Search

In attaining the cold plasma condition at the divertor target via gas puffing and/or nitrogen seeding, a large variation in the pressure-momentum and cooling losses has, for the first time, been demonstrated on JET with the ITER-like wall, leveraging novel target electron temperature measurements and JET’s unique capability for disentangling the main-ion-driven net dissipation effects from impurity radiation. These observations provide key insights into the dependence of volumetric losses on the dominant dissipation mechanisms, whether due to neutral-plasma interaction processes, impurity radiation induced detachment, and their interplay.While numerical code studies indicate that both volumetric cooling and pressure-momentum losses in the scrape-off layer (SOL) are essential for detachment to occur, experimental confirmation has been largely lacking, in large part due to diagnostic limitations. In the present assessment of L-mode discharges in the diagnostically optimized outer horizontal target configuration, the onset of pressure-momentum losses in nitrogen seeding scans in low recycling conditions is shown to occur at relatively high target electron temperature of 10 eV compared to pure deuterium pressure scans, leading to a more pronounced reduction in the target heat flux, and demonstrating for the first time the importance of momentum losses in impurity radiation dominated dissipative divertors in the absence of strong neutral-plasma interactions. Conversely, a relatively constant target heat flux response is observed above Te,t = 5 eV in pure deuterium density ramps, where the decrease in Te,t is balanced by a rise in target pressure prior to the ion flux rollover, coinciding with the pressure-momentum loss onset. While in practice the relative strength of each dissipation channel will be driven by integrated scenario requirements and core plasma performance constraints, this work demonstrates that there is no universal form of the fmom-loss and fcooling volumetric loss factors and their correlation. Moreover, the results confirm the theoretical and code predictions that both volumetric loss processes are required for detachment, while also resolving an apparent, but counter-intuitive, finding of the code results that volumetric power loss is insensitive to the presence of impurities, demonstrating that this is not the case experimentally. Further interpretive modelling is required to elucidate the dominant momentum and cooling loss channels in the observed trends.

Future fusion reactors require a safe, steady-state divertor operation. With deep divertor detachment, which is typically induced by impurity seeding, the radiation concentrates in a small region at the X-point or on closed flux surfaces above the X-point. This so-called X-point radiator (XPR) moves further inside the confined region with increasing seeding and the location can be actively controlled.At AUG, the parameter space for operation with an XPR was significantly extended, using active feedback on the XPR location. The XPR is observed in nearly the whole operational space of AUG in the high-densities or high collisionality regime. ELM suppression is consistently observed in all cases where the XPR was moved to a significant height above the X-point.Direct measurements of density and temperature from the region around the XPR using the new divertor Thomson scattering system at AUG indicate that the temperature at the location of the XPR remains high (>30eV) and only the region towards the X-point cools down further. In this cold XPR core, the temperature reduces to about 1eV.An XPR is also observed in TCV by the injection of nitrogen as extrinsic impurity. This highlights that the wall material (W for AUG, C for TCV) or machine size does not play a significant role for the existence of the regime. However, the scenario appears to be less stable in TCV. First experiments show the necessity of an active control for the XPR: Depending on the wall conditions and the nitrogen wall storage, the required nitrogen seeding level to achieve an XPR changes.Both, the low temperatures measured radially outside of the radiation zone at AUG, and the lower stability of the XPR regime at TCV with the presence of carbon are consistent with the predictions of a one-dimensional model of the XPR. However, the model would predict the development of the cold XPR core, and significant radiation at the X-point might already exist before reaching this cold temperature solution.

Measurements of the ion currents to and plasma conditions at the low-field side (LFS) divertor target plate in low-confinement mode plasmas in the JET ITER-like wall materials configuration show that the core plasma density required to detach the LFS divertor plasma is independent of the hydrogenic species protium, deuterium and tritium, and a 40 %/60 % deuterium–tritium mixture. This observation applies to a divertor plasma configuration with the LFS strike line connected to the horizontal part of the LFS divertor chosen because of its superior diagnostic coverage. The finding is independent of the operational status of the JET cryogenic pump. The electron temperature (Te) at the LFS strike line was markedly reduced from 25 eV to 5 eV over a narrow range of increasing core plasma density, and observed to be between 2 eV and 3 eV at the onset of detachment. The electron density (ne) peaks across the LFS plasma when Te at the target plate is 1 eV, and spatially moves to the X-point for higher core densities. The density limit was found approximately 20 % higher in protium than in tritium and deuterium–tritium 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 Pheat = 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 N2 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, θrad: γrad,totalmax=1-fELM×ΔWELMPheat1-θrad. About 50 % and 16 % of the ELM induced diamagnetic energy drop (ΔWELM) radiates during the ELM in JET-C and JET-ILW, respectively, which corresponds to the maximum total radiated powers of γrad,JET-Cmax=0.87 and γrad,JET-ILWmax=0.77. These values of the maximum of the radiative power fractions are in good agreement with γradmax experimentally observed in JET-C (90 %) and JET-ILW (75 %).

While current tokamak experiments are beginning to use real-time feedback control systems to manage the plasma exhaust, future tokamaks still require validation of theoretical models used to predict the threshold impurity concentration required to sufficiently reduce the power and particle fluxes to the divertor. This work exploits new spectroscopic measurements of the divertor nitrogen concentration, cN, in partially detached N2-seeded H-mode plasmas on ASDEX Upgrade (AUG) and JET with the ITER-Like Wall (JET-ILW) to test the parameter dependencies of the power flowing to the outer divertor, Pdiv,outer, and the separatrix electron density, ne,sep. A least-squares regression of the AUG measurements demonstrates that the threshold cN required for detachment scales as cN∝Pdiv,outer1.19±0.32ne,sep-2.77±0.36. This scaling of ne,sep is also consistent with the measurements from JET which, at constant Pdiv,outer, show cN∝ne,sep-2.43±0.27. The dependencies of Pdiv,outer and ne,sep is demonstrated over at least a factor of two change in both parameters and indicates a stronger dependence on ne,sep in comparison to the Lengyel model, which could be due to the assumption in this model that the heat flux channel width is independent of density. This first assessment of detachment scaling with impurity seeding highlights the need for further analysis of the systematic uncertainties of the measurement and more consistent scenarios from more tokamaks to investigate the machine size scaling.

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. Keywords: ASDEX upgrade, Magnetic perturbations, Divertor heat flux, MSC: 00-01, 99-00

Pairs of ASDEX Upgrade L-mode discharges with the toroidal magnetic field, BT, in the forward and reverse directions have been used to study the impact of neoclassical drifts on the divertor plasma conditions and detachment. The evolution of the peak heat flux and the total power loads onto both the outer and the inner targets depends significantly on the toroidal field direction: increasing the core plasma density affects mainly the heat loads in the BT 0 (favourable). Ion saturation current measurements show similar trends to those of the IR heat flux data. These discrepancies are not only caused by drifts but also by different levels of radiated power in the core, thus the power across the separatrix, Psep. Tomographic reconstructions show that Psep is not constant within the entire dataset. Finally, at Ip=0.8MA, a significant reduction of the peak heat flux is observed at both targets for both field directions. On the other hand, at Ip=0.6MA, a reduction of the peak heat flux is only observed for BT

SOLPS 5.0 simulations assuming convection-dominated radial ion transport show qualitative and quantitative agreement with measurements of detached high-field side (HFS) divertor conditions for unseeded low-power L-mode plasmas in ASDEX Upgrade, while simultaneously maintaining a reasonable match to the density and temperature measurements at the low-field side (LFS) and HFS midplanes within the scatter of the data. The decreased diffusive transport from the high-density region in the HFS divertor volume across the separatrix into the core enables spatial extension of the high-density front to above the X-point in agreement with spectroscopic measurements. The suppressed ion fuelling from this region into the core plasma allows increasing the neutral D2 fuelling to experimental levels, leading to agreement with the measured sub-divertor neutral fluxes within 30%. Detachment of the HFS divertor is observed as a significant decrease of the target ion flux and as reproduction of the characteristic roll-over behaviour of the integrated target ion current at increasing upstream density.

ASDEX Upgrade is planning the installation of a pair of in-vessel coils in order to study alternative divertor configurations, like an X-divertor (XD) or a Snowflake (SF) in a machine with a high heating power compared to its size. Two designs for the technical realization of these coils are currently proposed, one including z-shaped joints and one with a single spiral conductor. The magnetic fields produced by these conductors intrinsically have a non-axisymmetric component. In particular in the XD configuration, where very shallow field line incidence angles occur, these 3D fields have a strong impact on the power distribution on the outer divertor plate. The 2D and 3D plasma transport is studied by the SOLPS and initial EMC3-EIRENE code simulations, respectively, in order to support the decision making processes of the engineering design. Although several aspects still require further attention, this paper gives a status report. The strong toroidal asymmetries observed in the power deposition pattern for the spiral conductor seem to be strongly reduced, when highly dissipative divertor conditions are reached. The asymmetry for the z-joint design is substantially smaller than the spiral conductor even for low-dissipative conditions. Keywords: Power exhaust, Alternative divertor, ASDEX upgrade, EMC3-EIRENE, Tokamak, X-divertor,

The impact of the finite grid size in SOLPS-ITER edge plasma simulations is assessed for JET H-mode discharges with a metal wall. For a semi-horizontal divertor configuration it is shown that the separatrix density is at least 30% higher when a narrow scrape-off layer (SOL) grid width is chosen in SOLPS-ITER compared to the case for which the SOL grid width is maximised. The density increase is caused by kinetic neutrals being not confined inside the divertor region because of the reduced extent of the plasma grid. In this case, an enhanced level of reflections of energetic neutrals at the low-field side (LFS) metal divertor wall is observed. This leads to a shift of the ionisation source further upstream which must be accounted for as a numerical artefact. An overestimate in the cooling at the divertor entrance is observed in this case, identified by a reduced heat flux decay parameters λqdiv. Otherwise and further upstream the mid-plane heat decay length λq parameter is not affected by any change in divertor dissipation. This confirms the assumptions made for the ITER divertor design studies, i.e. that λq upstream is essentially set by the assumptions for the ratio radial to parallel heat conductivity. It is also shown that even for attached conditions the decay length relations λne > λTe > λq hold in the near-SOL upstream. Thus for interpretative edge plasma simulations one must take the (experimental) value of λne into account, rather than λq, as the former actually defines the required minimum upstream SOL grid extent. Keywords: edge plasma, modelling, SOLPS-ITER, JET, divertor

The possibility to safely operate tokamaks strongly depends on the capability to protect the divertor targets from excessive power load and erosion. Here, we will present a study of the JT-60SA Scenario 3 with a W wall to identify possible criticality in the transition from C to W wall. The assessment of the SOL and divertor plasma conditions is made with the SOLPS-ITER code suite. The simulations results show that safe operations are mainly limited by Te. In addition, the analysis of the balance equations highlights the importance of the impurity injection. It represents the main mechanism to drive the decrease in power load onto the outer target (POT) by impurity radiation and to achieve detachment via the collisions between D+ions and D2molecules.

A double-null configuration is being considered for the EU-DEMO, due to its potential benefits for power exhaust arising from the use of two active divertors and magnetically disconnected low- and high-field sides. Using systematic parameter scans in fluid simulations, we have investigated the divertor power exhaust in the EU-DEMO in a connected double-null configuration, and compared the edge plasma properties to those obtained in a single-null configuration under detached conditions anticipated for reactor operation. Neglecting drift effects and kinetic behaviour of the neutrals, no clear benefits of the double-null configuration could yet be identified for the radiation pattern and power mitigation on open field lines. Future work should address the aforementioned physics as well as the effect of the additional X-point on core radiation.

A physics and engineering analysis of alternative divertor configurations is carried out by examining benefits and problems by comparing the baseline single null solution with a Snowflake, an X- and a Super-X divertor. It is observed that alternative configurations can provide margin and resilience against large power fluctuations, but their engineering has intrinsic difficulties, especially in the balance between structural solidity and accessibility of the components and when the specific poloidal field coil positioning poses further constraints. A hybrid between the X- and Super-X divertor is proposed as a possible solution to the integration challenge.

Energy dissipation in the plasma edge is key for future tokamaks. The potential of neon as radiating seeding species in disconnected double null (DDN) configuration is assessed in EAST discharges in high confinement mode (H-mode). As the separation between the two separatrices in the studied DDN discharges is minimum 1.5 cm, the configuration is effectively a single null configuration, and the benefits of the double null topology are minimal. Neon seeding, on the other hand, has a favourable effect: both the target heat flux and the divertor temperature decrease more than five-fold with increased seeding rate in high-recycling conditions. Interpretive edge plasma simulations with SOLPS-ITER in support of ongoing transport analysis are presented. For the unseeded case the numerical results agree with the experimental data within a factor two for the target temperature conditions and measured neutral pressures in the active divertor. The key for achieving good agreement is a suitable selection of coefficients for anomalous transport and neutral conductances between the upper cryopump and the main chamber.

ASDEX Upgrade is preparing a hardware modification of its upper divertor in order to study alternative divertor configurations, like the X-divertor (XD) and the snowflake divertor (SF), that are discussed as a possible solution for the power exhaust problem. Experiments characterizing the current upper divertor in single-null (SN) configuration before and during detachment were carried out recently and interpreted by EMC3-EIRENE (Lunt et al., 2020) now including volumetric recombination. In continuation of these studies we here report on further simulations with this code extrapolating these conditions to the future upper divertor. For the same transport parameters, input power and upstream separatrix density for which the outer target (OT) of the SN is attached, the XD and SF show significant detachment at the OT accompanied by a reduction of the parallel heat flux by a factor of more than five. Despite the shallow field line incidence angles in the XD configuration the intrinsic 3D error fields from the current feeds only cause toroidal variations of the power fluxes of the order of 10%. With a hypothetical misalignment of the divertor coils by 3 cm substantial asymmetries in the power deposition profile are found, however, even those vanish when going to detached plasma conditions. In order to fully detach the plasma at the OT in the SF configuration impurities, here in the form of nitrogen, need to be puffed directly into the region of the secondary X-point in the simulation.

Recent TCV experiments confirm the predicted formation of an electric potential well, below the magnetic X-point, in configurations with unfavorable Btfield direction (ion ∇Bdrift away from the divertor), that substantially reshapes the typical divertor E×Bflow pattern. The local charge balance ∇⋅jin the private flux region (PFR) of diverted tokamak plasmas has been previously argued to be dominated by parallel and diamagnetic currents. This hypothesis is tested herein in TCV discharges by comparison with SOLPS-ITER simulations, fully accounting for drifts and currents. Simulated parallel currents correctly capture measured current profile characteristics for both targets and both Bt-directions, whereas those omitting drifts fail. It is shown that the resulting parallel currents dictate the electric fields in the PFR for low temperature (detached divertor) conditions resulting in locally negative electric plasma potential in configurations with unfavorable H-mode access.

SOLPS modeling of Ar seeding in ASDEX Upgrade H-mode plasma shows that the divertor compression of Ar increases with the seeding rate when the inner divertor is detached and the outer divertor is attached. This results from the redistribution of the Ar density as the inner divertor detaches. During the detachment of the inner divertor, the parallel friction and thermal forces on Ar increase in the inner divertor and decrease in the outer divertor. The net force of the two directs Ar from the low field side to the high field side of the plasma along the field line. The net force is found to be stronger after the detachment of the inner divertor in the simulation. Although the divertor compression of Ar increases, the ratio of radiated power by Ar inside the divertor over that inside the separatrix does not increase after the detachment. This is because the radiation efficiency, i.e. the radiated per Ar particle on average decreases with the seeding rate in the divertor and increases inside the separatrix. The radiation efficiency in the simulation is determined by the characteristic cooling curve of Ar and the transport effects.

A heat pulse experiment was carried out in the tokamak ASDEX Upgrade to estimate the stochastic layer width of a deuterium L-mode discharge with externally applied Magnetic Perturbations. The method relies on the deposition of ECRH pulses in the plasma edge while measuring the divertor target heat flux with high temporal resolution IR thermography and Langmuir probes. The experimental results were compared to simulations of the time dependent heat pulse propagation on a constant plasma background with the EMC3-EIRENE code package, using an ad-hoc screening model. If no screening was taken into account in the simulations a decrease in the characteristic heat pulse propagation time was observed, which shows that the heat transport is enhanced compared to the screened cases. No such enhancement was found in the experiment, indicating strong screening. In further simulations the effect of screening on the target fluxes was investigated for varying densities. For low densities it was found that screening reduces the strike line splitting strongly, while for higher densities no strong strike line splitting was found, independent of the screening degree. For strongly detached L-mode conditions with MPs experiments at AUG indicate that the lobe structures vanish completely. Keywords: ASDEX upgrade, Magnetic perturbations, Divertor heat flux, MSC: 00-01, 99-00

A series of neon seeded JET ELMy H-mode pulses is considered from the modeling as well as from the experimental point of view. For two different Ne seeding rates and two different D puffing gas levels the heating power, Pheat, is in the range 22–29.5 MW. The main focus is on the numerical reconstruction of the total radiated power (which mostly depends on the W concentration) and its distribution between core and divertor and of Zeff (which mostly depends on the Ne concentration). To model self-consistently the core and the SOL two input parameters had to be adjusted case by case: the SOL diffusivity, DSOL, and the core impurity inward pinch, vpinch. DSOL had to be increased with increasing ΓNe and the level of vpinch had to be changed, for any given ΓNe, according to the level of Pheat: it decreases with increasing Pheat. Since the ELM frequency, fELM, is experimentally correlated with Pheat, (it increases with Pheat) the impurity inward pinch can be seen as to depend on fELM. Therefore, to maintain a low vpinch level (i.e. high fELM) ΓNe/Pheat should not exceed a certain threshold, which slightly increases with the ΓD puffing rate. This might lead to a limitation in the viability of reducing the target heat load by Ne seeding at moderate ΓD, while keeping Zeff at acceptably low level. Keywords: Tokamak, Integrated modeling, Core plasma, Edge plasma

In the past at JET, with the MkI divertor, a systematic study of the influence of X-point height and poloidal flux expansion has been conducted [1,2] showing minor differences in the radiation distribution, whereas in [3] experiment and simulations have shown enhancement of detachment as the flux expansion was increased. More recently at JET, equipped with the ITER-like wall (ILW), radiative seeded scenarios have been studied and a maximum radiation fraction 75% has been achieved. EDGE2D-EIRENE [4–6] simulations [7,8] have already shown that the divertor heat fluxes can be reduced with N2 injection, qualitatively consistent with experimental observations [9], by adjusting the impurity injection rate to reproduce the measured divertor radiation. In this paper we will present edge predictive simulations on the assessment of effects of poloidal flux expansion and recycling on radiation distribution and X-point peaking on JET-ILW nitrogen seeded plasmas. Keywords: Power exhaust, Radiation, EDGE2D, EIRENE, JET, Nitrogen, Impurity seeding

The Swiss Plasma Center (SPC) is planning a divertor upgrade for the TCV tokamak. The upgrade aims at extending the research of conventional and alternative divertor configurations to operational scenarios and divertor regimes of greater relevance for a fusion reactor. The main elements of the upgrade are the installation of an in-vessel structure to form a divertor chamber of variable closure and enhanced diagnostic capabilities, an increase of the pumping capability of the divertor chamber and the addition of new divertor poloidal field coils. The project follows a staged approach and is carried out in parallel with an upgrade of the TCV heating system. First calculations using the EMC3-Eirene code indicate that realistic baffles together with the planned heating upgrade will allow for a significantly higher compression of neutral particles in the divertor, which is a prerequisite to test the power dissipation potential of various divertor configurations.

Divertor detachment in the TCV tokamak has been investigated through experiments and modelling. Density ramp experiments were carried out in ohmic heated L-mode pulses with the ion ∇B drift directed away from the primary X-point, similar to previous studies [1]. Before the roll-over in the ion current to the outer strike point, C III and Dα emission from the outer leg recede slowly from the strike point toward the X-point, at a rate of ∼2.0 × 10−19m/m−3 along the magnetic field as the electron temperature along the leg reduces with increasing density. Around the onset of detachment, the upstream density profile and outer target Dα profiles broaden, possibly leading to an increase in radiation in the SOL by increased interaction between the SOL and the carbon tiles lining the outer wall. The plasma conditions upstream and at various locations along the detached outer divertor leg have been characterised, and the consistency of this data has been checked with the interpretive OSM-EIRENE-DIVIMP suite of codes [2] and are broadly found to be consistent with measured Dγ/Dα emissivity profiles along the detached outer divertor leg.

An H-mode database of ASDEX Upgrade plasmas with improved diagnostics for a variety of different plasma conditions has been analysed to study the momentum removal in the Scrape-Off Layer (SOL). A strong reduction, up to a factor of 100, of the electron pressure is observed close to the separatrix. Sets of L-mode and H-mode like plasma simulations with the SOLPS5.0 code package have supported the interpretation of the experimental observations (e.g., apparent momentum gain, strong radial position dependence) and helped to clarify the role of ion-neutral interactions. The experimental data are in turn used to check the SOLPS predictions, which differ quantitatively, suggesting that important pieces of physics are still not successfully captured by the code. Keywords: Momentum removal, Scrape-off layer, SOLPS, Divertor detachment

The understanding of divertor physics and the evolution of divertor detachment is crucial for developing the capability to model power exhaust in current experiments and reliably predict it for future fusion devices. In simulations of ASDEX Upgrade, an experimentally observed region of high density in the high field side scrape-off layer has been recovered. Validated modeling with SOLPS5.0 shows that a detailed match of the high field side scrape-off layer plasma is not only important for local plasma parameters, but can lead to strong changes in global parameters. Drifts play a crucial role in lower-single null discharges with forward toroidal field (∇B→-drift pointing down). Their inclusion changes the spatial extent as well as the radial and poloidal gradients of the high field side high density. Adapted transport coefficients that take into account core fueling by plasma diffusion due to the presence of the high field side density and drift-driven radial fluxes now reconcile the modeled deuterium compression ratio, divertor neutral density, neutral radiation levels and deuterium fueling rates with experimental measurements. The onset of strong volume recombination in the simulations now allows to remove the previously necessary increase of perpendicular transport in the inner divertor from the simulations. Keywords: PSI-21: Edge modeling, Divertor plasma, ASDEX-Upgrade, Detachment, High field side high density, SOLPS

We discuss the benefits of installing a pair of in-vessel coils with currents |Ifx| ≲ 50 kAt in the upper divertor of ASDEX Upgrade (AUG) to study a series of ‘alternative’ divertor configurations, like the Snowflake (SF) and the X-divertor (XD), that are currently considered as alternative solutions for the power exhaust problem. The possibility of operating the standard lower single-null (SN) and double-null (DN) would be preserved. Potential effects to reduce the peak parallel- and/or perpendicular heat flux are predicted from a simple geometrical-diffusive model as well as by numerical EMC3-EIRENE simulations for pure deuterium attached conditions with spatially constant diffusion coefficients. Beyond that a series of other potential transport- and radiation related heat flux mitigation effects are identified and could be studied experimentally with the modified upper divertor in the high-power divertor Tokamak AUG. Keywords: Power exhaust, Alternative divertor, ASDEX Upgrade, EMC3-EIRENE, Tokamak, Flux expansion

Future tokamak reactors of conventional design will require high levels of exhaust power dissipation (more than 90% of the input power) if power densities at the divertor targets are to remain compatible with active cooling. Impurity seeded H-mode discharges in JET-ITER-like Wall (ILW) have reached a maximum radiative fraction (Frad) of ∼75%. Divertor Langmuir probe (LP) measurements in these discharges indicate, however, that less than ∼3% of the thermal plasma power reaches the targets, suggesting a missing channel for power loss. This paper presents experimental evidence from limiter LP for enhanced cross-field particle fluxes on the main chamber walls at high Frad. In H-mode nitrogen-seeded discharges with Frad increasing from ∼30% to up to ∼75%, the main chamber wall particle fluence rises by a factor ∼3 while the divertor plasma fluence drops by one order of magnitude. Contribution of main chamber wall particle losses to detachment, as suggested by EDGE2D-EIRENE modeling, is not sufficient to explain the magnitude of the observed divertor fluence reduction. An intermediate detached case obtained at Frad ∼ 60% with neon seeding is also presented. Heat loads were measured using the main chamber wall thermocouples. Comparison between thermocouple and bolometry measurements shows that the fraction of the input power transported to the main chamber wall remains below ∼5%, whatever the divertor detachment state is. Main chamber sputtering of beryllium by deuterium is reduced in detached conditions only on the low field side. If the fraction of power exhaust dissipated to the main chamber wall by cross-field transport in future reactors is similar to the JET-ILW levels, wall plasma power loading should not be an issue. However, other contributions such as charge exchange may be a problem.

Identification of the mechanisms for the H-mode density limit in machines with fully metallic walls, and their scaling to future devices is essential to find for these machines the optimal operational boundaries with the highest attainable density and confinement. Systematic investigations of H-mode density limit plasmas in experiments with deuterium external gas fuelling have been performed on machines with fully metallic walls, JET and AUG and results have been compared with one another.Basically, the operation phases are identical for both tokamaks: the stable H-mode phase, degrading H-mode phase, breakdown of the H-mode with energy confinement deterioration usually accompanied by a dithering cycling phase, followed by the l-mode phase. The observed H-mode density limit on both machines is found close to the Greenwald limit (n/nGW=0.8–1.1 in the observed magnetic configurations). The similar behavior of the radiation on both tokamaks demonstrates that the density limit (DL) is neither related to additional energy losses from the confined region by radiation, nor to an inward collapse of the hot discharge core induced by overcooling of the plasma periphery by radiation.It was observed on both machines that detachment, as well as the X-point MARFE itself, does not trigger a transition in the confinement regime and thus does not present a limit on the plasma density. It is the plasma confinement, most likely determined by edge parameters, which is ultimately responsible for the transition from H- to l-mode.The measured Greenwald fractions are found to be consistent with the predictions from different theoretical models [16,30] based on MHD instability theory in the near-SOL. Keywords: Density limit, H-mode, Detachment, Greenwald fraction, JET, ASDEX Upgrade

Future fusion reactors require a safe, steady state divertor operation. A possible solution for the power exhaust challenge is the detached divertor operation in scenarios with high radiated power fractions. The radiation can be increased by seeding impurities, such as N for dominant scrape-off-layer radiation, Ne or Ar for SOL and pedestal radiation and Kr for dominant core radiation.Recent experiments on two of the all-metal tokamaks, ASDEX Upgrade (AUG) and JET, demonstrate operation with high radiated power fractions and a fully-detached divertor by N, Ne or Kr seeding with a conventional divertor in a vertical target geometry. For both devices similar observations can be made. In the scenarios with the highest radiated power fraction, the dominant radiation originates from the confined region, in the case of N and Ne seeding concentrated in a region close to the X-point.Applying these seed impurities for highly radiative scenarios impacts local plasma parameters and alters the impurity transport in the pedestal region. Thus, plasma confinement and stability can be affected. A proper understanding of the effects by these impurities is required in order to predict the applicability of such scenarios for future devices.

Robust power exhaust schemes employing impurity seeding are needed for target operational scenarios in present day tokamak devices with metallic plasma-facing components (PFCs). For an electricity-producing fusion power plant at power density Psep/R>15MW/m divertor detachment is a requirement for heat load mitigation. 2D plasma edge transport codes like the SOLPS code as well as plasma-wall interaction (PWI) codes are key to disentangle relevant physical processes in power and particle exhaust. With increased quantitative credibility in such codes more realistic and physically sound estimates of the life-time expectations and performance of metallic PFCs can be accomplished for divertor conditions relevant for ITER and DEMO. An overview is given on the recent progress of plasma edge and PWI modelling activities for (carbon-free) metallic devices, that include results from JET with the ITER-like wall, ASDEX Upgrade and Alcator C-mod. It is observed that metallic devices offer an opportunity to progress the understanding of underlying plasma physics processes in the edge. The validation of models can be substantially improved by eliminating carbon from the experiment as well as from the numerical system with reduced degrees of freedom as no chemical sputtering from amorphous carbon layers and no carbon or hydro-carbon transport are present. With the absence of carbon as the primary plasma impurity and given the fact that the physics of the PWI at metallic walls is less complex it is possible to isolate the crucial plasma physics processes relevant for particle and power exhaust. For a reliable 2D dissipative plasma exhaust model these are: cross-field drifts, complete kinetic neutral physics, geometry effects (including main-chamber, divertor and sub-divertor structures), SOL transport reflecting also the non-diffusive nature of anomalous transport, as well as transport within the pedestal region in case of significant edge impurity radiation affecting pedestal pressure and hence Psep.

During detachment a structure of strongly enhanced density develops close to the inner target. Its dynamics is approximated by those of radiative fluctuations appearing at a similar position and studied by means of a time-delay-estimation technique in the ASDEX Upgrade tokamak. Compared to theory the dynamics can be described as follows: at increasing density the ionization front moves upstream to reduce ionization radiation in order to balance the increased recombination radiation. The recombination zone stays close to the target strike point. The parallel motion of the ionization front is determined by the perpendicular neutral motion. The divertor nose constitutes an obstacle for the perpendicular neutral flux from the target to the region above the X-point. Passing into this shadow the neutral flux above the X-point is strongly reduced, the ionization front fades away and the heat flux from upstream can increase the temperature in the recombination region, subsequently reducing recombination and reforming an ionization front below the X-point. A cyclic reformation of the ionization front propagating from below to above the X-point occurs leading to a fluctuation as observed in the experiment.

A global heat flux model based on a fractional derivative of plasma pressure is proposed for the heat transport in fusion plasmas. The degree of the fractional derivative of the heat flux, α, is defined through the power balance analysis of the steady state. The model was used to obtain the experimental values of α for a large database of the Joint European Torus (JET) carbon-wall as well as ITER like-wall plasmas. The fractional degrees of the electron heat flux are found to be α