V. Parail

RMP ELM suppression in DIII-D plasmas with ITER similar shapes and collisionalities

Large Type-I edge localized modes (ELMs) are completely eliminated with small n = 3 resonant magnetic perturbations (RMP) in low average triangularity, , plasmas and in ITER similar shaped (ISS) plasmas, , with ITER relevant collisionalities . Significant differences in the RMP requirements and in the properties of the ELM suppressed plasmas are found when comparing the two triangularities. In ISS plasmas, the current required to suppress ELMs is approximately 25% higher than in low average triangularity plasmas. It is also found that the width of the resonant q95 window required for ELM suppression is smaller in ISS plasmas than in low average triangularity plasmas. An analysis of the positions and widths of resonant magnetic islands across the pedestal region, in the absence of resonant field screening or a self-consistent plasma response, indicates that differences in the shape of the q profile may explain the need for higher RMP coil currents during ELM suppression in ISS plasmas. Changes in the pedestal profiles are compared for each plasma shape as well as with changes in the injected neutral beam power and the RMP amplitude. Implications of these results are discussed in terms of requirements for optimal ELM control coil designs and for establishing the physics basis needed in order to scale this approach to future burning plasma devices such as ITER.

Active control of type-I edge localized modes with n = 1 and n = 2 fields on JET

Recent experiments on JET have shown that type-I edge localized modes (ELMs) can be controlled by the application of static low n = 1 external magnetic perturbation fields produced by four external error field correction coils (EFCC) mounted far away from the plasma between the transformer limbs. When an n = 1 field with an amplitude of a few mT at the plasma edge (the normalized poloidal flux, ?, is larger than 0.95) is applied during the stationary phase of a type-I ELMy H-mode plasma, the ELM frequency rises from ~30?Hz up to ~120?Hz. The energy loss per ELM normalized to the total stored energy, ?WELM/W, decreased from 7% to values below the resolution limit of the diamagnetic measurement (<2%). Transport analysis using the TRANSP code shows up to 20% reduction in the thermal energy confinement time because of density pump-out, but when normalized to the IPB98(y, 2) scaling the confinement time shows almost no reduction. Stability analysis of controlled ELMs suggests that the operational point with n = 1 perturbation field moves from the intermediate-n peeling?ballooning boundary to the low-n peeling boundary, and the radial width of the most unstable mode reduced from ~3% down to ~1% of the normalized minor radius. The first results of ELM control with n = 2 fields on JET demonstrate that the frequency of ELMs can be increased by a factor of 3.5 with the present capability of the EFCC power supply. During the application of the n = 1, 2 fields, a reduction in the absolute ELM size (?WELM) and ELM peak heat fluxes on the divertor target by roughly the same factor as the increase in the ELM frequency has been observed. The reduction in heat flux is mainly due to the drop in particle flux rather than a change in the electron temperature. Similar plasma braking effects have been observed with n = 1 and n = 2 external fields when the same EFCC coil current was applied. Compensation of the density pump-out effect has been achieved by means of gas fuelling in low triangularity plasmas. An optimized fuelling rate to compensate the density pump-out effect has been identified. When the n = 1 field is applied in plasmas with reduced toroidal rotation and density due to increased TF ripple of 8%, both the magnitude of the toroidal braking and density pump-out are found to become smaller; however, the increase in the ELM frequency with the n = 1 field is still observed. Active ELM control by externally applied fields may offer an attractive method for next-generation tokamaks, e.g. ITER.

Physics of transport in tokamaks

This paper is an overview of recent results relating to turbulent particle and heat transport, and to the triggering of internal transport barriers (ITBs). The dependence of the turbulent particle pinch velocity on plasma parameters has been clarified and compared with experiment. Magnetic shear and collisionality are found to play a central role. Analysis of heat transport has made progress along two directions: dimensionless scaling laws, which are found to agree with the prediction for electrostatic turbulence, and analysis of modulation experiments, which provide a stringent test of transport models. Finally the formation of ITBs has been addressed by analysing electron transport barriers. It is confirmed that negative magnetic shear, combined with the Shafranov shift, is a robust stabilizing mechanism. However, some well established features of internal barriers are not explained by theory.

Characterization of pedestal parameters and edge localized mode energy losses in the Joint European Torus and predictions for the International Thermonuclear Experimental Reactor

This paper presents the experimental characterization of pedestal parameters, edge localized mode (ELM) energy, and particle losses from the main plasma and the corresponding ELM energy fluxes on plasma facing components for a series of dedicated experiments in the Joint European Torus (JET). From these experiments, it is demonstrated that the simple hypothesis relating the peeling-ballooning linear instability to ELM energy losses is not valid. Contrary to previous observations at lower triangularities, small energy losses at low collisionality have been obtained in regimes at high plasma triangularity and q95∼4.5, indicating that the edge plasma magnetohydrodynamic stability is linked with the transport mechanisms that lead to the loss of energy by conduction during type I ELMs. Measurements of the ELM energy fluxes on the divertor target show that their time scale is linked to the ion transport along the field and the formation of a high energy sheath, in agreement with kinetic modeling of ELMs. Higher...

Improved performance of ELMy H-modes at high density by plasma shaping in JET

We present the results of experiments in JET to study the effect of plasma shape on high density ELMy H-modes, with geometry of the magnetic boundary similar to that envisaged for the standard Q = 10 operation in ITER. The experiments described are single lower null plasmas, with standard q profile, neutral beam heating and gas fuelling, with average plasma triangularity ? calculated at the separatrix ~0.45-0.5 and elongation ?~1.75. In agreement with the previous results obtained in JET and other divertor Tokamaks, the thermal energy confinement time and the maximum density achievable in steady state for a given confinement enhancement factor increase with ?. The new experiments have confirmed and extended the earlier results, achieving a maximum line average density ne~1.1nGR for H98~0.96. In this plasma configuration, at 2.5?MA/2.7?T (q95~2.8), a line average density ~95%?nGR with H98 = 1 and ?N~2 are obtained, with plasma thermal stored energy content Wth being approximately constant with increasing density, as long as the discharge maintains Type I ELMs, up to nped~nGR (and ne~1.1nGR). A change in the Type I ELMs behaviour is observed for pedestal densities nped70%?nGR, with their frequency decreasing with density (at constant Psep), enhanced divertor D? emission and increased inter-ELM losses. We show that this change in the ELM character at high pedestal density is due to a change in transport and/or stability in the pedestal region, with the ELMs changing from Type I to mixed Type I and Type II. The similarity of these observations with those in the Type II ELM regime in ASDEX?Upgrade and with other small ELM regimes in DIII-D, JT-60U and Alcator C-MOD is discussed. Finally, we present the first results of experiments by studying in more detail the effects of the plasma boundary geometry, in particular by investigating separately the effect of the upper and lower triangularity, at high average ?. We show that the changes to the lower ? (or of the radial position of the x-point) affect the pedestal parameters, the size of ELM energy losses as well as the global energy confinement of the plasma.

Development of ITER 15 MA ELMy H-mode inductive scenario

The poloidal field (PF) coil system on ITER, which provides both feedforward and feedback control of plasma position, shape, and current, is a critical element for achieving mission performance. Analysis of PF capabilities has focused on the 15 MA Q = 10 scenario with a 300-500 s flattop burn phase. The operating space available for the 15 MA ELMy H-mode plasma discharges in ITER and upgrades to the PF coils or associated systems to establish confidence that ITER mission objectives can be reached have been identified. Time dependent self-consistent free-boundary calculations were performed to examine the impact of plasma variability, discharge programming, and plasma disturbances. Based on these calculations a new reference scenario was developed based upon a large bore initial plasma, early divertor transition, low level heating in L-mode, and a late H-mode onset. Equilibrium analyses for this scenario indicate that the original PF coil limitations do not allow low li (<0.8) operation or lower flux states, and the flattop burn durations were predicted to be less than the desired 400 s. This finding motivates the expansion of the operating space, considering several upgrade options to the PF coils. Analysis was also carried out to examine the feedback current reserve required in the CS and PF coils during a series of disturbances and a feasibility assessment of the 17 MA scenario was undertaken. Results of the studies show that the new scenario and modified PF system will allow a wide range of 15 MA 300-500 s operation and more limited but finite 17 MA operation.

Reduction of divertor heat load in JET ELMy H-modes using impurity seeding techniques

The main objective of this paper is investigation of methods for reduction of divertor heat loads in order to increase the lifetime of divertor tiles in future fusion reactors. Special emphasis is given to studies of reduction of transient heat loads due to edge localized modes (ELMs). Two methods are compared: argon seeded type-I ELMy H-modes and nitrogen seeded type-III ELMy H-modes. In both scenarios, the impurity seeding leads to a reduction in the pedestal energy and hence a reduction in the energy released by the ELM. This consequentially reduces the power load to the divertor targets. At high radiative power fractions in type-III ELMy H-modes, part of that released ELM energy (small ELMs, below 20 kJ) is dissipated by radiation in the scrape off layer (SOL). Modelling of the ELM mitigation supports the experimental findings. This ELM mitigation by radiative dissipation is not effective for larger ELMs. In between ELMs, the plasma is detached and radiates strongly from the X-point region. During an ELM, the nitrogen in the X-point and divertor region becomes ionized into more weakly radiating higher charge states and the plasma re-attaches for large ELMs. At JET, argon radiates predominantly in the main plasma and not so much in the cold divertor region. Hence, the effect of radiative dissipation of ELM heat fluxes by argon is very low due to the limited argon density in the divertor region. Nevertheless, both scenarios might be compatible with an integrated ITER scenario, with respect to acceptable divertor lifetime and acceptable confinement.

Progress towards steady-state operation and real-time control of internal transport barriers in JET

In JET, advanced tokamak research mainly focuses on plasmas with internal transport barriers (ITBs) that are strongly influenced by the current density profile. A previously developed optimized shear regime with low magnetic shear in the plasma centre has been extended to deeply negative magnetic shear configurations. High fusion performance with wide ITBs has been obtained transiently with negative central magnetic shear configuration: HIPB98(y,2) ~ 1.9, βN = 2.4 at Ip = 2.5 MA. At somewhat reduced performance, electron and ion ITBs have been sustained in full current drive operation with 1 MA of bootstrap current: HIPB98(y,2) ~ 1, βN = 1.7 at Ip = 2.0 MA. The ITBs were maintained for up to 11 s for the latter case. This duration, much larger than the energy confinement time (37 times larger), is already approaching a current resistive time. New real-time measurements and feedback control algorithms have been developed and implemented in JET for successfully controlling the ITB dynamics and the current density profile in the highly non-inductive current regime.

Towards fully non-inductive current drive operation in JET

Quasi-steady operation has been achieved at JET in the high-confinement regime with internal transport barriers (ITBs). The ITB has been maintained up to 11 s. This duration, much larger than the energy confinement time, is already approaching a current resistive time. The high-performance phase is limited only by plant constraints. The radial profiles of the thermal electron and ion pressures have steep gradients typically at mid-plasma radius. A large fraction of non-inductive current (above 80%) is sustained throughout the high-performance phase with a poloidal beta exceeding unity. The safety factor profile plays an important role in sustaining the ITB characteristics. In this regime where the self-generated bootstrap current (up to 1.0 MA) represents 50% of the total current, the resistive evolution of the non-monotonic q-profile is slowed down by using off-axis lower-hybrid current drive.

On the heating mix of ITER

This paper considers the heating mix of ITER for the two main scenarios. Presently, 73 MW of absorbed power are foreseen in the mix 20/33/20 for ECH, NBI and ICH. Given a sufficient edge stability, Q = 10-the goal of scenario 2-can be reached with 40MW power irrespective of the heating method but depends sensitively inter alia on the H-mode pedestal temperature, the density profile shape and on the characteristics of impurity transport. ICH preferentially heats the ions and would contribute specifically with Delta Q 0.5, and strong off-axis current drive (CD). The findings presented here are based on revised CD efficiencies gamma for ECCD and a detailed benchmark of several CD codes. With ECCD alone, the goals of scenario 4 can hardly be reached. Efficient off-axis CD is only possible with NBI. With beams, inductive discharges with f(ni) > 0.8 can be maintained for 3000 s. The conclusion of this study is that the present heating mix of ITER is appropriate. It provides the necessary actuators to induce in a flexible way the best possible scenarios. The development risks of NBI at 1 MeV can be reduced by operation at 0.85 MeV.