Divertor Target Research Articles

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.

Electron density pedestal behaviour in strike-point sweeping experiment on JET

Strike-point sweeping, a technique often used to spread heat loads on divertor targets, was employed in JET experiments for the first time to generate an edge-localized modulated particle source for investigating plasma fuelling and particle transport in the edge region. This approach was motivated by the possibility of achieving higher modulation frequencies than those available from traditional gas puff modulation at JET. Higher frequencies would enable the collection of more edge-localized information from the electron density response to the modulated particle source. Various sweeping frequencies, up to 18.5 Hz, were commissioned and utilized in the experiments. Both strong and weak electron density responses were observed in H-mode plasmas, depending on the strike-point configuration and the distance the strike-points moved during the sweep cycle. The electron density response exhibited complex and unconventional behaviour (compared to gas puff modulation), which presented challenges for interpretation. In this study, we analyse one experiment in detail using an optimization framework in which transport and particle source parameters are determined by fitting our forward model parameters to the experimental electron density measurements. We demonstrate that a consistent picture emerges and that our approach can provide new insights into these complex data. However, we note that while strike-point sweeping generates the desired modulated edge-localized particle source, it also modifies the properties of the edge transport barrier. Therefore, the strike-point sweeping methodology is a promising but challenging way to study edge particle transport and edge fuelling properties, requiring very precise measurements.

Modeling of plasma beta effects on the island divertor transport in the standard configuration of W7-X

The influence of plasma beta effects on the edge plasma transport in the Wendelstein 7-X standard configuration is studied systematically by using EMC3-EIRENE combined with a 3D equilibrium code named HINT. The magnetic topology changes induced by plasma beta effects are significantly reflected in plasma transport behaviors and heat flux patterns on divertor targets. After validating the modeling strategy by comparisons with experimental data, the extended simulations for high performance plasmas show that the threshold separatrix density for accessing the power detachment is reduced in higher beta plasmas. Compared with the vacuum field case, the impurity radiation distributions with finite beta effects are modified in the magnetic island region. The divertor heat flux is distributed more evenly along the toroidal direction on the strike line at the vertical target. The strike line on the horizontal target moves towards the pumping gap with an increase in the plasma beta. In addition, the different pressure profiles with the same central beta also result in a modified heat flux pattern on the divertor targets.

Separatrix parameters and core performances across the WEST L-mode database

WEST database analysis shows a correlation of the recycled neutral source around the separatrix with core performances. This observation questions the causality chain between particle source and turbulent transport up to the core in L-mode, high recycling plasmas, an unavoidable phase of all scenarios. The best core performances correlate with the lowest values of the density at the separatrix, , similarly to ASDEX Upgrade (AUG) tokamak and Joint European Torus (JET) tokamak in H-mode (Verdoolaege et al 2021 Nucl. Fusion 61 076006). Reflectometry in the midplane provides , while the temperature at the separatrix, is inferred by the ‘two-point model’ using Langmuir probe data on divertor targets. Lower separatrix resistivity does not correlate with better core performances, unlike H-mode observations (Eich et al 2020 Nucl. Fusion 60 056016). As expected in the presence of an efficient neutral source due to recycling fluxes, correlates with the D recycled particle flux at the divertor measured by visible spectroscopy. Coherently, at a given controlled central line integrated density , lower correlates with a larger density gradient around the separatrix as well as a larger global density peaking, , measured by interferometry. The latter correlates as well with lower collisionality in the core, similarly to JET and AUG H-modes (Angioni et al 2007 Nucl. Fusion 47 1326). The correlations reported allow phrasing the subsequent causality question: what is the interplay chain between low neutral recycling at the divertor plates, low density at the separatrix, high density peaking at the separatrix, high global density peaking, higher central temperature and better core energy confinement quality? Understanding the causality chain is essential to prepare ITER operation and design DEMO scenarios where the ratio of the divertor leg to the ionization length will be larger and where the pumped flux with respect to the plasma volume will be lower than presently operating tokamaks.

Compact Radiative Divertor Experiments at ASDEX Upgrade and Their Consequences for a Reactor.

We present a novel concept to tackle the power exhaust challenge of a magnetically confined fusion plasma. It relies on the prior establishment of an X-point radiator that dissipates a large fraction of the exhaust power before it reaches the divertor targets. Despite the spatial proximity of the magnetic X point to the confinement region, this singularity is far away from the hot fusion plasma in magnetic coordinates and therefore allows the coexistence of a cold and dense plasma with a high potential to radiate. In the compact radiative divertor (CRD) the target plates are placed close to this magnetic X point. We here report on high performance experiments in the ASDEX Upgrade tokamak that indicate the feasibility of this concept. Despite the shallow (projected) field line incidence angles of the order of θ_{⊥}=0.2°, no hot spots were observed on the target surface monitored by an IR camera, even at a maximum heating power of P_{heat}=15 MW. And even with the X point located exactly on the target surface and without density or impurity feedback control, the discharge remains stable, the confinement good (H_{98,y2}=1), hot spots absent, and the divertor in a detached state. In addition to its technical simplicity, the CRD scales beneficially to reactor-scale plasmas that would benefit from an increased volume of the confined plasma, more space for breeding blankets, smaller poloidal field coil currents, and-potentially-an increased vertical stability.

A reduced-model (nSOLT) simulation of neutral recycling effects on plasma turbulence in the divertor region of MAST-U

The 2D scrape-off-layer turbulence code (nSOLT), which includes 1D kinetic neutral–plasma interactions, is applied to study effects of neutral recycling on plasma turbulence for parameters illustrative of the MAST-U divertor region. Neutral recycling is modeled by injecting a fraction of the parallel plasma flux to the divertor back into the simulation domain as a source of Franck–Condon-distributed neutrals. Stationary sources, concentrated at the magnetic separatrix, model plasma streaming into the divertor region from the upstream scrape-off-layer and sustain plasma turbulence absent neutral recycling. Starting from one such no-neutrals equilibrium, we initiate recycling in a numerical experiment designed to diagnose and identify the effects of various neutral–plasma interactions on the divertor plasma, divertor turbulence, and plasma exhaust. The onset of recycling triggers an initial burst of enhanced cross field plasma transport that is quelled by ionization cooling and charge–exchange (CX) friction, with growing neutral pressure, leading to a quiescent, turbulence-free state. Diagnosis of this transient burst reveals that (1) the sudden increase in plasma density due to ionization dominates the onset of the burst, (2) electron cooling due to ionization increases collisionality and disconnects blob filaments from the sheath, and (3) CX friction drives tripole polarization of a blob that can dominate the curvature-driven dipole polarization, leading to the stagnation of blob propagation and reduced radial turbulent transport. It is shown that CX friction is negligible compared to sheath physics in determining equilibrium mean flow shearing rates, for parameters considered herein (specifically a short connection length to the divertor target), while it can significantly reduce interchange-instability growth rates.

Experiment with nitrogen seeding at the Globus-M2 tokamak

First experiment with nitrogen seeding has been performed at the compact spherical tokamak Globus-M2. Significant reduction of the electron temperature and the energy flux to the outer lower divertor target has been observed experimentally and reproduced in the modeling with the SOLPS-ITER code.

SOLPS-ITER modeling of deuterium throughput impact on the ITER SOL plasma

The impact of the ionization source in the ITER far SOL associated with main chamber deuterium puffing on the overall SOL performance is studied numerically using SOLPS-ITER modeling. It is found out that this source can have significant influence on the retention of the seeded impurity radiator (Ne) in the divertor and on plasma temperature in far SOL. The former is responsible for typical separatrix concentration of seeded impurity compatible with the required power handling limits on the divertor targets, while the latter is responsible for sputtering of divertor material. Both factors indicate the beneficial effect of high fuel throughput on separatrix and pedestal plasma composition.

Variation in the volumetric power and momentum losses in the JET-ILW scrape-off layer

In attaining the cold plasma condition at the divertor target via gas puffing and/or nitrogen seeding, a large variation in the pressure-momentum and cooling losses has, for the first time, been demonstrated on JET with the ITER-like wall, leveraging novel target electron temperature measurements and JET’s unique capability for disentangling the main-ion-driven net dissipation effects from impurity radiation. These observations provide key insights into the dependence of volumetric losses on the dominant dissipation mechanisms, whether due to neutral-plasma interaction processes, impurity radiation induced detachment, and their interplay.While numerical code studies indicate that both volumetric cooling and pressure-momentum losses in the scrape-off layer (SOL) are essential for detachment to occur, experimental confirmation has been largely lacking, in large part due to diagnostic limitations. In the present assessment of L-mode discharges in the diagnostically optimized outer horizontal target configuration, the onset of pressure-momentum losses in nitrogen seeding scans in low recycling conditions is shown to occur at relatively high target electron temperature of 10 eV compared to pure deuterium pressure scans, leading to a more pronounced reduction in the target heat flux, and demonstrating for the first time the importance of momentum losses in impurity radiation dominated dissipative divertors in the absence of strong neutral-plasma interactions. Conversely, a relatively constant target heat flux response is observed above Te,t = 5 eV in pure deuterium density ramps, where the decrease in Te,t is balanced by a rise in target pressure prior to the ion flux rollover, coinciding with the pressure-momentum loss onset. While in practice the relative strength of each dissipation channel will be driven by integrated scenario requirements and core plasma performance constraints, this work demonstrates that there is no universal form of the fmom-loss and fcooling volumetric loss factors and their correlation. Moreover, the results confirm the theoretical and code predictions that both volumetric loss processes are required for detachment, while also resolving an apparent, but counter-intuitive, finding of the code results that volumetric power loss is insensitive to the presence of impurities, demonstrating that this is not the case experimentally. Further interpretive modelling is required to elucidate the dominant momentum and cooling loss channels in the observed trends.

Manufacturing and testing of flat-type divertor mockup with advanced materials

During reactor operation, the divertor must withstand unprecedented simultaneous high heat fluxes and high-energy neutron irradiation. The extremely severe service environment of the divertor imposes a huge challenge to the bonding quality of divertor joints, i.e., the joints must withstand thermal, mechanical and neutron loads, as well as cyclic mode of operation. In this paper, potassium-doped tungsten (KW) is selected as the plasma facing material (PFM), oxygen-free copper (OFC) as the interlayer, oxide dispersion strengthened copper (ODS-Cu) alloy as the heat sink material, and reduced activation ferritic/martensitic (RAFM) steel as the structural material. In this study, a vacuum brazing technology is proposed and optimized to bond Cu and ODS-Cu alloy with the silver-free brazing material CuSnTi. The most appropriate brazing parameters are a brazing temperature of 940 °C and a holding time of 15 min. High-quality bonding interfaces have been successfully obtained by vacuum brazing technology, and the average shear strength of the as-obtained KW/Cu and ODS-Cu alloy joints is ∼268 MPa. And a fabrication route for manufacturing the flat-type divertor target based on brazing technology is set. For evaluating the reliability of the fabrication technologies under the reactor relevant condition, the high heat flux test at 20 MW/m2 for the as-manufactured flat-type KW/Cu/ODS-Cu/RAFM mockup is carried out by using the Electron-beam Material testing Scenario (EMS-60) with water cooling. This paper reports the improved vacuum brazing technology to connect Cu to ODS-Cu alloy and summarizes the production route, high heat flux (HHF) test, the pre and post non-destructive examination, and the surface results of the flat-type KW/Cu/ODS-Cu/RAFM mockup after the HHF test. The test results demonstrate that the mockup manufactured according to the fabrication route still have structural and interfacial integrity under cyclic high heat loads.

Development of continuous V-shaped structure for high heat flux components of flat-type divertor

The continuous V-shaped structure with hypervapotron for divertor target can improve the cooling performance and detachment in the strike point area. The design and manufacture of this structure has been described in this work. The continuous V-shaped structure in a flat-type divertor is beneficial for divertor detachment because it has a deep slot that can trap neutral particles. The target cooled by hypervapotron is designed to sustain stationary heat flux up to ∼ 10 MW· m−2. The hypervapotron structure is directly connected at the root of the V configuration and expands the potential candidate area of the strike point for better adaptability to different plasma configurations. The simulation results demonstrate that temperature of the structure meets the allowance values of the materials because the strike point hits at the inner tip of V-shaped structure, i.e. the extreme point. The heat-sink composite panel used is made of CuCrZr and 316L stainless steel and is formed by explosive welding. The V-shaped structure is made by bending, machining, and brazing. Finally, three continuous V-shaped structure mockups were fabricated and tested to demonstrate the feasibility of engineering design.

Characterisation of detachment in the MAST-U Super-X divertor using multi-wavelength imaging of 2D atomic and molecular emission processes

In this work, we provide the first 2D spatially resolved description of radiative detachment in MAST-U Super-X L-mode divertor plasmas. The Super-X magnetic configuration was designed to achieve reduced heat- and particle loads at the divertor target compared to conventional exhaust solutions. We use filtered camera imaging to reconstruct 2D emissivity profiles in the poloidal plane for multiple atomic and molecular emission lines and bands. A set of deuterium fuelling scans is discussed that, together, span attached to deeply detached divertor states observed in MAST-U. Emissivity profiles facilitate separate analysis of locked-mode induced split branches of the scrape-off layer. Molecular deuterium Fulcher band emission front tracking reveals that the deuterium electron-impact ionisation front, for which it serves a proxy, detaches at different upstream electron densities in the split branches. Upon detachment of this ionisation front, Balmer emission attributed to molecular activated recombination appears near-target. We report a simultaneous radial broadening of the emission leg, consistent with previous SOLPS-ITER modelling. With increased fuelling this emission region detaches, implying electron temperatures below eV. In this phase, 2D Balmer line ratio reconstruction indicates an onset of volumetric direct electron-ion recombination near-target. At the highest fuelling rates this emission region moves off-target, suggesting a drop in near-wall electron density accompanying the low temperatures.

Argon–seeded detachment during ELM control by RMPs in KSTAR

In this study, we demonstrate argon-seeded discharges that exhibited a detached divertor during the full suppression and mitigation of edge-localized modes (ELMs) by an International Thermonuclear Experimental Reactor-like, three-row resonant magnetic perturbation (RMP) configuration in KSTAR. During the ELM suppression phase, the peak heat flux on the divertor target was successfully reduced from 1.6 MW m−2–0.5 MW m−2 via argon seeding. Further, the ion saturation current densities corresponding to the particle fluxes on both targets were reduced by more than 50%. During the RMP grassy-ELM regime, a further reduction to 0.1 MW m−2 in the divertor heat load was successfully achieved. A highly localized radiation zone near the X-point was also observed during divertor detachment. The calculated degree of detachment based on the two-point model increased to levels of approximately 3 and 2.3 for the outer target and inner target cases, respectively. These results provide valuable information regarding the effect of mid-Z impurities on RMP-detachment-compatible discharges.

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 ).

Heat Pipe-Based DEMO Divertor Target Concept: High Heat Flux Performance Evaluation

The use of heat pipes (HP) for the DEMO in-vessel plasma-facing components (PFCs) has been considered because of their high capacity to transport the heat from a heat source to a heat sink by means of the vaporization and condensation of the working fluid inside and their ability to enlarge the heat transfer area of the cooling circuit substantially. Recent engineering studies conducted in the framework of the EUROfusion work package Divertor (Wen et al, 2021) indicate that it is possible to design a heat pipe with a capillary limit above 6 kW using a composite capillary structure (wherein axial grooves cover the adiabatic zone and the condenser, and sintered porous material covers the evaporator). This power level would correspond to an applied heat flux of 20 MW/m2, rendering such a design interesting with respect to a divertor target concept. To validate the results of the initial engineering analysis, several experiments have been conducted to evaluate the actual performance of the proposed heat pipe concept. The present contribution presents the experiment’s results regarding the examination of the operating limits of two different designs for an evaporator: one featuring a plain porous structure, and one featuring ribs and channels.

Edge turbulence transport during ELM suppression with n = 4 resonant magnetic perturbation on EAST

The edge turbulence characteristics and the induced radial transport have been investigated in edge localized mode (ELM) suppression by using the 4 resonant magnetic perturbation coils on EAST, with 3.6 and the electron collisionality 0.5. During ELM suppression, the edge turbulence is enhanced dramatically, as measured by the reciprocating probe and the poloidal correlation reflectometry. In the near SOL, the low frequency turbulence (<30 kHz) has a large fluctuation level and propagates along the ion diamagnetic drift direction with a speed of 0.35 km s−1 in the plasma frame; an electromagnetic mode around 120 kHz with a small (∼0.15 cm−1) appears when the ELM is suppressed; weak broadband turbulence between 40–120 kHz propagates in the electron diamagnetic drift direction with a velocity of 3.4 km s−1 in the plasma frame. During the ELM suppression, the radial turbulent particle flux, calculated in both the time and frequency domains, is much higher (can be up to five times) than that in the inter-ELM phase. Furthermore, the low frequency turbulence (<30 kHz) dominates the cross-field particle transport. The 120 kHz electromagnetic mode also contributes to outward particle flux, which is relatively small. A set of CGYRO simulations are performed to illustrate the nature of the 120 kHz electromagnetic mode and the low frequency turbulence, suggesting that the former is the micro-tearing mode and the latter is the ion temperature gradient mode. The bispectral analysis suggests a strong three-wave coupling between the low frequency and high frequency turbulence (>250 kHz), which could be beneficial to form the observed turbulent transport. The estimated upstream cross-field particle flux is consistent with the total particle flux deposited on divertor targets, demonstrating that the enhanced radial turbulent particle transport is an important mechanism for particle exhaust in ELM suppression.

Refining light impurity content estimates at the lower divertor based on experimental data in WEST

WEST is an actively cooled, long-pulse tokamak with nearly all plasma-facing components (PFC) made of tungsten (W). One of the aims of WEST is to study plasma operation with W PFCs in preparation for long-pulse operation on W divertor devices like ITER (Bucalossi et al., 2014) [1]. For long-pulse operation, the W impurity content and transport to the core plasma are critical concerns that require further measurement and interpretation in order to improve plasma performance and PFC durability. In low confinement mode (L-mode) operation, the sputtering of tungsten is not expected to be driven by the plasma itself (D), but by the main light impurities (C, B, O, N) observed at trace levels in WEST. Identifying the main W sources and understanding W transport in the SOL are key metrics for plasma performance, which are simulated with Scrape-Off Layer (SOL) impurity transport codes (SOLEDGE-ERO, OEDGE-DIVIMP). Previous studies have used O as proxy for light impurity with a global content varying from 3 to 5% in SOL transport simulations (Genova et al., 2021; Rooij et al., 2020; Klepper et al., 2022) [2–4]. A power scan was performed in a standard WEST magnetic equilibrium with LH injected power up to 5 MW. Density and temperature were measured with flush-mounted Langmuir probes (Dejarnac et al., 2021) [5]. The impurity fluxes are estimated by applying the number of ionizations per photons (S/XB) coefficients calculated with the collisional-radiative model “ColRadPy” (Johnson et al., 2019) [6] to the spectral radiance measured with the WEST visible spectroscopy system (Meyer et al., 2016) [7]. This work details how the low-Z impurity content (C, O) near the divertor target was refined using the visible spectroscopy for several impurity charge states, coupled with ColRadPy to infer the content of all impurity charge states at the divertor. First results shows that the main light impurity is C and not O as originally assumed. This inferred impurity content is then used to evaluate the W sputtering at the divertor, considering the sheath at the divertor. A quantitative comparison is performed with the W flux at the divertor measured with visible spectroscopy that shows a qualitative agreement, but not fully consistent at the strike point position. This, in turn, suggests that the impurity content is not complete with O and C only.

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..

Towards fast surrogate models for interpolation of tokamak edge plasmas

One of the major design limitations for tokamak fusion reactors is the heat load that can be sustained by the materials at the divertor target. Developing a full understanding of how machine or operation parameters affect the conditions at the divertor requires an enormous number of simulations. A promising approach to circumvent this is to use machine learning models trained on simulation data as surrogate models. Once trained such surrogate models can make fast predictions for any scenario in the design parameter space. In future such simulation based surrogate models could be used in system codes for rapid design studies of future fusion power plants. This work presents the first steps towards the development of such surrogate models for plasma exhaust and the datasets required for their training. Machine learning models like neural networks usually require several thousand data points for training, but the exact amount of data required varies from case to case. Due to the long runtimes of simulations we aim at finding the minimal amount of training data required. A preliminary dataset based on SOLPS-ITER simulations with varying tokamak design parameters, including the major radius, magnetic field strength and neutral density is constructed. To be able to generate more training data within reasonable computation time the simulations in the dataset use fluid neutral simulations and no fluid drift effects. The dataset is used to train a simple neural network and Gradient Boosted Regression Trees and test how the performance depends on the number of training simulations.