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Experiments at DIII-D use different divertor geometries to measure the effect of divertor closure on detachment onset. We compare matched H-mode discharges in the closed (upper) divertor and the open (lower) divertor and use as a detachment onset metric the rollover in peak ion flux to the target, Jsat, as a function of upstream pedestal top density, ne,ped. Measurements of electron temperature in the lower divertor confirm that this metric captures the transition from attached to detached divertor for this data set. Experiments at two different neutral beam injection powers demonstrate an ≈ 20–40% decrease in pedestal density at detachment onset in the closed case compared to the open. High-resolution edge density and temperature measurements and power balance analysis allow us to determine the difference in upstream separatrix density ne,sep at detachment onset, which we find to be ≈ 15–25% lower in the closed divertor case. The effect of closure on detachment seen in separatrix density is hence only about 60–75% the effect seen in pedestal density, with the remainder due to differences in ne,sep/ne,ped. Future devices are expected to have a scrape off layer that is opaque to neutrals, and hence will not have the same pedestal fueling to affect ne,sep/ne,ped; this suggests that the effect of closure on detachment onset seen in future devices will be less than that seen in current experiments, and closer to the magnitude of the effect we see here in the separatrix for a similar degree of divertor closure. Keywords: Tokamak, Divertor, Detachment, Fusion

A refined version of the Fundamenksi-Moulton 'free-streaming' model (FSM) for the dynamics of divertor density, particle flux, and heat flux during edge localized modes (ELMs) is presented. This model depends only on inter-ELM pedestal and divertor conditions and, crucially, incorporates particle recycling: a FSM with recycling model, FSRM. The effective particle recycling coefficient, Reff, is the only empirical fitting parameter in the FSRM. The predictions of the FSRM are systematically tested against a DIII-D database of ELM ion and energy fluence measurements and are shown to be consistent with the model across a wide range of pedestal and divertor conditions using a constant value of 0.96 for Reff . Predictions for W sputtering during ELMs are developed based on the FSRM. It is concluded that energetic free-streaming D+ ions and C6+ impurities are the dominant contributors to the intra-ELM gross erosion of W in the DIII-D divertor, i.e., recycling ions and impurities have relatively little impact on the total W sputtering rate. These calculations are also shown to be consistent with spectroscopic measurements of W gross erosion for three different pedestal conditions after incorporating the strong electron density dependence of the WI 400.8 nm ionizations/photon (S/XB) coefficient. Keywords: Tungsten sputtering, Tungsten Erosion, WI spectroscopy, Edge Localized Modes, Recycling

The structure of the edge plasma in a magnetic confinement system has a strong impact on the overall plasma performance. We uncover for the first time a magnetic-field-direction dependent density shelf, i.e., local flattening of the density radial profile near the magnetic separatrix, in high confinement plasmas with low edge collisionality in the DIII-D tokamak. The density shelf is correlated with a doubly peaked density profile near the divertor target plate, which tends to occur for operation with the ion $B\ifmmode\times\else\texttimes\fi{}\ensuremath{\nabla}B$ drift direction away from the $X$-point, as currently employed for DIII-D advanced tokamak scenarios. This double-peaked divertor plasma profile is connected via the $E\ifmmode\times\else\texttimes\fi{}B$ drifts, arising from a strong radial electric field induced by the radial electron temperature gradient near the divertor target. The drifts lead to the reversal of the poloidal flow above the divertor target, resulting in the formation of the density shelf. The edge density shelf can be further enhanced at higher heating power, preventing large, periodic bursts of the plasma, i.e., edge-localized modes, in the edge region, consistent with ideal magnetohydrodynamics calculations.

We present measurements and modeling of aluminum erosion and redeposition experiments in separate helium and deuterium low power, low density L-mode plasmas at the outer divertor strike point of DIII-D to provide a low-Z material benchmark dataset for tokamak erosion-deposition modeling codes. Coatings of Al ∼100nm thick were applied to ideal (smooth) and realistic (rough) surfaces and exposed to repeat plasma discharges using the DiMES probe. Redeposition in all cases was primarily in the downstream toroidal field direction, evident from both in-situ spectroscopic and post-mortem non-spectroscopic measurements. The gross Al erosion yield was estimated from film thickness change measurements of small area samples, and was found to be ∼40–70% of the expected erosion yield based on theoretical physical sputtering yields after including sputtering by a 1–3% carbon impurity. The multi-step redeposition and re-erosion process, and hence the measured net erosion yield and material migration patterns, were found to be influenced by the surface roughness and/or porosity. A time-dependent model of material migration accounting for deposit accumulation in hidden areas was developed to reproduce the measurements in these experiments and determine a redeposition probability distribution function for sputtered atoms. Keywords: Erosion, Deposition, Migration, Roughness, Aluminum, Mixed materials

Detailed measurements of the ne, Te, and Ti profiles in the vicinity of the separatrix of ELMing H-mode discharges have been used to examine plasma stability at the extreme edge of the plasma and assess stability dependent models of the heat flux width. The results are strongly contrary to the critical gradient model, which posits that a ballooning instability determines a gradient scale length related to the heat flux width. The results of this analysis are not sensitive to the choice of location to evaluate stability. Significantly, it is also found that the results are completely consistent with the heuristic drift model for the heat flux width. Here the edge pressure gradient scales with plasma density and is proportional to the pressure gradient inferred from the equilibrium in accordance with the predictions of that theory.

Aluminum (Al) is a convenient proxy for beryllium (Be) plasma material interaction studies since they have a number of physical and chemical similarities. Al samples were exposed at the lower outer strike point of an L-mode divertor plasma in DIII-D (conditions 7–11 × 1018 D-ions cm−2 s−1, Te = 12–47 eV). The gross erosion rate was directly measured using post-mortem ion beam analysis of small 1 mm-sized samples where local re-deposition was determined to be negligible. The gross erosion rate was also calculated using spectroscopic methods, but these rates greatly underestimate the direct (i.e. non-spectroscopic) measurement. The direct measured erosion yields were within the range of published D+ → Al ion beam sputtering yields. The ionizations per photon (S/XB) coefficients used in the spectroscopic analysis were determined in separate experiments using He plasmas at the PISCES-B linear plasma facility at UCSD. The measured S/XB coefficients were on average ∼6× higher than the theoretically calculated values.

The OEDGE code is used to model the outer divertor plasma for discharges from a density scan experiment on DIII-D with the objective of assessing EIRENE and ADAS hydrogenic emission atomic physics data for D α , D β and D γ for values of T e and n e characteristic of the range of divertor plasma conditions from attached to weakly detached. Confidence in these values is essential to spectroscopic interpretation of any experiment or modeling effort. Good agreement between experiment and calculated emissions is found for both EIRENE and ADAS calculated emission profiles, confirming their reliability for plasma conditions down to ∼1 eV. For the cold dense plasma conditions characteristic of detachment, it is found that the calculated emissions are especially sensitive to T e .

Recent DIII-D experiments assessed the snowflake divertor (SF) configuration in a radiative regime in H-mode discharges with D 2 D 2 seeding. The SF configuration was maintained for many energy confinement times (2–3 s) in H-mode discharges ( I p =1.2 I p = 1.2 MA, P NBI =4–5 P NBI = 4 – 5 MW, and B×∇B B × ∇ B down (favorable direction toward the divertor)), and found to be compatible with high performance operation (H98y2 ⩾ 1). The two studied SF configurations, the SF-plus and the SF-minus, have a small finite distance between the primary X-point and the secondary B p B p null located in the private flux region or the common flux region, respectively. In H-mode discharges with the SF configurations (cf. H-mode discharges with the standard divertor with similar conditions) the stored energy lost per the edge localized mode (ELM) was reduced, and significant divertor heat flux reduction between and during ELMs was observed over a range of collisionalities, from lower density conditions toward a higher density H-modes with the radiative SF divertor.

ITER will use the shaped inner column as startup/rampdown limiter. An exploratory experiment was performed in DIII-D using a similarly shaped inner wall limiter, looking for evidence of a short fall-off feature, λshort, near the last closed flux surface. Measurements were made on the high field side (HFS) using infrared thermography and a Langmuir swing-probe. In some cases clear evidence was found for a narrow feature λ short ∼ ρ pol D + ∼ a few mm. The ratio q | | 0 short / q | | 0 long was mostly between 0.1 and 0.6. Measurements made on the low field side and bottom side showed little clear indication of any narrow feature.

Edge fluid–plasma/kinetic–neutral modeling of well-diagnosed DIII-D experiments is performed in order to document in detail how well certain aspects of experimental measurements are reproduced within the model as the transition to detachment is approached. Results indicate, that at high densities near detachment onset, the poloidal temperature profile produced in the simulations agrees well with that measured in experiment. However, matching the heat flux in the model requires a significant increase in the radiated power compared to what is predicted using standard chemical sputtering rates. These results suggest that the model is adequate to predict the divertor temperature, provided that the discrepancy in radiated power level can be resolved.

Upgrades to core and divertor Thomson scattering (DTS) diagnostics at DIII-D have provided measurements of electron pressure profiles in the lower divertor from attached- to fully-detached divertor plasma conditions. Detailed, multistep sequences of discharges with increasing line-averaged density were run at several levels of Pinj. Strike point sweeping allowed 2D divertor characterization using DTS optimized to measure Te down to 0.5 eV. The ionization front at the onset of detachment is found to move upwards in a controlled manner consistent with the indication that scrape-off layer parallel power flux is converted from conducted to convective heat transport. Measurements of ne, Te and pe in the divertor versus Lparallel demonstrate a rapid transition from Te ⩾ 15 eV to ⩽3 eV occurring both at the outer strike point and upstream of the X-point. These observations provide a strong benchmark for ongoing modeling of divertor detachment for existing and future tokamak devices.

Simulations using the multi-fluid code UEDGE indicates that, in low confinement (L-mode) plasmas in DIII-D, poloidal projection of the ionization driven flows dominate poloidal particle flows in the divertor near the divertor plates, whereas E × B drift flows dominate the radial particle flows. In contrast, in high confinement (H-mode) conditions E × B drift flows dominate both poloidal and radial particle flows in the divertor in the vicinity of the strong gradient region near the separatrix. UEDGE indicates that the toroidal C2+ flow velocities in the divertor plasmas are mainly entrained within 30% to the background deuterium flow in both L- and H-mode plasmas in the plasma region where the CIII 465 nm emission is measured. Therefore, UEDGE indicates that the Doppler Coherence Imaging Spectroscopy (CIS), measuring the toroidal velocity of the C2+ ions, can provide insight to the deuterium flows in the divertor. Parallel-to-B velocity dominates the toroidal divertor flow; direct drift impact being less than 1%. Toroidal divertor flow is predicted to reverse when the magnetic field is reversed. This is explained by the parallel-B flow towards the nearest divertor plate corresponding to opposite toroidal directions in opposite toroidal field configurations.

Joint DIII-D/EAST experiments in the radio-frequency (RF) heated H-mode scheme with comparison to that of neutral beam (NB) heated H-mode scheme were carried out on DIII-D and EAST under similar conditions to examine the effect of heating scheme on scrape-off layer (SOL) width in H-mode plasmas for application to ITER. A dimensionally similar plasma equilibrium was used to match the EAST shape parameters. The divertor heat flux and SOL widths were measured with infra-red camera in DIII-D, while with divertor Langmuir probe array in EAST. It has been demonstrated on both DIII-D and EAST that RF-heated plasma has a broader SOL than NB-heated plasma when the edge electrons are effectively heated in low plasma current and low density regime with low edge collisionality. Detailed edge and pedestal profile analysis on DIII-D suggests that the low edge collisionality and ion orbit loss effect may account for the observed broadening. The joint experiment in DIII-D has also demonstrated the strong inverse dependence of SOL width on the plasma current in electron cyclotron heated (ECH) H-mode plasmas. Keywords: SOL width, Radio-frequency heating, Neutral beam heating, Low collisionality

The OEDGE code is used to model core fueling for attached L-mode plasmas and between edge localized modes (ELMs) for attached H-mode plasmas in DIII-D. Empirical plasma reconstruction has been used to determine the plasma conditions in these discharges. EIRENE is used to model the hydrogen recycling. Divertor recycling accounts for 65–100% of the core fueling. The fraction of the total divertor target flux ionized inside the separatrix ranges from 5% to 20%. The fraction of total wall flux ionized inside the separatrix ranges from 20% to 50%. Neutrals originating from wall regions closer to the separatrix are more likely to ionize in the confined plasma. Ionization in the confined plasma is concentrated below the midplane with peaks in the poloidal profiles just above the X -point. Radial core ionization in high density H-mode is peaked strongly near the separatrix.

This paper documents first work toward validation of BOUT++ nonlinear edge localized mode (ELM) simulations in X-point geometry, at experimental pedestal collisionality, against multiple diagnostic measurements of a well-characterized ELM event in DIII-D. The key to the BOUT++ simulations is the use of a hyper-resistivity model that effectively spreads the very thin current sheets that form in low collisionality nonlinear simulations, and allows for ELM driven magnetic reconnection at finite current density. Experimental ELM characterization includes multiple fast line-integrated diagnostic measurements revealing in–out divertor asymmetric response to ELMs, IRTV imaging at the divertor targets, visible emission in the divertor volume to test the extension of BOUT++ to X-point geometry, and forward modeling of new electron cyclotron emission imaging to test predictions of ELM filaments in the edge pedestal. Initial comparisons suggest optimized BOUT boundary conditions and model parameters, and show similarities between initial BOUT++ results and several measurements.

Pedestal fueling through edge recycling is examined with the interpretive OEDGE code for high-density discharges in DIII-D. A high current, high-density discharge is found to have a similar radial ion flux profile through the pedestal to a lower current, lower density discharge. The higher density discharge, however, has a greater density gradient indicating a pedestal particle diffusion coefficient that scales near linear with 1/Ip. The time dependence of density profile is taken into account in the analysis of a discharge with low frequency ELMs. The time-dependent analysis indicates that the inferred neutral ionization source is inadequate to account for the increase in the density profile between ELMs, implying an inward density convection, or density pinch, near the top of the pedestal.

An experiment aimed at benchmarking the ITER scrape-off layer (SOL) power width scaling in limited L-mode discharges has been conducted on DIII-D. Scans of the main scaling parameters were performed in an inner-wall-limited (IWL) magnetic configuration. Using the near-SOL density and temperature e-folding lengths, λ n , λ T , determined from reciprocating Langmuir probe measurements, SOL power flux density e-folding lengths, λ q , are derived. A few lower single null (LSN) discharges were also run for comparison. The results are generally in agreement with the ITER design assumptions, finding that λ n and λ T are correlated ( λ T ∼ 1.2 λ n ) and both λ n and λ T are on average 2.1–2.5 times larger in IWL configurations than in LSN. In moderate elongation (κ ∼ 1.4) IWL discharges, λ q is largest and agrees with the assumed ITER scaling within the estimated uncertainty (a factor of ∼2). In IWL discharges λ q measurements are consistent with the expectations of SOL power balance.

Unbalanced double null ELMy H-mode configurations in DIII-D are used to simulate the situation in ITER high triangularity, burning plasma magnetic equilibria, where the second X-point lies close to the top of the vacuum vessel, creating a secondary divertor region at the upper blanket modules. The measured plasma conditions in the outer secondary divertor closely duplicated those projected for ITER. 13 CH 4 was injected into the secondary outer divertor to simulate sputtering there. The majority of the 13 C found was in the secondary outer divertor. This material migration pattern is radically different than that observed for main wall 13 CH 4 injections into single null configurations where the deposition is primarily at the inner divertor. The implications for tritium codeposition resulting from sputtering at the secondary divertor in ITER are significant since release of tritium from Be co-deposits at the main wall bake temperature for ITER, 240 °C, is incomplete. The principal features of the measured 13 C deposition pattern have been replicated by the OEDGE interpretive code.

Experiments are performed on the DIII-D tokamak to determine the retention rate in an all graphite first-wall tokamak. A time-dependent particle balance analysis shows a majority of the fuel retention occurs during the initial Ohmic and L-mode phase of discharges, with peak fuel retention rates typically ∼2 × 1021 D/s. The retention rate can be zero within the experimental uncertainties ( ⩽ 0.6 - 1.2 × 10 20 D/s. Dynamic particle balance is also performed on discharges with resonant magnetic perturbation ELM suppression and shows no additional retention during the ELM-suppressed phase of the discharge.

Recent DIII-D experiments show that both magnetic balance and particle drifts are important to understanding how argon impurities accumulate in balanced and unbalanced double-null plasmas during ‘puff and pump’ radiating divertor operation. Unbalanced double-null shapes, which are biased in the direction opposite to the ion B × ∇ B drift direction, have produced the best result to-date in terms of coupling a radiating divertor approach with an H-mode plasma. The proximity to balanced double-null that can be used for puff-and-pump may depend on the width of the heat flux profile in the scrape-off layer. A comparison of plasma behavior in the open lower divertor to that of the more closed upper divertor determined that plasma density control was far more sensitive to the ion B × ∇ B drift direction than to the relative closure of the divertor.

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

Spectroscopy of high-n Balmer line transitions provides a means of measuring ne and Te in recombining plasmas [J. L. Terry et al., Phys. Plasmas 5, 1579 (1998)]. The relative intensities of Rydberg series lines near the ionization limit are a sensitive diagnostic of Te, for Te < 1.5 eV. Stark broadening of these same lines provides a measure of local ne and, with less accuracy, of Te. The accuracy of different theoretical models for Stark broadening [H.R. Griem, Spectral Line Broadening by Plasmas (Academic, New York, 1974); E. Oks, Stark Broadening of Hydrogen and Hydrogenlike Spectral Lines in Plasmas: The Physical Insight (Alpha Science International, Oxford, UK, 2006)] is evaluated by comparing values of ne and Te measured on DIII-D by divertor Thomson scattering (DTS) with those deduced from spectral profile analysis of Balmer series deuterium lines. In particular, the detailed dependence of line width on principal quantum number provides a sensitive metric for distinguishing which model best accords with experiment.

Use of carbon in tokamaks leads to a serious tritium retention issue due to co-deposition. To further investigate the processes involved, a detached ELMy H-mode (6.5 MW NBI) experiment was performed on DIII-D in which {sup 13}CH{sub 4} was puffed into the main vessel through the toroidally symmetric pumping plenum at the top of lower single-null discharges. Subsequently, the {sup 13}C content of tiles taken from the vessel wall was measured. The interpretive OEDGE code was used to model the results. The {sup 13}C deposition pattern could be reproduced, in general shape and magnitude, by assuming in the code the existence of a parallel flow and a radial pinch in the scrape-off layer. Parallel flows of Mach {approx} 0.3 toward the inner divertor and a radial pinch {approx}10 to 20 m/s (+ R-direction) were found to yield {sup 13}C deposition comparable to the experiment.

The evolution and performance limits for the pedestal in H-mode are dependent on the two main drive terms for instability: namely the edge pressure gradient and the edge current density. These terms are naturally coupled though neoclassical (Pfirsch-Schluter and bootstrap) effects. On DIII-D, local measurements of the edge current density are made using an injected lithium beam in conjunction with Zeeman polarimetry and compared with pressure profile measurements made with other diagnostics. These measurements have confirmed the close spatial and temporal correlation that exists between the measured current density and the edge pressure in H- and QH-mode pedestals, where substantial pressure gradients exist. In the present work we examine the changes in the measured edge current for DIII-D pedestals which have a range of values for the ion and electron collisionalities {v i *, v* e } due to fuelling effects. Such changes in the collisionality in the edge are expected to significantly alter the level of the bootstrap current from the value predicted from the collisionless limit and therefore should correspondingly alter the pedestal stability limits. We find a clear decrease in measured current as v increases, even for discharges having similar edge pressure gradients.

The quiescent H (QH) mode, an edge localized mode (ELM)-free and high-confinement mode, combines well with an internal transport barrier to form quiescent double barrier stationary state, high performance plasmas. The QH-mode edge pedestal pressure is similar to that seen in ELMing phases of the same discharge, with similar global energy confinement. The pedestal density in early ELMing phases of strongly pumped counter injection discharges drops, and a transition to QH-mode occurs, leading to lower calculated edge bootstrap current. The plasmas current ramp experiment and ELITE code modelling of edge stability suggest that QH-modes lie near an edge current stability boundary. At high triangularity, QH-mode discharges operate at higher pedestal density and pressure and have achieved ITER level values of βPED and ν*.

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.

Experiments are described that have increased our understanding of the transport and stability physics that set the H-mode edge pedestal width and height, determine the onset of Type-I edge localized modes (ELMs) and produce the nonlinear dynamics of the ELM perturbation in the pedestal and scrape-off layer (SOL). Models now exist for the ne pedestal profile and the pe height at the onset of Type-I ELMs, and progress has been made towards models of the Te pedestal width and nonlinear ELM evolution. Similarity experiments between DIII-D and JET suggested that neutral penetration physics plays an important role in the relationship between the width and height of the ne pedestal. Plasma physics appears to dominate in setting the Te pedestal width. Measured pedestal conditions including edge current at ELM onset agree with intermediate-n peeling?ballooning (P?B) stability predictions. Midplane ELM dynamics data show the predicted (P?B) structure at the ELM onset, large rapid variations of the SOL parameters and fast radial propagation in later phases, similar to features in nonlinear ELM simulations.

High temporal and spatial resolution measurements in the boundary of the DIII-D tokamak show that edge localized modes (ELMs) are composed of fast bursts of hot, dense plasma that travel radially starting at the separatrix at {approx}450 m/s and rotate in the scrape off layer (SOL), convecting particles and energy to the SOL and walls. The temperature and density in the ELM plasma initially correspond to those at the top of the density pedestal but decay with radius in the SOL. The temperature decay length ({approx}1.2-1.5 cm) is much shorter than the density decay length ({approx}3-8 cm), which in turn decreases with increasing pedestal density. The local particle and energy flux at the wall during the bursts are 10-50% ({approx} 1-2 x 10{sup 21} m{sup -2} s{sup -1}) and 1-2% ({approx} 20-30 kW/m{sup 2}) respectively of the LCFS average fluxes, indicating that particles are transported radially much more efficiently than heat.

Measurements and modeling of the 2D poloidal D α intensity distribution in DIII-D low density L-mode and medium density ELMy H-mode plasmas indicate that the core plasma is predominately fueled near the divertor X-point region. The neutral deuterium and ion carbon emission were measured in the divertor and inner main chamber scrape-off layer (SOL) using a plasma imaging technique, covering 85% of the poloidal cross-section. Typically, the peak emission in the inner main SOL at the tokamak midplane was three orders of magnitude lower than in the divertor. For discharges with the ion B × ∇ B drift direction toward the lower divertor the UEDGE/DEGAS codes predict strong core plasma fueling from the significantly higher density and lower temperature plasma calculated in the inner divertor leg. The concomitant carbon ion flow reversal in the inner divertor leg enhances the leakage of carbon from the divertor into the main SOL, and hence into the core.

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).

Edge localized mode (ELM) effects in the DIII-D pedestal and boundary plasmas were measured with multiple fast diagnostics in matched, lower single-null, ELMing H-mode discharges with the ion B × ∇B drift either toward or away from the divertor. Data show a strong dependence of the delay in inner vs. outer divertor ELM Dα on drift direction, and a weaker drift dependence of the inner vs. outer delay of Prad, in addition to the strong density dependence seen in previous work. Time dependent boundary plasma modeling during an ELM was done with the UEDGE code including a six-species fluid carbon model and the effect of B-field induced particle drifts. The ELM was modeled as an instantaneous, outer midplane peaked, increase of diffusion coefficients from the top of the pedestal to the SOL. The simulations show delays in the inner vs. outer divertor ELM transients that are similar to the measured delays at moderate density.

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...

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.

Changes in the divertor magnetic balance in DIII-D H-mode plasmas affects core, edge, and divertor plasma behavior. Both the pedestal density ne,PED and plasma stored energy WT were sensitive to changes in magnetic balance near the double-null (DN) configuration, e.g., both decreased 20–30% when the DN shifted to a slightly unbalanced DN, where the B ×∇B drift direction pointed away from the main X-point. Recycling at each of the four divertor targets was sensitive to changes in magnetic balance and the B ×∇B drift direction. The poloidal distribution of the recycling in DN is in qualitative agreement with the predictions of UEDGE modeling with particle drifts included. The particle flux at the inner divertor target is shown to be much more sensitive to magnetic balance than the particle flux at the outer divertor target near the DN shape. These results suggest possible advantages and drawbacks for balanced DN operation.

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 quiescent high confinement (QH-mode) and quiescent double barrier (QDB) modes in DIII-D have long-duration H-mode confinement without ELMs, possibly an alternative operating mode in future tokamaks for avoiding damage by ELMs. Instead of ELMs, there is an edge harmonic oscillation (EHO), which is a continuous electromagnetic mode with associated density fluctuations. The edge pedestal is similar to ELMing H-mode, but at very low density to date. We see C 6+ ion temperatures of 3–7 keV in scrape-off layer (SOL), 100 kV/m radial electric fields just inside the separatrix, and a hot area on the divertor baffle whose heating correlates with the presence of the EHO. We attribute the baffle heating to perturbation of trapped ion orbits by the EHO, allowing particles to strike the baffle. The outboard SOL is wider than the inboard, probably for lack of trapped ions on the inside.

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.

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.

DIII-D has recently demonstrated improved energy confinement by injecting neutral deuterium gas into high performance near-double null divertor (DND) plasmas during high power operation. Representative parameters for these plasmas are: q 95 = 6, P IN up to 15 MW, H 98 = 1.4–1.8, and β N = 2.5–4.0. The ion B × B direction is away from the primary X-point. While plasma conditions at lower to moderate power input (e.g. ≤11 MW) are shown to be favorable to successful puff-and-pump radiating divertor applications, particularly when using argon seeds, plasma behavior at higher powers (e.g. ≥14 MW) may make successful puff-and-pump operation more problematic. In contrast to lower powered high performance plasmas, both τ E and β N in the high power cases (≥14 MW) increased and ELM frequency decreased, as density was raised by deuterium gas injection. Improved performance in the higher power plasmas was tied to higher pedestal pressure, which according to peeling-ballooning mode stability analysis using the ELITE code could increase with density along the kink/peeling stability threshold, while the pedestal pressure gradient in the lower power discharges were limited by the ballooning threshold. This resulted in improved fueling efficiency and ≈10% higher τ E and β N than is normally observed in comparable high performance plasmas on DIII-D. Applying the puff-and-pump radiating divertor approach at moderate versus high power input is shown to result in a much different evolution in core and pedestal plasma behavior. We find that injecting deuterium gas into these highly powered DND plasmas may open up a new avenue for achieving elevated plasma performance, including better fueling, but the resulting higher density may also complicate application of a radiating divertor approach to heat flux reduction in present-day tokamaks, if scenarios involving second-harmonic electron cyclotron heating are used.

The effects of changes in core density on divertor electron temperature, density and heat flux when resonant magnetic perturbations (RMPs) are applied are presented, notably a reduction in RMP induced secondary radial peaks in the electron temperature profile at the target plate is observed when the core density is increased, which is consistent with modeling. RMPs is used here to indicate non-axisymmetric magnetic field perturbations, created using in-vessel control coils, which have at least one but typically many resonances with the rotational transform of the plasma (Evans et al 2006 Phys. Plasmas 13 056121). RMPs are found to alter inter-ELM heat flux to the divertor by modifying the core plasma density. It is shown that applying RMPs reduces the core density and increases the inter-ELM heat flux to both the inner and outer targets. Using gas puffing to return the core density to the pre-RMP levels more than eliminates the increase in inter-ELM heat flux, but a broadening of the heat flux to the outer target remains. These measurements were made at a single toroidal location, but the peak in the heat flux profile was found near the outer strike point where simulations indicate little toroidal variation should exist and tangentially viewing diagnostics showed no evidence of strong asymmetries. In experiments where divertor Thomson scattering measurements were available it is shown that local secondary peaks in the divertor electron temperature profile near the target plate are reduced as the core density is increased, while peaks in the divertor electron density profile near the target are increased. These trends observed in the divertor electron temperature and density are qualitatively reproduced by scanning the upstream density in EMC3-Eirene modeling. Measurements are presented showing that higher densities are needed to induce detachment of the outer strike point in a case where an increase in electron temperature, likely due to a change in MHD activity, is seen after RMPs are applied.

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.

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.

This paper examines the sensitivity of edge pedestal and divertor performance parameters to changes in the volume of a secondary divertor that tends to form inside the vacuum chamber in high triangularity plasma shapes. The sensitivity was examined by varying the vertical distance of the secondary X -point from the target plate, Z S X , while holding the primary X -point height fixed. The ion ∇ B drift was in the direction of the primary divertor. Discharges with and without active primary divertor cryo-pumping were examined. The effective rate of rise of the core density at the L–H transition increased 80% as Z S X was reduced. At high density achieved by gas injection, the core line-averaged density at the H–L back transition decreased by 30% as Z S X was reduced. Both of these results indicate that performance may be affected when core plasma screening of neutrals in the secondary divertor is reduced as Z S X decreases. The peak heat flux in the secondary divertor decreased as Z S X was reduced due to increased flux expansion until a threshold was reached at Z S X ∼3 cm where it remained constant. This indicated that the high recycling character of the plasma above the target in the secondary divertor was lost when the length of the outer divertor leg was reduced below a critical value.

Scrape-off layer plasma simulation modeling has been carried out to assess the effect of tightly closing the lower divertor in DIII-D, which at present is almost fully open, on the achievement of cold dissipative/detached divertor conditions. To isolate the impact of other factors on the divertor plasma solution and to make direct comparisons, most of the parameters including the meshes were kept as similar as possible. Only the neutral baffling was modified to compare a fully open divertor with a tightly closed one. The modeling shows that the tightly closed divertor greatly improves trapping of recycling neutrals, thereby increasing radiative and charge exchange losses in the divertor and reducing the electron temperature T et and deposited power density q dep at the target plate. Furthermore, the closed structure enables the divertor plasma to enter into highly dissipative and detached divertor conditions at a significantly lower upstream density. The effects of divertor closure on the neutral density and pressure, and their correlation with the divertor plasma conditions are also demonstrated. The effect of molecular D2−ion D+ elastic collisions and neutral–neutral collisions on the divertor plasma solution are assessed.

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.

Experimentally measured carbon line emissions and total radiated power distributions from the DIII-D divertor and Scrape-Off Layer (SOL) are compared to those calculated with the Monte Carlo Impurity (MCI) model. A UEDGE background plasma is used in MCI with the Roth and Garcia-Rosales (RG-R) chemical sputtering model and/or one of six physical sputtering models. While results from these simulations do not reproduce all of the features seen in the experimentally measured radiation patterns, the total radiated power calculated in MCI is in relatively good agreement with that measured by the DIII-D bolometric system when the Smith78 physical sputtering model is coupled to RG-R chemical sputtering in an unaltered UEDGE plasma. Alternatively, MCI simulations done with UEDGE background ion temperatures along the divertor target plates adjusted to better match those measured in the experiment resulted in three physical sputtering models which when coupled to the RG-R model gave a total radiated power that was within 10% of measured value.

Characteristics of the H-mode pedestal are studied in Type I ELM discharges with ITER cross-sectional shape and aspect ratio. The scaling of the width of the edge steep gradient region, δ , which is most consistent with the data are with the normalized edge pressure, ( β PED POL ) 0.4 . Fits of δ to a function of temperature, such as ρ POL , are ruled out in divertor pumping experiments. The edge pressure gradient is found to scale as would be expected from infinite n ballooning mode theory; however, the value of the pressure gradient exceeds the calculated first stable limit by more than a factor of 2 in some discharges. This high edge pressure gradient is consistent with access to the second stable regime for ideal ballooning for surfaces near the edge. In lower q discharges, including discharges at the ITER value of q , edge second stability requires significant edge current density. Transport simulations give edge bootstrap current of sufficient magnitude to open second stable access in these discharges. Ideal kink analysis using current density profiles including edge bootstrap current indicate that before the ELM these discharges may be unstable to low n , edge localized modes.

Under certain conditions, high temperature electrons diffusing across the separatrix near the midplane can propagate to the divertor without collision and produce a high energy tail in the electron distribution function near the separatrix. We evaluate a kinetic criterion along the field using measurements from divertor Thomson scattering and compare temperature measurements from target plate Langmuir probes to estimate the magnitude of the high energy tail. A Fokker–Planck kinetic model [O. Batishchev, Phys. Plasmas 4 (1998) 1672], which includes electron–electron, electron-ion, electron-neutral, and charge exchange interactions is used to estimate the size of the high energy tail. For the cases considered, we found less than 1% of non-thermal particles near the target plate at typical densities and power levels for the DIII-D scrape-off layer plasma. ELMs violate our kinetic criterion.