Abstract A novel multicode workflow to interpret collector probe deposition patterns in DIII-D has been developed. The components of the workflow consist of a detailed computer-aided-design file of the vessel wall and the scrape-off-layer (SOL) codes MAFOT, OSM, DIVIMP, and 3DLIM. A special-purpose toolkit enables passing the output of these codes among each other to provide a full-SOL picture of impurity transport. A demonstration of the workflow is described to support evidence of near-SOL tungsten parallel accumulation during trace W impurity experiments on DIII-D. Iteration between simulated deposition patterns in 3DLIM and DIVIMP predicts a region of elevated W density near the separatrix about halfway between the outboard midplane and the top of the plasma. This workflow will be used to better interpret collector probe experiments on DIII-D.

Separatrix-to-Wall Simulations of Impurity Transport with a Fully Three-Dimensional Wall in DIII-D

Tungsten (W) is used as the plasma-facing material in the divertor region of future fusion reactors, such as ITER; however, its concentration in the core plasma must be maintained at an extremely low level. W transport in the scrape-off layer (SOL), which is related to the source of core W contamination, has been extensively studied. In this study, the influence of E× B drift on the transport of W impurity in the SOL is studied via numerical simulations of a model case based on EAST upper single-null configuration with high recycling divertor plasma. W transport is simulated using DIVIMP on the background plasma obtained from scape-off layer plasma simulation-ITER simulation including drifts. The E× B drift of W ions is introduced based on the background electric field. Therefore, both the direct E× B drift effect of W ion and the indirect effect via background plasma on W transport in the SOL are studied. The influence on the flux of W impurities entering confined plasma across the last closed flux surface Γ enter is focused on, which is expected to be proportional to the core W concentration. Results reveal that Γ enter is mainly from the outer (inner) target under a favorable (unfavorable) toroidal field B T and can be increased by more than one order of magnitude compared with the case without drifts; this reflects the significant effect of E× B drift. The effects due to the background plasma and the poloidal and radial E× B drift of W ion, as well as the related mechanisms, are analyzed in detail for three stages of W transport in the SOL: effective sputtering from the target, leakage from the divertor, and entry into the confined plasma.

Simulation study of the influence of E× B drift on tungsten impurity transport in the scrape-off layer


 Numerical simulations explore a possible tightly baffled, long-legged divertor (TBLLD) concept in a future upgrade of the Tokamak à configuration variable (TCV). The SOLPS-ITER code package is used to compare the exhaust performance of several TBLLD configurations with results from unbaffled and baffled TCV configurations. The investigated TBLLDs feature a range of radial gaps between the separatrix and the divertor baffles, with a smaller gap resulting in tighter baffling. All modelled TBLLDs are predicted to lead to a denser and colder plasma in front of the targets and increase the power handling by factors of 2-3 compared to the present, baffled, divertor and by up to a factor of 12 compared to the original, unbaffled, configuration. This improved TBLLD performance is attributed to an increased neutral confinement with more plasma-neutral interactions in the divertor region. Both power handling capability and neutral confinement increase with tighter baffling. The core compatibility of TBLLDs with nitrogen seeding is also evaluated and the detachment window, with acceptable core pollution, for these TBLLDs is explored, showing a reduction of the required upstream impurity concentration to achieve detachment by up to 18% with tighter baffling.

https://iopscience.iop.org/article/10.1088/1741-4326/ace45f/pdf

Performance assessment of a tightly baffled, long-legged divertor configuration in TCV with SOLPS-ITER

Abstract Tungsten (W) impurity eroded from the divertor target can degrade the confinement performance of future fusion reactors. The mechanism of the effect of E × B drift on the W transport in the scrape‐off layer (SOL) was investigated with a fixed cross‐field anomalous particle diffusion coefficient for W ions ( D ⊥,W ) of 1 m 2 s −1 in our previous simulation work, which indicates that the W flux entering the confined region ( Γ enter ) and mean W concentration in the confined region () can be increased by more than one order of magnitude by drifts. In this work, the influence of the cross‐field anomalous diffusion on the W transport in the SOL is further investigated. The basic flow pattern of the W ion flux ( Γ W ) keeps when D ⊥,W varying from 0.5 to 3 m 2 s −1 , and Γ enter mainly comes from the hotter divertor region. Results show that D ⊥,W value affects the W leakage from divertor region and entry into confined region. Therefore, Γ enter and both decrease with increasing D ⊥,W . Mechanisms for cases w/wo drifts are compared. 

Simulation study of the influence of the cross‐field anomalous diffusion on the tungsten impurity transport in the scrape‐off layer with activated drifts

Helicon waves are thought to be promising in various tokamaks, such as DIII-D, because they can penetrate reactor-grade high-density cores and drive the off-axis current with higher efficiency. In the frequency regime ∼ 476 MHz, both slow electrostatic and fast electromagnetic helicon waves can coexist in DIII-D. If the antenna parasitically excites the slow mode, these waves can propagate along the magnetic field line into the scrape-off layer (SOL). Although the importance of the misalignment of the Faraday screen and the electron density in the SOL on the excitation and propagation of slow modes is well known, the conditions for minimizing slow mode excitation have yet to be optimized. Using the Petra-M simulation code in the 2D domain, we analyze the effects of the misalignment of the antenna in the poloidal direction, the misalignment of the Faraday screen in the toroidal direction, and the density in front of the antenna on slow mode generation. Our results suggest that the misalignment of the Faraday screen is a critical factor in reducing the slow mode and that the misalignment angle should be below ∼ 5° to minimize the slow wave excitation. When the electron density is higher than 3.5×1018 m−3 in the SOL, the generation of the slow mode from the antenna is minimized and unaffected by the misalignment of the Faraday screen. 

Full-wave simulations on helicon and parasitic excitation of slow waves near the edge plasma

Impurity transport modeling of the new tungsten (W)-coated, V-shaped small angle slot (SAS) divertor in the DIII-D tokamak was conducted using the SOLPS-ITER plasma edge code package and the DIVIMP impurity tracking code. The inboard baffle of the current SAS divertor will be shifted closer to the outboard baffle, creating a V-corner at the slot vertex. In addition, the outboard baffle will be coated with 10–15 μm of W for experiments studying high-Z sourcing and leakage in a closed divertor. Modeling of the ‘SAS-VW’ divertor predicts that these changes to the inner baffle will reduce W gross erosion by 40× relative to the existing SAS divertor when the outer strike point (OSP) is at the V-corner and the ion B × ∇B drift is towards the divertor, driven primarily by significant cooling near the slot vertex. Most W erosion in SAS-VW is expected to occur near the slot entrance, which may pose a higher risk to core contamination than W eroded deeper in the slot. Adding a new sheath-based prompt redeposition model outlined in Guterl et al (2021 Nucl. Mater. Energy 27 100948) increases the sensitivity of redeposition estimates to near-target plasma conditions and may provide more accurate predictions of net erosion. Moving the OSP outboard from the slot vertex ∼4 cm onto the W-coated region yields a 40× increase in the gross erosion rate and a 50% decrease in the core leakage fraction. Thus slight variations in strike point location may counteract the potential benefits of the tightly-baffled V slot on minimizing erosion. This impurity transport modeling provides useful guidance for future experiments on the SAS-VW divertor focused on high-Z erosion/redeposition, scrape-off layer transport, and core leakage.

Predicting tungsten erosion and leakage properties for the new V-shaped small angle slot divertor in DIII-D

The integration of a high-performance core and a dissipative divertor, or the so-called ‘core–edge integration,’ has been widely identified as a critical gap in the design of future fusion reactors. In this letter, we report, for the first time, direct experimental evidence of electron turbulence at the DIII-D H-mode pedestal that correlates with the broadening of the pedestal and thus facilitates core–edge integration. In agreement with gyrokinetic simulations, this electron turbulence is enhanced by high η e (η e = Ln /LT e, where Ln is the density scale length and LT e is the electron temperature scale length), which is due to a strong shift between the density and temperature pedestal profiles associated with a closed divertor. The modeled turbulence drives significant heat transport with a lower pressure gradient that may broaden the pedestal to a greater degree than the empirical and theoretically predicted pedestal width scalings. Such a wide pedestal, coupled with a closed divertor, enables us to achieve a good core–edge scenario that integrates a high-temperature low-collisionality pedestal (pedestal top temperature T e,ped > 0.8 keV and a pedestal top collisionality ν*ped < 1) under detached divertor conditions. This paves a new path toward solving the core–edge integration issue in future fusion reactors.

https://iopscience.iop.org/article/10.1088/1741-4326/acdfe2/pdf

Direct measurement of the electron turbulence-broadening edge transport barrier to facilitate core–edge integration in tokamak fusion plasmas

• The compact SAS divertor in DIII-D provides a good testbed for understanding the effects of a tightly closed divertor on particle and power dissipation, and for application to core-edge integration solutions. • A longer outer leg facilitates the achievement of divertor dissipation in the SAS divertor, while a shorter leg leads to higher divertor temperatures and requires higher upstream densities to achieve the same level of divertor detachment. • The divertor conditions strongly depend on the divertor target shaping and Bt direction. • In-slot gas puffing can reduce detachment onset density by about 10% when the outer strike point is placed at the inner slanted surface. • Completed divertor detachment and a high confinement core with normalized energy confinement factor H 98 >1.0 can be simultanesouly achieved with the SAS divertor, demonstrating the benefit of a closed divertor for exploration of core-edge integration. Dedicated experiments in DIII-D find that magnetic shaping and divertor target geometry significantly affect the divertor plasma conditions and divertor detachment process in the small-angle-slot (SAS) divertor. The compact SAS divertor in DIII-D provides a good testbed for understanding the effects of a tightly closed divertor on particle and power dissipation, and for application to core-edge integration solutions. A longer outer leg facilitates the achievement of divertor dissipation in the SAS divertor, while a shorter leg leads to higher electron temperatures near the divertor target plate and requires higher upstream densities to achieve the same level of divertor detachment. In addition, with the ion B × ∇ B drift away from the SAS divertor and the outer strike point (OSP) near the outer corner, the target temperature is lower for a particular upstream density than with the OSP on a slanted or flat surface, leading to lower heat flux even when the particle flux remains similar. In contrast, with the ion B × ∇ B drift into the SAS divertor, a strike point at the inner slanted surface exhibits a lower upstream density to achieve divertor detachment than a strike point either at the outer corner or the outer slanted target. Experimental results and SOLPS-ITER simulations with full drifts suggest the strong interplay between drift flows and the neutral distribution resulted from target shaping. Furthermore, in-slot gas puffing has been shown to achieve global divertor detachment with an onset density about 10% lower than that using main-chamber gas puffing when the outer strike point is placed at the inner slanted surface. Corresponding modelling reveals that the local gas puffing enhances the neutral ionization which potentially facilitates the achievement of divertor dissipation. However, such improvement diminishes when the strike point is at the outer corner, which also indicates the geometric dependence on divertor performance in the SAS divertor. Even with different strike point locations, complete divertor detachment with very low particle and heat fluxes at the divertor targets and a high confinement core with normalized energy confinement factor H 98 >1.0 can be simultanesouly achieved with the SAS divertor with ion B × ∇ B drift into SAS divertor, demonstrating the benefit of a closed divertor for exploration of core-edge integration. 

Dependence of particle and power dissipation on divertor geometry and plasma shaping in DIII-D small-angle-slot divertor

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Predicting tungsten erosion and leakage properties for the new V-shaped small angle slot divertor in DIII-D

Direct measurement of the electron turbulence-broadening edge transport barrier to facilitate core–edge integration in tokamak fusion plasmas

Dependence of particle and power dissipation on divertor geometry and plasma shaping in DIII-D small-angle-slot divertor