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X-Point Radiator (XPR) regimes have been obtained in WEST tokamak experiments with nitrogen seeding during the experimental campaigns of 2023 and 2024. These experiments showed the formation of a stable toroidal radiating ring near the X-point, similar to observations in other devices such as JET, ASDEX-Upgrade, TCV, and COMPASS. In WEST, the onset of this regime is associated with a sharp transition of the divertor plasma from hot to cold and dense conditions, with increased particle fluxes, indicating that the plasma is not detached. At the same time, core conditions are significantly improved. These scenarios were successfully controlled in WEST using an interferometry line-of-sight passing through the X-point. Interpretative modeling of these discharges with the SOLEDGE3X-EIRENE code reveals that a physics mechanism needed to stabilize WEST nitrogen XPRs is not present when driving the simulations at constant power. On the contrary, a stable XPR can be obtained by increasing the power injected at the time of the XPR onset to represent the reduction of W contamination, highlighting the need to describe the plasma dynamics and go toward integrated core–edge simulations.

The effect of thermodynamic nonequilibrium of hydrogen isotope recycling, which consists in significant deviation of the velocity distribution function (VDF) of neutral hydrogen isotope atoms from the Maxwellian VDF, on the passive signal of the charge-exchange recombination spectroscopy (CXRS diagnostics) of edge plasma in tokamaks is analyzed using the ITER tokamak reactor as an example. The calculations of the VDF for atomic and molecular deuterium are carried out using the three-dimensional Monte Carlo code EIRENE as a part of the two-dimensional transport code SOLPS and the semi-analytical Ballistic Model BM1D2V, which is one-dimensional in coordinate and two-dimensional in projections of atomic velocity. The source of level population of the Be IV beryllium ion, the line radiation of which is used in the CXRS diagnostics, is calculated for the above two methods of calculating the VDF and the average atomic kinetic energy, as well as in the case of neglecting the thermodynamic nonequilibrium of recycling, when the temperature of atoms is equal to the temperature of the plasma ions. The calculations of the cross section for the charge-exchange reaction of beryllium ions Be V on deuterium atoms using the ARSENY code and other codes and the erosion kinetics of the beryllium first wall using the three-dimensional code ERO2.0 are used. The calculations of the level populations of the Be IV beryllium ion using the nl-KinRyd code, the luminosity of the ions, and the passive charge-exchange radiation intensity on the observation chord are carried out for the three indicated cases of the VDF of deuterium atoms. A significant dependence of the results of the conducted predictive modeling on the thermodynamic nonequilibrium of recycling is shown. [ABSTRACT FROM AUTHOR]

The dithering H-mode phase, characterized by oscillations, is generally observed at input power values close to the L-H transition power threshold and low plasma collisionalities (low electron density and/or high plasma temperature). Measurements to characterize the dithering phase are presented for the low aspect ratio, high magnetic field tokamak, ST40. The dithering phase oscillation frequency is observed between 400 and 800 Hz and demonstrates an inverse relationship with core plasma density. Dithering phase H-modes are documented across a nonlinear, low-density power threshold operational space, with signature low- and high-density branches. The minimum power threshold for dithering H-mode access is measured at a core, line average electron density of 4.7(±0.5)×1019 m−3, close to a predicted value of 4.1(±0.4)×1019 m−3 from multimachine studies. ASTRA calculated values of power coupled to the ion species, at the dithering H-mode transition, exhibit a similar nonlinear dependence on density. This analysis points to the important contribution of the ion thermal channel to the L-H phase transition. The low-frequency plasma density and D-alpha dithers appear to be accompanied by sudden bursts of magnetohydrodynamic (MHD) activity. A simple model is tested to demonstrate a possible scenario of self-regulation among turbulence, zonal flows, pressure (density) gradient and MHD activities. [ABSTRACT FROM AUTHOR]

We report on comprehensive experimental studies of turbulence spreading in edge plasmas. These studies demonstrate the relation of turbulence spreading and entrainment to intermittent convective density fluctuation events or bursts (i.e. blobs and holes). The non-diffusive character of turbulence spreading is thus elucidated. The turbulence spreading velocity (or mean jet velocity) manifests a linear correlation with the skewness of density fluctuations, and increases with the auto-correlation time of density fluctuations. Turbulence spreading by positive density fluctuations is outward, while spreading by negative density fluctuations is inward. The degree of symmetry breaking between outward propagating blobs and inward propagating holes increases with the amplitude of density fluctuations. Thus, blob-hole asymmetry emerges as crucial to turbulence spreading. These results highlight the important role of intermittent convective events in conveying the spreading of turbulence, and constitute a fundamental challenge to existing diffusive models of spreading.

The mathematical model of a rarefied edge plasma is considered in this paper. The Vlasov–Poisson equations are solved based on the developed algorithm of successive iterations in time after a pulse change in one or more parameters of the problem. The problem is multidimensional, nonstationary, and multiparameter. The developed fundamental special techniques make it possible to carry out calculations without the use of supercomputers. Mathematical models of the edge plasma parameters near a flat structural element of a spacecraft moving in the ionosphere along a circular orbit are reported. [ABSTRACT FROM AUTHOR]

Estimates of plasma conditions in the far scrape-off layer (SOL) and of first wall (FW) fluxes in ITER are key input parameters to first wall erosion and impurity migration models, which are in turn involved in the assessment of FW panel lifetime and fuel retention studies. SOLEDGE3X up-to-the wall boundary plasma simulations are performed for ITER, based on an expected Pre-Fusion-Power-Operation (PFPO-1) scenario at PSOL= 20MW, including an impact study of enhanced far-SOL transport. This latter study concerns the possible formation of density shoulders, which are modelled here by applying an increase to the prescribed perpendicular particle and heat diffusivity coefficients maps in the far-SOL in the code, in order to flatten the density and temperature profiles there. Several kinds of such obtained “shoulders” are considered. A brief comparison with SOLPS-ITER is performed on the reference case with uniform coefficients, and shows good agreement. When far-SOL transport is increased, temperatures computed on the first wall rise to 20–30 eV for ions, and to 10–20 eV for electrons. It is also found that for first wall quantities of interest in the ITER machine, the assumed level of perpendicular transport in the far-SOL is the most relevant parameter, with the location at which transport is increased being much less important.

ITER Pre‐Fusion Power Operation 1 (PFPO‐1) phase low‐power plasmas are simulated with the SOLEDGE3X‐EIRENE code, a multi‐fluid edge plasma solver that can describe plasma conditions up to the machine's first wall, including a kinetic description of neutrals. Here, SOLEDGE3X is used in 2D mean‐field mode. A throughput scan is performed, reproducing a similar scan from the ITER SOLPS simulations database. The physics assumptions for these simulations are presented, in particular the plasma–neutral interaction model where molecule charge‐exchange and ion‐molecule elastic collisions have newly been added. Challenges encountered in such up‐to‐the‐wall simulations are discussed, along with remaining open questions. Results are presented with two foci: on targets, and on the rest of the first wall. On targets, conditions transition from attached to partially detached as expected, and toroidally symmetric heat fluxes (i.e. without 3D effects) remain below 5 MW/m2, within the divertor design limits. An increase in the parallel heat flux decay length λq with throughput is also observed. On the rest of the first wall, increasing puff rate does not only spread power deposited on the target but also increases heat load on the first wall. These fluxes remain however small (<100 kW/m2). [ABSTRACT FROM AUTHOR]

The multi-component fluid closure derived by Zhdanov (2002 Transport Processes in Multicomponent Plasma (London: Taylor and Francis)) is implemented in the fluid code SOLEDGE3X-EIRENE to deal with arbitrary edge plasma composition. The closure assumes no distinction between species such as light versus heavy species separation. The work of Zhdanov is rewritten in a matricial form in order to clearly link friction forces and heat fluxes to the different species velocities and temperature gradients. [ABSTRACT FROM AUTHOR]

The impact of generalized sheath boundary conditions, taking into account the sheath perturbations driven by the Alfvén waves, on the propagation of plasma filaments at the tokamak edge is considered. A model describing the filament dynamics and containing new boundary conditions is developed. Estimates of the model control parameter—the sheath impedance—on the plasma transport are obtained. Simulations of the filament dynamics under the conditions characteristic of edge plasmas in NSTX and Alcator C‑Mod tokamaks are performed. It is shown that the sheath impedance leads to an increase in the mean values of the electrostatic potential and advection velocities of filaments, in agreement with experimental observations, requiring, at the same time, a more detailed description of the Alfvén wave reflection from the conducting surfaces of the installations. [ABSTRACT FROM AUTHOR]

The plasma-facing material is one of the most critical issues in fusion devices, which is important for the improvement of the life time of divertor targets, the control of impurities and the main plasma parameters. The high- energy plasma results in the high particle flux and heat flux deposition on the divertor target, leading to the erosion o f target plates. The re fore, the materials of target plates have an effect on the edge plasma parameters. In this work, the distributions of electron density, electron temperature, and neutral particle density with different divertor materials ( lithium, boron, carbon, and tungsten) are investigated with the help o f the three-dimensional edge transport program EMC3-EIRENE. It is found that the downstream electron density with tungsten targets is significantly lower than that of the other three low-Z target mate ria ls, and the downstream electron temperature of the tungsten targets is higher. Similar distributions of the electron density and temperature are also observed in the mid- p lane region. [ABSTRACT FROM AUTHOR]

Background The energy produced by nuclear fusion on a Tokamak device is mainly exhausted through the divertor, its service life is directly affected by the interaction between huge heat flux from core and the divertor target. The large amount of impurity produced by the heat flux hitting the target leads to the reduction of the plasma confinement performance whilst pumping is an important means to control plasma density and impurity density. Purpose This study aims to investigate the influence of pumping on the heat load of the target plate which is of reference significance for the future experiment. Methods Based on the experimental parameters of the HL-2A Tokamak, SOLPS-ITER code was used to study the effect of pumping on the heat load of the divertor target under different upstream electron densities. Analysis was performed through density scanning to find the sensitive threshold whilst and atom-molecular collision process was applied to the effect of pumping on the distribution of plasma and neutral particle parameters in divertor region at different upstream electron density. Results Density scanning results show that pumping near the detachment threshold ( T e t O S P ~5 eV) has a greater effect on the thermal load of the target plate. When the pumping rate is 12 m3·s-1, 36 m3·s-1 and 96 m3·s-1 respectively, the miss threshold and thermal load peak of outer target plate are 1.11, 1.24, 1.39 and 1.37, 1.96, 2.54 times of those without pumping respectively. Conclusion It is found that the decreases of deuterium molecular density results in the energy of the collision reaction power decreases when the upstream electron density exceeds the detachment threshold, leads to the increase of the temperature and energy flow of the plasma in the target plate.

The impact of triangularity on edge plasma transport and turbulence is addressed from full 3D turbulence simulations performed with TOKAM3X. Flux driven fluid simulations are run on analytical magnetic equilibria generated with positive and negative triangularity δ in a bottom limiter configuration. The conservation of the energy is assured by the increase of the bottom limiter radial position from δ>0to δ

To understand the beneficial effects of impurity boron (B) on the plasma, the transport of B species in both atomic and ionic states in the edge plasma and the ensuing evolution of the background plasma are simulated during real-time B powder injection by using two-dimensional fluid code SOLPS-ITER. The initial background plasma profiles for the simulation are reconstructed based on an EAST upper-single null discharge (shot No. 70601). With the injection rate of B atoms, 4.0 × 1021 atoms/s, the simulation shows that effective charge number is less than 1.1 at the inner boundary R - Rsep = −5.2 cm (background plasma density, 5.1 × 1019 m−3) at the outer mid-plane indicating that the B impurity does not have much influence on the energy confinement. The present work finds that the impurity B exhibits very specific features in the edge plasma, mostly present in the narrow region just outside the separatrix. The particular distribution pattern can reduce the heat flux greatly, leading to the divertor detachment. The divertor detachment can be achieved easily under an injection rate of B powder in a reasonable range. This study suggests that the impurity B has a promising potential application in future tokamaks.

In this work, evolution of dust grains from different materials used in fusion energy installations was studied and a model was constructed that describes the generation of dust. The model accounts for the thermochemical, electrical, and other properties of the materials of the fusion reactor wall. It was shown that the dominant process that leads to the decrease in dust grain mass is the thermal evaporation, which is determined by the saturated vapor pressure at thermal equilibrium temperature. Estimates of the lifetime of dust grains from different materials were obtained depending on plasma parameters. The findings of this work can be useful for estimates of the length of penetration of dust grains into the reactor. The difference in the dynamics of particles from light and heavy elements was shown. Out of the four elements under study (Be, Ni, Mo, and W), nickel grains demonstrate the highest penetrating ability due to their long lifetime and moderate weight. [ABSTRACT FROM AUTHOR]

A new type of scrape-off layer plasma instability associated with radial inhomogeneity of plasma density, contact of plasma with conducting material surfaces, and neutral recycling is presented. In order to elucidate the physics of this instability, the simplest possible setup is analyzed. [ABSTRACT FROM AUTHOR]

This work describes integrated numerical modelling applied to DEMO discharges with tungsten wall and divertor, using the COREDIV code, which self-consistently solves 1D radial transport equations of plasma and impurities in the core region and 2D multi-fluid transport in the SOL. The model is self-consistent with respect to both the effects of impurities on the α-power level and the interaction between seeded (Ar) and intrinsic impurities (tungsten, helium). This work is to analyse the influence of the helium and hydrogen recycling on the He confinement time in DEMO reactor. Simulations have been done for argon seeding puff 5.0 × 10211/s. It is found that recycling coefficient of helium have strong influence on the He confinement time, which increases from 7.25 s to 26.9 s going from lowest to highest recycling coefficient, but it has small influence on the effective charge state and radiation in the SOL. Alpha power decreases only by about 20%, which is the effect of main plasma dilution. [ABSTRACT FROM AUTHOR]

Highlights • The COREDIV code has been used to simulate DEMO inductive discharges with different impurity seeding (Ar, Xe) and mix gasses. • For xenon seeding for case with standard radial diffusion in SOL, such regime of operation seems not to be possible. • For Ar and combined Ar+Xe seeding it is possible to achieve H-mode plasma operation with acceptable level of the power to the target, but is upper limit on Xe concentration. • The influence of the prompt re-deposition model on the DEMO working point is relatively weak. Abstract In this paper numerical simulations with COREDIV code, which self-consistently solves radial transport equations in the core region and 2D multi-fluid transport in the SOL of EU DEMO discharges in full tungsten environment (W divertor and wall) for H-mode scenario has been performed with different impurity seeding. The optimal impurity mix for reactor would require the use of high Z impurity (low dilution, higher P α) in core and low/medium Z impurity radiating in the SOL region. In this paper we focus on investigations how the operational domain of EU DEMO can be influenced by different seeding (Ar, Xe) and mix gasses. In case of Xenon seeding, H-mode operation can NOT be achieved for standard radial diffusion (D SOL = 0.5m2/s) in the SOL. In the simulation, we have considered combined seeding of Ar (good radiator in SOL) and Xe (good radiator in core). In order to find the optimal impurity mix, simulations have been performed in such a way that for a few fixed levels of Xe puff, Ar puff has been increased from zero to the maximum value allowed by the code stability. For combined seeding: xenon + argon working point for EU DEMO can be found. [ABSTRACT FROM AUTHOR]

Abstract The long-term operations of the EAST experiments result in a substantial redeposition of impurities on the first wall, which is considered as the potential impurity source during the discharge. The transport properties of lithium impurity in the scrape-off layer (SOL) of EAST have been studied by the three-dimensional (3D) edge transport code, the EMC3-EIRENE. The neutral lithium impurities from four impurity source positions on the first wall have been used in the modelling. The detailed analysis of the relationship between the impurity source positions and the density profiles of the lithium impurity for different charge states has been carried out. It is found that the density profiles of the Li1+ ions are determined by the impurity source positions, whereas the density profiles of the Li2+ and Li3+ ions are not related to the first wall source positions. [ABSTRACT FROM AUTHOR]