Search

We report on recent experiments on DIII-D that examined the effects that variations in the parallel connection length in the scrape-off layer (SOL), L||, and the radial location of the outer divertor target, RTAR, have on divertor plasma properties. Two-point modeling of the SOL plasma predicts that larger values of L|| and RTAR should lower temperature and raise density at the outer divertor target for fixed upstream separatrix density and temperature, i.e., nTAR ∝ [RTAR]2[L||]6/7 and TTAR ∝ [RTAR]−2[L||]−4/7. The dependence of nTAR and TTAR on L|| was consistent with our data, but the dependence of nTAR and TTAR on RTAR was not. The surprising result that the divertor plasma parameters did not depend on RTAR in the predicted way may be due to convected heat flux, driven by escaping neutrals, in the more open configuration of the larger RTAR cases. Modeling results using the SOLPS code support this postulate.

Predictions of the recently developed paleoclassical transport model are compared with data from many toroidal plasma experiments: electron heat diffusivity in DIII-D, C-Mod and NSTX ohmic and near-ohmic plasmas; transport modeling of DIII-D ohmic-level discharges and of the RTP ECH 'stair-step' experiments with eITBs at low order rational surfaces; investigation of a strong eITB in JT-60U; H-mode Te edge pedestal properties in DIII-D; and electron heat diffusivities in non-tokamak experiments (NSTX/ST, MST/RFP, SSPX/spheromak). The radial electron heat transport predicted by the paleoclassical model is found to agree with a wide variety of ohmic-level experimental results and to set the lower limit (within a factor {approx} 2) for the radial electron heat transport in most resistive, current-carrying toroidal plasmas -- unless it is exceeded by fluctuation-induced transport, which often occurs in the edge of L-mode plasmas and when the electron temperature is high ({approx}>T{sub e}{sup crit} {approx}B{sup 2/3}{bar {alpha}}{sup 1/2} keV) because then paleoclassical transport becomes less than gyro-Bohm-level anomalous transport.

A radiating divertor approach was successfully applied to high performance 'hybrid' plasmas [M.R. Wade, et al., Proc. 20th IAEA Fusion Energy Conf., Vilamoura, (2004)]. Our technique included: (1) injecting argon near the outer divertor target, (2) enhancing the plasma flow into the inner and outer divertors by a combination of particle pumping and deuterium gas puffing upstream of the divertor targets, and (3) isolating the inner divertor from the outer by a structure in the private flux region. Good hybrid conditions were maintained, as the peak heat flux at the outer divertor target was reduced by a factor of 2.5; the peak heat flux at the inner target decreased by 20%. This difference was caused by a higher concentration of argon at the outer target than at the inner target. Argon accumulation in the main plasma was modest (n{sub AR}/n{sub e} {le}0.004 on axis), although the argon profile was more peaked than the electron profile.

Recent experiments on DIII-D point to the importance of two factors in determining how effectively the deuterium particle inventory in a tokamak plasma can be controlled through pumping at the divertor target(s): (1) the divertor magnetic balance, i.e. the degree to which the divertor topology is single-null or double-null (DN) and (2) the direction of the of B × ∇B ion drift with respect to the X-point(s). Changes in divertor magnetic balance near the DN shape have a much stronger effect on the particle exhaust rate at the inner divertor target(s) than on the particle exhaust rate at the outer divertor target(s). The particle exhaust rate for the DN shape is strongest at the outer strike point opposite the B × ∇B ion particle drift direction. Our data suggests that the presence of B × ∇B and E × B ion particle drifts in the scrape-off layer and divertor(s) play an important role in the particle exhaust rates of DN and near-DN plasmas. Particle exhaust rates are shown to depend strongly on the edge (pedestal) density. These results have implications for particle control in ITER and other future tokamaks.

Far Scrape-Off Layer (SOL) and near-wall plasma parameters in DIII-D depend strongly on the discharge parameters and confinement regime. In L-mode discharges cross-field transport increases with the average discharge density and flattens far SOL profiles, thus increasing plasma contact with the low field side (LFS) main chamber wall. In H-mode between edge localized modes (ELMs) the plasma?wall contact is weaker than in L-mode. During ELM fluxes of particles and heat to the LFS wall increase transiently above the L-mode values. Depending on the discharge conditions, ELMs are responsible for 30?90% of the net ion flux to the outboard chamber wall. ELMs in high density discharges feature intermittent transport events similar to those observed in L-mode and attributed to blobs of dense hot plasma formed inside the separatrix and propagating radially outwards. Though the blobs decay with radius, some of them survive long enough to reach the outer wall and possibly cause sputtering. In lower density H-modes, ELMs can feature blobs of pedestal density propagating all the way to the outer wall.

The pioneering research on the Doublet-III (DIII) tokamak and its upgrade the DIII-D has contributed significantly to understanding of the physics of divertor plasmas and the development of the mod...

Studies of the H-mode pedestal in the DIII-D tokamak are presented. The global energy confinement increases as the plasma pressure on top of the pedestal increases. The best empirical description for a pedestal width parameter is {delta}{sub pe} [proportional to] ({beta}{sub pol}{sup PED}){sup 0.4}, where {delta}{sub pe} is the width of the electron pressure pedestal and {beta}{sub pol}{sup PED} is the poloidal beta at the top of the pedestal. The edge profiles of electron density n{sub e}, electron temperature T{sub e}, and ion temperature T{sub i} can all have different shapes. Thus, a simple width scaling for the edge might not exist, and studies of the physics of individual profiles have been initiated. A model for the n{sub e} profile, based on self-consistent treatment of edge particle sources and edge particle transport, agrees with several experimental observations. The steep gradient region for the T{sub e} profile often extends farther into the plasma than the n{sub e} pedestal step. Magnetohydrodynamic stability provides the ultimate limits to the evolution of the pedestal and usually leads to edge instabilities called edge-localized modes (ELMs). However, the absence of ELMs in a regime called the Quiescent H-mode shows that large pedestals can be produced withoutmore » ELMs.« less

Recent experiments on DIII-D point to the importance of two factors in determining the rate at which deuterium particles can be pumped at the divertor target(s): (1) the divertor magnetic balance, i.e., the degree to which the divertor topology is single-null (SN) or double-null (DN), and (2) the direction of the of B x {del}B ion drift with respect to the X-point(s). Changes in divertor magnetic balance near the DN shape have a much stronger effect on the pumping rate at the inner divertor target(s) than on the pumping rate at the outer divertor target(s). The behavior in the particle pumping observed at the inner and outer divertor target(s) in the DN and near-DN shapes suggests a redistribution of particles that would be expected in the presence of B x {del}B and E x B ion particle drifts in the scrapeoff layer (SOL) and divertor(s).

A series of hydrogen and deuterium discharges are analyzed with fluid plasma and Monte Carlo neutrals codes. Comparison of poloidally averaged radial distributions of core neutral density and ionization with analytic solutions of 1-D plasma and neutrals continuity equations support the hypothesis that the width of the density pedestal is largely determined by the neutral source. The increased neutral penetration depth that arises from multiple charge exchange can be included in the analytic model with radially dependent scale lengths. The scale length in the analytic model depends on the neutral fluid velocity which increases across the divertor and pedestal as the neutral atoms charge exchange with the higher temperature background ions. The neutral penetration depth and corresponding density pedestal width depend sensitively on the neutral temperature and the degree of ion-neutral temperature equilibration.

The most important mechanisms for introducing carbon into the DIII-D divertors [Nucl. Fusion 42 (2002) 614] are physical and chemical sputtering. Previous investigations have indicated that operating conditions where one or the other of these is dominant can be distinguished by using CD and C{sub 2} emissions to infer C I influxes from dissociation of hydrocarbons and comparing to measured C I influxes. The present work extends these results through detailed analysis of the C I spectral line shapes. In general, it is found that the profiles are actually asymmetric and have shifted peaks. These features are interpreted as originating from a combination of an anisotropic velocity distribution from physical sputtering (the Thompson model) and an isotropic distribution from molecular dissociation. The present study utilizes pure helium plasmas to benchmark C I spectral profiles arising from physical sputtering alone.

Far scrape-off layer (SOL) plasma parameters in DIII-D depend strongly on the discharge density and confinement regime. In L-mode, cross-field transport increases with the average discharge density and elevates the far SOL density, thus increasing plasma-wall contact. Far SOL density near the low field side (LFS) of the main chamber wall also increases with decreasing plasma current and with decreasing outer wall gap. In H-mode, between edge localized modes (ELMs), plasma-wall contact is weaker than in L-mode. During ELMs plasma fluxes to the LFS wall increase to, or above the L-mode levels. A large fraction of the net cross-field fluxes is convected through the SOL by large amplitude intermittent transport events. In high-density L-mode and during ELMs in H-mode, intermittent events propagate all the way to the LFS wall and may cause sputtering.

We present the dependence of the neutral pressure rise on strike point geometry in the inner and outer plenums of the upper divertor in the DIII-D tokamak. The upper outer plenum pressure increases in the proximity of the plenum opening in a manner similar to previous reports of the lower outer plenum pressure. The inner plenum pressure measurement is modulated up to a factor of two by edge-localized modes (ELMs) particle flux, due to the small plenum volume and resulting short equilibration time constant. Reasonable agreement is obtained between the measured dependence of the outer plenum pressure on strike point position and simulations with an analytic neutral transport model, using the time-averaged (i.e. long compared with the ELM period) divertor profiles as inputs. The modelling of inner plenum pressure data with time-averaged profiles traces through the middle of the data range. In addition, modelling with two other sets of profiles, one averaged in the vicinity of ELMs and the other between ELMs, covers the upper and lower envelopes of the data. These new data sets represent an independent test and the agreement confirms that the basic physics in the model is responsible for the dependence of pressure rise on strike point geometry in the long mean free path limit.

Pedestal studies in DIII-D find that the width of the region of steep gradient in the H-mode density is comparable with the neutral penetration length, as computed from a simple analytic model. This model has analytic solutions for the edge plasma and neutral density profiles, which are obtained from the coupled particle continuity equations for electrons and deuterium atoms. In its range of validity (edge temperature between 40 and 500 eV), the analytic model quantitatively predicts the observed decrease in the width as the pedestal density increases and the observed strong increase in the gradient of the density as the pedestal density increases. The model successfully predicts that L-mode and H-mode profiles with the same pedestal density have gradients that differ by less than a factor of 2. The width of the density barrier, measured from the edge of the electron temperature barrier, is the lower limit for the observed width of the temperature barrier. These results support the hypothesis that particle fuelling is an important part of the physics that determines the structure of the H-mode transport barrier.

Based on semianalytic modeling of experimental H-mode density profiles, a pedestal density limit is computed for a prototypical next generation burning plasma. For a range of input assumptions, the projected limit lies near or above Greenwald density limit [M. Greenwald, J. L. Terry, S. M. Wolfe, S. Ejima, M. G. Bell, S. M. Kaye, and G. H. Neilson, Nucl. Fusion 28, 2199 (1988)]. Two separate models are developed and utilized: the first model determines the edge density profile from consideration of cross-field transport in the steep gradient region and the distribution of fueling sources. It is shown that the observed shape of the density profile in H-mode can be explained with constant particle diffusivity within the separatrix. The second model determines the separatrix density above which the conduction limited scrape-off layer (SOL) transport is no longer sustainable. These results imply that further increases in pedestal density (while maintaining separatrix density below the power balance limit) are...

Key DIII-D advanced tokamak (AT) experimental and modelling results are applied to examine the physics and control issues for ITER to operate in a negative central shear (NCS) AT scenario. The effects of a finite edge pressure pedestal and current density are included based on the DIII-D experimental profiles. Ideal and resistive stability analyses demonstrate that feedback control of resistive wall modes by rotational drive or flux conserving intelligent coils is crucial for these AT configurations to operate at attractive βN values in the range 3.0–3.5. Vertical stability and halo current analyses show that reliable disruption mitigation is essential and mitigation control using an impurity gas can significantly reduce the local mechanical stress to an acceptable level. Core transport and turbulence analyses indicate that control of the rotational shear profile is essential to reduce the pedestal temperature required for high β. Consideration of edge stability and core transport suggests that a sufficiently wide pedestal is necessary for the projected fusion performance. Heat flux analyses indicate that, with core-only radiation enhancement, the outboard peak divertor heat load is near the design limit of 10 MW m−2. Detached operation may be necessary to reduce the heat flux to a more manageable level. Evaluation of the ITER pulse length using a local step response approach indicates that the 3000 s ITER long-pulse scenario is probably both necessary and sufficient for demonstration of local current profile control.

The reduction in size of Type I edge localized modes (ELMs) with increasing density is explored in DIII-D [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] for the purpose of studying the underlying transport of ELM energy. The separate convective and conductive transport of energy due to an ELM is determined by Thomson scattering measurements of electron density and temperature in the pedestal. The conductive transport from the pedestal during an ELM decreases with increasing density, while the convective transport remains nearly constant. The scaling of the ELM energy loss is compared with an edge stability model. The role of the divertor sheath in limiting energy loss from the pedestal during an ELM is explored. Evidence of outward radial transport to the midplane wall during an ELM is also presented.

Measurements from balanced or slightly unbalanced, double-null, ELMing H-mode plasmas show that the upper, more closed, divertor strike point particle fluxes are larger than the lower, more open, divertor strike point particle fluxes for comparable conditions. Comparison of the upper and lower outer strike point peak particle fluxes measured with Langmuir probes during scans of the magnetic balance shows an asymmetry around the magnetic balance point which is consistent with both the enhanced upper recycling and the expected direction of private region E × B drifts. Comparison of the inner and outer peak particle fluxes in the upper divertor shows a dependence on magnetic field direction that is also consistent with the private region E × B drift direction.

We report on DIII–D data that reveal the underlying processes responsible for transport of energy and particles from the edge pedestal to the divertor target during Type I edge-localized modes (ELMs). The separate convective and conductive transport of energy due to an ELM is determined by Thomson scattering measurements of electron density and temperature in the pedestal. Conductive transport is measured as a drop in pedestal temperature and decreases with increasing density. The convective transport of energy, measured as a loss of density from the pedestal, however, remains constant as a function of density. From the SOL, ELM energy is quickly carried to the divertor target. An expected sheath limit to the ELM heat flux set by the slower arrival of pedestal ions is overcome by additional ionization of neutrals generated from the divertor target as evidenced by a fast, ∼100 μs, rise in divertor density. A large in/out asymmetry of the divertor ELM heat flux is observed at high density, but becomes nearly symmetric at low density.

The 2-D fluid code UEDGE was used to analyze DIII-D experiments to determine the role of neutrals in core fueling, core impurities, and also the H-mode pedestal structure. We compared the effects of divertor closure on the fueling rate and impurity density of high-triangularity, H-mode plasmas. UEDGE simulations indicate that the decrease in both deuterium core fueling (≈15–20%) and core carbon density (≈15–30%) with the closed divertor compared to the open divertor configuration is due to greater divertor screening of neutrals. We also compared UEDGE results with a simple analytic model of the H-mode pedestal structure [Nucl. Fusion 42 (2002) 52]. The model predicts both the width and gradient of the transport barrier in n e as a function of the pedestal density. The more sophisticated UEDGE simulations of H-mode discharges corroborate the simple analytic model, which is consistent with the hypothesis that fueling processes play a role in H-mode transport barrier formation.

The energy lost from the pedestal region due to an average ELM in DIII-D is determined from changes to the electron density and temperature profiles as measured by Thomson scattering. The ELM energy due to loss of temperature in the pedestal is associated with conduction and is found to decrease with increasing pedestal density. The ELM energy from lost pedestal density, or convective transport, remains constant as a function of density. The scaling of the two transport channels, conduction and convection, are examined in terms of parallel transport processes in the scrape-off-layer and divertor.

Cross-field fluctuation-driven transport is studied in edge and scrape-off layer (SOL) plasmas in the DIII-D tokamak using a fast reciprocating Langmuir probe array allowing local measurements of the fluctuation-driven particle and heat fluxes. Two different non-diffusive mechanisms that can contribute strongly to the cross-field transport in the SOL of high-density discharges are identified and compared. The first of these involves intermittent transport events that are observed at the plasma separatrix and in the SOL. Intermittence has qualitatively similar character in L-mode and ELM-free H-mode. Low-amplitude ELMs observed in high-density H-mode produce in the SOL periods with cross-field transport enhanced to L-mode levels and featuring intermittent events similar to those in L-mode. The intermittent transport events are compatible with the concept of plasma filaments propagating across the SOL due to E×B drifts. The intermittent character of the transport in the SOL is also in agreement with predictions of the non-linear numerical simulations performed with an imposed driving flux. Another type of non-diffusive transport is often seen in high-density H-modes with prolonged ELM-free periods, where the transport near the separatrix is dominated by quasi-coherent modes driving particle and/or heat fluxes exceeding L-mode levels. These modes may play an important role by providing particle and/or heat exhaust between ELMs.

An analytic model, derived from coupled continuity equations for the electron and neutral deuterium densities, is consistent with many features of edge electron density profiles in the DIII-D tokamak [J. L. Luxon et al., Plasma Physics and Controlled Fusion Research, 1986, Vol. I (International Atomic Energy Agency, Vienna, 1987), p. 159]. For an assumed constant particle diffusion coefficient, the model shows that particle transport and neutral fueling produce electron and neutral density profiles that have the same characteristic scale lengths at the plasma edge. For systematic variations of density in high-mode (H-mode) discharges, the model predicts that the width of the electron density transport barrier decreases and the maximum gradient increases, as observed in the experiments. The widths computed from the model agree quantitatively with the experimental widths for conditions in which the model is valid. These results support models of transport barrier formation in which the H-mode particle barri...

An analytic model for the width of the electron density transport barrier shows qualitative and quantitative agreement with systematic data obtained in the DIII-D tokamak. The width of the barrier in H-mode decreases as the pedestal density increases. Moreover, for L-mode and H-mode density profiles with the same pedestal density, the gradients of the barriers are of similar magnitude even though the temperature and pressure gradients are much larger in H-mode. These results are consistent with the hypothesis that neutral penetration provides important physics which helps to set the width of the H-mode barrier for the electron density.

Densities of up to 40% above the Greenwald limit are reproducibly achieved in high confinement (HITER89P = 2) ELMing H mode discharges. Simultaneous gas fuelling and divertor pumping were used to obtain these results. Confinement of these discharges, similar to moderate density H mode, is characterized by a stiff temperature profile, and is therefore sensitive to the density profile. A particle transport model is presented that explains the roles of divertor pumping and geometry for access to high densities. The energy loss per ELM at high density is a factor of five lower than the predictions of an earlier scaling, based on data from lower density discharges.

Particle pumping characteristics were compared between pumping from the inner side private flux region (IPP) and pumping from both sides of the private flux region (BPP) in the JT-60U W shaped divertor, and between JT-60U IPP and pumping in the DIII-D lower baffled divertor. The pumping flux for BPP is smaller than that for IPP by about a factor of 2 with weak in-out asymmetry of recycling neutral flux and by a factor of 3.5-6.5 with strong in-out asymmetry. The reduction of the pumping flux for BPP is consistent with Monte Carlo simulations, where backflow at the outer pumping slot is observed due to in-out recycling asymmetry. The pumping flux in DIII-D at Ip = 0.8 MA and BT = 1.6 T is comparable to or smaller than that for JT-60U IPP at Ip = 1.0 MA, BT = 3.8 T and Ip = 1.5 MA, BT = 3.5 T in the same density regime. In the DIII-D divertor with pumping from the private flux region, the pumping flux decreases with increasing in-out asymmetry. The pumping flux normalized by the integrated Dα emission over the whole plasma exhibits a similar dependence on the distance between the pumping slot and the strike point in JT-60U IPP and the DIII-D lower divertor with pumping through the outer divertor plasma region.

Significant progress in obtaining high performance discharges lasting many energy confinement times in the DIII-D tokamak has been realized in recent experimental campaigns. Normalized performance ∼10 has been sustained for more than 5>E with qmin > 1.5. (The normalized performance is measured by the product βNH89, indicating the proximity to the conventional β limits and energy confinement quality, respectively.) These H mode discharges have an ELMing edge and β < 5%. The limit to increasing β is a resistive wall mode, rather than the tearing modes as previously observed. Confinement remains good despite qmin > 1. The global parameters were chosen to optimize the potential for fully non-inductive current sustainment at high performance, which is a key program goal for the DIII-D facility. Measurement of the current density and loop voltage profiles indicate that ≈75% of the current in the present discharges is sustained non-inductively. The remaining ohmic current is localized near the half-radius. The electron cyclotron heating system is being upgraded to replace this remaining current with ECCD. Density and β control, which are essential for operating advanced tokamak discharges, were demonstrated in ELMing H mode discharges with βNH89 ≈ 7 for up to 6.3 s or ≈34τE. These discharges appear to have stationary current profiles with qmin ≈ 1.05 in agreement with the current profile relaxation time ≈1.8 s.

The DIII–D program has completed a series of density control and plasma core confinement experiments this past year. These experiments were designed to investigate the performance of baffled and open divertors with single-null plasmas and particle control in double-null plasmas. The experiments utilized all three of the DIII–D divertor assemblies located in the lower outer corner, the upper outer corner, and the upper inner corner of the vessel, which were installed last year. Each divertor consists of a liquid helium cryopump, a shielded protective ring, and a gas puff system. The divertors were designed to optimize pumping performance and to withstand the electromagnetic loads from both halo and toroidal, induced currents. With theoretical pumping speeds varying from 15,000 to 32,000 l/s, the cryopumps, combined with the baffle structures, collect particles and prevent them from recirculating back into the plasma core. The intent of the gas puff systems is to inject neutral gases in and around the divertors to minimize the heat flux on the divertors, minimizing the impurities generated by the excessive heating of the divertor graphite tiles. This hardware permits either single- or double-null plasma experiments and enables continued research of well confined high beta divertor plasmas with noninductive current drive, which is one of the primary research goals of DIII–D.

Significant progress has been made in obtaining high-performance discharges for many energy confinement times in the DIII-D tokamak [J. L. Luxon et al., Plasma Physics and Controlled Fusion Research (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159]. Normalized performance (measured by the product of βNH89 and indicative of the proximity to both conventional β limits and energy confinement quality, respectively) ∼10 has been sustained for >5 τE with qmin>1.5. These edge localized modes (ELMing) H-mode discharges have β∼5%, which is limited by the onset of resistive wall modes slightly above the ideal no-wall n=1 limit, with approximately 75% of the current driven noninductively. The remaining Ohmic current is localized near the half-radius. The DIII-D electron cyclotron heating system is being upgraded to replace this inductively driven current with localized electron cyclotron current drive (ECCD). Density control, which is required for effective ECCD, has been successfully demonstrated in long-pulse high-performance ELMing H-mode discharges with βNH89∼7 for up to 6.3 s. In plasma shapes compatible with good density control in the present divertor configuration, the achieved βN is somewhat less than that in the high βNH89=10 discharges.

Particle confinement was compared for the high confinement mode plasmas with the edge localized mode in the Japan Atomic Energy Research Institute Tokamak-60 Upgrade (JT-60U) [S. Ishida, JT-60 Team, Nucl. Fusion 39, 1211 (1999)] and the DIII-D tokamak [J. L. Luxon et al., Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159] considering separate confinement times for particles supplied by neutral beam injection (NBI) (center fueling) and by recycling and gas-puffing (edge fueling). Similar dependence on the NBI power was obtained in JT-60U and DIII-D. The particle confinement time for center fueling in DIII-D was smaller by a factor of 4 in the low density discharges and by a factor of 1.8 in the high density discharges than JT-60U scaling, respectively, suggesting the stronger dependence on the density in DIII-D. The particle confinement time for edge fueling in DIII-D was comparable with JT-60U scaling in the low density discharges...

ELMing (edge-localized) H-mode discharges with densities as high as 40% above the Greenwald density and good energy confinement, HITER-89P=2, were obtained with D2 gas puffing on DIII-D [Chan et al., Proceedings of the 16th IAEA Conference, Montreal (International Atomic Energy Agency, Vienna, 1996), Vol. 1, p. 95]. These discharges have performance comparable to the best pellet fueled DIII-D discharges. Spontaneous peaking of the density profile was an important factor in obtaining high energy confinement. Without density profile peaking, the energy confinement at high density degraded with reduction in the H-mode pedestal pressure under the stiff temperature profile conditions observed at high density on DIII-D. Reduction in the pedestal pressure was associated with loss of access to the second stable regime for ideal ballooning modes at the edge, and change in the edge-localized mode (ELM) instability from a low to high toroidal mode number. Gyrokinetic stability calculations indicate that the core of ...

The energy released at each edge-localized mode (ELM) is found to decrease in relation to the pedestal pressure, by more than a factor of five, as the line-averaged density in DIII-D H-mode is raised from about half the Greenwald density limit to near the Greenwald limit. The pedestal pressure remains nearly constant over this range demonstrating an attractive regime for future larger tokamaks. The reduction in ELM energy, in both low and high triangularity configurations, is seen to scale more with the pedestal electron temperature than the pedestal density. At low density both the electron density and temperature inside the separatrix drop due to the ELM instability; however at high density the density perturbation remains similar while the temperature profile is unaffected. ELMs at high density are also characterized by smaller magnetic fluctuations consistent with a higher toroidal mode number ELM instability.

We report on recent experiments for which the magnetic balance of highly triangular ( δ ≈0.8), unpumped H-mode plasmas was varied. Changes in divertor heat loading and particle flux were observed when the magnetic configuration was varied from a balanced double-null (DN) divertor to a slightly unbalanced DN divertor. For attached plasmas, the variation in heat flux sharing between divertors is very sensitive near balanced DN. This sensitivity can be shown to be consistent with the measured scrape-off width of the parallel divertor heat flux density, λ q ∥ . At magnetic balance we find that the peak heat flux density at the divertor in the ∇ B ion drift direction is twice that of the other divertor. Most of the heat flux go to the outboard divertor targets in a balanced double-null, where the peak heat flux density at the outer divertor targets may exceed that of the inner divertor targets by tenfold. However, the variation of the peak particle flux density between divertors is less sensitive to changes in magnetic balance. These particle and heat flux `asymmetries' in DN plasmas are consistent with the presence of E × B poloidal particle drifts in the scrape-off layer and private flux region [1] . Regardless of how the divertors were magnetically balanced, D 2 gas puffing always reduced energy confinement to the range τ E / τ E 89 P ≈1.3–1.6. When this energy confinement range was reached, τ E / τ E 89 P remained nearly constant up to near the H-mode density limit.

A major upgrade of the DIII-D divertor, with the goal of enhancing impurity and density control and increasing the thermal pulse length limit of advanced tokamak (AT) plasmas has been successfully completed and commissioned. The integrated system that includes independent cryopumps at both the inner and the outer legs of the divertor, private flux region and outboard baffles, and improved graphite divertor armor, has been successfully applied to a variety of plasma conditions. Comparison of similar discharges before and after the upgrades show that with the new divertor the core plasma neutral source and carbon content are lower by as much as 50%. Calculations supported by preliminary infra-red (IR) camera measurements show that the new graphite armor design increases the limit on the discharge duration, due to temperature of the tile edges reaching sublimation point, by an order of magnitude. With the new system we have been able to control the density of high confinement H-mode plasmas to less than 1/3 of the Greenwald limit. It is observed that with divertor pumping during the current ramp phase the wall particle inventory and consequently the density rise after the H-mode transition can be significantly reduced.

The characteristics of the H-mode are studied in discharges with varying triangularity and squareness. The pressure at the top of the H-mode pedestal increases strongly with triangularity, primarily due to an increase in the margin by which the edge pressure gradient exceeds the ideal ballooning mode first stability limit. Two models are considered for how the edge may exceed the ballooning mode limit. In one model, access to the ballooning mode second stable regime allows the edge pressure gradient and associated bootstrap current to continue to increase until an edge localized, low toroidal mode number, ideal kink mode is destabilized. In the second model, the finite width of the H-mode transport barrier and diamagnetic effects raise the pressure gradient limit above the ballooning mode limit. We observe a weak inverse dependence of the width of the H-mode transport barrier, Δ, on triangularity relative to the previously obtained scaling Δ∝(βPEP)1/2. The energy loss for Type I ELMs increases with triangularity in proportion to the pedestal energy increase. At low density, the energy confinement of high-triangularity discharges is larger than discharges with low triangularity, as a result of an increase in the energy in the H-mode pedestal. At high density, both the change in pedestal pressure and the response of the density profile are found to play a role in setting the energy confinement. The highest energy confinement at high density was obtained in low-triangularity discharges.

Many theoretical models show a direct connection between energy transport and particle transport. For a complete understanding of transport processes both energy and particle transport must be understood. A discussion is presented of how energy and particle transport should be related in terms of the relative importance of the diagonal and off-diagonal terms in the equation for the fluxes. A model for particle transport is discussed and it is shown that this model can describe measured density profiles in the DIII-D tokamak under a wide range of DIII-D parameters. This model is obtained from a Lagrangian formulation of the kinetic equation and utilizes the assumption that transport takes place while approximately conserving the first and second adiabatic invariants. The measured results utilize the improved diagnostic capability of DIII-D with an improved capability of measuring current density profiles and particle density profiles. This model is then extended to include energy transport. It is experimentally observed that the particle diffusivity and the thermal diffusivity do not differ greatly and have roughly the same radial dependence. This is discussed and compared with the model. A brief discussion of the effect of internal transport barriers is included.

Recent progress towards an increased understanding of the physical processes in the divertor and scrape-off layer (SOL) plasmas in DIII-D has been made possible by a combination of new diagnostics, improved computational models and changes in divertor geometry. The work focused primarily on ELMing H mode discharges. The physics of partially detached divertor plasmas, with divertor heat flux reduction by divertor radiation enhancement using D2 puffing, was studied in two dimensions, and a model of the heat and particle transport was developed that includes conduction, convection, ionization, recombination and flows. Plasma and impurity particle flows were measured with Mach probes and spectroscopy and compared with the UEDGE model. The model now includes self-consistent calculations of carbon impurities. Impurity radiation was increased in the divertor and SOL with `puff and pump' techniques using SOL D2 puffing, divertor cryopumping and argon puffing. The important physical processes in plasma-wall interactions were examined with a DiMES (divertor material evaluation system) probe, plasma characterization near the divertor plate and the REDEP code. Experiments comparing single null plasma operation in baffled and open divertors demonstrated a change in the edge plasma profiles. These results are consistent with a reduction in the core ionization source calculated with UEDGE. Divertor particle control in ELMing H mode with pumping and baffling resulted in a reduction in H mode core densities to ne/nGr ≈ 0.25 (with nGr the Greenwald density). Divertor particle exhaust and heat flux were studied as the plasma shape was varied from a lower single null to a balanced double null, and finally to an upper single null.

A set of continuously gas fueled DIII-D [Plasma Physics and Controlled Fusion Research (International Atomic Energy Agency, Vienna, 1986) p. 159] high confinement mode (H-mode) shots at three neutral beam injection powers and two values of magnetic field (q95) have been analyzed to investigate the evolution of plasma edge conditions leading to the formation of X-point MARFEs (multifaceted asymmetric radiation from edge), which are followed by an H–L transition that constitutes a practical density limit for this class of shots in DIII-D. The MARFE initiation is found to be caused by a combination of a sharply increasing concentration of recycling neutrals in the plasma edge, decreasing edge inverse temperature and density gradient scale lengths, and increasing edge density. The line-average density at which MARFEs precipitates increases with increasing injection power and with decreasing q95. It is not clear whether the formation of the X-point MARFE causes or merely precedes the H–L transition.

A first-flight neutral transport model to describe the dependence of pump plenum neutral pressure on plasma parameters and magnetic geometry is presented. It is shown that the model is in excellent agreement with neutral pressure data from a low recycling DIII-D tokamak discharge. It is also shown that the main contribution to plenum pressure arises from the part of the ion particle flux profile which is closest to the plenum entrance. This work illustrates the sufficiency of a simple model in divertor plenum hardware design studies to maximize particle exhaust for density control.

We report new experimental results with the Radiative Divertor Project-outer baCe (RDP-OB) and cryopump in both upper single-null (USN) and double-null (DN) ELMing H-mode discharges. The baCed divertor reduced the core ionization (2‐2.5·), in reasonable agreement with predictions from UEDGE/DEGAS modeling (3.75·). The upper cryopump achieved density control of ne/ngw 0.22 (line density/Greenwald density) with Zeff 2 in high-d plasmas. The measured exhaust is comparable to the lower pump, except at lower core electron densities (ne < 5 · 10 19 m ˇ3 ). EAcient impurity exhaust was obtained with deuterium SOL flow. Preliminary experiments with DN operation has shown that the particle exhaust to the upper pump depends on the up/down magnetic balance. Preliminary experiments indicate that the DN exhaust is roughly 40‐50% of the USN exhaust at ne 4 · 10 19 m ˇ3 . ” 1999 Elsevier Science B.V. All rights reserved.

The role of neutrals in triggering the H–L back transition in high density ELMing H-mode plasmas is explored in double-null (DN) and single-null (SN) divertors. We propose that the neutral particle buildup below the X-point may play an important role in triggering the H–L transition at high density.

Experiments on DIII-D have demonstrated the efficacy of using induced scrape-off-layer (SOL) flow to preferentially enrich impurities in the divertor plasma. This SOL flow is produced through simultaneous deuterium gas injection at the midplane and divertor exhaust. Using this SOL flow, an improvement in enrichment (defined as the ratio of impurity fraction in the divertor to that in the plasma core) has been observed for all impurities in trace-level experiments (i.e., impurity level is non-perturbative), with the degree of improvement increasing with impurity atomic number. In the case of argon, exhaust gas enrichment using modest SOL flow is as high as 17. Using this induced SOL flow technique and argon injection, radiative plasmas have been produced that combine high radiation losses ( P rad / P input > 70%), low core fuel dilution ( Z eff τ E > 1.0 τ E,ITER93H ).

Pellet injection has been used on the DIII-D tokamak to study density limits and particle transport in H-mode and inner wall limited L-mode plasmas. These experiments have provided a variety of conditions to examine the fueling efficiency of pellets injected into DIII-D plasmas. The fueling efficiency, defined as the total increase in number of plasma electrons divided by the number of pellet fuel atoms, is determined by measurements of density profiles before and just after pellet injection. We have found that there is a decrease in the pellet fueling efficiency with increased neutral beam injection power. The pellet penetration depth also decreases with increased neutral beam injection power so that, in general, fueling efficiency increases with penetration depth. The fueling efficiency is generally 25% lower in ELMing H-mode discharges than in L-mode due to an expulsion of particles with a pellet triggered ELM. A comparison with fueling efficiency data from other tokamaks shows similar behavior.

We have studied the processes limiting plasma density and successfully achieved discharges with density ∼50% above the empirical Greenwald scaling with H-mode energy confinement. This was accomplished by density profile control, enabled through pellet injection and divertor pumping. Several density-limiting mechanisms, including MARFEs and pellet loss processes, were successfully avoided. Finally we have looked in detail at the first and most common density limit process in DIII-D, total divertor detachment, and found that the local upstream separatrix density n sep, det e at detachment onset (partial detachment) increases with the scrape-off layer heating power, Pheat, i.e. nsep,dete∼P0.8heat. This is in marked contrast to the line-average density at detachment which is insensitive to the heating power. Our data are in reasonable agreement with the Borass model, which predicted that the upstream density at detachment would increase as P0.7heat.

This paper summarizes results from a two-dimensional (2D) physics analysis of the transition to and stable operation of the partially detached divertor (PDD) regime induced by deuterium injection in DIII-D. The analysis [1] shows that PDD operation is characterized by a radiation zone near the X-point at -15 eV which reduces the energy flux into the divertor and thereby also reduces the target plate heat flux, an ionization zone below the X-point which provides a deuterium ion source to fuel parallel flow down the outer divertor leg, an ion-neutral interaction zone in the outer leg which removes momentum and energy from the flow and finally a volume recombination zone above the target plate which reduces the particle flux to the low levels measured on the plates and thereby also contributes to reduction in target plate heat flux.

In a magnetically diverted tokamak, the scrape-off layer (SOL) and divertor plasma separates the first wall from 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 we will 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. We will also discuss the results of experiments designed to influence the distribution of impurities in the plasma using enhanced SOL plasma flow. Extensive modelling 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.

A series of experiments was conducted on the DIII-D tokamak [J. L. Luxon and L. G. Davis, Fusion Technol. 8, 441 (1985)] to investigate the physical processes which limit density in high confinement mode (H-mode) discharges. The typical H-mode to low confinement mode (L-mode) transition limit at high density near the empirical Greenwald density limit [M. Greenwald et al., Nucl. Fusion 28, 2199 (1988)] was avoided by divertor pumping, which reduced divertor neutral pressure and prevented formation of a high density, intense radiation zone (MARFE) near the X-point. It was determined that the density decay time after pellet injection was independent of density relative to the Greenwald limit and increased nonlinearly with the plasma current. Magnetohydrodynamic (MHD) activity in pellet-fueled plasmas was observed at all power levels, and often caused unacceptable confinement degradation, except when the neutral beam injected (NBI) power was ⩽3 MW. Formation of MARFEs on closed field lines was avoided with lo...

Recent measurements of the two-dimensional (2-D) spatial profiles of divertor plasma density, temperature, and emissivity in the DIII-D tokamak [J. Luxon et al., in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion (International Atomic Energy Agency, Vienna, 1987), p. 159] under highly radiating conditions are presented. Data are obtained using a divertor Thomson scattering system and other diagnostics optimized for measuring the high electron densities and low temperatures in these detached divertor plasmas (ne⩽1021 m−3, 0.5 eV⩽Te). D2 gas injection in the divertor increases the plasma radiation and lowers Te to less than 2 eV in most of the divertor volume. Modeling shows that this temperature is low enough to allow ion–neutral collisions, charge exchange, and volume recombination to play significant roles in reducing the plasma pressure along the magnetic separatrix by a factor of 3–5, consistent with the measurements. Absolutely calibrated vacuum ultravi...

Electron density, temperature and parallel pressure measurements at several locations along field lines connecting the midplane scrapeoff layer (SOL) with the outer divertor are presented for both attached and partially-detached divertor cases: Ip = 1.4 MA, q95 = 4.2 and Pinput ∼ 6.7 MW under ELMing H-mode conditions. At the onset of the Partially Detached Divertor (PDD), a high density, low temperature plasma forms in the divertor SOL (divertor MARFE). The electron pressure drops by a factor of 2 between the midplane separatrix and the X-point and then an additional ∼ 3–5 times between the X-point and the outboard separatrix strike point. These results are in contrast to the attached (non-PDD) case, where electron pressure in the SOL is reduced by, at most, a factor of two between the midplane and the divertor target. Divertor MARFEs generally have only marginal adverse impact on important H-mode characteristics, such as confinement time. In fact, PDD discharges at low input power (i.e., approximately twice the L-H-mode threshold power) maintain good H-mode characteristics until a high density, low temperature plasma abruptly forms inside the separatrix near the X-point (X-point MARFE). Concurrent with the appearance of this X-point MARFE is a degradation in both energy confinement and the plasma fueling rate and an increase in the carbon impurity concentration inside the core plasma. The formation of the X-point MARFE is consistent with a thermal instability resulting from the temperature dependence of the carbon radiative cooling rate in the range ∼ 7–30 eV.

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.