Divertor Target Plates Research Articles

Numerical study of the erosion of the EAST tungsten divertor targets caused by edge localized modes

Edge localized modes (ELMs) in company with high-confinement mode (H-mode) will release high energy plasma fluxes to the scrape of layer (SOL). Large portions of these high heat fluxes will eventually irradiate the divertor target plates, and may erode, even melt them. In this paper, we develope a one-dimensional heat conductivity model including evaporation, radiation, melting processes of tungsten to study the erosion of the divertor tungsten targets caused by ELMs in EAST at the current and possible future operation parameters. Based on both experimental data of heat fluxes on the carbon-fibre composites divertor in EAST and possible future data of high heat fluxes, the surface temperature of slab-shaped tungsten is evaluated numerically by solving the one-dimensional model. It is found that the current Type I ELMs do not cause any noticeable changes of the tungsten target, the surface temperature being raised only several tens of degrees. Simulation results show that ELMs will not become a problem for EAST tungsten wall for the time being and the near future as long as much more severe transient events, e.g., disruption, can be avoided. When deposition energy is increased to 1 MJ/m2 with a duration of 600 μs, the tungsten plate will melt for a layer as thick as 6.8 μm.

HL-2M Divertor Geometry Exploration with SOLPS5.0

One of the critical issues to be solved for HL-2M is the power and particle exhaust. Divertor target plate geometry strongly influences the plasma profiles by controlling the neutral recycling pattern, which has in turn a strong effect on the symmetry and stability of the divertor plasma and finally on the whole edge region. The numerical simulation software SOLPS5.0 Package is used to design and explore the divertor target plates for HL-2M. We choose two divertor geometries, and assess the heat flux on the target plates and first wall, then further discuss the divertor plasma parameters, and how private flux baffling affects both neutral recirculation pattern and pumping efficiency.

An overview of recent HL-2A experiments

For the first time supersonic molecular beam injection (SMBI) and cluster jet injection (CJI) were applied to mitigate edge-localized modes (ELMs) in HL-2A successfully. The ELM frequency increased by a factor of 2–3 and the heat flux on the divertor target plates decreased by 50% on average after SMBI or CJI. Energetic particle induced modes were observed in different frequency ranges with high-power electron cyclotron resonance heating (ECRH). The high frequency (200–350 kHz) of the modes with a relatively small amplitude was close to the gap frequency of the toroidicity-induced Alfvén eigenmode. The coexistent multi-mode magnetic structures in the high-temperature and low-collision plasma could affect the plasma transport dramatically. Long-lived saturated ideal magnetohydrodynamic instabilities during strong neutral beam injection heating could be suppressed by high-power ECRH. The absolute rate of nonlinear energy transfer between turbulence and zonal flows was measured and the secondary mode competition between low-frequency (LF) zonal flows (ZFs) and geodesic acoustic modes (GAMs) was identified, which demonstrated that ZFs played an important role in the L–H transition. The spontaneously generated E × B shear flow was identified to be responsible for the generation of a large-scale coherent structure (LSCS), which provided unambiguous experimental evidence for the LSCS generation mechanism. New meso-scale electric potential fluctuations (MSEFs) at frequency f ∼ 10.5 kHz with two components of n = 0 and m/n = 6/2 were also identified in the edge plasmas for the first time. The MSEFs coexisted and interacted with magnetic islands of m/n = 6/2, turbulence and LF ZFs.

Equilibrium and vertical-instability considerations for vertical strike-point shifts on the ITER divertor targets

The study of operation with raised strike points on the first ITER divertor target plates is motivated by the need to gain experience with operation with strike points on tungsten (W) surfaces during the non-active phases (in the case of an initial carbon fibre composite (CFC)/W divertor); or (if ITER begins with a full-W divertor), to gain experience with plasma control and transients while operating with raised strike points to avoid damaging the baseline strike regions in preparation for the nuclear phase, and to provide a means for operation should damage occur in the baseline strike zone. For operation with raised strike points, we use the Corsica code to investigate the range of possible H- and L-mode equilibria, with emphasis on the maximum plasma current, achievable shapes, etc. With raised strike points the maximum achievable plasma current is close to 14 MA. The operating space (βp − li) for raised strike points has been studied. The size of the βp − li operating space shrinks (compared to using standard strike-point positions) at 14 MA. For 12 MA, however, the operating space is not affected when using raised strike points. For equilibria with elevated strike points (at roughly the CFC/W transitions, following the 2007 ITER Design Review) the vertical-instability growth-rates at high plasma current (14 MA) are somewhat high but are within the 20 s−1 which studies indicate are controllable in ITER. At lower currents (12 MA) in H-mode, the vertical-instability growth rates stay below 10.0 s−1 for most of βp − li space. At 12 MA in H-mode, multiple equilibria which meet our constraints have been found in overlapping regions of the βp − li operating space.

Radiative Divertor Plasma Behavior in L- and H-Mode Discharges with Argon Injection in EAST

Introducing strong radiative impurities into divertor plasmas has been considered as an important way to mitigate the peak heat load at the divertor target plate for ITER, and will be employed in EAST for high power long pulse operations. To this end, radiative divertor experiments were explored under both low (L) and high (H) - mode confinement regimes, for the first time in EAST, with the injection of argon and its mixture (25% Ar in D2). The Ar injection greatly reduced particle and heat fluxes to the divertor in L-mode discharges, achieving nearly complete detached divertor plasma regimes for both single null (SN) and double null (DN) configurations, without increasing the core impurity content. In particular, the peak heat flux was reduced by a factor of ∼6, significantly reducing the intrinsic in-out divertor asymmetry for DN, as seen by both the new infra-red camera and the Langmuir probes at the divertor target. Promising results have also been obtained in the H-modes with argon seeding, demonstrating a significant increase in the frequency and decrease in the amplitude of the edge localized modes (ELMs), thus reducing both particle and heat loads caused by the ELMs. This will be further explored in the next experimental campaign with increasing heating power for long pulse operations.

Design and cooling of the edge segments of the DEMO divertor target plates

In support of shadowing of the divertor target plate edges in toroidal direction against damage caused by the incident particles, the fingers at the boundary of the target plate should ideally form a flat surface. The reference cooling fingers are of hexagonal shape and when assembled together, their edge boundary cannot be flat. Therefore, the boundary segments need to be designed in a different way. Three possible designs are investigated: non-symmetric pentagonal fingers and two square-shaped fingers of different sizes, all cooled by the same type of concentric cartridge as in the reference design. Their heat transfer performance is analyzed from the point of view of maximum allowable temperature of the thimble structure. The computational fluid dynamics (CFD) analysis is performed to obtain the minimum mass flow rate of the coolant which is necessary to keep the structure's temperature below the permissible limit at an acceptable pressure loss.

Simulation Study for Divertor Geometry and Gas Puffing to Handle Huge Exhaust Power in HL-2M with SOLPS5.0

One of the critical issues to be solved for HL-2M is the power exhaust. Divertor target plate geometry strongly influences the plasma profiles by controlling the neutral recycling pattern, which in turn has a strong effect on the symmetry and stability of the divertor plasma and finally on the whole edge region. The numerical simulation SOLPS5.0 package is used to design and explore the divertor target plates for HL-2M. We start with the choice of a proper target plate geometry, which has a smaller incidence angle in the permissible space, and then discuss the method of gas puffing to reduce the heat flux density on the target and the effects of gas puffing on the divertor plasma performance.

Elastic–plastic analysis of the steel-to-tungsten transition joint for a high performance divertor

The use of tungsten as a plasma-facing material necessitates a transition joint to the oxide dispersion strengthened (ODS) steel or ferritic steel (FS) structural material of the primary coolant loop at the end of the divertor target plate where the surface heat flux is very low. A critical issue in the transition joints is the coefficient of thermal expansion (CTE) mismatch between the tungsten (or tungsten-alloy) and ODS steel, which can lead to unacceptably high thermal stresses during steady state and ratcheting during cyclic loads. Detailed 2D and 3D thermo-mechanical analyses were conducted to study the behavior of a transition from tungsten to FS with an intermediate layer of tantalum, located outside of the high heat flux region. The results include plastic strains under various loading conditions including fabrication processes, warm and cold shutdown, and allow for plastic behaviors leading to stress relaxation. The accumulation of plastic deformation may cause ratcheting. Modifications were proposed to the transition joint design in order to eliminate stress concentration and ratcheting under cyclic loading. The results of the modified design exhibited less plastic deformation in the joints as well as no ratcheting caused by warm and cold shutdown.

Temperature Distribution and Heat Flux on the EAST Divertor Targets in H-Mode

An infrared camera (IR) has been put into operation in the Experimental Advanced Superconducting Tokamak (EAST), which is used to measure the temperature distribution on the surface of lower divertor target plates. With a finite difference method, the heat flux onto the divertor target plates is calculated from the surface temperature profile. The high confinement mode (H-mode) with type-III edge localized modes (ELMs) has been obtained with about 1 MW lower-hybrid wave power on the EAST in the autumn experiment in 2010. The analyzed H-mode discharges were lower single null X-point diverted discharges with a density range of < ne > (1 ∼ 4) × 1019 m−3. The surface temperature of the inner target plate increases with heating power. The peak temperature on the surface of target plates is lower than 200°C with about 2.4 MW heating power. Comparison among the heat flux profiles occurring in different phases in the same discharge has been performed. It indicates that the heat flux profile obviously changes from the ohmic phase to the H-mode phase, and the full width at half maximum (FWHM) of the heat flux profile is the narrowest during the ELM-free H-phase. On the outer target plate, the peak heat flux exceeds 2 MW/m2 during the ELMy H-mode phase, whereas it is only about 0.8 MW/m2 during the ELM-free phase in the same discharge.

Modelling of radiative divertor operation towards detachment in experimental advanced superconducting tokamak

In order to actively control power load on the divertor target plates and study the effect of radiative divertor on plasma parameters in divertor plasmas and heat fluxes to the targets, dedicated experiments with Ar impurity seeding have been performed on experimental advanced superconducting tokamak in typical L-mode discharge with single null divertor configuration, ohmic heating power of 0.5 MW, and lower hybrid wave heating power of 1.0 MW. Ar is puffed into the divertor plasma at the outer target plate near the separatrix strike point with the puffing rate 1.26×1020 s−1. The radiative divertor is formed during the Ar puffing. The SOL/divertor plasma in the L-mode discharge with radiative divertor has been modelled by using SOLPS5.2 code package [V. Rozhansky et al., Nucl. Fusion 49, 025007 (2009)]. The modelling shows the cooling of the divertor plasma due to Ar seeding and is compared with the experimental measurement. The changes of peak electron temperature and heat fluxes at the targets with the shot time from the modelling results are similar to the experimental measurement before and during the Ar impurity seeding, but there is a major difference in time scales when Ar affects the plasma in between experiment and modelling.

Calorimetry of the JET ITER-Like Wall components

As part of the ILW project, new diagnostics have been installed in order to protect the plasma-facing components (PFCs). Here we present the diagnostics used to monitor the PFC temperature, thermocouples and cameras, and assess the consistency of their measurements. In dedicated limited L-mode plasmas, the surface of the limiter tiles are heated up to 900 °C. The comparison of surface temperature measurements from IR and near IR cameras, which have been calibrated against a black-body source, leads to a Be emissivity of 0.18, comparable with the theoretical one. Energy calculation derived from thermocouples, which are embedded in both limiters and divertor target plates (W-coated CFC), is compared to a 1D model based on thermal quadrupole approach (benchmarked with an ANSYS model) associated to an inversion computation. The analysis of 20 pulses shows that a good energy balance is achieved within the error bar of the model, assessed to be of 30%.

Simulation of tokamak SOL and divertor region including heat flux mitigation by gas puffing

Two-dimensional (2D), scrape-off layer (SOL)-divertor transport simulations are performed using the integrated plasma-neutral-impurity code KTRAN developed at Seoul National University. Firstly, the code is applied to reproduce a National Spherical Torus eXperiment (NSTX) discharge by using the prescribed transport coefficients and the boundary conditions obtained from the experiment. The plasma density, the heat flux on the divertor plate, and the D α emission rate profiles from the numerical simulation are found to follow experimental trends qualitatively. Secondly, predictive simulations are carried out for the baseline operation mode in Korea Superconducting Tokamak Advanced Research (KSTAR) to predict the heat flux on the divertor target plates. The stationary peak heat flux in the KSTAR baseline operation mode is expected to be 6.5 MW/m2 in the case of an orthogonal divertor. To study the mitigation of the heat flux, we investigated the puffing effects of deuterium and argon gases. The puffing position is assumed to be in front of the strike point at the outer lower divertor plate. In the simulations, mitigation of the peak heat flux at the divertor target plates is found to occur when the gas puffing rate exceeds certain values, ∼1.0 × 1020 /s and ∼5.0 × 1018 /s for deuterium and argon, respectively. Multi-charged impurity transport is also investigated for both NSTX and KSTAR SOL and divertor regions.

Particle and power deposition on divertor targets in EAST H-mode plasmas

The effects of edge-localized modes (ELMs) on divertor particle and heat fluxes were investigated for the first time in the Experimental Advanced Superconducting Tokamak (EAST). The experiments were carried out with both double null and lower single null divertor configurations, and comparisons were made between the H-mode plasmas with lower hybrid current drive (LHCD) and those with combined ion cyclotron resonance heating (ICRH). The particle and heat flux profiles between and during ELMs were obtained from Langmuir triple-probe arrays embedded in the divertor target plates. And isolated ELMs were chosen for analysis in order to reduce the uncertainty resulting from the influence of fast electrons on Langmuir triple-probe evaluation during ELMs. The power deposition obtained from Langmuir triple probes was consistent with that from the divertor infra-red camera during an ELM-free period. It was demonstrated that ELM-induced radial transport predominantly originated from the low-field side region, in good agreement with the ballooning-like transport model and experimental results of other tokamaks. ELMs significantly enhanced the divertor particle and heat fluxes, without significantly broadening the SOL width and plasma-wetted area on the divertor target in both LHCD and LHCD + ICRH H-modes, thus posing a great challenge for the next-step high-power, long-pulse operation in EAST. Increasing the divertor-wetted area was also observed to reduce the peak heat flux and particle recycling at the divertor target, hence facilitating long-pulse H-mode operation. The particle and heat flux profiles during ELMs appeared to exhibit multiple peak structures, and were analysed in terms of the behaviour of ELM filaments and the flux tubes induced by modified magnetic topology during ELMs.

Module of lithium divertor for KTM tokamak

Activity on projects of ITER and DEMO reactors has shown that solution of problems of divertor target plates and other plasma facing elements (PFEs) based on the solid plasma facing materials cause serious difficulties. Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium–metal with low Z. Application of lithium will allow to create a self-renewal and MHD stable liquid metal surface of the in-vessel devices possessing practically unlimited service life; to reduce power flux due to intensive re-irradiation on lithium atoms in plasma periphery that will essentially facilitate a problem of heat removal from PFE; to reduce Zeff of plasma to minimally possible level close to 1; to exclude tritium accumulation, that is provided with absence of dust products and an opportunity of the active control of the tritium contents in liquid lithium. Realization of these advantages is based on use of so-called lithium capillary-porous system (CPS) – new material in which liquid lithium fill a solid matrix from porous material. The progress in development of lithium technology and also activity in lithium experiments in the tokamaks TFTR, T-11M, T-10, FTU, NSTX, HT-7 and stellarator TJ II permits of solving the problems in development of steady-state operating lithium divertor module project for Kazakhstan tokamak KTM. At present the lithium divertor module for KTM tokamak is under development in the framework of ISTC project # K-1561. Initial heating up to 200°C and lithium surface temperature stabilization during plasma interaction in the range of 350–550°C will be provided by external system for thermal stabilization due to circulation of the Na–K heat transfer media. Lithium filled tungsten felt is offered as the base plasma facing material of divertor. Development, creation and experimental research of lithium divertor model for KTM will allow to solve existing problems and to fulfill the basic approaches to designing of lithium divertor and in-vessel elements of new fusion reactor generation, to investigate plasma physics aspects of lithium influence, to develop technology of work with lithium in tokamak conditions. Results of this project addresses to the progress in the field of fusion neutrons source and fusion energy source creation.

Controlled tungsten melting and droplet ejection studies in ASDEX Upgrade

Tungsten rods of 1×1×3 mm3 were exposed in single H-mode discharges at the outer divertor target plate of ASDEX Upgrade using the divertor manipulator system. Melting of the W rod at a pre-defined time was induced by moving the initially far away outer strike point close to the W-rod position. Visible light emissions of both the W pin and consecutively ejected W droplets were recorded by two fast cameras with crossed viewing cones. The time evolution of the local W source at the pin location was measured by spectroscopic observation of the WI line emission at 400.9 nm and compared to the subsequent increase of tungsten concentration in the confined plasma derived from tungsten vacuum UV line emission. Combining these measurements with the total amount of released tungsten due to the pin melt events and ejected droplets allowed us to derive an estimate of the screening factor for this type of tungsten source. The resulting values of the tungsten divertor retention in the range 10–20 agree with those found in previous studies using a W source of sublimated W(CO)6 vapour at the same exposure location. Ejected droplets were found to be always accelerated in the general direction of the plasma flow, attributed to friction forces and to rocket forces. Furthermore, the vertically inclined target plates cause the droplets, which are repelled by the target plate surface potential due to their electric charge, to move upwards against gravity due to the centrifugal force component parallel to the target plate.

Modeling of Impurity Transport in Edge Plasmas and Tritium Codeposition on Plasma Facing Walls in ITER

In order to simulate carbon deposition profile in the divertor of ITER, long-distance transport in the scrape-off-layer and divertor plasma of carbon and hydrocarbons eroded from the divertor target plates are modeled. Physically eroded carbons dominate a sharp profile on the outer target plate, whereas at the inner target plate, a very small redeposition is observed. Chemically eroded hydrocarbons produce a redeposition on the dome area as well as both inner and outer target plates. Assuming tritium content in the redeposited layers, tritium co-deposition profile on the inner and outer target plates and dome is estimated, which allows us to predict the long-term tritium retention in the divertor of ITER.

MHD Equilibrium Configuration Reconstructions for HL-2A Tokamak

The EFIT (Equilibrium Fitting) code is modified for the equilibrium configuration reconstruction in HL-2A. Signals from Langmuir probe (LP) at the divertor target plates are employed in the reconstruction of divertor configurations. The results show that discharge #2895 starts with a limiter configuration and develops gradually into a divertor configuration after t = 230 ms. This transition process is clearly demonstrated by the LP signals for the reconstruction. The profiles of plasma parameters such as safety factor q, pressure and current density as well as the evolution of major shape parameters of plasma, such as the boundary magnetic fluxes, the positions of both x-point and magnetic axis, are calculated from the reconstructed configurations. The possibility to apply the method to the swing of strike point on the target plate is discussed.

Modelling studies of SOL-divertor plasmas in EAST tokamak with ohmic shots

Modelling studies of SOL-divertor plasmas are carried out using the edge plasma code package B2.5/Eirene-SOLPS5.0. The modelling results are compared with the experimental data of ohmic shots in the EAST tokamak. The profiles of plasma temperature, density, ion flux at the divertor target plates from the modelling are compared with the Langmuir probe profiles. Dα and CIII emissivities from the modelling are also matched to the experimental measurement. The model treats plasma species of deuterium (D) and carbon (C) impurities in a real divertor geometry and an magnetohydrodynamic equilibrium from an experimental shot. Neutral particles of D0 and C0 are traced with Eirene. Properties of low recycling, high recycling and detachment in the divertor plasmas during the density ramp-up are studied at first for an ohmic shot in a discontacted double null configuration. For two ohmic shots with a single null configuration, modelling results of profiles of electron density ne, electron temperature Te, total heat flux Qt and ion saturation flux Γsat are similar to the experimental measurement. The calculated profiles at some positions on the target plates, however, differ from the measured profiles; the calculated ne and Te are one–two times lower than the measured ne and Te, and the calculated Γsat is three–four times lower than the measured Γsat. The calculated Qt profiles along the outer target plate are in better agreement with the measured profiles of Qt, but the peak values of the Qt profiles from the modelling differ from the experimental measurement, which may result from the low spatial resolution of the experimental measurement and the difficulties of measuring peak values with narrow widths.

Optimization of isotope composition of fusion plasmas

The radial profiles of deuteron and triton densities in a fusion tokamak reactor are computed by taking into account anomalous transport due to ion temperature gradient and trapped electron drift instabilities in the plasma core and neoclassical transport in plateau-banana regimes, residual anomalous diffusion and convection triggered by edge localized modes in the edge transport barrier. The charged particle sources due to influx of neutral particles through the separatrix and ablation of frozen pellets in the plasma interior are optimized by searching for conditions providing the minimal outflow of unburned tritons to divertor target plates and their exhaust by pumping systems. It is demonstrated that such conditions correspond to fuelling of deuterium plasma component with gas puffing and of tritium with pellets and a ratio of deuteron/triton densities close to 2/1 at the separatrix. Effects of the absolute level and the ion mass dependence of neoclassical and anomalous transport contributions, of the position where neutrals enter the confined volume through the separatrix on the radial profiles of the ion densities are investigated.

Scaling of the power exhaust channel in Alcator C-Mod

Parametric dependences of the heat flux footprint on the outer divertor target plate are explored in EDA H-mode and ohmic L-mode plasmas over a wide range of parameters with attached plasma conditions. Heat flux profile shapes are found to be independent of toroidal field strength, independent of power flow along magnetic field lines and insensitive to x-point topology (single-null versus double-null). The magnitudes and widths closely follow that of the “upstream” pressure profile, which are correlated to plasma thermal energy content and plasma current. Heat flux decay lengths near the strike-point in H- and L-mode plasmas scale approximately with the inverse of plasma current, with a diminished dependence at high collisionality in L-mode. Consistent with previous studies, pressure gradients in the boundary scale with plasma current squared, holding the magnetohydrodynamic ballooning parameter approximately invariant at fixed collisionality—strong evidence that critical-gradient transport physics plays a key role in setting the power exhaust channel.