Edge Plasma Transport Research Articles

Modeling of plasma facing component erosion, impurity migration, dust transport and melting processes at JET-ILW

An overview of the modeling approaches, validation methods and recent main results of analysis and modeling activities related to the plasma-surface interaction (PSI) in JET-ILW experiments, including the recent H/D/T campaigns, is presented in this paper. Code applications to JET experiments improve general erosion/migration/retention prediction capabilities as well as various physics extensions, for instance a treatment of dust particles transport and a detailed description of melting and splashing of PFC induced by transient events at JET. 2D plasma edge transport codes like the SOLPS-ITER code as well as PSI codes are key to realistic description of relevant physical processes in power and particle exhaust. Validation of the PSI and edge transport models across JET experiments considering various effects (isotope effects, first wall geometry, including detailed 3D shaping of plasma-facing components, self-sputtering, thermo-forces, physical and chemically assisted physical sputtering formation of W and Be hydrides) is very important for predictive simulations of W and Be erosion and migration in ITER as well as for increasing quantitative credibility of the models. JET also presents a perfect test-bed for the investigation and modeling of melt material dynamics and its splashing and droplet ejection mechanisms. We attribute the second group of processes rather to transient events as for the steady state and, thus, treat those as independent additions outside the interplay with the first group.

Simulation study of the influence of drifts on the upstream and target heat flux width under different B T directions

The heat flux width (λq ) is a key parameter determining the heat load at divertor targets. In recent years, drifts have been found to play a remarkable role in the edge plasma transport, while its influence on λq has not been well understood. Investigations of the influence of drifts on λq , systematic simulations using the SOLPS-ITER code are performed in this work. The statistics of the simulation results show that the drift under favorable/unfavorable B T tends to increase the λq in the outer/inner side and decrease the λq in the other side, which is consistent with the experiment observations. At the upstream and the target, the mechanisms of the influence of the drifts on λq are different. The upstream λq (λq ,u) is directly affected by the drift-induced convective heat flux, while λq at the target (λq ,t) is indirectly influenced through heat conduction (in the high-recycling regime) and the sheath (in the detached regime) due to the change of plasma parameters there. Furthermore, the synergetic effect of geometry and drift under favorable B T leads to an anomalously large λq ,t in the inner side at high density.

Modeling of tungsten impurity transport and distribution in EAST based on multi-fluid and kinetic Monte Carlo simulations

In recent years, the materials of plasma facing components, such as divertor target plates, domes, and outer walls of tokamaks, such as ASDEX Upgrade, WEST, JET, EAST, and ITER, have been changed from carbon to tungsten because of its lower erosion and tritium retention rates. Impurities are produced by interactions between the plasma and the first wall. This study provides an investigation to simulate the transport and distribution of tungsten impurities in the edge plasma on EAST. The 2D multi-fluid edge plasma transport code SOLPS-ITER and 2D kinetic Monte Carlo impurity transport code DIVIMP were used in the simulations. The multi-fluid model in SOLPS-ITER and the kinetic Monte Carlo model in DIVIMP were employed to treat tungsten impurity ions. The 2D density contour distributions in the computational region and the 1D density radial profiles at the inner and outer midplanes of tungsten impurity particles with ionization states (W0–W+74) and the total tungsten particles with all charge states were obtained. With the heating power 1.5 MW and the line-averaged plasma density 2 × 1019 m−3, the results from SOLPS-ITER and DIVIMP show that the maximum density of tungsten ion with single ionization state is about 1014 m−3 and the total density of tungsten impurities with all charge states is about 1015 m−3 at the core boundary. To the best of our knowledge, the simulation results from SOLPS-ITER and DIVIMP are compared for the first time to benchmark SOLPS-ITER with the multi-fluid mode and DIVIMP with the kinetic model for tungsten impurity transport. The density distributions of tungsten impurities with different ionization states from SOLPS-ITER and DIVIMP are highly similar, and good agreement can be found under similar conditions involved in the calculation. From the comparison benchmark between SOLPS-ITER and DIVIMP for tungsten impurity transport, it can be concluded that the impurity transport approximation used by DIVIMP is good.

Modeling transient edge plasma transport with dynamic recycling

The work presents numerical simulation studies of the role that dynamic plasma recycling on the main wall and divertor target surfaces plays in transient edge plasma transport phenomena, such as edge localized modes (ELMs). The studies are performed by coupling the edge plasma transport code UEDGE [Rognlien et al., J. Nucl. Mater. 196–198, 347 (1992)] and the wall reaction–diffusion transport code FACE [Smirnov et al., Fusion Sci. Technol. 71, 75 (2017)]. The two-dimensional, time-dependent, two-way coupling of the codes, in a realistic tokamak geometry, is accomplished using the Integrated Plasma Simulator framework [Elwasif et al., in 18th Euromicro Conference on Parallel, Distributed and Network-Based Processing (PDP 2010), Pisa, Italy (IEEE, 2010), pp. 419–427] for all modeled material plasma boundaries. The simulations show that dynamic plasma recycling has substantially different characteristics on the main wall and on the divertor plates. It is demonstrated that during an ELM cycle the outer wall can dynamically absorb and release a number of particles comparable to that expelled by the ELM from the core plasma, by far exceeding the dynamic retention capacity of the divertor surfaces. The resulting evolution of the edge and divertor plasma conditions during an ELM cycle is analyzed.

Toward plasma drifts in EMC3: Implementation of gradient, divergence, and particle tracing schemes

AbstractThis paper presents a first implementation of gradient, divergence, and particle tracing schemes for the EMC3 code, a stochastic 3D plasma fluid code widely employed for edge plasma and impurity transport modeling in tokamaks and stellarators. These schemes are essential to accommodate plasma drift computations, which are currently absent in the code. Plasma drifts have been recognized to significantly influence transport of particles and energy, and their inclusion in future code upgrades will substantially enhance the code's predictive capabilities. For gradient and divergence calculations, we introduce a second‐order least‐squares gradient scheme. We confirm the second‐order convergence properties and assess the accuracy of several analytical test cases in the presence of synthetic noise. In the second part of this paper, we employ the validated gradient scheme in a fourth‐order Runge–Kutta particle tracing scheme to trace a particle through a generic drift velocity field. The impact of synthetic noise on the scheme's performance is investigated by evaluating various error metrics. We find that the implemented schemes function as intended and exhibit sufficient accuracy to enable drift computations.

First results of EAST edge modeling with SOLPS‐ITER 3.2.0 on Extended Grid

AbstractThe processes in far scrape‐off layer (SOL) and the plasma interaction with the first‐wall (FW) elements may notably affect the tokamak discharge, since they define fuel and impurity recycling, material erosion and redeposition, wall surface heating, etc. For a long time, the far SOL description in most plasma edge transport codes was insufficient or absent at all, and so particle and heat fluxes onto the FW (except divertor plates) were out of consideration. Recently some codes, for example, SOLEDGE and SOLPS‐ITER are upgraded allowing for the extension of the computational grid up to real walls and for corresponding account of the vacuum vessel shape and all in‐vessel elements. The new release of SOLPS‐ITER (the version 3.2.0) required a development of a new code data structure and new approach to numerical approximation of fluid equations compatible with unstructured non‐orthogonal computational grid. Intensive testing of the new code in different conditions is still required. In the present contribution, such a testing is performed for the EAST disconnected double null (DDN) L‐mode discharge. For the first time, the SOLPS‐ITER 3.2.0 modeling results with drifts and currents turned on are presented, and a comparison to former SOLPS‐ITER version (3.0.8) is performed. The far SOL transport and its effects on the discharge performance are studied by comparing the computational results obtained on several meshes which differ by their width in equatorial midplane. A single null (SN) one (with mesh width limited by distance to the secondary separatrix) was examined versus two DDN meshes (one with actual and one with artificially extended targets to make mesh wider) and a true unstructured (the widest) mesh. The notable difference in results obtained on different meshes appears in those places where plasma density does not vanishes at the computational domain boundaries. For the cases on true unstructured (the widest) mesh, the particle and heat fluxes onto central column, limiters, far SOL part of targets, dome umbrella and other EAST far SOL in‐vessel structures are calculated for the first time by SOLPS‐ITER, allowing assessment of the plasma interaction with those surfaces.

Numerical study of divertor detachment in TCV H-mode scenarios

The effect of divertor closure and nitrogen seeding on the detachment process has been studied by performing 2D numerical simulations of tokamak à configuration variable (TCV) H-mode divertor scenarios with the SOLEDGE3X-EIRENE edge plasma transport code. The simulations reveal that, in the cases with only deuterium gas fuelling, detachment occurs at a similar level of divertor neutral pressure (0.76 Pa), despite the difference in divertor closure achieved by changing the length of the outer baffle. Nitrogen reduces the target temperature with little effect on the upstream density and momentum loss but drops the upstream pressure, leading to a decrease in the target particle flux and divertor neutral pressure. Furthermore, when the radiation front starts to move up from the outer target, the peak parallel heat flux level at the outer target remains approximately the same ( MW m−2), regardless of whether it is a deuterium fuelling scan, a nitrogen seeding scan at a fixed fuelling rate, or a change in the length of the outer baffle. An empirical partial detachment qualifier calibrated on AUG experimental data was compared with the TCV simulations. The results show good agreement in detachment state prediction, indicating the potential of this detachment qualifier to be applied in devices of different sizes, emphasizing the combined influence of input power entering the divertor, neutral pressure, and the concentration and species of impurity in achieving divertor detachment.

Numerical studies of plasma edge in W7-X with 3D FINDIF code

Abstract Modelling of the plasma transport for inherently three-dimensional (3D) problems as in stellarators requires dedicated complex codes. FINDIF is a 3D multifluid plasma edge transport code that has been previously successfully used for the analysis of energy transport in the TEXTOR-DED tokamak [1], where 3D perturbations led to an ergodic structure of field lines in the plasma edge. The ongoing efforts to apply it meaningfully to Wendelstein 7-X (W7-X) plasma problems resulted in advancements in the main model and accompanying tools for mesh generation and post-processing. In order to verify the applicability of the code and to compare with the reported simulation (EMC3-EIRENE) and experimental (OP1.1) results, a series of simulations for varying plasma density, temperature and anomalous transport coefficients as well as for fixed input power were performed. The connection length pattern of FINDIF traced magnetic field lines on the limiter was reproduced and its impact on heat loads was confirmed. An increase in the peak heat load on the limiter with a rise in plasma density, temperature and anomalous plasma transport coefficients was observed. The decay lengths of density, electron temperature and heat flux did not change with density, and were decreasing with temperature and increasing with anomalous plasma transport coefficient, which was compared to the simple scrape-off layer (SOL) model.

Spontaneous formation of a transport barrier in helium plasma in a tokamak with circular configuration

We report on the first experimental observation of a spontaneously formed transport barrier in the tokamak with a circular configuration in helium plasmas. There was no external polarization of the plasma by electric field or other technique to form the barrier as it is typically used in tokamaks with circular plasma. In general, the transport barriers play an important role in plasma confinement especially in tokamaks with divertor configuration. In our experiments, we clearly observe distinct characteristics of a transport barrier, including a steep gradient of the electron temperature and an enhanced radial electric field along with the change in the plasma potential, floating potential, and electron temperature fluctuation. The electron temperature and the plasma potential are obtained by a combination of the ball-pen and Langmuir probe measurements with high temporal resolution on a shot-to-shot basis. This first experimental observation of the spontaneously formed transport barrier might bring new possibilities to obtain a fusion-relevant study of the edge plasma parameters and transport in helium plasmas even on small tokamaks.

Electromagnetic turbulence simulation of tokamak edge plasma dynamics and divertor heat load during thermal quench

The edge plasma turbulence and transport dynamics, as well as the divertor power loads during the thermal-quench phase of tokamak disruptions, are numerically investigated with BOUT++’s flux-driven six-field electromagnetic turbulence model. Here, transient yet intense particle and energy sources are applied at the pedestal top to mimic the plasma power drive at the edge induced by a core thermal collapse, which flattens the core temperature profile. Interesting features, such as surging of divertor heat load (up to 50 times) and broadening of heat-flux width (up to four times) on the outer-divertor target plate, are observed in the simulation, in qualitative agreement with experimental observations. The dramatic changes in divertor heat load and width are due to the enhanced plasma turbulence activities inside the separatrix. Two cross-field transport mechanisms, namely, the E × B turbulent convection and the stochastic parallel advection/conduction, are identified to play important roles in this process. First, an elevated edge pressure gradient drives instabilities and subsequent turbulence in the entire pedestal region. The enhanced turbulence not only transports particles and energy radially across the separatrix via the E × B convection, which causes the initial divertor heat-load burst, but it also induces amplified magnetic fluctuation . Once themagnetic fluctuation is large enough to break the magnetic flux surface, magnetic flutter effect provides an additional radial transport channel. In the late stage of our simulation, reaches to 10−4 level that completely breaks magnetic flux surfaces such that stochastic field lines are directly connecting pedestal top plasma to the divertor target plates or first wall, further contributing to the divertor heat-flux width broadening.

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

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

Modeling of plasma beta effects on the island divertor transport in the standard configuration of W7-X

The influence of plasma beta effects on the edge plasma transport in the Wendelstein 7-X standard configuration is studied systematically by using EMC3-EIRENE combined with a 3D equilibrium code named HINT. The magnetic topology changes induced by plasma beta effects are significantly reflected in plasma transport behaviors and heat flux patterns on divertor targets. After validating the modeling strategy by comparisons with experimental data, the extended simulations for high performance plasmas show that the threshold separatrix density for accessing the power detachment is reduced in higher beta plasmas. Compared with the vacuum field case, the impurity radiation distributions with finite beta effects are modified in the magnetic island region. The divertor heat flux is distributed more evenly along the toroidal direction on the strike line at the vertical target. The strike line on the horizontal target moves towards the pumping gap with an increase in the plasma beta. In addition, the different pressure profiles with the same central beta also result in a modified heat flux pattern on the divertor targets.

Simulation of plasma transport in the linear plasma device MPS-LD by SOLPS-ITER

• The direct imposing particle source is the suitable method to model the plasma source for the SOLPS-ITER modeling of plasma transport in linear device. • The MPS-LD can achieve plasma detachment by increasing the equivalent H + injection rate (source strength). • The combination of high heating power and short device length significantly raise the plasma parameters near the target. Linear plasma devices are the ideal devices to simulate the divertor conditions of tokamak experimentally. The Multiple Plasma Simulation Linear Device (MPS-LD) is under construction at Dalian University of Technology with a focus of studying the plasma-material interactions (PMIs) and edge plasma transport. Before the experiment, numerical modeling plays a crucial role in predicting the main plasma parameters and discharge performance. To this end, simulations are carried out by using scrape-off layer plasma simulation code SOLPS-ITER for the designing studies of MPS-LD in the present work. The effects of particle source simulation method, heating power, and device chamber length on the plasma are investigated systemically. The simulation demonstrates that the direct imposing particle source in the source region is suitable for handling the plasma source. The radiation region, recombination front region and recombination region are formed during plasma transport. Enhancing the plasma source strength can promote the achievement of plasma detachment. The helicon source with a 6kW radio frequency power source can only obtain the maximum electron density of 5 × 10 18 m -3 and electron temperature of 2 eV in the vicinity of the target. Raising the heating power and shortening the distance from the plasma source to the target can significantly increase the plasma parameters at the target.

Simulation study of particle transport by weakly coherent mode in the Alcator C-Mod tokamak

A simulation study has been conducted of the physical mechanisms behind the weakly coherent mode (WCM) and its produced particle transport in the I-mode edge plasmas by using the BOUT++ code. The WCM is identified in our simulations by its poloidal and radial distributions as well as its frequency and wavenumber spectra. Its produced radial particle flux is calculated and compared with the experimental value. The good agreement indicates that the WCM is an important particle transport channel in the I-mode pedestal. It is found that the WCM can transport particles across the strong outer shear layer of the E r well established in the formation of I-mode, based on which a possible explanation is provided why I-mode does not feature a density pedestal. The key point lies in the change of the cross-phase between the electric potential and density fluctuations induced by the E × B Doppler shift. In the strong shear layer, although the electric potential fluctuation is significantly suppressed, the cross-phase is close to π/2, resulting in a strong drive of the density fluctuation and particle transport. To identify the physical nature of the WCM, a linear dispersion relation for drift Alfvén modes is derived in the slab geometry. A drift Alfvén wave instability is found to have similar dependence to the simulated linear instability behind the WCM on the resistivity and the parallel electron pressure gradient and thermal force terms in the parallel Ohm’s law.

Toward holistic understanding of the ITER-like resonant magnetic perturbation (RMP) ELM control on KSTAR

KSTAR has clarified a set of unresolved 3D physics issues utilizing the ITER-like in-vessel, three-row, resonant magnetic perturbation (RMP) configurations. Since RMP-driven, edge-localized-modes (ELMs)-crash control elevates the divertor heat flux peak through its impact on edge plasma parameters and transport, a series of intentionally misaligned RMP configurations (IMCs) have been explored to investigate the relationship between RMP ELM control and divertor heat fluxes, while searching for an ideal IMC that could be favorable in both aspects. First of all, the contrasting influence of kink vs anti-kink phasing on the ELM-crash suppression has been articulated, demonstrating the synergistic benefit of ‘kink’ phasing on ELM-crash-suppression. On the other hand, the three-row IMC in the anti-kink phasing becomes more insensitive to the ELM-crashes at the sub-marginal level of RMP, consistent with theory. Meanwhile, the divertor ‘wetted’ area of ELM-crash-suppression gets narrower than that of ELM-crash-mitigation, suggesting that ELM-crash-mitigation remains advantageous over ELM-crash-suppression in terms of time-averaged divertor thermal loading. In comparison, based on a set of two-row IMCs, no evidence of divertor heat flux broadening was found during ELM-crash-suppression, supporting a hypothesis that the dispersal of the divertor heat flux in three-row IMCs cannot be driven by helically structured two-row RMPs alone. Among ITER-like three-rows, lower two-row RMPs have been found to be much more effective in suppressing the ELM-crashes than upper two-row RMPs. Although it is quite preliminary, the up/down asymmetric dependence of RMP coupling may be generically attributed to lower-single-null plasmas. Such a holistic understanding of RMP-driven, ELM-crash-control in KSTAR is expected not only to elucidate various subtle points in the vicinity of ELM-crash-suppression, but also to clarify the relevant divertor thermal loading issues for ITER and beyond.

Confinement improvement during detached phase with RMP application in deuterium plasmas of LHD

In order to explore the compatibility of good core plasma performance with divertor heat load mitigation, the interaction between cold edge plasma and core plasma transport, including the edge transport barrier (ETB), has been analysed in the divertor detachment discharges of deuterium plasmas in LHD with resonant magnetic perturbation (RMP) field application. The RMP application introduces a widened edge stochastic layer and sharp boundary in the magnetic field structure between the confinement region and the edge stochastic layer. The widened edge stochastic layer enhances impurity radiation and provides stable detachment operation as compared with the case without RMP. It is found that ETB is formed at the confinement boundary at the onset of detachment transition. However, as the detachment deepens, the resistive pressure gradient-driven MHD mode is excited, which degrades the ETB. At the same time, however, the core transport decreases to keep global plasma stored energy (W p) unchanged, showing clear core-edge coupling. After a gradual increase of density fluctuation during the MHD activity, a spontaneous increase of W p and the recovery of ETB are observed while the detachment is maintained. Then, the coherent MHD mode ceases and ELM-like bursts appear. In the improved mode, impurity decontamination occurs, and the divertor heat load increases slightly. Key controlling physical processes in the interplay between core and cold edge plasma are discussed. A comparison between deuterium and hydrogen plasmas shows that hydrogen plasmas exhibit similar features to the deuterium ones in terms of density and magnetic fluctuations, impurity decontamination towards higher confinement, etc. But most of the features are modest in the hydrogen plasmas and thus no clear confinement mode transition with clear ETB formation is defined. Better global confinement is obtained in the deuterium plasmas than the hydrogen ones at a higher radiation level.

Line identification of extreme ultraviolet spectra of Mo V to Mo XVIII in EAST tokamak

The presence of high-<i>Z</i> impurities in magnetically confined fusion devices has different influences on the confinement property of the plasma due to the high cooling rate of high-<i>Z</i> impurities. The first wall of EAST is equipped with molybdenum tiles, molybdenum particles sputtered from inevitable plasma-wall interaction enter into the plasma and become high-<i>Z</i> impurity. In this paper, four fast-time-response extreme ultraviolet (EUV) spectrometers, a system which is upgraded in the EAST 2021 campaign, are used to monitor the line emission from impurity ions in the 5–500 Å wavelength range simultaneously. The in-situ wavelength calibration is carried out accurately using several well-known emission lines of low- and medium-Z impurity ions. The observed spectral lines are carefully identified based on the National Institute of Standards Technology (NIST) database, previously published experimental data and the time evolution of the normalized line intensity of emission lines from impurity ions. At the lower electron temperature (<i>T</i><sub>e0</sub> = 1.5 keV), the EUV spectra emitted from molybdenum ions in the range of 5–485 Å are systematically identified in EAST discharges accompanied with spontaneous sputtering events. As a result, two unresolved transition arrays of molybdenum spectra composed of Mo<sup>19+</sup>-Mo<sup>24+</sup> (Mo XX-Mo XXV) and Mo<sup>16+</sup>-Mo<sup>29+</sup> (Mo XVII-Mo XXX) are observed in the ranges of 15–30 Å and 65–95 Å. In addition, several spectral lines of lower molybdenum ions of Mo<sup>4+</sup>-Mo<sup>17+</sup> (Mo V-Mo XVIII) in the ranges of 27–60 Å and 120–485 Å are observed and identified on EAST for the first time, including a few strong and isolated forbidden and resonant lines, e.g. Mo XII at 329.414 Å, 336.639 Å and 381.125 Å, Mo XIII at 340.909 Å and 352.994 Å, Mo XIV at 373.647 Å and 423.576 Å, Mo XV at 50.448 Å, 57.927 Å and 58.832 Å. Six spectral lines are newly observed in the range of 27–32 Å, i.e. (27.21 ± 0.01) Å, (27.37 ± 0.01) Å, (28.99 ± 0.01) Å, (30.81 ± 0.01) Å, (31.54 ± 0.01) Å and (31.83 ± 0.01) Å, which may be Mo XV-Mo XVIII spectral lines. As a result, twelve strong and isolated spectral lines are chosen in routine observation for impurity transport physical study. The identification of these spectral lines not only enriches the molybdenum atom database, but also provides a solid experimental data base for magnetically confined devices to study the behavior and transport in core and edge plasmas of high-Z impurity.

Influence of the X-point location on edge plasma transport in the J-TEXT tokamak with a high-field-side single-null divertor

High-density experiments in the high-field-side mid-plane single-null divertor configuration have been performed for the first time on J-TEXT. The experiments show an increase in the highest central channel line-averaged density from to , while the X-point moves away from the target by increasing the divertor coil current. The corresponding Greenwald fraction rises from 0.50 to 0.79. For the impurity transport, the density normalized radiation intensity (absolute extreme ultraviolet and soft x-ray) of the central channel density decreased significantly (>50%) with an increase in the plasma density. To better understand the underlying physics mechanisms, the 3D edge Monte Carlo code coupled with EIRENE (EMC3- EIRENE) has been implemented for the first time on J-TEXT. The simulation results show good agreement with the experimental findings. As the X-point moves away from the target, the divertor power decay length drops and the scrape-off layer impurity screening effect is enhanced.

Progress in edge plasma turbulence modelling—hierarchy of models from 2D transport application to 3D fluid simulations in realistic tokamak geometry

This contribution presents the recent effort at CEA and French federation for fusion to simulate edge plasma transport with the new code SOLEDGE3X. The latter can be used both as a 2D transport code or as a 3D turbulence code. It makes possible simulating edge plasma up to the first wall including the complex wall geometry. It also includes neutral recycling and impurity sputtering, seeding and transport. In order to improve turbulence description in transport simulation, a reduced model for turbulence intensity prediction has been derived and implemented, based on ‘k-epsilon’ like models from the neutral fluid community. Applications to a JET L-mode detached plasma and to a WEST plasma are used as illustration of the code abilities.

Bolometer tomography on Wendelstein 7-X for study of radiation asymmetry

The algorithm for bolometer tomography at Wendelstein 7-X (W7-X) has been recently improved using a novel regularization functional, based on relative gradient smoothing of the sought radiation profile. It has been validated using radiation patterns provided by 3D modeling under real plasma conditions as phantoms and then applied to bolometer measurements performed during the first divertor operation phase of W7-X. The following results are presented: (1) edge-localized 2D radiation patterns with clearly resolved magnetic island radiation structures, (2) an up-down asymmetry in the impurity radiation that is not captured by the 3D edge plasma transport modeling, (3) reversal of the asymmetry with reversed magnetic field direction. Further analysis reveals a poloidal variation of the emissivity in the outer confined plasma region with a field-direction dependent asymmetry, also supported by the soft x-ray measurements. This asymmetry is considered to be related to asymmetric impurity distributions, driven by the pronounced ion-impurity friction force at the plasma edge where the collisionality of the W7-X plasma is sufficiently high to develop impurity asymmetry as predicted by neoclassical theory of parallel impurity transport.