Divertor Flux Research Articles

Overview of the first Wendelstein 7-X long pulse campaign with fully water-cooled plasma facing components

After a long device enhancement phase, scientific operation resumed in 2022. The main new device components are the water cooling of all plasma facing components and the new water-cooled high heat flux divertor units. Water cooling allowed for the first long-pulse operation campaign. A maximum discharge length of 8 min was achieved with a total heating energy of 1.3 GJ. Safe divertor operation was demonstrated in attached and detached mode. Stable detachment is readily achieved in some magnetic configurations but requires impurity seeding in configurations with small magnetic pitch angle within the edge islands. Progress was made in the characterization of transport mechanisms across edge magnetic islands: Measurement of the potential distribution and flow pattern reveals that the islands are associated with a strong poloidal drift, which leads to rapid convection of energy and particles from the last closed flux surface into the scrape-off layer. Using the upgraded plasma heating systems, advanced heating scenarios were developed, which provide improved energy confinement comparable to the scenario, in which the record triple product for stellarators was achieved in the previous operation campaign. However, a magnetic configuration-dependent critical heating power limit of the electron cyclotron resonance heating was observed. Exceeding the respective power limit leads to a degradation of the confinement.

Modelling of power exhaust in TCV positive and negative triangularity L-mode plasmas

L-mode negative triangularity (NT) operation is a promising alternative to the positive triangularity (PT) H-mode as a high-confinement edge localised mode-free operational regime. In this work, two TCV Ohmic L-mode core density ramps with opposite triangularity δ≃±0.3 are investigated using SOLPS-ITER modelling. This numerical study aims to investigate the power exhaust differences between NT and PT focusing, in particular, on the geometrical effect of triangularity. To disentangle the latter from differences related to cross-field transport, anomalous diffusivities for particle ( DnAN ) and energy ( χe/iAN ) transport are fixed to the same values in PT and NT. The simulation results clearly show dissimilar transport and accumulation of neutral particles in the scrape-off layer for the two configurations. This gives rise to different ionization sources in the edge and divertor regions and produces differences in the poloidal and cross-field fluxes, ultimately leading to different power and particle divertor fluxes in the two configurations. Simulations recover the experimental feature of a hotter and attached outer target ( Te,OSPNT≳5eV ) in the NT scenario compared to the PT counterpart.

Emulation techniques for scenario and classical control design of tokamak plasmas

The optimization of scenarios and design of real-time-control in tokamaks, especially for machines still in design phase, requires a comprehensive exploration of solutions to the Grad–Shafranov (GS) equation over a high-dimensional space of plasma and coil parameters. Emulators can bypass the numerical issues in the GS equation, if a large enough library of equilibria is available. We train an ensemble of neural networks to emulate the typical shape-control targets (separatrix at midplane, X-points, divertor strike point, flux expansion, and poloidal beta) as a function of plasma parameters and active coil currents for the range of plasma configurations relevant to spherical tokamaks with a super-X divertor, with percent-level accuracy. This allows a quick calculation of the classical-control shape matrices, potentially allowing real-time calculation at any point in a shot with submillisecond latency. We devise a hyperparameter sampler to select the optimal network architectures and quantify uncertainties on the model predictions. To generate the relevant training set, we devise a Markov-chain Monte Carlo algorithm to produce large libraries of forward Grad–Shafranov solutions without the need for user intervention. The algorithm promotes equilibria with desirable properties, while avoiding parameter combinations resulting in problematic profiles or numerical issues in the integration of the GS equation.

Design and simulation of a superconducting machine excitation system taking into account the three-dimensional magnetic leakage

The optimized design of a new high-temperature superconducting rotating pole machine is presented. Its main structural feature is the use of a double stator core which separates the synchronous machine pole shoe from the pole body to rotate separately as the machine rotor, allowing the superconducting coil to operate in a stationary state. The inner stator core, the stationary dewar and the rotor core together form the excitation system of the machine. The excitation coil windings adopt a rectangular cross-section, with flux divertor strategically placed between the high-temperature superconducting coils. This configuration aims to modulate the background magnetic field, specifically reducing the perpendicular magnetic field component. This mitigation minimizes the impact of ambient magnetic fields on the superconducting coil’s current carrying capacity, ensuring an optimized magnetic field environment for its operation. Through the integration of these modifications, the technical and economic parameters of the enhanced high-temperature superconducting machine have been significantly improved. The optimization of design, coupled with detailed calculations of the 3D electromagnetic field, was achieved utilizing the commercial software Ansys EM module.

Integrated control of edge localized modes and divertor flux using mixed toroidal harmonic resonant magnetic perturbations in EAST

Mixed harmonic resonant magnetic perturbations (RMPs) for integrated edge localized modes (ELMs) and divertor flux control are demonstrated in EAST target plasmas of low input torque and normalized beta βN∼ 1.7–1.9, which are close to the equivalent value in ITER high Q operation. The applied RMPs are designed to combine a static harmonic of the toroidal mode number n = 3 with a static or rotating harmonic of n = 2. ELM suppression is achieved without a drop of plasma energy confinement, and tungsten concentration is effectively reduced during the application of RMPs. With mixed harmonics, the toroidal varying steady state heat and particle fluxes on the divertor target can be modified with the rotating n = 2 harmonic, which agrees with the numerical modeling of three-dimensional magnetic topology, with plasma responses being taken into account. ELM suppression correlates with the times of larger n = 3 response with mixed n = 2 and n = 3 RMPs. The mixture of harmonics and the rotating n = 2 harmonic does not require additional coil current because the variation is only in the upper-lower coil current phase space. These results further affirm the effectiveness of integrated ELM and divertor flux control using RMPs with mixed harmonics and improve the understanding of the role of plasma responses in ELM suppression.

Preparing the operation of Wendelstein 7-X in the steady-state regime

The Wendelstein 7-X stellarator (W7-X) started operation in December 2015 and has performed three experimental campaigns up to November 2018. Since then the device has been completed to the planned configuration (most importantly with predominantly actively cooled first wall, including the High Heat Flux divertor and a divertor cryo pump) to be able to perform stationary plasma programs. After this completion pFhase, at the end of 2021, the commissioning has been performed and scientific operation has started in the fall of 2022.

The physics basis to integrate an MHD stable, high-power hybrid scenario to a cool divertor for steady-state reactor operation

Coupling a high-performance core to a low heat flux divertor is a crucial step for ITER and a Fusion Pilot Plant or DEMO. Experiments in DIII-D recently expanded the steady-state hybrid scenario to high density and divertor impurity injection to study the feasibility of a radiating mantle solution. This work presents the physics basis for trade-offs between density, current drive and stability to tearing modes (TMs) at high β. EC power is crucial to tailor the plasma profiles into a passively stable state, and to eject impurities from the core. Off-axis EC depositions decrease the heating efficiency, but calculated electron heat transport coefficients show that this effect is partially mitigated by improved confinement inside the EC deposition. Additionally, the reduction in pressure is recovered by increasing the density. This favourable scaling of confinement with density was observed in high power plasmas for years, and this work provides a comprehensive explanation. ELITE predictions indicate that a path in peeling-ballooning stability opens up for certain conditions of density, power, q 95 and shaping, allowing the edge pressure to continue increasing without encountering a limit. In the core, calculated anomalous fast-ion diffusion coefficients are consistent with density fluctuation measurements in the toroidicity-induced Alfvén eigenmode range, showing that smaller fast-ion losses contribute to the enhanced confinement at high density. The edge integration study shows that divertor heat loads can be reduced with Ne and Ar injection, but this eventually triggers a cascade of n = 1, 2, 3 core TMs. We can now show that impurity radiation in the core is small and it is not the cause for the drop in confinement at high Ar and Ne injection rates. The overlap between the core TMs is consistent with the loss of pressure as estimated by the Belt model for the coupled rational surfaces. Optimization of these trade-offs has achieved plasmas with sustained H 98y2 = 1.7, f GW = 0.7 and ∼85% mantle radiation. The scenario and its variations at higher density and on- vs off-axis EC heating has been studied as a candidate for an integrated solution for several reactor designs, such as ITER, ARC, and the ARIES-ACT1 case, showing promising results in terms of fusion power and gain.

Simulation of Liquid Lithium Divertor Geometry Using SOLPS-ITER

Plasma-facing component (PFC) geometries are evaluated for a Fusion Nuclear Science Facility (FNSF) design to find solutions that are compatible with flowing liquid lithium components while satisfying requirements from the perspective of plasma and neutral particle transport. Flowing liquid metal (LM) divertor systems offer important advantages due to continual regeneration of the surface material, but may require modifications from standard optimized tokamak divertor designs. Scrape Off Layer Plasma Simulator for the International Thermonuclear Experimental Reactor (SOLPS-ITER) simulations are used to compare standard vertical target, open, and balanced baffled geometries. It is found that the open geometry offers little control of neutral particles and has narrow operating windows where the divertor fluxes can be lowered to acceptable levels, while maintaining the required upstream density. The addition of neutral baffles allows greater control over the divertor conditions with less impact on the upstream density. The divertor neutral pressure in the baffled geometry is slightly higher than the open divertor, but not as large with a vertical target. A first-pass coupling of the SOLPS solution with sheath and implantation/erosion models to compute the lithium emission shows that the lithium is largely confined to the near-surface region due to the small ionization mean free path and large main ion flow toward the surface. The lithium source was varied by three orders of magnitude. At the highest rate, when the sourced lithium approaches the level of the main ion recycling flux, a high core lithium concentration (~10%) is observed with only moderate power dissipation to affect the divertor plasma.

JINTRAC integrated simulations of ITER scenarios including fuelling and divertor power flux control for H, He and DT plasmas

We have modelled self-consistently how to most efficiently fuel ITER hydrogen (H), helium (He) and deuterium–tritium (DT) plasmas with gas and/or pellets with the integrated core and 2D SOL/divertor suite of codes JINTRAC. This paper presents the first overview of full integrated simulations from core to divertor of ITER scenarios following their evolution from X-point formation, through L-mode, L–H transition, steady-state H-mode, H–L transition and current ramp-down. Our simulations respect all ITER operational limits, maintaining the target power loads below 10 MW m−2 by timely gas fuelling or Ne seeding. For the pre-fusion plasma operation (PFPO) phase our aim was to develop robust scenarios and our simulations show that commissioning and operation of the ITER neutral beam (NB) to full power should be possible in 15 MA/5.3 T L-mode H plasmas with pellet fuelling and 20 MW of ECRH. For He plasmas gas fuelling alone allows access to H-mode at 7.5 MA/2.65 T with 53–73 MW of additional heating, since after application of NB and during the L–H transition, the modelled density build-up quickly reduces the NB shine-through losses to acceptable levels. This should allow the characterisation of ITER H-mode plasmas and the demonstration of ELM control schemes in PFPO-2. In ITER DT plasmas we varied the fuelling and heating schemes to achieve a target fusion gain of Q = 10 and to exit the plasma from such conditions with acceptable divertor loads. The use of pellets in DT can provide a faster increase of the density in L-modes, but it is not essential for unrestricted NB operation due to the lower shine-through losses compared to H. During the H–L transition and current ramp-down, gas fuelling and Ne seeding are required to keep the divertor power loads under the engineering limits but accurate control over radiation is crucial to prevent the plasma becoming thermally unstable.

Divertor geometry modeling with the SOLPS-ITER code for reactor concepts with liquid metal divertors

• The SOLPS-ITER code has been applied to the analysis of the scrape-off layer (SOL) plasma associated with the fast flow Lithium (Li) divertor design for Fusion Nuclear Science Facility (FNSF). • Two flat divertor configurations are investigated and we found that simplified open geometry presents challenges for neutral control and impurity retention, but balanced geometry with baffling appears promising. • Lithium (Li) is sourced into the balanced configuration and scanning over a large range to understand what level of Li emission is needed to influence the divertor and upstream plasma conditions. • At low sourcing level ( ϕ Li ≤ 1 × 10 23 / s ), the Li shows a minor effect on the plasma and is well confined in the Private flux region (PFR) and divertor surfaces. • At high Li sourcing level ( ϕ Li > 1 × 10 24 / s ), the Li strongly affects the upstream and divertor plasma conditions and leads to strong detachment by increasing momentum and power losses. The SOLPS-ITER code has been applied to the analysis of the scrape-off layer (SOL) plasma associated with the fast flow Lithium (Li) divertor design for Fusion Nuclear Science Facility (FNSF). Two divertor configurations are investigated to determine puff levels that provide robust operational windows that meet the FNSF design requirements on upstream density and divertor flux with a flat divertor for flowing liquid metal (LM). Neon (Ne) is introduced to control the heat flux to the divertors while Deuterium (D 2 ) puffing is performed to maintain sufficiently high upstream density. We found that a simplified open geometry presents challenges for neutral control and impurity retention, but a balanced geometry with baffling is found to have acceptable puff operational windows. Lithium (Li) is sourced into the balanced configuration uniformly along the divertor surfaces and scanning over a large range to understand what level of Li emission is needed to influence the divertor and upstream plasma conditions. At low sourcing level ( ϕ Li ≤ 1 × 10 23 / s ), the Li shows a minor effect on the plasma and is well confined in the private flux region (PFR) and divertor surfaces. At moderate sourcing level ( 1 × 10 23 < ϕ Li ≤ 1 × 10 24 / s ), the Li starts affecting the upstream and divertor plasmas, and Li radiation and electron cooling are almost linearly increasing with the Li sourcing level. At high Li sourcing level ( ϕ Li > 1 × 10 24 / s ), the Li strongly affects the upstream and divertor plasma conditions and leads to strong detachment by increasing momentum and power losses. At high sourcing level, Li reaches above the x-point that can result in a substantial Li radiation in the core and sufficient dissipation to strongly affect the divertor and upstream plasmas.

First snowflake divertor experiments in MAST-U tokamak

First snowflake (SF) divertor experiments in the MAST-U tokamak demonstrated steady-state snowflake-plus divertor configurations in 450 kA ohmic L-mode plasmas. The SF divertor configuration features a second poloidal field (PF) null in the divertor region close by or overlapping with the main X-point. The resulting low PF region and two additional divertor legs (strike points) may lead to additional power and particle flux sharing via a hypothesized convective cell, and increased plasma-wetted area and radiation. The free-boundary Grad–Shafranov equilibrium code FIESTA was used to design SF configurations with several inter-null distances and orientations. In the experiment, the SF configurations with inter-null distances 0.13–0.20 m and lasting 0.2–0.3 s were obtained. Parallel connection lengths between the midplane and the outer strike point in the SF configurations (evaluated at field lines 1–2 mm from the separatrix in the midplane) were 25–30 m, higher than in the standard divertor (20–25 m) or the Super-X divertor (25 m). Diagnostic measurements highlighted salient SF features. The infra-red video bolometer diagnostic showed that the radiated power peaking in the PF null region was not as pronounces as in the standard divertor. Divertor ion fluxes measured by target Langmuir probes showed increased ion flux in the plate region where a secondary SF strike point landed, concomitantly with the SF formation. These measurements may suggest that some particle and heat redistribution was taking place in the convective SF zone. The first SF experiments provide a basis for future SF studies in MAST-U tokamak with higher input power, improved plasma control and diagnostic measurements, to be compared with the modeling predictions of plasma convective SF mixing and lower density strike point detachment threshold. • First snowflake (SF) divertor experiments in the MAST-U tokamak demonstrated steady-state snowflake-plus divertor configurations with inter-null distances 0.13–0.20 m in 450 kA ohmic L-mode plasmas. • The free-boundary Grad–Shafranov equilibrium code FIESTA was used to design SF configurations with several inter-null distances and orientations. • Divertor geometry (i.e., parallel connection lengths, divertor flux expansion and area expansion) in the SF-plus divertor fvorably comared to the standard divertor • Divertor measurements with the infra-red video bolometer diagnostic and target Langmuir probes may suggest that some particle and heat redistribution took place in the convective SF zone.

Non-linear MHD modelling of edge localized modes suppression by resonant magnetic perturbations in ITER

Edge localized modes (ELMs) suppression by resonant magnetic perturbations (RMPs) was studied with the non-linear magneto-hydro-dynamic (MHD) code JOREK for the ITER H-mode scenarios at 15 MA, 12.5 MA, 10 MA/5.3 T. The main aim of this work was to demonstrate that ELMs can be suppressed by RMPs while the divertor 3D footprints of heat and particle fluxes remain within divertor material limits. The unstable peeling–ballooning modes responsible for ELMs without RMPs were modelled first for each scenario using numerically accessible parameters for ITER. Then the stabilization of ELMs by RMPs was modelled with the same parameters. RMP spectra, optimized by the linear MHD MARS-F code, with main toroidal harmonics N = 2, N = 3, N = 4 have been used as boundary conditions of the computational domain of JOREK, including realistic RMP coils, main plasma, scrape off layer (SOL) divertor and realistic first wall. The model includes all relevant plasma flows: toroidal rotation, two fluid diamagnetic effects and neoclassical poloidal friction. With RMPs, the main toroidal harmonic and the non-linearly coupled harmonics remain dominant at the plasma edge, producing saturated modes and a continuous MHD turbulent transport thereby avoiding ELM crashes in all scenarios considered here. The threshold for ELM suppression was found at a maximum RMP coils current of 45 kAt–60 kAt compared to the coils maximum capability of 90 kAt. In the high beta poloidal steady-state 10 MA/5.3 T scenario, a rotating QH-mode without ELMs was observed even without RMPs. In this scenario with RMPs N = 3, N = 4 at 20 kAt maximum current in RMP coils, similar QH-mode behaviour was observed however with dominant edge harmonic corresponding to the main toroidal number of RMPs. The present MHD modelling was limited in time by few tens of ms after RMPs were switched on until the magnetic energy of the modes saturates. As a consequence the thermal energy was still evolving on this time scale, far from the ITER confinement time scale and hence only the form of 3D footprints on the divertor targets can be indicated within this set-up. Also note, that the divertor physics was missing in this model, so realistic values of fluxes are out of reach in this modelling. However the stationary 3D divertor and particle fluxes could be simply extrapolated from these results to the stationary situation considering that a large power fraction should be radiated in the core and SOL and only about 50 MW power is going to the divertor, which is an arbitrary, but reasonable number used here. The 3D footprints with RMPs show the characteristic splitting with the main RMP toroidal symmetry. The maximum radial extension of the footprints typically was ∼20 cm in inner divertor and ∼40 cm in outer divertor with stationary heat fluxes decreasing further out from the initial strike point from ∼5 MW m−2 to ∼1 MW m−2 assuming a total power in the divertor and walls is 50 MW. The heat fluxes remain within the divertor target and baffle areas, however with rather small margin in the outer divertor which could be an issue for the first wall especially in transient regimes when part of the plasma thermal energy is released due to switching on the RMP coils. This fact should be considered when RMPs are applied with a more favorable application before or soon after the L–H transition, although optimization is required to avoid increasing the L–H power threshold with RMPs.

Recent progress in the assessment of irradiation effects for in-vessel fusion materials: tungsten and copper alloys

The present contribution highlights results of the recent irradiation campaigns applied to screen mechanical properties of advanced tungsten and copper-based materials—main candidates for the application in the plasma-facing components (PFC) in the European DEMO, which has also been presented at 28th IAEA fusion energy conference. The main challenges in the formulated irradiation programme were linked to: (I) assessment of the ductile-to-brittle transition temperature of newly developed tungsten-based materials; (ii) investigation of an industrial pure tungsten grade under high temperature irradiation, reflecting operational conditions in the high flux divertor region; (iii) assessment of the high temperature strength of CuCrZr-based alloys and composites developed to enable the extension of the operational window for the heat sink materials. The development and choice of the advanced materials is driven naturally by the need to extend the operation temperature/fluence window thereby enlarging the design space for PFCs. The obtained results helped identifying the prospective tungsten and copper-based material grades as well as yielded a number of unexpected results pointing at severe degradation of the mechanical properties due to the irradiation. The results are discussed along with the highlights of the microstructural examination. An outlook for near future investigations involving in-depth post-irradiation examination and further irradiation campaigns is provided.

Analyses of monoblocks divertor at one-sided high heat flux using RPI wall boiling model

The divertor is one of the core components in the Tokamak magnetic confinement nuclear fusion device, which is responsible for eliminating high-flux energy flow and particle flow. The divertor in China Fusion Engineering Test Reactor (CFETR) is designed to withstand a steady-state thermal load of 10 MW/m2 and a transient thermal load up to 20 MW/m2. The water-cooled scheme is a promising candidate scheme. The model under high heat flux divertor tube system was established by using RPI wall boiling model and the Eulerian multiphase model. On this basis, the fluid-thermal-structure coupling method is used to simulate the complete structure of divertor blocks. In the steady-state condition, three heat transfer patterns are found and compared in the circumferential and axial directions, and methods to enhance heat transfer (including flow velocity and tube structure) are further evaluated. The results reveal the superiority of the screw tube and the swirl tube in the enhancement of heat transfer with high heat flow. For unsteady conditions, the response of temperature and stress for different components are different. The change of flow boundary quickly affects the whole system, while the thermal shock only poses a great threat to external components, and its influence on internal components is lagging and slight.

Integrated ELM and divertor power flux control using RMPs with low input torque in EAST in support of the ITER research plan

Experiments have been carried out at the EAST tokamak to study ITER-relevant scenario integration issues, related to edge localized mode (ELM) control in H-mode plasmas by the application of three-dimensional (3D) resonant magnetic perturbations (RMPs), which have a large impact on the execution of the ITER research plan. The EAST experiments have successfully demonstrated ELM suppression at normalized torque inputs similar to ITER. The application of RMP fields with high toroidal mode number (n = 4) reduces the impact of ELM control on energy and particle confinement compared to those use lower n (n = 1, 2) RMPs. Injection of successive pellets is found to be effective in increasing the plasma density in ELM-suppressed H-modes and reducing the divertor power without triggering large ELMs at EAST. Access to high recycling and radiative divertor conditions while maintaining ELM suppression has been demonstrated in EAST by the use of gas fuelling and neon impurity seeding. Both approaches have been found to be effective in reducing power fluxes to the divertor strike points in near-separatrix lobes for both n = 2 and n = 4 RMPs. However, reduction of power fluxes in off-separatrix lobes is only effective for n = 4 RMP application, which is consistent with magnetic topology modelling (including plasma response) results showing a shallow penetration into the confined plasma region of field lines connected to these lobes compared to n = 2. The EAST results support the use of high n 3D fields for ELM suppression in ITER high Q DT scenarios since they provide optimum integration features regarding energy and particle confinement, pellet fuelling, radiative divertor operation while eliminating ELM transient power loads and being compatible with low torque input.

Mechanism of the active divertor flux control by the supersonic-molecular-beam-injection with lower hybrid wave-induced magnetic perturbations on the EAST tokamak

The redistribution of the divertor flux caused by the synergy of the supersonic-molecular-beam-injection (SMBI) and the magnetic perturbations induced by lower hybrid waves (LHWs), has been observed on the Experimental Advanced Superconducting Tokamak (EAST) (Li et al 2013 Nature Phys. 9 817). To reveal the physical mechanism, simulations with good agreements to the experimental findings are first performed by utilizing a self-consistent fluid 3D edge plasma Monte-Carlo code coupled to a kinetic neutral particle transport code. The ions and electrons originating from the ionization of injected neutral particles in the plasma edge flow along the magnetic flux tube towards to the divertor, thus directly increasing the divertor flux on the split strike lines in the footprint. Combining this with the multi-lobe structure of the edge magnetic topology, actively controlling the divertor flux can be realized by adjusting the SMBI position or the phase of the magnetic perturbations.

Investigation of edge impurity transport and divertor fluxes by toroidally localized lithium injection on EAST with EMC3-EIRENE

Three-dimensional (3D) edge fluid transport code EMC3-EIRENE has been employed to study lithium transport in the scrape-off layer (SOL) of EAST with the toroidally localized lithium injection experiment. The simulated profiles of the density of the Li ions show that the Li1+ ions mainly populate at the upper private region, while the Li2+ and Li3+ ions distribute at the upper X-point and upstream SOL regions. The synthetic 3D line-integrated LiII emission image obtained by EMC3-EIRENE is in reasonable agreement with the experimental profile measured by the CCD camera system on EAST. The impacts of the poloidal locations of Li injection (LI) have been studied to evaluate the variation of the Li impurity transport and emission behaviors, which shows a strong dependence on the poloidal position of LI. It is found that the LI at the outer midplane is the most effective location to radiate the power and reduce the heat flux deposition on the divertor targets, in comparison with the other poloidal injection locations.

Performance assessment of long-legged tightly-baffled divertor geometries in the ARC reactor concept

Extremely intense power exhaust channels are projected for tokamak-based fusion power reactors; a means to handle them remains to be demonstrated. Advanced divertor configurations have been proposed as potential solutions. Recent modelling of tightly baffled, long-legged divertor geometries for the divertor test tokamak concept, ADX, has shown that these concepts may access passively stable, fully detached regimes over a broad range of parameters. The question remains as to how such divertors may perform in a reactor setting. To explore this, numerical simulations are performed with UEDGE for the long-legged divertor geometry proposed for the ARC pilot plant conceptual design—a device with projected heat flux power width () of 0.4 mm and power exhaust of 93 MW—first for a simplified Super-X divertor configuration (SXD) and then for the actual X-point target divertor (XPTD) being proposed. It is found that the SXD, combined with 0.5% fixed-fraction neon impurity concentration, can produce passively stable, detached divertor regimes for power exhausts in the range of 80–108 MW—fully accommodating ARC’s power exhaust. The XPTD configuration is found to reduce the strike-point temperature by a factor of ∼10 compared to the SXD for small separations (∼1.4) between main and divertor X-point magnetic flux surfaces. Even greater potential reductions are identified for reducing separations to ∼1 or less. The power handling response is found to be insensitive to the level of cross-field convective or diffusive transport assumed in the divertor leg. By raising the separatrix density by a factor of 1.5, stable fully detached divertor solutions are obtained that fully accommodate the ARC exhaust power without impurity seeding. To our knowledge, this is the first time an impurity-free divertor power handling scenario has been obtained in edge modelling for a tokamak fusion power reactor with of 0.4 mm.

Simulation study of the impurity radiation in the quasi-snowflake divertor with Ar seeding for CFETR

The snowflake divertor is considered as a possible candidate to exhaust the huge amount of power from core plasma for future fusion reactors. In this work, the radiative quasi-snowflake (QSF) divertor with argon seeding was studied by SOLPS5.0 simulation for the China Fusion Engineering Test Reactor. The relation between radiation power, Prad, electron density, ne, and effective ion charge, Zeff, was established based on the density and radiation fraction scan for both QSF and the ITER-like divertor (ILD). The comparison of radiative efficiency shows that QSF divertor is able to radiate more power with much lower impurity concentration than the ILD, when upstream density is relatively high. Further analyses are performed for the low (ne,sep ~ 2.3 × 1019 m−3) and high density cases (ne,sep ~ 4.5 × 1019 m−3). For low density cases, outer divertor plasma of QSF is firstly detached due to a much pronounced increase of outer divertor flux expansion, while the inner divertor detaches at first for the ILD configuration. Furthermore, the degree of detachment is much higher for the outer divertor in the QSF configuration under similar Prad, which can be beneficial for impurity screening in the outer divertor. For high density cases, both divertor targets of QSF are detached with relatively lower radiated power while the outer target of ILD is still attached. Zeff at outer mid-plane for QSF is obviously lower than that for ILD with similar Prad, which means better impurity screening in QSF.

Impact of a resonant magnetic perturbation field on impurity radiation, divertor footprint, and core plasma transport in attached and detached plasmas in the Large Helical Device

The effects of resonant magnetic perturbation (RMP) field on impurity radiation, divertor footprint distribution, and core plasma transport are investigated in the detachment discharges of the Large Helical Device (LHD). The RMP with m/n = 1/1 mode creates an edge magnetic island in the stochastic layer, which enhances the impurity emission from low charge states, C2+ and C3+, and then triggers a detachment transition. Emission from the higher charge states, C4+ and C5+, implies enhanced penetration of impurities during the detachment phase with RMP. The toroidal divertor particle flux distribution exhibits n = 1 mode structure in both the attached and detached phases, but with a large toroidal phase shift between the two phases. The distribution in the attached phase is well correlated with the magnetic footprint of field line connection length calculated by the vacuum approximation. During the detached phase, however, the phase shift is not well explained by the vacuum approximation, where a significant plasma response to the external RMP is observed. The energy confinement time becomes systematically shorter with RMP application due to the shrinkage of plasma volume caused by the edge magnetic island. On the other hand, the pressure profile during detachment with RMP is found to be more peaked than without RMP. The analysis using the core transport code TASK3D, considering the heating profiles of neutral beam injection, shows no significant transport degradation during detachment with RMP application, even with the enhanced radiation, reduced divertor flux, and possible impurity penetration.