67 Matching Results

Search Results

Advanced search parameters have been applied.

Poloidal pressure gradients, divertor detachment and marfes

Description: Because the radiation power density from a marfe scales approximately as the square of its plasma pressure, and since increased radiation would aid divertor detachment for high power tokamaks, this paper identifies regions that might permit locally increased plasma pressure in steady state. The magnetic and dynamic (flow) constraints of magneto-hydrodynamics (MHD) are examined for self-consistent locally increased pressure equilibria, in both the magnetically open tokamak scrape-off layer (SOL) and the closed surfaces just inside the last closed flux surface. In most tokamak geometries it is difficult to recycle particles at a sufficient rate to sustain high pressure marfes, but they might be possible near a divertor X-point.
Date: November 1, 1997
Creator: Schaffer, M.J.
Partner: UNT Libraries Government Documents Department

Helical-D pinch

Description: A stabilized pinch configuration is described, consisting of a D-shaped plasma cross section wrapped tightly around a guiding axis. The {open_quotes}helical-D{close_quotes} geometry produces a very large axial (toroidal) transform of magnetic line direction that reverses the pitch of the magnetic lines without the need of azimuthal (poloidal) plasma current. Thus, there is no need of a {open_quotes}dynamo{close_quotes} process and its associated fluctuations. The resulting configuration has the high magnetic shear and pitch reversal of the reversed field pinch (RFP). (Pitch = P = qR, where R = major radius). A helical-D pinch might demonstrate good confinement at q << 1.
Date: August 1, 1997
Creator: Schaffer, M.J.
Partner: UNT Libraries Government Documents Department

Engineering features of ISX

Description: ISX, an Impurity Study Experiment, is presently being designed at Oak Ridge National Laboratory as a joint scientific effort between ORNL and General Atomic Company. ISX is a moderate size tokamak dedicated to the study of impurity production, diffusion, and control. The significant engineering features of this device are discussed. (auth)
Date: January 1, 1975
Creator: Lousteau, D.C.; Jernigan, T.C.; Schaffer, M.J. & Hussung, R.O.
Partner: UNT Libraries Government Documents Department

Plasma flow in the DIII-D divertor

Description: Indications that flows in the divertor can exhibit complex behavior have been obtained from 2-D modeling but so far remain mostly unconfirmed by experiment. An important feature of flow physics is that of flow reversal. Flow reversal has been predicted analytically and it is expected when the ionization source arising from neutral or impurity ionization in the divertor region is large, creating a high pressure zone. Plasma flows arise to equilibrate the pressure. A radiative divertor regime has been proposed in order to reduce the heat and particle fluxes to the divertor target plates. In this regime, the energy and momentum of the plasma are dissipated into neutral gas introduced in the divertor region, cooling the plasma by collisional, radiative and other atomic processes so that the plasma becomes detached from the target plates. These regimes have been the subject of extensive studies in DIII-D to evaluate their energy and particle transport properties, but only recently it has been proposed that the energy transport over large regions of the divertor must be dominated by convection instead of conduction. It is therefore important to understand the role of the plasma conditions and geometry on determining the region of convection-dominated plasma in order to properly control the heat and particle fluxes to the target plates and hence, divertor performance. The authors have observed complex structures in the deuterium ion flows in the DIII-D divertor. Features observed include reverse flow, convective flow over a large volume of the divertor and stagnant flow. They have measured large gradients in the plasma potential across the separatrix in the divertor and determined that these gradients induce poloidal flows that can potentially affect the particle balance in the divertor.
Date: July 1998
Creator: Boedo, J. A.; Porter, G. D. & Schaffer, M. J.
Partner: UNT Libraries Government Documents Department

The electrical insulation of the DIII-D advanced divertor electrode

Description: The electrode for biasing experiments on the DIII-D tokamak was installed in the summer of 1990 and biasing experiments have shown positive results. For the electrode, electrical insulation had to provide voltage standoff in the DIII-D divertor environment of neutral pressures in the range of 10{sup {minus}8} to 5 {times} 10{sup {minus}2} torr, variable magnetic fields, and in the presence of ionizing radiation. The electrical insulation system was designed and tested in air and vacuum for voltages up to 3 kV. In this paper, we provide an update on our operating experience, problems encountered, and improvements to the system. Electrical breakdown of some components has occurred during tokamak operations and transient voltages, up to 5 kV, have been observed. The original concept for insulating the water and electrical feeds for the electrode, a thin layer of woven ceramic cloth insulation between the feeds and a ground plane to keep out stray plasma, was found to be prone to failure. A new scheme of rigid ceramic insulators surrounded by a ground plane was designed and is being implemented. Another problem was arcs from vessel potential surfaces to the electrode in several locations where vessel ground existed within 1 cm of the electrode. The arc traveled in a small crack between two insulators. Careful attention has been paid to closing this and other small gaps in the insulation. Coatings on the surface of plasma facing insulators have been found to be electrically conductive. Grooves are being machined into the insulators to give areas shadowed from the coating source. Tests are being done to demonstrate the design concepts in both vacuum and glow discharge environments. Plasma sprayed ceramic coatings were also tested to determine the voltage standoff capability in a glow plasma discharge. The results of these tests will be discussed. 2 refs., ...
Date: October 1, 1991
Creator: Smith, J.P.; Schaffer, M.J. & Hyatt, A.W.
Partner: UNT Libraries Government Documents Department

Bias-sustained shield plasma

Description: Divertor biasing may provide a method for density and impurity control by enhancing the shielding efficiency of the scrape-off layer. The idea is to make the scrape-off plasma denser and thicker by heating it with a bias-driven current, and by inducing a radial E [times] B drift. If the bias is applied to flux surfaces at the outer edge of the usual scrape-off layer, a new layer of plasma can be added which is sustained by the bias-supplied power. A simple theoretical model will be presented which shows that there is a threshold condition which must be satisfied in order for the bias-heated plasma to be self-sustaining. The bias-sustained plasma must also be opaque enough to neutrals in order for it to be fueled by a gas puff, which means that it win serve as a shield to the core plasma against neutral impurities and hydrogen. Experiments performed on DIII-D have demonstrated both a modification of the central nickel impurity concentration and an increase in the ionization of hydrogen within the scrape-off layer due to biasing.
Date: September 1, 1992
Creator: Staebler, G.M.; Hyatt, A.W.; Schaffer, M.J. & Mahdavi, M.A.
Partner: UNT Libraries Government Documents Department

Modeling of DIII-D noble gas puff and pump experiments

Description: Previous DIII-D experiments that induced a D{sup +} flow in the scrap-off layer (SOL) showed that this flow increased the divertor concentration of extrinsically injected impurities (neon, argon). These impurity fueling and exhaust (or puff and pump) experiments raise a number of modeling issues: the effect of edge-localized modes (ELMs) in regulating impurity core accumulation; the particle balance of the extrinsic impurities; the relation between divertor and plenum enrichment; and the effect of features unique to the present DIII-D Advanced Divertor configuration, specifically, the localized back-conductance of D{sub 2} and impurities from the baffle plenum in the outboard divertor region. To aid in understanding the relations between these processes, models have been improved: for core impurity transport to include ELM effects, and for divertor models to treat helium, neon, and argon transport with DIII-D--specific configuration effects. The models have been used to analyze a series of experiments in which neon and argon were first continuously injected (in the divertor private flux region) for 1.5 s, and then exhausted by the DIII-D cryopumping system. Deuterium was puffed at rates of 80 Torr L/s and 150 Torr L/s from the midplane and the divertor private region in these experiments. Results of the simulations are given.
Date: August 1, 1997
Creator: Hogan, J.T.; Wade, M.; Maingi, R.; Owen, L.; Schaffer, M. & West, P.
Partner: UNT Libraries Government Documents Department

Divertor E X B Plasma Convection in DIII-D

Description: Extensive two-dimensional measurements of plasma potential in the DIII-D tokamak divertor region are reported for standard (ion VB{sub T} drift toward divertor X-point) and reversed B{sub T} directions; for low (L) and high (H) confinement modes; and for partially detached divertor mode. The data are consistent with recent computational modeling identifying E x B{sub T} circulation, due to potentials sustained by plasma gradients, as the main cause of divertor plasma sensitivity to B{sub T} direction.
Date: July 1, 1999
Creator: Boedo, J.A.; Schaffer, M.J.; Maingi, M.; Lasnier, C.J. & Watkins, J.G.
Partner: UNT Libraries Government Documents Department


Description: Detailed measurements in two dimensions by probes and Thomson scattering reveal unexpected local electric potential and electron pressure (p{sub e}) maxima near the divertor X-point in L-mode plasmas in the DIII-D tokamak [J.L. Luxon and L.G. Davis, Fusion Technol. 8, 441 (1985)]. The potential drives E x B circulation about the X-point, thereby exchanging plasma between closed and open magnetic surfaces at rates that can be comparable to the total cross-separatrix transport. The potential is consistent with the classical parallel Ohm's law. A simple model is proposed to explain the pressure and potential hills in low power, nearly detached plasmas. Recent two-dimensional edge transport modeling with plasma drifts also shows X-point pressure and potential hills but by a different mechanism. These experimental and theoretical results demonstrate that low power tokamak plasmas can be far from poloidal uniformity in a boundary layer just inside the separatrix. Additional data, though preliminary and incomplete, suggest that E x B circulation across the separatrix might be a common feature of low confinement behavior.
Date: November 1, 2000
Partner: UNT Libraries Government Documents Department

Calculation of the Thermal Footprint of Resonant Magnetic Perturbations in DIII-D

Description: The effect of resonant magnetic perturbations on heat transport in DIII-D H-mode plasmas has been calculated by combining the TRIP3D field-line tracing code with the E3D two-fluid transport code. Simulations show that the divertor heat flux distribution becomes non-axisymmetric because heat flux is efficiently guided to the divertor along the three-dimensional invariant manifolds of the magnetic field. Calculations demonstrate that heat flux is spread over a wider area of the divertor target, thereby reducing the peak heat flux delivered during steady-state operation. Filtered optical cameras have observed non-axisymmetric particle fluxes at the strike-point and Langmuir probes have observed non-axisymmetric floating potentials. On the other hand, the predicted magnitude of stochastic thermal transport is too large to match the pedestal plasma profiles measured by Thomson scattering and charge exchange recombination spectroscopy. The Braginskii thermal conductivity overestimates the expected heat transport in the pedestal because the mean free path is longer than estimates of the parallel thermal correlation length, and collisionless transport models are probably required for accurate description. However, even the collisionless estimates for electron thermal transport are too large by one to two orders of magnitude. Thus, it is likely that another mechanism such as rotational screening of resonant perturbations limits the stochastic region and reduces transport inside of the pedestal.
Date: September 14, 2007
Creator: Joseph, I; Evans, T; Moyer, R; Fenstermacher, M; Groth, M; Kasilov, S et al.
Partner: UNT Libraries Government Documents Department

Stochastic Transport Modeling of Resonant Magnetic Perturbations in DIII-D

Description: Three-dimensional two-fluid simulations of heat transport due to resonant magnetic perturbations of tokamaks have been computed by coupling the TRIP3D field line tracing code to the E3D edge transport code. The predicted electron temperature contours follow the new separatrix represented by the perturbed invariant manifold structure of the X-point in qualitative agreement with X-point TV observations. However, preliminary modeling predicts that the resulting stochastic heat transport is greater than that measured in low-collisionality ELM suppression experiments in DIII-D H-mode plasmas. While improved determination of transport coefficients is definitely required, possible explanations include plasma screening of resonant perturbations, invalid treatment of the edge as a fluid, or insufficient understanding of stochastic heat transport.
Date: June 1, 2006
Creator: Joseph, I; Moyer, R A; Evans, T E; Schaffer, M J; Runov, A M; Schneider, R et al.
Partner: UNT Libraries Government Documents Department

Comparison of Edge Plasma Perturbation During ELM Control Using One vs Two Toroidal Rows of RMP Coils in ITER Similar Shaped Plasmas on DIII-D

Description: Large Type-I edge localized modes (ELMs) were suppressed by n = 3 resonant magnetic perturbations (RMPs) from a set of internal coils (I-coil) in plasmas with an ITER similar shape at the ITER pedestal collisionality, {nu}*{sub e} {approx} 0.1 and low edge safety factor (q{sub 95} {approx} 3.6), with either a single toroidal row of the internal RMP coils or two poloidally separated rows of coils. ELM suppression with a single row of internal coils was achieved at approximately the same q{sub 95} surface-averaged perturbation field as with two rows of coils, but required higher current per coil. Maintaining complete suppression of ELMs using n = 3 RMPs from a single toroidal row of internal coils was less robust to variations in input neutral beam injection torque than previous ELM suppression cases using both rows of internal coils. With either configuration of RMP coils, maximum ELM size is correlated with the width of the edge region having good overlap of the magnetic islands from vacuum field calculations.
Date: May 21, 2008
Creator: Fenstermacher, M E; Evans, T E; Osborne, T H; Schaffer, M J; deGrassie, J S; Gohil, P et al.
Partner: UNT Libraries Government Documents Department

Direct measurement of divertor exhaust neo enrichment in DIII-D

Description: We report first direct measurements of divertor exhaust gas impurity enrichment, {eta}{sub exh}=(exhaust impurity concentration){divided_by}(core impurity concentration), for both unpumped and D{sub 2} puff-with-divertor-pump conditions. The experiment was performed with neutral beam heated, ELMing H-mode, single-null diverted deuterium plasmas with matched core and exhaust parameters in the DIII-D tokamak. Neon gas impurity was puffed into the divertor. Neon density was measured in the exhaust by a specially modified Penning gauge and in the core by absolute charge exchange recombination spectroscopy. Neon particle accounting indicates that much of the puffed neon entered a temporary unmeasured reservoir, inferred to be the graphite divertor target, which makes direct measurements necessary to calculate divertor enrichments. D{sub 2} puff into the SOL (scrape-off layer) with pumping increased {eta}{sub exh} threefold over either unpumped conditions or D{sub 2} puff directly into the divertor with pumping. These results show that SOL flow plays an important role in divertor exhaust impurity enrichment.
Date: June 1, 1996
Creator: Schaffer, M.J.; Wade, M.R.; Maingi, R.; Monier-Garbet, P.; West, W.P.; Whyte, D.G. et al.
Partner: UNT Libraries Government Documents Department

Scientific basis and engineering design to accommodate disruption and halo current loads for the DIII-D tokamak

Description: Plasma disruptions and halo current events apply sudden impulsive forces to the interior structures and vacuum vessel walls of tokamaks. These forces arise when induced toroidal currents and attached poloidal halo currents in plasma facing components interact with the poloidal and toroidal magnetic fields respectively. Increasing understanding of plasma disruptions and halo current events has been developed from experiments on DIII-D and other machines. Although the understanding has improved, these events must be planned for in system design because there is no assurance that these events can be eliminated in the operation of tokamaks. Increased understanding has allowed an improved focus of engineering designs.
Date: October 1996
Creator: Anderson, P. M.; Bozek, A. S.; Hollerbach, M. A.; Humphreys, D. A.; Luxon, J. L.; Reis, E. E. et al.
Partner: UNT Libraries Government Documents Department

Plasma pressure and flows during divertor detachment

Description: MHD theory applied to tokamak plasma scrape-off layer (SOL) equilibria requires Pfirsch-Schlueter current, which, because the magnetic lines are open, normally closes through electrically conducting divertor or limiter components. During detached divertor operation the Pfirsch-Schlueter current path to the divertor target is sometimes blocked, in which case theory predicts that the plasma develops a poloidal pressure gradient around the upstream SOL and a corresponding parallel flow, in order to satisfy all the conditions of MHD equilibrium. This paper reports the only known examples of detached diverted plasma in the DIII-D tokamak with blocked Pfirsch-Schlueter current, and they show no clear SOL poloidal pressure differences. However, the predicted pressure differences are small, near the limit of detectability with the available diagnostics. In the more usual DIII-D partially detached divertor operation mode, the Pfirsch-Schlueter current appears to never be blocked, and no unusual poloidal pressure differences are observed, as expected. Finally, a local overpressure is observed just inside the magnetic separatrix near the X-point in both attached and detached Ohmically heated plasmas.
Date: August 1, 1998
Creator: Schaffer, M.J.; Brooks, N.H.; Boedo, J.A.; Isler, R.C. & Moyer, R.A.
Partner: UNT Libraries Government Documents Department

Observation of SOL Current Correlated with MHD Activity in NBI-heated DIII-D Tokamak Discharges

Description: This work investigates the potential roles played by the scrape-off-layer current (SOLC) in MHD activity of tokamak plasmas, including effects on stability. SOLCs are found during MHD activity that are: (1) slowly growing after a mode-locking-like event, (2) oscillating in the several kHz range and phase-locked with magnetic and electron temperature oscillations, (3) rapidly growing with a sub-ms time scale during a thermal collapse and a current quench, and (4) spiky in temporal behavior and correlated with spiky features in Da signals commonly identified with the edge localized mode (ELM). These SOLCs are found to be an integral part of the MHD activity, with a propensity to flow in a toroidally non-axisymmetric pattern and with magnitude potentially large enough to play a role in the MHD stability. Candidate mechanisms that can drive these SOLCs are identified: (a) toroidally non-axisymmetric thermoelectric potential, (b) electromotive force (EMF) from MHD activity, and (c) flux swing, both toroidal and poloidal, of the plasma column. An effect is found, stemming from the shear in the field line pitch angle, that mitigates the efficacy of a toroidally non-axisymmetric SOLC to generate a toroidally non-axisymmetric error field. Other potential magnetic consequences of the SOLC are identified: (i) its error field can introduce complications in feedback control schemes for stabilizing MHD activity and (ii) its toroidally non-axisymmetric field can be falsely identified as an axisymmetric field by the tokamak control logic and in equilibrium reconstruction. The radial profile of a SOLC observed during a quiescent discharge period is determined, and found to possess polarity reversals as a function of radial distance.
Date: March 26, 2004
Creator: Takahashi, H.; Fredrickson, E.D.; Schaffer, M.J.; Austin, M.E.; Evans, T.E.; Lao, L.L. et al.
Partner: UNT Libraries Government Documents Department


Description: Small non-axisymmetric magnetic fields are known to cause serious loss of stability in tokamaks leading to loss of confinement and abrupt termination of plasma current (disruptions). The best known examples are the locked mode and the resistive wall mode. Understanding of the underlying field anomalies (departures in the hardware-related fields from ideal toroidal and poloidal fields on a single axis) and the interaction of the plasma with them is crucial to tokamak development. Results of both locked mode experiments and resistive wall mode experiments done in DIII-D tokamak plasmas have been interpreted to indicate the presence of a significant anomalous field. New measurements of the magnetic field anomalies of the hardware systems have been made on DIII-D. The measured field anomalies due to the plasma shaping coils in DIII-D are smaller than previously reported. Additional evaluations of systematic errors have been made. New measurements of the anomalous fields of the ohmic heating and toroidal coils have been added. Such detailed in situ measurements of the fields of a tokamak are unique. The anomalous fields from all of the coils are one third of the values indicated from the stability experiments. These results indicate limitations in the understanding of the interaction of the plasma with the external field. They indicate that it may not be possible to deduce the anomalous fields in a tokamak from plasma experiments and that we may not have the basis needed to project the error field requirements of future tokamaks.
Date: February 1, 2003
Partner: UNT Libraries Government Documents Department

Numerical Study of the Formation, Ion Spin-up and Nonlinear Stability Properties of Field-reversed Configurations

Description: Results of three-dimensional numerical simulations of field-reversed configurations (FRCs) are presented. Emphasis of this work is on the nonlinear evolution of magnetohydrodynamic (MHD) instabilities in kinetic FRCs and the new FRC formation method by the counter-helicity spheromak merging. Kinetic simulations show nonlinear saturation of the n = 1 tilt mode, where n is the toroidal mode number. The n = 2 and n = 3 rotational modes are observed to grow during the nonlinear phase of the tilt instability due to the ion spin-up in the toroidal direction. The ion toroidal spin-up is shown to be related to the resistive decay of the internal flux, and the resulting loss of particle confinement. Three-dimensional MHD simulations of counter-helicity spheromak merging and FRC formation show good agreement with results from the SSX-FRC experiment. Simulations show formation of an FRC in about 30 Alfven times for typical experimental parameters. The growth rate of the n = 1 tilt mode is shown to be significantly reduced compared to the MHD growth rate due to the large plasma viscosity and field-line-tying effects.
Date: November 12, 2004
Creator: Belova, E.V.; Davidson, R.C.; Ji, H.; Yamada, M.; Cothran, C.D.; Brown, M.R. et al.
Partner: UNT Libraries Government Documents Department

The design and fabrication of a toroidally continuous cryocondensation pump for the DIII-D Advanced Divertor

Description: A cryocondensation pump will be installed in the baffle chamber of the DIII-D tokamak in the spring of 1992. The design is complete and fabrication of this pump is in progress. The purpose of the pump is to study plasma density control by pumping the divertor. The pump is toroidally continuous, approximately 10 m long, in the lower outer corner of the vacuum vessel interior. It consists of a 1 m{sup 2} liquid helium cooled surface surrounded by a liquid nitrogen cooled shield to limit the heat load on the helium cooled surface. The stainless steel liquid nitrogen shell has a copper coating on it to enhance thermal conductivity, but the coating is broken to keep the toroidal electrical resistance high. The liquid nitrogen cooled surface is surrounded by a radiation/particle shield to prevent energetic particles from impacting and releasing condensed water molecules. The whole pump is supported off the water cooled vacuum vessel wall. Key design considerations were: how to accommodate the temperature differences between the various components, developing low heat leak paths for the various supports, and maintaining electrical insulation in a low pressure environment in the presence of induced voltage spikes. A single point ground for the system was used to limit disruption induced currents and the resulting electro-mechanical forces on the pump. A testing program was used to develop coating techniques to enhance heat transfer and emissivity of the various surfaces. Fabrication tests were done to determine the best method of attaching the liquid nitrogen flow tubes to their shield surfaces. A prototype sector of the pump was built to verify fabrication and assembly techniques.
Date: November 1, 1991
Creator: Smith, J. P.; Baxi, C. B.; Reis, E.; Schaffer, M. J.; Schaubel, K. M. & Menon, M. M.
Partner: UNT Libraries Government Documents Department

Engineering, installation, testing, and initial operation of the DIII-D Advanced Divertor

Description: The Advanced Divertor (AD) for General Atomics tokamak, DIII-D, was installed in the summer of 1990. The AD has enabled two classes of physics experiments to be run: divertor biasing and divertor baffling. Both are new experiments for DIII-D. The AD has two principal components: (1) a continuous ring electrode; and (2) a toroidally symmetric baffle. The tokamak can be run in bias baffle or standard DIII-D divertor modes by accurate positioning of the outer divertor strike point through the use of the DIII-D control system. The paper covers design, analysis, fabrication, installation, instrumentation, testing, initial operation, and future plans for the Advanced Divertor from an engineering viewpoint. 2 refs., 5 figs.
Date: September 1, 1990
Creator: Andersen, P.M.; Baxi, C.B.; Reis, E.E.; Schaffer, M.J. & Smith, J.P.
Partner: UNT Libraries Government Documents Department

Fast Ion Effects During Test Blanket Module Simulation Experiments in DIII-D

Description: Fast beam-ion losses were studied in DIII-D in the presence of a scaled mockup of two Test Blanket Modules (TBM) for ITER. Heating of the protective tiles on the front of the TBM surface was found when neutral beams were injected and the TBM fields were engaged. The fast-ion core confinement was not significantly affected. Different orbit-following codes predict the formation of a hot spot on the TBM surface arising from beam-ions deposited near the edge of the plasma. The codes are in good agreement with each other on the total power deposited at the hot spot predicting an increase in power with decreasing separation between the plasma edge and the TBM surface. A thermal analysis of the heat flow through the tiles shows that the simulated power can account for the measured tile temperature rise. The thermal analysis, however, is very sensitive to the details of the localization of the hot spot which is predicted to be different among the various codes.
Date: June 3, 2011
Creator: Kramer, G. J.; Ellis, R.; Gorelenkova, M.; Heidbrink, W. W.; Kurki-Suonio, T.; Nazikian, R. et al.
Partner: UNT Libraries Government Documents Department