Outer Divertor Region Research Articles

Modelling study of divertor W leakage for different divertor conditions with Ne seeding in EAST tokamak

Abstract Sequential SOLPS-ITER and DIVIMP simulations are carried out for tungsten (W) edge transport during neon (Ne) injection in EAST, elucidating the pathway of W impurity leakage from the divertor to the core plasma under different dissipative divertor conditions. Divertor conditions from low-recycling to detachment are generated by modulating the Ne injection rate in the SOLPS simulation. With the drift velocities incorporated in DIVIMP, W transport under the drift effect is investigated. For low-density L-mode plasma conditions with favorable Bt direction, the majority of W source originates from the outer divertor target and leaks upstream through the near-SOL EB drift reversed region at both inner and outer divertors. The EB drift is demonstrated to have a significant effect on W leakage and the effect diminishes with the increase of Ne injection rate. Compared to the D2 injection cases, Ne injection helps to achieve a strong dissipative divertor condition at a much lower plasma density, resulting in a higher W leakage ability due to the smaller friction with the background plasma. For Ne injection cases under high-density H-mode conditions, W leaks mostly through the outer divertor region while the inner divertor is fully detached. Compared to the L-mode cases, the shorter power decay lengths in the scrap-off layer (SOL) of the H-mode cases result in smaller parallel EB drift velocities especially in the middle and far SOL region. With the increase of the Ne injection rate, the decrease of the EB drift in the middle SOL leads to a wider near-separatrix W leakage path. Therefore, a worse divertor W screening is expected, which is consistent with the previous published experimental observations [1].

Investigating the effect of BT direction on W source-to-core pathways during the SAS-VW campaign on DIII-D

Experiments using the V-shaped closed slot tungsten (W) coated SAS-VW divertor in DIII-D studied the effects of the BT direction on core contamination of eroded tungsten from a closed slot divertor configuration. Core W content is inferred using soft-X ray tomography (SXR) and vacuum ultraviolet spectroscopy (SPRED), while W divertor erosion is inferred from visible spectroscopy of W emission (400.9 nm) measured by in-slot filterscopes (filtered photo-multipliers). Post-mortem analysis from the campaign discovered tile misalignment leading to suspected pronounced leading-edge erosion in the unfavorable BT direction (ion B→×∇B→ drift away from divertor) likely not captured by diagnostics. However, empirical findings show up to ∼ 2-3x larger core contamination in the favorable BT direction even considering no additional W erosion from leading edges. A “source-to-core efficiency factor” is derived to estimate the effects of leading-edge erosion and compare W contamination for two pairs of H-mode discharges in opposite BT directions. While having differing absolute parameters, similar core impurity density gradients suggest comparable core impurity transport. These results show that favorable BT may have stronger source-to-core pathways for W impurities sourced from the outer divertor region. Possible explanations could include the effects of E→×B→ drifts on W transport in the scrape-off-layer (SOL) as well as previously determined fast SOL inner target directed flows in favorable BT.

Drifts effect on the divertor W leakage mechanisms under different dissipative divertor conditions of EAST

A new drift module has been developed in the DIVIMP code, enabling two-dimensional simulations of tungsten (W) transport in the edge plasma with full drifts included. By using the SOLPS-DIVIMP code package, the impact of drifts on W transport and screening has been investigated for various levels of dissipative divertor conditions and different divertor geometric configurations in EAST. Simulation results reveal that E⃗×B⃗ drifts can enhance the W leakage by more than one order of magnitude. Under the favorable Bt direction, W erosion mainly occurs on the outer divertor target, making the W leakage from the outer divertor region the dominator. A leakage path from the near-scrape-off layer (SOL) region is revealed by the modeling results. In the leakage path, both the ion temperature gradient force and the reversed poloidal E⃗×B⃗ drift are pointing upstream. With the radial E⃗×B⃗ drift pushing W ions from the well-screened far-SOL region to the near-SOL region, the leakage from the near-SOL region becomes significant. As the divertor condition varies from the low-recycling regime to the deep detachment regime, the decrease of the ion temperature gradient velocity and poloidal E⃗×B⃗ drift velocity narrows the width of the near-SOL leakage tunnel and thus enhances W screening. While under the unfavorable Bt , W erosion and leakage from the inner divertor target matters, the leakage mechanism especially the leakage path from the near-SOL region is similar as the favorable Bt cases. Furthermore, the effect of different divertor geometries on the W screening has been investigated. The configuration with the outer strike point (OSP) on the horizontal divertor plate is proved to narrow the near-SOL leakage tunnel, and thus the unreversed poloidal E⃗×B⃗ drift pointing to the divertor target dominates and helps to enhance the divertor W screening. For the same D 2 puffing rate, the W leakage ability of cases with the OSP on the horizontal target can be more than 10 times weaker than the cases with the OSP on the vertical target, especially when the divertor is detached.

Toroidal modeling of runaway electron loss due to 3D fields in ITER

Mitigation of runaway electrons (REs) by three-dimensional (3D) magnetic field perturbations is numerically investigated for the ITER 15 MA baseline D–T scenario, utilizing the MARS-F code (Liu et al Phys. Plasmas 7 3681) with a drift orbit test particle tracing module. Considered are two types of 3D fields: the n = 3 (n is the toroidal mode number) resonant magnetic perturbation (RMP) utilized for the purpose of controlling the edge localized modes in ITER, and perturbations generated by the n = 1 magneto-hydrodynamic (MHD) instabilities in a post-disruption plasma. The RMP field, applied to a pre-disruption plasma, is found to be moderately effective in mitigating the RE seeds in ITER when vacuum field model is assumed. Up to ∼40% loss fraction is possible at 90 kA-turn coil current. The mitigation efficiency is however substantially reduced, down to less than 5%, when the plasma response is taken into account. This is due to strong screening of the resonant magnetic field components by the plasma response resulting in much less field line stochasticity. On the other hand, the MARS-F modeling, based on the DINA-simulated post-disruption equilibria, shows that the n = 1 resistive kink instabilities develop in these plasmas, as the edge safety factor q a evolves and drops below integer numbers. RE mitigation by these MHD instabilities is sensitive to the eigenmode structure. The best mitigation is achieved as q a drops below 3, when a global kink instability occurs that encompasses both internal and external components. This global instability is found to be capable of mitigating over 80% MeV-level passing RE orbits at a field perturbation |δB|/B 0 that is comparable to that observed in DIII-D experiments, and full mitigation if the perturbation amplitude is doubled. The ‘wetted’ area on the ITER limiting surface, due to MHD instability induced RE loss, generally increases with the perturbation amplitude (together with increasing loss fraction). At the highest perturbation level assumed in this study, the wetted area reaches ∼60% of the total limiting surface area. The lost RE orbits mainly strike the outer divertor region of the limiting surface, with some fraction also hitting a wide area along the inboard side of the surface.

Self-sustained divertor oscillations in ASDEX Upgrade

Alternating radiation phenomena between the inner and outer divertor regions in the sub kHz range are investigated in the tokamak ASDEX Upgrade. While the inner divertor oscillates between the onset and fluctuating state of detachment, the outer divertor oscillates between conditions where it can maintain high recycling conditions or not. The detachment state of the inner divertor determines the magnitude of the neutral flux through the private flux region, thus sets the recycling conditions at the outer divertor. In return these recycling conditions determine the particle content in the divertor, hence the detachment state at the inner divertor.

Exploring SF- in-out asymmetry and detachment bifurcation in HL-2M with E × B by SOLPS

Abstract In this paper, the SF- (with the second X-point located near the outer target surface) in-out asymmetry and detachment bifurcation in HL-2M were explored by the edge plasma code SOLPS5.1 including electric but neglecting magnetic drifts. The modeling results reveal that the radial electric drift component (Epol × B) can play a key role on inducing the SF- in-out asymmetry due to the effective carbon (C) screen capacity and high heat dissipation. The Epol × B drift can enhance the poloidal electric drift component (Er × B), and their synergistic effects can induce a marked in-out asymmetry. Besides, the results showed that the detachment bifurcation in SF- also could occur under given conditions. The presence of four factors (including E × B, C impurity, normal-Bt, and relatively small radial anomalous transport coefficients) forms the necessary conditions for producing the detachment bifurcation. The root reason for the detachment bifurcation formation is due to the strong interaction of Te, E×B, and C impurity radiation loss in the outer divertor region with normal-Bt direction. The detachment bifurcation could be removed or mitigated by weakening their interaction. Finally, the time-dependent detachment bifurcation process is simulated in this work to better understand its physical picture.

Integrated infrared and visible tangential wide-angle viewing systems for surface temperature measurement and discharge monitoring in EAST

Two newly developed integrated infrared and visible tangential wide-angle (WA) viewing systems (VS) have been mounted in the Experimental Advanced Superconducting Tokamak (EAST) to provide temperature measurement and hot spot monitoring of the plasma facing components (PFC) for machine protection and further physics studies of various plasma parameters such as the scrap-off layer power fall-off length λq, the peak heat flux on the divertor target qpeak etc. The systems use endoscopes with reflective optics to obtain a wide-angle view (49.5°×62°). Via endoscope, the 3-5 μm middle wave infrared camera and the visible camera with spectral range of 400−700 nm, the WAVS could provide simultaneous real time infrared and visible image data of the first wall components in the same field of view (FOV). Some key components e.g. the 4.6 GHz lower hybrid wave (LHW) antenna, the ion cyclotron resonance frequency (ICRF) antenna, the high field side region of neutral beam injection in port F, and approximately 78° more toroidal coverage of outer divertor regions with 27° more toroidal coverage of inner divertor regions are firstly covered by the camera system. This paper presents the diagnostic description of the newly developed WAVS and preliminary experimental results observed including fast-ion losses to the inner wall during neutral beam injection (NBI), single edge localized mode (ELM) observation on upper outer divertor, hot spots on antenna guide limiters and temperature and heat flux distribution on the tungsten and graphite divertor targets in EAST.

ERO modeling and analysis of tungsten erosion and migration from a toroidally symmetric source in the DIII-D divertor

A toroidally symmetric tungsten ring inserted in the lower outer divertor of DIII-D was exposed to 25 repeated, attached L-mode shots in reverse- configuration. Radial profiles of the W gross erosion flux inferred in situ from spectroscopic measurements of the WI line (400.9 nm) during these experiments are well reproduced by ERO-D3D simulations of carbon and tungsten impurity erosion, transport and redeposition in the outer divertor region. Tungsten gross erosion is mainly induced by physical sputtering of tungsten by carbon impurities. The outward radial transport of carbon impurities in the outer divertor is shown to be mainly governed by drifts in the sheath region. In addition, the erosion and redeposition of carbon on tungsten, induced by the implantation of carbon into tungsten modeled with the homogeneous mixed material model, increases the effective flux of carbon impurities onto the tungsten ring (carbon recycling on tungsten). The dynamics of carbon implantation in tungsten is shown to be consistent with the plasma shot duration in DIII-D. Moreover, it is shown that the localized deposition of tungsten measured experimentally in the outboard region away from the tungsten ring is caused by the long-range radial transport of tungsten impurities in the outer divertor region induced by the interplay between poloidal and radial drifts. Such experimental measurements might provide direct quantitative estimations of tungsten net erosion. The modeling and analysis of carbon and tungsten erosion and redeposition presented in this paper demonstrates that various physical mechanisms and their synergistic effects need to be taken into account to accurately describe erosion, transport and redeposition of impurities in tokamak divertors.

Neutral density estimation in the ASDEX upgrade divertor from deuterium emissivity measurements during detachment and shoulder formation

Neutral density distribution in the divertor region of ASDEX Upgrade is estimated in different discharge conditions. The 2D divertor emission of Dα and Dγ is obtained with a tomographic inversion of their brightness measured by two cameras. From these emissions, 2D maps of neutrals density n0 can be obtained by a simple model that takes into account both excitation and recombination processes populating deuterium n = 3 and n = 5 levels. In attached discharges, the neutral density n0 is larger in the inner divertor, and when increasing core ne its maximum moves toward the dome and the outer divertor region. In particular, the modification of the n0 distribution is studied in plasmas where the mean electron density in the core increases, moving from an attached to a detached phase. In these discharges, the outboard midplane scrape off layer electron density increases and its profile flattens, with the formation of the so called density shoulder. The modification of the neutral density is followed during this evolution, showing that both the shape and the absolute value of the n0 distribution change in time, suggesting that they play a role in the two phenomena of shoulder formation and detachment. Moreover, in the detached phase, also the recombination contributes to the neutral deuterium emission.

Initial studies of plasma facing component surface conditioning in the national spherical tokamak experiment upgrade with the materials analysis particle probe

An innovative PFC diagnostic, the Materials Analysis Particle Probe (MAPP) was used to study the chemistry of Plasma Facing Components (PFCs) in the National Spherical Tokamak Experiment Upgrade (NSTX-U). NSTX-U used boronization as conditioning strategy during the 2015–2016 experimental campaign. Deposition with ∼9.1g of deuterated tri-methyl boron (d-TMB) in a helium glow discharge resulted in coatings with an equivalent thickness of 7.0nm at the lower outer divertor region. MAPP was used to capture for the first time the in-vacuo evolution of the chemical state of ATJ graphite PFCs after boronization and plasma exposure via X-ray Photoelectron Spectroscopy (XPS) on a day-to-day basis. The XPS data shows the formation of B4C on the PFCs surface during boronization. We observed the gradual increase of the B2O3 fraction in the coatings with plasma exposures. In contrast, we measured the formation of non-stoichiometric oxides when the samples were only exposed to residual gases and an argon vent. MAPP data revealed erosion and oxidation of deposited boron coatings over the course of tens of shots (a time resolution improved several orders-of-magnitude compared to post-campaign PFC characterization) that is correlated with the transitory nature of plasma performance enhancement with boron conditioning.

Measurements of ion cyclotron parametric decay of lower hybrid waves at the high-field side of Alcator C-Mod

Ion cyclotron parametric decay instability (PDI) of lower hybrid (LH) waves is surveyed using edge Langmuir probes on the Alcator C-Mod tokamak. The measurement is carried out simultaneously at the high-field side (HFS) and low-field side (LFS) mid-plane of the tokamak, as well as in the outer divertor region. Different LH spectra are observed depending on the location of the probes and magnetic configuration in L-mode plasmas, with drift direction downward. In lower single null (LSN) plasmas, strong ion cyclotron PDI occurring at the HFS is observed for the first time. This instability is characterized by a frequency separation in sidebands corresponding to the ion cyclotron frequency (ωci) near the HFS scrape-off layer and develops with threshold-like behavior as density increases. In inner wall limited (IWL) plasmas, this HFS instability shows a higher density threshold compared with that in LSN plasmas. The pump width becomes broadened even in the absence of the sidebands. In upper single null plasmas with similar plasma parameters, ion cyclotron PDI sidebands have a frequency separation corresponding to ωci near the LFS and are weaker than those observed in the LSN and IWL plasmas. Correlation between the onset of ion cyclotron PDI and the observed loss of lower hybrid current drive efficiency (Wallace et al 2012 Phys. Plasmas 19 062505) is discussed.

Arc erosion on W plasma facing components in ASDEX Upgrade

Abstract The occurrence of arc traces on plasma phasing components show a strong local variation. Strong erosion is observed at the inner baffle region, where arcing is more dominate than physical sputtering and in total 2 g of tungsten are eroded during one campaign. On the other hand this is about a factor of ten less than the net physical sputtering in the outer divertor region. Droplet production by arcing is a direct dust source, which covers 15% of the eroded tungsten. The penetration of these droplets, which have a typical diametre of 2 μm, into the core plasma is still unclear. Laboratory measurements of the light emission by arcing using a fast camera confirm former data, which show that arcs are active during ELMs.

Comparison of kinetic and fluid models for tungsten impurity transport using IMPGYRO and SOLPS

A benchmark calculation of tungsten impurity transport comparing kinetic modeling with IMPGYRO and multi-fluid modeling with SOLPS has been performed. The primary source of tungsten has been set at the outer mid-plane in the calculations. In both cases, impurity densities are relatively large in the SOL region near the top of the machine (top-SOL) compared to the divertor region mainly due to the effect of the thermal force. In the outer divertor region, however, a significant difference between the two codes has been observed, which could be explained by the effect of the drifts. The drifts are switched off in SOLPS in this study, while they are naturally taken into account in IMPGYRO. The drifts prevent impurities from being transported toward the outer divertor region.

Simulation Study of an Extended Divertor Leg for Heat Control in the SlimCS DEMO Reactor

AbstractPower handling in the divertor is one of the most important issues for fusion tokamak reactor design. In this study, the effect of the divertor leg length on power handling in the SlimCS DEMO reactor is investigated for future divertor design. Simulation with the SONIC code suite shows enhancement of recycling in the outer divertor region due to effects of the long divertor leg and the deep V‐shaped corner. As a result, the ion and electron temperature at the outer divertor significantly decreases. In addition, the peak of the Ar impurity radiation, which is located close to the outer divertor target in the standard divertor geometry, moves upstream. The former effect decreases the heat load to the outer divertor due to ion conduction and convection, while the latter reduces the heat load due to the impurity radiation. The peak of the total heat load on the outer divertor decreases by about 30 % (© 2012 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

Interpretation of Be migration studies at JET and validation of an integrated numerical model for plasma impurity transport and wall composition dynamics

Fusion experiments with wall cladding of different chemical elements face the problem of material mixing by the plasma–wall interaction. Understanding of experiments that investigate this long term material re-distribution require an iterative approach that takes the local erosion, the global transport of material by the scrape-off layer (SOL) plasma and the local re-deposition into account. We present a model that combines parameterized DIVIMP calculations and a simplified sputtering model to simulate the time evolution of first wall material migration in a fusion experiment. The simulations are benchmarked by Be evaporation experiments performed previously at JET. The time evolution of the spectroscopically measured Be erosion fluxes for 800s total plasma time at the main wall and inner divertor can be reproduced by the model using an extended calculation grid for DIVIMP and an increased cross-field diffusivity for impurity ions in the SOL. Discrepancies were found in the outer divertor region.

Scrape-off layer plasmas for ITER with 2nd X-point and convective transport effects

Plasma fluxes to the divertor region in ITER near the magnetic separatrix have been modeled extensively in the past. The smaller, but potentially very important fluxes to the main chamber and outer divertor regions are the focus of the present paper. Two main additions to the usual transport modeling are investigated: namely, convective radial transport from intermittent, rapidly propagating ‘blob’ events, and inclusion of the magnetic flux-surface region beyond the second X-point that actually contacts the main-chamber wall. The two-dimensional fluid transport code UEDGE is use to model the plasma, while the energy spectrum of charge-exchange neutrals to the main chamber wall is calculated by the DEGAS 2 Monte Carlo code. Additionally, the spatial distribution of beryllium sputtered from the main chamber wall is determined in the fluid limit.

Surface studies of tungsten erosion and deposition in JT-60U

In order to study tungsten erosion and migration in JT-60U, 13 W tiles have been installed in the outer divertor region and tungsten deposition on graphite tiles was measured. Dense local tungsten deposition was observed on a CFC tile toroidally adjacent to the W tiles, which resulted from prompt ionization and short range migration of tungsten along field lines. Tungsten deposition with relatively high surface density was found on an inner divertor tile around standard inner strike positions and on an outer wing tile of a dome. On the outer wing tile, tungsten deposition was relatively high compared with carbon deposition. In addition, roughly uniform tungsten depth distribution near the upper edge of the inner divertor tile was observed. This could be due to lift-up of strike point positions in selected 25 shots and tungsten flow in the SOL plasma.

Numerical analysis of the SOL/divertor plasma flow with the effect of drifts

The effects of drifts on parallel plasma flows have been studied. Numerical calculations using a 2D edge plasma simulation code (the SOLPS 5.0 code) have been done for the analysis of JT-60U W-shaped divertor plasmas in the detached state. Although significant effects of drifts on parallel plasma flows have not been observed in the inner and outer divertor regions, the flow reversal due to drifts has been observed at the outer mid-plane. The Pfirsch–Schlüter flow possibly explains the flow reversal.

Molecular contribution to the Dα emission in the divertor of the ASDEX Upgrade experiment

Spectral line emission from the deuterium Balmer α line and from molecular deuterium in the Fulcher band, d 3 Π u→a 3 ∑ g +, was studied in the outer divertor region of the ASDEX Upgrade plasma with a pair of relatively calibrated spectrometers. A dependence of the signals on the proximity of the plasma strike point and on divertor gas puffing was noted. The rotational spectrum of the lowest diagonal vibrational molecular bands indicated rotational temperatures of ca. 1000 K and hence, through calculation of relative upper state populations, a ground state vibrational temperature of ca. 4000 K. Calculating the particle fluxes, the molecular deuterium was found to contribute up to 1% of the Dα intensity, the relative importance of the molecules at some divertor heights was affected by the position of the strike point.

Tritium distribution on plasma-facing tiles from ASDEX Upgrade

Tritium surface distribution on the plasma-facing tiles used in the divertor region and the central column of ASDEX Upgrade was measured by Tritium Imaging Plate Technique. The tritium intensity is high at the dome baffle underneath the X point, the outer divertor region and the mid-plane or lower side of the central column. In contrast, little tritium was retained in the carbon deposition found in the inner strike point module tiles. Such profile is very similar to that observed in JT-60U, indicating that long-term tritium retention in ASDEX Upgrade was dominated by the implantation of energetic tritons escaping from the main plasma without losing their initial energy (1MeV) rather than the codeposition of eroded carbon and hydrogen isotopes. However, such tritium profile was modified by carbon deposition and thermal load, particularly on the divertor target tiles and the central column.