Outer Divertor Target Research Articles

Dependence of divertor asymmetries on the toroidal magnetic field

A study of the effect of toroidal magnetic field (Bt) on the divertor asymmetries is carried out with a plasma transport code under BOUT++ framework with the magnetic equilibria from EAST and C-Mod. For the simulation cases with drifts, the density is larger at the inner divertor target, while the temperature and heat flux are larger at the outer divertor target. The ratio of the total particle flux at the outer target to that of the inner target increases with increasing Bt, while the ratio of the total heat flux at the outer target to that of the inner target decreases with increasing Bt. The in–out divertor asymmetries of both total particle and heat fluxes get weaker with increasing Bt. Further analysis shows that the edge radial transport induced by drifts is much weaker for the simulation case with higher Bt, indicating that drift-driven divertor asymmetries may be less important for future tokamaks with high Bt.

The effect of CD4 injection on WD molecule sputtering at the divertor target in EAST

Tungsten (W) is one of the preferred plasma-facing materials for future fusion reactors because of the advantages of high melting point, low physical sputtering rate and low fuel retention. The lifetime and service performance of W materials are limited by erosion processes. In recent years, tungsten deuteride molecular (WD) sputtering has been discovered in tokamak. Studying WD sputtering helps to better understand the erosion of tungsten materials in tokamak. EAST adopts the method of injecting impurity gas in the divertor to alleviate the particle flux and heat flux bombarding the tungsten divertor, and CD4 is one of the impurities used. CD4 injection experiments on EAST show that after CD4 is injected from the upper outer (UO) divertor target, despite that the electron temperature (Te) around the strike point on the UO divertor target drops to about 8 eV, strong WD sputtering occurs in the far scrape-off layer (SOL) region. The divertor probe data shows that after CD4 injection, the particle flux from the upstream plasma onto the divertor split into two bands, creating a new particle flux strike point in the far SOL region. The energy of incident particles at the original strike point is significantly reduced, while the energy of incident particles at the far SOL region is enhanced. The edge electron density profile at mid-plane indicates that CD4 injection causes an increase of density in the SOL, which could improve the coupling of Lower-Hybrid Wave (LHW) with the edge plasma as evidenced by the decreasing reflected power of LHW. The increase in absorbed power of LHW could change the particle and energy transport in the edge plasma, thus plays an important role in the enhanced WD sputtering in the far SOL region after CD4 injection.

Evaluation of local erosion and deposition on the W-monoblock of JA-DEMO divertor

We developed and performed a 3D Monte-Carlo simulation on local erosion and deposition in outer divertor target of JA-DEMO. The local erosion and deposition were evaluated considering the divertor plasma parameters and the roof shape of W-monoblocks. First, for local erosion simulation, the trajectories of D ions and impurity (He and Ar) ions were traced before irradiating the monoblock surface. The physical sputtering was calculated by these ions irradiation, considering their charge states, incident-energies and -angles. The highest sputtering flux was produced by multiply-charged Ar ions (Ar3+–Ar5+). The sputtering flux strongly depended on the poloidal positions along the divertor target. Although almost no erosion (sputtering) was obtained in low-temperature zone (∼1 eV), the erosion occurred in mid- (∼10 eV) and high-temperature (>10 eV) zone. Second, for local deposition simulation, the trajectories of W atoms emitted from the monoblock surface were traced. After transport (including ionization and recombination) in the plasma, W atoms and ions were deposited on the monoblock surface. The deposition occurred even on the area where plasmas did not irradiate (shadow). This was due to perpendicular diffusion to magnetic field lines, gyro-orbit effects for W ions and straight-line motion for neutral W atoms. The poloidal distribution of the deposition fluxes was similar to that of the sputtering fluxes. In low-temperature zone, no deposition flux occurred due to very low sputtering flux. The deposition flux in mid-temperature zone increased significantly and that in high-temperature zone remained high.

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

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

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.

Survey of tungsten gross erosion from main plasma facing components in WEST during a L-mode high fluence campaign

An initial high fluence campaign was performed in WEST, in 2023, on the newly installed actively cooled tungsten divertor composed of ITER-grade monoblocks. The campaign consisted in the repetition of a 60 s long Deuterium L-mode pulse in attached divertor conditions, cumulating over 10000s of plasma exposure. A maximum deuterium fluence of approximately 5⋅1026m−2 was reached in the outer strike point region, representative of a few high performance ITER pulses. Gross tungsten erosion inferred from visible spectroscopy shows that the most eroded plasma facing component is the inner divertor target with rates ten times larger than on the outer divertor target. The outer midplane tungsten bumpers, located a few centimeters from the plasma, show gross erosion rates two times lower than at the outer divertor. We conclude that the outer midplane bumpers have a negligible contribution to the long range tungsten migration and deposition onto the lower divertor. The cumulated gross erosion rate on the inner divertor translates in an effective gross erosion thickness of about 20μm, while it is about 2μm for the outer divertor. Strikingly, these orderings coincide with the thickness of deposits found locally on the divertor : the exposed surfaces of high field side monoblocks are covered with several tens of μm tungsten deposits, while on the lower field side, few μm thin tungsten deposits are only found on the magnetically shadowed parts of monoblocks. The strong impact of those deposits on WEST operation, namely perturbation of surface temperature measurement with infra-red thermography, and the emission of flakes causing radiative perturbation of the confined plasma, calls for anticipating similar issues in ITER. In particular, the start of research operation shall consider the definition of a divertor erosion budget in order to anticipate the formation of deleterious deposits.

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.

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

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

Comparison of partial and deep energy detachment behaviors with Ar seeding on EAST new corner slot divertor

It is necessary for future fusion reactor to reduce the heat fluxes on the entire divertor target, especially if view of long pulse high performance operation. In recent EAST experiments, partial energy detachment without confinement degradation, and deep energy detachment with protection of the entire divertor target have both been confirmed on EAST corner slot divertor by argon (Ar) seeding, which can provide reference for the divertor protection on future fusion reactors. In the deep energy detachment state, the electron temperature T et along entire lower outer divertor target decreases to less than 10 eV and heat fluxes are also strongly mitigated with peak heat flux reduction of more than 90%. Compared to the attached state, there is a moderate confinement degradation with H 98,y2 from ∼1 to ∼0.9 because of Ar radiation in the core region. This confinement degradation can be avoided in the partial energy detachment state, where the radiative power losses in the core are reduced. The experiment and SOLPS-ITER simulation results show that there is no decrease of particle flux js on the divertor target in the partial energy detachment state because the momentum loss in the SOL region is not strong enough. With increasing Ar seeding, there is a js decrease in the deep energy detachment state. The increases of momentum and power losses in the SOL region, and the decrease of upstream pressure all contribute to the js reduction.

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.

Erosion estimates for the divertor and main wall components from STEP

The tungsten erosion within Spherical Tokamak for Energy Production (STEP) assuming tungsten main wall and tungsten divertor has been estimated with ERO at the inner and outer divertor, at the inner and outer midplane and at the outboard baffle entrance. Plasma parameters are based on SOLPS simulations applying argon puffing for edge cooling. The plasma parameter range covers peak electron temperatures T e between 3 and 25 eV in the divertor. At the inner midplane T e ∼ 13 eV, at the outer midplane ∼7 eV and at the outboard baffle entrance between 1 eV and 4 eV. The modelled peak gross erosion is highest in the divertor with up to 1E19 W m−2 s−1 within the inner and 7E19 W m−2 s−1 in the outer one for the plasma parameter range studied. At the main wall the gross erosion is about 2E18 W m−2 s−1 at the inner midplane and 1.3E17 W m−2 s−1 at the outer one. However, tungsten deposition within the divertor is much larger with amounts between 88% and 98% and only between 10% and 60% at the midplane. At all locations studied, tungsten erosion due to deuterium ions is negligibly small compared to the erosion by argon ions. Erosion due to deuterium atoms has been studied for the outer midplane and is there at least four times smaller than the erosion due to argon ions. The simulations have been performed considering singly ionised Ar. However, according to the SOLPS runs the mean charge of Ar impinging the surfaces is about two at the locations of largest erosion, which leads to an increase of the gross erosion by a factor between 1.5 and 5 with the largest increase occurring at the outer divertor target.

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.

Evaluation of ITER divertor shunts as a synthetic diagnostic for detachment control

Reliable diagnostics that measure the detached state of the ITER divertor plasma will be necessary to control heat flux to the divertor targets during steady state, burning plasma operation. This paper conducts an initial exploration into the feasibility of the divertor shunt diagnostic as a lightweight, robust, and real-time detachment sensor. This diagnostic is a set of shunt lead pairs that measure the voltage drop along the divertor cassette body, from which the plasma scrape-off layer (SOL) current is calculated. Using SOLPS-ITER simulations for control-relevant ITER plasma scenarios, the thermoelectric current magnitude along the SOL is shown to decrease significantly with the onset of partial detachment at the outer divertor target. Electromagnetic modelling of a simplified divertor cassette is used to develop a control-oriented inductance-resistance circuit model, from which SOL currents can be calculated from shunt pair voltage measurements. The sensitivity and frequency-response of the resulting system indicates that the diagnostic will accurately measure SOL thermoelectric currents during ITER operation. These currents will be a good measure of the detached state of the divertor plasma, making the divertor shunt diagnostic a potentially extremely valuable and physically robust sensor for real-time detachment control.

Modeling study of divertor particle flow pattern and in–out plasma density asymmetry due to drifts with SOLPS and BOUT++

A study of the effects of drifts on the particle flow pattern and in–out divertor plasma density asymmetry for L-mode and H-mode plasmas is carried out for EAST discharges by the edge plasma transport codes SOLPS and BOUT++ . The simulation of L-mode plasmas is done by SOLPS while the simulation of H-mode plasmas is done by BOUT++ . The toroidal magnetic field direction for the simulated discharge is artificially reversed in the codes to study the effects of different drift directions on the divertor particle flow pattern and the in–out asymmetry of divertor plasma density. The divertor particle flows induced by diamagnetic and E × B drifts are found to have similar directions in the divertor region for the same discharge. The directions of the flows induced by drifts would be reversed with the reverse of toroidal magnetic field direction. The diamagnetic drift seems to have no effect on the in–out asymmetry of divertor plasma density due to its divergence-free nature. However, the E × B drift could result in a pronounced asymmetry of plasma density between the inner and outer divertor targets. The density in–out asymmetry caused by E × B drift is reversed with the reverse of E × B drift flow direction. Detailed analysis shows that the radial component of the E × B drift flow is the main cause of density asymmetry. The results from the simulation of H-mode plasmas with BOUT++ are similar to those of the L-mode plasmas with SOLPS except that the drift effects seem to be slightly larger in the H-mode plasmas compared to the L-mode plasmas.

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.

Parameter dependencies of the separatrix density in low triangularity L-mode and H-mode JET-ILW plasmas

The midplane electron separatrix density, n e,sep, in JET-ILW L-mode and H-mode low triangularity deuterium fuelled plasmas exhibits a strong explicit dependence on the averaged outer divertor target electron temperature, n e,sep ∼ T e,ot −1/2. This dependence is reproduced by analytic reversed two point model (rev-2PM), and arises from parallel pressure balance, as well as the ratio of the power and momentum volumetric loss factors, (1 − f cooling)/(1 − f mom-loss). Quantifying the influence of the (1 − f cooling) and (1 − f mom-loss) loss factors on n e,sep has been enabled by measurement estimates of these quantities from L-mode density (fueling) ramps in the outer horizontal, VH(C), and vertical target, VV, divertor configurations. Rev-2PM n e,sep estimates from the extended H-mode and more limited L-mode datasets are recovered to within ±25% of the measurements, with a scaling factor applied to account for use of , an averaged quantity, rather than flux tube resolved target values. Both the (1 − f cooling) and (1 − f mom-loss) trends and recovery of n e,sep using the rev-2PM formatting are reproduced in EDGE2D-EIRENE L-mode-like and H-mode-like density scan simulations. The general lack of a divertor configuration effect in the JET-ILW n e,sep trends can be attributed to a significant influence of main chamber recycling, which has been shown in the EDGE2D-EIRENE results to moderate n e,sep with respect to changes in divertor neutral leakage imposed by changes in the divertor configuration. The unified n e,sep vs trends can, however, be broken if large modifications to the divertor geometry (e.g. complete removal of the outer divertor baffle structure) are introduced in the model. The more pronounced high-field side high density region formation in the VH(C) configuration with reduced clearance to the separatrix does not appear to have a significant influence on the outer midplane separatrix and pedestal parameters when mapped to , although conditions at the inner midplane could not be assessed.

Effect of the E × B drift on the redistribution of the divertor particle flux in the HL-2A ECRH plasmas

Double-peaked distribution (DPD) of particle flux has only been observed on the outer divertor target in electron cyclotron resonance heating deuterium plasmas with B×∇B toward the X-point in the HL-2A tokamak using high spatiotemporal Langmuir probe arrays. The experimental results demonstrate that the formation of the DPD is mainly due to the enhanced poloidal E×B drift flow stimulated in the divertor region, which is dependent on the plasma density, heating power, and divertor structure. The experimental results are qualitatively consistent with the SOLPS simulation. The experiment also shows that the formation of the DPD might be related to the enhanced cross field transport in the far scrape-off layer. This experimental findings presented here reveal the crucial role played by the synergistic effect of poloidal E × B drift flow and the closed divertor structure in the redistribution of the particle flux, which provides a potential way for the control of high heat flux in future fusion devices.

Modeling of the impact of neon seeding on the detachment in EAST by SOLPS-ITER

In the EAST radiative divertor experiments, stable detachment and divertor protection have been achieved with mixed impurity gas seeding. Experiments indicate that neon (Ne) is an appropriate external impurity for plasma detachment in EAST. In this work, the physical mechanism of plasma detachment with Ne seeding in EAST is studied by using the SOLPS-ITER code package. The electron temperature Tet on the upper outer divertor target decreases to below 10 eV with Ne seeding, which is mainly due to the power loss and upstream power drop caused by the impurity radiation. For Tet < 10 eV, the significant drop of the poloidal saturation ion current js,p occurs around the upper outer strike point. The simulation results show that the js,p drop is mainly driven by the rise of the momentum loss and the upstream pressure loss. The increase in the momentum loss is mainly caused by the plasma–neutral interactions in the region from the X-point to the upper outer divertor target. The upstream pressure loss around the X-point is related to the decrease in the upstream electron temperature Teu, which originates from the increased radiation loss with Ne seeding.

Physics of the electric field in the scrape-off layer in ASDEX Upgrade L-mode discharges and comparison to experiments

The electric field in the Scrape-Off Layer (SOL) of fusion plasmas is a key quantity affecting the local plasma transport and possibly also the overall plasma confinement. However, the physics determining the SOL electric field is experimentally not well investigated. In this study a systematic experimental analysis of the electric field in AUG L-mode discharges at various plasma currents, densities and heating powers is presented. In particular, the relation of the electric field to the divertor condition, as measured by Langmuir probes, is analyzed in detail by applying a simplified form of Ohm’s law. The analysis shows that the peak value of the radial electric field E r in the near SOL measured by Doppler reflectometry decreases from about 8 kV/m at the lowest densities to -2 kV/m at the highest densities, which is accompanied by a flattening of the outer divertor target electron temperature profiles. The electric field obtained by integrating Ohm’s law from the divertor target to the midplane agrees with E r in the near SOL within the uncertainties, evidencing the quantitative validity of the underlying model. Based on the findings about the relation between the electric field and the target conditions, a scaling to obtain the maximum of E r in the SOL in terms of upstream parameters is developed. • The radial electric field E r in the SOL of AUG L-mode discharges is analyzed. • It is shown that E r decreases with increasing Greenwald fraction. • A model for E r using outer target plasma measurements is presented. • A scaling for E r , based on the 2-point model, is introduced.

Design of a Novel Variable Geometry Divertor for Tokamaks

The divertor is a key component of fusion reactors, allowing exhaust of gas, impurities, and helium ash to preserve plasma purity. The divertor geometry strongly affects plasma performance, and it is designed to be compatible with different plasma shapes in present-day fusion experiments. We present a novel concept for a variable geometry divertor, in which the divertor baffle tiles are reorientable by external actuation. Implementation of this concept in a medium-sized research tokamak would uniquely provide the flexibility to tailor divertor geometry to the plasma configuration and also enable study of the effect of divertor closure on plasma performance. To ensure compatibility with typical tokamak operations, the adjustable divertor must withstand the effects of significant mechanical and thermal stresses such as MW/m2-scale heat fluxes and large electromagnetic fields, e.g., disruption forces. The technological solutions for actuation mechanisms, cooling system, gas baffling and plasma-facing components are assessed. A functional reduced-scale model with movable outer divertor target baffle tiles is developed and the actuation mechanism is tested.