H. Reimerdes

Neoclassical tearing modes

Neoclassical tearing modes are one of the most serious concerns for operation on a next-step tokamak device. The modes occur on present tokamaks at normalized pressure (beta (N)) values comparable to those envisaged For baseline scenarios in future devices, such as ITER-FEAT. Further, empirical scalings based on data from many of the present machines point to much lower thresholds on a larger device. However, physics-based models indicate an important role for the seed island mechanisms, which may in fact give rise to increased stability on larger devices-i.e. if the seed island width (required to trigger the NTM) falls below the critical levels required. Fits based on these models suggest this is the case, bur are too badly constrained at present to make reliable predictions, and the physics is complex, making quantitative theoretical calculation difficult. Further experiments are required to examine the scaling of the seed, as well as to identify the role and relative sizes of the stabilizing terms that set the critical size for mode growth. In the event that the modes are unavoidable, promising feedback stabilization techniques are being developed with the use of localized RF current drive to change the stability properties of the plasma. Further work is needed to demonstrate sustained access to higher beta (N) and provide data to refine models. This paper reviews the underlying physics and key issues, commenting on the present status of understanding and further work required.

Fast seeding of NTMs by sawtooth crashes in TCV and their preemption using ECRH

Many tokamaks have observed that sawteeth of sufficient duration may trigger neoclassical tearing modes (NTMs) that lead to plasma performance degradation. In this paper, TCVs ability to accurately control the period of individual sawteeth is exploited, using localized electron cyclotron resonance heating and current drive (ECRH and ECCD), to trigger NTMs under controlled conditions, providing an excellent environment for the study of the seeding of NTMs by sawtooth crashes. The TCV experiments show evidence of a fast formation of seed islands with poloidal/toroidal mode numbers m/n = 3/2 and 2/1 within a few tens of microseconds following the sawtooth crash. Crashes of sawteeth with a longer period duration are observed to generate larger seed islands but also increase the plasma stability to conventional tearing modes. While these two effects compete, the NTM stability is found to decrease with increasing sawtooth period. The seed island size can be reduced and, thereby, the NTM stability improved, by increasing the value of the safety factor q95. Alternatively, NTM stability can be increased by application of preemptive ECRH at the resonant surface of the NTM. Preemptive ECRH is found to enlarge the plasma operational domain by improving the conventional tearing stability and by reducing the coupling between the driving (m/n = 1/1 or 2/2) and the driven modes (m/n = 2/1 or 3/2), resulting in smaller sawtooth generated seed islands. The efficiency of preemptive ECRH increases when sufficient ECRH power is applied in a short time interval prior to the sawtooth crash.

Magnetohydrodynamic helical structures in nominally axisymmetric low-shear tokamak plasmas

The primary goal of hybrid scenarios in tokamaks is to enable high performance operation with large plasma currents whilst avoiding MHD instabilities. However, if a local minimum in the safety factor is allowed to approach unity, the energy required to overcome stabilizing magnetic field line bending is very small, and as a consequence, large MHD structures can be created, with typically dominant m = n = 1 helical component. If there is no exact q = 1 rational surface the essential character of these modes can be modelled assuming ideal nested magnetic flux surfaces. The methods used to characterize these structures include linear and non-linear ideal MHD stability calculations which evaluate the departure from an axisymmetric plasma state, and also equilibrium calculations using a 3D equilibrium code. While these approaches agree favourably for simulations of ITER relevant hybrid regimes in this paper, the relevance of the ideal MHD model itself is tested through empirical examination of helical states in MAST and TCV. While long lived modes in MAST do not have island structures, some of the continuous mode oscillations exhibited in high elongation experiments in TCV indicate that resistivity may play a role in further weakening the ability of the tokamak core to remain axisymmetric. The simulations and experiments consistently highlight the need to control the safety factor in hybrid scenarios planned for future fusion grade tokamaks such as ITER.

X-point position dependence of edge intrinsic toroidal rotation on the Tokamak à Configuration Variablea)

Recent theoretical work predicts intrinsic toroidal rotation in the tokamak edge to depend strongly on the normalized major radial position of the X-point. With this motivation, we conducted a series of Ohmic L-mode shots on the Tokamak a Configuration Variable, moving the X-point from the inboard to the outboard edge of the last closed flux surface in both lower and upper single null configurations. The edge toroidal rotation evolved from strongly co-current for an inboard X-point to either vanishing or counter-current for an outboard X-point, in agreement with the theoretical expectations. The whole rotation profile shifted roughly rigidly with the edge rotation, resulting in variation of the peak core rotation by more than a factor of two. Core rotation reversals had little effect on the edge rotation. Edge rotation was slightly more counter-current for unfavorable than favorable ∇B drift discharges.

High-beta, steady-state hybrid scenario on DIII-D

The potential of the hybrid scenario (first developed as an advanced inductive scenario for high fluence) as a regime for high-beta, steady-state plasmas is demonstrated on the DIII-D tokamak. These experiments show that the beneficial characteristics of hybrids, namely safety factor >= 1 with low central magnetic shear, high stability limits and excellent confinement, are maintained when strong central current drive (electron cyclotron and neutral beam) is applied to increase the calculated non-inductive fraction to approximate to 100% (approximate to 50% bootstrap current). The best discharges achieve normalized beta of 3.4, IPB98(y,2) confinement factor of 1.4, surface loop voltage of 0.01 V, and nearly equal electron and ion temperatures at low collisionality. A 0D physics model shows that steady-state hybrid operation with Q(fus) similar to 5 is feasible in FDF and ITER. The advantage of the hybrid scenario as an advanced tokamak regime is that the external current drive can be deposited near the plasma axis where the efficiency is high; additionally, good alignment between the current drive and plasma current profiles is not necessary as the poloidal magnetic flux pumping self-organizes the current density profile in hybrids with an m/n = 3/2 tearing mode.

Real-Time Multi-EC-Actuator MHD Control on TCV

Real-time control of multiple plasma actuators is a requirement in advanced tokamaks; for example for burn control, plasma current profile control and MHD stabilization - EC wave absorption is ideally suited especially for the latter. For example, on ITER 24 EC sources can be switched between 54 inputs at the torus. In the torus, 5 launchers direct the power to various locations across the plasma profile via 11 steerable mirrors. For optimal usage of the available power, the aiming and polarization of the beams must be adapted to the plasma configuration and the needs of the scenario. Since the EC system performs many competing tasks, present day systems should demonstrate the ability of an EC plant to deal with several targets in parallel and/or to switch smoothly between goals to attain overall satisfaction. Recently TCV has taken a first step towards such a demonstration. Several EC launchers are used simultaneously to regulate the sawtooth period and to preempt m/n = 3/2 NTMs, by controlling the power levels. In parallel, a second algorithm stabilizes any NTM that saturates [1]. These and real-time MHD control experiments on ELMs [2] are presented.

Contributions of Electron Cyclotron Waves to Performance in Advanced Regimes on DIII‐D

High‐power electron cyclotron (EC) waves are used to increase performance in several Advanced Tokamak (AT) regimes on DIII‐D where there is a simultaneous need for high noninductive current and high beta. In the Quiescent High‐confinement mode (QH‐mode), a direct measurement of the electron cyclotron current drive (ECCD) profile is made using modulation techniques, and a trapped electron mode (TEM) dominated regime with core Te>Ti is created. In the “highqmin” AT scenario, ECCD provides part of the off‐axis noninductive current and helps to produce a tearing stable equilibrium. In the hybrid regime, strong central current drive from EC waves and other sources increases the noninductive current fraction to ≈100%. Surprisingly, the core safety factor remains above unity, meaning good alignment between the current drive profile and the desired plasma current profile is not necessary in this scenario.

Influences of multiple low-n modes on n=1 resistive wall mode identification and feedback control

It is well known in theory that even after the n=1 resistive wall mode (RWM) is suppressed, the other low-n modes, such as n=2 or 3, can appear sequentially, as β increases. In recent DIII-D experiments [J. L. Luxon, Nucl. Fusion 42, 614 (2002)], we found such an example that supports the theoretical prediction: while the n=1 mode was suppressed, an n=3 mode grew dominant, leading to a β collapse. The n=1 RWM suppression was likely due to a combination of rotational stabilization and n=1 RWM feedback. The multiple RWM identification was performed using an expanded matched filter, where n=1 and n=3 RWM basis vectors are simultaneously considered. Taking advantage of the expanded matched filter, we found that an n=3 mode following an edge-localized-mode burst grew almost linearly for several milliseconds without being hindered. This n=3 mode appeared responsible for the β collapse (down to the n=3 no-wall limit), as well as for a drop in toroidal rotation. A preliminary analysis suggests that the identity o...

Enhanced (E)over-right-arrow x (B)over-right-arrow drift effects in the TCV snowflake divertor

Measurements of various plasma parameters at the divertor targets of snowflake (SF) and conventional single-null configurations indicate an enhanced effect of the drift in the scrape-off layer of plasmas in the SF configuration. Plasma boundary transport simulations using the EMC3-Eirene code show that the poloidal gradients of the kinetic profiles in the vicinity of the null-point of a SF divertor are substantially larger than those of a conventional single-null configuration. These gradients are expected to drive larger flows in the SF divertor and are thought to be responsible for the formation of the double-peaked particle and heat flux target profiles observed experimentally. Experiments in forward and reversed toroidal magnetic field directions further support this conclusion. The formation of such a double-peaked profiles is enhanced at higher plasma densities and may have beneficial effects on the divertor heat loads since they lead to broader target profiles and lower peak heat fluxes.Measurements of various plasma parameters at the divertor targets of snowflake (SF) and conventional single-null configurations indicate an enhanced effect of the (E) over right arrow x (B) over right arrow drift in the scrape-off layer of plasmas in the SF configuration. Plasma boundary transport simulations using the EMC3-Eirene code show that the poloidal gradients of the kinetic profiles in the vicinity of the null-point of a SF divertor are substantially larger than those of a conventional single-null configuration. These gradients are expected to drive larger (E) over right arrow x (B) over right arrow flows in the SF divertor and are thought to be responsible for the formation of the double-peaked particle and heat flux target profiles observed experimentally. Experiments in forward and reversed toroidal magnetic field directions further support this conclusion. The formation of such a double-peaked profiles is enhanced at higher plasma densities and may have beneficial effects on the divertor heat loads since they lead to broader target profiles and lower peak heat fluxes.

Enhanced

Measurements of various plasma parameters at the divertor targets of snowflake (SF) and conventional single-null configurations indicate an enhanced effect of the drift in the scrape-off layer of plasmas in the SF configuration. Plasma boundary transport simulations using the EMC3-Eirene code show that the poloidal gradients of the kinetic profiles in the vicinity of the null-point of a SF divertor are substantially larger than those of a conventional single-null configuration. These gradients are expected to drive larger flows in the SF divertor and are thought to be responsible for the formation of the double-peaked particle and heat flux target profiles observed experimentally. Experiments in forward and reversed toroidal magnetic field directions further support this conclusion. The formation of such a double-peaked profiles is enhanced at higher plasma densities and may have beneficial effects on the divertor heat loads since they lead to broader target profiles and lower peak heat fluxes.Measurements of various plasma parameters at the divertor targets of snowflake (SF) and conventional single-null configurations indicate an enhanced effect of the (E) over right arrow x (B) over right arrow drift in the scrape-off layer of plasmas in the SF configuration. Plasma boundary transport simulations using the EMC3-Eirene code show that the poloidal gradients of the kinetic profiles in the vicinity of the null-point of a SF divertor are substantially larger than those of a conventional single-null configuration. These gradients are expected to drive larger (E) over right arrow x (B) over right arrow flows in the SF divertor and are thought to be responsible for the formation of the double-peaked particle and heat flux target profiles observed experimentally. Experiments in forward and reversed toroidal magnetic field directions further support this conclusion. The formation of such a double-peaked profiles is enhanced at higher plasma densities and may have beneficial effects on the divertor heat loads since they lead to broader target profiles and lower peak heat fluxes.