EMC3-EIRENE Simulations Research Articles

Impact of spatially varying transport coefficients in EMC3-Eirene simulations of W7-X and assessment of drifts

Abstract Modelling the scrape-off layer of a stellarator is challenging due to the complex magnetic 3D geometry. The here presented study analyses simulations of the scrape-off layer (SOL) of the stellarator Wendelstein 7-X (W7-X) using spatially varying diffusion coefficients for the magnetic standard configuration, extending our previous study (Bold et al 2022 Nucl. Fusion 62 106011). Comparing the EMC3-Eirene simulations with experimental observations, an inconsistency between the strike-line width (SLW) and the upstream parameters was observed. While to match the experimental SLW a particle diffusion coefficient D ≈ 0.2 m2 s−1 is needed, D ≈ 1 m2 s−1 is needed to get experimental separatrix temperatures of 50 eV at the given experimental heating power. We asses the impact of physically motivated spatially varying transport coeffients. Agreement with experimental data can be improved, but various differences remain. We show that drifts are expected to help overcome the discrepancies and, thus, the development of SOL transport models including drifts is a necessary next step to study the SOL transport of the W7-X stellarator.

First experimental confirmation of island SOL geometry effects in a high radiation regime on W7-X

Abstract This work characterizes the detachment behavior and radiation characteristics of the low
iota configuration in the Wendelstein 7-X (W7-X) stellarator. The island scrape-off layer (SOL) of the
low iota has a poloidal mode number of 6 islands surrounding the last closed flux surface (LCFS).
The island geometry of the low iota configuration is significantly different to that of the standard
magnetic field configuration, whose detachment characteristics have already been described in previous
work[2, 3, 4]. Experimental results show that the radiation pattern in the low iota configuration
is starkly different to that of the standard magnetic field configuration, with radiation concentrated
at the island SOL O-points, rather than the X-points. Additionally, this O-point localized radiation
pattern is associated with unstable detachment, with both radiation oscillations in experiments and the
lack of a self-consistent plasma solution at high radiated power fraction in EMC3-Eirene simulations.
EMC3-Eirene simulations are used to understand the radiation distribution. It was found that the O-
point localized radiation arises first from local impurity accumulation near the parallel flow stagnation,
which is located close to the geometrical center of the island (”O-point”). The local cooling in this
region leads to plasma condensation in the islands in closest magnetic connection to the divertor target
plates. The heat source to this region of the island, which is thermally isolated from the upstream heat
source in terms of parallel transport, must arise via perpendicular heat transport. This heat source is
expected to be large for the low iota configuration due to its very small internal island field line pitch.
This work highlights the importance (complementary to previous work, e. g. [5, 6]) of the internal
island field line pitch not only on the radiation pattern, but also the detachment performance of the
island divertor.

Analysis of the neutral fluxes in the divertor region of Wendelstein 7-X under attached and detached conditions using EMC3-EIRENE

This paper analyzes the neutral fluxes in the divertor region of the W7-X standard configuration for different input powers, both under attached and detached conditions. The performed analysis is conducted through EMC3-EIRENE simulations. They show the importance of the horizontal divertor to generate neutrals, and resolve the neutral plugging in the divertor region. Simulations of detached cases show a decrease in the number of generated neutrals compared to the attached simulations, in addition to a higher fraction of the ion flux arriving on the baffles during detachment. As the ionization takes place further inside the plasma during detachment, a larger percentage of the generated neutral particles leave the divertor as neutrals. The leakage in the poloidal and toroidal direction increases, just as the fraction of collected particles at the pumping gap. The fraction of pumped particles increases with a factor two, but stays below one percent. This demonstrates that detachment with the current target geometry, although it improves the power exhaust, is not yet leading to an increased particle exhaust.

Performance of tungsten plasma facing components in the stellarator experiment W7-X: Recent results from the first OP2 campaign

The transition to reactor-relevant materials for the plasma facing components (PFCs) is an important and necessary step to provide a proof of principle that the stellarator concept can meet the requirements of a future fusion reactor by demonstrating high performance steady-state operation. As a first step to gain experience with tungsten as plasma-facing material in the Wendelstein 7-X (W7-X) stellarator, graphite tiles coated with an approximately 10 µm MedC tungsten layer (NILPRP Bucharest) were installed to complete the ECRH beam dump area in two of the five plasma vessel modules over an area of approximately one square meter each. In addition, tungsten baffle tiles are installed (40 tiles in total) with either bulk tungsten as part of NBI shine-through target or with a tungsten heavy alloy (W95-Ni3.5-Cu1.5) to replace the graphite tiles that were previously thermally overloaded. Based on an advanced diffusive field line tracing method and EMC3-Eirene simulations, the overloaded baffle tiles were redesigned by making the tiles thinner (i.e. moving the plasma-facing surface (PFS) away from the hot plasma region) and by reducing the local angle of incidence through toroidal displacement of the watershed. Significant erosion of the tungsten tiles can only be expected if sputtering by impurity ions such as carbon or oxygen ions contributes. However, the resulting central concentration of tungsten and the corresponding radiation losses are expected to be marginal. The expected deposition of carbon on the tungsten surfaces in the baffle regions mitigates further the possible tungsten enrichment in the core plasma. In OP2.1, no adverse effects of the installed tungsten PFCs on the plasma performance were observed during normal plasma operation. With the design changes made in the baffle area, the predicted heat load reductions could be experimentally confirmed.

Overview over the neutral gas pressures in Wendelstein 7-X during divertor operation under boronized wall conditions

During the first test divertor campaign of the stellarator experiment Wendelstein 7-X (Pedersen et al 2022 Nucl. Fusion 62 042022), OP1.2b, 13 neutral gas pressure gauges collected data in different locations in the plasma vessel, enabling a detailed analysis of the neutral gas pressures, the compression ratios and the particle exhaust rates via the turbomolecular pumps in the different magnetic field configurations. In Wendelstein 7-X, the edge magnetic islands are intersected by the divertor target plates and used to create a plasma-wall interface. As the number and position of the magnetic islands varies in different magnetic field configurations, the position of the strike line on the target plates and thus the neutral gas pressure in the subdivertor differs between the configurations. Neutral gas pressures on the order of few 10−4 mbar were measured in the subdivertor region. The highest neutral gas pressure of mbar was obtained in the so-called high iota configuration featuring four edge magnetic islands per cross section. The neutral particle flux through the pumping gaps into the subdivertor volume was provided by EMC3-EIRENE simulations and allowed to analyze the relation between the particle flux entering the subdivertor and the pressure distribution in the subdivertor. Finite element simulations in ANSYS provide a detailed picture of the pressure distribution in the subdivertor volume and agree with the neutral gas pressure measurements in the subdivertor in the standard configuration featuring an island chain of 5 edge magnetic islands. Surprisingly high neutral gas pressures that were not predicted by the simulation were measured in the subdivertor region away from the main strike line for discharges in the most used magnetic configuration, the standard configuration. While the pressure ratio between the two sections of the subdivertor volume, the low and high iota section is 0.06 in high iota configuration, a ratio of 2–5 was obtained in the other configurations, indicating significant particle loads and exhaust rates on the high iota section of the subdivertor in magnetic configurations with the main strike line on the low iota divertor targets.

Simulations of lithium impurity transport with a liquid lithium divertor using EMC3-EIRENE and ITCD

Simulations of lithium (Li) impurity transport behaviour with a liquid lithium divertor on the EAST tokamak have been performed using the fluid code EMC3-EIRENE and the Monte Carlo code ITCD. The interactions between particles and the wall surface, like the gyration-induced scrape-off effect, sheath acceleration, kinetic self-sputtering and reflection, are typically ignored or simplified in fluid models. In order to uncover the impacts of the scrape-off effect, sheath acceleration, kinetic self-sputtering and reflection on Li distribution, detailed analysis of the density distributions of Li1+ and Li2+ ions at the divertor region have been carried out by ITCD modelling. The newly-upgraded ITCD code can principally reproduce the distribution patterns of Li ions as well as EMC3-EIRENE under identical plasma conditions on EAST. However, due to the scrape-off effect and sheath acceleration, the densities of Li1+ and Li2+ ions simulated by ITCD are evidently reduced compared to the EMC3-EIRENE simulations. The kinetic self-sputtering by redeposited Li particles in ITCD can increase the densities of the Li ions. Moreover, the kinetic reflection database calculated by SRIM code has been employed by ITCD, which leads to an increase in the Li densities compared to the reflection approach used in EMC3-EIRENE. Overall, by adding the aforesaid physical effects in ITCD, the total number of Li ions shows an obvious reduction compared to the EMC3-EIRENE modelling.

Evaluation of edge transport and core accumulation of tungsten for CFETR with EMC3-EIRENE and STRAHL

The edge transport and core accumulation of tungsten (W) particles on China Fusion Engineering Test Reactor (CFETR) have been studied by integrated modelling consisting of EMC3-EIRENE and STRAHL codes. The edge transport and power dissipation of W particles are simulated by EMC3-EIRENE. An in–out asymmetry of W(1–28)+ ions density has been revealed in the in- and out-board divertor regions. This is mainly due to the stronger reversal flow velocity of W ions at the outboard divertor. The upward flow of W ions near the separatrix leads to a moderate W impurity leakage from the divertor on CFETR compared to the existing full W device ASDEX Upgrade due to the high plasma density near the CFETR divertor targets. Further, the density distribution and radiation loss of W ions in the core region are investigated by STRAHL code. The high charge-state W(29–60)+ and W(61–74)+ ions mainly reside in the regions of ΨN = 0.20–0.98 and 0.00–0.90 (ΨN is the normalized poloidal magnetic flux), respectively. The W induced energy dissipation in different regions is assessed according to both STRAHL and EMC3-EIRENE simulations. Particularly, the impacts of the W core radiation on the operation regime are discussed according to the H-mode threshold scaling law proposed by Martin et al (2008 J. Phys.: Conf. Ser. 123 012033) for the baseline plasma on CFETR. Further, parameter studies on the pinch velocity (v imp) and diffusion coefficient (D imp) have been performed to check their impacts on the operation regime of CFETR. A three-fold increase of v imp/D imp results in a higher W core energy loss, which can lead to the transition from H-mode back to L-mode.

EMC3–EIRENE simulations of neon impurity seeding effects on heat flux distribution on CFETR

The numerical modelling of the heat flux distribution with neon impurity seeding on China fusion engineering test reactor has been performed by the three-dimensional (3D) edge transport code EMC3–EIRENE. The maximum heat flux on divertor targets is about 18 MW m−2 without impurity seeding under the input power of 200 MW entering into the scrape-off layer. In order to mitigate the heat loads below 10 MW m−2, neon impurity seeded at different poloidal positions has been investigated to understand the properties of impurity concentration and heat load distributions for a single toroidal injection location. The majority of the studied neon injections gives rise to a toroidally asymmetric profile of heat load deposition on the in- or out-board divertor targets. The heat loads cannot be reduced below 10 MW m−2 along the whole torus for a single toroidal injection location. In order to achieve the heat load mitigation (<10 MW m−2) along the entire torus, modelling of sole and simultaneous multi-toroidal neon injections near the in- and out-board strike points has been stimulated, which indicates that the simultaneous multi-toroidal neon injections show a better heat flux mitigation on both in- and out-board divertor targets. The maximum heat flux can be reduced below 7 MW m−2 on divertor targets for the studied scenarios of the simultaneous multi-toroidal neon injections.

2D measurements of parallel counter-streaming flows in the W7-X scrape-off layer for attached and detached plasmas

Investigations of particle parallel flow velocities have been carried out for the scrape-off layer (SOL) of the Wendelstein 7-X (W7-X) stellarator, in order to gain insights on the SOL transport properties during attached and detached plasma scenarios. The experimental evidence is based on the coherence imaging spectroscopy (CIS) diagnostic, able to measure 2D impurity emission intensity and flow velocity. The impurity monitored by CIS is C2+, characterized by a line-emission intensity observed to be linearly proportional to the total plasma radiated power in both attached and detached plasmas. The related C2+ velocity shows a strong dependence on the line-averaged electron density while remaining insensitive to the input power. During attached plasmas, the velocity increases with increasing line-averaged density. The tendency reverses in the transition to and during detachment, in which the velocity decreases by at least a factor of 2. The sharp drop in velocity, together with a rise in line-emission intensity, is reliably correlated to the detachment transition and can therefore be used as one of its signatures. The impurity flow velocity appears to be well coupled with the main ions’ one, thus implying the dominant role of impurity-main ion friction in the parallelimpurity transport dynamics. In view of this SOL impurity transport regime, the CIS measurement results are here interpreted with the help of EMC3-Eirene simulations, and their major trends are already explainable with a simple 1D fluid model.

Near- and far scrape-off layer transport studies in detached, small-ELM ASDEX Upgrade discharges by means of EMC3-EIRENE

We compare various diagnostics characterizing an ASDEX Upgrade upper single-null discharge to EMC3-EIRENE simulations now including volumetric recombination and main chamber plasma-wall interaction but not yet drifts. The discharge is in a small-ELM regime and is approaching detachment due to a density ramp and nitrogen seeding. Time/ELM averaged upstream density, electron and ion temperature measurements as well as downstream ion saturation current, electron temperature and power flux measurements match the simulation within the error bars before and after reaching detachment, qualifying the code for this regime. Although the Ion Cyclotron Resonance Heating (ICRH) antenna limiters are 35 mm away from the separatrix, i.e. ten times the typical near-SOL power fall-off length 3.8 mm, they receive power fluxes of several hundreds of kW exceeding the ones to the outer target in the detached phase and strongly contribute to the particle fueling. This is explained by a substantially larger far-SOL power fall-off length 25 mm possibly caused by the small ELMs in the detached phase and the occurrence of a density shoulder.

Conduction-based model of the Scrape-Off Layer power sharing between inner and outer divertor in diverted low-density tokamak plasmas

Abstract A simple analytic model for the repartition of the Scrape-Off Layer (SOL) exhaust power between the inner and outer divertors in a diverted low-density tokamak plasma is introduced. Electron heat conduction is assumed to dominate the heat transport, from the outboard mid-plane to the divertor targets, with no heat sinks or sources in the SOL. Both divertor channels are in the attached, high-recycling regime. The model is in reasonable qualitative agreement with recent TCV experimental data and EMC3-Eirene simulations. For the Single Null divertor, it reproduces the experimentally observed increase in the power ratio between the inner and outer divertor plates of TCV, P in / P out , with increasing the outer divertor leg length or the outer target flux expansion. For the Snowflake Minus configuration, it reproduces the observed variation of P in / P out with X-point separation, although only for the reversed magnetic field direction. Within the model limitations, it provides a basic understanding of the power sharing in alternative divertor geometries.

Effect of magnetic perturbation fields on power decay length in EMC3-EIRENE simulations and comparison to experiment in ASDEX upgrade

Abstract Previous simulation results for the tokamak ASDEX Upgrade have shown that Magnetic Perturbation (MP) fields could lead to a decrease of the peak heat flux onto the divertor targets. However, experimentally, such a decrease has not been observed. To understand this effect, a well diagnosed ASDEX Upgrade discharge was simulated with the three dimensional transport code EMC3-EIRENE, where an ad hoc model was applied to mimic the screening field created by the plasma in response to the external MP field. The simulations show that only for sufficiently weak screening a broadening of the plasma wetted area occurs. The heat flux broadening effect was then studied more systematically by simulations where heat and particle diffusion coefficients were scanned for three different MP field strengths. The scans show that only for sufficiently low diffusion coefficients and high MP field strengths, a significant decrease of the peak heat flux can be achieved. Furthermore, the effect of a kink response, which is commonly observed during ELM suppression discharges in ASDEX Upgrade, was investigated with EMC3-EIRENE simulations. The simulation results indicate, that while a kink response can deform the midplane profiles strongly, the effect on the power decay length may still be small.

EMC3-EIRENE modeling of edge plasma to improve the ICRF coupling with local gas puffing in DEMO

We report the first 3D EMC3-EIRENE simulations of scrape-off layer (SOL) plasma in DEMO. EMC3-EIRENE is a 3D edge plasma fluid and neutral particle transport code. Effects of local gas puffing on the SOL density and ion cyclotron range of frequencies (ICRF) power coupling have been studied. A pure deuterium plasma is simulated and the puffing/fueling gas is deuterium gas. The gas puffing cases investigated include divertor, top, midplane and antenna gas puffing. The ICRF antenna is distributed through 360° toroidally, and poloidally located in the outer top of the vessel. The results show that toroidally distributed but poloidally localized antenna gas puffing increases the density in front of the antenna most significantly, while top or midplane gas puffing increases this antenna density to a moderate level. Influences of gas puff rate and particle transport parameters on the SOL density are investigated. The parameter scans indicate that the shift of the fast wave cut-off density position to the antenna depends almost linearly on the total gas puff rate. To achieve a significant antenna density increase, the antenna gas puff rate should be at the level of 1.8 × 1023 el s−1 (i.e. 1.0 × 1022 el s−1 per tokamak segment). As the total gas puff rate increases to 4.0 × 1023 el s−1, the evanescent layer of the fast wave (with = 2 m−1) almost vanishes. Moreover, it is found that antenna gas puffing reduces the local power flux to the main chamber wall near the gas valve due to a local decrease of the SOL temperature.

Doppler coherence imaging of divertor and SOL flows in ASDEX upgrade and Wendelstein 7-X

Measurement of ion dynamics is of great importance to the study of plasma exhaust and impurity particle flows in the scrape-off-layer (SOL) and divertor in magnetic confinement experiments. Doppler coherence imaging spectroscopy (CIS) is a passive optical diagnostic that produces 2D images of line-integrated ion flow velocity. Doppler CIS flow measurements of neutral deuterium and impurities were conducted for the first time in the divertor of the medium-sized tokamak experiment ASDEX Upgrade. A detailed sightline, emission and magnetic field analysis was undertaken to identify location and direction of flows. Under the assumption of toroidal axisymmetry, they revealed mainly parallel impurity flows in the proximity of the X-point and target plates on the order of 20 km s−1. Two Doppler CIS systems are currently set up for the optimized stellarator Wendelstein 7-X (W7-X). The main task of this diagnostic will be the measurement of impurity ion flows in one of the island divertors of W7-X. EMC3-EIRENE simulations have been carried out to estimate the flow behaviour in the W7-X SOL.

Numerical characterization of the edge transport conditions and limiter fluxes of the HIDRA stellarator

The Hybrid Illinois Device for Research and Applications (HIDRA) is a new device for education and Plasma-Material Interaction research at the University of Illinois at Urbana-Champaign. In advance of its first operational campaign, EMC3-EIRENE simulations have been run on the device. EMC3-EIRENE has been modified to calculate a per-plasma-cell relaxed Bohm-like diffusivity simultaneously with the electron temperature at each iteration. In our characterization, the electron temperature, diffusivity, heat fluxes, and particle fluxes have been obtained for varying power levels on a HIDRA magnetic grid, and scaling laws have been extracted, using constraints from previous experimental data taken when the device was operated in Germany (WEGA facility). Peak electron temperatures and heat fluxes were seen to follow a power-law dependence on the deposited radiofrequency (RF) power of type f(PRF)∝aPRFb, with typical exponents in the range of b ∼0.55 to 0.60. Higher magnetic fields have the tendency to linearize the heat flux dependence on the RF power, with exponents in the range of b ∼ 0.75. Particle fluxes are seen to saturate first, and then slightly decline for RF powers above 120 kW in the low-field case and 180 kW in the high-field case.

TCV experiments towards the development of a plasma exhaust solution

Research towards a plasma exhaust solution for a fusion power plant aims at validating edge physics models, strengthening predictive capabilities and improving the divertor configuration. The TCV tokamak is extensively used to investigate the extent that geometric configuration modifications can affect plasma exhaust performance. Recent TCV experiments continue previous detachment studies of Ohmically heated L-mode plasmas in standard single-null configurations, benefitting from a range of improved diagnostic capabilities. Studies were extended to nitrogen seeding and an entire suite of alternative magnetic configurations, including flux flaring towards the target (X divertor), increasing the outer target radius (Super-X) and movement of a secondary x-point inside the vessel (X-point target) as well as the entire range of snowflake configurations. Nitrogen seeding into a snowflake minus configuration demonstrated a regime with strong radiation in the large region between the two x-points, confirming EMC3-Eirene simulations, and opening a promising path towards highly radiating regimes with limited adverse effects on core performance.

Investigation of heat flux deposition on divertor target on the Large Helical Device with EMC3-EIRENE modelling

The measured divertor heat flux profiles are compared to the EMC3-EIRENE simulations for two different times of an LHD discharge, corresponding to higher and lower edge temperatures. The relation between the three-dimensional magnetic field structure and the heat flux distributions on the divertor has been analysed. The modelled heat flux for the lower plasma temperature case has a better agreement with the experimental result obtained by the Langmuir probes, which shows a qualitative reproduction of the experimental profile shape. However, the heat flux distribution for the high plasma temperature case shows a different behaviour between the simulation results and the experimental measurements. The detailed analysis of the heat flux distribution for the higher temperature case which has a larger discrepancy has been performed, both quantitatively and qualitatively. The radiation of the eroded impurity from divertor target plates has a minor effect on the heat flux distribution. Non-uniform cross-field transport coefficients are used in the simulations and its impact on the heat flux distributions is discussed for the case of the high plasma temperature.

Effects of outer top gas injection on ICRF coupling in ASDEX Upgrade: towards modelling of ITER gas injection

The influence of outer top gas injection on the scrape-off layer (SOL) density and ion cyclotron range of frequency (ICRF) coupling has been studied in ASDEX Upgrade (AUG) L-mode plasmas for the first time. The three-dimensional (3D) edge plasma fluid and neutral transport code EMC3-EIRENE is used to simulate the SOL plasma density, and the 3D wave code RAPLICASOL is used to compute the ICRF coupling resistance with the calculated density. Improvements have been made in the EMC3-EIRENE simulations by fitting transport parameters separately for each gas puffing case. It is found that the calculated local density profiles and coupling resistances are in good agreement with the experimental ones. The results indicate that the SOL density increase depends sensitively on the spreading of the injected outer top gas. If more gas enters into the main chamber through the paths near the top of vessel, the SOL density increase will be more toroidally uniform; if more gas chooses the paths closer to the mid-plane, then the SOL density increase will be more local and more significant. Among the various local gas puffing methods, the mid-plane gas valve close to the antenna is still the best option in terms of improving ICRF coupling. Differences between the outer top gas puffing in AUG and the outer top gas puffing in ITER are briefly summarized. Instructive suggestions for ITER and future plans for ITER gas injection simulations are discussed.

Effects of varying stochastic layer on edge plasma and impurity transport in 3D EMC3-EIRENE simulations of LHD

The transport properties of edge plasma and carbon impurity in the stochastic layer of the Large Helical Device (LHD) have been studied with the three-dimensional edge transport code EMC3-EIRENE. The recent development of computational meshes, which store the magnetic configurations, has been used to carry out the studies of the impact of the magnetic field structure on the edge plasma and impurity transport in the stochastic layer. The horizontal variation of the magnetic axis position (Rax) from 3.60m to 3.90m leads to more open flux channels connecting to the divertor and thus a stronger particle exhaust on the target, which results in a higher erosion of divertor target. Furthermore, the outboard divertor target plates suffer from the stronger erosion compared to the inboard divertor target plates for Rax=3.90m, which is in contrast to the case of Rax=3.60m. It is also found that the stronger CIV line emission is obtained at the outboard X-point region when the magnetic axis location moves outwards horizontally.

Recent progress towards a quantitative description of filamentary SOL transport

A summary of recent results on filamentary transport, mostly obtained with the ASDEX-Upgrade tokamak (AUG), is presented and discussed in an attempt to produce a coherent picture of scrape-off layer (SOL) filamentary transport. A clear correlation is found between L-mode density shoulder formation in the outer midplane and a transition between the sheath-limited and the inertial filamentary regimes. Divertor collisionality is found to be the parameter triggering the transition. A clear reduction of the ion temperature takes place in the far SOL after the transition, both for the background and the filaments. This coincides with a strong variation of the ion temperature distribution, which deviates from Gaussianity and becomes dominated by a strong peak below 5 eV. The filament transition mechanism triggered by a critical value of collisionality seems to be generally applicable to inter-ELM H-mode plasmas, although a secondary threshold related to deuterium fueling is observed. EMC3-EIRENE simulations of neutral dynamics show that an ionization front near the main chamber wall is formed after the shoulder formation. Finally, a clear increase of SOL opacity to neutrals is observed, associated with the shoulder formation. A common SOL transport framework is proposed to account for all these results, and their potential implications for future generation devices are discussed.