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A modeling study is reported using new 2D data from DIII-D tokamak divertor plasmas and improved 2D transport model that includes large cross-field drifts for the numerically difficult low anomalous transport regime associated with the H-mode. The data set, which spans a range of plasma densities for both forward and reverse toroidal magnetic field (Bt), is provided by divertor Thomson scattering (DTS). Measurements utilizing X-point sweeping give corresponding 2D profiles of electron temperature (Te) and density (ne) across both divertor legs for individual discharges. The simulations focus on the open magnetic field-line regions, though they also include a small region of closed field lines. The calculations show the same features of in/out divertor plasma asymmetries as measured in the experiment, with the normal Bt direction (ion ∇B drift toward the X-point) having higher ne and lower Te in the inner divertor leg than outer. Corresponding emission data for total radiated power shows a strong inner-divertor/outer-divertor asymmetry that is reproduced by the simulations. These 2D UEDGE transport simulations are enabled for steep-gradient H-mode conditions by newly implemented algorithms to control isolated grid-scale irregularities.

PLASMA-SURFACE INTERACTIONS 21 — Proceedings of the 21st International Conference on Plasma-Surface Interactions in Controlled Fusion Devices Kanazawa, Japan May 26-30, 2014 The triggering of rapid small edge localized modes (ELMs) by high frequency pellet injection has been proposed as a method to prevent large naturally occurring ELMs that can erode the ITER plasma facing components (PFCs). Deuterium pellet injection has been used to successfully demonstrate the ondemand triggering of edge localized modes (ELMs) at much higher rates and with much smaller intensity than natural ELMs. The proposed hypothesis for the triggering mechanism of ELMs by pellets is the local pressure perturbation resulting from reheating of the pellet cloud that can exceed the local high-n ballooning mode threshold where the pellet is injected. Nonlinear MHD simulations of the pellet ELM triggering show destabilization of high-n ballooning modes by such a local pressure perturbation. A review of the recent pellet ELM triggering results from ASDEX Upgrade (AUG), DIII-D, and JET reveals that a number of uncertainties about this ELM mitigation technique still remain. These include the heat flux impact pattern on the divertor and wall from pellet triggered and natural ELMs, the necessary pellet size and injection location to reliably trigger ELMs, and the level of fueling to be expected from ELM triggering pellets and synergy with larger fueling pellets. The implications of these issues for pellet ELM mitigation in ITER and its impact on the PFCs are presented along with the design features of the pellet injection system for ITER. 2014 Elsevier B.V. All rights reserved. 1. Introduction The H-mode confinement regime of plasma operation is planned for ITER in order to achieve high fusion performance. It is characterized by a steep pressure gradient and ‘‘pedestal’’ at the plasma edge that is expected to lead to the quasi-periodic instability of edgelocalized modes (ELMs) [1]. ELMs expel periodic bursts of particles and energy from the plasma, which if large enough can pose a serious threat to the PFCs by erosion and melting from the high heat fluxes and

Recent experiments on DIII-D have increased confidence in the ability to suppress edge-localized modes (ELMs) using edge-resonant magnetic perturbations (RMPs) in ITER, including an improved physics basis for the edge response to RMPs as well as expansion of RMP ELM suppression to more ITER-like conditions. Complete ELM suppression has been achieved utilizing n = 3 RMPs in the ITER baseline scenario. In addition, RMP ELM suppression has been expanded to include plasmas with helium concentrations near 25% and the use of n = 2 RMPs. Analysis of the kinetic profile response suggests that ELM suppression is correlated with the co-alignment of the ω⊥e = 0 location, an n = 3 rational surface, and the top of the pedestal. Modelling predicts that such a co-alignment could potentially lead to island (or island chain) formation just inside the top of the pedestal, inhibiting the growth of the pedestal and thereby maintaining the ELM-free state. Detailed analysis of data obtained during toroidal phase variations of the applied n = 3 RMPs have provided further evidence of an island-like structure at the top of the pedestal. In addition, nearly matched discharges with co-neutral-beam-injection (co-NBI) and counter-NBI have demonstrated the importance of the presence of the ω⊥e = 0 location for ELM suppression. In the counter-NBI cases, the toroidal rotation profile is such that there is no ω⊥e = 0 location and ELMs are not suppressed in conditions in which ELM suppression is generally observed with co-NBI.

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.

Using resonant magnetic perturbations with toroidal mode number n = 3, we have produced H-mode discharges without edge localized modes (ELMs) which run with constant density and radiated power for periods up to about 2550 ms (17 energy confinement times). These ELM suppression results are achieved at pedestal collisionalities close to those desired for next step burning plasma experiments such as ITER and provide a means of eliminating the rapid erosion of divertor components in such machines which could be caused by giant ELMs. The ELM suppression is due to an enhancement in the edge particle transport which reduces the edge pressure gradient and pedestal current density below the threshold for peeling-ballooning modes. These n = 3 magnetic perturbations provide a means of active control of edge plasma transport.

Large 70 Hz Type-I edge localized modes (ELMs) are converted into small 130 Hz oscillations using edge resonant magnetic perturbations (RMPs) from a coil with currents ⩽0.4% Ip in double null DIII-D plasmas. When the RMP is properly phased with respect to the background field errors, all but a few isolated ELM-like events are suppressed. The impulsive pedestal energy loss ΔEELM/Δt1/2 to the scrape-of layer is reduced a factor of ⩾20 relative to the Type-I ELMs and the core confinement is unaffected by the perturbation field. Significant changes in the properties of the ELMs are also observed when edge RMPs are applied to lower single null plasmas but the nature of these changes are much more complex. Both lower single null and double null plasmas are being studied to determine how ELM control techniques based on the application of edge RMPs can be expected to scale to future devices such as ITER.

Recently a number of major, unanticipated effects have been reported in tokamak edge research raising the question of whether we understand the controlling physics of the edge. This report is on the first part – here focused on the outer divertor – of a systematic study of the simplest possible edge plasma – no ELMs, no detachment, etc. – for a set of 10 repeat, highly diagnosed, single-null, divertor discharges in DIII-D. For almost the entire, extensive data set so far evaluated, the matches of experiment and model are so close as to imply that the controlling processes at the outer divertor for these simple plasma conditions have probably been correctly identified and quantitatively characterized in the model. The principal anomaly flagged so far relates to measurements of Te near the target, potentially pointing to a deficiency in our understanding of sheath physics in the tokamak environment.

Modeling of transient impurity transport during large ELMs is used to explore basic processes which may determine ELM-averaged enrichment. The b2-Eirene code (solps4), used for DIII-D geometry, suggests that a complex sequence can occur during an ELM cycle in which a transiently detached phase, with relatively low enrichment, can occur even under nominally attached conditions. A slower recovery phase then follows, in which the effect of induced scrape-off layer flows can increase in importance. The model results are compared with available fast time-scale measurements. The observed increased enrichment with higher Z is similar to trends in basic particle reflection properties. Neon recycling processes may thus introduce a significant history effect, as illustrated by analysis of continuous, unforced neon accumulation in a DIII-D discharge with a well-characterized operational history.

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.

The UEDGE modeling code has been used to study the effect of varying the carbon yield from the plasma facing surfaces on the core plasma carbon contamination in DIII-D. The model of the lower single-null, ELMing H-mode plasma shows a remarkably weak dependence of the core carbon concentration over an approximate factor of two variation in the source. This weak dependence is in agreement with the analysis of spectroscopic data from DIII-D [1]. Examination of the carbon transport shows a general flow pattern of carbon as follows: (1) parallel flow from the divertors to the near scrape off layer (SOL) near the separatrix, (2) cross field diffusion from the near SOL to the far SOL (near the wall), and (3) parallel flow from the far SOL to the far region of the inner divertor. The carbon flux from the divertors to the near SOL drops as the sputtering rate is reduced. In the far SOL, background plasma parameters adjust in small ways to produce an increasing carbon density with decreasing sputtering yield. This increasing density of carbon in the far SOL is consistent with a reduction in the parallel velocity of carbon ions flowing from the far SOL back to the inner divertor. Since the carbon density near the separatrix is constant as the sputtering yield is reduced, the increasing density in the far SOL reduces the radial gradient and therefore the diffusive radial flow. A balance in the outward radial diffusive flow from the near SOL and the flow from the divertor into the near SOL maintains the carbon density in the near SOL nearly constant, even though the carbon throughput changes.

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.

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.

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.

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 comprehensive set of experiments has been carried out on the DIII-D tokamak to measure the 2-D (R,Z) structure of the divertor plasma in a systematic way using new diagnostics. Measurements cover the divertor radially from inside the X-point to the outer target plate and vertically from the target plate to above the X-point. Identical, repeatable shots were made, each having radial sweeps of the X-point and divertor strike points, to allow complete plasma and radiation profile measurements. Data have been obtained in ohmic, L-mode, ELMing H-mode, and reversed B{sub T} operation ({gradient}B drift away from the X-point). In addition, complete measurements were made of radiative divertor plasmas with a Partially Detached Divertor (PDD) induced by D{sub 2} injection and with a Radiating Mantle induced by Impurity injection (RMI) using neon and nitrogen. The data set includes first observations of the radial and poloidal profiles of the X-point, inner and outer leg plasmas in PDD and RMI radiative divertor operation. Preliminary data analysis shows that intrinsic impurities play a critical role in determining the SOL and divertor conditions.

Recent experiments on DIII-D have demonstrated that having a toroidally-monochromatic spectral content of edge-resonant magnetic perturbations (RMPs) is not a necessary condition for suppression of edge localized modes (ELMs). Robust ELM suppression has been reproducibly obtained on DIII-D during experiments in which various non-axisymmetric coil loops were turned off pseudo-randomly producing a variety of n=1 ?> , n= 2 ?> , and n= 3 ?> spectral contributions. It was shown that RMP ELM suppression could be achieved with as few as 5 out of 12 internal coil loops (I-coils) on DIII-D at similar coil currents and with good plasma confinement. Linear MHD plasma response (m3dc1, ipec, mars) and vacuum (surfmn, trip3d) modelling have been performed in order to understand the effects of the perturbation spectrum on the plasma response and ELM suppression. The results suggest that reduction of the dominant n= 3 ?> perturbation field is compensated by increased n= 2 ?> field in the plasma that may lead to RMP ELM suppression at lower levels of n= 3 ?> perturbative magnetic flux from the I-coils. These results provide additional confidence that ITER may be capable of RMP ELM suppression in the event of multiple internal coil failures.

This paper examines the extensions of the physics and engineering guidelines for the International Thermonuclear Experimental Reactor (ITER) device needed for acceptable operating points for a steady-state tokamak power reactor. Noninductive current drive is provided in steady state by high-energy neutral beam injection in the plasma core, lower hybrid slow waves in the outer regions of the plasma, and bootstrap current. Three different levels of extension of the ITER physics/engineering guidelines, with differing assumptions on the possible plasma beta, elongation, and aspect ratio, are considered for power reactor applications. Plasma gain Q = fusion power/input power in excess of 20 and average neutron wall fluxes from 2.3 to 3.6 MW/m{sup 2} are predicted in devices with major radii varying from 7.0 to 6.0 m and aspect ratios from 2.9 to 4.3.

OAK-B135 A stochastic magnetic boundary, produced by an externally applied edge resonant magnetic perturbation, is used to suppress large edge localized modes (ELMs) in high confinement (H-mode) plasmas. The resulting H-mode displays rapid, small oscillations with a bursty character modulated by a coherent 130 Hz envelope. The H-mode transport barrier is unaffected by the stochastic boundary. The core confinement of these discharges is unaffected, despite a three-fold drop in the toroidal rotation in the plasma core. These results demonstrate that stochastic boundaries are compatible with H-modes and may be attractive for ELM control in next-step burning fusion tokamaks.

High temporal and spatial resolution measurements in the boundary of the DIII-D tokamak show that edge-localized modes (ELMs) are produced in the low field side, are poloidally localized and are composed of fast bursts (∼20 to 40μs long) of hot, dense plasma on a background of less dense, colder plasma (∼5×1018m−3, 50 eV) possibly created by the bursts themselves. The ELMs travel radially in the scrape-off layer (SOL), starting at the separatrix at ∼450m∕s, and slow down to ∼150m∕s near the wall, 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 quickly decay with radius in the SOL. The temperature decay length (∼1.2 to 1.5 cm) is much shorter than the density decay length (∼3 to 8 cm), and the latter decreases with increasing pedestal (and SOL) density. The local particle and energy flux (assuming Ti=Te) at the midplane wall during the bursts are 10% to 50% (∼1 to 2×1021m−2s−1) and 1% to...

The physics and engineering guidelines for the ITER device are shown to lead to viable physics operating points for a steady state tokamak power reactor. Non-inductive current drive is provided in steady state by high energy neutral beam injection in the plasma core, lower hybrid slow waves in the outer regions of the plasma and bootstrap current. Plasma gain Q(/equivalent to/fusion power/input power) in excess of 20 and average neutron wall loading, approx. 2.0 MW/m/sup 2/ are predicted in a device with major radius, R/sub 0/ = 7.5 m and minor radius, a = 2.8 m. 15 refs., 3 figs., 3 tabs.

It is desirable for the plasma operating points of future Engineering Test Reactor (ETR) tokamaks to be in parameter regimes that are inherently stable to thermal fluctuations; in other words, thermal equilibrium is maintained by properties of the power balance terms themselves without an active burn control system. Methodologies are presented for calculating thermally stable operating points and scenarios to achieve these conditions. Results are given for an ETR tokamak with major radius R/sub 0/ = 5.8 m in both the ignition and current-drive modes. Burn dynamics simulations and discussion of critical issues are also presented. The analyses are general and should be applicable to a wide variety of deuterium-tritium burning tokamaks.

The extensions of the physics and engineering guidelines for the ITER device needed for acceptable operating points for a steady-state tokamak power reactor are examined. Non-inductive current drive is provided in steady state by high-energy neutral-beam injection in the plasma core, lower-hybrid slow waves in the outer regions of the plasma and (30%) bootstrap current. Three different levels of extension of the ITER physics/engineering guide-lines, with differing assumptions on the possible plasma beta, elongation and aspect ratio, are considered for power-reactor applications. Plasma gain, Q/sub p/ = fusion power/input power, in excess of 20 and average neutron wall fluxes from 2.3 to 3.6 MW/m/sup 2/ are predicted in devices with major radii varying from 7.0 to 6.0 m and aspect ratios from 2.9 to 4. 3. Peak divertor heat fluxes range up to 12.2 MW/m/sup 2/ which is somewhat higher than the current ITER design limit of 10 MW/m/sup 2/ with a magnetically swept divertor. These designs were selected on the basis of improvements in physics/engineering consistent with time scales for development of future reactors. The design re-optimization on the basis of cost-of-electricity (COE) was then examined using a reactor systems model. This analysis generally verified the original estimates formore » the required extensions of the ITER guidelines. Cost of electricity is projected to be less than 66 mills/kW/sub e/h in all of the configurations. The smallest reactor, which has the largest neutron wall flux and mass power density, yields the lowest COE, 56 mills/kW/sub e/h. 37 refs., 8 figs., 11 tabs.« less

Though helium plasmas are one option for the low activation phase of ITER, little effort has thus far been devoted to studying them in a large, diverted tokamak. A recent campaign on JET has therefore sought to address some of the important questions related to helium operation (He concentrations near 90%) in single null configuruations, particularly with regard to edge and divertor physics. This contribution compiles a selection of results from these experiments, in which, in each case, discharges have been chosen to match as closely as possible previous, well characterised D plasmas in both L and ELMing H-modes. These matched pulses are used to draw conclusions regarding the principle source and location of carbon production in D plasmas, to compare and contrast the mechanisms of the density limit and the detachment process in D and He, to investigate the nature of cross-field power transport in the SOL and to gain insight into the process by which ELM energy is transported to the divertor targets. (C) 2003 Elsevier Science B.V. All rights reserved.

This paper examines the radiation profiles and corresponding ionization source profiles of carbon and main plasma ions in matched helium and deuterium. L-mode plasmas in JET. The radiation intensities from C1+, C2+ and C3+ in the helium plasmas showed reduction by factors of 8, 10 and 25, respectively along the inner SOL and divertor leg compared with the deuterium cases. However, the emission in the outer divertor leg was less than a factor of 2 lower in helium. Using photon efficiencies calculated by the UEDGE code for the spectrometer lines of sight, the calculated source rates of C1+ and C3+ along the inner SOL and inner divertor were reduced by factors of 4 and 20, respectively in the helium plasmas. In the outer divertor leg the source rate of C3+ was reduced by a factor of 10 but the C1+ source rate did not change in helium. These measurements are consistent with a model that chemical sputtering of carbon dominates the source from the inner wall and inner divertor in deuterium L-mode plasmas while physical sputtering appears to dominate the source from the outer divertor. (C) 2003 Elsevier Science B.V. All rights reserved.

The electrical conductivity for a weakly ionized magnetized plasma in which the external magnetic field is subject to random fluctuations is calculated using a stochastic differential equation approach. Calculation of the dc conductivity for a magnetized plasma with random fluctuations in the magnitude of the external magnetic field is presented as an example. The effect of the random fluctuations in the magnetic field appears as a modification of the effective collision frequency of the plasma. The results for other cases involving ac electrical conductivity and diffusion coefficient showing similar modification of the effective collision frequency of the plasma are also presented.