Divertor Target Plates Research Articles

Detachment plasma achieved based on active temperature feedback system in EAST

The infrared (IR) thermography diagnostic has recently been upgraded to achieve real-time temperature measurement and feedback capability in the experimental advanced superconducting tokamak (EAST). A new control and data acquisition program for IR camera is developed with online temperature calibration and compensation. Reflective memory (RFM) is configured in the data acquisition system for real-time data transmission to the plasma control system (PCS). Based on the upgraded IR thermography diagnostic, detachment plasma has been achieved for the first time in EAST with active temperature feedback. The divertor impurity seeding system is used to inject a sequence of short neon impurity pulses with 50% D2 to increase the edge radiation of the plasma particles. The temperature of the divertor target plates decreases to the preset target value and the divertor detachment has been realized with no serious reduction of plasma stored energy and confinement observed in the temperature feedback control phase. As a new feedback control method, the IR temperature feedback control shows great application prospects for the temperature and heat flux control of the first wall components in future long pulse operation of EAST.

First Evidence of Local E×B Drift in the Divertor Influencing the Structure and Stability of Confined Plasma near the Edge of Fusion Devices.

The structure of the edge plasma in a magnetic confinement system has a strong impact on the overall plasma performance. We uncover for the first time a magnetic-field-direction dependent density shelf, i.e., local flattening of the density radial profile near the magnetic separatrix, in high confinement plasmas with low edge collisionality in the DIII-D tokamak. The density shelf is correlated with a doubly peaked density profile near the divertor target plate, which tends to occur for operation with the ion B×∇B drift direction away from the X-point, as currently employed for DIII-D advanced tokamak scenarios. This double-peaked divertor plasma profile is connected via the E×B drifts, arising from a strong radial electric field induced by the radial electron temperature gradient near the divertor target. The drifts lead to the reversal of the poloidal flow above the divertor target, resulting in the formation of the density shelf. The edge density shelf can be further enhanced at higher heating power, preventing large, periodic bursts of the plasma, i.e., edge-localized modes, in the edge region, consistent with ideal magnetohydrodynamics calculations.

Time-dependent modeling of coupled plasma-wall dynamics

Effects of dynamic wall outgassing on divertor plasma during an edge localized mode (ELM)-like heat pulse are modeled using the newly coupled edge plasma code UEDGE and wall reaction-diffusion code FACE. Different divertor regimes are simulated depending on the degree of the plasma detachment. It is shown that hydrogen outgassing from the divertor target plate induced by ELM pulse in the semi-detached regime can trigger deep plasma detachment. The induced detachment does not spontaneously reverse after the pulse due to ineffective implantation of low energy plasma hydrogen in the target material. We also show that plasma neutrals play an essential role in heat transport in a detached divertor. The modeled hydrogen dynamics in the wall material demonstrates that the heat pulse leads to enhanced hydrogen transport toward the wall bulk.

Disconnection of scrape off layer turbulence between the outer midplane and divertor target plate in NSTX

The connection between scrape off layer (SOL) turbulence at the outer midplane and at the divertor target plate is characterized in the National Spherical Torus Experiment (NSTX). Divertor fluctuations due to midplane blobs are studied by imaging their helical intersection with the divertor plate. Divertor fluctuation levels gradually decrease approaching the outer strike point. Divertor radial turbulence scale lengths as well as radial and poloidal turbulence propagation velocities are comparable to those on the low field side midplane. Strong correlation between divertor and midplane turbulence is observed in the far SOL. The correlation progressively decreases towards the separatrix, along with the decrease in the measured divertor fluctuations. The reduction in correlation is consistent with an electrostatic two-region blob model (Myra et al 2006 Phys. Plasmas 13 112502) with perpendicular ion polarization currents due to X-point geometry causing filament disconnection.

Preliminary Design of ITER Divertor Langmuir Probe System

The primary aim of the ITER divertor Langmuir probe system is to measure the plasma parameters at the divertor target plates. Saturation ion flux coming from the direct-current biased probe mode is used for advanced machine control, and the swept double-probe mode is recommended to measure electron temperature and density for physics studies. The design of the probe system includes three parts. First, tungsten Langmuir probes are mounted on the side of the target plates for collecting current from plasma, and thermomechanical simulation results show the design of the probe is robust and can survive under harsh working environments. Second, the electronics consists of the power supply, mode switching, and signal conditioning box and is used for driving Langmuir probes in different operation modes to obtain expected plasma information. Third, the functions of instrument and control include publishing configuration; monitoring and control; calibration; data acquisition; communication with the control, data access and communication (CODAC) system; and real-time ion flux measurement at the divertor target. The system design also complies with ITER’s technical practices, standards, and codes.

Neural network regression approaches to reconstruct properties of magnetic configuration from Wendelstein 7-X modeled heat load patterns

Convective heat loads onto the plasma facing components of magnetic confinement devices contain information about edge magnetic field properties which are not yet fully exploited. Machine learning approaches are a promising technique to automatically extract information about such properties from heat load images. In this study, we present the successful reconstruction of proxies for two independent, important edge magnetic field properties given simulated heat load images on the Wendelstein 7-X divertor target plates. Six different artificial neural network architectures from shallow and simple feed-forward fully-connected neural network to deep Inception ResNets with 24 223 to 804 804 free parameters are investigated. The relative reconstruction error is between and with calculation times on the order of milliseconds. A competing benchmark method without machine learning reaches slightly smaller errors but exceeds the calculation time by three orders of magnitude. The experiments demonstrate that machine learning is also a powerful tool in this particular field of nuclear fusion research and deep convolutional neural networks are identified as favorable algorithms for the stated problem. The findings of this paper build a basis for future real time discharge optimization and control by means of machine learning methods.

Modeling the effect of divertor closure on plasma detachment for new divertor design of EAST by SOLPS

One of the major challenges for the design and operation of next-step high-power steady-state fusion devices is to develop and validate advanced divertor solutions for handling power exhaust, while maintaining acceptable divertor target plate erosion. This requires to access divertor detachment at relatively low main plasma densities compatible with current drive and high plasma confinement. Recently a new initiative has been launched on the EAST superconducting tokamak to develop a new divertor for evaluating boundary plasma solutions applicable to the next step fusion experiments beyond ITER. In the present work, four different divertor configurations with different degrees of closure are evaluated through detailed modeling with the SOLPS plasma boundary code. The modeling results show that increasing divertor closure can significantly trap more neutrals from to with the same upstream separatrix density and hence facilitate the onset of detachment decreasing from to Moreover, with increasing divertor closure the divertor radiated power is also increased from 200 to 450 kW and the peak heat flux density at the divertor target is reduced from 9.3 to 3.8 MW m−2. Therefore, increasing divertor closure can benefit divertor operation for EAST and may be used as one of the scientific metrics for the divertor design of future fusion devices.

Development of a 1D thermofluid code for divertor target plate modeling

The Oxford Divertor Thermo-Fluid Code (OxDTFC) is a numerical model to assess and compare the steady state thermo-fluid performance of the cooling systems present in the divertor target plate region of a fusion tokamak. Two water-cooled divertor target plate concepts have been used to demonstrate the model: an ITER-like monoblock system and the High Pressure Jet Cascade (HPJC) modular concept, developed by the University of Oxford. Validation of the model was carried out by comparing results for the ITER-like monoblock system with experimental and numerical data in the literature. This comparison showed that the model could quickly and accurately replicate divertor target plate thermo-fluid results across various coolant flow rates, incident heat fluxes, and geometric parameters. OxDTFC is used to compare the performance of the ITER-like monoblock and HPJC systems under equivalent target plate power loading. In this analysis, for equivalent peak heat sink temperatures and thermal loading conditions, the HPJC cooling system was found to have pumping power requirements of ∼65% and coolant flow rates of ∼50% less than a comparable ITER-like monoblock system.

DIII-D research towards establishing the scientific basis for future fusion reactors

DIII-D research is addressing critical challenges in preparation for ITER and the next generation of fusion devices through focusing on plasma physics fundamentals that underpin key fusion goals, understanding the interaction of disparate core and boundary plasma physics, and developing integrated scenarios for achieving high performance fusion regimes. Fundamental investigations into fusion energy science find that anomalous dissipation of runaway electrons (RE) that arise following a disruption is likely due to interactions with RE-driven kinetic instabilities, some of which have been directly observed, opening a new avenue for RE energy dissipation using naturally excited waves. Dimensionless parameter scaling of intrinsic rotation and gyrokinetic simulations give a predicted ITER rotation profile with significant turbulence stabilization. Coherence imaging spectroscopy confirms near sonic flow throughout the divertor towards the target, which may account for the convection-dominated parallel heat flux. Core-boundary integration studies show that the small angle slot divertor achieves detachment at lower density and extends plasma cooling across the divertor target plate, which is essential for controlling heat flux and erosion. The Super H-mode regime has been extended to high plasma current (2.0 MA) and density to achieve very high pedestal pressures (~30 kPa) and stored energy (3.2 MJ) with H98y2 ≈ 1.6–2.4. In scenario work, the ITER baseline Q = 10 scenario with zero injected torque is found to have a fusion gain metric independent of current between q95 = 2.8–3.7, and a lower limit of pedestal rotation for RMP ELM suppression has been found. In the wide pedestal QH-mode regime that exhibits improved performance and no ELMs, the start-up counter torque has been eliminated so that the entire discharge uses ≈0 injected torque and the operating space is more ITER-relevant. Finally, the high- (⩽3.8) hybrid scenario has been extended to the high-density levels necessary for radiating divertor operation, achieving ~40% divertor heat flux reduction using either argon or neon with Ptot up to 15 MW.

Comparison of measurements with different types of divertor Langmuir probe in EAST tokamak

Divertor Langmuir probes (Div-LPs), embedded in the divertor target plates, are the important diagnostic tools to measure the plasma parameters in EAST divertor region. A series of measurement methods, i.e., single-probe, double-probe and triple-probe schemes, are applied and compared for the Div-LPs on EAST, to get the electron temperature Te and the electron density ne at the same poloidal position, respectively. It is found that the electron temperature Te and the electron density ne measured by the single-probe and the double-probe schemes are basically the same during the plasma discharge phase. Further analysis of the electron energy distribution function (EEDF) represented by the first derivative curve method shows that the electrons conform to a Maxwell distribution, which indicate that the three methods can be applied in the EAST divertor region. Besides, the domed and flush-mounted probe tips for measuring the plasma parameters are also compared, and the results show a good match. The sheath potential coefficient Λ is also calculated from the I-V characteristic curves by two ways, i.e. Λp=(Vplasma-Vf)/Te and ΛI= ln (|Ies/Iis|), and compared with the traditional value Λ=2.8, the results indicate that the ∼1.5ΛI and Λ=2.8 are available to replace the Λp under the different strength of toroidal magnetic field in EAST tokamak.

Calculation of the heat flux in the lower divertor target plate using an infrared camera diagnostic system on the experimental advanced superconducting tokamak

During the discharging of Tokamak devices, interactions between the core plasma and plasma-facing components (PFCs) may cause exorbitant heat deposition in the latter. This poses a grave threat to the lifetimes of PFCs materials. An infrared (IR) diagnostic system consisting of an IR camera and an endoscope was installed on an Experimental Advanced Superconducting Tokamak (EAST) to monitor the surface temperature of the lower divertor target plate (LDTP) and to calculate the corresponding heat flux based on its surface temperature and physical structure, via the finite element method. First, the temperature obtained by the IR camera was calibrated against the temperature measured by the built-in thermocouple of EAST under baking conditions to determine the true temperature of the LDTP. Next, based on the finite element method, a target plate model was built and a discretization of the modeling domain was carried out. Then, a heat conduction equation and boundary conditions were determined. Finally, the heat flux was calculated. The new numerical tool provided results similar to those for DFLUX; this is important for future work on related physical processes and heat flux control.

EDGE2D-EIRENE predictions of molecular emission in DIII-D high-recycling divertor plasmas

Abstract The contributions of deuterium molecular emission to the total deuterium radiation was assessed in DIII-D ohmically-confined plasmas in high-recycling divertor conditions. Radial profiles of the deuterium Ly-α line intensity across the low-field side divertor leg were obtained with the recently installed divertor Survey Poor Resolution, Extended Spectrometer [1]. A high-resolution spectrometer was used to measure the poloidal profiles of the deuterium Balmer-α and the deuterium Fulcher-α band intensity in the visible wavelength range. The scrape-off layer plasma and neutral distributions were simulated using the edge fluid EDGE2D-EIRENE [2], and the numerical solutions constrained utilizing Thomson scattering and Langmuir probe measurements at the low-field side midplane and the divertor target plate. The studies show that for these conditions molecular emission plays a negligible role in the total radiative power balance of the low-field side divertor, but molecular processes are important when evaluating deuterium Balmer-α line intensity for code-experiment validation.

Heat flux distribution with lower hybrid current drive in the experimental advanced superconducting tokamak

Lower hybrid current drive (LHCD) considerably affects the heat flux distribution. This study analyzed the lower divertor target plate (LDTP) with different LHCD powers and frequencies in the experimental advanced superconducting tokamak (EAST). The analysis provided a detailed explanation of the heat flux at specific times (i.e., transformation of magnetic field configuration, LHCD starting and operating periods) and specific target locations (i.e., original strike point and second peak heat flux area). The average heat flux, average peak heat flux, and maximum peak heat flux during operation of different LHCD powers and frequencies were determined and compared. The sizes of the heat fluxes at the specific target locations were compared at different LHCD powers and frequencies. In addition to the heat flux distribution under different LHCD powers and frequencies, the heat flux distributions in electron cyclotron resonance heating (ECRH) + LHCD and in LHCD only were investigated to study the effect of ECRH on heat flux distribution. Detailed analyses of the heat flux distribution under different conditions were conducted to provide a reference for actual engineering applications.

Multiphysics analysis of W7-X control coils

The quasi-symmetric fivefold modular Wendelstein 7-X (W7-X) stellarator consists of three groups of coil systems, i.e. superconducting magnet, trim coil and control coil systems. The control coil system contains ten identical 3D shaped control coils (CC) situated behind the baffle plates of corresponding divertor units, and is designated to rectify the error field and to sweep hot spots on the divertor target plates. The CC is wound from copper conductor with a square cross section of 16 mm × 16 mm and a water cooling hollow of Ø 8 mm. The control coil system was installed in W7-X in 2015, and the integral commissioning has been done in parallel with the completion of W7-X. During the operation phase (OP 1.2a) with limited plasma heating energy, a leakage in one of the CC cooling water plug-in was found and dictates a detailed transient thermal analysis of CC to determine the allowable operation time without cooling water flow. The paper presents the transient thermal analysis and is followed by a detailed finite element mechanical analysis with the consideration of temperature gradient loads, dead weight and electromagnetic forces. Moreover, the transient thermal and mechanical performance of actively cooled CC to be intensively operated during steady state operation phase (OP 2) are also analyzed and evaluated with the same FE model.

Surface temperature measurement of EAST divertor between 20 and 200 °C using the emissivity model prior

During the discharge of the Experimental Advanced Superconducting Tokamak (EAST), the divertor target plate (DTP) accumulates a large amount of heat and suffers irreversible damages, so building of a temperature measurement system is crucial for guaranteeing the steady running of EAST. In this work, a radiation thermometer system which mainly consist of a high-speed infrared camera and an endoscope is employed to get surface temperature of upper divertor target plate (UDTP) between 20 and 200 °C. The spectral response function, the geometric factor and the radiation coming from the inner parts of the measurement system are obtained using a reference blackbody placed inside EAST based on multi-temperature calibration method. Meanwhile, in order to reduce temperature measurement error aroused by the emissivity of the divertor material, a nonlinear model is built between 20 and 200 °C in this paper to describe the dependence of spectral band emissivity on temperature. Finally the surface temperature of UDTP is obtained by solving the temperature equation using this emissivity model prior. Based on this method, the surface temperature of UDTP between 20 and 200 °C is measured and it can follow the changes of heating power in EAST real time. Compared with the traditional method in which the emissivity of the divertor material is deemed to a determinate constant, this method can efficiently improve the precision of temperature measurement.

Effects of the gas puffing neutral on the plasma parameters in the end-cell of GAMMA 10/PDX by using the multi-fluid code “LINDA”

This research numerically investigates the behavior of plasma flow in GAMMA 10/PDX. The impact of impurity gases seeding into the divertor region on the plasma parameters and the energy loss processes has been investigated numerically by using the multi-fluid code “LINDA”. The neutral particles are injected into the bulk hydrogen plasma and the upstream plasma parameters on the z-axis are ni=1.0×1019m−3, Ti=100eV,Te=30eV. A reduction in the plasma temperature has been observed with the increasing H neutral injection. The ionization and Charge-exchange (CX) power loss increase almost linearly with the increasing H neutral density, which induce a reduction in the plasma temperature. During H seeding, Ar and Ne are seeded simultaneously to observe the influence of Ar and Ne on the electron and ion energy loss mechanism. In this study, the fraction of impurity density to upstream plasma density (nz/ni) is used to be 10%. A comparison between Ar and Ne shows that Ar is effective on the electron cooling. The radiation power loss of Ar is found to be higher than that of Ne. The radiation cooling effect and CX loss significantly reduces the electron and ion temperature, respectively. A slight influence of Ar and Ne seeding on the ion energy loss terms has been observed. The electron energy loss terms increase significantly during impurity gas Ar and Ne seeding. These outcomes represent the impact of gas seeding for reducing the plasma energy on the divertor target plates.

Chemically assisted physical sputtering of Tungsten: Identification via the [formula omitted] transition of WD in TEXTOR and ASDEX Upgrade plasmas

Tungsten (W) is a preferred plasma-facing material for areas of high particle and power impact in present day and future fusion devices. The lifetime of W plasma-facing components (PFCs) under steady-state conditions is limited by erosion processes induced by energetic impinging particles such as impurities (carbon C, oxygen O, nitrogen N, etc.) and hydrogen isotopes above the threshold energy for Physical Sputtering (PS). We discovered for the first time, in addition to the bare physical sputtering process of W, a second W erosion mechanism at PFCs in TEXTOR (limiter surfaces) and ASDEX Upgrade (divertor target plates) during deuterium plasma bombardment. The tungsten deuteride molecule WD has been identified spectroscopically via the 6Π→6Σ+ transition in parallel to ordinary neutral W line emission of the WI transition (5d5(6S)6s 7S3 → 5d5(6S)6p 7P4) at λ=400.9nm used as measure to quantify the bare PS process or, more precisely, the gross erosion of W. We identified the underlying process for the molecular release as Chemically Assisted Physical Sputtering (CAPS) as observed also in the case of beryllium or lanthanum. Measurements in TEXTOR and ASDEX Upgrade showed a dependence of the WD band emission on the surface temperature - connected to the deuterium content in the near W surface - as well as on the flux and energy of impinging energetic particles. A quantification of the released WD is not yet possible due to lack of appropriate molecular data for the conversion of photons into particles, but the spatially resolved ratio of atomic (WI) and molecular photon flux (WD) indicates that the release takes place at the same time, but differently distributed along the target plates and limiters reflecting the variation in the described critical parameters (impinging ion flux, ion impact energy, and material temperature) along the surface. The plasma conditions in front of the interaction zone are in both cases fully ionising with an electron temperature Te > 10eV, thus, recombination into WD can be excluded.

Managing leading edges during assembly of the Wendelstein 7-X divertor

Divertor target plates of magnetic confinement experiments are usually intersected by the magnetic field under a shallow angle. If steps are present in the target surface, the resulting leading edges will receive a strongly increased heat load as compared with the surrounding target surface. A very precise alignment of the target components is required to limit the height of leading edges in all magnetic configurations. In preparation of the initial divertor operation phase of the Wendelstein 7-X stellarator, we combined a model calculating the thermal load pattern on the targets for each magnetic configuration with a time-dependent model for heat transport within the targets, with a model for carbon sublimation and with measurements of step heights between target components to estimate the carbon sublimation rates to be expected. In a few locations of the as-built and as-installed divertor, these were found to be critically high, such that a limitation of plasma performance might have resulted. By re-aligning one type of target component according to the results of our modelling, the expected carbon sublimation rates were strongly reduced to uncritical levels.

Deuterium retention and erosion in liquid Sn samples exposed to D2 and Ar plasmas in GyM device

The use of tin (Sn) as a liquid metal for plasma facing components has been recently proposed as a solution to the high heat load issue on the divertor target plates in nuclear fusion reactors. Due to its low vapor pressure, low reactivity with hydrogen and good resilience to neutron impact, tin is a good candidate as plasma facing component. However its high atomic number poses concerns about plasma contamination.In this paper two fundamental aspects have been investigated: deuterium retention and erosion fluxes from the Sn surface towards the plasma. The samples were exposed to plasma inside the linear machine GyM in magnetic cusp configuration. This setup permits to expose free liquid specimens without the need for the Capillary Porous System. Moreover it permits to lower the magnetic field in order to increase Sn Larmor radius and consequently limit Sn re-deposition in erosion experiments.Ex-situ analyses by ion beam diagnostics on solid samples exposed to deuterium plasma have proved that the amount of retained atomic deuterium is very low, approximately 0.18 at% estimated by Nuclear Reaction Analysis and 0.25 at% estimated by Elastic Recoil Detection Analysis.In the framework of erosion studies, the spectroscopic parameter S/XB was evaluated in Ar plasma for the SnI line at 380.1 nm by Optical Emission Spectroscopy and mass loss measurements in the 5–11 eV Te range, at a density ne ∼ 1.5 × 1011 cm−3. An average value of 150 ± 23 was obtained.

Divertor leg filaments in NSTX-U

This paper presents the first observation of divertor localized turbulence in the NSTX-Upgrade spherical tokamak. Previous work in NSTX discharges (Maqueda et al 2010 Nucl. Fusion 50 075002) described fluctuations on the divertor target due to blobs generated on the low field side midplane. In the region that shows disconnection from upstream turbulence (private flux region and proximity of the strike point), divertor-localized fluctuations are observed via imaging of C III and D-α emission in diverted L-mode discharges and are characterized in this paper. Field-aligned filaments connected to the divertor target plate are radially localized at the separatrix on the outer divertor leg and in the private flux region on the inner divertor leg. These are limited to the region below the X-point with fluctuation levels up to 10%–20%. The filaments have comparable poloidal and radial correlation lengths (10–100 ion gyroradii ) and parallel correlation lengths of several meters. Toroidal mode numbers are in the range of 10–30 and 2–10 for outer and inner leg filaments. The poloidal motion of outer leg filaments has the same direction and comparable magnitude to advection due to drift calculated by the multi-fluid edge transport code UEDGE with inclusion of cross field drifts. Disconnection between inner and outer leg filaments as well as the absence of correlation with upstream turbulence support the hypothesis of X-point disconnection for divertor leg filaments. Simulations with the ArbiTER linear eigenvalue solver and a resistive-balooning model were performed with a simulation grid limited to the divertor legs. Instabilities localized to the bad curvature side of the divertor legs were observed, in qualitative agreement with the experiment.