High Density H-mode Research Articles

Modelling study of divertor W leakage for different divertor conditions with Ne seeding in EAST tokamak

Abstract Sequential SOLPS-ITER and DIVIMP simulations are carried out for tungsten (W) edge transport during neon (Ne) injection in EAST, elucidating the pathway of W impurity leakage from the divertor to the core plasma under different dissipative divertor conditions. Divertor conditions from low-recycling to detachment are generated by modulating the Ne injection rate in the SOLPS simulation. With the drift velocities incorporated in DIVIMP, W transport under the drift effect is investigated. For low-density L-mode plasma conditions with favorable Bt direction, the majority of W source originates from the outer divertor target and leaks upstream through the near-SOL EB drift reversed region at both inner and outer divertors. The EB drift is demonstrated to have a significant effect on W leakage and the effect diminishes with the increase of Ne injection rate. Compared to the D2 injection cases, Ne injection helps to achieve a strong dissipative divertor condition at a much lower plasma density, resulting in a higher W leakage ability due to the smaller friction with the background plasma. For Ne injection cases under high-density H-mode conditions, W leaks mostly through the outer divertor region while the inner divertor is fully detached. Compared to the L-mode cases, the shorter power decay lengths in the scrap-off layer (SOL) of the H-mode cases result in smaller parallel EB drift velocities especially in the middle and far SOL region. With the increase of the Ne injection rate, the decrease of the EB drift in the middle SOL leads to a wider near-separatrix W leakage path. Therefore, a worse divertor W screening is expected, which is consistent with the previous published experimental observations [1].

The effect of plasma shaping on high density H-mode SOL profiles and fluctuations in TCV

The impact of plasma shaping on the properties of high density H-mode scrape-off layer (SOL) profiles and transport at the outer midplane has been investigated on Tokamakà configuration variable. The experimental dataset has been acquired by evolving the upper triangularity while keeping the other parameters constant. The scan comprises δup values between 0.0 and 0.6, excluding negative triangularity scenarios. Within this study, a transition from type-I edge localised modes to the quasi-continuous exhaust regime takes place from low to high δup . The modification of the upstream SOL profiles has been assessed, in terms of separatrix quantities, within the αt turbulence control parameter theoretical framework (Eich et al 2020 Nucl. Fusion 60 056016). The target parallel heat load and the upstream near-SOL density profiles have been shown to broaden significantly for increasing αt . Correspondingly, in the far SOL a density shoulder formation is observed when moving from low to high δup . These behaviours have been correlated with an enhancement of the SOL fluctuation level, as registered by wall-mounted Langmuir probes as well as the thermal helium beam diagnostic. Specifically, both the background and the filamentary-induced fluctuating parts of the first wall ion saturation current signal are larger at higher δup , with filaments being ejected more frequently into the SOL. Comparison of two pulses at the extremes of the δup scan range, but with otherwise same input parameters, shows that the midplane neutral pressure does not change much during the H-mode phase of the discharge. This indicates that indirect effects of the change in geometry, linked to first wall recycling sources, should not play a significant role. The total core radiation increases at high δup , on account of a stronger plasma–wall interaction and resulting larger carbon impurity intake from the first wall. This is likely associated to the enhanced first wall fluctuations, as well as a smaller outer gap and the close-to-double-null magnetic topology at high shaping.

First results of high density H-mode operation in metal-wall EAST tokamak

In metal-wall EAST superconducting tokamak, H-mode operation with plasma density close to the Greenwald density limit nGW has been achieved with radio frequency and NBI heating for the first time. Both gas puffing from horizontal plane and HFS pellet fueling were used for density ramp-up during the experiment. The confinement of H-mode gradually deteriorates with plasma density increasing. And the H-L transition can be observed after heating power dropping or pellet injection. In the density range of (0.6–1)×nGW, the divertor detachment occurs and causes an obvious confinement degradation. The maximum accessible density ne, max in H-mode phase deviates from the Greenwald scaling. It has been observed that the fraction ne, max/nGW is almost independent of the total heating power, but a high heating power is helpful to extend the duration of high density H-mode. And ne,max/nGW has an increase relation with the safety factor q95 varied by changing plasma current. Besides, it is also found that the discharges with low plasma current have a higher ne, max/nGW than those with high plasma current. All these findings will provide a good reference to the high density plasma operation for future metal-wall fusion devices.

The radiated power limit in impurity seeded JET-ILW plasmas

The total radiated fraction is examined in high density H-mode plasmas (Greenwald fraction of about 85%) in JET by the variation of the auxiliary heating power of P heat =14 MW-29 MW. An achieved radiation fraction of about 75% at most has been observed in JET-ILW, which is less than the highest achievable (≈90%) fraction in JET-C during the high radiative power scenarios with N 2 seeding. It is shown that the maximal achievable total radiation fraction averaged over ELM cycles has a strong dependence on the radiation efficiency of the ELM energy, θ rad : γ rad , t o t a l max = 1 - f ELM × Δ W ELM P heat 1 - θ rad . About 50% and 16% of the ELM induced diamagnetic energy drop (ΔW ELM ) radiates during the ELM in JET-C and JET-ILW, respectively, which corresponds to the maximum total radiated powers of γ rad , J E T - C max = 0.87 and γ rad , J E T - I L W max = 0.77 . These values of the maximum of the radiative power fractions are in good agreement with γ rad max experimentally observed in JET-C (90%) and JET-ILW (75%). PACS numbers: 52.55.Rk, 52.25.Vy, 52.40.Hf, 52.55.Fa

Dependence of scrape-off layer profiles and turbulence on gas fuelling in high density H-mode regimes in TCV

A set of high density, highly shaped H-mode discharges has been performed in the TCV tokamak with the aim of assessing the effects of increasing divertor neutral recycling on the properties of upstream inter-ELM scrape-off layer (SOL) profiles and transport. An increase of divertor neutral pressure has been correlated with the evolution of separatrix properties and turbulence level. The latter has been quantified by means of the α t parameter introduced in (Eich 2020 Nucl. Fusion 60 056016), describing the contribution of resistive-interchange turbulence in the SOL relative to drift wave transport. The analysis reveals a general broadening of the upstream SOL profiles as α t increases, with the SOL power width measured by the vertical IR thermography system increasing significantly. In a similar way, the upstream density profile widens in the near SOL, whereas in the far SOL a density shoulder is observed to progressively form and increase in amplitude. This behaviour is associated with an enhancement of far SOL turbulent transport in the form of blob-filaments travelling radially faster across the far SOL and becoming bigger at higher α t. The detected filaments, evaluated from the fast reciprocating probe at the outer midplane, are determined to mostly belong to the resistive ballooning and resistive X-point regimes.

Impact of divertor configuration on recycling neutral fluxes for ITER-like wall in JET H-mode plasmas

In recent years it has been well known that in JET, with the ITER-like wall, the performance of high-power H-mode plasmas depends strongly on the magnetic topology of the divertor. This is generally attributed to the effect of the magnetic field shaping on the neutral flux transport and pumping, which—in high density H-mode plasmas—determine the pedestal properties and the global confinement. In this work we have analysed the spatial distribution and the dynamic behaviour of the Dα-emission for different magnetic configurations. Experimental observations indicate that for certain configurations, the surface temperature and the Dα—emission anomalously increase on top of the inner divertor, which points to thermal outgassing there. This is also the region where most beryllium co-deposits accumulate and most deuterium becomes trapped. The overheating at this region far from the strike point (SP) is observed to happen in magnetic configurations with reduced distance between the divertor material surface and the separatrix (clearance). The neutral flux that appears at the upper inner divertor during a few milliseconds after the ELM-crash, is more than an order of magnitude larger than the gas puffing rate and dominates over all other regions. Finally, a preliminary study describes how this thermal fuel outgassing from the co-deposited layers could be used intentionally as a wall-conditioning in JET technique with plasmas that focus their particle and heat flux there. This could be used as a complementary wall isotope control technique and more specifically for tritium recovery from the upper inner divertor where most fuel-trapping beryllium co-deposits accumulate in JET ITER-like wall.

Peculiarity of highly radiating multi-impurity seeded H-mode plasmas on JET with ITER-like wall

On JET with fully metallic first wall, highly radiative conditions with N2, Ne and Ar as well as their mixture as radiators are approached in high density H-mode plasmas. The confinement increases from H98(y,2) = 0.65 in unseeded pulses with γrad ∼ 30% to a value of H98(y,2) = 0.75 at γrad ∼ 50% with N2 injection. A degradation of the pedestal profile is compensated by steeper core ne and Te profiles. Further increase of γrad with increase of the N2 seeding rate leads to a moderate confinement degradation which can be avoided by applying of combined impurity seeding. The enhancement of the plasma performance for the radiation fractions beyond 55% with the maximum value of H98(y,2) = 0.78 is reached with combined N2 + Ne and N2 + Ar impurity injections. The observed intense, strongly localized radiation at the X-point inside the confined plasma in the scenarios with the highest radiated power fraction is interlinked to complete divertor detachment. In the JET-ILW, the X-point radiation is stable, reproducible and reversible.

Influence of high magnetic field on access to stationary H-modes and pedestal characteristics in Alcator C-Mod

Recent Alcator C-Mod experiments have explored access to and characteristics of H-modes at magnetic fields approaching 8 T, the highest field achieved to date in a diverted tokamak. The H-modes originated from L-mode densities ranging from to , allowing insight into the density dependence of the H-mode power threshold at high magnetic field. This dependence is compared to predictions from the ITPA scaling law ([]), finding that the law is approximately accurate at 7.8 T. However, the law underpredicted the high density H-mode threshold at lower magnetic field in previous C-Mod experiments ([]), suggesting that the overall dependence of the threshold on magnetic field is weaker than predicted by the scaling law. The threshold data at 7.8 T also indicates that the onset of a low density branch at this magnetic field on C-Mod occurs below , which is lower than predicted by an existing model for low density branch onset. The H-modes achieved steady-state densities ranging from to , and higher transient densities, and had values of q95 from 3.3 to 6.0. This parameter range allowed the achievement of all three types of H-mode routinely observed at lower magnetic field on C-Mod: the stationary, ELM-suppressed Enhanced Dα (EDA) regime, seen at high densities and high values of q95; the nonstationary ELM-free regime, seen at lower densities and values of q95; and the ELMy regime, seen at low density, moderate q95, and specialized plasma shape. The parameter space in which these regimes occur at 7.8 T is consistent with lower magnetic field experience. Pressure pedestal height at 7.8 T is compared to EPED [, ] predictions, and a scaling law for EDA density pedestal height developed between 4.5 T and 6.0 T is updated to include fields from 2.7 T to 7.8 T. Overall, this analysis increases confidence in the use of low magnetic field experience to predict some elements of high magnetic field tokamak behavior.

On the E-H transition in inductively coupled radio frequency oxygen plasmas: II. Electronegativity and the impact on particle kinetics

In this series of two papers we present results about the E-H transition of an inductively coupled oxygen discharge driven at radio frequency (13.56 MHz) for different total gas pressures. The mode transition from the low density E-mode to the high density H-mode is studied using comprehensive plasma diagnostics. The measured electron density can be used to distinguish between the different operation modes. This paper focuses on the determination of the negative atomic ion density and the electronegativity by two experimental methods and global rate equation calculation. As a result, the electronegativity significantly decreases over two orders of magnitude from about 25 in the E-mode to about 0.1 in the H-mode. The temporal behavior of the electronegativity in pulsed ICP shows that the negative atomic ion density reaches a steady state after 10 ms. Negative atomic ions are mainly produced by the dissociative attachment with the molecular ground state. The ion–ion recombination with the positive molecular ions and the collisional detachment with the singlet molecular metastables contribute significantly to the loss of the negative atomic ions.

Density limits investigation and high density operation in EAST tokamak

Increasing the density in a tokamak is limited by the so-called density limit, which is generally performed as an appearance of disruption causing loss of plasma confinement, or a degradation of high confinement mode which could further lead to a H → L transition. The L-mode and H-mode density limit has been investigated in EAST tokamak. Experimental results suggest that density limits could be triggered by either edge cooling or excessive central radiation. The L-mode density limit disruption is generally triggered by edge cooling, which leads to the current profile shrinkage and then destabilizes a 2/1 tearing mode, ultimately resulting in a disruption. The L-mode density limit scaling agrees well with the Greenwald limit in EAST. The observed H-mode density limit in EAST is an operational-space limit with a value of . High density H-mode heated by neutral beam injection (NBI) and lower hybrid current drive (LHCD) are analyzed, respectively. The constancy of the edge density gradients in H-mode indicates a critical limit caused perhaps by e.g. ballooning induced transport. The maximum density is accessed at the H → L transition which is generally caused by the excessive core radiation due to high Z impurities (Fe, Cu). Operating at a high density () is favorable for suppressing the beam shine through NBI. High density H-mode up to could be sustained by 2 MW 4.6 GHz LHCD alone, and its current drive efficiency is studied. Statistics show that good control of impurities and recycling facilitate high density operation. With careful control of these factors, high density up to 0.93 stable H-mode operation was carried out heated by 1.7 MW LHCD and 1.9 MW ion cyclotron resonance heating with supersonic molecular beam injection fueling.

Improved confinement in high-density H-modes via modification of the plasma boundary with lower hybrid wavesa)

Injecting Lower Hybrid Range of Frequency (LHRF) waves into Alcator C-Mod's high-density H-mode plasmas has led to enhanced global energy confinement by increasing pedestal temperature and pressure gradients, decreasing the separatrix density, modifying the pedestal radial electric field and rotation, and decreasing edge turbulence. These experiments indicate that edge LHRF can be used as an actuator to increase energy confinement via modification of boundary quantities. H98-factor increases of up to ∼35% (e.g., H98 from 0.75 to 1.0) are seen when moderate amounts of LH power (PLH/Ptot ∼ 0.15) are applied to H-modes of densities n¯e ∼ 3 × 1020 m−3, corresponding to values ∼0.5 of the Greenwald density. However, the magnitude of the improvement is reduced if the confinement quality of the target H-mode plasma is already good (i.e., H98target ∼ 1). Ray-tracing modeling and accessibility calculations for the LH waves indicate that they do not penetrate to the core. The LHRF power appears to be deposited in p...

Comparative analysis of core heat transport of JET high density H-mode plasmas in carbon wall and ITER-like wall

A consistent deterioration of global confinement in H-mode experiments has been observed in JET [] following the replacement of all carbon plasma facing components (PFCs) with an all metal (‘ITER-like’) wall (ILW). This has been correlated to the observed degradation of the pedestal confinement, as lower electron temperature (Te) values are routinely measured at the top of the edge barrier region. A comparative investigation of core heat transport in JET-ILW and JET-CW (carbon wall) discharges has been performed, to assess whether core confinement has also been affected by the wall change.The results presented here have been obtained by analysing a set of discharges consisting of high density JET-ILW H-mode plasmas and comparing them against their counterpart discharges in JET-CW having similar global operational parameters. The set contains 10 baseline () discharge-pairs with 2.7 T toroidal magnetic field, 2.5 MA plasma current, and 14 to 17 MW of neutral beam injection (NBI) heating.Based on a Te profile analysis using high resolution Thomson scattering (HRTS) data, the Te profile peaking (i.e. core Te (ρ = 0.3) / edge Te (ρ = 0.7)) is found to be similar, and weakly dependent on edge Te, for both JET-ILW and JET-CW discharges. When ILW discharges are seeded with N2, core and edge Te both increase to maintain a similar peaking factor.The change in core confinement is addressed with interpretative TRANSP simulations. It is found that JET-ILW H-mode plasmas have higher NBI power deposition to electrons and lower NBI power deposition to ions as compared to the JET-CW counterparts. This is an effect of the lower electron temperature at the top of the pedestal. As a result, the core electron energy confinement time is reduced in JET-ILW discharges, but the core ion energy confinement time is not decreased. Overall, the core energy confinement is found to be the same in the JET-ILW discharges compared to the JET-CW counterparts.

Detailed OEDGE modeling of core-pedestal fueling in DIII-D

Abstract The OEDGE code is used to model core fueling for attached L-mode plasmas and between edge localized modes (ELMs) for attached H-mode plasmas in DIII-D. Empirical plasma reconstruction has been used to determine the plasma conditions in these discharges. EIRENE is used to model the hydrogen recycling. Divertor recycling accounts for 65–100% of the core fueling. The fraction of the total divertor target flux ionized inside the separatrix ranges from 5% to 20%. The fraction of total wall flux ionized inside the separatrix ranges from 20% to 50%. Neutrals originating from wall regions closer to the separatrix are more likely to ionize in the confined plasma. Ionization in the confined plasma is concentrated below the midplane with peaks in the poloidal profiles just above the X -point. Radial core ionization in high density H-mode is peaked strongly near the separatrix.

Investigations of impurity seeding and radiation control for long-pulse and high-density H-mode plasmas in JT-60U

Rseduction of heat loading appropriate for the plasma facing components such as the divertor is crucial for a fusion reactor. Power handling by large radiative power loss has been studied in long pulse ELMy H-mode discharges on JT-60U (τd = 30–35 s). Case 1 is argon (Ar) seeding into standard ELMy H-mode plasmas, where large radiation loss in the confined region of the main plasma caused a change in ELM characteristics from Type-I to Type-III. Case 2 is a combination of Ar and nitrogen (Ne) gas seeding into Type-I ELMy H-mode plasmas with an internal transport barrier (ITB). For case 1, large radiation loss both from the main plasma and from the divertor was produced, and operation of Type-III ELMs was preferable to a reduction in ELM energy loss fraction (WELM/Wdia) to 0.15%. Both transient and steady-state heat loadings were reduced. Relatively good energy confinement (HH98y2 = 0.87 − 0.75) with large frad (Prad/Pabs > 0.8) and divertor plasma detachment was sustained continuously for 13.5 s. For case 2, with reduced Ar seeding to the main plasma and increased divertor radiation with Ne seeding, the ELMy H-mode plasma with an ITB had better energy confinement (HH98y2 = 0.95 − 0.8), which was sustained continuously for 12 s. The radiated power was increased primarily in the divertor (), which was produced both by seeded Ne ions and by carbon influx due to transient (ELM) and steady-state heat loadings in the attached divertor. Reduction in the heat loading was not enough, thus enhancement of the radiated power in the divertor will be necessary for the formation of the divertor detachment.

Edge plasma physics issues for the Fusion Advanced Studies Torus (FAST) in reactor relevant conditions

To have reliable predictions of the thermal loads on the divertor plates and of the core plasma purity in the proposed Fusion Advanced Studies Torus (FAST) tokamak, numerical self-consistent simulations have been made for the H-mode and steady-state scenario by using the 2D multi-fluid code COREDIV. In the simulations full W plasma facing components, foreseen for basic operation, as well as liquid lithium divertor targets have been considered. Impurity seeding, for reducing divertor heat loads, was allowed. The overall picture shows that, marginally in the intermediate and, necessarily in the high density H-mode scenarios (average density 〈n e〉 = 2 × 1020 m−3 and 5 × 1020 m−3, respectively), impurity seeding should be foreseen with W as target material; however, only a small amount of Ar (0.03% atomic concentration), not affecting the core purity, is sufficient to maintain the divertor peak loads below 18 MW m−2, which represents the safety limit for the W monoblock technology, presently accepted for the ITER divertor tiles. Li always needs additional impurities for decreasing divertor heat loads. At low plasma densities (but ⩾1.3 × 1020 m−3), typical of steady-state regimes, W alone is effective in dissipating the input power by radiative losses, without excessive core contamination. Impurity seeding would lead to excessive W sputtering by Ar and too high Z eff. The impact of unmitigated giant (1.5 MJ) type I edge localized modes on the W divertor targets was also analysed: the resulting maximum energy load of 1 MJ m−2, larger than the tolerable one by a factor of 3, seems not difficult to recover by foreseen mitigation tools.

Dynamic and static deuterium inventory in ASDEX Upgrade with tungsten first wall

Hydrogen retention in the divertor tokamak ASDEX Upgrade is studied by surface analysis and gas balances methods. Comparing carbon and tungsten plasma facing components (PFCs), the deuterium content of deposits at the divertor plates has dropped by a factor of 13. With tungsten PFCs only 0.7% of the puffed hydrogen is retained, including a significant amount deeply implanted in the tungsten coatings at the outer divertor. Gas balances for ITER relevant high density H-mode discharges lead to a hydrogen retention averaged over a discharge of 8.2 ± 3.3% for tungsten PFCs and 23 ± 7% for carbon PFCs. For tungsten PFCs wall saturation is observed, i.e. only 1.5 ± 3.5% of the puffed gas is retained after reaching steady state conditions. Within the error bars all hydrogen outgasses within 15 min.

On impurity handling in high performance stellarator/heliotron plasmas

The Large Helical Device (LHD) and Wendelstein 7-X (W7-X, under construction) are experiments specially designed to demonstrate long-pulse (quasi steady state) operation, which is an intrinsic property of stellarators and heliotrons. Significant progress has been made in establishing high performance plasmas. A crucial point is the increasing impurity confinement at high density observed at several machines (TJ-II, W7-AS, LHD) which can lead to impurity accumulation and early pulse termination by radiation collapse. In addition, theoretical predictions for non-axisymmetric configurations predict the absence of impurity screening by ion temperature gradients in standard ion-root plasmas. Nevertheless, scenarios were found where impurity accumulation was successfully avoided in LHD and W7-AS due to the onset of friction forces in the (high density and low temperature) scrape-off-layer (SOL), the generation of magnetic islands at the plasma boundary and to a certain degree also by edge localized modes, flushing out impurities and reducing the net impurity influx into the core. In both the W7-AS high density H-mode regime and in the case of application of sufficient electron cyclotron radiation heating power a reduction in impurity core confinement was observed. The exploration of such purification mechanisms is a demanding task for successful steady-state operation. Impurity transport at the plasma edge/SOL was identified to play a major role for the global impurity behaviour in addition to the core confinement.

Scrape-off layer transport and deposition studies in DIII-D

Trace C13H4 injection experiments into the main scrape-off layer (SOL) of low density L-mode and high-density H-mode plasmas have been performed in the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] to mimic the transport and deposition of carbon arising from a main chamber sputtering source. These experiments indicated entrainment of the injected carbon in plasma flow in the main SOL, and transport toward the inner divertor. Ex situ surface analysis showed enhanced C13 surface concentration at the corner formed by the divertor floor and the angled target plate of the inner divertor in L-mode; in H-mode high surface concentration was found both at the corner and along the surface bounding the private flux region inboard of the outer strike point. Interpretative modeling was made consistent with these experimental results by imposing a parallel carbon ion flow in the main SOL toward the inner target, and a radial pinch toward the separatrix. Predictive modeling carried out to better understand the underlying plasma transport processes suggests that the deuterium flow in the main SOL is related to the degree of detachment of the inner divertor leg. These simulations show that carbon ions are entrained with the deuteron flow in the main SOL via frictional coupling, but higher charge-state carbon ions may be suspended upstream of the inner divertor X-point region due to balance of the friction force and the ion temperature gradient force.

A new edge-based energy scaling law for W7-AS and its implications for boundary-island divertor operation

Within this paper it is established that the total plasma energy W for Wendelstein 7-AS plasmas is uniquely related to the density near the separatrix nes, for an extremely wide range of conditions, in the fashion W=5.47nes0.7[kJ,1019m-3]. This expression holds for peaked core density profiles (Normal Confinement, NC) as well as those with an edge transport barrier (High Density H-mode, HDH). It encompasses absorbed powers P=0.26–2.4MW over a line-averaged density range n¯e=1–40×1019m-3 (Pecrh=0.26–1.5MW: n¯e=1–10×1019m-3; Pnbi=0.7–2.4MW: n¯e=7–40×1019m-3), including even attached/detached conditions at the target plates. A consequence of this scaling for a boundary-island divertor – which is an inherent feature of the Wendelstein stellarator line, requiring a large nes to effect detachment – is that detachment on W7-AS can be expected and is only found for high-confinement situations, i.e. the HDH-mode.

Impurity Dynamics in Nonaxisymmetric Plasmas

The stellarator impurity transport code has been developed to describe the evolution of the impurity concentration and convective and diffusive fluxes of different charge states in time and space for given background plasma profiles in nonaxisymmetric devices. An extended model of neoclassical transport coefficients obtained by benchmarking of various methods has been employed for calculation of the radial electric field and for description of impurity ions. Calculations were performed mainly for light impurity species for background plasma profiles in high-density long-pulse Large Helical Device (LHD) discharges with and without an externally induced island at the edge and for W7-AS discharges with low and high confinement. It is shown that in the frame of neoclassical theory, the forces due to the radial electric field, the temperature gradient (convective terms), and the density gradient (diffusive term) mainly determine the impurity dynamics and eventually, together with atomic processes, the radial distribution of each ionization stage. Calculations show that in LHD discharges a different sign of the electric field (measured in experiment) within the island ensures the effective pumping of impurities within the island and their screening from penetration into the bulk plasma. It is shown that in the frame of purely neoclassical theory, the retention of impurities at the plasma edge, seen in the high-density H-mode of operation in W7-AS, cannot be explained.