The effect of CD4 injection on WD molecule sputtering at the divertor target in EAST

In this work, the turbulent transport in the edge plasma and Scrape-Off Layer (SOL) region of IR-T1 tokamak at the presence of biased limiter has been investigated and analyzed. The time and radial evolution of floating potential, electric field and turbulent transport have been measured by using two arrays of the Langmuir probes in both the radial and poloidal directions. The analyses have been done by the Fast Fourier Transport method and spectral features of them are obtained with the help of the standard Auto-Correlation technique and modified covariance power spectral density estimate. The probability distribution function and actual transfer function magnitude of the radial and poloidal turbulent transport (Γr and Γp) have been investigated and compared in the edge plasma and SOL region. Also the histogram of turbulent transport has been analyzed and compared in the edge and SOL at presence of positive limiter biasing. The results show that in the edge plasma poloidal turbulent transport (Γp) is about of 60 % more than SOL region whereas radial turbulent transport (Γr) is about of 40 % less. During the application of positive biasing, it was found that Γr in the IR-T1 reduces by about 80 % in the edge plasma and 45–50 % in the SOL. Increase of Γp is about of 50 % after applied positive biasing in the edge while it increases 70 % nearly, in the SOL. Consequently, the improvement in confinement can be obtained for positive limiter biasing.

Plasma flows and their profiles in edge and scrape-off layer (SOL) regions of a tokamak plasma are important as these can modify the interchange plasma turbulence. These flows have been investigated in the presence of neutral gas numerically using the two-dimensional model equations. A reduction of poloidal flows, radial particle, and energy fluxes has been observed in the edge and SOL regions by the presence of the neutral gas. The reduction of radial flux is due to the reduction in the radial velocity of the plasma blob. We have investigated Reynolds stress in the presence of the gas. It is found that in the presence of ion-neutral collisions, the magnitude of the stress decreases. The gas also reduces the diamagnetic drift frequency in the edge and SOL regions.

Experiments in EAST have concentrated on studying the internal transport barrier (ITB) regime in high normalized beta () discharges, where the study of the compatibility between ITB dynamics and divertor plasmas is an important step for future steady-state and high-performance plasma operations. In this work, the characteristics of the divertor particle flux and their responses to ITB dynamics have been studied in high- discharges. In order to describe the characteristics, the ITB duration is divided into two phases: phase I (i.e. ITB formation) and phase II (i.e. ITB degradation), according to the variation of plasma stored energy. In phase I, the particle flux near the inner strike point (SP) is continuously enhanced during the inter-edge-localized mode (ELM) phase in both the lower single-null and upper single-null configurations. The total particle flux in the scrape-off layer (SOL) region reveals a similar trend with an increase of the flux near the SP. However, in the private flux region (PFR) the total particle flux shows a reduction. Meanwhile, a movement of the SP away from the divertor corner is also observed during the ITB formation. In phase II, the particle flux near the inner SP, the total particle flux in the inner SOL and PFR region recover to their initial level before ITB formation, respectively. Additionally, the particle decay length is obviously reduced in phase I and then gradually enhanced in phase II. The continuous variation of the particle flux at the inner divertor target is in accordance with a compression of the magnetic flux surfaces due to ITB formation, which increases the gradient of electron density in the edge region. It is indicative that the ITB has a feasible impact on the behavior of divertor plasmas by means of increasing the Shafranov shift during the inter-ELM phase.

SOLPS modelling of ELMing H-mode

The paper investigates the effect of pedestal turbulence, i.e. coherent mode (CM), on the divertor total particle flux, the particle flux distribution and heat flux distribution, and their decay lengths in the scrape-off layer (SOL) region on EAST tokamak. The experimental evidence indicates the coherence between the pedestal CM and the divertor particle flux fluctuation, which implies that the pedestal CM can affect the divertor plasma behavior. The increase of the particle flux decay length correlates with the density radial decay length (λ n,SOL) in the SOL region where the λ n,SOL is largely affected by the change of pedestal CM amplitude. A quantitative statistical analysis shows that the divertor total particle flux and in the SOL region increase with increasing amplitude of the electron density fluctuation and the electron temperature fluctuation induced by the pedestal CM. The temporal decay process of the total particle flux is interrupted by the appearance of CM at the pedestal. Meanwhile, the spatial distributions of the divertor particle flux and heat flux are enhanced when the pedestal CM appears, which are observed in both the upper single null and lower single null configurations.

A plasma materials interaction code, EDDY, is modified to calculate the erosion and deposition patterns on divertor plates of fusion devices. An area in the plate is divided into segments, where dynamic erosion and deposition processes on the plate are simulated, and the ionisation and dissociation processes of released impurities in a divertor plasma are followed. Using the modified code, the erosion and deposition patterns on inner and outer vertical targets (graphite) in a designed fusion device, ITER, in a semi-detached plasma condition and on a divertor target plate in JT-60U are calculated. An erosion zone and a deposition zone are obtained in a private flux region and a scrape-off layer (SOL) region, respectively, on the outer vertical target. Near the strike point in the SOL region on the inner vertical target where the plasma temperature is low (<5 eV), the erosion dominates due to low ionisation and dissociation rates of chemically sputtered hydrocarbons if the same chemical sputtering yield (0.04) as for the outer vertical target is assumed. Physical sputtering is much less important for the patterns, in particular, near the strike points and in the private flux region. The two-dimensional plots of redeposited hydrocarbons and carbons for both targets reveal the dependence of the local redeposition process on the temperature and density of the plasma and the energy of the released particles.

Modification of a Monte Carlo simulation code, Erosion and Deposition based on DYnamic model (EDDY), for plasma-surface interactions in a designed tokamak, International Thermonuclear Experimental Reactor-Fusion Energy Advanced Tokamak (ITER-FEAT), and its application for erosion and redeposition of a carbon target in the divertor are presented. The modified EDDY code allows us to treat the deposition of plasma impurities and the prompt redeposition of sputtered atoms and molecules on the target surface. At elevated temperatures, furthermore, the impurity diffusion inside the target and chemical sputtering of carbon are taken into account. In the ITER-FEAT, physical sputtering of the divertor target is very small in the scrape-off layer (SOL) region, and chemical sputtering dominates the erosion near the strike point and in the private flux region. Prompt redeposition strongly suppresses the sputtering of the target and plasma carbon impurity deposits on it. As a result, no erosion is calculated in the SOL region and a thick deposition layer is produced near the strike point. A narrow erosion zone remains only in the private flux region. Furthermore, radial distributions of each particle species released in the plasma and their redeposition profiles on the surface are discussed.

The effects of nitrogen gas seeding in the edge and scrape-off layer (SOL) regions of a tokamak plasma are studied through 2D fluid simulations using the BOUT++ code. Proper account is taken of the presence of multiple charged states of nitrogen ions due to ionization, recombination, and dissociation processes, and a self-consistent study of the interaction of these ions with the turbulent plasma in the edge and SOL regions is carried out. The self-consistent model includes the effects of polarization drifts of the main plasma and impurity ions for determining the plasma vorticity. Nitrogen seeding is found to modify the turbulence as well as to influence the profiles of the equilibrium plasma density and the electron temperature. The densities of N3+ to N5+ ions are found to be relatively higher than the other charged states. This is understood and further validated by a 0D simulation. The radial profiles of these impurity ions are mapped, and their radiation energy losses are estimated. The radial profile of the radiation losses is maximum near to the edge-to-SOL transition region and becomes broader in the edge region than the SOL region.

Understanding helium (He) plasma-induced tungsten (W) surface modifications, and the effect of irradiation defects on He plasma-induced W surface modifications under a real divertor environment are important for the operation of fusion reactors. In this study, two iron (Fe) ions pre-irradiated W samples with dislocation loops and voids, and two unirradiated W samples were exposed to He plasma at a divertor leg position of the Large Helical Device (LHD). The gross erosion rate are 1.0×1020 atoms m-2 s-1, 1.0×1020 atoms m-2 s-1, 9.3×1019 atoms m-2 s-1, 7.4×1019 atoms m-2 s-1 for W9, W10, W11, W12, respectively. The surface of each sample after the exposure was different at the strike point and the two regions on either side of the strike point, the scrape-off layer (SOL) region and the private region. The typical He plasma-induced structures at the SOL region were stripe structure, sawtooth structure, and no undulating structure, which are collectively called He-structures in the present study. At the strike point, the typical He plasma-induced structure was dense W protrusions. At the private flux region, the typical He plasma-induced structures are semi-formed He-structures. The formations of these structures were depended on the grain orientation. And pinholes were observed on these structures. No significant difference in the He plasma-induced structures was found between the pre-irradiated W and the unirradiated W. The formations of no undulating structure and stripe structure were discussed based on the observed semi-formed He-structures.

Several experiments on different machines and in different fast wave (FW) heating regimes, such as hydrogen minority heating and high harmonic fast waves, have found strong interactions between radio-frequency (RF) waves and the scrape-off layer (SOL) region. This paper examines the propagation and the power loss in the SOL by using the full wave code AORSA, in which the edge plasma beyond the last closed flux surface (LCFS) is included in the solution domain and a collisional damping parameter is used as a proxy to represent the real, and most likely nonlinear, damping processes. 3D AORSA results for the National Spherical Torus eXperiment (NSTX), where a full antenna spectrum is reconstructed, are shown, confirming the same behavior found for a single toroidal mode results in Bertelli et al, Nucl. Fusion, 54 083004, 2014, namely, a strong transition to higher SOL power losses (driven by the RF field) when the FW cut-off is moved away from in front of the antenna by increasing the edge density. Additionally, full wave simulations have been extended to “conventional” tokamaks with higher aspect ratios, such as the DIII-D, Alcator C-Mod, and EAST devices. DIII-D results show similar behavior found in NSTX and NSTX-U, consistent with previous DIII-D experimental observations. In contrast, a different behavior has been found for Alcator C-Mod and EAST, which operate in the minority heating regime unlike NSTX/NSTX-U and DIII-D, which operate in the mid/high harmonic regime. A substantial discussion of some of the main aspects, such as (i) the pitch angle of the magnetic field; (ii) minority heating vs. mid/high harmonic regimes is presented showing the different behavior of the RF field in the SOL region for NSTX-U scenarios with different plasma current. Finally, the preliminary results of the impact of the SOL region on the evaluation of the helicon current drive efficiency in DIII-D is presented for the first time and briefly compared with the different regimes mentioned above.

Based on the variations of the static pressure along the magnetic field line in different divertor operation regimes, the effects of the divertor operation regimes on the plasma parallel flow at the edge of a tokamak are investigated using a one-dimensional fluid model. In low recycling regime, the variation of the static pressure along the field line is obvious from the scrape-off layer (SOL) region near the X-point, and the variation tendency is the same as that of the density. The Mach number of the plasma parallel flow increases along the magnetic field line and the variation is from gentle to sharp. In high recycling regime, the static pressure does not change much except in the near divertor plate region, as a result, the Mach number of the plasma parallel flow varies gently in the SOL region and the most of the divertor region, and it increases rapidly in the near divertor plate region. The variation of the static pressure in weak divertor detachment regime is similar to that in high recycling regime, but the static pressure shows decrease tendency in the SOL region near the X-point, consequently, the Mach number of the plasma parallel flow at X-point is larger than that in high recycling regime. In strong divertor detachment regime, static pressure decreases obviously in the SOL region and away from the divertor plate region, where the static pressure decreases rapidly, and a high Mach plasma parallel flow is observed. Static pressure decreasing while dynamic pressure increasing to keep the total pressure conservation is shown to be a possible cause of the high Mach parallel flow.

The effect of divertor magnetic balance on H-mode performance in DIII-D

Particle transport towards the scrape-off layer (SOL) region and the low-frequency instabilities occurring in the edge plasma have been studied in a small research Tokamak. The results obtained show that there exists enhanced particle transport in the plasma. The drift-wave-like instabilities observed in that region are related to this enhanced particle transport to the edge region.

The plasma parameters in the scrape-off layer (SOL) region and the particle flux at the divertor target are measured by using a fast reciprocating Langmuir probe assembly (FRLPA) and a fixed edge Langmuir probe array (ELPA) in the KSTAR. The e-folding lengths of the plasma density (n e ) and the electron temperature (T e ), λ ne and λ Te , are obtained from the radial profile measurement using the FRLPA at the outboard mid-plane during ohmic and H-mode discharges. Particle fluxes measured by using the ELPA at the divertor target are mapped to the outboard mid-plane, and the fluxes are quite well matched to the radial profile of the particle flux from the FRLPA measurement. Finally, the peak heat flux at the divertor target during edge localized modes (ELMs) can be estimated to be up to 1.0 MWm−2 for the neutral beam power of ~ 2.7 MW by using T e at the separatrix and λ Te from the FRLPA measurements. In this work, the experimental results from the probe measurements in the KSTAR are presented.

Two dimensional interchange turbulence has been used to describe neon gas interaction with tokamak plasma, and is able to include anomalous transport effects self-consistently in the edge and scrape-off layer (SOL) regions. Model equations related to the plasma turbulence coupled with neon gas have been solved numerically using the BOUT++ framework code. Two different fluid models of the neon gas have been investigated, and in this context a few results related to the Aditya tokamak are presented. Numerical results indicate that ionization and radiative cooling processes modify the plasma turbulence. The existence of neon ions and their radial profile in the edge and SOL regions are the most significant results in this work. This new result has been investigated in detail in the presence of the interchange plasma turbulence. Other relevant results such as an increase of plasma density and a small change of electron temperature are also presented. In turbulence saturated states the electron temperature can increase slightly even in the presence of radiative cooling. Neon ions in the presence of polarization drift and radiative cooling modify the radial electric field and its radial shear. Radial electron energy flux has been investigated in the context of poloidal velocity shear and turbulence decorrelation rates. An increase of the electron energy confinement time has been observed numerically, mainly in the SOL region in the presence of gas seeding. Experimental results show an increase of global energy confinement time only when the gas seeding becomes effective in changing the global plasma parameters.