Divertor Surface Research Articles

Divertor geometry modeling with the SOLPS-ITER code for reactor concepts with liquid metal divertors

• The SOLPS-ITER code has been applied to the analysis of the scrape-off layer (SOL) plasma associated with the fast flow Lithium (Li) divertor design for Fusion Nuclear Science Facility (FNSF). • Two flat divertor configurations are investigated and we found that simplified open geometry presents challenges for neutral control and impurity retention, but balanced geometry with baffling appears promising. • Lithium (Li) is sourced into the balanced configuration and scanning over a large range to understand what level of Li emission is needed to influence the divertor and upstream plasma conditions. • At low sourcing level ( ϕ Li ≤ 1 × 10 23 / s ), the Li shows a minor effect on the plasma and is well confined in the Private flux region (PFR) and divertor surfaces. • At high Li sourcing level ( ϕ Li > 1 × 10 24 / s ), the Li strongly affects the upstream and divertor plasma conditions and leads to strong detachment by increasing momentum and power losses. The SOLPS-ITER code has been applied to the analysis of the scrape-off layer (SOL) plasma associated with the fast flow Lithium (Li) divertor design for Fusion Nuclear Science Facility (FNSF). Two divertor configurations are investigated to determine puff levels that provide robust operational windows that meet the FNSF design requirements on upstream density and divertor flux with a flat divertor for flowing liquid metal (LM). Neon (Ne) is introduced to control the heat flux to the divertors while Deuterium (D 2 ) puffing is performed to maintain sufficiently high upstream density. We found that a simplified open geometry presents challenges for neutral control and impurity retention, but a balanced geometry with baffling is found to have acceptable puff operational windows. Lithium (Li) is sourced into the balanced configuration uniformly along the divertor surfaces and scanning over a large range to understand what level of Li emission is needed to influence the divertor and upstream plasma conditions. At low sourcing level ( ϕ Li ≤ 1 × 10 23 / s ), the Li shows a minor effect on the plasma and is well confined in the private flux region (PFR) and divertor surfaces. At moderate sourcing level ( 1 × 10 23 < ϕ Li ≤ 1 × 10 24 / s ), the Li starts affecting the upstream and divertor plasmas, and Li radiation and electron cooling are almost linearly increasing with the Li sourcing level. At high Li sourcing level ( ϕ Li > 1 × 10 24 / s ), the Li strongly affects the upstream and divertor plasma conditions and leads to strong detachment by increasing momentum and power losses. At high sourcing level, Li reaches above the x-point that can result in a substantial Li radiation in the core and sufficient dissipation to strongly affect the divertor and upstream plasmas.

Computational investigation of incident ion angles and material erosion at rough graphite and silicon carbide divertor surfaces

We present a computational investigation of the dependence of material erosion on the incident ion angle at rough graphite and silicon carbide divertor surfaces. Ion angle distributions (IADs) for D plasmas at NSTX-U and DIII-D divertors were calculated by an equation-of-motion model that traces the ion trajectories in the sheath. Then, the effective sputtering yields and ion shadowed area fractions were calculated by a Monte Carlo micro-patterning and roughness code that applied the calculated IADs to surface topographic data that were obtained from optical confocal microscopy of rough graphite and SiC divertor surfaces from NSTX-U and DIII-D experiments. The calculations found that the effective sputtering yields, the sputtering pattern, and the shadowed area are determined by the detailed surface topology rather than the root mean square roughness RRMS, which represents deviations from a flat surface. The suppression of the effective sputtering yields for rough surfaces compared to the yield for a smooth surface was accounted for by the change of the mean local incident ion angle (LIIA) ⟨θ′⟩. The mean surface inclination angle distribution (SIAD) ⟨δ⟩ was found to be a useful parameter to estimate the LIIA from the calculated IADs. We report global empirical formulas for the mean LIIA and fraction of the area shadowed from the main ions for D plasmas for rough surfaces with B-field incident angles α = 85°–89° as a function of the mean SIAD ⟨δ⟩. We propose the use of the mean LIIA ⟨θ′⟩ to estimate the sputtering yield for rough surfaces from the angular dependence of the sputtering yield.

Control of tungsten impurity source and edge transport using different gas injection with full tungsten divertor on EAST

• With the newly installed EAST lower W divertor, W source and edge transport are experimentally investigated for different Ne and D 2 gas injection conditions. • W erosion on the whole divertor target can be well suppressed or mitigated by suitable edge Ne or D 2 injection when divertor detachment is achieved. • A W screening enhancement is obtained during the attached divertor condition due to D 2 injection. • After the divertor detachment, the core W content is observed to be kept in a high level, although W source from the divertor is strongly reduced. Tungsten (W) erosion and edge transport are investigated for EAST L-mode discharges with different gas injection. It is found that W erosion can be suppressed or mitigated by Ne or D 2 seeding when divertor detachment is achieved. Compared to edge D 2 fueling, Ne seeding from the divertor target is favorable for full detachment condition and thus W erosion suppression. Increasing the upstream plasma density by edge D 2 fueling can affect the divertor condition, which may lead to a W erosion mitigation. D 2 puffing at divertor target is less effective on increasing the upstream plasma density than OMP D 2 puffing, and thus it is less effective on reduction of W erosion rates. W gross erosion profiles with different amount of injected D 2 at the divertor are reproduced by a mixed material W erosion model, which indicates that there exist a Li-C overlayer on the W surface of EAST divertor. Ne and D 2 injection are also found to have different impact on the normalized core W density. For the attached divertor condition, divertor Ne seeding will increase W leakage, but a suitable D 2 fueling from divertor target can strengthen the edge W screening. After the divertor detachment, although the W source is dramatically reduced, the W core density is kept in a high level for the Ne seeding discharges, and even increased for the D 2 fueling discharges.

Performance of tungsten nitride compound surfaces to resist sputtering under intense irradiation in nuclear fusion reactors

The formation of tungsten nitride (WNx) films has been experimentally detected on the surface of tungsten divertors in fusion reactors after seeding cooling systems using nitrogen, and hence the sputtering resistance of the WNx film is crucial for the stable operation of the fusion reactor. By performing first-principles calculations, we investigated the sputtering resistance of W2N1, W1N1, and W2N3 surfaces, respectively. It was found that the tungsten nitride surfaces have good sputtering resistance. Especially, the sputtering resistance of the nitrogen-rich (001) surface of W1N1 is better than that of pure tungsten. However, hydrogenation at surface N atoms will weaken the chemical bonds between the N atoms and their own nearest-neighbors in the surface, facilitating the escape of the N atoms from the irradiated surface, even forming ammonia molecules. Our efforts not only reveal the performance of sputtering resistance of WNx surfaces, but also show the influence of hydrogen on the sputtering-resistant properties of WNx, which are valuable for assessing the state and stability of WNx films on divertor surfaces.

Observation of low temperature VUV tungsten emission in JET divertor plasmas

The properties of tungsten make it ideal for use as a plasma facing surface in the divertor of large plasma machines such as JET and ITER. However, the intense heat and particle fluxes that fall on the divertor surfaces lead to its release from these surfaces into the plasma and it is necessary to model its transport from the divertor and plasma edge into the plasma core. This requires measurement of spectral features over a wide temperature range. In large machines the W influx is often determined from W I line intensities, there being few measurements of discrete W lines from other low ionization stages. Their observation is highly desirable because the transport of neutral W differs markedly from that of the W ions. A change in the line-of-sight of a VUV survey spectrometer on JET to view directly into the divertor has led to the observation of numerous discrete low temperature W lines in the VUV spectral region. The spectrum of an intense influx in which W IV to W VIII features are observed has been analysed in order to provide spectral classifications so that these lines can be used for diagnostic purposes. The first observation of a VUV low temperature W magnetic dipole (M1) transition is reported for the W VIII ionization stage. The analysis shows where further line identifications are needed and that the provision of the highest quality atomic data for these ionization stages is desirable. W VI and W VII line intensity measurements are used to illustrate their use in determining the electron temperature of the emitting plasma region and the W concentration. Dependences of the W line intensities on plasma parameters shows the optimal conditions for the W release and suggests the site of its release.

Determination of the characteristic magnetic pre-sheath length at divertor surfaces using micro-engineered targets on DiMES at DIII-D

The magnetic pre-sheath (MPS) length, L MPS, is a critical parameter to define the sheath potential, which controls the ion trajectory of low-Z species (D, T, He, and C), as well as the prompt re-deposition of high-Z species. To determine L MPS, we fabricated micro-trenches (30 × 30 × 4 μm) via focused ion beam milling on a silicon surface and exposed them to L-mode deuterium plasmas in DIII-D via the divertor material evaluation system (DiMES) removable sample exposure probe. The areal distribution of impurity depositions, mainly consisting of carbon, was measured by energy-dispersive x-ray spectroscopy (EDS) to reveal the deuterium ion shadowing effect on the trench floors. The carbon deposition profiles showed that the erosion was maximized for the azimuthal direction of φ = −40° (referenced to the toroidal magnetic field direction) as well as the polar angle of θ = 80°. A Monte Carlo equation-of-motion (EOM) model, based on a collisionless MPS, was used to calculate the azimuthal and polar deuterium ion angle distributions (IADs) at the surface for a range of L MPS = k × ρ i, where ρ i is the ion gyro radius and k = 0.5–4. Then, gross erosion profiles were calculated by a Monte Carlo micro-patterning and roughness (MPR) code for ion sputtering using as input the calculated azimuthal and polar IADs for each value of k. Good agreement with the experimental C deposition profiles was obtained for the case k = 2.5–3.5. This result is consistent with a previous kinetic modeling prediction of k ∼ 3, as well as previous analytical investigations that predicted the L MPS to be several ion gyro radii. A validation of theoretical sheath models supports its applicability to ITER and pilot plant divertors to successfully predict plasma–materials interactions.

Conceptual design of reciprocating probes and material-testing manipulator for tokamak COMPASS Upgrade

Three new in-vessel manipulators are designed and built for the new COMPASS Upgrade tokamak with uniquely high vessel temperature (250–500 °C) and heat flux density (perpendicular to divertor surface q ⊥ ∼ 80 MW/m2 and q ∥ ∼ GW/m2 at separatrix), which challenges the edge plasma diagnostics. Here we show their detailed engineering designs supported by heat conduction and mechanical models. Deep reciprocation of electrostatic probes near the separatrix should be possible by optimizing older concepts in (a) the head and probe geometry, (b) strongly increasing the deceleration up to 100× gravity by springs and strengthening the manipulator mechanical structure. One reciprocates close to the region of edge plasma influx (the outer midplane), the other at the plasma sink (between the outer divertor strike point and X-point), for studying the plasma divertor (impurity-seeded) detachment and liquid metal vapor transport. Both probe heads are equipped with a set of ball-pen and Langmuir probes, measuring reliably and extremely fast (10−6 s) local (1 mm resolution) plasma potential, density, electron temperature and heat flux and even ion temperature with 10−5 s resolution. The divertor manipulator (without reciprocation) will place various material test targets at the outer divertor. Unique will be its capability to increase 15× the surface heat flux with respect to the surrounding tungsten tiles just by controllable surface inclination of the test targets. We plan to test liquid metal targets where such inclined surface was found critical to achieve the desired mode with lithium vapor shielding. Even in the conservative expected performance of COMPASS Upgrade, we predict to reach and survive the EU DEMO relevant heat fluxes.

Design and physics basis for the upcoming DIII-D SAS-VW campaign to quantify tungsten leakage and transport in a new slot divertor geometry

A set of experiments are planned to exploit the high SOL collisionality enabled by a tightly baffled slot divertor geometry to suppress tungsten leakage in DIII-D. A toroidal row of graphite tiles from the Small Angle Slot (SAS) divertor is being coated with 10–15 μm of tungsten. New spectroscopic viewing chords with in-vacuo optics will measure the W gross erosion source from the divertor surface with high spatial and temporal resolution. In parallel, the bottom of the SAS divertor is changed from a flat to a ‘V’ shape. New SOLPS-ITER/DIVIMP simulations conducted with drifts using the planned ‘V’ shape predict a substantial reduction in W sourcing and SOL accumulation in either B × ∇B direction relative to either the old SAS divertor shape or the open, lower divertor. Dedicated studies are planned to carefully characterize the level of W sourcing, leakage, and scrape-off-layer (SOL) accumulation in DIII-D over a wide range of plasma scenarios. Various actuators will be assessed for their efficacy in further reducing high-Z impurity sources and leakage from the slot divertor geometry. This coupled code-experiment validation effort will be used to stress-test physics models and build confidence in extrapolations to advanced, high-Z divertor geometries for next-step devices.

Calibration-free laser-based spectroscopic study of Sn-based alloys

The elemental quantification of liquid metal divertor (LMD) surface is important for understanding the material erosion, migration, re-deposition, and fuel retention in Plasma-Facing Components (PFCs). Currently, LMD are attractive candidates for the short- and long-term operation of fusion devices like DEMO. Liquid metals can provide self-cooling, self-replenishing plasma-facing surfaces requiring very little upkeep. In a previous work, we studied Li and LiSn layers deposited on attachment screws in the COMPASS tokamak by means of Calibration-Free Laser-Induced Breakdown Spectroscopy (CF-LIBS). Several problems were encountered related to the detection of Sn in LiSn. Thus, in the present work, we are optimizing the experimental conditions for the detection of Sn I-II and Pb I-II in Pb-containing Sn-based alloys, performing the quantification of Pb in traces and in bulk quantities using CF-LIBS approach.

Effect of single-legged coil on 3D plasma boundary corrugation in EAST

A single-legged coil behind the lower divertor and covering a 120° toroidal angle is utilized in a recent EAST discharge, for the purpose of increasing the wetted area of the divertor surface by locally modifying the magnetic field near the X-point. The plasma response, in particular, the plasma boundary surface corrugation due to the single-legged coil current, is modeled by the updated MARS-F code, by computing the plasma displacement for all important toroidal harmonics (n = 1, 2, 4 and 5) associated with the partial toroidal coverage by the coil. The plasma response produced by the single-legged coil is found to be non-local and is of the kink-peeling type. For a reference EAST plasma with a lower single-null magnetic configuration, the plasma boundary corrugation near the X-point, produced by the upper single-legged coil, is about twice as large as that produced by the lower single-legged coil, despite the proximity of the latter to the X-point.

Simulation of the SPARC plasma boundary with the UEDGE code

In this work, the UEDGE edge transport code is used to examine conditions in the SPARC divertor and edge plasma for various levels of carbon impurity and power from the core (P SOL). A double-null magnetic configuration is simulated assuming up-down symmetry in geometry and physics. The anomalous heat and particle transport coefficients are tuned to match empirical predictions for SPARC’s midplane density profiles, target plate heat flux profiles, and inner/outer divertor power sharing. Convective transport is included on the low-field side, while on the high-field side the transport is modeled as purely diffusive. Hydrogen neutrals are modeled as a fluid with inertial effects, and a carbon impurity is included using the fixed-fraction model. We find that detachment induced by impurity seeding could significantly reduce the heat flux to the divertor surfaces in the SPARC tokamak. At P SOL = 28 MW (the value predicted for SPARC’s full-power H-mode scenario) cases with both divertor legs detached were obtained with a carbon impurity fraction between 0.3%–1.4%, far below Z eff limits for SPARC. When the plasma in the outer leg is detached, the peak heat flux density perpendicular to the target plate is below 1 MW m−2, electron and ion temperatures are less than 1.5 eV, and momentum detachment is observed. However, the detachment state is found to be sensitive to the side-wall boundary conditions, the level of neutral pumping, and the target plate tilt. Finally, a broadly similar SOLPS simulation of SPARC is used to assess the appropriateness of the simpler impurity and neutral models used in UEDGE.

Simple, general, realistic, robust, analytic tokamak equilibria. Part 1. Limiter and divertor tokamaks

Tokamak equilibria have been derived that are analytic solutions to the Grad–Shafranov equation. This paper, Part 1, describes a wide range of such equilibria including smooth limiter surfaces, double- and single-null divertor surfaces, arbitrary aspect ratio, elongation, triangularity and beta. Part 2 generalizes the analysis to include edge pedestals and toroidal flow.

Simple, general, realistic, robust, analytic tokamak equilibria. Part 2. Pedestals and flow

Part 1 described a wide range of analytic tokamak equilibria modelling smooth limiter surfaces, double- and single-null divertor surfaces, arbitrary aspect ratio, elongation, triangularity and beta. Part 2 generalizes the analysis to further include edge pedestals and toroidal flow. Specifically, edge pedestals are allowed in the pressure, pressure gradient and toroidal current density. Also, an edge-localized contribution to the bootstrap current is treated. In terms of flow, analytic solutions are obtained for two cases: a $\gamma = 2$ adiabatic and a $\gamma = \infty $ incompressible energy conservation relation.

Predictive modeling of a lithium vapor box divertor in NSTX-U using SOLPS-ITER

Abstract The unmitigated heat flux in attached operation of a fusion power plant is predicted to be destructive to any solid divertor surface. Detachment, whereby the plasma pressure drops significantly before reaching the divertor target thus greatly reducing the heat flux and sputtering, will be necessary to ensure adequate lifetime of plasma facing components (PFCs). The lithium vapor box divertor aims to detach the divertor plasma via evaporating and condensing lithium surfaces. By evaporating lithium near or at the divertor plate and condensing it closer to the main chamber, a lithium vapor density gradient can be created. This density gradient ties energy losses to poloidal distance between the target and the detachment point. The radiation zone is then prevented from reaching the X-point as the lithium ionization rate decreases when the detachment front moves away from the divertor target. Here we present Scrape Off Layer Plasma Simulator (SOLPS) simulations of a lithium vapor box divertor using an NSTX-U equilibrium and PFC geometry. The parameters for the core boundary conditions, gas puff intensity, and heat and particle transport coefficients are chosen to match experimental values. Acceptable agreement with experimental Scrape-Off Layer (SOL) widths is found, indicating a reasonable choice of transport coefficients. In predictive simulations, lithium is added via evaporation at the target. Predictions for peak heat fluxes and upstream impurity concentration are given for a variety of evaporation rates. Target electron temperature is predicted to be able to be reduced to recombination levels (below 1 eV) for lithium evaporation rates of 1 ⋅ 1 0 23 Li/s, indicating detachment. Peak heat flux at the lower outer target could be reduced by as much as a factor of six while maintaining upstream lithium fractions below 2%. The prevention of lithium from reaching the midplane is shown to be due to an increase in frictional forces acting on the lithium from a deuterium gas puff. Lithium is also shown to be redeposited close to the evaporator which is favorable for initial tests and future capillary porous systems.

Observation of surface deformation of tungsten exposed to single pulsed high heat flux and magnetic field for divertor design

Tungsten (W) is one of candidate surface materials for a plasma facing component (PFC) in a magnetically confinement fusion reactor such as a divertor from the perspective of its high melting point, high thermal conductivity, low tritium retention and low sputtering yield. In the fusion reactor in operating, the divertor surface is exposed by the transient heat load generated by plasma disruption or ELM. The heat load is sufficiently high to melt the W surface. Because of the repetition of melting and solidification, it is concerned that W surface deformation occurs, the droplets scatter and its heat resistance degrades. These are critical issues for developing the fusion reactor. In previous studies, the in-situ observations of the W specimen surface during irradiating by a pulsed laser simulating the heat load were conducted using a HSV camera. However, PFCs lies under intense magnetic field in the actual fusion-reactor condition. Magnetic field affects the motion of fluid by the effect of generated Lorentz force. Therefore, this study aims to clarify the melting and solidification behavior of W under the intense magnetic field. In order to simulate the heat load, the Nd/YAG laser, which has almost flatten spatial profile and rectangular temporal profile, was irradiated on a W specimen. The power density of the laser was 3.12 GW/m2 and 3.9 GW/m2 and the pulse width was 3 [ms]. Helmholtz-like electromagnetic coils were applied for generating pulsed magnetic field. The magnetic density was also changed in the range from 0 to 4 [T] in this study. From observation of the light emitted from the heated W surface using a high-speed video camera and measurement of irradiation traces, the behavior of molten W and the effect of the magnetic field will be reported and discussed.

Real-time estimation and control of divertor surface heat flux on the DIII-D tokamak

Future tokamaks will require robust technologies for the mitigation of heat exhaust onto the plasma-facing components. As a first step towards this development, a system has been developed at DIII-D that estimates and controls in real-time the heat flux to the PFCs. Real-time estimation of the peak power flux from this model-based approach is validated with off-line infra-red measurements for various DIII-D plasma discharges. A nonlinear free-boundary simulation code (GSevolve) is used for simulating the closed loop response and for the off-line determination of the control parameters. The implementation and first experimental results of the application during the DIII-D plasma campaign are reported.

Beryllium-driven structural evolution at the divertor surface

Erosion of the beryllium first wall material in tokamak reactors has been shown to result in transport and deposition on the tungsten divertor. Experimental studies of beryllium implantation in tungsten indicate that mixed W–Be intermetallic deposits can form, which have lower melting temperatures than tungsten and can trap tritium at higher rates. To better understand the formation and growth rate of these intermetallics, cumulative molecular dynamics (MD) simulations of both high and low energy beryllium deposition in tungsten were performed. In both cases, a W–Be mixed material layer (MML) emerged at the surface within several nanoseconds, either through energetic implantation or a thermally-activated exchange mechanism, respectively. While some ordering of the material into intermetallics occurred, fully ordered structures did not emerge from the deposition simulations. Targeted MD simulations of the MML to further study the rate of Be diffusion and intermetallic growth rates indicate that for both cases, the gradual re-structuring of the material into an ordered intermetallic layer is beyond accessible MD time scales(⩽1 μs). However, the rapid formation of the MML within nanoseconds indicates that beryllium deposition can influence other plasma species interactions at the surface and begin to alter the tungsten material properties. Therefore, beryllium deposition on the divertor surface, even in small amounts, is likely to cause significant changes in plasma-surface interactions and will need to be considered in future studies.

Experimental verification of ion impact angle distribution at divertor surfaces using micro-engineered targets on DiMES at DIII-D

• The first detailed experimental verification of polar ion impact angle distribution (IAD). • Specially fabricated micro-trenches were exposed to L-mode D plasmas at DIII-D DiMES. • Calculated polar and azimuthal IADs were combined with a Monte-Carlo MPR code. • MPR revealed the areal erosion yield on the trench floor due to D sputtering. • Good agreement between MPR erosion and EDS net C deposition profiles was seen. • Such agreement strongly verifies the IADs, and the modeling of the Chodura sheath. We report the first detailed experimental verification of the polar deuterium ion impact angle distribution (IAD) on the DIII-D divertor surface in L-mode plasmas using micro-engineered trenches in samples mounted on the DiMES probe. These trenches were fabricated via focused ion beam (FIB) milling of a silicon surface partially coated with aluminum. The sample surfaces were exposed to eight repeat L-mode deuterium discharges (30 s total exposure time). The samples were examined by scanning electron microscopy (SEM), which revealed changes on the trench floor due to material deposition and evidence for shadowing of the incident deuterium ions by the trench walls. The areal distribution of carbon and aluminum deposition was measured by energy-dispersive X-ray spectroscopy (EDS). One-dimensional profiles of this deposition are in agreement with net erosion profiles calculated from a Monte Carlo micro-patterning and roughness (MPR) code for ion sputtering using as input the polar and azimuthal deuterium IADs reported previously (Chrobak et al., Nucl. Fusion 58, 106019 (2018)). The deposition profiles verified the characteristic shape of the polar IADs, which have a broad maximum from 79° to 86°, over the experimental range of 68°–83°, where 0° is the surface normal direction.

Thermal loads in gaps between ITER divertor monoblocks: First lessons learnt from WEST

In the WEST tokamak, ITER-like divertor targets consisting of tungsten monoblocks bonded via an OHFC-Cu compliance layer to CuCrZr cooling tubes were exposed to plasma during the 2018 experimental campaign in which modest heating power was available. Up to 2.5 MW/m2 divertor surface heat flux was attained. Inspection of the components after the campaign revealed a wide variety of damage at both leading and trailing monoblock edges, and at the optical hot spots which are the projections along magnetic field lines of the toroidal gaps between monoblocks onto the poloidal leading edges. Cracking, deformation, and melting occurred. Consideration of the large body of past work on high heat flux testing, combined with the expected loading conditions in WEST, suggests that fractures form during the first transient events such as disruptions. Deformation occurs under subsequent exposure to steady state heat loads. Nearly identical damage was observed on reciprocating probes made of W-La(10%) alloy under measured irradiation conditions, lending credence to this hypothesis.

EDGE2D-EIRENE modeling of the impact of wall materials on core edge, scrape-off layer and divertor parameters

A series of EDGE2D-EIRENE cases was run in different wall materials of JET L-mode plasma with the input power close to the H-mode power threshold in ITER-like wall (ILW), with tungsten (W) covering the divertor, and beryllium (Be)—main chamber surfaces. Cases with all carbon (C) and all W walls were run at the same input power as in ILW cases. Contrary to expectations based on the idea that deeper penetration of neutrals re-emitted from surfaces with higher atomic mass into the plasma core raises core density, EDGE2D-EIRENE cases had the highest density in the C wall case. This is explained via the effect of the ‘power deficit’ in the divertor, as recycling neutrals deposit a fraction of their power into walls, thereby cooling the divertor plasma. This effect is stronger in C compared to W divertor surfaces. The plasma cooling in the divertor increases plasma density and neutral recycling, with more neutrals penetrating into the core. Metal wall (ILW and W wall) EDGE2D-EIRENE cases were found to have larger radial electric field (E r) both inside and outside of the separatrix, in line with the mechanism of the E × B shear turbulence suppression and experimental observations of lower H-mode power threshold in metal wall machines. The originally expected trend for higher core densities in metal wall machines, based on the effect of deeper penetration of neutrals re-emitted from higher atomic mass surfaces, was recovered in much lower density EDGE2D-EIRENE cases, where the effect of the power deficit in the divertor was less important. Subtle effects of interplay among mechanisms involving neutral circulation in the divertor and in the main chamber wall are considered to explain differences between ILW and W wall cases.