Plasma Boundary Research Articles

SOLEDGE3X full vessel plasma boundary simulations of ITER non-active phase plasmas

Abstract The onset of detachment in the ITER machine is analyzed in this work through the help of 2D-axisymmetric boundary plasma simulations with the SOLEDGE3X-EIRENE code, which features a numerical domain for the plasma solver extending up to the first wall. The plasma boundary is computed in scenarios from the first non-active phase of ITER, in pure H and at 20 MW. This set of simulations is used in two aspects: first, to study the plasma detachment in the divertor, and second, the plasma conditions, fluxes, and beryllium erosion at the first wall. Here, the code results are also compared to those obtained with the well-established SOLPS-ITER code, which includes a plasma numerical domain only covering the main SOL. Results show an increase in the SOL width $\lambda_q$ with increasing density, and a detailed analysis is carried out, for the first time, on each of the different plasma-neutral interactions in the code's physics model in EIRENE. The gross beryllium erosion rates of first wall panels are estimated from 2D simulations, with the aim of assessing their sensitivity to two parameters: the divertor density regime, and the presence of density shoulders in the far-SOL formed by enhanced perpendicular transport at this location. The erosion contributions from neutrals and ions are considered in each case, and the charge-exchange atoms fluxes and energy distributions are provided, highlighting the two atom populations (cold and charge-exchange).

Simulations of divertor designs that spatially separate power and particle exhaust using mid-leg divertor particle pumping

Predictive design modeling of a Dissipation-Focused Divertor for future operation in DIII-D reveals that increasing the poloidal distance of the pump duct entrance from the target surface along the low-field side divertor baffle increases neutral compression and modifies the spatial distribution of power dissipation. With a divertor pump located mid-leg between the target and the X-point, SOLPS-ITER boundary plasma simulations without drifts predict the formation of a dense neutral cloud near the target with > 30x higher neutral compression in detachment, a more stable detachment front located further from the target, and ∼ 25 % lower outer midplane separatrix density required for detachment onset, compared to a pump located in the scrape-off layer at the target surface. Up to 19 MW of power flowing into the divertors is modeled using the following two numerical implementations for particle pumping: a specified fraction of particles incident on variable wall sections of the plasma grid is removed from the computational domain (so-called albedo pumping), and a pump duct is modeled which includes dynamics of kinetic neutrals in the duct. The simulations show that the detachment front is located between the divertor target and the X-point and is relatively stable near the pump entrance, without a strong dependence on gas puff rate or injected power. The mid-leg pump design spatially separates the two primary functions of a divertor (power handling and particle exhaust), with the majority of power dissipation occurring near the target plate and particle exhaust taking place further upstream. The benefit of enhanced dissipation using mid-leg pumping comes at the cost of a higher outer midplane separatrix density for a given amount of particle injection.

Real-time plasma boundary shape reconstruction using visible camera on EAST tokamak

Abstract Accurate plasma shape reconstruction is crucial for the stable operation of tokamak devices. In this study, a plasma optical boundary reconstruction system is constructed on the Experimental Advanced Superconducting Tokamak (EAST), and real-time shape reconstruction based on an optical method is realized for the first time on EAST. First, a boundary extraction algorithm based on gray features is designed to extract the optical boundary from plasma optical images. Then, to reconstruct the optical boundary in the tokamak coordinate system, a camera calibration algorithm is developed based on feature point matching, and a coordinate mapping algorithm is designed based on plasma geometric features. Then, the system reconstructs the plasma boundary shape based on images acquired by a camera. Finally, the system is deployed in real-time on EAST, and its reconstruction rate meets the requirement of the EAST plasma shape control system. This study validates the feasibility of using an optical boundary shape reconstruction method on a tokamak device for real-time plasma shape reconstruction, providing a new shape reconstruction approach during steady-state discharge in future fusion devices.

Cross-Scale Energy Transfer from Fluid-Scale Alfvén Waves to Kinetic-Scale Ion Acoustic Waves in the Earth's Magnetopause Boundary Layer.

In space plasmas, large-amplitude Alfvén waves can drive compressive perturbations, accelerate ion beams, and lead to plasma heating and the excitation of ion acoustic waves at kinetic scales. This energy channeling from fluid to kinetic scales represents a complementary path to the classical turbulent cascade. Here, we present observational and computational evidence to validate this hypothesis by simultaneously resolving the fluid-scale Alfvén waves, kinetic-scale ion acoustic waves, and their imprints on ion velocity distributions in the Earth's magnetopause boundary layer. We show that two coexisting compressive modes, driven by the magnetic pressure gradients of Alfvén waves, not only accelerate the ion tail population to the Alfvén velocity, but also heat the ion core population near the ion acoustic velocity and generate Debye-scale ion acoustic waves. Thus, Alfvén-acoustic energy channeling emerges as a viable mechanism for plasma heating near plasma boundaries where large-amplitude Alfvén waves are present.

Impact of a free normal velocity boundary on the modelling of external MHD modes

Free normal-flow (NF) conditions at the plasma boundary are shown to be essential for three-dimensional magnetohydrodynamic (MHD) simulations to agree with linear stability theory. A comparative verification study is presented between two different formulations of the boundary conditions (BCs) for velocity perturbations: (i) fully consistent free NF implementation and (ii) fixed NF formulation, neglecting flow perturbations at the numerical boundary. Numerical results are compared with consolidated figures of merit from the linear theory of external kink modes. We consider two classes of initial equilibria presenting different numerical challenges: a uniform current channel surrounded by pure vacuum and a shaped Wesson-like equilibrium, with smooth (polynomial) radial dependency. Only the fully consistent free NF formulation is invariably accurate in modelling the plasma interface at the numerical boundary, without the need of enforcing a pseudovacuum region at the edge of the simulation domain, as in most analogous past studies. This study employs the cylindrical code SPECYL (Cappello & Biskamp, Nucl. Fusion, vol. 36, no. 5, 1996, p. 571) that solves a full-MHD model without pressure gradients, whose fully consistent resistive wall module with free NF BCs was recently successfully verified against the independent code PIXIE3D (Spinicci et al., AIP Adv., vol. 13, no. 9, 2023, p. 095111).

Application of a large wall electric probe (LWEP) for plasma and ambient gas studies

Abstract The operation of a large wall electric probe (LWEP) is analyzed, where the probe surface area in contact with the plasma is only a few times smaller than the area of the plasma boundaries. Compared to a small standard Langmuir probe (SLP), the LWEP may have significantly greater resolution and sensitivity, but could substantially perturb the plasma and distort the measured properties. A target application for the LWEP is where the sensitivity of the measurement is critical, but the effects of the plasma perturbation is minimal. A specific case where a LWEP may outperform a SLP is the measurement of a nonlocal electron distribution function at energies of electron free diffusion, that is, at energies significantly exceeding thermal electron energies, in order to obtain information about the parameters of the plasma or ambient gas. Examination of the LWEP analysis process has revealed that, depending on the plasma volume and probe configurations, it may be necessary to measure either the first or second derivatives of the probe current with respect to the probe potential to derive the energetic part of the electron energy distribution function.

Dynamic modeling of air–metal plasma mixture of single-turn coil with erosion at megaGauss magnetic field

Single-turn coil (STC) is a destructive pulsed magnet aiming at 100–300 T magnetic field. Due to the high discharge current, the conductor of STC is heated rapidly and undergoes melting and vaporization, leading to the generation of supersonic air–metal vapor mixed plasma jet and the magneto-fluid effect. In this study, the mixed plasma mass-transfer and fluid dynamic characteristics are modeled at megaGauss magnetic field, high temperature, high pressure, and supersonic conductor shock deformation. The collision integral method is employed to calculate the fluid transport properties. In addition, a boundary constraint model of fluid–structure interaction (FSI) compatible with both fluid wall boundary condition and plasma jet entrance condition and a model to simultaneously solve the thermal ionization and high electric field ionization of the mixed vapor are proposed. As the result, the distributions of plasma electrical conductivity, current density, electron, heavy particles, temperature, air body load, and velocity are derived. Especially, the region of highest electrical conductivity is not the air domain near the inner surface of the conductor with the highest electron density and the highest magnetic field, but the air domain near the outer surface of the conductor with the relatively higher electron density and lower magnetic field.

Impact of the surface on He(23S1) and He2(a3Σu+) metastables densities in atmospheric pressure RF plasma

Abstract The dynamic of helium metastable formation along an atmospheric pressure helium plasma channel with different bounding surfaces is analysed. The densities of He(23S1) and He2(a 3 Σ u + ) are measured using optical absorption spectroscopy. A simple model for helium metastable creation, quenching by water impurities and destruction upon surface impact, is developed. The model shows a very good agreement if one assumes that surfaces mainly control secondary electron emission, affecting the local heating at the plasma boundary sheath. This, in turn, changes the rate for He(23S1) and He2(a 3 Σ u + ) conversion. The helium metastable induced desorption of adsorbed water causes a decay of the metastable density along the plasma channel due to quenching.

Manipulating density pedestal structure to improve core–edge integration towards low collisionality

Abstract DIII-D experiments have achieved promising core–edge integrated plasma scenarios which combine a high-temperature low-collisionality pedestal (pedestal top temperature Te ,ped > 0.8 keV and collisionality ν*ped < 1) with a partially detached divertor by leveraging the benefits of a low-density-gradient pedestal in a closed divertor. It is found that with a closed divertor and high heating power, strong gas puffing to achieve detachment moves the peak density gradient outward with respect to the maximum gradient of electron temperature and reduces the density gradient at the pedestal region, which correlates with shallow pedestal fuelling due to the closed divertor geometry. In high-current plasmas in particular, the pedestal top density is found to change little with gas puffing while the separatrix, density increases to allow access for divertor detachment. The separation between density and temperature pedestals results in a high- ηe well above the electron-temperature-gradient stability threshold. Electron turbulence is found to be enhanced in the pedestal and correlated with high ηe resulting from the pedestal shift. The pedestal is wider than the EPED scaling. A revised empirical width scaling is derived based on the combination of EPED scaling with ηe and highlights the important role of additional turbulence on the pedestal structure. The wide temperature pedestal facilitates the achievement of a high-temperature, low-collisionality pedestal and high global performance. Simultaneously, the outward shift of the density pedestal facilitates access to detached divertor conditions with low temperature and heat flux towards the target plate. This approach may be promising for closing the core–edge integration gap for future fusion reactors, which may have a weak-gradient density pedestal due to the highly opaque boundary plasmas.

Power handling in a highly-radiative negative triangularity pilot plant

Abstract This work explores detailed power handling solutions for a class of high-field, highly-radiative negative triangularity (NT) reactors based around the MANTA concept [Rutherford et al. (2024)]. The divertor design is kept as simple as possible, opting for a standard divertor with standard leg length. FreeGS is used to create an equilibrium for the boundary region, prioritizing a short outer leg length of only ∼50 cm (∼40% of the minor radius). The UEDGE code package is used for the boundary plasma solution, to track plasma temperatures and fluxes to the divertor targets. It is found that for PSOL = 25 MW and nsep = 0.96 × 1020 m−3, conditions consistent with initial core transport modeling, little additional power mitigation is necessary. For a fixed impurity fraction of just 0.13\% Ne in the plasma, the peak heat flux density at the more heavily loaded outer targets falls to 7.8 MW/m2, while the electron temperature, Te, remains just under 5 eV. Scans around the parameter space reveal that even at densities lower than in the primary operating scenario, PSOL can be increased up to 50 MW, so long as a slightly higher fraction of extrinsic radiator is used. With less than 1% neon (Ne) impurity content, the divertor still experiences less than 10 MW/m2 at the outer target. Design of the plasma-facing components includes a close-fitting vacuum vessel with a tungsten inner surface as well as FLiBe-carrying cooling channels fashioned into the VV wall directly behind the divertor targets. For the seeded heat flux profile, Ansys Fluent heat transfer simulations estimate that the outer target temperature remains at just below 1550◦C. Initial scoping of advanced divertor designs shows that for an X-divertor, detachment of the outer target becomes much simpler, and plasma fluxes to the targets drop considerably with only 0.01% Ne content.

First SOLEDGE3X-EIRENE simulations of the ITER Neon seeded burning plasma boundary up to the first wall

Boundary plasma simulations are essential to estimate expected divertor and first wall (FW) heat and particle loads on ITER during burning plasma operation. A key missing feature of existing SOLPS simulations (Pitts et al., 2019) is the absence of a plasma solution out to the main chamber walls, essential to self-consistently estimate the gross sputtering of wall material. Here, SOLEDGE3X is applied for the first time to obtain up-to-the wall burning plasma solutions of the ITER boundary plasma at the nominal PSOL = 100 MW of the main SOLPS database simulations, including He ash, Ne seeding but without fluid drifts. Compared with the most recent SOLPS-ITER simulations, our simulations show differences in the exact impurity distribution, but the key results for divertor and wall heat flux remain consistent. In the context of the ITER re-baselining exercise (Pitts, 2024), in which the Be FW armour is proposed to be exchanged for tungsten (W), estimates of W wall sources are key to the assessment of likely core contamination and hence impact on fusion gain. We compare the W gross erosion rates due to the different species excluding W self-sputtering. For the cases simulated spanning 0.27%–0.47% separatrix-averaged Ne concentration and 7.5×1022s−1−1.95×1023s−1 D fuelling, Ne8+ remains the largest contributor to the sputtering flux with the largest source being the outer divertor and baffle. The species-wise contribution to W sputtering changes with fuelling with sputtering due to lower Ne charge states being significant at low D fuelling. In general, the gross W sputtering source is found to decrease with increase in D fuelling and increase with increased Ne seeding.

Control-Oriented Free-Boundary Equilibrium Solver for Tokamaks

A free-boundary equilibrium solver for an axisymmetric tokamak geometry was developed based on the finite difference method and Picard iteration in a rectangular computational area. The solver can run either in forward mode, where external coil currents are prescribed until the converged magnetic flux function ψ(R,Z) map is achieved, or in inverse mode, where the desired plasma boundary, with or without an X-point, is prescribed to determine the required coil currents. The equilibrium solutions are made consistent with prescribed plasma parameters, such as the total plasma current, poloidal beta, or safety factor at a specified flux surface. To verify the mathematical correctness and accuracy of the solver, the solution obtained using this numerical solver was compared with that from an analytic fixed-boundary equilibrium solver based on the EAST geometry. Additionally, the proposed solver was benchmarked against another numerical solver based on the finite-element and Newton-iteration methods in a triangular-based mesh. Finally, the proposed solver was compared with equilibrium reconstruction results in DIII-D experiments.

Stellarator optimization with constraints

In this work we consider the problem of optimizing a stellarator subject to hard constraints on the design variables and physics properties of the equilibrium. We survey current numerical methods for handling these constraints, and summarize a number of methods from the wider optimization community that have not been used extensively for stellarator optimization thus far. We demonstrate the utility of new methods of constrained optimization by optimizing a quasi-axisymmetric stellarator for favourable physics properties while preventing strong shaping of the plasma boundary, which can be difficult to create with external current sources.

Loss of energetic particles due to feedback control of resistive wall mode in HL-3

Effects of three-dimensional (3D) magnetic field perturbations due to feedback control of an unstable ( is toroidal mode number) resistive wall mode (RWM) on the energetic particle (EP) losses are systematically investigated for the HL-3 tokamak. The MARS-F (Liu et al 2000 Phys. Plasmas 7 3681) code, facilitated by the test particle guiding center tracing module REORBIT, is utilized for the study. The RWM is found to generally produce no EP loss for co-current particles in HL-3. Assuming the same perturbation level at the sensor location for the close-loop system, feedback produces nearly the same loss of counter-current EPs compared to the open-loop case. Assuming however that the sensor signal is ten times smaller in the close-loop system than the open-loop counter part (reflecting the fact that the RWM is more stable with feedback), the counter-current EP loss is found significantly reduced in the former. Most of EP losses occur only for particles launched close to the plasma edge, while particles launched further away from the plasma boundary experience much less loss. The strike points of lost EPs on the HL-3 limiting surface become more scattered for particles launched closer to the plasma boundary. Taking into account the full gyro-orbit of particles while approaching the limiting surface, REORBIT finds slightly enhanced loss fraction.

Electron transport barrier and high confinement in configurations with internal islands close to the plasma edge of W7-X

Abstract The low magnetic shear in the Wendelstein 7-X (W7-X) stellarator makes it feasible to shape the separatrix by the large islands constituting an island-divertor, and this can be exploited to access various magnetic configurations, including samples of different internal island sizes and locations. To investigate the configuration effects on the plasma confinement, a configuration scan was performed by changing the coil currents to vary the rotational transform between values 5 / 4 and 5 / 6 at the plasma boundary with different power levels (2, 4, 6 MW) of electron cyclotron resonance heating (ECRH) at a maximum plasma density of 8 × 10 19 m − 2 . neutral beam injection (NBI) heating was also applied during some configurations of the scan to create a density ramp and access high densities beyond the X2 ECRH cutoff. For the magnetic configurations, where the 5 / 5 and 5 / 6 island chains were moved closer to separatrix but remaining inside the last closed flux surface, the electron cyclotron emission shows that an electron temperature, T e , pedestal develops already during ECRH heated plasma buildup phase indicating a transport barrier, and the barrier sustains irrespective of changed plasma heating conditions such as NBI in the later part of discharge. The transport barrier is broken by subsequent fast crashes, observed with multiple plasma diagnostics with characteristics such as tokamak edge localized modes, and the corresponding crash amplitude and frequency vary with plasma pressure. The impact of the transport barrier on plasma confinement can be seen through the increased core T e profile, which could be responsible for the overall increase in the stored diamagnetic energy by approximately 10% for these configurations. After the plasma heating is terminated, a backwards transition to a degraded confinement state is also observed. These observations indicate a configuration triggered high confinement mode in low shear W7-X. This work focuses on the occurrence of this transport barrier for different magnetic configurations and its relation to internal magnetic islands.

Coupled model for liquid lithium plasma facing components

Numerical analysis provides the design choice and operating window of liquid metal Plasma Facing Components (PFC) concepts. Coupled analysis of boundary plasma together with the surrounding boundary structures is required. To achieve this goal, PPPL is developing a comprehensive multi-physics model for modeling of PFCs in fusion devices. The model includes the fluid-kinetic code SOLPS-ITER and the flow and heat transfer code CFX from ANSYS. SOLPS-ITER was augmented with a liquid metal boundary condition algorithm, allowing direct two-way coupling of the plasma analysis with the two-dimensional analytical slab flow model which includes heat convection in the liquid metal PFC. The target heat flux resulting from this coupled analysis is used as a boundary condition for detailed 3D Computational Fluid Dynamics (CFD) Magneto Hydro Dynamics (MHD) and heat transfer analysis. A new formulation of MHD equations is introduced in the numerical procedure ensuring current conservation of the discretized equations. Results of the 3D analysis are used for final validation of the coupled model. A PFC design where a porous wall is used to stabilize the liquid metal surface, while MHD drive is used to push the liquid metal flow inside the PFC, will be investigated in the regimes where vapor shielding is created for enhanced volumetric plasma heat dissipation.

FLARE: field line analysis and reconstruction for 3D boundary plasma modeling

Abstract The FLARE code is a magnetic mesh generator that is integrated within a suite of tools for the analysis of the magnetic geometry in toroidal fusion devices. A magnetic mesh is constructed from field line segments and permits fast reconstruction of field lines in 3D boundary plasma codes such as EMC3-EIRENE. Both intrinsically non-axisymmetric configurations (stellarators) and those with symmetry breaking perturbations of an axisymmetric equilibrium (tokamaks) are supported. The code itself is written in Modern Fortran with MPI support for parallel computing, and it incorporates object-oriented programming for the definition of the magnetic field and the material surface geometry. Extended derived types for a number of different magnetohydrodynamic equilibrium and plasma response models are implemented. The core element of FLARE is a field line tracer with adaptive step-size control, and this is integrated into tools for the construction of Poincaré maps and invariant manifolds of X-points. A collection of high-level procedures that generate output files for visualization is build on top of that. The analysis modules are build with Python frontends that facilitate customization of tasks and/or scripting of parameter scans.

TECXY simulations of the power exhaust in the multi-impurity plasma of DTT reactor

Reduction of the heat load to plasma-facing components is a crucial problem for future fusion reactors like Divertor Test Tokamak (DTT). Mitigation of the power load via increased plasma radiation with the use of puffed ions of impurities would be one way to mitigate power in the scrape-off layer. This paper presents a numerical investigation of the impact of seeded impurities on the radiation pattern and the power load to the divertor plates of the high-field DTT reactor in the single null (SN) configuration. The simulations have been done with the use of the TECXY code, which solves multi-species plasma transport equations for multiple impurity species simultaneously and all associated ionization stages in a two-dimensional poloidal geometry. TECXY represents the model of plasma transport in the scrape-off layer region by a classical set of transport equations of multi-species plasma derived by Braginskii. The paper aims to compare the mitigation capabilities of neon and argon impurities seeded in the plasma of the DTT device and to obtain a significant energy flux reduction to the target plate at the smallest possible impurity concentration. Performed investigations showed the effects of neon and argon impurities seeding separately for constant electron density at the separatrix. It has been found that the decrease in electron temperature on the divertor plates up to 3 eV at the outer and the inner divertor plates and the peak power load below 15 MW m−2 at the outer divertor plate can be achieved with argon seeding and much lower impurity concentration than that in the case of neon impurity seeding. Studies have shown the complexity of the effect of neon and argon impurities on the boundary plasma. It was found that the reduction of temperature and the power on both divertor plates was the most effective for the high upstream plasma densities. The results show also that diffusive perpendicular transport strongly affects impurity radiation and thus plasma condition at divertor plates.

Realization of a gas puff imaging system on the Wendelstein 7-X stellarator.

A system for studying the spatiotemporal dynamics of fluctuations in the boundary of the W7-X plasma using the "Gas-Puff Imaging" (GPI) technique has been designed, constructed, installed, and operated. This GPI system addresses a number of challenges specific to long-pulse superconducting devices, such as W7-X, including the long distance between the plasma and the vacuum vessel wall, the long distance between the plasma and diagnostic ports, the range of last closed flux surface (LCFS) locations for different magnetic configurations in W7-X, and management of heat loads on the system's plasma-facing components. The system features a pair of "converging-diverging" nozzles for partially collimating the gas puffed locally ≈135mm radially outboard of the plasma boundary, a pop-up turning mirror for viewing the gas puff emission from the side (which also acts as a shutter for the re-entrant vacuum window), and a high-throughput optical system that collects visible emission resulting from the interaction between the puffed gas and the plasma and directs it along a water-cooled re-entrant tube directly onto the 8 × 16 pixel detector array of the fast camera. The DEGAS 2 neutral code was used to simulate the Hα (656nm) and HeI (587nm) line emission expected from well-characterized gas-puffs of H2 and He and excited within typical edge plasma profiles in W7-X, thereby predicting line brightnesses used to reduce the risks associated with system sensitivity and placement of the field of view. Operation of GPI on W7-X shows excellent signal-to-noise ratios (>100 at 2 Mframes/s) over the field of view for minimally perturbing gas puffs. The GPI system provides detailed measurements of the two-dimensional (radial and poloidal) dynamics of plasma fluctuations in the W7-X edge and scrape-off layer and in and around the magnetic islands outside the LCFS that make up the island divertor configuration employed on W7-X.

Complex electron-ion-plasma surface modification of high-alloy stainless steel

The work is devoted to identification and analysis of patterns of change in the elemental and phase composition, defective substructure, mecha­nical (microhardness) and tribological (wear resistance and friction coefficient) properties of stainless high-chromium steel subjected to complex processing, combining vacuum irradiation of the samples surface layer with an intense pulsed electron beam of submillisecond exposure duration and subsequent nitriding under electron-ionic heating conditions. High-chromium steel AISI 310S, which in the initial state is a polycrystalline aggregate based on γ-iron, was used as the research material. Pulsed electron beam treatment of steel was carried out on a “SOLO” installation equipped with an electron source with a plasma cathode based on a low-pressure pulsed arc discharge with grid stabilization of the cathode plasma boundary and an open anode plasma boundary. Steel nitriding was carried out on a “TRIO” installation with a chamber size of 600×600×600 mm, equipped with a switching unit to implement the electron-ionic processing mode. Nitriding was carried out at 723, 793, and 873 K temperatures for 1, 3 and 5 h. It was found that electron-ionic nitriding of the samples pre-irradiated with an electron beam (10 J/cm2, 200 μs, 3 pulses at 723 and 793 K for 3 h) is accompanied by the formation of a ceramic layer containing only iron and chromium nitrides. The highest values of steel wear resistance after electron-ionic nitriding, exceeding the wear resistance of the initial steel by more than 700 times, are observed at nitriding parameters of 793 K, 3 h.