside a confinement region bears two roles, one is the reduction of the thermal load on a diverter plate and others due to the plasma particles lost from the confinement region, and another is the restriction of the back flow of cold plasma and impurities generated outside the confinement region to a bulk plasma region. In this study, the new method of controlling plasma outside a confinement region called magnetic diverter is considered. To the plasma particles advancing along magnetic force lines, the reflection and capture of the plasma particles occur in the region of orbital stochasticity, and the thermal load on an end plate and the reverse flow to a bulk plasma region are restricted. The numerical computation model used regarding the particle control utilizing the orbital stochasticity and the results of calculating the orbit of plasma particles in a magnetic field are reported. (K.I.)

Control of plasma position in the CASTOR tokamak

International Nuclear Information System (INIS)

Valovic, M.

1988-11-01

A simple servo-system designed for plasma position control in the CASTOR tokamak is described. Both radial and vertical plasma displacements were minimized using two servo-loops consisting of detection coils, a conventional electric controller and an amplifier operated as an unipolar voltage-controlled current source. To ensure the optimum conditions in the start-up phase of the discharge, currents in the servo-systems were externally preprogrammed. The prescribed plasma position was maintained with the accuracy of 3 mm. The feedback control improves plasma parameters, e.g. it removes the positional disruption at the end of the tokamak discharge. (J.U.). 4 figs., 3 refs

The first experiment of MPD Jet injection into GAMMA 10 plasma

International Nuclear Information System (INIS)

Ichimura, Kazuya; Nakashima, Yousuke; Takeda, Hisato

2014-01-01

Results of the first experiment of short pulse plasma injection by MPD (magneto plasma dynamic) Jet into GAMMA 10/PDX's longer pulse plasma are reported. In the experiment, a new method for plasma start-up without using plasma guns was applied. In this method, the main plasma of GAMMA 10/PDX was produced by ECRH (electron cyclotron resonance heating) and ICRF (ion cyclotron range of frequency). Then, MPD Jet plasma was injected into the main plasma along magnetic field line. As a result, density of the main plasma was increased and the end-loss flux was doubled. Flow velocity of the plasmoid injected by the MPD Jet was evaluated from the change of plasma density in each cell of the tandem mirror. The result indicated that the flow speed is several km/s. It is found that the plasmoid worked as strong fueling device which dramatically raises the density of plasma. Therefore injection of MPD Jet plasma into tandem mirror can be a useful tool to study physical phenomena of divertor and PWI. (author)

The Material Plasma Exposure eXperiment (MPEX)

Science.gov (United States)

Rapp, J.; Biewer, T. M.; Bigelow, T. S.; Canik, J.; Caughman, J. B. O.; Duckworth, R. C.; Goulding, R. H.; Hillis, D. L.; Lore, J. D.; Lumsdaine, A.; McGinnis, W. D.; Meitner, S. J.; Owen, L. W.; Shaw, G. C.; Luo, G.-N.

2014-10-01

Next generation plasma generators have to be able to access the plasma conditions expected on the divertor targets in ITER and future devices. The Material Plasma Exposure eXperiment (MPEX) will address this regime with electron temperatures of 1--10 eV and electron densities of 1021--1020 m-3. The resulting heat fluxes are about 10 MW/m2. MPEX is designed to deliver those plasma conditions with a novel Radio Frequency plasma source able to produce high density plasmas and heat electron and ions separately with Electron Bernstein Wave (EBW) heating and Ion Cyclotron Resonance Heating (ICRH). Preliminary modeling has been used for pre-design studies of MPEX. MPEX will be capable to expose neutron irradiated samples. In this concept targets will be irradiated in ORNL's High Flux Isotope Reactor (HFIR) or possibly at the Spallation Neutron Source (SNS) and then subsequently (after a sufficient long cool-down period) exposed to fusion reactor relevant plasmas in MPEX. The current state of the pre-design of MPEX including the concept of handling irradiated samples will be presented. ORNL is managed by UT-Battelle, LLC, for the U.S. DOE under Contract DE-AC-05-00OR22725.

Status of the new WEST plasma control system

International Nuclear Information System (INIS)

Ravene, Nathalie; Nouailletas, Rémy; Signoret, Jacqueline; Guillerminet, Bernard; Treutterrer, Wolfgang; Spring, Anett; Masand, Harish; Dhongde, Jasraj; Bhandarkar, Manisha; Rapson, Chris; Laqua, Heike; Lewerentz, Marc; Moreau, Philippe; Brémond, Sylvain; Allegretti, Ludovic; Raupp, Gerhard; Werner, Andreas; Laurent, François Saint; Nardon, Eric

2016-01-01

The WEST (W – for Tungsten – Environment in Steady state Tokamak) project is aiming at minimizing technology and operational risks of a full tungsten actively cooled divertor on ITER. It was started in 2013 and consists in transforming the Tore Supra tokamak into an X point divertor device, while taking advantage of its long discharge capability. To operate the next coming operations of WEST, new controllers are required. These developments are an opportunity to develop a new Plasma Control System (PCS) architecture featuring build-in real time handling of both plasma and plants events, thus addressing key ITER needs. The Tore Supra PCS will be refurbished including a new Pulse Schedule Editor (PSE). The main idea is to use a time segmented approach to describe the pulse schedule with a full integration of event handling both on PCS and PSE. Further to detailed requirement specifications and architecture design, two software tools were selected to define and execute a whole plasma discharge defined as a set of time segments. The PCS real-time framework (RTF) is based on an upgraded version of the AUG framework, called DCS (Discharge Control System). The PSE is the Xedit application used on WEGA and under further development for W7-X facility. This paper reports on the status of the new WEST PCS developments. The on-going developments to adapt DCS to the Tore Supra Control infrastructure networks (new real-time network, chronology system and pulse supervision) will be reported. The required preparations for the use of Xedit will be presented, mainly the appropriate formal description of the WEST control system and the implementation of the mapping between the Xedit experiment configuration and DCS configuration files.

Status of the new WEST plasma control system

Energy Technology Data Exchange (ETDEWEB)

Ravene, Nathalie, E-mail: nathalie.ravenel@gmail.com [IRFM, CEA, F-13108 Saint Paul lez Durance (France); Nouailletas, Rémy; Signoret, Jacqueline; Guillerminet, Bernard [IRFM, CEA, F-13108 Saint Paul lez Durance (France); Treutterrer, Wolfgang [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, 85748 Garching (Germany); Spring, Anett [Max-Planck-Institut für Plasmaphysik, Teilinstitut Greifswald, Wendelsteinstraße 1, D-17491 Greifswald (Germany); Masand, Harish; Dhongde, Jasraj; Bhandarkar, Manisha [Institute for Plasma Research (IPR), Near Indira Bridge, Bhat, Gandhinagar, 382 428 Gujarat (India); Rapson, Chris [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, 85748 Garching (Germany); Laqua, Heike; Lewerentz, Marc [Max-Planck-Institut für Plasmaphysik, Teilinstitut Greifswald, Wendelsteinstraße 1, D-17491 Greifswald (Germany); Moreau, Philippe; Brémond, Sylvain; Allegretti, Ludovic [IRFM, CEA, F-13108 Saint Paul lez Durance (France); Raupp, Gerhard [Max-Planck-Institut für Plasmaphysik, Boltzmannstraße 2, 85748 Garching (Germany); Werner, Andreas [Max-Planck-Institut für Plasmaphysik, Teilinstitut Greifswald, Wendelsteinstraße 1, D-17491 Greifswald (Germany); Laurent, François Saint; Nardon, Eric [IRFM, CEA, F-13108 Saint Paul lez Durance (France)

2016-11-15

The WEST (W – for Tungsten – Environment in Steady state Tokamak) project is aiming at minimizing technology and operational risks of a full tungsten actively cooled divertor on ITER. It was started in 2013 and consists in transforming the Tore Supra tokamak into an X point divertor device, while taking advantage of its long discharge capability. To operate the next coming operations of WEST, new controllers are required. These developments are an opportunity to develop a new Plasma Control System (PCS) architecture featuring build-in real time handling of both plasma and plants events, thus addressing key ITER needs. The Tore Supra PCS will be refurbished including a new Pulse Schedule Editor (PSE). The main idea is to use a time segmented approach to describe the pulse schedule with a full integration of event handling both on PCS and PSE. Further to detailed requirement specifications and architecture design, two software tools were selected to define and execute a whole plasma discharge defined as a set of time segments. The PCS real-time framework (RTF) is based on an upgraded version of the AUG framework, called DCS (Discharge Control System). The PSE is the Xedit application used on WEGA and under further development for W7-X facility. This paper reports on the status of the new WEST PCS developments. The on-going developments to adapt DCS to the Tore Supra Control infrastructure networks (new real-time network, chronology system and pulse supervision) will be reported. The required preparations for the use of Xedit will be presented, mainly the appropriate formal description of the WEST control system and the implementation of the mapping between the Xedit experiment configuration and DCS configuration files.

Tip Clearance Control Using Plasma Actuators

Science.gov (United States)

2007-03-01

Clearance Control Using Plasma Actuators 4 posed by Denton (1993). A number of investigators have used partial shrouds, or " winglet " designs to...SDBD actuator Plasma enhanced aerodynamics has been demonstrated in a range of applications involving sepa- ration control, lift enhancement, drag... aerodynamic benefits of a squealer tip geometry. Specifically, the squealer tip is known to reduce the discharge coefficient of the tip gap, thereby

Structure formation in turbulent plasmas - test of nonlinear processes in plasma experiments

International Nuclear Information System (INIS)

Itoh, S.-I.; Yagi, Masatoshi; Inagaki, Shigeru

2009-01-01

Full text: Recent developments in plasma physics, either in the fusion research in a new era of ITER, or in space and in astro-physics, the world-wide and focused research has been developed on the subject of structural formation in turbulent plasma being associated with electro-magnetic field formation. Keys for the progress were a change of the physics view from the 'linear, local and deterministic' picture to the description based on 'nonlinear instability, nonlocal interaction and probabilistic excitation' for the turbulent state, and the integration of the theory-simulation-experiment. In this presentation, we first briefly summarize the theory of microscopic turbulence and mesoscale fluctuations and selection rules. In addition, the statistical formation of large-scale structure/deformation by turbulence is addressed. Then, the experimental measurements of the mesoscale structures (e.g., zonal flows, zonal fields, streamer and transport interface) and of the nonlinear interactions among them in turbulent plasmas are reported. Confirmations by, and new challenges from, the experiments are overviewed. Work supported by the Grant-in-Aid for Specially-Promoted Research (16002005). (author)

Proceeding of JSPS-CAS Core University Program seminar on production and control of high performance plasmas with advanced plasma heating and diagnostic systems

International Nuclear Information System (INIS)

Gao Xiang; Morita, Shigeru

2011-02-01

The JSPS-CAS Core University Program (CUP) seminar on 'Production and control of high performance plasmas with advanced plasma heating and diagnostic systems' took place in Guilin Bravo Hotel, Guilin, China, 1-4 November 2010. This seminar was organized in the framework of CUP in the field of plasma and nuclear fusion. Two special talks and 46 oral talks were presented in the seminar including 36 Chinese, 18 Japanese and 4 Korean attendees. Production and control of high performance plasmas is a crucial issue for realizing an advanced nuclear fusion reactor in addition to developments of advanced plasma heating and diagnostics. This seminar was motivated along the issues. Results in the field of fusion experiments obtained through CUP activities during recent two years were summarized. Possible direction of future collaboration and further encouragement of scientific activity of younger scientists were also discussed in this seminar with future experimental plans in both countries. (author)

Tokamak-7 operation in experiments with a plasma

International Nuclear Information System (INIS)

Buzanki, V.V.; Bychkov, A.V.; Denisov, V.F.

1982-01-01

The results of experiments with plasma at the Tokamak-7 (T-7) device are presented. The experiments have been carried out with a constant diaphragm of 31,5 cm radius and two movable graphite diaphragms at the 26-28 cm plasma filament radius and 1,6-1,9 T magnetic field. Two stable regimes with 150 and 200 kA and 250 ms discharge current length have been investigated. It is shown that the strongest poloidal filed perturhations have been observed at the beginning of the discharge. Electron plasma temperature Tsub(e) has been determined from the spectrum analysis of soft X radiation by the foil method. Stable plasma regimes with current up to 200 kA, bypass voltage being equal 1,58V electron density -0,5-5,0 x 10 13 cm -3 , Tsub(e)=1,1-1,3 keV ion temperature-490 eV. The range between discharge pulses has reached 3 min. at the discharge current-240 kA. No considerable effect of magnetic field variables on the superconducting magnetic system has been observed

Power supply controlled for plasma torch generation

International Nuclear Information System (INIS)

Diaz Z, S.

1996-01-01

The high density of energy furnished by thermal plasma is profited in a wide range of applications, such as those related with welding fusion, spray coating and at the present in waste destruction. The waste destruction by plasma is a very attractive process because the remaining products are formed by inert glassy grains and non-toxic gases. The main characteristics of thermal plasmas are presented in this work. Techniques based on power electronics are utilized to achieve a good performance in thermal plasma generation. This work shown the design and construction of three phase control system for electric supply of thermal plasma torch, with 250 kw of capacity, as a part of the project named 'Destruction of hazard wastes by thermal plasma' actually working in the Instituto Nacional de Investigaciones Nucleares (ININ). The characteristics of thermal plasma and its generation are treated in the first chapter. The A C controllers by thyristors applied in three phase arrays are described in the chapter II, talking into account the power transformer, rectifiers bank and aliasing coil. The chapter III is dedicated in the design of the trigger module which controls the plasma current by varying the trigger angle of the SCR's; the protection and isolating unit are also presented in this chapter. The results and conclusions are discussed in chapter IV. (Author)

Reactive gas control of non-stable plasma conditions

International Nuclear Information System (INIS)

Bellido-Gonzalez, V.; Daniel, B.; Counsell, J.; Monaghan, D.

2006-01-01

Most industrial plasma processes are dependant upon the control of plasma properties for repeatable and reliable production. The speed of production and range of properties achieved depend on the degree of control. Process control involves all the aspects of the vacuum equipment, substrate preparation, plasma source condition, power supplies, process drift, valves (inputs/outputs), signal and data processing and the user's understanding and ability. In many cases, some of the processes which involve the manufacturing of interesting coating structures, require a precise control of the process in a reactive environment [S.J. Nadel, P. Greene, 'High rate sputtering technology for throughput and quality', International Glass Review, Issue 3, 2001, p. 45. ]. Commonly in these circumstances the plasma is not stable if all the inputs and outputs of the system were to remain constant. The ideal situation is to move a process from set-point A to B in zero time and maintain the monitored signal with a fluctuation equal to zero. In a 'real' process that's not possible but improvements in the time response and energy delivery could be achieved with an appropriate algorithm structure. In this paper an advanced multichannel reactive plasma gas control system is presented. The new controller offers both high-speed gas control combined with a very flexible control structure. The controller uses plasma emission monitoring, target voltage or any process sensor monitoring as the input into a high-speed control algorithm for gas input. The control algorithm and parameters can be tuned to different process requirements in order to optimize response times

Control of plasma density distribution via wireless power transfer in an inductively coupled plasma

International Nuclear Information System (INIS)

Lee, Hee-Jin; Lee, Hyo-Chang; Kim, Young-Cheol; Chung, Chin-Wook

2013-01-01

With an enlargement of the wafer size, development of large-area plasma sources and control of plasma density distribution are required. To control the spatial distribution of the plasma density, wireless power transfer is applied to an inductively coupled plasma for the first time. An inner powered antenna and an outer resonant coil connected to a variable capacitor are placed on the top of the chamber. As the self-resonance frequency ω r of the resonant coil is adjusted, the power transfer rate from the inner powered coil to the outer resonant coil is changed and the dramatic evolution of the plasma density profile is measured. As ω r of the outer resonant coil changes from the non-resonant condition (where ω r is not the driving angular frequency ω rf ) to the resonant condition (where ω r = ω rf ), the plasma density profile evolves from a convex shape with maximal plasma density at the radial center into a concave shape with maximal plasma density in the vicinity of the resonant antenna coil. This result shows that the plasma density distribution can be successfully controlled via wireless resonance power transfer. (fast track communication)

Advanced real-time control systems for magnetically confined fusion plasmas

International Nuclear Information System (INIS)

Goncalves, B.; Sousa, J.; Fernandes, H.; Rodrigues, A.P.; Carvalho, B.B.; Neto, A.; Varandas, C.A.F.

2008-01-01

Real-time control of magnetically confined plasmas is a critical issue for the safety, operation and high performance scientific exploitation of the experimental devices on regimes beyond the current operation frontiers. The number of parameters and the data volumes used for the plasma properties identification scale normally not only with the machine size but also with the technology improvements, leading to a great complexity of the plant system. A strong computational power and fast communication infrastructure are needed to handle in real-time this information, allowing just-in-time decisions to achieve the fusion critical plasma conditions. These advanced control systems require a tiered infrastructure including the hardware layer, the signal-processing middleware, real-time timing and data transport, the real-time operating system tools and drivers, the framework for code development, simulation, deployment and experiment parameterization and the human real-time plasma condition monitoring and management. This approach is being implemented at CFN by offering a vertical solution for the forthcoming challenges, including ITER, the first experimental fusion reactor. A given set of tools and systems are described on this paper, namely: (i) an ATCA based hardware multiple-input-multiple-output (MIMO) platform, PCI and PCIe acquisition and control modules; (ii) FPGA and DSP parallelized signal processing algorithms; (iii) a signal data and event distribution system over a 2.5/10Gb optical network with sub-microsecond latencies; (iv) RTAI and Linux drivers; and (v) the FireSignal, FusionTalk, SDAS FireCalc application tools. (author)

Electromagnetic Wave Transmittance Control using Anisotropic Plasma Lattice

Science.gov (United States)

Matlis, Eric; Corke, Thomas; Hoffman, Anthony

2017-11-01

Experiments of transmission through a lattice array of plasma columns have shown an absorption band close to the plasma frequency at 14 GHz. The beam was oriented at a 35° incident angle to the planar plasma cell. These experiments were designed to determine if the observed absorption was the result of the isotropic plasma medium or that of an anisotropic metamaterial. Transmission of the microwave energy was not consistent with an isotropic material in which absorption would monotonically increase below the plasma frequency. The experimental results are supported by an anisotropic model which was developed for the plasma permittivity using an effective medium approximation. The plasma columns were modeled as uniform rods with permittivity described by a Drude model while the components of the permittivity tensor was calculated using the Maxwell-Garnett effective medium theory. Electron densities of n = 4 x1012 cm-3 were assumed which is consistent with prior experimental measurements. This model confirms the existence of non-zero imaginary wave vector k in a narrow region centered about 14 GHz.

Experiments on CT plasma merging in the CTCC-1

International Nuclear Information System (INIS)

Watanabe, K.; Ikegami, K.; Nishikawa, M.; Ozaki, A.; Satomi, N.; Uyama, T.

1982-01-01

A compact toroid (CT) plasma merging experiment has been tried preliminarily in the CTCC-1 experiment as a method for further-heating of CT, on producing two CT plasmas in the flux conserver successively. Two CT plasmas were observed really to merge with each other in the flux conserver. In the merging process, it is found that the field reconnection of surface magnetic field lines of CT is completed until 30 μs after the second CT injection, but magnetic field lines around the center of CT merge slowly, taking about 100 μs. Experimental results indicate that merging of CT results in doubled addition of toroidal fluxes and no-addition of poloidal fluxes

Active feedback control of kink modes in tokamaks: 3D VALEN modeling and HBT-EP experiments

International Nuclear Information System (INIS)

Maurer, D.A.

2002-01-01

Significant progress in the development of active feedback control as a robust technique for the suppression of the wall stabilized external kink or resistive wall mode (RWM) in tokamaks has been achieved through a combination of modeling and experiments. Results from the 3D feedback modeling code VALEN, which serves as the primary analysis and feedback control design tool for RWM studies on the HBT-EP and DIII-D experiments, are in good agreement with observations. VALEN modeling of proposed advanced control system designs on HBT-EP, DIII-D, NSTX, and FIRE are predicted to approach the ideal wall beta limit in agreement with design principles based on simple single mode analytic theory of RWM feedback control. Benchmark experiments on HBT-EP have shown suppression of plasma disruption at rational edge q values using active feedback control in agreement with model predictions. In addition, the observation in HBT-EP of the plasma amplification of static resonant magnetic fields in plasmas marginally stable to the RWM is in agreement with theory. (author)

Local magnetic divertor for control of the plasma--limiter interaction in a tokamak

International Nuclear Information System (INIS)

Zweben, S.J.; Liewer, P.C.; Gould, R.W.

1984-01-01

An experiment is described in which plasma flow to a tokamak limiter is controlled through the use of a local toroidal divertor coil mounted inside the limiter itself. This coil produces a local perturbed field B/sub C/ approximately equal to the local unperturbed toroidal field B/sub T/approx. =3 kG, such that when B/sub C/ adds to B/sub T/ the field lines move into the limiter and the local plasma flow to it increases by a factor as great as 1.6, and when B/sub C/ subtracts from B/sub T/ the field lines move away from the limiter and the local plasma flow to it decreases by as much as a factor of 4. A simple theoretical model is used to interpret these results. Since these changes occur without significantly affecting global plasma confinement, such a control scheme may be useful for optimizing the performance of pumped limiters

High speed and high functional inverter power supplies for plasma generation and control, and their performance

International Nuclear Information System (INIS)

Uesugi, Yoshihiko; Razzak, Mohammad A.; Kondo, Kenji; Kikuchi, Yusuke; Takamura, Shuichi; Imai, Takahiro; Toyoda, Mitsuhiro

2003-01-01

The Rapid development of high power and high speed semiconductor switching devices has led to their various applications in related plasma fields. Especially, a high speed inverter power supply can be used as an RF power source instead of conventional linear amplifiers and a power supply to control the magnetic field in a fusion plasma device. In this paper, RF thermal plasma production and plasma heating experiments are described emphasis placed on using a static induction transistor inverter at a frequency range between 200 kHz and 2.5 MHz as an RF power supply. Efficient thermal plasma production is achieved experimentally by using a flexible and easily operated high power semiconductor inverter power supply. Insulated gate bipolar transistor (IGBT) inverter power supplies driven by a high speed digital signal processor are applied as tokamak joule coil and vertical coil power supplies to control plasma current waveform and plasma equilibrium. Output characteristics, such as the arbitrary bipolar waveform generation of a pulse width modulation (PWM) inverter using digital signal processor (DSP) can be successfully applied to tokamak power supplies for flexible plasma current operation and fast position control of a small tokamak. (author)

Beta II compact torus experiment plasma equilibrium and power balance

International Nuclear Information System (INIS)

Turner, W.C.; Goldenbaum, G.C.; Granneman, E.H.A.; Prono, D.S.; Hartman, C.W.; Taska, J.

1982-01-01

In this paper we follow up some of our earlier work that showed the compact torus (CT) plasma equilibrium produced by a magnetized coaxial plasma gun is nearly force free and that impurity radiation plays a dominant role in determining the decay time of plasma currents in present generation experiments

Fundamental investigations of capacitive radio frequency plasmas: simulations and experiments

International Nuclear Information System (INIS)

Donkó, Z; Derzsi, A; Hartmann, P; Korolov, I; Schulze, J; Czarnetzki, U; Schüngel, E

2012-01-01

Capacitive radio frequency (RF) discharge plasmas have been serving hi-tech industry (e.g. chip and solar cell manufacturing, realization of biocompatible surfaces) for several years. Nonetheless, their complex modes of operation are not fully understood and represent topics of high interest. The understanding of these phenomena is aided by modern diagnostic techniques and computer simulations. From the industrial point of view the control of ion properties is of particular interest; possibilities of independent control of the ion flux and the ion energy have been utilized via excitation of the discharges with multiple frequencies. ‘Classical’ dual-frequency (DF) discharges (where two significantly different driving frequencies are used), as well as discharges driven by a base frequency and its higher harmonic(s) have been analyzed thoroughly. It has been recognized that the second solution results in an electrically induced asymmetry (electrical asymmetry effect), which provides the basis for the control of the mean ion energy. This paper reviews recent advances on studies of the different electron heating mechanisms, on the possibilities of the separate control of ion energy and ion flux in DF discharges, on the effects of secondary electrons, as well as on the non-linear behavior (self-generated resonant current oscillations) of capacitive RF plasmas. The work is based on a synergistic approach of theoretical modeling, experiments and kinetic simulations based on the particle-in-cell approach. (paper)

Use of plasma waves to create in Tokamaks quasi-stationary conditions required for controlled fusion

International Nuclear Information System (INIS)

Moreau, D.

1993-04-01

In this thesis are studied the coupling of hybrid waves to the plasma, multijunction antennas, hybrid wave stochastic propagation, fast wave current drive and lower-hybrid current drive experiments in Tore Supra and Jet. The possibility of decoupling current density profile and temperature give one more degree of freedom for the control of plasma in a configuration which is not very flexible

Feedback control and stabilization experiments on the Texas Experimental Tokamak (TEXT)

International Nuclear Information System (INIS)

Uckan, T.; Carreras, B.A.; Richards, B.; Wootton, A.J.; Bengtson, R.D.; Bravenec, R.; Li, G.X.; Hurwitz, P.D.; Phillips, P.E.; Rowan, W.L.

1994-06-01

Plasma edge feedback experiments on the Texas Experimental Tokamak (TEXT) have been successful in controlling the edge plasma potential fluctuation level. The feedback wave-launcher, consisting of electrostatic probes located in the shadow of the limiter, is driven by the local edge potential fluctuations. In general, the edge potential fluctuations are modified in a broad frequency band. Moreover, it is observed that the potential fluctuations can be reduced (≤100 kHz) without enhancing other modes, or excited (10 to 12 kHz), depending on the phase difference between the driver and the launcher signal, and gain of the system. This turbulence modification is achieved not only locally but also halfway around the torus and has about 2 cm of poloidal extent. Experiments on the characterization of the global plasma parameters with the edge feedback are discussed. Effects of the edge feedback on the estimated fluctuation-induced radial particle flux as well as on the local plasma parameters are presented

Robust control design for the plasma horizontal position control on J-TEXT Tokamak

International Nuclear Information System (INIS)

Yu, W.Z.; Chen, Z.P.; Zhuang, G.; Wang, Z.J.

2013-01-01

It is extremely important for tokamak to control the plasma position during routine discharge. However, the model of plasma in tokamak usually contains much of the uncertainty, such as structured uncertainties and unmodeled dynamics. Compared with the traditional PID control approach, robust control theory is more suitable to handle this problem. In the paper, we propose a H ∞ robust control scheme to control the horizontal position of plasma during the flat-top phase of discharge on Joint Texas Experimental Tokamak (J-TEXT) tokamak. First, the model of our plant for plasma horizontal position control is obtained from the position equilibrium equations. Then the H ∞ robust control framework is used to synthesize the controller. Based on this, an H ∞ controller is designed to minimize the regulation/tracking error. Finally, a comparison study is conducted between the optimized H ∞ robust controller and the traditional PID controller in simulations. The simulation results of the H ∞ robust controller show a significant improvement of the performance with respect to those obtained with traditional PID controller, which is currently used on our machine

Stability and Control of Burning Tokamak Plasmas with Resistive Walls: Final Technical Report

Energy Technology Data Exchange (ETDEWEB)

Miller, George [Univ. of Tulsa, OK (United States); Brennan, Dylan [Princeton Univ., NJ (United States); Cole, Andrew [Columbia Univ., New York, NY (United States); Finn, John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-10-02

This project is focused on theoretical and computational development for quantitative prediction of the stability and control of the equilibrium state evolution in toroidal burning plasmas, including its interaction with the surrounding resistive wall. The stability of long pulse burning plasmas is highly sensitive to the physics of resonant layers in the plasma, sources of momentum and flow, kinetic effects of energetic particles, and boundary conditions at the wall, including feedback control and error fields. In ITER in particular, the low toroidal flow equilibrium state, sustained primarily by energetic alpha particles from fusion reactions, will require the consideration of all of these key elements to predict quantitatively the stability and evolution. The principal investigators on this project have performed theoretical and computational analyses, guided by analytic modeling, to address this physics in realistic configurations. The overall goal has been to understand the key physics mechanisms that describe stable toroidal burning plasmas under active feedback control. Several relevant achievements have occurred during this project, leading to publications and invited conference presentations. In theoretical efforts, with the physics of the resonant layers, resistive wall, and toroidal momentum transport included, this study has extended from cylindrical resistive plasma - resistive wall models with feedback control to toroidal geometry with strong shaping to study mode coupling effects on the stability. These results have given insight into combined tearing and resistive wall mode behavior in simulations and experiment, while enabling a rapid exploration of plasma parameter space, to identify possible domains of interest for large plasma codes to investigate in more detail. Resonant field amplification and quasilinear torques in the presence of error fields and velocity shear have also been investigated. Here it was found, surprisingly, that the Maxwell

Control of plasma layer in a fusion reactor correlated to DC motor control using PSO-ANFIS

International Nuclear Information System (INIS)

Mahapatra, Sakuntala; Daniel, Raju; Dey, Deep Narayan

2013-01-01

Plasma position and shape control is very crucial for the overall performance of the fusion reactor such as Tokamak. The quality of the discharge in the Saskatchewan TORus-Modified (STOR-M) tokamak is strongly related to the position of the plasma column within the discharge vessel. If the plasma column approaches too near the wall, then either minor or complete disruption occurs. Consequently it is necessary to be able to control dynamically the position of the plasma column throughout the entire discharge. Now a day's most fusion reactor employs the traditional PID controller for the confinement of plasma layer. Fuzzy logic is used for the control of Plasma layer. In this paper we have used the hybrid of PSO-ANFIS technique to control the speed of a DC motor. We have used two input parameters like speed, torque and output is firing angle. In our work first order Sugeno fuzzy model is taken with three rules and the parameters of Gaussian membership function is controlled by the PSO technique. PSO-ANFIS speed controller obtains better dynamic behavior and superior performance of the DC motor speed control. Similar approach can be correlated to the control of plasma layer. For the plasma control two inputs can be taken as plasma position ΔH and the plasma current and the single output, the control decision u(t). (author)

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)

Real-time Linux operating system for plasma control on FTU--implementation advantages and first experimental results

International Nuclear Information System (INIS)

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

2004-01-01

In this paper, we report on the experiment carried out at the Frascati Tokamak Upgrade (FTU) on the porting of the plasma control system (PCS) from a LynxOS architecture to an open source Linux real-time architecture. The old LynxOS system was implemented on a VME/PPC604r embedded controller guaranteeing successful plasma position, density and current control. The new RTAI-Linux operating system has shown to easily adapt to the VME hardware via a VME/INTELx86 embedded controller. The advantages of the new solution versus the old one are not limited to the reduced cost of the new architecture (based on the open-source characteristic of the RTAI architecture) but also enhanced by the response time of the real-time system which, also through an optimization of the real-time code, has been reduced from 150 μs (LynxOS) to 70 μs (RTAI). The new real-time operating system is also shown to be suitable for new extended control activities, whose implementation is also possible based on the reduced duty cycle duration, which leaves space for the real-time implementation of nonlinear control laws. We report here on recent experiments related to the optimization of the coupling between additional radiofrequency power and plasma

Real-time Linux operating system for plasma control on FTU--implementation advantages and first experimental results

Energy Technology Data Exchange (ETDEWEB)

Vitale, V. E-mail: vitale@frascati.enea.it; Centioli, C.; Iannone, F.; Mazza, G.; Panella, M.; Pangione, L.; Podda, S.; Zaccarian, L

2004-06-01

In this paper, we report on the experiment carried out at the Frascati Tokamak Upgrade (FTU) on the porting of the plasma control system (PCS) from a LynxOS architecture to an open source Linux real-time architecture. The old LynxOS system was implemented on a VME/PPC604r embedded controller guaranteeing successful plasma position, density and current control. The new RTAI-Linux operating system has shown to easily adapt to the VME hardware via a VME/INTELx86 embedded controller. The advantages of the new solution versus the old one are not limited to the reduced cost of the new architecture (based on the open-source characteristic of the RTAI architecture) but also enhanced by the response time of the real-time system which, also through an optimization of the real-time code, has been reduced from 150 {mu}s (LynxOS) to 70 {mu}s (RTAI). The new real-time operating system is also shown to be suitable for new extended control activities, whose implementation is also possible based on the reduced duty cycle duration, which leaves space for the real-time implementation of nonlinear control laws. We report here on recent experiments related to the optimization of the coupling between additional radiofrequency power and plasma.

Particle control and plasma performance in the Lithium Tokamak eXperiment

Energy Technology Data Exchange (ETDEWEB)

Majeski, R.; Abrams, T.; Boyle, D.; Granstedt, E.; Hare, J.; Jacobson, C. M.; Kaita, R.; Kozub, T.; LeBlanc, B.; Lundberg, D. P.; Lucia, M.; Merino, E.; Schmitt, J.; Stotler, D. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Biewer, T. M.; Canik, J. M.; Gray, T. K.; Maingi, R.; McLean, A. G. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Kubota, S. [University of California at Los Angeles, Los Angeles, California 90095 (United States); and others

2013-05-15

The Lithium Tokamak eXperiment is a small, low aspect ratio tokamak [Majeski et al., Nucl. Fusion 49, 055014 (2009)], which is fitted with a stainless steel-clad copper liner, conformal to the last closed flux surface. The liner can be heated to 350 °C. Several gas fueling systems, including supersonic gas injection and molecular cluster injection, have been studied and produce fueling efficiencies up to 35%. Discharges are strongly affected by wall conditioning. Discharges without lithium wall coatings are limited to plasma currents of order 10 kA, and discharge durations of order 5 ms. With solid lithium coatings discharge currents exceed 70 kA, and discharge durations exceed 30 ms. Heating the lithium wall coating, however, results in a prompt degradation of the discharge, at the melting point of lithium. These results suggest that the simplest approach to implementing liquid lithium walls in a tokamak—thin, evaporated, liquefied coatings of lithium—does not produce an adequately clean surface.

Laser light scatter experiments on plasma focus plant

International Nuclear Information System (INIS)

Wenzel, N.

1985-01-01

The plasma focus plant is an experiment on nuclear fusion, which is distinguished by a high neutron yield. Constituting an important method of diagnosis in plasma focussing, the laser light scatter method makes it possible, apart from finding the electron temperature and density, to determine the ion temperature resolved according to time and place and further, to study the occurrence of micro-turbulent effects. Starting from the theoretical basis, this dissertation describes light scatter measurements with ruby lasers on the POSEIDON plasma focus. They are given, together with earlier measurements on the Frascati 1 MJ plant and the Heidelberg 12 KJ plant. The development of the plasma parameters and the occurrence of superthermal light scatter events are discussed in connection with the dynamics of the plasma and the neutron emission characteristics of the individual plants. The results support the view that the thermo-nuclear neutron production at the plasma focus is negligible. Although the importance of micro-turbulent mechanisms in producing neutrons cannot be finally judged, important guidelines are given for the spatial and time relationships with plasma dynamics, plasma parameters and neutron emission. The work concludes with a comparison of all light scatter measurements at the plasma focus described in the literature. (orig.) [de

Stability and control of resistive wall modes in high beta, low rotation DIII-D plasmas

International Nuclear Information System (INIS)

Garofalo, A.M.; Jackson, G.L.; Haye, R.J. La; Okabayashi, M.; Reimerdes, H.; Strait, E.J.; Ferron, J.R.; Groebner, R.J.; In, Y.; Lanctot, M.J.; Matsunaga, G.; Navratil, G.A.; Solomon, W.M.; Takahashi, H.; Takechi, M.; Turnbull, A.D.

2007-01-01

Recent high-β DIII-D (Luxon J.L. 2002 Nucl. Fusion 42 64) experiments with the new capability of balanced neutral beam injection show that the resistive wall mode (RWM) remains stable when the plasma rotation is lowered to a fraction of a per cent of the Alfven frequency by reducing the injection of angular momentum in discharges with minimized magnetic field errors. Previous DIII-D experiments yielded a high plasma rotation threshold (of order a few per cent of the Alfven frequency) for RWM stabilization when resonant magnetic braking was applied to lower the plasma rotation. We propose that the previously observed rotation threshold can be explained as the entrance into a forbidden band of rotation that results from torque balance including the resonant field amplification by the stable RWM. Resonant braking can also occur naturally in a plasma subject to magnetic instabilities with a zero frequency component, such as edge localized modes. In DIII-D, robust RWM stabilization can be achieved using simultaneous feedback control of the two sets of non-axisymmetric coils. Slow feedback control of the external coils is used for dynamic error field correction; fast feedback control of the internal non-axisymmetric coils provides RWM stabilization during transient periods of low rotation. This method of active control of the n = 1 RWM has opened access to new regimes of high performance in DIII-D. Very high plasma pressure combined with elevated q min for high bootstrap current fraction, and internal transport barriers for high energy confinement, are sustained for almost 2 s, or 10 energy confinement times, suggesting a possible path to high fusion performance, steady-state tokamak scenarios

Plasma confinement of Nagoya high-beta toroidal-pinch experiments

International Nuclear Information System (INIS)

Hirano, K.; Kitagawa, S.; Wakatani, M.; Kita, Y.; Yamada, S.; Yamaguchi, S.; Sato, K.; Aizawa, T.; Osanai, Y.; Noda, N.

1977-01-01

Two different types of high-β toroidal pinch experiments, STP [1] and CCT [2,3], have been done to study the confinement of the plasma produced by a theta-pinch. The STP is an axisymmetric toroidal pinch of high-β tokamak type, while the CCT consists of multiply connected periodic toroidal traps. Internal current-carrying copper rings are essential to the CCT. Since both apparatuses use the same fast capacitor bank system, they produce rather similar plasma temperatures and densities. The observed laser scattering temperature and density is about 50 eV and 4x10 15 cm -3 , respectively, when the filling pressure is 5 mtorr. In the STP experiment, strong correlations are found between the βsub(p) value and the amplitude of m=2 mode. It has a minimum around the value of βsub(p) of 0.8. The disruptive instability is observed to expand the pinched plasma column without lowering the plasma temperature. Just before the disruption begins, the q value around the magnetic axis becomes far less than 1 and an increase of the amplitude of m=2 mode is seen. The CCT also shows rapid plasma expansion just before the magnetic field reaches its maximum. Then the trap is filled up with the plasma by this irreversible expansion and stable plasma confinement is achieved. The energy confinement time of the CCT is found to be about 35 μs. (author)

Initial measurements of two- and three-dimensional ordering, waves, and plasma filamentation in the Magnetized Dusty Plasma Experiment

Energy Technology Data Exchange (ETDEWEB)

Thomas, Edward, E-mail: etjr@auburn.edu; Konopka, Uwe [Physics Department, Auburn University, Auburn, Alabama 36849 (United States); Merlino, Robert L. [Department of Physics and Astronomy, The University of Iowa, Iowa City, Iowa 52242 (United States); Rosenberg, Marlene [Department of Electrical and Computer Engineering, University of California, San Diego, La Jolla, California 92093 (United States)

2016-05-15

The Magnetized Dusty Plasma Experiment at Auburn University has been operational for over one year. In that time, a number of experiments have been performed at magnetic fields up to B = 2.5 T to explore the interaction between magnetized plasmas and charged, micron-sized dust particles. This paper reports on the initial results from studies of: (a) the formation of imposed, ordered structures, (b) the properties of dust wave waves in a rotating frame, and (c) the generation of plasma filaments.

A note on supersonic flow control with nanosecond plasma actuator

Science.gov (United States)

Zheng, J. G.; Cui, Y. D.; Li, J.; Khoo, B. C.

2018-04-01

A concept study on supersonic flow control using nanosecond pulsed plasma actuator is conducted by means of numerical simulation. The nanosecond plasma discharge is characterized by the generation of a micro-shock wave in ambient air and a residual heat in the discharge volume arising from the rapid heating of near-surface gas by the quick discharge. The residual heat has been found to be essential for the flow separation control over aerodynamic bodies like airfoil and backward-facing step. In this study, novel experiment is designed to utilize the other flow feature from discharge, i.e., instant shock wave, to control supersonic flow through shock-shock interaction. Both bow shock in front of a blunt body and attached shock anchored at the tip of supersonic projectile are manipulated via the discharged-induced shock wave in an appropriate manner. It is observed that drag on the blunt body is reduced appreciably. Meanwhile, a lateral force on sharp-edged projectile is produced, which can steer the body and give it an effective angle of attack. This opens a promising possibility for extending the applicability of this flow control technique in supersonic flow regime.

Shear optimization experiments with current profile control on JET

International Nuclear Information System (INIS)

1997-01-01

A record performance on JET has been obtained with shear optimization scenarios. A neutron yield of 5.6x10 16 s -1 in deuterium discharges, and a global energy confinement improvement above the ITER-89 L-mode scaling with H ≤ 2.5 in L-mode and H ≤ 3 in H-mode have been achieved. The tailoring of plasma current, density and heating power waveforms and current profile control with lower hybrid current drive and ICRF phasing have been essential. Internal energy, particle and momentum transport barriers develop spontaneously upon heating above a threshold power of about 15 MW with neutral beams and ICRH into a low-density target plasma, with a wide central region of slightly negative or flat magnetic shear with q > 1 everywhere. An additional H-mode transition can also raise the pressure in the region between internal and edge transport barriers. The ion heat conductivity falls to the neoclassical level in the improved core confinement region. Pressure profile control through power deposition feedback control makes it possible to work close to the marginal stability boundary for pressure-driven MHD modes. First experiments in deuterium/tritium plasmas, with up to 75% tritium target concentration, have established internal transport barriers already with heating powers at the lowest threshold of pure deuterium plasmas, resulting in a fusion power output of P fusion = 2 MW. (author)

Edge localized modes control: experiment and theory

International Nuclear Information System (INIS)

Becoulet, M.; Huysmans, G.; Thomas, P.; Joffrin, E.; Rimini, F.; Monier-Garbet, P.; Grosman, A.; Ghendrih, P.; Parail, V.; Lomas, P.; Matthews, G.; Wilson, H.; Gryaznevich, M.; Counsell, G.; Loarte, A.; Saibene, G.; Sartori, R.; Leonard, A.; Snyder, P.; Evans, T.; Gohil, P.; Moyer, R.; Kamada, Y.; Oyama, N.; Hatae, T.; Kamiya, K.; Degeling, A.; Martin, Y.; Lister, J.; Rapp, J.; Perez, C.; Lang, P.; Chankin, A; Eich, T.; Sips, A.; Stober, J.; Horton, L.; Kallenbach, A.; Suttrop, W.; Saarelma, S.; Cowley, S.; Loennroth, J.; Shimada, M.; Polevoi, A.; Federici, G.

2005-01-01

The paper reviews recent theoretical and experimental results focussing on the identification of the key factors controlling ELM energy and particle losses both in natural ELMs and in the presence of external controlling mechanisms. Present experiment and theory pointed out the benefit of the high plasma shaping, high q 95 and high pedestal density in reducing the ELM affected area and conductive energy losses in Type I ELMs. Small benign ELMs regimes in present machines (EDA, HRS, Type II, Grassy, QH, Type III in impurity seeded discharges at high δ ) and their relevance for ITER are reviewed. Recent studies of active control of ELMs using stochastic boundaries, small pellets and edge current generation are presented

Confinement projections for the Burning Plasma Experiment (BPX)

International Nuclear Information System (INIS)

Goldston, R.J.; Bateman, G.; Kaye, S.M.; Perkins, F.W.; Pomphrey, N.; Stotler, D.P.; Zarnstorff, M.C.; Porkolab, M.; Reidel, K.S.; Stambaugh, R.D.; Waltz, R.E.

1991-01-01

The mission of the Burning Plasma Experiment (BPX, formerly CIT) is to study the physics of self-heated fusion plasmas (Q = 5 to ignition), and to demonstrate the production of substantial amounts of fusion power (P fus = 100 to 500 MW). Confinement projections for BPX have been made on the basis of (1) dimensional extrapolation (2) theory-based modeling calibrated to experiment, and (3) statistical scaling from the available empirical data base. The results of all three approaches, discussed in this paper, roughly coincide. We presently view the third approach, statistical scaling, as the most reliable means for projecting the confinement performance of BPX, and especially for assessing the uncertainty in the projection. 11 refs., 2 figs., 1 tab

Characteristics of pulsed plasma synthetic jet and its control effect on supersonic flow

Directory of Open Access Journals (Sweden)

Di Jin

2015-02-01

Full Text Available The plasma synthetic jet is a novel flow control approach which is currently being studied. In this paper its characteristic and control effect on supersonic flow is investigated both experimentally and numerically. In the experiment, the formation of plasma synthetic jet and its propagation velocity in quiescent air are recorded and calculated with time resolved schlieren method. The jet velocity is up to 100 m/s and no remarkable difference has been found after changing discharge parameters. When applied in Mach 2 supersonic flow, an obvious shockwave can be observed. In the modeling of electrical heating, the arc domain is not defined as an initial condition with fixed temperature or pressure, but a source term with time-varying input power density, which is expected to better describe the influence of heating process. Velocity variation with different heating efficiencies is presented and discussed and a peak velocity of 850 m/s is achieved in still air with heating power density of 5.0 × 1012 W/m3. For more details on the interaction between plasma synthetic jet and supersonic flow, the plasma synthetic jet induced shockwave and the disturbances in the boundary layer are numerically researched. All the results have demonstrated the control authority of plasma synthetic jet onto supersonic flow.

MTX [Microwave Tokamak Experiment] diagnostic and auxiliary systems for confinement, transport, and plasma physics studies

International Nuclear Information System (INIS)

Hooper, E.B.; Allen, S.L.; Casper, T.A.; Thomassen, K.I.

1989-01-01

This note describes the diagnostics and auxiliary systems on the Microwave Tokamak Experiment (MTX) for confinement, transport, and other plasma physics studies. It is intended as a reference on the installed and planned hardware on the machine for those who need more familiarity with this equipment. Combined with the tokamak itself, these systems define the opportunities and capabilities for experiments in the MTX facility. We also illustrate how these instruments and equipment are to be used in carrying out the MTX Operations Plan. Near term goals for MTX are focussed on the absorption and heating by the microwave beam from the FEL, but the Plan also includes using the facility to study fundamental phenomena in the plasma, to control MHD activity, and to drive current noninductively

Microsecond plasma opening switch experiments on GIT-4

Energy Technology Data Exchange (ETDEWEB)

Bystritskij, V M; Lisitsyn, I V; Sinebryukhov, A A; Sinebryukhov, V A [Russian Academy of Sciences, Tomsk (Russian Federation). Inst. of Electrophysics; Kim, A A; Kokshenev, V A; Koval` chuk, B M [Russian Academy of Sciences, Tomsk (Russian Federation). High Current Electronics Inst.

1997-12-31

The plasma opening switch (POS) operation at the current level up to 2 MA was studied at the terawatt power GIT-4 generator. The experiments are described in which the electrode diameter and the strength of the applied magnetic field were varied, and different plasma sources were used. It is shown that the high voltage / low impedance switch operation can be achieved if the linear current density at the POS cathode does not exceed 20 kA/cm. This value limits the maximum cathode diameter of the magnetically insulated transmission line. The anode diameter is limited by the requirement of no gap closure with a dense electrode plasma. The application of external magnetic field decreases the plasma density necessary for achieving a long POS conduction time operation regime. (J.U.). 1 tab., 4 refs.

Microsecond plasma opening switch experiments on GIT-4

International Nuclear Information System (INIS)

Bystritskij, V.M.; Lisitsyn, I.V.; Sinebryukhov, A.A.; Sinebryukhov, V.A.; Kim, A.A.; Kokshenev, V.A.; Koval'chuk, B.M.

1996-01-01

The plasma opening switch (POS) operation at the current level up to 2 MA was studied at the terawatt power GIT-4 generator. The experiments are described in which the electrode diameter and the strength of the applied magnetic field were varied, and different plasma sources were used. It is shown that the high voltage / low impedance switch operation can be achieved if the linear current density at the POS cathode does not exceed 20 kA/cm. This value limits the maximum cathode diameter of the magnetically insulated transmission line. The anode diameter is limited by the requirement of no gap closure with a dense electrode plasma. The application of external magnetic field decreases the plasma density necessary for achieving a long POS conduction time operation regime. (J.U.). 1 tab., 4 refs

Nonlinear plasma experiments in geospace with gigawatts of RF power at HAARP

Energy Technology Data Exchange (ETDEWEB)

Sheerin, J. P., E-mail: jsheerin@emich.edu [Physics and Astronomy, Eastern Michigan Univ., Ypsilanti, MI 48197 (United States); Cohen, Morris B., E-mail: mcohen@gatech.edu [Electrical and Computer Engineering, Georgia Tech, Atlanta, GA 30332-0250 (United States)

2015-12-10

The ionosphere is the ionized uppermost layer of our atmosphere (from 70 – 500 km altitude) where free electron densities yield peak critical frequencies in the HF (3 – 30 MHz) range. The ionosphere thus provides a quiescent plasma target, stable on timescales of minutes, for a whole host of active plasma experiments. High power RF experiments on ionospheric plasma conducted in the U.S. have been reported since 1970. The largest HF transmitter built to date is the HAARP phased-array HF transmitter near Gakona, Alaska which can deliver up to 3.6 Gigawatts (ERP) of CW RF power in the range of 2.8 – 10 MHz to the ionosphere with microsecond pointing, power modulation, and frequency agility. With an ionospheric background thermal energy in the range of only 0.1 eV, this amount of power gives access to the highest regimes of the nonlinearity (RF intensity to thermal pressure) ratio. HAARP’s unique features have enabled the conduct of a number of unique nonlinear plasma experiments in the interaction region of overdense ionospheric plasma including generation of artificial aurorae, artificial ionization layers, VLF wave-particle interactions in the magnetosphere, parametric instabilities, stimulated electromagnetic emissions (SEE), strong Langmuir turbulence (SLT) and suprathermal electron acceleration. Diagnostics include the Modular UHF Ionospheric Radar (MUIR) sited at HAARP, the SuperDARN-Kodiak HF radar, spacecraft radio beacons, HF receivers to record stimulated electromagnetic emissions (SEE) and telescopes and cameras for optical emissions. We report on short timescale ponderomotive overshoot effects, artificial field-aligned irregularities (AFAI), the aspect angle dependence of the intensity of the HF-enhanced plasma line, and production of suprathermal electrons. One of the primary missions of HAARP, has been the generation of ELF (300 – 3000 Hz) and VLF (3 – 30 kHz) radio waves which are guided to global distances in the Earth

Computations in plasma physics

International Nuclear Information System (INIS)

Cohen, B.I.; Killeen, J.

1984-01-01

A review of computer application in plasma physics is presented. Computer contribution to the investigation of magnetic and inertial confinement of a plasma and charged particle beam propagation is described. Typical utilization of computer for simulation and control of laboratory and cosmic experiments with a plasma and for data accumulation in these experiments is considered. Basic computational methods applied in plasma physics are discussed. Future trends of computer utilization in plasma reseaches are considered in terms of an increasing role of microprocessors and high-speed data plotters and the necessity of more powerful computer application

Static and dynamic control of plasma equilibrium in a Tokamak

International Nuclear Information System (INIS)

Blum, J.; Dei Cas, R.

1979-01-01

We are dealing here with the problem of controlling the plasma boundary and its displacements. Static control consists in determining the currents in the external coils of the Tokamak so that the plasma boundary has certain fixed characteristics: radial position, vertical elongation, desired shape. A self-consistent method is proposed here, considering a free plasma boundary, and using the techniques of optimal control of distributed parameter systems to solve the problem. The dynamic control problem considered in the second part of the paper is the control of the plasma radial displacements. An elaborate system of preprogramming and feedback control has been developed to ensure equilibrium and stability of the horizontal plasma motions. Optimal control techniques have been used to calculate the optimal primary coils configuration, the preprogramming voltages and the feedback gains. A new stability diagrams has been obtained which takes into account the erosion of the plasma by the limiter. All these calculations have been applied successfully to TFR 600 where thin liner and the presence of an iron core make the problem of stabilization of the radial displacements very difficult

Magnetic configuration control of ITER plasmas

International Nuclear Information System (INIS)

Albanese, R.; Mattei, M.; Portone, A.; Ambrosino, G.; Artaserse, G.; Crisanti, F.; De Tommasi, G.; Fresa, R.; Sartori, F.; Villone, F.

2007-01-01

The aim of this paper is to present some new tools used to review the capability of the ITER Poloidal Field (PF) system in controlling the broad range of plasma configurations presently forecasted during ITER operation. The attention is focused on the axi-symmetric aspects of plasma magnetic configuration control since they pose the greatest challenges in terms of control power and they have the largest impact on machine capital cost. Some preliminary results obtained during ongoing activities in collaboration between ENEA/CREATE and EFDA are presented. The paper is divided in two main parts devoted, respectively, to the presentation of a procedure for the PF current optimisation during the scenario, and of a software environment for the study of the PF system capabilities using the plasma linearized response. The proposed PF current optimisation procedure is then used to assess Scenario 2 design, also taking into account the presence of axisymmetric eddy currents and possible variations of poloidal beta and internal inductance. The numerical linear model based tool derived from the JET oriented eXtreme Shape Controller (XSC) tools is finally used to obtain results on the strike point sweeping in ITER

Control of ROS and RNS productions in liquid in atmospheric pressure plasma-jet system

Science.gov (United States)

Uchida, Giichiro; Ito, Taiki; Takenaka, Kosuke; Ikeda, Junichiro; Setsuhara, Yuichi

2016-09-01

Non-thermal plasma jets are of current interest in biomedical applications such as wound disinfection and even treatment of cancer tumors. Beneficial therapeutic effects in medical applications are attributed to excited species of oxygen and nitrogen from air. However, to control the production of these species in the plasma jet is difficult because their production is strongly dependent on concentration of nitrogen and oxygen from ambient air into the plasma jet. In this study, we analyze the discharge characteristics and the ROS and RNS productions in liquid in low- and high-frequency plasma-jet systems. Our experiments demonstrated the marked effects of surrounding gas near the plasma jet on ROS and RNS productions in liquid. By controlling the surround gas, the O2 and N2 main plasma jets are selectively produced even in open air. We also show that the concentration ratio of NO2- to H2O2 in liquid is precisely tuned from 0 to 0.18 in deionized water by changing N2 gas ratio (N2 / (N2 +O2)) in the main discharge gas, where high NO2- ratio is obtained at N2 gas ratio at N2 / (N2 +O2) = 0 . 8 . The low-frequency plasma jet with controlled surrounding gas is an effective plasma source for ROS and RNS productions in liquid, and can be a useful tool for biomedical applications. This study was partly supported by a Grant-in-Aid for Scientific Research on Innovative Areas ``Plasma Medical Innovation'' (24108003) from the Ministry of Education, Culture, Sports, Science and Technology, Japan (MEXT).

Full Tokamak discharge simulation and kinetic plasma profile control for ITER

International Nuclear Information System (INIS)

Hee Kim, S.

2009-10-01

Understanding non-linearly coupled physics between plasma transport and free-boundary equilibrium evolution is essential to operating future tokamak devices, such as ITER and DEMO, in the advanced tokamak operation regimes. To study the non-linearly coupled physics, we need a simulation tool which can self-consistently calculate all the main plasma physics, taking the operational constraints into account. As the main part of this thesis work, we have developed a full tokamak discharge simulator by combining a non-linear free-boundary plasma equilibrium evolution code, DINA-CH, and an advanced transport modelling code, CRONOS. This tokamak discharge simulator has been used to study the feasibility of ITER operation scenarios and several specific issues related to ITER operation. In parallel, DINA-CH has been used to study free-boundary physics questions, such as the magnetic triggering of edge localized modes (ELMs) and plasma dynamic response to disturbances. One of the very challenging tasks in ITER, the active control of kinetic plasma profiles, has also been studied. In the part devoted to free-boundary tokamak discharge simulations, we have studied dynamic responses of the free-boundary plasma equilibrium to either external voltage perturbations or internal plasma disturbances using DINA-CH. Firstly, the opposite plasma behaviour observed in the magnetic triggering of ELMs between TCV and ASDEX Upgrade has been investigated. Both plasmas experience similar local flux surface expansions near the upper G-coil set and passive stabilization loop (PSL) when the ELMs are triggered, due to the presence of the PSLs located inside the vacuum vessel of ASDEX Upgrade. Secondly, plasma dynamic responses to strong disturbances anticipated in ITER are examined to study the capability of the feedback control system in rejecting the disturbances. Specified uncontrolled ELMs were controllable with the feedback control systems. However, the specifications for fast H-L mode

A Toroidally Symmetric Plasma Simulation code for design of position and shape control on tokamak plasmas

International Nuclear Information System (INIS)

Takase, Haruhiko; Senda, Ikuo

1999-01-01

A Toroidally Symmetric Plasma Simulation (TSPS) code has been developed for investigating the position and shape control on tokamak plasmas. The analyses of three-dimensional eddy currents on the conducting components around the plasma and the two-dimensional magneto-hydrodynamic (MHD) equilibrium are taken into account in this code. The code can analyze the plasma position and shape control during the minor disruption in which the deformation of plasma is not negligible. Using the ITER (International Thermonuclear Experimental Reactor) parameters, some examples of calculations are shown in this paper. (author)

Formation of toroidal pre-heat plasma without residual magnetic field for high-beta pinch experiments

International Nuclear Information System (INIS)

Ikeda, Nagayasu; Tamaru, Ken; Nagata, Akiyoshi.

1979-01-01

Formation of toroidal pre-heat plasma was studied. The pre-heat plasma without residual magnetic field was made by chopping the current for pre-heat, A small toroidal-pinch system was used for the experiment. The magnetic field was measured with a magnetic probe. One turn loop was used for the measurement of the toroidal one-turn electric field. A pair of Rogoski coil was used for the measurement of plasma current. The dependence of residual magnetic field on chopping time was measured. By fast chopping of the primary current in the pre-heating circuit, the poloidal magnetic field was reduced to several percent within 5 microsecond. After chopping, no instability was observed in the principal discharge plasma produced within several microsecond. As the conclusion, it can be said that the control of residual field can be made by current chopping. (Kato, T.)

Proton imaging of hohlraum plasma stagnation in inertial-confinement-fusion experiments

International Nuclear Information System (INIS)

Li, C.K.; Séguin, F.H.; Frenje, J.A.; Sinenian, N.; Rosenberg, M.J.; Manuel, M.J.-E; Rinderknecht, H.G.; Zylstra, A.B.; Petrasso, R.D.; Amendt, P.A.; Landen, O.L.; Mackinnon, A.J.; Town, R.P.J.; Wilks, S.C.; Betti, R.; Meyerhofer, D.D.; Soures, J.M.; Hund, J.; Kilkenny, J.D.; Nikroo, A.

2013-01-01

Proton radiography of the spatial structure and temporal evolution of plasma blowing off from a hohlraum wall reveals how the fill gas compresses the wall blow-off, inhibits plasma jet formation and impedes plasma stagnation in the hohlraum interior. The roles of spontaneously generated electric and magnetic fields in hohlraum dynamics and capsule implosions are demonstrated. The heat flux is shown to rapidly convect the magnetic field due to the Nernst effect, which is shown to be ∼10 times faster than convection by the plasma fluid from expanded wall blow-off (v N ∼ 10v). This leads to inhibition of heat transfer from the gas region in the laser beam paths to the surrounding cold gas, resulting in a local plasma temperature increase. The experiments show that interpenetration of the two materials (gas and wall) occurs due to the classical Rayleigh–Taylor instability as the lighter, decelerating ionized fill gas pushes against the heavier, expanding gold wall blow-off. This experiment provides physics insight into the effects of fill gas on x-ray-driven implosions, and would impact the ongoing ignition experiments at the National Ignition Facility. (paper)

Towards the conceptual design of the ITER real-time plasma control system

International Nuclear Information System (INIS)

Winter, A.; Makijarvi, P.; Simrock, S.; Snipes, J.A.; Wallander, A.; Zabeo, L.

2014-01-01

Highlights: • We describe the main control areas and interfaces for the ITER real-time plasma control system and the current state of their design. • An overview is given for the implementation strategy for the plasma control system as part of the ITER control, data access and communication system. • Current efforts on the creation of simulation and development tools are presented. - Abstract: ITER will be the world's largest magnetic confinement tokamak fusion device and is currently under construction in southern France. The ITER Plasma Control System (PCS) is a fundamental component of the ITER Control, Data Access and Communication system (CODAC). It will control the evolution of all plasma parameters that are necessary to operate ITER throughout all phases of the discharge. The design and implementation of the PCS poses a number of unique challenges. The timescales of phenomena to be controlled spans three orders of magnitude, ranging from a few milliseconds to seconds. Novel control schemes, which have not been implemented at present-day machines need to be developed, and control schemes that are only done as demonstration experiments today will have to become routine. In addition, advances in computing technology and available physics models make the implementation of real-time or faster-than-real-time predictive calculations to forecast and subsequently to avoid disruptions or undesired plasma regimes feasible. This requires the PCS design to be adaptable in real-time to the results of these forecasting algorithms. A further novel feature is a sophisticated event handling system, which provides a means to deal with plasma related events (such as MHD instabilities or L-H transitions) or component failure. Finally, the schedule for design and implementation poses another challenge. The beginning of ITER operation will be in late 2020, but the conceptual design activity of the PCS has already commenced as required by the on-going development of

Overview of the preliminary design of the ITER plasma control system

Science.gov (United States)

Snipes, J. A.; Albanese, R.; Ambrosino, G.; Ambrosino, R.; Amoskov, V.; Blanken, T. C.; Bremond, S.; Cinque, M.; de Tommasi, G.; de Vries, P. C.; Eidietis, N.; Felici, F.; Felton, R.; Ferron, J.; Formisano, A.; Gribov, Y.; Hosokawa, M.; Hyatt, A.; Humphreys, D.; Jackson, G.; Kavin, A.; Khayrutdinov, R.; Kim, D.; Kim, S. H.; Konovalov, S.; Lamzin, E.; Lehnen, M.; Lukash, V.; Lomas, P.; Mattei, M.; Mineev, A.; Moreau, P.; Neu, G.; Nouailletas, R.; Pautasso, G.; Pironti, A.; Rapson, C.; Raupp, G.; Ravensbergen, T.; Rimini, F.; Schneider, M.; Travere, J.-M.; Treutterer, W.; Villone, F.; Walker, M.; Welander, A.; Winter, A.; Zabeo, L.

2017-12-01

An overview of the preliminary design of the ITER plasma control system (PCS) is described here, which focusses on the needs for 1st plasma and early plasma operation in hydrogen/helium (H/He) up to a plasma current of 15 MA with moderate auxiliary heating power in low confinement mode (L-mode). Candidate control schemes for basic magnetic control, including divertor operation and kinetic control of the electron density with gas puffing and pellet injection, were developed. Commissioning of the auxiliary heating systems is included as well as support functions for stray field topology and real-time plasma boundary reconstruction. Initial exception handling schemes for faults of essential plant systems and for disruption protection were developed. The PCS architecture was also developed to be capable of handling basic control for early commissioning and the advanced control functions that will be needed for future high performance operation. A plasma control simulator is also being developed to test and validate control schemes. To handle the complexity of the ITER PCS, a systems engineering approach has been adopted with the development of a plasma control database to keep track of all control requirements.

Self-tuning control studies of the plasma vertical position problem

International Nuclear Information System (INIS)

Zheng, Guang Lin; Wellstead, P.E.; Browne, M.L.

1993-01-01

The plasma vertical position system in a tokamak device can be open-loop unstable with time-varying dynamics, such that the instability increases with system dynamical changes. Time-varying unstable dynamics makes the plasma vertical position a particularly difficult one to control with traditional fixed-coefficient controllers. A self-tuning technique offers a new solution of the plasma vertical position control problem by an adaptive control approach. Specifically, the self-tuning controller automatically tunes the controller parameters without an a priori knowledge of the system dynamics and continuously tracks dynamical changes within the system, thereby providing the system with auto-tuning and adaptive tuning capabilities. An overview of the self-tuning methods is given, and their applicability to a simulation of the Joint European Torus (JET) vertical plasma positions system is illustrated. Specifically, the applicability of pole-assignment and generalized predictive control self-tuning methods to the vertical plasma position system is demonstrated. 26 refs., 16 figs., 1 tab

Wind tunnel experiments on flow separation control of an Unmanned Air Vehicle by nanosecond discharge plasma aerodynamic actuation

International Nuclear Information System (INIS)

Chen Kang; Liang Hua

2016-01-01

Plasma flow control (PFC) is a new kind of active flow control technology, which can improve the aerodynamic performances of aircrafts remarkably. The flow separation control of an unmanned air vehicle (UAV) by nanosecond discharge plasma aerodynamic actuation (NDPAA) is investigated experimentally in this paper. Experimental results show that the applied voltages for both the nanosecond discharge and the millisecond discharge are nearly the same, but the current for nanosecond discharge (30 A) is much bigger than that for millisecond discharge (0.1 A). The flow field induced by the NDPAA is similar to a shock wave upward, and has a maximal velocity of less than 0.5 m/s. Fast heating effect for nanosecond discharge induces shock waves in the quiescent air. The lasting time of the shock waves is about 80 μs and its spread velocity is nearly 380 m/s. By using the NDPAA, the flow separation on the suction side of the UAV can be totally suppressed and the critical stall angle of attack increases from 20° to 27° with a maximal lift coefficient increment of 11.24%. The flow separation can be suppressed when the discharge voltage is larger than the threshold value, and the optimum operation frequency for the NDPAA is the one which makes the Strouhal number equal one. The NDPAA is more effective than the millisecond discharge plasma aerodynamic actuation (MDPAA) in boundary layer flow control. The main mechanism for nanosecond discharge is shock effect. Shock effect is more effective in flow control than momentum effect in high speed flow control. (paper)

Controlling the Plasma-Polymerization Process of N-Vinyl-2-pyrrolidone

DEFF Research Database (Denmark)

Norrman, Kion; Winther-Jensen, Bjørn

2005-01-01

N-vinyl-2-pyrrolidone was plasma-polymerized on glass substrates using a pulsed AC plasma. Pulsed AC plasma produces a chemical surface structure different from that produced by conventional RF plasma; this is ascribed to the different power regimes used. A high degree of control over the structure...... of the chemical surface was obtained using pulsed AC plasma, as shown by ToF-SIMS. It is demonstrated how the experimental conditions to some extent control the chemical structure of the plasma-polymerized film, e.g., film thickness, density of post-plasma-polymerized oligomeric chains, and the density of intact...

Plasma density control with ergodic divertor on Tore Supra; Controle de la densite du plasma en presence du divertor ergodique dans le tokamak Tore Supra

Energy Technology Data Exchange (ETDEWEB)

Meslin, B

1998-04-30

Plasma density control on the tokamak Tore Supra is important for the optimization of every experimental scenario dealing with the improvement of plasma performances. Specific conditions are required both in the plasma bulk and at the edge. Within the framework of the present study, a magnetic configuration is used in the e plasma edge of Tore Supra: the ergodic divertor configuration. A magnetic perturbation which is resonant with the permanent field destroys the plasma confinement locally, opening the field lines onto the material components. They aim of the study is the characterization of the edge density in every relevant scenario for Tore Supra. The first part of this work is dedicated to density and temperature measurements by a series of fixed Langmuir probes located at the very edge of the plasma. Thanks to them, density regimes have been put in evidence during experiments where the volume averaged density , an usual control parameter of the plasma, was varied. The analysis of heat and particle transport through the plasma edge region explains the mechanisms leading to those regimes. The essential factor in our analysis is the dependence of the electron conductivity and ionization depth on temperature. While heat conduction governs the heat transport, the edge density varies linearly according to . Below a critical temperature, reached when the ion flux amplification at constant power density is large enough, a parallel temperature gradient appears leading to a density gradient in the opposite direction in order to maintain the pressure constant along the field lines. A high recycling regime is obtained and the edge density varies like {sup 3}. The pressure conservation is no more satisfied during the detachment of the plasma, which is characterized by a high neutral density at low temperatures leading to a ion momentum loss by friction against the neutrals. The edge density drops in those conditions. These regimes are similar

SPQR II: A beam-plasma interaction experiment

International Nuclear Information System (INIS)

Bimbot, R.; Della-Negra, S.; Gardes, D.

1986-01-01

SPQR II is an interaction experiment designed to probe energy -and charge-exchange of C/sup n/ + ions at 2 MeV/a.m.u., flowing through a fully ionized plasma column of hydrogen with nl-script = 10 19 e-cm -2 at T = 5 eV

Current control for magnetized plasma in direct-current plasma-immersion ion implantation

International Nuclear Information System (INIS)

Tang Deli; Chu, Paul K.

2003-01-01

A method to control the ion current in direct-current plasma-immersion ion implantation (PIII) is reported for low-pressure magnetized inductively coupled plasma. The ion current can be conveniently adjusted by applying bias voltage to the conducting grid that separates plasma formation and implantation (ion acceleration) zones without the need to alter the rf input power, gas flux, or other operating conditions. The ion current that diminishes with an increase in grid bias in magnetized plasmas can be varied from 48 to 1 mA by increasing the grid voltage from 0 to 70 V at -50 kV sample bias and 0.5 mTorr hydrogen pressure. High implantation voltage and monoenergetic immersion implantation can now be achieved by controlling the ion current without varying the macroscopic plasma parameters. The experimental results and interpretation of the effects are presented in this letter. This technique is very attractive for PIII of planar samples that require on-the-fly adjustment of the implantation current at high implantation voltage but low substrate temperature. In some applications such as hydrogen PIII-ion cut, it may obviate the need for complicated sample cooling devices that must work at high voltage

ISTTOK plasma control with the tomography diagnostic

Energy Technology Data Exchange (ETDEWEB)

Fernandes, H.; Caralho, P.J.; Duarte, P.; Pereira, T.; Coelho, R.; Silva, C. [Association Euratom/IST, Institute of Plasmas and Nuclear Fusion, Technology Graduate Institute, P-1049-001 Lisbon (Portugal)

2011-07-01

A real-time plasma position control system is mandatory to achieve long duration (up to 250 ms), Alternating Current (AC) discharges on the ISTTOK tokamak. Such a system has been used for some time supported only on magnetic field diagnostic data. However, this system does not function accurately when the plasma current is low, rendering it inoperative during the plasma current reversal. A tomography diagnostic with 3 pinhole cameras and 8 silicone photodiode channels per camera was installed and customized to supply alternative plasma position to be used for plasma position control. As no filtering is applied, most of the radiation detected is in the visible/near-UV range. This system (i) executes a tomographic reconstruction, (ii) determines the average emissivity position from it, (iii) calculates the shift from the required position and (iv) supplies the vertical field power supply unit with the desired current value, all in less than 100 {mu}s. The horizontal magnetic field power supply unit is expected to be included in the system and will have no impact in the process time. This paper presents the tomography diagnostic architecture together with results of its scientific exploitation in ISTTOK AC discharges, where it has proven to be capable of supplying an accurate plasma position during the current reversal. The use of the tomography diagnostic for plasma position overcomes some limitations of the magnetic diagnostics, but poses challenges of its own such as blindness to plasma current direction. (authors)

Comparison between gas puffing and supersonic molecular beam injection in plasma density feedback experiments in EAST

International Nuclear Information System (INIS)

Zheng, Xingwei; Li, Jiangang; Hu, Jiansheng; Li, Jiahong; Ding, Rui; Cao, Bin; Wu, Jinhua

2013-01-01

To achieve desirable plasma density control, a supersonic molecular beam injection (SMBI) feedback control system has been developed recently for the EAST tokamak. The performance of the SMBI and gas puffing (GP) feedback systems were used and compared. The performance of pulse width mode is better than that of pulse amplitude mode when GP was used for density feedback control. During one-day experiments, the variation of gas input and wall retention can be clarified into two stages. In the first stage the retention ratio is as high as 80–90%, and the gas input is about an order of 10 22 D 2 . However, in the second stage, the retention ratio is at a range of 50–70%. The gas input of a single discharge is small and the net wall retention grows slowly. The results of the SMBI feedback control experiment was analyzed. The shorter delay time of SMBI makes it faster at feeding back control the plasma density. The result showed that, compared with GP, the gas input of SMBI was decreased ∼30% and the wall retention was reduced ∼40%. This shows SMBI's advantage for the long pulse high density discharges in EAST. (paper)

Electron Bernstein wave experiments in a over-dense reversed field pinch plasma

International Nuclear Information System (INIS)

Forest, C. B.; Anderson, J.K.; Cengher, M.; Chattopadhyay, P.K.; Carter, M.; Harvey, R.W.; Pinsker, R.I.; Smirnov, A.P.

2003-01-01

Experiments and theoretical work show that it is possible to couple power to the EBW in an RFP, and that these waves may be suitable for driving current. The main results of our work thus far are: (1) A coupling theory for a phased array of waveguides is developed and compared to experiment. Both O and X mode polarizations can be used; in general coupling for both is optimized for obliquely launched waves. (2) The surface impedance and reflection coefficients have been measured for EBWs launched by waveguide antennas on the edge of MST. Emission and coupling measurements are both consistent with theoretical models and the measured density gradients at the plasma edge. In particular, the coupling showed a strong asymmetry in N Φ for X-mode launch. (3) Black-body levels of emission have been observed in the ECRF from over-dense MST plasmas, which by reciprocity indicate that coupling to the EBW is possible with external antennas. Emission is preferentially polarized in the X-mode and is affected by density fluctuations at the plasma edge. Mode conversion efficiencies as high as 75% have been observed. (4) Ray tracing of EBW waves, coupled to Fokker Planck calculations show that localized, efficient current drive is possible. Current drive is possible by choosing the poloidal angle of the launching antenna to control the N of the wave. (authors)

Using plasma waves to create in tokamaks the necessary quasi-stationary conditions for controlled fusion

International Nuclear Information System (INIS)

Moreau, D.

1993-04-01

It is studied, on the one hand, how using hybrid waves with frequency near from lower hybrid frequency in fusion plasma. Works about coupling waves in plasma (chap.I), their propagation and response of the plasma to the absorption of the waves (chap.II). This method is the most effective until today. Because of limits, it has been investigated, on the other hand, fast magnetosonic wave to control current density in the centre of the discharge in a reactor or a very hot plasma. Theoretical study (chap.III) and experimental results (chap.IV) are presented. Experiments are in progress or planned in following tokamaks: D3-D (USA), JET (Europe), TORE SUPRA (France), JT-60 (Japan). figs. refs. tabs

Laboratory experiments on plasma jets in a magnetic field using high-power lasers

Directory of Open Access Journals (Sweden)

Nishio K.

2013-11-01

Full Text Available The experiments to simulate astrophysical jet generation are performed using Gekko XII (GXII HIPER laser system at the Institute of Laser Engineering. In the experiments a fast plasma flow generated by shooting a CH plane (10 μm thickness is observed at the rear side of the plane. By separating the focal spot of the main beams, a non-uniform plasma is generated. The non-uniform plasma flow in an external magnetic field (0.2∼0.3 T perpendicular to the plasma is more collimated than that without the external magnetic field. The plasma β, the ratio between the plasma and magnetic pressure, is ≫ 1, and the magnetic Reynolds number is ∼150 in the collimated plasma. It is considered that the magnetic field is distorted by the plasma flow and enhances the jet collimation.

The web-based user interface for EAST plasma control system

Energy Technology Data Exchange (ETDEWEB)

Zhang, R.R., E-mail: rrzhang@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Anhui (China); Xiao, B.J. [Institute of Plasma Physics, Chinese Academy of Sciences, Anhui (China); School of Nuclear Science and Technology, University of Science and Technology of China, Anhui (China); Yuan, Q.P. [Institute of Plasma Physics, Chinese Academy of Sciences, Anhui (China); Yang, F. [Institute of Plasma Physics, Chinese Academy of Sciences, Anhui (China); Department of Computer Science, Anhui Medical University, Anhui (China); Zhang, Y. [Institute of Plasma Physics, Chinese Academy of Sciences, Anhui (China); Johnson, R.D.; Penaflor, B.G. [General Atomics, DIII-D National Fusion Facility, San Diego, CA (United States)

2014-05-15

The plasma control system (PCS) plays a vital role at EAST for fusion science experiments. Its software application consists of two main parts: an IDL graphical user interface for setting a large number of plasma parameters to specify each discharge, several programs for performing the real-time feedback control and managing the whole control system. The PCS user interface can be used from any X11 Windows client with privileged access to the PCS computer system. However, remote access to the PCS system via the IDL user interface becomes an extreme inconvenience due to the high network latency to draw or operate the interfaces. In order to realize lower latency for remote access to the PCS system, a web-based system has been developed for EAST recently. The setup data are retrieved from the PCS system and client-side JavaScript draws the interfaces into the user's browser. The user settings are also sent back to the PCS system for controlling discharges. These technologies allow the web-based user interface to be viewed by authorized users with a web browser and have it communicate with PCS server processes directly. It works together with the IDL interface and provides a new way to aid remote participation.

The web-based user interface for EAST plasma control system

International Nuclear Information System (INIS)

Zhang, R.R.; Xiao, B.J.; Yuan, Q.P.; Yang, F.; Zhang, Y.; Johnson, R.D.; Penaflor, B.G.

2014-01-01

The plasma control system (PCS) plays a vital role at EAST for fusion science experiments. Its software application consists of two main parts: an IDL graphical user interface for setting a large number of plasma parameters to specify each discharge, several programs for performing the real-time feedback control and managing the whole control system. The PCS user interface can be used from any X11 Windows client with privileged access to the PCS computer system. However, remote access to the PCS system via the IDL user interface becomes an extreme inconvenience due to the high network latency to draw or operate the interfaces. In order to realize lower latency for remote access to the PCS system, a web-based system has been developed for EAST recently. The setup data are retrieved from the PCS system and client-side JavaScript draws the interfaces into the user's browser. The user settings are also sent back to the PCS system for controlling discharges. These technologies allow the web-based user interface to be viewed by authorized users with a web browser and have it communicate with PCS server processes directly. It works together with the IDL interface and provides a new way to aid remote participation

Preliminary experiments on wastes degradation by thermal plasma

International Nuclear Information System (INIS)

Cota S, G.; Pacheco S, J.; Segovia R, A.; Pena E, R.; Merlo S, L.

1996-01-01

This work presents the fundamental aspects involved in the installation and start up of an experimental equipment for the hazardous wastes degradation using the thermal plasma technology. It is mentioned about the form in which the thermal plasma is generated and the characteristics that its make to be an appropriate technology for the hazardous wastes degradation. Just as the installed structures for to realize the experiments and results of the first studies on degradation, using nylon as problem sample. (Author)

First experiments probing the collision of parallel magnetic fields using laser-produced plasmas

Energy Technology Data Exchange (ETDEWEB)

Rosenberg, M. J., E-mail: mros@lle.rochester.edu; Li, C. K.; Séguin, F. H.; Frenje, J. A.; Petrasso, R. D. [Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Fox, W. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Igumenshchev, I.; Stoeckl, C.; Glebov, V. [Laboratory for Laser Energetics, University of Rochester, Rochester, New York 14623 (United States); Town, R. P. J. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States)

2015-04-15

Novel experiments to study the strongly-driven collision of parallel magnetic fields in β ∼ 10, laser-produced plasmas have been conducted using monoenergetic proton radiography. These experiments were designed to probe the process of magnetic flux pileup, which has been identified in prior laser-plasma experiments as a key physical mechanism in the reconnection of anti-parallel magnetic fields when the reconnection inflow is dominated by strong plasma flows. In the present experiments using colliding plasmas carrying parallel magnetic fields, the magnetic flux is found to be conserved and slightly compressed in the collision region. Two-dimensional (2D) particle-in-cell simulations predict a stronger flux compression and amplification of the magnetic field strength, and this discrepancy is attributed to the three-dimensional (3D) collision geometry. Future experiments may drive a stronger collision and further explore flux pileup in the context of the strongly-driven interaction of magnetic fields.

Strategies for the plasma position and shape control in IGNITOR

International Nuclear Information System (INIS)

Ramogida, G.; Alladio, F.; Albanese, R.

2006-01-01

The control of the plasma position and shape is a crucial issue in IGNITOR as in every compact, high field, elongated tokamak. The capability of the Poloidal Field Coil system, as presently designed, to provide an effective vertical stabilization of the plasma has been investigated using the CREATE L response model [R. Albanese, F. Villone, '' The Linearized CREATE L Plasma Response Model for the Control of Current, Position and Shape in Tokamaks '', Nucl. Fus., vol. 38, p. 723 (1998)]. This linearized MHD model assumes an axisymmetric deformable plasma described by few global parameters. An optimization of the vertical position control strategy has been carried out and the most effective coil combination has been selected to stabilize the plasma while fulfilling engineering constraints on the coils and minimizing the required power and voltage. The two pairs of coils selected for the vertical control will be fed up with up-down anti-symmetric currents provided by a dedicated supply and overlapped to the scenario currents. The growth rate of the vertical instability and the power required by the active stabilization system have been estimated with this model, indicating that it is possible to design a control system able to guarantee a stability region that includes the most interesting operation conditions. An assessment of the requirements for the plasma cross section shape control has been carried out considering independent perturbations of the plasma global parameters as disturbances and showing that the undesired shape modification rejection is possible with the present PFC and power supply system. The PF coils have been ranked with respect to their capability to restore the shape modifications due to different plasma disturbances and the most effective coil combination, that minimizes recovery time and voltage required, has been selected. In order to have additional means to monitor and control the centre of the plasma column, under demanding conditions

Control strategy for plasma equilibrium in a tokamak

International Nuclear Information System (INIS)

Miskell, R.V.

1975-01-01

The dynamic control of the plasma position within the torus of a Tokamak fusion device is a significant factor in the development of nuclear fusion as an energy source. This investigation develops a state variable model of a TOKAMAK thermonuclear device, suitable for application of modern control theory techniques. The model considers eddy currents in the conducting shell surrounding the torus and the classical Shafranov equilibrium equation. The equations necessary to characterize the operating conditions of a TOKAMAK are cast in state variable form. Two control variables are selected, the vertical field current and the plasma temperature. The figure of merit chosen minimizes the shift of the plasma within the torus and considers position perturbations necessary to maintain the dense and hotter portions of the plasma profile in the center of the torus, i.e., overcome uneven poloidal fields due to the toroidal geometry. The model uses a Kalman filter to estimate unmeasured state variables, and uses the second variation of the calculus of variations to maintain an optimal control path. (Diss. Abstr. Int., B)

JACoW Safety instrumented systems and the AWAKE plasma control as a use case

CERN Document Server

Blanco Viñuela, Enrique; Fernández Adiego, Borja; Speroni, Roberto

2018-01-01

Safety is likely the most critical concern in many process industries, yet there is a general uncertainty on the proper engineering to reduce the risks and ensure the safety of persons or material at the same time as providing the process control system. Some of the reasons for this misperception are unclear requirements, lack of functional safety engineering knowledge or incorrect protection functionalities attributed to the BPCS (Basic Process Control System). Occasionally the control engineers are not aware of the hazards inherent to an industrial process and this causes an incorrect design of the overall controls. This paper illustrates the engineering of the SIS (Safety Instrumented System) and the BPCS of the plasma vapour controls of the AWAKE R&D; project, the first proton-driven plasma wakefield acceleration experiment in the world. The controls design and implementation refers to the IEC61511/ISA84 standard, including technological choices, design, operation and maintenance. Finally, the publica...

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

Energy Technology Data Exchange (ETDEWEB)

Bremond, S

1995-10-18

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

Control of magnetohydrodynamic stability by phase space engineering of energetic ions in tokamak plasmas.

Science.gov (United States)

Graves, J P; Chapman, I T; Coda, S; Lennholm, M; Albergante, M; Jucker, M

2012-01-10

Virtually collisionless magnetic mirror-trapped energetic ion populations often partially stabilize internally driven magnetohydrodynamic disturbances in the magnetosphere and in toroidal laboratory plasma devices such as the tokamak. This results in less frequent but dangerously enlarged plasma reorganization. Unique to the toroidal magnetic configuration are confined 'circulating' energetic particles that are not mirror trapped. Here we show that a newly discovered effect from hybrid kinetic-magnetohydrodynamic theory has been exploited in sophisticated phase space engineering techniques for controlling stability in the tokamak. These theoretical predictions have been confirmed, and the technique successfully applied in the Joint European Torus. Manipulation of auxiliary ion heating systems can create an asymmetry in the distribution of energetic circulating ions in the velocity orientated along magnetic field lines. We show the first experiments in which large sawtooth collapses have been controlled by this technique, and neoclassical tearing modes avoided, in high-performance reactor-relevant plasmas.

Baseball II-T, a new target plasma startup experiment

International Nuclear Information System (INIS)

Chargin, A.; Denhoy, B.; Frank, A.; Thomas, S.

1975-01-01

A brief description is given of modifications and additions to the existing Baseball II experiment. These changes will make it possible to study target plasma buildup in a steady-state magnetic field. This experiment, now called Baseball II-T + will use a pellet generator to deliver ammonia pellets into the center of the magnetic mirror field where they will be heated with a 300-J, 50-ns, CO 2 laser. The plasma created by this method will have a density of approximately 10 13 cm -3 and a temperature of about 1 keV. This target plasma will be used for neutral beam injection startup studies with a 50-A, 20-keV neutral beam. Later, the beam power will be increased to study buildup. With ion injection energies of up to 50 keV, it may be possible to achieve etatau as high as 10 12 cm -3 s. The new components necessary to achieve these goals are described

EURATOM-CEA association contributions to the 26. EPS conference on controlled fusion and plasma physics, Maastricht

International Nuclear Information System (INIS)

1999-10-01

This report references the EURATOM-CEA association contributions presented at the 26. EPS conference on controlled fusion and plasma physics, in Maastricht (Netherlands) the 14-18 June 1999. Two invited papers and 24 contributed papers are proposed. They deal with: tokamak devices; particle recirculation in ergodic divertor; current profile control and MHD stability in Tore Supra discharges; edge-plasma control by the ergodic divertor; electron heat transport in stochastic magnetic layer; bolometry and radiated power; particle collection by ergodic divertor; study and simulation of plasma impurities; line shape modelling for plasma edge conditions; dynamical study of the radial structure of the fluctuations measured by reciprocating Langmuir probe in Tore Supra; up-down asymmetry of density fluctuations; Halo currents in a circular tokamak; real time measurement of the position, density, profile and current profile at Tore Supra; poloidal rotation measurement by reflectometry; interpretation of q-profile dependence of the LH power deposition profile during LHCD experiments; ICFR plasma production and optimization; improved core electron confinement; measurement of hard X-ray emission profile; modelling of shear effects on thermal and particles transport; ion turbulence; current drive generation based on autoresonance and intermittent trapping mechanisms. (A.L.B.)

Development in Diagnostics Application to Control Advanced Tokamak Plasma

International Nuclear Information System (INIS)

Koide, Y.

2008-01-01

For continuous operation expected in DEMO, all the plasma current must be non-inductively driven, with self-generated neoclassical bootstrap current being maximized. The control of such steady state high performance tokamak plasma (so-called 'Advanced Tokamak Plasma') is a challenge because of the strong coupling between the current density, the pressure profile and MHD stability. In considering diagnostic needs for the advanced tokamak research, diagnostics for MHD are the most fundamental, since discharges which violate the MHD stability criteria either disrupt or have significantly reduced confinement. This report deals with the development in diagnostic application to control advanced tokamak plasma, with emphasized on recent progress in active feedback control of the current profile and the pressure profile under DEMO-relevant high bootstrap-current fraction. In addition, issues in application of the present-day actuators and diagnostics for the advanced control to DEMO will be briefly addressed, where port space for the advanced control may be limited so as to keep sufficient tritium breeding ratio (TBR)

Plasma physics

International Nuclear Information System (INIS)

1979-01-01

This report contains the papers delivered at the AEB - Natal University summer school on plasma physics held in Durban during January 1979. The following topics were discussed: Tokamak devices; MHD stability; trapped particles in tori; Tokamak results and experiments; operating regime of the AEB Tokamak; Tokamak equilibrium; high beta Tokamak equilibria; ideal Tokamak stability; resistive MHD instabilities; Tokamak diagnostics; Tokamak control and data acquisition; feedback control of Tokamaks; heating and refuelling; neutral beam injection; radio frequency heating; nonlinear drift wave induced plasma transport; toroidal plasma boundary layers; microinstabilities and injected beams and quasilinear theory of the ion acoustic instability

Divertor experiments in a toroidal plasma, with E x B drift due to an applied radial electric field

International Nuclear Information System (INIS)

Strait, E.J.

1979-09-01

It is proposed that the E x B drift arising from an externally applied electric field could be used in a tokamak or other toroidal magnetic plasma confinement device to remove plasma and impurities from the region near the wall and reduce the amount of plasma striking the wall. This could either augment or replace a conventional magnetic field divertor. Among the possible advantages of this scheme are easy external control over the rate of removal of plasma, more rapid removal than the naturally occurring rate in a magnetic divertor, and simplification of construction if the magnetic divertor is eliminated. Results of several related experiments performed in the Wisconsin Levitated Octupole are presented

Laboratory and space experiments as a key to the plasma universe

International Nuclear Information System (INIS)

Faelthammar, C.G.

1993-08-01

Almost all of the known matter in our universe is in the state of plasma. Because of the complexity of the plasma state, a reliable understanding has to be built on empirical knowledge, since theoretical models easily become misleading unless guided by experiment or observation. Cosmical plasmas cover a vast range of densities and temperatures, but in important respects they can be classified into three main categories: high, medium, and low density plasmas. The ability of a plasma to carry electric current is very different in different kinds of plasma, varying from high density plasmas, where the ordinary Ohms law applies to low density plasmas, where no local macroscopic relation needs to exist between electric field and current density. According to classical formulas, the electrical conductivity of many plasmas should be practically infinite. But on the basis of laboratory experiments and in situ measurements in space we now know that in important cases the plasmas ability to carry electric current can be reduced by many powers of ten, and even collisionless plasmas may support significant magnetic-field aligned electric fields. A small number of processes responsible for this have been identified. They include anomalous resistivity, magnetic mirror effect and electric double layers. One of the consequences is possible violation of the frozen field condition, which greatly simplifies the analysis but can be dangerously misleading. Another is the possibility of extremely efficient release of magnetically stored energy. Cosmical plasmas have a strong tendency to form filamentary and cellular structures, which complicates their theoretical description by making homogeneous models inappropriate. In situ observations in the Earths magnetosphere have revealed completely unexpected and still not fully understood chemical separation processes that are likely to be important also in astrophysical plasmas. 108 refs

Asymmetry of edge plasma turbulence in biasing experiments on tokamak TF-2

International Nuclear Information System (INIS)

Budaev, V.P.

1994-01-01

It was observed in tokamaks the suppression of edge turbulence causes by setting a radial electric field at the plasma edge. The poloidal plasma rotation governed by this electric field is likely to result in changes in edge convention and poloidal asymmetry, however there is no experimental evidence about that of the experimental database concerning the biasing and conditions of edge plasma electrostatic turbulence excitation is not still complete. Also a relation between macroscopic convection and small-scale electrostatic turbulence have not yet revealed both in biasing and non biasing plasmas. In this paper results from biasing experiments carried on on ohmically heated tokamak TF-2 are presented. Changes in both equilibrium and fluctuated edge plasma parameters also convection and turbulence driven particle flux were demonstrated in probe measurements with biasing of electrode immersed within Last Closed Flux Surface (LCFS). Poloidal edge plasma structure and charge in asymmetry have demonstrated in the biasing experiments. (author). 6 refs, 4 figs

To the problem of electron temperature control in plasma

Energy Technology Data Exchange (ETDEWEB)

Galechyan, G.A. [Institute of Applied Problem of Physics, Yerevan (Armenia); Anna, P.R. [Raritan Valley Community College, Somerville, NJ (United States)

1995-12-31

One of the main problems in low temperature plasma is control plasma parameters at fixed values of current and gas pressure in the discharge. It is known that an increase in the intensity of sound wave directed along the positive column to values in excess of a definite threshold leads to essential rise of the temperature of electrons. However, no less important is the reduction of electron temperature in the discharge down to the value less than that in plasma in the absence external influence. It is known that to reduce the electron temperature in the plasma of CO{sub 2} laser, easily ionizable admixture are usually introduced in the discharge area with the view of increasing the overpopulation. In the present work we shall show that the value of electron temperature can be reduced by varying of sound wave intensity at its lower values. The experiment was performed on an experimental setup consisted of the tube with length 52 cm and diameter 9.8 cm, two electrodes placed at the distance of 27 cm from each other. An electrodynamical radiator of sound wave was fastened to one of tube ends. Fastened to the flange at the opposite end was a microphone for the control of sound wave parameters. The studies were performed in range of pressures from 40 to 180 Torr and discharge currents from 40 to 110 mA. The intensity of sound wave was varied from 74 to 92 dB. The measurement made at the first resonance frequency f = 150 Hz of sound in the discharge tube, at which a quarter of wave length keep within the length of the tube. The measurement of longitudinal electric field voltage in plasma of positive column was conducted with the help of two probes according to the compensation method. Besides, the measurement of gas temperature in the discharge were taken. Two thermocouple sensors were arranged at the distance of 8 cm from the anode, one of them being installed on the discharge tube axis, the second-fixed the tube wall.

To the problem of electron temperature control in plasma

International Nuclear Information System (INIS)

Galechyan, G.A.; Anna, P.R.

1995-01-01

One of the main problems in low temperature plasma is control plasma parameters at fixed values of current and gas pressure in the discharge. It is known that an increase in the intensity of sound wave directed along the positive column to values in excess of a definite threshold leads to essential rise of the temperature of electrons. However, no less important is the reduction of electron temperature in the discharge down to the value less than that in plasma in the absence external influence. It is known that to reduce the electron temperature in the plasma of CO 2 laser, easily ionizable admixture are usually introduced in the discharge area with the view of increasing the overpopulation. In the present work we shall show that the value of electron temperature can be reduced by varying of sound wave intensity at its lower values. The experiment was performed on an experimental setup consisted of the tube with length 52 cm and diameter 9.8 cm, two electrodes placed at the distance of 27 cm from each other. An electrodynamical radiator of sound wave was fastened to one of tube ends. Fastened to the flange at the opposite end was a microphone for the control of sound wave parameters. The studies were performed in range of pressures from 40 to 180 Torr and discharge currents from 40 to 110 mA. The intensity of sound wave was varied from 74 to 92 dB. The measurement made at the first resonance frequency f = 150 Hz of sound in the discharge tube, at which a quarter of wave length keep within the length of the tube. The measurement of longitudinal electric field voltage in plasma of positive column was conducted with the help of two probes according to the compensation method. Besides, the measurement of gas temperature in the discharge were taken. Two thermocouple sensors were arranged at the distance of 8 cm from the anode, one of them being installed on the discharge tube axis, the second-fixed the tube wall

The design of remote participation platform for EAST plasma control

Energy Technology Data Exchange (ETDEWEB)

Yuan, Q.P., E-mail: qpyuan@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Xiao, B.J. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science & Technology of China, Hefei (China); Zhang, R.R. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Chai, W.T.; Liu, J.; Xiao, R.; Zhou, Z.C.; Pei, X.F. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science & Technology of China, Hefei (China)

2016-11-15

Highlights: • The remote participation platform for EAST plasma control is composed of real time control service and scenario management. • The web based interface has been developed for supporting remote participation. • The functionality module has been designed and assistant tools have been developed. - Abstract: EAST has become a physics experimental platform for high parameter and steady-state long-pulse plasma operation. A new remote participation platform for EAST plasma control is designed, which is composed of gatekeeper system, web-based user interface system, discharge scenario management system, online simulation system and data interface with on-site plasma control system (PCS). The identification and access privilege of remote participator is validated by the gatekeeper system. Only authorized users can set control parameters for next shot plasma control or making discharge scenario for future shot through WebPCS which is a web-based user interface and designed based on B/S structure. The systematic architecture design and preliminary deployment of such remote platform will be presented in this paper.

The Plasma Wave Experiment (PWE) on board the Arase (ERG) satellite

Science.gov (United States)

Kasahara, Yoshiya; Kasaba, Yasumasa; Kojima, Hirotsugu; Yagitani, Satoshi; Ishisaka, Keigo; Kumamoto, Atsushi; Tsuchiya, Fuminori; Ozaki, Mitsunori; Matsuda, Shoya; Imachi, Tomohiko; Miyoshi, Yoshizumi; Hikishima, Mitsuru; Katoh, Yuto; Ota, Mamoru; Shoji, Masafumi; Matsuoka, Ayako; Shinohara, Iku

2018-05-01

The Exploration of energization and Radiation in Geospace (ERG) project aims to study acceleration and loss mechanisms of relativistic electrons around the Earth. The Arase (ERG) satellite was launched on December 20, 2016, to explore in the heart of the Earth's radiation belt. In the present paper, we introduce the specifications of the Plasma Wave Experiment (PWE) on board the Arase satellite. In the inner magnetosphere, plasma waves, such as the whistler-mode chorus, electromagnetic ion cyclotron wave, and magnetosonic wave, are expected to interact with particles over a wide energy range and contribute to high-energy particle loss and/or acceleration processes. Thermal plasma density is another key parameter because it controls the dispersion relation of plasma waves, which affects wave-particle interaction conditions and wave propagation characteristics. The DC electric field also plays an important role in controlling the global dynamics of the inner magnetosphere. The PWE, which consists of an orthogonal electric field sensor (WPT; wire probe antenna), a triaxial magnetic sensor (MSC; magnetic search coil), and receivers named electric field detector (EFD), waveform capture and onboard frequency analyzer (WFC/OFA), and high-frequency analyzer (HFA), was developed to measure the DC electric field and plasma waves in the inner magnetosphere. Using these sensors and receivers, the PWE covers a wide frequency range from DC to 10 MHz for electric fields and from a few Hz to 100 kHz for magnetic fields. We produce continuous ELF/VLF/HF range wave spectra and ELF range waveforms for 24 h each day. We also produce spectral matrices as continuous data for wave direction finding. In addition, we intermittently produce two types of waveform burst data, "chorus burst" and "EMIC burst." We also input raw waveform data into the software-type wave-particle interaction analyzer (S-WPIA), which derives direct correlation between waves and particles. Finally, we introduce our

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

Real-time control environment for the RFX experiment

International Nuclear Information System (INIS)

Barana, O.; Cavinato, M.; Luchetta, A.; Manduchi, G.; Taliercio, C.

2005-01-01

A comprehensive set of control schemes can be presently implemented on RFX due to the enhanced load assembly and renewed power supply system. The schemes include: plasma equilibrium control and resistive wall mode stabilization, aiming at controlling actively the discharge when the passive action of the shell vanishes; the rotation of the localised helical deformation to minimize the enhanced plasma-wall interaction; the MHD mode control and the 'intelligent shell', aiming at achieving a better comprehension of the underlying physics. To the purpose, an integrated, distributed, digital system has been developed consisting of a set of computing nodes. Each node can act either as pre-processing or control station, the former acquiring raw data and computing intermediate control parameters, the latter executing control algorithms and driving the power amplifiers. An overview of the system architecture is presented in the paper with reference to the software real-time environment providing both basic functions, such as data read-out and real-time communication, and useful tools to program control algorithms, to perform simulations and to commission the system. To simulate the control schemes, the real-time environment is extended to include a so called 'simulation mode', in which the real-time nodes exchange their input/output signals with one station running a suitable model of the experiment, for instance the two dimensional FEM code MAXFEA in the case of the equilibrium control. In this way the control system can be tested offline and the time needed for the commissioning of algorithms reduced

Ignition and burn control characteristics of thermonuclear plasmas

International Nuclear Information System (INIS)

Chaniotakis, E.A.

1990-01-01

Achieving the long sought goal of fusion energy requires the attainment of an ignited and controlled thermonuclear plasma. Obtaining an ignited plasma in a tokamak device requires consideration of both the physics of the plasma and the engineering of the machine. With the aide of completely analytical procedure optimized and ignited tokamaks are obtained under various physics assumptions. These designs show the possible advantage of tokamaks characterized by high (∼4.5) aspect ratio, and high (∼15 T) toroidal magnetic field. The control of an ignited plasma is investigated by using auxiliary power modulation. With auxiliary power stable operating points can be created with Q ∼50. Recognizing the need for a fast 1 1/2-D transport model for studying profile effects the plasma transport equations are solved using variational methods. A computer model based on the variational method has been developed. This model solves the 1 1/2-D transport equation very fast with little loss of accuracy. 74 refs., 70 figs., 8 tabs

Introduction to plasma physics and controlled fusion

CERN Document Server

Chen, Francis F

2016-01-01

The third edition of this classic text presents a complete introduction to plasma physics and controlled fusion, written by one of the pioneering scientists in this expanding field.  It offers both a simple and intuitive discussion of the basic concepts of the subject matter and an insight into the challenging problems of current research. This outstanding text offers students a painless introduction to this important field; for teachers, a large collection of problems; and for researchers, a concise review of the fundamentals as well as original treatments of a number of topics never before explained so clearly.  In a wholly lucid manner the second edition covered charged-particle motions, plasmas as fluids, kinetic theory, and nonlinear effects.  For the third edition, two new chapters have been added to incorporate discussion of more recent advances in the field.  The new chapter 9 on Special Plasmas covers non-neutral plasmas, pure electron plasmas, solid and ultra-cold plasmas, pair-ion plasmas, d...

Plasma flow switch and foil implosion experiments on Pegasus II

International Nuclear Information System (INIS)

Cochrane, J.C.; Bartsch, R.R.; Benage, J.R.; Forman, P.R.; Gribble, R.F.; Ladish, J.S.; Oona, H.; Parker, J.V.; Scudder, D.W.; Shlachter, J.S.; Wysocki, F.J.

1993-01-01

Pegasus II is the upgraded version of Pegasus, a pulsed power machine used in the Los Alamos AGEX (Above Ground EXperiments) program. A goal of the program is to produce an intense (> 100 TW) source of soft x-rays from the thermalization of the kinetic energy of a 1 to 10 MJ plasma implosion. The radiation pulse should have a maximum duration of several 10's of nanoseconds and will be used in the study of fusion conditions and material properties. The radiating plasma source will be generated by the thermalization of the kinetic energy of an imploding cylindrical, thin, metallic foil. This paper addresses experiments done on a capacitor bank to develop a switch (plasma flow switch) to switch the bank current into the load at peak current. This allows efficient coupling of bank energy into foil kinetic energy

Translation experiment of a plasma with field reversed configuration

International Nuclear Information System (INIS)

Tanjyo, Masayasu; Okada, Shigefumi; Ito, Yoshifumi; Kako, Masashi; Ohi, Shoichi

1984-01-01

Experiments to translate the FRC plasma from is formation area (pinch coil) into two kinds of metal vessels (magnetic flux conservers) with larger and smaller bore than that of the pinch coil have been carried out in OCT with an aim of improving the particle confinement time tau sub(N) by increasing xsub(s) (ratio of the plasma radius to that of the conducting wall). Demonstrated were successful translations of the plasma into both vessels. The xsub(s) of the translated plasma increased to 0.6 in the larger bore vessel and to 0.7 in the smaller one from 0.4 of the source plasma in the pinch coil. With the increase in xsub(s), tau sub(N) and also decay time of the trapped magnetic flux are extended from 15 - 20 μs of the source plasma to 50 - 80 μs. The tau sub(N) is found to have stronger dependence on xsub(s) than on rsub(s). During the translation phase, almost half of the total particle and the plasma energy are lost. The plasma volume is, therefore, about half of that expected from the analysis on the ideal translation process. It is also found that the translation process is nearly isothermal as is expected from the analysis. (author)

Simultaneous real-time control of the current and pressure profiles in JET: experiments and modelling

Energy Technology Data Exchange (ETDEWEB)

Mazon, D.; Laborde, L.; Litaudon, X.; Moreau, D.; Zabeo, L.; Joffrin, E. [Association Euratom-CEA Cadarache (DSM/DRFC), 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Murari, A. [Consorzio RFX Association Euratom-ENEA, Padova (Italy); Ariola, M.; Albanese, R.; Tommasi, G. de; Pironti, A. [Association Euratom-ENEA, CREATE, Napoly (Italy); Moreau, D. [EFDA-JET CSU, Culham Science Centre, Abingdon, OX (United Kingdom); Tala, T. [Euratom-Tekes Association, VTT Processes (Finland); Crisanti, F.; Pericoli-Ridolfini, V.; Tuccillo, A. [Association Euratom-ENEA, C.R. Frascati (Italy); Baar, M. de; Vries, P. de [Euratom-FOM Association, TEC Cluster, Nieuwegein (Netherlands); De la Luna, E. [Euratom-Ciemat Association (Spain); Felton, R.; Corrigan, G. [Euratom-UKAEA Association, Culham Science Centre, Abingdon (United Kingdom)

2004-07-01

Real-time control of the plasma profiles (current density, pressure and flow) is one of the major issues for sustaining internal transport barriers (ITB) in a high performance plasma, with a large bootstrap current fraction. We have recently investigated the experimental and numerical aspects of the simultaneous control of the current and pressure profiles in JET ITB discharges. The current density and the electron temperature were successfully controlled via the safety factor profile (or via its inverse the tau-profile) and the {rho}{sup *}{sub Te} profile respectively. The results of these new studies are presented. With only a limited number of actuators, the technique aims at minimizing an integral square error signal which combines the 2 profiles, rather than attempting to control plasma parameters at some given radii with great precision. The resulting fuzziness of the control scheme allows the plasma to relax towards a physically accessible non-linear state which may not be accurately known in advance, but is close enough to the requested one to provide the required plasma performance. Closed loop experiments have allow to reach different target q and {rho}{sup *}{sub Te} profiles, and to some degree, to displace the region of maximum electron temperature gradient. The control has also shown some robustness in front of rapid transients.

Method of controlling plasma discharge in a thermonuclear device

International Nuclear Information System (INIS)

Kawasaki, Kozo; Ishida, Takayuki; Takemaru, Koichi; Kawasaki, Takahide.

1982-01-01

Purpose: To prolong the plasma discharging period by previously increasing the temperature at the thick portion of a vacuum container prior to the plasma discharge to thereby decrease the temperature difference caused by the plasma discharge between the thick portion and the bellows. Method: Temperature values at the outer surface of the thick portion and the bellows of a vacuum container detected by temperature sensors are applied to the input processing section of a temperature control device, and baking control is carried out by way of the output processing section so that each of the portions of the vacuum container may be maintained at the temperature set by the temperature setting section based on the calculation performed in the control processing section. By previously increasing the temperature β at the thick portion higher by about 100 0 C than the temperature α for the bellows in the baking treatment prior to the plasma discharge, the plasma discharge period during which the temperature levels at both of the portions are reversed after the plasma discharge and the temperature difference arrives at a predetermined level i.g., of 100 0 C can significantly be prolonged as compared with the case where the plasma discharge is started at the same temperature for both of the portions. (Yoshino, Y.)

Implementation strategy for the ITER plasma control system

International Nuclear Information System (INIS)

Winter, A.; Ambrosino, G.; Bauvir, B.; De Tommasi, G.; Humphreys, D.A.; Mattei, M.; Neto, A.; Raupp, G.; Snipes, J.A.; Stephen, A.V.; Treutterer, W.; Walker, M.L.; Zabeo, L.

2015-01-01

This paper gives an overview of the scope and context of the CODAC high-level real-time applications (Supervision and Plasma Control) and presents the strategy and current state of design of the tools to support the implementation. A real-time framework, which is currently under development with strong support of the worldwide fusion community will not only support the implementation of plasma control strategies with the extensive exception handling and forecasting functionality foreseen for ITER, but also integrated commissioning, orchestration and supervision as well as the real-time needs of ITER plant system developers. A second cornerstone in the implementation strategy is the development of a powerful simulation environment (Plasma Control System Simulation Platform – PCSSP) to design and verify control strategies, event handling and orchestration and automation. The development of PCSSP is currently under contract and this paper will also give an overview of its current state of development.

Implementation strategy for the ITER plasma control system

Energy Technology Data Exchange (ETDEWEB)

Winter, A., E-mail: axel.winter@iter.org [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Ambrosino, G. [CREATE/Università di Napoli Federico II, Dip. Ingegneria Elettrica e delle Tecnologie dell’Informazione (Italy); Bauvir, B. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); De Tommasi, G. [CREATE/Università di Napoli Federico II, Dip. Ingegneria Elettrica e delle Tecnologie dell’Informazione (Italy); Humphreys, D.A. [General Atomics, San Diego, CA (United States); Mattei, M. [CREATE/Seconda Università di Napoli, Dip. Ingegneria Industriale e dell’Informazione (Italy); Neto, A. [Fusion for Energy, Barcelona (Spain); Raupp, G. [Max Planck Institute for Plasma Physics, EURATOM Association, Garching (Germany); Snipes, J.A. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France); Stephen, A.V. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon (United Kingdom); Treutterer, W. [Max Planck Institute for Plasma Physics, EURATOM Association, Garching (Germany); Walker, M.L. [General Atomics, San Diego, CA (United States); Zabeo, L. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St Paul Lez Durance Cedex (France)

2015-10-15

This paper gives an overview of the scope and context of the CODAC high-level real-time applications (Supervision and Plasma Control) and presents the strategy and current state of design of the tools to support the implementation. A real-time framework, which is currently under development with strong support of the worldwide fusion community will not only support the implementation of plasma control strategies with the extensive exception handling and forecasting functionality foreseen for ITER, but also integrated commissioning, orchestration and supervision as well as the real-time needs of ITER plant system developers. A second cornerstone in the implementation strategy is the development of a powerful simulation environment (Plasma Control System Simulation Platform – PCSSP) to design and verify control strategies, event handling and orchestration and automation. The development of PCSSP is currently under contract and this paper will also give an overview of its current state of development.

Pre-study of burn control in Tokamak reactor experiments

International Nuclear Information System (INIS)

Elevant, T.; Anderson, D.; Hamnen, H.; Lisak, M.

1991-04-01

Findings from a general study of issues associated with control of burning fusion plasmas are reported, and applications to ITER are given. A number of control variables are discussed. A zerodimensional system has been developed and stability against coupled temperature and density variations are studied. Also space dependent energy balance and transition to thermonuclear burn are analysed as well as maximum obtainable Q-values under subignited operation conditions. Control designs with different input-output strategies are analysed and numerically simulated, and a numerical experiment of system identification is made. Requirements on diagnostics are discussed and areas for further studies are identified. (au) (64 refs.)

Prospects for observing the magnetorotational instability in the plasma Couette experiment

Science.gov (United States)

Flanagan, K.; Clark, M.; Collins, C.; Cooper, C. M.; Khalzov, I. V.; Wallace, J.; Forest, C. B.

2015-08-01

Many astrophysical disks, such as protoplanetary disks, are in a regime where non-ideal, plasma-specific magnetohydrodynamic (MHD) effects can significantly influence the behaviour of the magnetorotational instability (MRI). The possibility of studying these effects in the plasma Couette experiment (PCX) is discussed. An incompressible, dissipative global stability analysis is developed to include plasma-specific two-fluid effects and neutral collisions, which are inherently absent in analyses of Taylor-Couette flows (TCFs) in liquid metal experiments. It is shown that with boundary driven flows, a ion-neutral collision drag body force significantly affects the azimuthal velocity profile, thus limiting the flows to regime where the MRI is not present. Electrically driven flow (EDF) is proposed as an alternative body force flow drive in which the MRI can destabilize at more easily achievable plasma parameters. Scenarios for reaching MRI relevant parameter space and necessary hardware upgrades are described.

Surface Treatment of a Lithium Limiter for Spherical Torus Plasma Experiments

Energy Technology Data Exchange (ETDEWEB)

Kaita, R.; Majeski, R.; Doerner, R.; Antar, G.; Timberlake, J.; Spaleta, J.; Hoffman, D.; Jones, B.; Munsat, T.; Kugel, H.; Taylor, G.; Stutman, D.; Soukhanovskii, V.; Maingi, R.; Molesa, S.; Efthimion, P.; Menard, J.; Finkenthal, M.; Luckhardt, S.

2001-03-20

The concept of a flowing lithium first wall for a fusion reactor may lead to a significant advance in reactor design, since it could virtually eliminate the concerns with power density and erosion, tritium retention, and cooling associated with solid walls. As part of investigations to determine the feasibility of this approach, plasma interaction questions in a toroidal plasma geometry are being addressed in the Current Drive eXperiment-Upgrade (CDX-U) spherical torus (ST). The first experiments involved a toroidally local lithium limiter (L3). Measurements of pumpout rates indicated that deuterium pumping was greater for the L3 compared to conventional boron carbide limiters. The difference in the pumpout rates between the two limiter types decreased with plasma exposure, but argon glow discharge cleaning was able to restore the pumping effectiveness of the L3. At no point, however, was the extremely low recycling regime reported in previous lithium experiments achieved. This may be due to the much larger lithium surfaces that were exposed to the plasma in the earlier work. The possibility will be studied in the next set of CDX-U experiments, which are to be conducted with a large area, fully toroidal lithium limiter.

Surface Treatment of a Lithium Limiter for Spherical Torus Plasma Experiments

International Nuclear Information System (INIS)

Kaita, R.; Majeski, R.; Doerner, R.; Antar, G.; Timberlake, J.; Spaleta, J.; Hoffman, D.; Jones, B.; Munsat, T.; Kugel, H.; Taylor, G.; Stutman, D.; Soukhanovskii, V.; Maingi, R.; Molesa, S.; Efthimion, P.; Menard, J.; Finkenthal, M.; Luckhardt, S.

2001-01-01

The concept of a flowing lithium first wall for a fusion reactor may lead to a significant advance in reactor design, since it could virtually eliminate the concerns with power density and erosion, tritium retention, and cooling associated with solid walls. As part of investigations to determine the feasibility of this approach, plasma interaction questions in a toroidal plasma geometry are being addressed in the Current Drive eXperiment-Upgrade (CDX-U) spherical torus (ST). The first experiments involved a toroidally local lithium limiter (L3). Measurements of pumpout rates indicated that deuterium pumping was greater for the L3 compared to conventional boron carbide limiters. The difference in the pumpout rates between the two limiter types decreased with plasma exposure, but argon glow discharge cleaning was able to restore the pumping effectiveness of the L3. At no point, however, was the extremely low recycling regime reported in previous lithium experiments achieved. This may be due to the much larger lithium surfaces that were exposed to the plasma in the earlier work. The possibility will be studied in the next set of CDX-U experiments, which are to be conducted with a large area, fully toroidal lithium limiter

Complex astrophysical experiments relating to jets, solar loops, and water ice dusty plasma

Science.gov (United States)

Bellan, P. M.; Zhai, X.; Chai, K. B.; Ha, B. N.

2015-10-01

> Recent results of three astrophysically relevant experiments at Caltech are summarized. In the first experiment magnetohydrodynamically driven plasma jets simulate astrophysical jets that undergo a kink instability. Lateral acceleration of the kinking jet spawns a Rayleigh-Taylor instability, which in turn spawns a magnetic reconnection. Particle heating and a burst of waves are observed in association with the reconnection. The second experiment uses a slightly different setup to produce an expanding arched plasma loop which is similar to a solar corona loop. It is shown that the plasma in this loop results from jets originating from the electrodes. The possibility of a transition from slow to fast expansion as a result of the expanding loop breaking free of an externally imposed strapping magnetic field is investigated. The third and completely different experiment creates a weakly ionized plasma with liquid nitrogen cooled electrodes. Water vapour injected into this plasma forms water ice grains that in general are ellipsoidal and not spheroidal. The water ice grains can become quite long (up to several hundred microns) and self-organize so that they are evenly spaced and vertically aligned.

Novel magnetic controlled plasma sputtering method

International Nuclear Information System (INIS)

Axelevich, A.; Rabinovich, E.; Golan, G.

1996-01-01

A novel method to improve thin film vacuum sputtering is presented. This method is capable of controlling the sputtering plasma via an external set of magnets, in a similar fashion to the tetrode sputtering method. The main advantage of the Magnetic Controlled Plasma Sputtering (MCPS) is its ability to independently control all deposition parameters without any interference or cross-talk. Deposition rate, using the MCPS, is found to be almost twice the rate of triode and tetrode sputtering techniques. Experimental results using the MCPS to deposit Ni layers are described. It was demonstrated that using the MCPS method the ion beam intensity at the target is a result of the interaction of a homogeneous external magnetic field and the controlling magnetic fields. The MCPS method was therefore found to be beneficial for the production of pure stoichiometric thin solid films with high reproducibility. This method could be used for the production of compound thin films as well. (authors)

Laser Induced Fluorescence Diagnostic for the Plasma Couette Experiment

Science.gov (United States)

Katz, Noam; Skiff, Fred; Collins, Cami; Weisberg, Dave; Wallace, John; Clark, Mike; Garot, Kristine; Forest, Cary

2010-11-01

The Plasma Couette Experiment (PCX) at U. Wisconsin-Madison consists of a rotating high-beta plasma and is well-suited to the study of flow-driven, astrophysically-relevant plasma phenomena. PCX confinement relies on alternating rings of 1kG permanent magnets and the rotation is driven by electrode rings, interspersed between the magnets, which provide an azimuthal ExB. I will discuss the development of a laser-induced fluorescence diagnostic (LIF) to characterize the ion distribution function of argon plasmas in PCX. The LIF system--which will be scanned radially--will be used to calibrate internal Mach probes, as well as to measure the time-resolved velocity profile, ion temperature and density non-perturbatively. These diagnostics will be applied to study the magneto-rotational instability in a plasma, as well as the buoyancy instability thought to be involved in producing the solar magnetic field. This work is supported by NSF and DOE.

EURATOM-CEA Association Contributions to the 16. European Conference on Controlled Fusion and Plasma Physics

International Nuclear Information System (INIS)

1989-01-01

The contributions to the 16th European Conference on controlled fusion and Plasma Physics are presented. The following subjects, concerning Tore Supra, are discussed: runaway electrons dynamics and confinement; spectroscopic studies of plasma surface interactions; ergodic divertor experiments; magnetic field structure and transport induced by the ergodic divertor; fast ions losses during neutral beam injection; current profile control by electron-cyclotron and lower-hybrid waves; and electromagnetic analysis of the lower hybrid system. The report also includes studies on: a possible explanation for the runaway energy limit (resonant interaction with the ripple field); thermal equilibrium of the edge plasma with an ergodic divertor; neutral confinement in pump limiter with a throat; microtearing turbulence and heat transport; toroidal coupling and frequency spectrum of tearing modes; collisionless fast ion dynamics in tokamaks; variational description of lower hybrid wave propagation and absorption in tokamaks; magnetodrift turbulence and disruptions; specific turbulence associated with sawtooth relaxations in TFR plasmas; detailed structure of the q profile around q = 1 in JET; turbulence propagation during pellet injection; tokamak reactor concept with 100% bootstrap current; optimization of a steady state tokamak driven by lower hybrid waves; and thermodesorption of graphite exposed to a deuterium plasma

Plasma under control: Advanced solutions and perspectives for plasma flux management in material treatment and nanosynthesis

Science.gov (United States)

Baranov, O.; Bazaka, K.; Kersten, H.; Keidar, M.; Cvelbar, U.; Xu, S.; Levchenko, I.

2017-12-01

Given the vast number of strategies used to control the behavior of laboratory and industrially relevant plasmas for material processing and other state-of-the-art applications, a potential user may find themselves overwhelmed with the diversity of physical configurations used to generate and control plasmas. Apparently, a need for clearly defined, physics-based classification of the presently available spectrum of plasma technologies is pressing, and the critically summary of the individual advantages, unique benefits, and challenges against key application criteria is a vital prerequisite for the further progress. To facilitate selection of the technological solutions that provide the best match to the needs of the end user, this work systematically explores plasma setups, focusing on the most significant family of the processes—control of plasma fluxes—which determine the distribution and delivery of mass and energy to the surfaces of materials being processed and synthesized. A novel classification based on the incorporation of substrates into plasma-generating circuitry is also proposed and illustrated by its application to a wide variety of plasma reactors, where the effect of substrate incorporation on the plasma fluxes is emphasized. With the key process and material parameters, such as growth and modification rates, phase transitions, crystallinity, density of lattice defects, and others being linked to plasma and energy fluxes, this review offers direction to physicists, engineers, and materials scientists engaged in the design and development of instrumentation for plasma processing and diagnostics, where the selection of the correct tools is critical for the advancement of emerging and high-performance applications.

Progress of plasma experiments and superconducting technology in LHD

International Nuclear Information System (INIS)

Motojima, O.; Sakakibara, S.; Imagawa, S.; Sagara, A.; Seki, T.; Mutoh, T.; Morisaki, T.; Komori, A.; Ohyabu, N.; Yamada, H.

2006-01-01

The large helical device is a heliotron device with L = 2 and M = 10 continuous helical coils and three pairs of poloidal coils, and all of coils are superconductive. Since the experiments started in 1998, the development of engineering technologies and the demonstration of large-superconducting-machine operations have greatly contributed to an understanding of physics in currentless plasmas and a verification of the capability of fully steady-state operation. In recent plasma experiments, the steady state and high-beta experiments, which are the most important subjects for the realization of attractive fusion reactors, have progressed remarkably and produced two world-record parameters, i.e. the highest average beta of 4.5% in helical devices and the highest total input energy of 1.6 GJ in all magnetic confinement devices. No degradation has been observed in the coil performance, and stable cryogenic operational schemes at 4.4 K have been established. The physics and engineering results from the LHD experiment directly contribute to the design study for a D-T fusion demo reactor FFHR with a LHD-type heliotron configuration

SPQR II: A beam-plasma interaction experiment

Science.gov (United States)

Bimbot, R.; Della-Negra, S.; Gardès, D.; Rivet, M. F.; Fleurier, C.; Dumax, B.; Hoffman, D. H. H.; Weyrich, K.; Deutsch, C.; Maynard, G.

1986-01-01

SPQR II is an interaction experiment designed to probe energy -and charge-exchange of Cn+ ions at 2 MeV/a.m.u., flowing through a fully ionized plasma column of hydrogen with nℓ=1019 e-cm-2 at T=5 eV. One expects a factor of two enhanced stopping over the cold gas case.

Surface temperature: A key parameter to control the propanethiol plasma polymer chemistry

Energy Technology Data Exchange (ETDEWEB)

Thiry, Damien, E-mail: damien.thiry@umons.ac.be; Aparicio, Francisco J. [Chimie des Interactions Plasma Surface (ChIPS), CIRMAP, Université de Mons, 23 Place du Parc, B-7000 Mons (Belgium); Laha, Priya; Terryn, Herman [Research Group Electrochemical and Surface Engineering (SURF), Department of Materials and Chemistry (MACH), Pleinlaan 2, 1050 Brussel (Belgium); Snyders, Rony [Chimie des Interactions Plasma Surface (ChIPS), CIRMAP, Université de Mons, 23 Place du Parc, B-7000 Mons, Belgium and Materia Nova Research Center, Parc Initialis, B-7000 Mons (Belgium)

2014-09-01

In this work, the influence of the substrate temperature (T{sub s}) on the chemical composition of propanethiol plasma polymers was investigated for a given set of plasma conditions. In a first study, a decrease in the atomic sulfur content (at. %S) with the deposition time (t{sub d}) was observed. This behavior is explained by the heating of the growing film during deposition process, limiting the incorporation of stable sulfur-based molecules produced in the plasma. Experiments carried out by controlling the substrate temperature support this hypothesis. On the other hand, an empirical law relating the T{sub s} and the at. %S was established. This allows for the formation of gradient layer presenting a heterogeneous chemical composition along the thickness, as determined by depth profile analysis combining X-ray photoelectron spectroscopy and C{sub 60} ion gun sputtering. The experimental data fit with the one predicted from our empiric description. The whole set of our results provide new insights in the relationship between the substrate temperature and the sulfur content in sulfur-based plasma polymers, essential for future developments.

Final Technical Report: Numerical and Experimental Investigation of Turbulent Transport Control via Shaping of Radial Plasma Flow Profiles

Energy Technology Data Exchange (ETDEWEB)

Schuster, Eugenio

2014-05-02

The strong coupling between the different physical variables involved in the plasma transport phenomenon and the high complexity of its dynamics call for a model-based, multivariable approach to profile control where those predictive models could be exploited. The overall objective of this project has been to extend the existing body of work by investigating numerically and experimentally active control of unstable fluctuations, including fully developed turbulence and the associated cross-field particle transport, via manipulation of flow profiles in a magnetized laboratory plasma device. Fluctuations and particle transport can be monitored by an array of electrostatic probes, and Ex B flow profiles can be controlled via a set of biased concentric ring electrodes that terminate the plasma column. The goals of the proposed research have been threefold: i- to develop a predictive code to simulate plasma transport in the linear HELCAT (HELicon-CAThode) plasma device at the University of New Mexico (UNM), where the experimental component of the proposed research has been carried out; ii- to establish the feasibility of using advanced model-based control algorithms to control cross-field turbulence-driven particle transport through appropriate manipulation of radial plasma flow profiles, iii- to investigate the fundamental nonlinear dynamics of turbulence and transport physics. Lehigh University (LU), including Prof. Eugenio Schuster and one full-time graduate student, has been primarily responsible for control-oriented modeling and model-based control design. Undergraduate students have also participated in this project through the National Science Foundation Research Experience for Undergraduate (REU) program. The main goal of the LU Plasma Control Group has been to study the feasibility of controlling turbulence-driven transport by shaping the radial poloidal flow profile (i.e., by controlling flow shear) via biased concentric ring electrodes.

Physics and chemistry of plasma pollution control technology

International Nuclear Information System (INIS)

Chang, J S

2008-01-01

Gaseous pollution control technologies for acid gases (NO x , SO x , etc), volatile organic compounds, greenhouse gases, ozone layer depleting substances, etc have been commercialized based on catalysis, incineration and adsorption methods. However, non-thermal plasma techniques based on electron beams and corona discharges are becoming significant due to advantages such as lower costs, higher removal efficiency and smaller space volume. In order to commercialize this new technology, the pollution gas removal rate, energy efficiency of removal, pressure drop of reactors and useable by-product production rates must be improved and identification of major fundamental processes and optimizations of reactor and power supply for an integrated system must be investigated. In this work, the chemistry and physics of plasma pollution control are discussed and the limitation of this type of plasma is outlined based on the plasma parameters.

Plasma physics and controlled nuclear fusion research

International Nuclear Information System (INIS)

1980-01-01

Full text: During the last decade, growing efforts have been devoted to studying the possible forms an electricity-producing thermonuclear reactor might take and the various technical problems that will have to be overcome. Previous IAEA Conferences took place in Salzburg (1961), Culham (1965), Novosibirsk (1968), Madison (1971), Tokyo (1974), Berchtesgaden (1976) and Innsbruck (1978) The exchange of information that has characterized this series of meetings is an important example of international co-operation and has contributed substantially to progress in controlled fusion research. The results of experiments in major research establishments, as well as the growing scientific insights in the field of plasma physics, give hope that the realization of nuclear fusion will be made possible on a larger scale and beyond the laboratory stage by the end of this century. The increase of the duration of existing tokamak discharges requires solution of the impurity control problem. First results from the new big machines equipped with the poloidal divertor recently came into operation. PDX (USA) and ASDEX (F.R. of Germany) show that various divertor configurations can be established and maintained and that the divertors function in the predicted manner. The reduction of high-Z impurities on these machines by a factor 10 was achieved. As a result of extensive research on radio-frequency (RF) plasma heating on tokamaks: PLT (USA), TFR (France), JFT-2 (Japan), the efficiency of this attractive method of plasma heating comparable to neutral beam heating was demonstrated. It was shown that the density of the input power of about 5-10 kW/cm 2 is achievable and this limit is high enough for application to reactor-like machines. One of the inspiring results reported at the conference was the achievement of value (the ratio of plasma pressure to magnetic field pressure) of ∼ 3% on tokamaks T-11 (USSR) and ISX-B (USA). It is important to note that this value exceeds the

Progress and plan of KSTAR plasma control system upgrade

Energy Technology Data Exchange (ETDEWEB)

Hahn, Sang-hee, E-mail: hahn76@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Kim, Y.J. [National Fusion Research Institute, Daejeon (Korea, Republic of); Penaflor, B.G. [General Atomics, San Diego, CA (United States); Bak, J.G.; Han, H.; Hong, J.S.; Jeon, Y.M.; Jeong, J.H.; Joung, M.; Juhn, J.W.; Kim, J.S.; Kim, H.S.; Lee, W.R.; Woo, M.H. [National Fusion Research Institute, Daejeon (Korea, Republic of); Eidietis, N.W.; Ferron, J.R.; Humphreys, D.A.; Hyatt, A.; Johnson, R.D.; Piglowski, D.A. [General Atomics, San Diego, CA (United States); and others

2016-11-15

Highlights: • Recent achievements of the KSTAR plasma control system are described. • Requirements and results of the testbed system for the future upgrade of the KSTAR plasma control system are presented. • An overview of the upgrade layout based is given. - Abstract: The plasma control system (PCS) has been one of essential systems in annual KSTAR plasma campaigns: starting from a single-process version in 2008, extensive upgrades are done through the previous 7 years in order to achieve major goals of KSTAR performance enhancement. Major implementations are explained in this paper. In consequences of successive upgrades, the present KSTAR PCS is able to achieve ∼48 s of 500 kA plasma pulses with full real-time shaping controls and real-time NB power controls. It has become a huge system capable of dealing with 8 separate categories of algorithms, 26 actuators directly controllable during the shot, and real-time data communication units consisting of +180 analog channels and +600 digital input/outputs through the reflective memory (RFM) network. The next upgrade of the KSTAR PCS is planned in 2015 before the campaign. An overview of the upgrade layout will be given for this paper. The real-time system box is planned to use the CERN MRG-Realtime OS, an ITER-compatible standard operating system. New hardware is developed for faster real-time streaming system for future installations of actuators/diagnostics.

Experimental studies on beam-plasma interaction

International Nuclear Information System (INIS)

Kiwamoto, Y.

1977-01-01

Beam-handling technology has reached now at such a level as to enable highly controlled experiments of beam-plasma interaction. Varieties of hypotheses and suppositions about the beam propagation and interaction in space plasma can be proved and often be corrected by examining the specific processes in laboratory plasma. The experiments performed in this way by the author are briefed: ion beam instability in unmagnetized plasma; ion beam instability perpendicular to magnetic field; and electron beam instability. (Mori, K.)

Plasma control using neural network and optical emission spectroscopy

International Nuclear Information System (INIS)

Kim, Byungwhan; Bae, Jung Ki; Hong, Wan-Shick

2005-01-01

Due to high sensitivity to process parameters, plasma processes should be tightly controlled. For plasma control, a predictive model was constructed using a neural network and optical emission spectroscopy (OES). Principal component analysis (PCA) was used to reduce OES dimensionality. This approach was applied to an oxide plasma etching conducted in a CHF 3 /CF 4 magnetically enhanced reactive ion plasma. The etch process was systematically characterized by means of a statistical experimental design. Three etch outputs (etch rate, profile angle, and etch rate nonuniformity) were modeled using three different approaches, including conventional, OES, and PCA-OES models. For all etch outputs, OES models demonstrated improved predictions over the conventional or PCA-OES models. Compared to conventional models, OES models yielded an improvement of more than 25% in modeling profile angle and etch rate nonuniformtiy. More than 40% improvement over PCA-OES model was achieved in modeling etch rate and profile angle. These results demonstrate that nonreduced in situ data are more beneficial than reduced one in constructing plasma control model

Linear quadratic Gaussian controller design for plasma current, position and shape control system in ITER

International Nuclear Information System (INIS)

Belyakov, V.; Kavin, A.; Rumyantsev, E.; Kharitonov, V.; Misenov, B.; Ovsyannikov, A.; Ovsyannikov, D.; Veremei, E.; Zhabko, A.; Mitrishkin, Y.

1999-01-01

This paper is focused on the linear quadratic Gaussian (LQG) controller synthesis methodology for the ITER plasma current, position and shape control system as well as power derivative management system. It has been shown that some poloidal field (PF) coils have less influence on reference plasma-wall gaps control during plasma disturbances and hence they have been used to reduce total control power derivative by means of the additional non-linear feedback. The design has been done on the basis of linear models. Simulation was provided for non-linear model and results are presented and discussed. (orig.)

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Field experiments and laboratory study of plasma turbulence and effects on EM wave propagation

International Nuclear Information System (INIS)

Lee, M.C.; Kuo, S.P.

1990-01-01

Both active experiments in space and laboratory experiments with plasma chambers have been planned to investigate plasma turbulence and effects on electromagnetic wave propagation. Plasma turbulence can be generated by intense waves or occur inherently with the production of plasmas. The turbulence effects to be singled out for investigation include nonlinear mode conversion process and turbulence scattering of electromagnetic waves by plasma density fluctuations. The authors have shown theoretically that plasma density fluctuations can render the nonlinear mode conversion of electromagnetic waves into lower hybrid waves, leading to anomalous absorption of waves in magnetoplasmas. The observed spectral broadening of VLF waves is the evidence of the occurrence of this process. Since the density fluctuations may have a broad range of scale lengths, this process is effective in weakening the electromagnetic waves in a wideband. In addition, plasma density fluctuations can scatter waves and diversify the electromagnetic energy. Schemes of generating plasma turbulence and the diagnoses of plasma effects are discussed

Noncircular plasma shape analysis in long-pulse current drive experiment in TRIAM-1M

International Nuclear Information System (INIS)

Minooka, Mayumi; Kawasaki, Shoji; Jotaki, Eriko; Moriyama, Shin-ichi; Nagao, Akihiro; Nakamura, Kazuo; Hiraki, Naoji; Nakamura, Yukio; Itoh, Satoshi

1991-01-01

Plasma cross section was noncircularized and the plasma shape was analyzed in order to study the characteristics of the plasma in long-pulse current drive experiments in high-field superconducting tokamak TRIAM-1M. Filament approximation method was adopted, since on-line processing by data processing computer is possible. The experiments of the noncircularization were carried out during 30-to 60-sec discharges. As a result, it became clear that D-shape plasma of elongation ratio 1.4 was maintained stably. By the analysis the internal inductance and poloidal beta were assessed, and so informations about the plasma current profile and internal pressure were obtained. (author)

Impact of gas puffing location on density control and plasma parameters in TJ-II

International Nuclear Information System (INIS)

Tabares, F.L.; Garcia-Cortes, I.; Estrada, T.; Tafalla, D.; Hidalgo, A.; Ferreira, J.A.; Pastor, I.; Herranz, J.; Ascasibar, E.

2005-01-01

Under pure Electron Cyclotron Resonance Heating (ECRH) conditions in TJ-II plasmas (P<300 kW, 53.2 GHz, 2nd harmonic X-mode, power density < 25 W/m''3), plasma start-up and good density control are obtained only by the proper combination of wall conditions and gas puffing characteristics. Such a control is particularly critical for the optimisation of the NBI power transfer to the target plasmas. The relatively low cut-off limit is easily reached due not only to the unfavourable wall/puffing-fuelling ratio but also due to the steep density profiles developed during the Enhanced Particle Confinement (EPC) modes. These modes are triggered by the gas puffing waveform, and they cannot be achieved for high iota magnetic configurations in TJ-II. Comparative experiments under metallic and boronised wall conditions have shown that the sensitivity of the EPC modes to the puffing rate is at least partially related to the energy balance at the plasma periphery under central heating scenarios. In this work, the impact of gas-fuelling location on the plasma parameters and density control is described. For that purpose, three different fuelling locations have been investigated; broad distribution from a side ports, localized injection from long tubes at different poloidal positions and highly localized injection through a movable limiter. Edge density and temperature profiles from a broad set of diagnostics (atomic beams, reflectometry, Thompson Scattering ECE, etc...) are analysed and compared. It has been found that preventing from transition to the EPC mode is achieved by using slow puffing rates, while neutral penetration into the plasma core can be enhanced for highly localized gas sources. Wall inventory, however, has been found to pl ay a dominant role in the fuelling of the plasma under most conditions. (author)

Remote device control and monitor system for the LHD deuterium experiments

Energy Technology Data Exchange (ETDEWEB)

Nakanishi, Hideya, E-mail: nakanisi@nifs.ac.jp [National Institute for Fusion Science (NIFS), Toki, Gifu 509-5292 (Japan); Dept. Fusion Science, SOKENDAI (The Graduate University for Advanced Studies), Toki, Gifu 509-5292 (Japan); Ohsuna, Masaki; Ito, Tatsuki; Nonomura, Miki; Imazu, Setsuo; Emoto, Masahiko; Iwata, Chie; Yoshida, Masanobu; Yokota, Mitsuhiro; Maeno, Hiroya; Aoyagi, Miwa; Ogawa, Hideki; Nakamura, Osamu; Morita, Yoshitaka; Inoue, Tomoyuki; Watanabe, Kiyomasa [National Institute for Fusion Science (NIFS), Toki, Gifu 509-5292 (Japan); Ida, Katsumi; Ishiguro, Seiji; Kaneko, Osamu [National Institute for Fusion Science (NIFS), Toki, Gifu 509-5292 (Japan); Dept. Fusion Science, SOKENDAI (The Graduate University for Advanced Studies), Toki, Gifu 509-5292 (Japan)

2016-11-15

Highlights: • Device remote control will be significant for the LHD deuterium experiments. • A central management GUI to control the power distribution for devices. • For safety, power management is separated from operational commanding. • Wi-Fi was tested and found to be not reliable with fusion plasmas. - Abstract: Upon beginning the LHD deuterium experiment, the opportunity for maintenance work in the torus hall will be conspicuously reduced such that all instruments must be controlled remotely. The LHD data acquisition (DAQ) and archiving system have been using about 110 DAQ front-end, and the DAQ central control and monitor system has been implemented for their remote management. This system is based on the “multi-agent” model whose communication protocol has been unified. Since DAQ front-end electronics would suffer from the “single-event effect” (SEE) of D-D neutrons, software-based remote operation might become ineffective, and then securely intercepting or recycling the electrical power of the device would be indispensable for recovering from a non-responding fault condition. In this study, a centralized control and monitor system has been developed for a number of power distribution units (PDUs). This system adopts the plug-in structure in which the plug-in modules can absorb the differences among the commercial products of numerous vendors. The combination of the above-mentioned functionalities has led to realizing the flexible and highly reliable remote control infrastructure for the plasma diagnostics and the device management in LHD.

Taming Instabilities in Plasma Discharges

International Nuclear Information System (INIS)

Klinger, T.; Krahnstover, N. O.; Mausbach, T.; Piel, A.

2000-01-01

Recent experimental work on taming instabilities in plasma discharges is discussed. Instead of suppressing instabilities, it is desired to achieve control over their dynamics, done by perturbing appropriately the current flow in the external circuit of the discharge. Different discrete and continuous feedback as well as open-loop control schemes are applied. Chaotic oscillations in plasma diodes are controlled using the OGY discrete feedback scheme. This is demonstrated both in experiment and computer simulation. Weakly developed ionization wave turbulence is tamed by continuous feedback control. Open-loop control of stochastic fluctuations - stochastic resonance - is demonstrated in a thermionic plasma diode. (author)

Development of high energy pulsed plasma simulator for plasma-lithium trench experiment

Science.gov (United States)

Jung, Soonwook

To simulate detrimental events in a tokamak and provide a test-stand for a liquid lithium infused trench (LiMIT) device, a pulsed plasma source utilizing a theta pinch in conjunction with a coaxial plasma accelerator has been developed. An overall objective of the project is to develop a compact device that can produce 100 MW/m2 to 1 GW/m2 of plasma heat flux (a typical heat flux level in a major fusion device) in ~ 100 mus (≤ 0.1 MJ/m2) for a liquid lithium plasma facing component research. The existing theta pinch device, DEVeX, was built and operated for study on lithium vapor shielding effect. However, a typical plasma energy of 3 - 4 kJ/m2 is too low to study an interaction of plasma and plasma facing components in fusion devices. No or little preionized plasma, ringing of magnetic field, collisions of high energy particles with background gas have been reported as the main issues. Therefore, DEVeX is reconfigured to mitigate these issues. The new device is mainly composed of a plasma gun for a preionization source, a theta pinch for heating, and guiding magnets for a better plasma transportation. Each component will be driven by capacitor banks and controlled by high voltage / current switches. Several diagnostics including triple Langmuir probe, calorimeter, optical emission measurement, Rogowski coil, flux loop, and fast ionization gauge are used to characterize the new device. A coaxial plasma gun is manufactured and installed in the previous theta pinch chamber. The plasma gun is equipped with 500 uF capacitor and a gas puff valve. The increase of the plasma velocity with the plasma gun capacitor voltage is consistent with the theoretical predictions and the velocity is located between the snowplow model and the weak - coupling limit. Plasma energies measured with the calorimeter ranges from 0.02 - 0.065 MJ/m2 and increases with the voltage at the capacitor bank. A cross-check between the plasma energy measured with the calorimeter and the triple probe

Tokamak plasma current disruption infrared control system

International Nuclear Information System (INIS)

Kugel, H.W.; Ulrickson, M.

1987-01-01

This patent describes a device for magnetically confining a plasma driven by a plasma current and contained within a toroidal vacuum chamber, the device having an inner toroidal limiter on an inside wall of the vacuum chamber and an arrangement for the rapid prediction and control in real time of a major plasma disruption. The arrangement is described which includes: scanning means sensitive to infrared radiation emanating from within the vacuum chamber, the infrared radiation indicating the temperature along a vertical profile of the inner toroidal limiter. The scanning means is arranged to observe the infrared radiation and to produce in response thereto an electrical scanning output signal representative of a time scan of temperature along the vertical profile; detection means for analyzing the scanning output signal to detect a first peaked temperature excursion occurring along the profile of the inner toroidal limiter, and to produce a detection output signal in repsonse thereto, the detection output signal indicating a real time prediction of a subsequent major plasma disruption; and plasma current reduction means for reducing the plasma current driving the plasma, in response to the detection output signal and in anticipation of a subsequent major plasma disruption

ICRF Mode Conversion Current Drive for Plasma Stability Control in Tokamaks

International Nuclear Information System (INIS)

Grekov, D.; Kock, R.; Lyssoivan, A.; Noterdaeme, J. M.; Ongena, J.

2007-01-01

There is a substantial incentive for the International Thermonuclear Experimental Reactor (ITER) to operate at the highest attainable beta (plasma pressure normalized to magnetic pressure), a point emphasized by requirements of attractive economics in a reactor. Recent experiments aiming at stationary high beta discharges in tokamak plasmas have shown that maximum achievable beta value is often limited by the onset of instabilities at rational magnetic surfaces (neoclassical tearing modes). So, methods of effective control of these instabilities have to be developed. One possible way for neoclassical tearing modes control is an external current drive in the island to locally replace the missing bootstrap current and thus to suppress the instability. Also, a significant control of the sawtooth behaviour was demonstrated when the magnetic shear was modified by driven current at the magnetic surface where safety factor equals to 1. In the ion cyclotron range of frequencies (ICRF), the mode conversion regime can be used to drive the local external current near the position of the fast-to-slow wave conversion layer, thus providing an efficient means of plasma stability control. The slow wave energy is effectively absorbed in the vicinity of mode conversion layer by electrons with such parallel to confining magnetic field velocities that the Landau resonance condition is satisfied. For parameters of present day tokamaks and for ITER parameters the slow wave phase velocity is rather low, so the large ratio of momentum to energy content would yield high current drive efficiency. In order to achieve noticeable current drive effect, it is necessary to create asymmetry in the ICRF power absorption between top and bottom parts of the plasma minor cross-section. Such asymmetric electron heating may be realized using: - shifted from the torus midplane ICRF antenna in TEXTOR tokamak; - plasma displacement in vertical direction that is feasible in ASDEX-Upgrade; - the

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.)

Automation of Aditya tokamak plasma position control DC power supply

Energy Technology Data Exchange (ETDEWEB)

Arambhadiya, Bharat, E-mail: bharat@ipr.res.in; Raj, Harshita; Tanna, R.L.; Edappala, Praveenlal; Rajpal, Rachana; Ghosh, Joydeep; Chattopadhyay, P.K.; Kalal, M.B.

2016-11-15

Highlights: • Plasma position control is very essential for obtaining repeatable high temperature, high-density discharges of longer durations in tokomak. • The present capacitor bank has limitations of maximum current capacity and position control beyond 200 ms. • The installation of a separate set of coils and a DC power supply can control the plasma position beyond 200 ms. • A high power thyristor (T588N1200) triggers for DC current pulse of 300 A fires precisely at required positions to modify plasma position. • The commissioning is done for the automated in-house, quick and reliable solution. - Abstract: Plasma position control is essential for obtaining repeatable high temperature, high-density discharges of longer duration in tokamaks. Recently, a set of external coils is installed in the vertical field mode configuration to control the radial plasma position in ADITYA tokamak. The existing capacitor bank cannot provide the required current pulse beyond 200 ms for position control. This motivated to have a DC power supply of 500 A to provide current pulse beyond 200 ms for the position control. The automatization of the DC power supply mandated interfaces with the plasma control system, Aditya Pulse Power supply, and Data acquisition system for coordinated discharge operation. A high current thyristor circuit and a timer circuit have been developed for controlling the power supply automatically for charging vertical field coils of Aditya tokamak. Key protection interlocks implemented in the development ensure machine and occupational safety. Fiber-optic trans-receiver isolates the power supply with other subsystems, while analog channel is optically isolated. Commissioning and testing established proper synchronization of the power supply with tokamak operation. The paper discusses the automation of the DC power supply with main circuit components, timing control, and testing results.

Plasma physics and controlled fusion research during half a century

Energy Technology Data Exchange (ETDEWEB)

Lehnert, Bo

2001-06-01

A review is given on the historical development of research on plasma physics and controlled fusion. The potentialities are outlined for fusion of light atomic nuclei, with respect to the available energy resources and the environmental properties. Various approaches in the research on controlled fusion are further described, as well as the present state of investigation and future perspectives, being based on the use of a hot plasma in a fusion reactor. Special reference is given to the part of this work which has been conducted in Sweden, merely to identify its place within the general historical development. Considerable progress has been made in fusion research during the last decades. Temperatures above the limit for ignition of self-sustained fusion reactions, i.e. at more than hundred million degrees, have been reached in large experiments and under conditions where the fusion power generation is comparable to the power losses. An energy producing fusion reactor could in principle be realized already today, but it would not become technically and economically efficient when being based on the present state of art. Future international research has therefore to be conducted along broad lines, with necessary ingredients of basic investigations and new ideas.

Plasma physics and controlled fusion research during half a century

International Nuclear Information System (INIS)

Lehnert, Bo

2001-06-01

A review is given on the historical development of research on plasma physics and controlled fusion. The potentialities are outlined for fusion of light atomic nuclei, with respect to the available energy resources and the environmental properties. Various approaches in the research on controlled fusion are further described, as well as the present state of investigation and future perspectives, being based on the use of a hot plasma in a fusion reactor. Special reference is given to the part of this work which has been conducted in Sweden, merely to identify its place within the general historical development. Considerable progress has been made in fusion research during the last decades. Temperatures above the limit for ignition of self-sustained fusion reactions, i.e. at more than hundred million degrees, have been reached in large experiments and under conditions where the fusion power generation is comparable to the power losses. An energy producing fusion reactor could in principle be realized already today, but it would not become technically and economically efficient when being based on the present state of art. Future international research has therefore to be conducted along broad lines, with necessary ingredients of basic investigations and new ideas

Development of a flight simulator for the control of plasma discharges

International Nuclear Information System (INIS)

Ravenel, N.; Artaud, J.F.; Bremond, S.; Guillerminet, B.; Huynh, P.; Moreau, P.; Signoret, J.

2010-01-01

The feedback control of fusion experiments in tokamak devices is entering a new area driven by the increase of control requirements for obtaining burning plasmas under safe operation conditions. A project aiming at setting up a flight simulator for the development of advanced controllers has started last year at CEA. This simulator will reuse most of the components of the European Integrated Tokamak Modelling (ITM) simulation platform. Thus, it will benefit from the development made by the task force and it will be able to offer a development platform for the new controllers of present day European tokamaks and future machines. This paper provides an overview of the architecture of the simulator. The functional specifications of the simulator have been defined and the needs in interface implementation are analysed as well.

EURATOM-CEA association contributions to the 26. EPS conference on controlled fusion and plasma physics, Maastricht

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-10-15

This report references the EURATOM-CEA association contributions presented at the 26. EPS conference on controlled fusion and plasma physics, in Maastricht (Netherlands) the 14-18 June 1999. Two invited papers and 24 contributed papers are proposed. They deal with: tokamak devices; particle recirculation in ergodic divertor; current profile control and MHD stability in Tore Supra discharges; edge-plasma control by the ergodic divertor; electron heat transport in stochastic magnetic layer; bolometry and radiated power; particle collection by ergodic divertor; study and simulation of pa impurities; line shape modelling for plasma edge conditions; dynamical study of the radial structure of the fluctuations measured by reciprocating Langmuir probe in Tore Supra; up-down asymmetry of density fluctuations; Halo currents in a circular tokamak; real time measurement of the position, density, profile and current profile at Tore Supra; poloidal rotation measurement by reflectometry; interpretation of q-profile dependence of the LH power deposition profile during LHCD experiments; ICFR plasma production and optimization; improved core electron confinement; measurement of hard X-ray emission profile; modelling of shear effects on thermal and particles transport; ion turbulence; current drive generation based on autoresonance and intermittent trapping mechanisms. (A.L.B.)

First laser-plasma interaction and hohlraum experiments on the National Ignition Facility

Energy Technology Data Exchange (ETDEWEB)

Dewald, E L; Glenzer, S H; Landen, O L; Suter, L J; Jones, O S; Schein, J; Froula, D; Divol, L; Campbell, K; Schneider, M S; Holder, J; McDonald, J W; Niemann, C; Mackinnon, A J; Hammel, B A [Lawrence Livermore National Laboratory, PO Box 808, Livermore, CA 94550 (United States)

2005-12-15

Recently the first laser-plasma interaction and hohlraum experiments have been performed at the National Ignition Facility (NIF) in support of indirect drive inertial confinement fusion designs. The effects of laser beam smoothing by spectral dispersion and polarization smoothing on the intense (2 x 10{sup 15} W cm{sup -2}) beam propagation in gas-filled tubes has been studied at up to 7 mm plasma scales as found in indirect drive gas filled ignition hohlraum designs. These experiments have shown the expected full propagation without filamentation and beam break up when using full laser smoothing. In addition, vacuum hohlraums have been irradiated with laser powers up to 6 TW, 1-9 ns pulse lengths and energies up to 17 kJ to activate several diagnostics, to study the hohlraum radiation temperature scaling with the laser power and hohlraum size, and to make contact with hohlraum experiments performed at the Nova and Omega laser facilities. Subsequently, novel long laser pulse hohlraum experiments have tested models of hohlraum plasma filling and long pulse hohlraum radiation production. The validity of the plasma filling assessment using in analytical models and radiation hydrodynamics calculations with the code LASNEX has been proven in these studies. The comparison of these results with modelling will be discussed.

From the conceptual design to the first simulation of the new WEST plasma control system

Energy Technology Data Exchange (ETDEWEB)

Nouailletas, R., E-mail: remy.nouailletas@cea.fr [IRFM, CEA, F-13108 Saint Paul lez Durance (France); Ravenel, N.; Signoret, J. [IRFM, CEA, F-13108 Saint Paul lez Durance (France); Treutterer, W. [Max Planck Institute for Plasma Physics, Boltzmannstr. 2, 85748 Garching (Germany); Spring, A.; Lewerentz, M. [Max Planck Institute for Plasma Physics, Wendeksteinstr. 1, 17491 Greifswald (Germany); Rapson, C.J. [Max Planck Institute for Plasma Physics, Boltzmannstr. 2, 85748 Garching (Germany); Masand, H.; Dhongde, J. [Institute for Plasma Research (IPR), Near Indira Bridge, Bhat, Gandhinagar 382 428, Gujarat (India); Moreau, P.; Guillerminet, B.; Brémond, S.; Allegretti, L. [IRFM, CEA, F-13108 Saint Paul lez Durance (France); Raupp, G. [Max Planck Institute for Plasma Physics, Boltzmannstr. 2, 85748 Garching (Germany); Werner, A. [Max Planck Institute for Plasma Physics, Wendeksteinstr. 1, 17491 Greifswald (Germany); Saint Laurent, F.; Nardon, E. [IRFM, CEA, F-13108 Saint Paul lez Durance (France); Bhandarkar, M. [Institute for Plasma Research (IPR), Near Indira Bridge, Bhat, Gandhinagar 382 428, Gujarat (India)

2015-10-15

Highlights: • We propose an overview of the future control system of the Tore Supra in WEST configuration. • The control system will be based on DCS (Discharge Control System) of ASDEX Upgrade. • The Pulse Schedule Editor will be based on the experiment program editor of the future W7X facility. • The operation of this new system is illustrated by an example based on a simple plasma current/loop voltage control. - Abstract: The configuration of the Tore Supra WEST project leads to control challenges and event handling close to those of ITER from a plasma scenario point of view (X-point configuration, H mode, long duration pulse) and from a machine protection point of view (metallic environment). Based on previous conceptual studies and to meet the WEST requirements, a sub-project will implement a new plasma control system (PCS) and a new pulse schedule editor (PSE). The main idea is to use a segment approach to describe the pulse scheduling with a full integration of event handling both on the PCS and on the PSE. After detailed specification work, it has been shown that the real-time framework called DCS (Discharge Control System) which is currently used on ASDEX upgrade fulfills the requirements and could be integrated into the WEST global control infrastructure. For the PSE, the Xedit tool, developed for the future W7X facility, has been chosen. This contribution will begin by a short explanation of the concepts proposed for the control of the plasma and the handling of events during the plasma discharge. Then it will focus on the new centralized architecture of the new Tore Supra PCS and an operating principle example showing the efficiency of the approach to handle normal and off-normal events. This later point will illustrate the required modifications of DCS and Xedit to fit with the Tore Supra Control infrastructure.

A Proton-Driven Plasma Wakefield Acceleration experiment at CERN

CERN Multimedia

The AWAKE Collaboration has been formed in order to demonstrate protondriven plasma wakefield acceleration for the first time. This technology could lead to future colliders of high energy but of a much reduced length compared to proposed linear accelerators. The SPS proton beam in the CNGS facility will be injected into a 10m plasma cell where the long proton bunches will be modulated into significantly shorter micro-bunches. These micro-bunches will then initiate a strong wakefield in the plasma with peak fields above 1 GV/m that will be harnessed to accelerate a bunch of electrons from about 20MeV to the GeV scale within a few meters. The experimental program is based on detailed numerical simulations of beam and plasma interactions. The main accelerator components, the experimental area and infrastructure required as well as the plasma cell and the diagnostic equipment are discussed in detail. First protons to the experiment are expected at the end of 2016 and this will be followed by an initial 3–4 yea...

Feedback control and stabilization experiments on the Texas Experimental Tokamak (TEXT)

International Nuclear Information System (INIS)

Uckan, T.; Richards, B.; Wootton, A.J.; Bengtson, R.D.; Bravenec, R.; Carreras, B.A.; Li, G.X.; Hurwitz, P.; Phillips, P.E.; Rowan, W.L.; Tsui, H.Y.W.; Uglum, J.R.; Wen, Y.; Winslow, D.

1995-01-01

Plasma edge feedback experiments on the Texas Experimental Tokamak (TEXT) have been successful in controlling the edge plasma potential fluctuation level. The feedback wave-launcher is driven by the local edge potential fluctuations. The edge potential fluctuations are modified in a broad frequency band. Moreover, the potential fluctuations can be reduced (≤100 kHz) without enhancing other modes, or excited (10 to 12 kHz), depending on the phase difference between the driver and the launcher signal, and gain of the system. This turbulence modification is achieved not only locally but also halfway around the torus and has about 2 cm of poloidal extent. The local plasma parameters at the edge and the estimated fluctuation-induced radial particle flux are somewhat affected by the edge feedback. ((orig.))

Power supply controlled for plasma torch generation; Fuente de alimentacion controlada para la generacion de un plasma

Energy Technology Data Exchange (ETDEWEB)

Diaz Z, S

1997-12-31

The high density of energy furnished by thermal plasma is profited in a wide range of applications, such as those related with welding fusion, spray coating and at the present in waste destruction. The waste destruction by plasma is a very attractive process because the remaining products are formed by inert glassy grains and non-toxic gases. The main characteristics of thermal plasmas are presented in this work. Techniques based on power electronics are utilized to achieve a good performance in thermal plasma generation. This work shown the design and construction of three phase control system for electric supply of thermal plasma torch, with 250 kw of capacity, as a part of the project named `Destruction of hazard wastes by thermal plasma` actually working in the Instituto Nacional de Investigaciones Nucleares (ININ). The characteristics of thermal plasma and its generation are treated in the first chapter. The A C controllers by thyristors applied in three phase arrays are described in the chapter II, talking into account the power transformer, rectifiers bank and aliasing coil. The chapter III is dedicated in the design of the trigger module which controls the plasma current by varying the trigger angle of the SCR`s; the protection and isolating unit are also presented in this chapter. The results and conclusions are discussed in chapter IV. (Author).

Langmuir probe-based observables for plasma-turbulence code validation and application to the TORPEX basic plasma physics experiment

International Nuclear Information System (INIS)

Ricci, Paolo; Theiler, C.; Fasoli, A.; Furno, I.; Labit, B.; Mueller, S. H.; Podesta, M.; Poli, F. M.

2009-01-01

The methodology for plasma-turbulence code validation is discussed, with focus on the quantities to use for the simulation-experiment comparison, i.e., the validation observables, and application to the TORPEX basic plasma physics experiment [A. Fasoli et al., Phys. Plasmas 13, 055902 (2006)]. The considered validation observables are deduced from Langmuir probe measurements and are ordered into a primacy hierarchy, according to the number of model assumptions and to the combinations of measurements needed to form each of them. The lowest levels of the primacy hierarchy correspond to observables that require the lowest number of model assumptions and measurement combinations, such as the statistical and spectral properties of the ion saturation current time trace, while at the highest levels, quantities such as particle transport are considered. The comparison of the observables at the lowest levels in the hierarchy is more stringent than at the highest levels. Examples of the use of the proposed observables are applied to a specific TORPEX plasma configuration characterized by interchange-driven turbulence.

Plasma focus system: Design, construction and experiments

International Nuclear Information System (INIS)

Alacakir, A.; Akguen, Y.; Boeluekdemir, A. S.

2007-01-01

The aim of this work is to construct a compact experimental system for fusion research. The design, construction and experiments of the 3 kJ Mather type plasma focus machine is described. This machine is established for neutron yield and fast neutron radiography by D-D reaction which is given by D + D→ 3 He (0.82 MeV) + n (2.45 MeV) . Investigation of the geometry of plasma focus machine in the presence of high voltage drive and vacuum system setup is shown. 108 neutron per pulse and 200 kA peak current is obtained for many shots. Scintillator screen for fast neutron imaging, sensitive to 2.45 MeV neutrons, is also manufactured in our labs. Structural neutron shielding computations for safety is also completed

Feedback control of plasma equilibrium with control system aided by personal computer on the JIPP T-IIU tokamak

International Nuclear Information System (INIS)

Tsuzuki, T.; Toi, K.; Matsuura, K.

1991-04-01

A feedback control system aided by a personal computer is developed to maintain plasma position on the required position in the JIPP T-IIU tokamak. The personal computer enables to adjust various control parameters easily. In this control system, a control demand for driving the power supply of feedback controlled vertical field coils is composed to be proportional to a total plasma current. This system has been successfully employed throughout the discharge where the plasma current substantially changes from zero to hundreds of kiloamperes, because the feedback control can be done, being independent of the plasma current. The analysis of this feedback control system taken into account of digital sampling agrees well with the experimental results. (author)

Control of supersonic axisymmetric base flows using passive splitter plates and pulsed plasma actuators

Science.gov (United States)

Reedy, Todd Mitchell

An experimental investigation evaluating the effects of flow control on the near-wake downstream of a blunt-based axisymmetric body in supersonic flow has been conducted. To better understand and control the physical phenomena that govern these massively separated high-speed flows, this research examined both passive and active flow-control methodologies designed to alter the stability characteristics and structure of the near-wake. The passive control investigation consisted of inserting splitter plates into the recirculation region. The active control technique utilized energy deposition from multiple electric-arc plasma discharges placed around the base. The flow-control authority of both methodologies was evaluated with experimental diagnostics including particle image velocimetry, schlieren photography, surface flow visualization, pressure-sensitive paint, and discrete surface pressure measurements. Using a blowdown-type wind tunnel reconstructed specifically for these studies, baseline axisymmetric experiments without control were conducted for a nominal approach Mach number of 2.5. In addition to traditional base pressure measurements, mean velocity and turbulence quantities were acquired using two-component, planar particle image velocimetry. As a result, substantial insight was gained regarding the time-averaged and instantaneous near-wake flow fields. This dataset will supplement the previous benchmark point-wise laser Doppler velocimetry data of Herrin and Dutton (1994) for comparison with new computational predictive techniques. Next, experiments were conducted to study the effects of passive triangular splitter plates placed in the recirculation region behind a blunt-based axisymmetric body. By dividing the near-wake into 1/2, 1/3, and 1/4 cylindrical regions, the time-averaged base pressure distribution, time-series pressure fluctuations, and presumably the stability characteristics were altered. While the spatial base pressure distribution was

Recent Progress on the magnetic turbulence experiment at the Bryn Mawr Plasma Laboratory

Science.gov (United States)

Schaffner, D. A.; Cartagena-Sanchez, C. A.; Johnson, H. K.; Fahim, L. E.; Fiedler-Kawaguchi, C.; Douglas-Mann, E.

2017-10-01

Recent progress is reported on the construction, implementation and testing of the magnetic turbulence experiment at the Bryn Mawr Plasma Laboratory (BMPL). The experiment at the BMPL consists of an ( 300 μs) long coaxial plasma gun discharge that injects magnetic helicity into a flux-conserving chamber in a process akin to sustained slow-formation of spheromaks. A 24cm by 2m cylindrical chamber has been constructed with a high density axial port array to enable detailed simultaneous spatial measurements of magnetic and plasma fluctuations. Careful positioning of the magnetic structure produced by the three separately pulsed coils (one internal, two external) are preformed to optimize for continuous injection of turbulent magnetized plasma. High frequency calibration of magnetic probes is also underway using a power amplifier.

Long conduction time plasma opening switch experiments at Sandia National Laboratories

International Nuclear Information System (INIS)

Savage, M.E.; Simpson, W.W.; Cooper, G.W.; Usher, M.A.

1993-01-01

Sandia National Laboratories has undertaken an ambitious program to reduce the size and cost of large pulsed power drivers. The program basis is inductive energy storage and Plasma Opening Switches (POS). Inductive energy storage has well known advantages, including increased efficiency and reduced stress on the vacuum interface. The Sandia approach is to retain the reliable and efficient Marx generator and the temporal pulse compression of the water dielectric capacitor. A triggered closing switch, developed at Sandia, transfers the capacitor charge into the energy storage inductor. This approach has several advantages, including relaxed requirements on Marx jitter and inductance, and much faster current risetime in the energy storage inductor. The POS itself is the key to the Sandia program. The switch design uses an auxiliary magnetic field to inject the plasma and hold it in place during conduction. After opening begins, the self magnetic field of the power pulse pushes on the plasma to increase the opened gap. The authors use magnetic pressure because they desire POS gaps of several cm. Typical plasma opening switches do not achieve large gaps. Improved opening allows more efficient transfer to loads. They present results from recent experiments at Sandia. Their driver presently supplies 650 kA with a 240 ns risetime to the input of the POS. The storage inductor is a 17 Ohm magnetically insulated transmission line (MITL) that is five meters long. They discuss the ways in which magnetic field influences the POS, and the ways in which they control the magnetic fields

Numerical modeling of the plasma ring acceleration experiment

International Nuclear Information System (INIS)

Eddleman, J.L.; Hammer, J.H.; Hartman, C.W.

1987-01-01

Modeling of the LLNL RACE experiment and its many applications has necessitated the development and use of a wide array of computational tools. The two-dimensional MHD code, HAM, has been used to model the formation of a compact torus plasma ring in a magnetized coaxial gun and its subsequent acceleration by an additional applied toroidal field. Features included in the 2-D calculations are self-consistent models for (1) the time-dependent poloidal field produced by a capacitor bank discharge through a solenoid field coil (located either inside the gun inner electrode or outside the outer gun electrode) and the associated diffusion of magnetic flux through neighboring conductors, (2) gas flow into the gun annular region from a simulated puffed gas valve plenum, (3) formation and motion of a current sheet produced by J x B forces resulting from discharge of the gun capacitor bank through the plasma load between the coaxial gun electrodes, (4) the subsequent stretching and reconnection of the poloidal field lines to form a compact torus plasma ring, and (5) finally the discharge of the accelerator capacitor bank producing an additional toroidal field for acceleration of the plasma ring. The code has been extended to include various models for gas breakdown, plasma anomalous resistivity, and mass entrainment from ablation of electrode material

Update on the DIII-D ECH system: experiments, gyrotrons, advanced diagnostics, and controls

Directory of Open Access Journals (Sweden)

Lohr John

2017-01-01

Full Text Available The ECH system on DIII-D is continuing to be upgraded, while simultaneously being operated nearly daily for plasma experiments. The latest major hardware addition is a new 117.5 GHz gyrotron, which generated 1.7 MW for short pulses during factory testing. A new gyrotron control system based on Field Programmable Gate Array (FPGA technology with very high speed system data acquisition has significantly increased the flexibility and reliability of individual gyrotron operation. We have improved the performance of the fast mirror scanning, both by increasing the scan speeds and by adding new algorithms for controlling the aiming using commands generated by the Plasma Control System (PCS. The system is used for transport studies, ELM control, current profile control, non-inductive current generation, suppression of MHD modes, startup assist, plasma density control, and other applications. A program of protective measures, which has been in place for more than two years, has eliminated damage to hardware and diagnostics caused by overdense operation. Other activities not directly related to fusion research have used the ECH system to test components, study methods for improving production of semiconductor junctions and materials, and test the feasibility of using ground based microwave systems to power satellites into orbit.

Update on the DIII-D ECH system: experiments, gyrotrons, advanced diagnostics, and controls

Science.gov (United States)

Lohr, John; Brambila, Rigoberto; Cengher, Mirela; Gorelov, Yuri; Grosnickle, William; Moeller, Charles; Ponce, Dan; Torrezan, Antonio; Ives, Lawrence; Reed, Michael; Blank, Monica; Felch, Kevin; Parisuaña, Claudia; LeViness, Alexandra

2017-08-01

The ECH system on DIII-D is continuing to be upgraded, while simultaneously being operated nearly daily for plasma experiments. The latest major hardware addition is a new 117.5 GHz gyrotron, which generated 1.7 MW for short pulses during factory testing. A new gyrotron control system based on Field Programmable Gate Array (FPGA) technology with very high speed system data acquisition has significantly increased the flexibility and reliability of individual gyrotron operation. We have improved the performance of the fast mirror scanning, both by increasing the scan speeds and by adding new algorithms for controlling the aiming using commands generated by the Plasma Control System (PCS). The system is used for transport studies, ELM control, current profile control, non-inductive current generation, suppression of MHD modes, startup assist, plasma density control, and other applications. A program of protective measures, which has been in place for more than two years, has eliminated damage to hardware and diagnostics caused by overdense operation. Other activities not directly related to fusion research have used the ECH system to test components, study methods for improving production of semiconductor junctions and materials, and test the feasibility of using ground based microwave systems to power satellites into orbit.

High speed photography diagnostics in laser-plasma interaction experiments

International Nuclear Information System (INIS)

Andre, M.L.

1988-01-01

The authors report on their effort in the development of techniques involved in laser-plasma experiments. This includes not only laser technology but also diagnostics studies and targets design and fabrication. Among the different kind of diagnostics currently used are high speed streak cameras, fast oscilloscopes and detectors sensitive in the i.r., visible, the u.v. region and the x-rays. In this presentation the authors describe the three high power lasers which are still in operation (P 102, OctAL and PHEBUS) and the main diagnostics used to characterize the plasma

A review of low density porous materials used in laser plasma experiments

Science.gov (United States)

Nagai, Keiji; Musgrave, Christopher S. A.; Nazarov, Wigen

2018-03-01

This review describes and categorizes the synthesis and properties of low density porous materials, which are commonly referred to as foams and are utilized for laser plasma experiments. By focusing a high-power laser on a small target composed of these materials, high energy and density states can be produced. In the past decade or so, various new target fabrication techniques have been developed by many laboratories that use high energy lasers and consequently, many publications and reviews followed these developments. However, the emphasis so far has been on targets that did not utilize low density porous materials. This review therefore, attempts to redress this balance and endeavors to review low density materials used in laser plasma experiments in recent years. The emphasis of this review will be on aspects of low density materials that are of relevance to high energy laser plasma experiments. Aspects of low density materials such as densities, elemental compositions, macroscopic structures, nanostructures, and characterization of these materials will be covered. Also, there will be a brief mention of how these aspects affect the results in laser plasma experiments and the constrictions that these requirements put on the fabrication of low density materials relevant to this field. This review is written from the chemists' point of view to aid physicists and the new comers to this field.

Plasma shape experiments for an optimized tokamak

International Nuclear Information System (INIS)

Hyatt, A.W.; Osborne, T.H.; Lazarus, E.A.

1994-07-01

In this paper we present results from recent experiments at DIII-D which measured the plasma stability and confinement performance product, βτ E , in one previously studied and three new plasma shapes. One important goal of these experiments was to identify performance vs shape trends which would identify a shape compatible with both high performance and the planned effort to decrease the power flux to the divertor floor using a closed ''slot'' divertor geometry. power flux to the divertor floor using a closed ''slot'' divertor geometry. The closed divertor hardware must be designed for a reduced set of plasma shapes, so care must be taken to choose the shape that optimizes βτ E and divertor performance. The four shapes studied form a matrix of moderate and high elongations (κ congruent 1.8 and 2.1) and low and high triangularities (δ congruent 0.3 and 0.9). All configurations were double-null diverted (DND), held fixed during a shot, with neutral beam heating. The shapes span a range of X-point locations compatible with the envisioned closed divertor. We find that from shape to shape, a shot's transient normalized performance, β N H, where β N ≡ β/(I p )/aB T and H ≡ τ E /τ E ITER-89P , increases strongly with triangularity, but depends only weakly on elongation. However, the normalized performance during quasi stationary ELMing H-mode, to which these discharges eventually relax, is insensitive to both triangularity and elongation. The moderate elongation, high triangularity DND shape is shown to be near optimum for future studies on DIII-D

Plasma shape experiments for an optimized tokamak

Energy Technology Data Exchange (ETDEWEB)

Hyatt, A.W.; Osborne, T.H. [General Atomics, San Diego, CA (United States); Lazarus, E.A. [Oak Ridge National Lab., TN (United States)

1994-12-31

In this paper we present results from recent experiments at DIII-D which measured the plasma stability and confinement performance product, {beta}{sub {tau}E}, in one previously studied and three new plasma shapes. One important goal of these experiments was to identify performance vs shape trends which would identify a shape compatible with both high performance and the planned effort to decrease the power flux to the divertor floor using a closed `slot` divertor geometry. The closed divertor hardware must be designed for a reduced set of plasma shapes, so care must be taken to choose the shape that optimizes {beta}{sub {tau}E} and divertor performance. The four shapes studied form a matrix of moderate and high elongations ({kappa} {approx_equal} 1.8 and 2.1) and low and high triangularities ({delta} {approx_equal} 0.3 and 0.9). All configurations were double-null diverted (DND), held fixed during a shot, with neutral beam heating. The shapes span a range of X-point locations compatible with the envisioned closed divertor. We find that from shape to shape, a shot`s transient normalized performance, {beta}{sub N}H, where {beta}{sub N} = {beta}/(I{sub p}/aB{sub T}) and H = {tau}{sub E}/{tau}{sub E}{sup ITER-89P}, increases strongly with triangularity, but depends only weakly on elongation. However, the normalized performance during quasi stationary ELMing H-mode, to which these discharges eventually relax, is insensitive to both triangularity and elongation. The moderate elongation, high triangularity DND shape is shown to be near optimum for future studies on DIII-D. (author) 7 refs., 7 figs.

Experiments and Simulations of ITER-like Plasmas in Alcator C-Mod

International Nuclear Information System (INIS)

Wilson, R.; Kessel, C.E.; Wolfe, S.; Hutchinson, I.H.; Bonoli, P.; Fiore, C.; Hubbard, A.E.; Hughes, J.; Lin, Y.; Ma, Y.; Mikkelsen, D.; Reinke, M.; Scott, S.; Sips, A.C.C.; Wukitch, S.

2010-01-01

Alcator C-Mod is performing ITER-like experiments to benchmark and verify projections to 15 MA ELMy H-mode Inductive ITER discharges. The main focus has been on the transient ramp phases. The plasma current in C-Mod is 1.3 MA and toroidal field is 5.4 T. Both Ohmic and ion cyclotron (ICRF) heated discharges are examined. Plasma current rampup experiments have demonstrated that (ICRF and LH) heating in the rise phase can save voltseconds (V-s), as was predicted for ITER by simulations, but showed that the ICRF had no effect on the current profile versus Ohmic discharges. Rampdown experiments show an overcurrent in the Ohmic coil (OH) at the H to L transition, which can be mitigated by remaining in H-mode into the rampdown. Experiments have shown that when the EDA H-mode is preserved well into the rampdown phase, the density and temperature pedestal heights decrease during the plasma current rampdown. Simulations of the full C-Mod discharges have been done with the Tokamak Simulation Code (TSC) and the Coppi-Tang energy transport model is used with modified settings to provide the best fit to the experimental electron temperature profile. Other transport models have been examined also.

DIII-D Integrated plasma control solutions for ITER and next-generation tokamaks

International Nuclear Information System (INIS)

Humphreys, D.A.; Ferron, J.R.; Hyatt, A.W.; La Haye, R.J.; Leuer, J.A.; Penaflor, B.G.; Walker, M.L.; Welander, A.S.; In, Y.

2008-01-01

Plasma control design approaches and solutions developed at DIII-D to address its control-intensive advanced tokamak (AT) mission are applicable to many problems facing ITER and other next-generation devices. A systematic approach to algorithm design, termed 'integrated plasma control,' enables new tokamak controllers to be applied operationally with minimal machine time required for tuning. Such high confidence plasma control algorithms are designed using relatively simple ('control-level') models validated against experimental response data and are verified in simulation prior to operational use. A key element of DIII-D integrated plasma control, also required in the ITER baseline control approach, is the ability to verify both controller performance and implementation by running simulations that connect directly to the actual plasma control system (PCS) that is used to operate the tokamak itself. The DIII-D PCS comprises a powerful and flexible C-based realtime code and programming infrastructure, as well as an arbitrarily scalable hardware and realtime network architecture. This software infrastructure provides a general platform for implementation and verification of realtime algorithms with arbitrary complexity, limited only by speed of execution requirements. We present a complete suite of tools (known collectively as TokSys) supporting the integrated plasma control design process, along with recent examples of control algorithms designed for the DIII-D PCS. The use of validated physics-based models and a systematic model-based design and verification process enables these control solutions to be directly applied to ITER and other next-generation tokamaks

Numerical Experiments Providing New Insights into Plasma Focus Fusion Devices

Directory of Open Access Journals (Sweden)

Sing Lee

2010-04-01

Full Text Available Recent extensive and systematic numerical experiments have uncovered new insights into plasma focus fusion devices including the following: (1 a plasma current limitation effect, as device static inductance is reduced towards very small values; (2 scaling laws of neutron yield and soft x-ray yield as functions of storage energies and currents; (3 a global scaling law for neutron yield as a function of storage energy combining experimental and numerical data showing that scaling deterioration has probably been interpreted as neutron ‘saturation’; and (4 a fundamental cause of neutron ‘saturation’. The ground-breaking insights thus gained may completely change the directions of plasma focus fusion research.

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

Hypervelocity dust particle impacts observed by the Giotto Magnetometer and Plasma Experiments

OpenAIRE

Neubauer, F. M.; Glassmeier, K. H.; Coates, A. J.; Goldstein, R.; Acuña, M. H.; Musmann, G.

1990-01-01

We report thirteen very short events in the magnetic field of the inner magnetic pile‐up region of comet Halley observed by the Giotto magnetometer experiment together with simultaneous plasma data obtained by the Johnstone plasma analyzer and the ion mass spectrometer experiments. The events are due to dust impacts in the milligram range on the spacecraft at the relative velocity between the cemetery dust and the spacecraft of 68 km/sec. They are generally consistent with dust impact events ...

Strategies for the plasma position and shape control in IGNITOR