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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.

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.

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.

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.

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.

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.

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.

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.