Divertor Target Plates Research Articles

Disconnection of edge coherent modes between the outer midplane and divertor target in EAST H-mode plasma

Abstract The edge coherent mode (ECM) is considered a highly attractive pedestal mode as it prolongs the duration of edge-localized modes (ELMs), increases particle and impurity transport without significantly affecting energy transport, and operates compatibly with high-performance plasma discharges. The ECM can also be detected by Langmuir probes on the divertor target plate, indicating that it extends from the pedestal region into the SOL, resulting a connection to the divertor target plate through magnetic field line. In this work, the distribution of ECM on the divertor target plates are investigated through the analysis of 215 upper single null discharges on the EAST tokamak. The coherence analysis of plasma fluctuation between the ECE signal at pedestal region of outer midplane and the ion saturation current measured by Langmuir probes in the divertor region reveals that the ECM is hardly detected by the divertor probe close to the outer strike point (OSP) but can be observed at far SOL. This indicates that there exists an ECM quiescent region in the near SOL on divertor plate and the extent of the quiescent region in poloidal flux coordinate (Δ) has been statistically analyzed. A pronounced relation between Δ and triangularity (δ) has been observed, i.e. Δ increasing with δ. Further analysis shows that this relation could be related to averaged magnetic shear in the SOL. This result is consistent with the physical picture that the strong magnetic shear close to X-point significantly squeezes the cross-section of flux-tubes down to scales where collisions dominate.

Toroidal distribution of heat load on castellated plasma facing components for lower divertor target in EAST

In tokamak devices, the performance of plasma facing components (PFCs) under high heat loads is a key challenge for achieving long-pulse and high-performance plasma operations. The lower divertor of EAST adopts an ITER-like cassette modules (CMs) structure. However, regular distribution of damage along the toroidal direction have been observed on the divertor target plates, which not only affects the lifespan of PFCs but also limits the long-pulse and high-performance plasma operations. Therefore, investigating the toroidal distribution of the heat load on the divertor target in EAST is crucial for understanding the mechanism of damage on the target plate surface. The toroidal characteristics of heat loads on CMs of the lower divertor in EAST was analyzed using a three-dimensional magnetic field line tracing program (PFCFlux). For the inner vertical target (IVT) and outer vertical target (OVT) within a single CM, the heat load varies monotonically along the toroidal direction. In contrast, the heat load distribution on the outer horizontal target (OHT) is uniform, except at the chamfers, which is directly correlated with the target’s geometric structure and the magnetic field configuration. The surface heat load distribution exhibits toroidal non-uniformity, with the heat deposition pattern being generally consistent and displaying periodic variations. The toroidally uneven heat load distribution is attributed to the change in the magnetic field’s tilt angle. Additionally, the peak heat load is observed at the chamfer positions between cassettes, and an analysis is conducted on the relationship between peak heat load and the size of the chamfer. Smaller tilt angles result in lower surface heat load. Furthermore, a comparative analysis of peak heat loads at the chamfer under different plasma current conditions reveals that higher plasma currents generally result in increased peak heat loads. This investigation into the toroidal heat load distribution on the divertor surface contributes to understanding of the damage mechanism of W PFC as the target plate and provides valuable data for divertor design of future fusion devices.

Dependence of ELM instability on separatrix density in EAST long-pulse H-mode plasmas

Abstract The transition from small edge-localized modes (ELMs) to large ELMs has been repetitively observed in minute-scale long-pulse high-confinement mode (H-mode) discharges in 2017 EAST campaign. The appearance of large ELMs is found to be strongly correlated with the decrease in separatrix density due to the gradual decrease in fuel recycling during the long-pulse H-mode operations (LPHOs). By numerical scan of separatrix density with a fixed temperature profile, it has been found that the dependence of ELM instability on separatrix density is related to the competition between the ion diamagnetic stabilizing effect and destabilizing effect of pressure gradient and current density in the pedestal region, shedding light on a comprehensive understanding of different roles of separatrix density in ELM instability observed in EAST experiments. With a high separatrix density, the ideal ballooning mode could be destabilized near the separatrix, which is thought to help achieve small ELMs in EAST LPHOs. In 2021 EAST campaign, the experiment of large ELMs control has been performed through actively changing fuel recycling by moving strike point location (SPL) on the lower tungsten divertor target plate. It has been demonstrated that the mitigation of large ELMs is strongly correlated with the significant increase in separatrix density, which is thought to be attributed to a higher ionization source in the scrape-off layer (SOL) region by SOLPS-ITER simulation. The high ionization source in the SOL region is believed to provide a strong fueling effect near the separatrix and thus raise the local density, which is considered as an important reason for triggering ballooning instabilities near the separatrix and achieving small ELMs.

A long-pulse small edge-localized-mode high-confinement plasma with detachment feedback control by floating potential in an experimental advanced superconducting tokamak in a metal wall environment

Abstract One of the key challenges facing the magnetic fusion research is to demonstrate the compatibility between high confinement and radiative divertor in long-pulse discharges with a metal wall environment. A small edge localized mode (ELM) high confinement plasma is obtained in the upgraded lower divertor of Experimental Advanced Superconducting Tokamak (EAST) with an energy confinement factor H98~1.1 and a Greenwald density fraction fGW ~ 0.65 maintained for 26 s, and periodical detachment is achieved through active control of neon impurity seeding in this long-pulse discharge. For the divertor region, partial detachment is achieved periodically on the outer divertor target plates with the plasma temperature near the outer strike point decreased to below 5 eV and peak surface temperature on the outer divertor target plates maintained below 350C. The peak heat flux of the lower outer divertor decreases significantly and its profile along the target becomes very flat in the detached state. Two low-frequency (<10 kHz) fluctuations that are related to rippling mode caused by a resistive instability appear in the detached state. For the pedestal region, the electron pressure profile is flatter and ELM amplitude is smaller in the detached state than that in the attached state. Edge coherent mode (ECM) appears in the attached state and disappears in the detached state. To achieve this experimentally, a new impurity seeding feedback control scheme is applied, where the floating potential measured by divertor Langmuir probes is used as the feedback sensor, which is more reliable in the long-pulse discharges with high heat fluxes, thus more suitable for application in future devices. This work provides a new approach for the actively controlled radiative divertor as a solution to the divertor heat loads of the future fusion reactors.

Design of a 3D-printed liquid lithium divertor target plate and its interaction with high-density plasma

A liquid Li divertor is a promising alternative for future fusion devices. In this work a new divertor model is proposed, which is processed by 3D-printing technology to accurately control the size of the internal capillary structure. At a steady-state heat load of 10 MW m−2, the thermal stress of the tungsten target is within the bearing range of tungsten by finite-element simulation. In order to evaluate the wicking ability of the capillary structure, the wicking process at 600 °C was simulated by FLUENT. The result was identical to that of the corresponding experiments. Within 1 s, liquid lithium was wicked to the target surface by the capillary structure of the target and quickly spread on the target surface. During the wicking process, the average wicking mass rate of lithium should reach 0.062 g s−1, which could even supplement the evaporation requirement of liquid lithium under an environment > 950 °C. Irradiation experiments under different plasma discharge currents were carried out in a linear plasma device (SCU-PSI), and the evolution of the vapor cloud during plasma irradiation was analyzed. It was found that the target temperature tends to plateau despite the gradually increased input current, indicating that the vapor shielding effect is gradually enhanced. The irradiation experiment also confirmed that the 3D-printed tungsten structure has better heat consumption performance than a tungsten mesh structure or multichannel structure. These results reveal the application potential and feasibility of a 3D-printed porous capillary structure in plasma-facing components and provide a reference for further liquid−solid combined target designs.

Modeling of tungsten impurity transport and distribution in EAST based on multi-fluid and kinetic Monte Carlo simulations

In recent years, the materials of plasma facing components, such as divertor target plates, domes, and outer walls of tokamaks, such as ASDEX Upgrade, WEST, JET, EAST, and ITER, have been changed from carbon to tungsten because of its lower erosion and tritium retention rates. Impurities are produced by interactions between the plasma and the first wall. This study provides an investigation to simulate the transport and distribution of tungsten impurities in the edge plasma on EAST. The 2D multi-fluid edge plasma transport code SOLPS-ITER and 2D kinetic Monte Carlo impurity transport code DIVIMP were used in the simulations. The multi-fluid model in SOLPS-ITER and the kinetic Monte Carlo model in DIVIMP were employed to treat tungsten impurity ions. The 2D density contour distributions in the computational region and the 1D density radial profiles at the inner and outer midplanes of tungsten impurity particles with ionization states (W0–W+74) and the total tungsten particles with all charge states were obtained. With the heating power 1.5 MW and the line-averaged plasma density 2 × 1019 m−3, the results from SOLPS-ITER and DIVIMP show that the maximum density of tungsten ion with single ionization state is about 1014 m−3 and the total density of tungsten impurities with all charge states is about 1015 m−3 at the core boundary. To the best of our knowledge, the simulation results from SOLPS-ITER and DIVIMP are compared for the first time to benchmark SOLPS-ITER with the multi-fluid mode and DIVIMP with the kinetic model for tungsten impurity transport. The density distributions of tungsten impurities with different ionization states from SOLPS-ITER and DIVIMP are highly similar, and good agreement can be found under similar conditions involved in the calculation. From the comparison benchmark between SOLPS-ITER and DIVIMP for tungsten impurity transport, it can be concluded that the impurity transport approximation used by DIVIMP is good.

Turbulence simulations with BOUT++ by using SOLPS grids for SOLPS/BOUT++ coupling

AbstractThe coupling of transport code SOLPS with the turbulence code BOUT++ was reported in Reference [D. R. Zhang et al., Phys. Plasmas 26, 012508 (2019)], while the grids of SOLPS and BOUT++ are not completely consistent with each other, especially in the divertor region. In the present work, a method of replacing the grids of BOUT++ with the grids of SOLPS is proposed to make the simulation region fully consistent with each other for the SOLPS/BOUT++ coupling. A SOLPS grid file is generated with an MHD equilibrium and used in BOUT++ code to simulate the profiles of plasma density, ion temperature, and electron temperature with the six‐field two‐fluid model. The profiles of the main plasma parameters simulated with the SOLPS grids are similar with the profiles simulated with the BOUT++ grids at the midplane, while the profiles are deformed compared with the profiles simulated with the BOUT++ grids at the outer divertor target because of the differences of the distributions of SOLPS grids and BOUT++ grids in the divertor region. The radial particle transport coefficient and heat transport coefficients are also calculated by using the BOUT++ code with the two grids, and the comparisons of the radial particle transport coefficient and heat transport coefficients simulated with the two grids at the midplane and outer divertor target plate are discussed.

Modeling of heating scheme impacts on neon seeded plasmas with the resonant magnetic perturbation application in the large helical device: ECR and NBI heating

Plasma parameter profiles along the effective minor radius of flux surfaces are modeled for the conditions with neon impurity puffing and external resonant magnetic perturbations at the plasma edge in the large helical device. For the electron cyclotron heating scenario, calculations qualitatively reproduce stable steady states with plasma detached from divertor target plates and radiation losses exceeding 90% of the heating power. It is unlikely with neutral beam injection heating that the plasma collapses if the radiation approaches a critical level noticeably below 100%. The latter is explained by the dependence of neutral beam attenuation on the plasma ion effective charge.

On the use of error field correction coils in JET

This work describes the usage of Error Field Correction Coil (EFCC) system [Barlow I. et al. 2001 Fusion Eng. Des. 58–59 189], which is a set of 4 coils located external to the vessel of the JET device, with the aim of introducing non-axisymmetric n = 1 magnetic field perturbations in various targeted plasma experiments. Besides being used to characterize and correct the intrinsic error field, the EFCCs system has been exploited to i) probe plasma stability in high-βN regimes through the MHD spectroscopy technique, allowing the identification of resistive-wall mode stable operational scenarios, ii) study radiation asymmetry when testing the shattered pellet injector as the prime candidate for ITER disruption mitigation system, and iii) control edge localized modes, permitting a reduction in heat fluxes and carbon erosion of the divertor target plates.

Effects of toroidally-distributed-divertor biasing on scrape-off-layer (SOL) current drive, divertor particle flux and fast electron confinement in the QUEST spherical tokamak

A novel divertor biasing by four biasing plates distributed toroidally (TDDB) on the upper divertor target plate is applied to low density tokamak plasmas started-up by 28 GHz 2nd harmonic electron cyclotron current drive (ECCD) in a quasi-double null configuration of the QUEST spherical tokamak (ST). In the ST plasmas of line averaged electron density < ne > ∼0.7–1 × 1018 m−3, about 20%–40% of the current Ibias driven by ∼85 V sawtooth bias voltage reaches the lower divertor plate along the biased scrape-off-layer (SOL) flux tube as the SOL current ISOL . The fact ISOL is noticeably lower than Ibias indicates an appreciable leakage of parallel current from the biased SOL flux tube. The leakage currents in the toroidal and radial directions are confirmed by detection of them using the unbiased plates in the TDDB experiments. From the ion saturation current density profile obtained by a divertor Langmuir probe array, the fall-off lengths of divertor particle flux are estimated together with the strike line position. Total particle flux to the upper divertor, evaluated by the integrated ion saturation current density profile is reduced by up to 45% during positive biasing of the TDDB, depending on the position of the strike line to the biased plate. In addition, the TDDB also induces a noticeable loss of fast electrons produced by ECCD, leading to an ∼2% reduction in the maximum toroidal current of the ST plasma compared to a shot without the TDDB. Reduction of the divertor particle flux and enhancement of the fast electron losses are thought to be dominantly caused by E ×B drift induced by the TDDB. In the present experimental conditions, the effects of magnetic perturbations produced by the SOL currents on the fast electron losses can be neglected because of a too small SOL current.

Electromagnetic turbulence simulation of tokamak edge plasma dynamics and divertor heat load during thermal quench

The edge plasma turbulence and transport dynamics, as well as the divertor power loads during the thermal-quench phase of tokamak disruptions, are numerically investigated with BOUT++’s flux-driven six-field electromagnetic turbulence model. Here, transient yet intense particle and energy sources are applied at the pedestal top to mimic the plasma power drive at the edge induced by a core thermal collapse, which flattens the core temperature profile. Interesting features, such as surging of divertor heat load (up to 50 times) and broadening of heat-flux width (up to four times) on the outer-divertor target plate, are observed in the simulation, in qualitative agreement with experimental observations. The dramatic changes in divertor heat load and width are due to the enhanced plasma turbulence activities inside the separatrix. Two cross-field transport mechanisms, namely, the E × B turbulent convection and the stochastic parallel advection/conduction, are identified to play important roles in this process. First, an elevated edge pressure gradient drives instabilities and subsequent turbulence in the entire pedestal region. The enhanced turbulence not only transports particles and energy radially across the separatrix via the E × B convection, which causes the initial divertor heat-load burst, but it also induces amplified magnetic fluctuation . Once themagnetic fluctuation is large enough to break the magnetic flux surface, magnetic flutter effect provides an additional radial transport channel. In the late stage of our simulation, reaches to 10−4 level that completely breaks magnetic flux surfaces such that stochastic field lines are directly connecting pedestal top plasma to the divertor target plates or first wall, further contributing to the divertor heat-flux width broadening.

An innovative divertor concept, the fish tail divertor, for reducing the surface temperature on the divertor target plate in EAST tokamak experiments

An innovative divertor concept, the fish tail divertor, is proposed in this paper, aimed at reducing the surface temperature on the tokamak divertor plate as well as that due to the edge localized modes. This new concept has been implemented in experiments to demonstrate its capability of strike point sweeping on the plate at a frequency range from 10 to 100 Hz by using an oscillating magnetic field. A strike point movement of 5–6 cm is achieved by applying a coil current of several percent of plasma current, leading to a significant reduction of divertor surface temperature. The result indicates a possible application in a fusion reactor.

Long-pulse H-mode operation with stored-energy monitoring for detachment feedback control with a new lower tungsten divertor in EAST

One of the key challenges for future fusion research is to mitigate the high steady-state heat load on the divertor target plates, and divertor detachment offers a promising solution. The Experimental Advanced Superconducting Tokamak (EAST) has developed several feedback control methods for divertor detachment. However, when an off-normal event momentarily disturbs the main plasma, impurity seeding may still be conducted by these methods for detachment, which probably drives the main plasma further away from its stable equilibrium or even causes the disruption. These off-normal events include excessive impurity seeding, loss of heating and dust droplets, which are not rare in current tokamak experiments, especially in long-pulse operation. To compensate for the drawbacks of these methods, we propose and develop a module of stored-energy monitoring to ensure stable plasmas in long-pulse operation. The stored energy usually decreases when the main plasma is away from its stable equilibrium, which is suitable for monitoring the state of the main plasma. Once the stored energy falls below a certain threshold, the module actively switches off the impurity seeding system. Without impurity seeding, the main plasma can recover with the increase in the stored energy. Only when the stored energy exceeds another threshold does the module switch on the impurity seeding to continue the detachment operation. The module function has been verified during the EAST radiative divertor experiments in the newly-upgraded lower tungsten divertor. A typical ∼20 s discharge in grassy-ELM H-mode regime with ∼5 MW source heating power is demonstrated with divertor partial detachment and good energy confinement by active impurity seeding (50% neon, 50% D2). The energy confinement factor is maintained at a high level, i.e. The electron temperature in the core region only has a slight change after the impurity seeding, while the electron density has a ∼10% increase. Furthermore, the ion temperature near the axis also has a remarkable increase. These achievements provide an important demonstration of the actively controlled radiative divertor mitigating the heat loads with good core confinement, which is an essential step toward steady-state operation of fusion reactors.

Investigation of ELM-related Larmor ion flux into toroidal gaps of divertor target plates

A detailed assessment of the thermo-mechanical limits of the International Tokamak Experimental Reactor (ITER) divertor with respect to potential excessive local transient heat loads due to edge localised modes (ELMs) has revealed a particular power loading scenario arising from the fact that ELM ions expelled from the upstream pedestal region will arrive at the divertor target plates without substantial thermalisation. As a consequence of their Larmor gyration around magnetic field lines, they are able to penetrate toroidal gaps between individual monoblocks of the target plate structure and can deliver rather intense heat loads to monoblock side faces near the gap entrance. To verify that this ELM-induced loading, predicted by both ion orbit simulations and particle in cell simulations, really does occur, two dedicated experiments have been performed on the ASDEX Upgrade tokamak. In both experiments a model toroidal gap structure of similar dimensions to those of the ITER divertor target monoblocks was exposed to a series of identical H-mode discharges with strong type-I ELMs. The effects arising from the gyro motion of hot ELM ions were identified by inverting, in the second experiment, the directions of both toroidal field and plasma current, thus reversing the ion gyration direction. The local distribution of incident ion flux on the gap side faces was quantified by pre- and post-exposure analysis of platinum marker layers to determine quantitatively the erosion rate of the platinum marker. The results fully confirm the ion orbit code predictions with respect to the penetration depth of incident ions with gyro orbits of similar or larger radius than the gap width. Moreover, the results confirm that ELM ions do indeed arrive at the divertor with their typical pedestal energies and also allow conclusions to be drawn regarding the corresponding intra-ELM ion particle and power flux, which is not easy to quantify using Langmuir probes.

Overview over the neutral gas pressures in Wendelstein 7-X during divertor operation under boronized wall conditions

During the first test divertor campaign of the stellarator experiment Wendelstein 7-X (Pedersen et al 2022 Nucl. Fusion 62 042022), OP1.2b, 13 neutral gas pressure gauges collected data in different locations in the plasma vessel, enabling a detailed analysis of the neutral gas pressures, the compression ratios and the particle exhaust rates via the turbomolecular pumps in the different magnetic field configurations. In Wendelstein 7-X, the edge magnetic islands are intersected by the divertor target plates and used to create a plasma-wall interface. As the number and position of the magnetic islands varies in different magnetic field configurations, the position of the strike line on the target plates and thus the neutral gas pressure in the subdivertor differs between the configurations. Neutral gas pressures on the order of few 10−4 mbar were measured in the subdivertor region. The highest neutral gas pressure of mbar was obtained in the so-called high iota configuration featuring four edge magnetic islands per cross section. The neutral particle flux through the pumping gaps into the subdivertor volume was provided by EMC3-EIRENE simulations and allowed to analyze the relation between the particle flux entering the subdivertor and the pressure distribution in the subdivertor. Finite element simulations in ANSYS provide a detailed picture of the pressure distribution in the subdivertor volume and agree with the neutral gas pressure measurements in the subdivertor in the standard configuration featuring an island chain of 5 edge magnetic islands. Surprisingly high neutral gas pressures that were not predicted by the simulation were measured in the subdivertor region away from the main strike line for discharges in the most used magnetic configuration, the standard configuration. While the pressure ratio between the two sections of the subdivertor volume, the low and high iota section is 0.06 in high iota configuration, a ratio of 2–5 was obtained in the other configurations, indicating significant particle loads and exhaust rates on the high iota section of the subdivertor in magnetic configurations with the main strike line on the low iota divertor targets.

Hot Spot Cause Analysis and Shaping Optimization Design of the Monoblock Divertor Target Plate in EAST

The monoblock divertor target plate (MDTP) is a mainstream divertor target plate. MDTP was installed in EAST and in WEST and will be used in ITER. Local high-temperature hot spots (HS) were observed on MDTP during a plasma experiment. HS will reduce the lifetime of MDTP. In this paper, the causes of HS on MDTP are determined through theoretical analysis and are verified by numerical simulations. The HS on MDTP seem to be caused by small high-density heat load areas on the toroidal and poloidal direction surfaces facing the incident direction of the plasma strike line (PSL) of the MDTP tungsten block. When toroidal HS and poloidal HS appear simultaneously, a super local high-temperature HS will be formed at the corner (facing the incident direction of PSL) of the MDTP tungsten block. The HS on MDTP can be eliminated by optimizing the geometry of the MDTP tungsten block, when the plasma configuration is determined. A method and the scope of application of the method, which can be used for tungsten block geometry optimization, are given in this paper. In order to facilitate the selection of a divertor configuration, the heat flux–carrying performance of the optimized MDTP was evaluated. In order to attain a maximum temperature within MDTP of less than 900 K, it was found that if the poloidal incidence angle between PSL and MDTP can be stably controlled as 5 deg (or 35 deg), MDTP can directly withstand PSL with a peak heat flux density of no more than 90 (or 40 ).

Characterisation of divertor detachment onset in JET-ILW hydrogen, deuterium, tritium and deuterium–tritium low-confinement mode plasmas

Measurements of the ion currents to and plasma conditions at the low-field side (LFS) divertor target plate in low-confinement mode plasmas in the JET ITER-like wall materials configuration show that the core plasma density required to detach the LFS divertor plasma is independent of the hydrogenic species protium, deuterium and tritium, and a 40%/60% deuterium-tritium mixture. This observation applies to a divertor plasma configuration with the LFS strike line connected to the horizontal part of the LFS divertor chosen because of its superior diagnostic coverage. The finding is independent of the operational status of the JET cryogenic pump. The electron temperature (Te) at the LFS strike line was markedly reduced from 25 eV to 5 eV over a narrow range of increasing core plasma density, and observed to be between 2 eV and 3 eV at the onset of detachment. The electron density (ne) peaks across the LFS plasma when Te at the target plate is 1 eV, and spatially moves to the X-point for higher core densities. The density limit was found approximately 20% higher in protium than in tritium and deuterium-tritium plasmas.

Characterization of the peak temperature on upper W divertor target during D and He plasma discharges in EAST

• The peak temperature on observed targets during both D and He discharges increase with the increase of heating power. • During the same heating power, the peak temperature during D discharges is generally lower than that during He discharges no matter L and H modes. • The sharp temperature leap induced by ELM induced transient heat load during D discharges is usually 2 times higher than that during He discharges. During the tokamak plasma discharge, the strong interaction between the edge plasma and the wall material leads to the overheating of the surface of divertor target plate and the possible melting of the material, which will seriously influence the safe operation of the device. Furthermore, transient heat load induced by edge localized modes (ELMs) is expected to cause surface damage of tungsten (W) based plasma-facing materials (PFMs) in future fusion devices. In order to ensure safe steady-state operation, it is prerequisite to obtain the peak temperature as well as distribution on the divertor target plate in real time. In EAST, a high temporal and spatial resolution infrared camera (IR) located in H port was built to observe the outer target on the L port of upper W divertor. Currently, the maximum temperature measured by IR camera has been characterized and analyzed during deuterium (D) and helium (He) plasma discharges. In addition, the ELM-induced temperature rise during H mode operations for D and He plasma discharges was specially studied. It is found that the maximum temperature on observed targets during both D and He discharges increase with the increase of heating power. In case of the same heating power, the maximum temperature during D discharges is generally lower than that during He discharges no matter L and H modes, which is more obvious when the total heating power exceeds 3 MW. However, the sharp temperature rise by ELM induced transient heat load during D discharges is usually higher than that during He discharges. Such qualitative and statistical analysis on the maximum temperature on upper outer divertor provide important reference for future operation of EAST and other tokamaks..

Simulation of effect of thermionic emission on magnetized sheath near target plate of tungsten divertor

The high confinement mode (H-mode) is a preferred operation mode of tokamak devices in the future, but the burst of edge localized mode (ELM) will sharply increase the heat load deposited on the divertor target, raising the target temperature rapidly and strengthening surface thermionic emission. In this paper, a one-dimensional fluid model is used to simulate the influence of thermionic emission on the characteristics of the magnetized sheath. The results show that the amplitude of float potential and the electric field strength both decrease under the action of thermionic emission. Plenty of thermionic emission electrons leave the target, resulting in a region with negative charge density near the target plate, and the magnetized sheath is divided into two parts: ion sheath and electron sheath. In the electron sheath, with the rise of the target surface temperature, electrons accumulated in front of the target also increase, the potential distribution is non-monotonic, and a “virtual cathode” structure appears. The reverse electric field formed near the target will confine the thermionic emission electrons leaving the target and slow down the ion movement, leading to a decrease of the ion energy deposited on the target. With the increase of the angle between the magnetic field and the target normal, the potential of the magnetized sheath and the proportion of the thickness of the electron sheath in the magnetized sheath both increase. The virtual cathode potential decreases, the temperature of the target required to form the virtual cathode rises.

E × B flow driven electron temperature bifurcation in a closed slot divertor with ion B × ∇B away from the X-point in the DIII-D tokamak

An electron temperature bifurcation is observed in the small angle slot divertor, which has been developed to enhance neutral cooling across the divertor target by coupling a closed slot structure with appropriate target shaping. Experiments in the DIII-D tokamak and associated SOLPS-ITER modeling with full drifts find a strong interplay between drifts and divertor geometry on divertor dissipation. The coupling of divertor geometry and drift flows can strongly affect the path towards divertor detachment onset as the plasma density is raised. With the strike point on the inner slanted surface and ion B × ∇B away from the magnetic X-point, bifurcative transitions were observed with sharp decrease of T e towards detachment onset both experimentally and computationally. This differs from the situation for the open divertor where the T e cliff was only observed for ion B × ∇B towards the X-point. SOLPS-ITER modeling with full drifts demonstrates that the magnitude of the E × B drift flow is comparable with the main plasma flow. The reversal of both the poloidal and radial E × B flows near the strike point leads to rapid density accumulation right near the separatrix, which results in bifurcative step transition of divertor conditions with cold plasma across the entire divertor target plate. These results indicate that the interplay between geometry and drifts should be fully taken into account in future fusion reactor divertor designs.