2014 Nuclear Fusion - announcing the shortlist

Local gyrokinetic calculations of the logarithmic gradients at mid-radius of both electron and boron densities in ASDEX Upgrade H-mode plasmas are presented and compared with the experimental observations. The experimental results show that both the electron and the boron density profiles increase their peaking in response to the addition of central electron cyclotron heating over a background of neutral beam injection (NBI) heating. The boron density profiles are always less peaked than the electron density profiles in the confinement region, and are flat or even slightly hollow in the presence of NBI heating only. The experimental behaviours are well reproduced by the theoretical predictions. The agreement allows the identification, through theoretical modelling, of the transport mechanisms responsible for the observed dependences. In particular, the observed increase in the logarithmic electron density gradient with increasing central electron heating is explained by a concurrent reduction of the outward pure convection and an increase in the inward thermodiffusion. In addition, it is found that the plasma toroidal rotation velocity and its radial gradient play a non-negligible role in the turbulent boron transport, and allow the prediction of a decrease in boron peaking with increasing rotation velocity, which is consistent with the experimental observations.

Turbulent transport of toroidal momentum is investigated in global linear gyrokinetic simulations. The poloidal tilt of the global mode structure arising from the radial variation of the equilibrium (profile shearing) is shown to induce non-diagonal non-pinch momentum transport (residual stress). Local simulations performed at finite radial wave vector show that the effect is mainly due to the antisymmetric radial component of the magnetic drift. The residual stress resulting from profile shearing enhances co-current rotation for ion temperature gradient turbulence and counter-current rotation for trapped electron mode turbulence.

Behaviour and characteristics of W plasma-facing components under impinging high heat fluxes are investigated in view of the material choices for the divertor in future devices such as ITER and DEMO. Experiments have been carried out in the plasma edge of the TEXTOR tokamak to study melt-layer motion, macroscopic tungsten erosion from the melt layer as well as the changes in material properties such as grain size and abundance of voids or bubbles. The parallel heat flux at the radial position of the plasma-facing components (PFCs) in the plasma ranges around q_|| ∼ 45 MW m⁻² allowing samples to be exposed at an impact angle of 35° to 20–30 MW m⁻². Melt-layer motion perpendicular to the magnetic field is observed following a Lorentz force originating from thermoelectric emission of the hot sample. Up to 3 g of molten W are redistributed forming mountain-like structures at the edge of the sample. The typical melt-layer thickness is 1–1.5 mm. Those hills are, due to the changes in the local geometry, particularly susceptible to even higher heat fluxes of up to the full q_||. Locally the temperature can reach up to 6000 K, high levels of evaporation are causing significant erosion in the form of continuous fine-spray (∼1 × 10²⁴ atoms m⁻² s⁻¹). Strong evaporation cooling is observed hindering the further heating of the samples. In addition, the formation of ligaments and splashes occurs several times during the melt phase ejecting droplets in the order of several 10 µm up to 100 µm probably caused by an instability evolving in the melt. In terms of material degradation several aspects are considered: formation of leading edges by redistributed melt, bubble formation and recrystallization. Bubbles are occurring in sizes between 1 and 200 µm while recrystallization increases the grain size up to 1.5 mm. The power-handling capabilities are thus severely degraded. Melting of tungsten (W) in future devices is highly unfavourable and needs to be avoided especially in light of uncontrolled transients and possible unshaped PFCs

In the fully tungsten clad ASDEX Upgrade, the sputtering rates of tungsten have been determined at all relevant plasma facing components using fast spectroscopic measurements with temporal resolution down to 0.5 ms. The sputtering strongly increases during an edge-localized mode (ELM) and the ELMs are often the dominant cause of tungsten sputtering. A modelling approach was employed to calculate the tungsten source at the limiters and the resulting tungsten density at the pedestal top inside the H-mode edge transport barrier (ETB). In the ETB, it is assumed that tungsten transport is collisional, i.e. behaves like other impurities. The collisional transport leads to strong inward drifts and steep density gradients in the ETB, which are flattened during an ELM causing an efflux of tungsten. The collisional transport in the ETB is also calculated for typical ITER conditions and the resulting tungsten density profiles as well as the transport of the helium ash through the ETB are evaluated.

A comprehensive suite of diagnostics has allowed detailed measurements of the Alfvén eigenmode (AE) spatial structure and subsequent fast-ion transport in the ASDEX Upgrade (AUG) tokamak [1]. Reversed shear Alfvén eigenmodes (RSAEs) and toroidal induced Alfvén eigenmodes (TAEs) have been driven unstable by fast ions from ICRH as well as NBI origin. In ICRF heated plasmas, diffusive and convective fast-ion losses induced by AEs have been characterized in fast-ion phase space. While single RSAEs and TAEs eject resonant fast ions in a convective process directly proportional to the fluctuation amplitude, δB/B, the overlapping of multiple RSAE and TAE spatial structures and wave–particle resonances leads to a large diffusive loss, scaling as (δB/B)². In beam heated discharges, coherent fast-ion losses have been observed primarily due to TAEs. Core localized, low amplitude NBI driven RSAEs have not been observed to cause significant coherent fast-ion losses. The temporal evolution of the confined fast-ion profile in the presence of RSAEs and TAEs has been monitored with high spatial and temporal resolution. A large drop in the central fast-ion density due to many RSAEs has been observed as q_min passes through an integer. The AE radial and poloidal structures have been obtained with unprecedented details using a fast SXR as well as 1D and 2D ECE radiometers. GOURDON and HAGIS simulations have been performed to identify the orbit topology of the escaping ions and study the transport mechanisms. Both passing and trapped ions are strongly redistributed by AEs.

The application of static, non-axisymmetric, nonresonant magnetic fields (NRMFs) to high beta DIII-D plasmas has allowed sustained operation with a quiescent H-mode (QH-mode) edge and both toroidal rotation and neutral beam injected torque near zero. Previous studies have shown that QH-mode operation can be accessed only if sufficient radial shear in the plasma flow is produced near the plasma edge. In past experiments, this flow shear was produced using neutral beam injection (NBI) to provide toroidal torque. In recent experiments, this torque was nearly completely replaced by the torque from applied NRMFs. The application of the NRMFs does not degrade the global energy confinement of the plasma. Conversely, the experiments show that the energy confinement quality increases with lower plasma rotation. Furthermore, the NRMF torque increases plasma resilience to locked modes at low rotation. These results open a path towards QH-mode utilization as an edge-localized mode (ELM)-stable H-mode in the self-heated burning plasma scenario, where toroidal momentum input from NBI may be small or absent.

Disruption mitigation is mandatory for ITER in order to reduce forces, to mitigate heat loads during the thermal quench and to avoid runaway electrons (REs). A fast disruption mitigation valve has been installed at JET to study mitigation by massive gas injection. Different gas species and amounts have been investigated with respect to timescales and mitigation efficiency. We discuss the mitigation of halo currents as well as sideways forces during vertical displacement events, the mitigation of heat loads by increased energy dissipation through radiation, the heat loads which could arise by asymmetric radiation and the suppression of REs.

Direction reversals of intrinsic toroidal rotation have been observed in Alcator C-Mod ohmic L-mode plasmas following modest electron density or toroidal magnetic field ramps. The reversal process occurs in the plasma interior, inside of the q = 3/2 surface. For low density plasmas, the rotation is in the co-current direction, and can reverse to the counter-current direction following an increase in the electron density above a certain threshold. Reversals from the co- to counter-current direction are correlated with a sharp decrease in density fluctuations with k_R ⩾ 2 cm⁻¹ and with frequencies above 70 kHz. The density at which the rotation reverses increases linearly with plasma current, and decreases with increasing magnetic field. There is a strong correlation between the reversal density and the density at which the global ohmic L-mode energy confinement changes from the linear to the saturated regime.

We develop and test a model, EPED1.6, for the H-mode pedestal height and width based upon two fundamental and calculable constraints: (1) onset of non-local peeling–ballooning modes at low to intermediate mode number, (2) onset of nearly local kinetic ballooning modes at high mode number. Calculation of these two constraints allows a unique, predictive determination of both pedestal height and width. The present version of the model is first principles, in that no parameters are fit to observations, and includes important non-ideal effects. Extensive successful comparisons with existing experiments on multiple tokamaks, including experiments where predictions were made prior to the experiment, are presented, and predictions for ITER are discussed.

Steady-state handling of divertor heat flux is a critical issue for ITER and future conventional and spherical tokamaks with compact high-power density divertors. A novel 'snowflake' divertor (SFD) configuration was theoretically predicted to have significant magnetic geometry benefits for divertor heat flux mitigation, such as an increased plasma-wetted area and a higher divertor volume available for volumetric power and momentum loss processes, as compared with the standard divertor. Both a significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with core H-mode confinement in discharges with the SFD using only a minimal set of poloidal field coils.

The first high-confinement mode (H-mode) with type-III edge localized modes at an H factor of H_IPB98(y,2) ∼ 1 has been obtained with about 1 MW lower hybrid wave power on the EAST superconducting tokamak. The first H-mode plasma appeared after wall conditioning by lithium (Li) evaporation before plasma breakdown and the real-time injection of fine Li powder into the plasma edge. The threshold power for H-mode access follows the international tokamak scaling even in the low density range and a threshold in density has been identified. With increasing accumulation of deposited Li the H-mode duration was gradually extended up to 3.6 s corresponding to ∼30 confinement times, limited only by currently attainable durations of the plasma current flat top. Finally, it was observed that neutral density near the lower X-point was progressively reduced by a factor of 4 with increasing Li accumulation, which is considered the main mechanism for the H-mode power threshold reduction by the Li wall coatings.