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The power exhaust concept and an appropriate divertor design are common critical issues for tokamak DEMO design activities which have been carried out in Europe, Japan, China, Korea and the USA. Conventional divertor concepts and power exhaust studies for recent DEMO designs (Pfusion = 1 – 2 GW, Rp = 7 – 9 m) are reviewed from the viewpoints of the plasma physics issues and the divertor engineering design. Radiative cooling is a common approach for the power fusion scenario. Requirements on the main plasma radiation fraction (fradmain = Pradmain/Pheat) and the plasma performance constrain the divertor design concept. Different challenges contribute to optimizing the future DEMO designs: for example, (i) increasing the main plasma radiation fraction for ITER-level Psep/Rp designs and simplifying the divertor geometry, and (ii) extending ITER divertor geometry with increasing divertor radiation (Praddiv) for larger Psep/Rp ≥ 25MWm−1 designs. Power exhaust simulations with large Psep = 150 – 300 MW have been performed using integrated divertor codes considering an ITER-based divertor geometry with longer leg length (1.6 – 1.7 m), as in a common baseline design. Geometry effects (ITER like geometry or more open one without baffle) on the plasma detachment profile and the required divertor radiation fraction (fraddiv = Praddiv/Psep) were key aspects of these studies. All simulations showed that the divertor plasma detachment were extended widely across the target plate with a reduction in the peak heat load of qtarget ≤ 10 MWm−2 for the large fraddiv = 0.7 – 0.8, while the peak qtarget location and value were noticeably different in the partially detached divertor. Simulation results also demonstrated that radial diffusion coefficients of the heat and particle fluxes were critical parameters for DEMO divertor design, and that effects of plasma drifts on outboard-enhanced asymmetry of the heat flux, suggested the need for longer divertor leg to ensure the existence of a detached divertor operation with qtarget ≤ 10 MWm−2.Integrated design of the water cooled divertor target, cassette body (CB) and cooling pipe routing has been developed for each DEMO concept, based on the ITER-like tungsten monoblock (W-MB) with Cu-alloy cooling pipes. Engineering design adequate under higher neutron irradiation condition was required. Therefore, inlet coolant temperature (Tcool) was increased. In current designs, it still shows a large potential variation between 70 °C and 200 °C. The influence of thermal softening on the Cu-alloy (CuCrZr) pipe was fostered near the strike-point when the high qtarget of ∼10 MWm−2 was studied. Improved technologies for high heat flux components based on the ITER W-MB unit have been developed for EU-DEMO. Different coolant conditions (low- and high-Tcool) were provided for Cu-alloy and reduced activation ferritic martensitic (RAFM) steel heat sink units, respectively. The high-Tcool coolant was also considered for the CB and supporting structures. Appropriate conditions for the high-Tcool coolant, i.e. 180 °C/ 5 MPa (EU-DEMO) and 290 °C/ 15 MPa (JA-DEMO, CFETR and K-DEMO), will be determined in the future optimizations of the divertor and DEMO design.

The neutral-induced material erosion is still an unsolved issue for ITER and future reactors, which influences the lifetime of the first wall material and tritium retention rate. A low energy neutral particle analyzer (LENPA) and a quartz crystal microbalance (QMB) were recently installed together on the equatorial port of sector H in EAST to study neutral-induced material erosion. The QMB can provide real-time and in-situ measurements on the erosion process of the wall material, while the neutral energy spectrum given by LENPA can help to calculate the material erosion rate for comparison with QMB results. In the 2021 EAST summer campaign, neutral-induced Al erosion was successfully studied with the two available diagnostics for the first time. In a series of long-pulse discharges, the average Al erosion rates measured by the QMB are consistent with the theoretical calculations using the LENPA data. Furthermore, the changes in plasma density, heating power and the real-time lithium powder injection result in different neutral flux and energy, which lead to changes in the neutral-induced material erosion rates.

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 D2 seeding when divertor detachment is achieved. Compared to edge D2 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 D2 fueling can affect the divertor condition, which may lead to a W erosion mitigation. D2 puffing at divertor target is less effective on increasing the upstream plasma density than OMP D2 puffing, and thus it is less effective on reduction of W erosion rates. W gross erosion profiles with different amount of injected D2 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 D2 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 D2 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 D2 fueling discharges.

The neutral energy spectrums with an energy range of 20–3000 eV was measured by the low energy neutral particle analyzer (LENPA) in the 2021 EAST summer campaign. The proportion of the neutral particles below 1000 eV measured by LENPA is more than 85 %. In the ohmic discharges, the integrated neutral flux increases with increasing the line-averaged density. However, the neural particles with higher energy (1000–3000 eV) are insensitive to the line-averaged density. In addition, the neutral flux in the energy range of 20–1000 eV of the discharges with supersonic molecular beam injection (SMBI) was lower than those without SMBI, which was different from neutrals in the range of 1000–3000 eV. The neutral flux increases with higher auxiliary heating power in all energy range. Compared to the ohmic discharges, the mean energy of neutral particles in auxiliary heated discharges is increased significantly.

Physical sputtering caused by particle bombardment is believed to be the main erosion mechanism of W materials in fusion devices, in which W atoms are the sputtering products. However, the tungsten deuteride molecule (WD) spectra have been observed in both TEXTOR and ASDEX Upgrade, which was believed to be the product of chemically assisted physical sputtering (CAPS), a new sputtering mechanism that has been proposed in recent years. In this paper, we report the spectroscopic observation of WD molecules in the EAST W divertor. The behaviors of WD molecules sputtering are compared with W atoms sputtering via the spectral measurements of the ro-vibrational band emission of WD 6Π → 6Σ+ in the spectral range between 673 nm and 678 nm and the WI line emission at 400.9 nm. The physical sputtering characters and chemical sputtering characters of WD molecule sputtering were confirmed in EAST. The measurements in EAST showed that there is an energy threshold for WD molecules sputtering and that the sputtering energy threshold of WD molecules is smaller than that of W atoms. Furthermore, the dependence of WD molecule sputtering efficiency (the absolute WD photon flux (PhWD) normalized to the particle flux (Γion) reaching the target with a constant Te) on impact energy and heat flux were studied, presenting significantly different behaviors compared with W atom sputtering. In addition, it is found that the decrease of WD sputtering efficiency with the heat flux is accompanied by the rise of the photon flux of Dδ (410.06 nm) normalized to the particle flux (Γion) at the divertor target, which may imply the enhanced deuterium desorption at the W surface. The further increase of heat flux hitting the target surface could elevate the surface temperature and benefit the deuterium release from the surface, which may influence the formation of WD molecules at the surface layer.

As net erosion of tungsten plasma-facing components in tokamak divertors is strongly reduced by tungsten prompt redeposition, predicting and monitoring tungsten prompt redeposition is crucial to estimate the lifetime of tungsten PFCs and the source of tungsten impurities in tokamak divertors. To that end, a theoretical and numerical analysis of tungsten prompt redeposition near divertor targets is presented. It is shown that the reduction of the Larmor radius of tungsten impurities due to multiple ionizations during the first gyro-orbit of those impurities does not significantly lower tungsten prompt redeposition, as the sheath electric field is much stronger than the Lorentz force in the sheath region because of the large inertia of tungsten particles. As a consequence, only tungsten impurities which are ionized near or beyond the sheath entrance avoid prompt redeposition. Tungsten prompt redeposition is therefore mainly governed by the ratio of the ionization mean-free path of sputtered neutral tungsten particles over the sheath width. This governing parameter has similar values for divertor plasma conditions in DIII-D experiments and in plasma conditions expected in the far-SOL of the ITER divertor, indicating that investigations of tungsten erosion in smaller devices like DIII-D are relevant for future fusion reactors like ITER. Furthermore, this governing parameter scales linearly with the magnetic field strength. Tungsten net erosion in the divertor region might thus significantly increase in fusion devices operating at high magnetic field. It is also observed that the fraction of tungsten impurities which promptly redeposit is only correlated to the fraction of impurities in the highest charge state reached during the first gyro-orbit. As a consequence, in-situ monitoring of tungsten prompt redeposition in divertors requires monitoring photon emissions associated with the ionizations of tungsten impurities in charge states Z>2+, typically W3+, W4+and W5+for plasma conditions expected in the far-SOL of the ITER divertor.

Hot spot is a serious challenge limiting long pulse operation with LHCD (lower hybrid current drive) in tokamak. In order to mitigate the hot spot in guard limiter and improve LHCD capability, effect of edge plasma density on inducing hot spot and current drive has been studied in EAST. The temperature in the guard limiter of the LH antenna, inducing hot spot directly, increases with edge density and LH power. Studies show that the hot spot is mainly ascribed to the heat flux in front of LH antenna. Further simulation indicates that such spot corresponds to the peak position of edge density due to local LH electric field. In addition, due to the stronger parametric instability (PI) behavior in the case of higher edge density, the current drive capability decreases with edge density. Strike-point splitting behaviors appear with density increase, in agreement with current profiles in the edge region and the reduction of total driven current, suggesting that more power is deposited in edge region, which does more contribution to hot spot. Above studies offer one possible idea to optimize the edge density so as to satisfy the coupling, mitigate the heat flux in the guard limiter, and improve current drive capability for fusion device.

Two new quartz crystal microbalance (QMB) systems have been installed at different locations in the mid-plane of EAST to measure erosion and deposition of the first wall materials. The two QMBs with different distances from the plasma show complex and distinct erosion and deposition behavior. The QMB in the mid-plane H port closer to the plasma shows a relatively higher net erosion result with an average erosion rate of 1.34 ng/(cm2s). The normal discharges mainly contribute to the total net erosion and the maximum erosion amount reaches up to 110 ng/cm2. Contrarily, a slight net deposition of 0.11 ng/(cm2s) is obtained for the QMB in the mid-plane G port further away from the plasma and the erosion and deposition amount of most discharges are much lower due to lower particle flux. The disruptive discharges lead to a total net deposition rate of 0.48 ng/(cm2s). Measurement results by the QMBs are demonstrated by a good agreement with the post-mortem analysis by Rutherford Backscattering Spectrometry (RBS).

To evaluate the erosion effects of neutral particles on the first wall materials during tokamak operation, aluminum (Al) coated samples were exposed to a sequence of plasma discharges in 2018 and 2019 EAST campaigns. In order to avoid the incidence of charged particles, the samples were installed in a molybdenum (Mo) cylindrical sleeve and positioned far from the main plasma. The erosion thicknesses of Al layer were measured by Rutherford Backscattering Spectrometry (RBS) before and after the plasma exposure. The average net erosion rates of the samples in the two campaigns are around 1.18 × 1013 atoms cm−2 s−1 and 4.64 × 1013 atoms cm−2 s−1, respectively. The different erosion rate may mainly result from different plasma density and heating power. The theoretical calculated erosion rates using the neutral distribution under different density conditions are in fair agreement with the experimental results.

Rare isotope tracer injection represents an established way for following particle transport in fusion research, but a tracer experiment is only as good as the tracer can be followed and separated from the common isotope. Physical models for transport analysis such as ERO2.0 require accurate total amounts of deposition with high spatial resolution. This work describes an extended high throughput NRA post analysis approach together with the surface analysis results of 2 technically rough graphite-tiles of 150x150 mm2 surface area exposed to D and He plasmas in EAST.The applied 1.43 MeV deuteron beam NRA offers percent range accuracy for layer thicknesses ranging from few 10 nm to 20 µm. Optimisation of the beam line and beam optics enables operation with ≤20 nA without reaching critical radiation levels, resulting in a measurement time of 5–10 s per spot. This enabled acquiring in total 3070 NRA points on the central 46 × 46 mm2 cut-outs of each sample and an adjacent part of equal size within 12 h of analysis time.The layers contained 50–60% D, ~2% 12C, and ~3% O. The central 20 mm radius around the injection hole contains 98% of the total 13C deposition. NRA finds up to 5.8 ± 0.9 × 1022 13C/m2 close to the hole with an e-folding fall-off length of about 3–10 mm in every radial direction. In the D discharge 24 ± 2% and 16 ± 2% in He of the injected 13C are found on the tiles with triangle shaped deposition profiles following ExB. The comparison to colour fringe analysis and SIMS reveals the particular strength of D-NRA regarding roughness, thick layers, and the separation of 12C and 13C deposits, but in compatible regions all methods agree. Compared to these, the NRA yields additional information indicating depositing species and physical processes.

The OEDGE code is used to model tungsten erosion and transport for experiments with toroidal rings of high-Z metal tiles in the DIII-D tokamak. Such modeling is needed for both experimental and diagnostic design to have estimates of the expected core and edge tungsten density and to understand the various factors contributing to the uncertainties in these calculations. OEDGE simulations are performed using the planned experimental magnetic geometries and plasma conditions typical of both L-mode and inter-ELM H-mode discharges in DIII-D. OEDGE plasma reconstruction based on specific representative discharges for similar geometries is used to determine the plasma conditions applied to tungsten plasma impurity simulations. A new model for tungsten erosion in OEDGE was developed which imports charge-state resolved carbon impurity fluxes and impact energies from a separate OEDGE run which models the carbon production, transport and deposition for the same plasma conditions as the tungsten simulations. These values are then used to calculate the gross tungsten physical sputtering due to carbon plasma impurities which is then added to any sputtering by deuterium ions; tungsten self-sputtering is also included. The code results are found to be dependent on the following factors: divertor geometry and closure, the choice of cross-field anomalous transport coefficients, divertor plasma conditions (affecting both tungsten source strength and transport), the choice of tungsten atomic physics data used in the model (in particular ionization rate for W-atoms), and the model of the carbon flux and energy used for calculating the tungsten source due to sputtering. Core tungsten density is found to be of order 1015m−3 (excluding effects of any core transport barrier and with significant variability depending on the other factors mentioned) with density decaying into the scrape off layer. For the typical core density in the plasma conditions examined of 2 to 4 ×1019m−3, this represents a concentration on the order of 5 × 10−5.

Carbon deposition on tungsten between ELMs was investigated in DIII-D in semi-attached/detached H-mode plasma conditions using fixed outer strike point (OSP) positions. Carbon deposition during plasma exposure of tungsten was monitored in-situ by measuring the reflectivity of the tungsten sample surface. No significant carbon deposition, i.e., without strong variations of the reflectivity, was observed during these experiments including discharges at high densities. In contrast, ERO modeling predicts a significant carbon deposition on the tungsten surface for those high density plasma conditions. The surface reflectivity decreases with methane injection, consistent with increased carbon coverage, as expected. The sweeping of OSP leads to a pronounced increase of the surface reflectivity, suggesting that the strike point sweeping may provide an effective means to remove carbon coating from tungsten surface. The ERO modeling however predicts again a regime of carbon deposition for these experiments. The discrepancies between carbon deposition regime predicted by the ERO model and the experimental observations suggest that carbon erosion during ELMs may significantly affect carbon deposition on tungsten.

We present measurements and modeling of aluminum erosion and redeposition experiments in separate helium and deuterium low power, low density L-mode plasmas at the outer divertor strike point of DIII-D to provide a low-Z material benchmark dataset for tokamak erosion-deposition modeling codes. Coatings of Al ∼100nm thick were applied to ideal (smooth) and realistic (rough) surfaces and exposed to repeat plasma discharges using the DiMES probe. Redeposition in all cases was primarily in the downstream toroidal field direction, evident from both in-situ spectroscopic and post-mortem non-spectroscopic measurements. The gross Al erosion yield was estimated from film thickness change measurements of small area samples, and was found to be ∼40–70% of the expected erosion yield based on theoretical physical sputtering yields after including sputtering by a 1–3% carbon impurity. The multi-step redeposition and re-erosion process, and hence the measured net erosion yield and material migration patterns, were found to be influenced by the surface roughness and/or porosity. A time-dependent model of material migration accounting for deposit accumulation in hidden areas was developed to reproduce the measurements in these experiments and determine a redeposition probability distribution function for sputtered atoms. Keywords: Erosion, Deposition, Migration, Roughness, Aluminum, Mixed materials

Tracer injection experiments in TEXTOR with MoF6 and WF6 lead to local deposition of about 6% for Mo and about 1% for W relative to the injected amount of Mo and W atoms. Modelling of these experiments has been done with ERO applying updated data for physical sputtering. The dissociation of the injected molecules has been treated in a simplified manner due to the lack of dissociation rate coefficients. However, with this it was possible to reproduce the observed radial penetration of Mo and W atoms into the plasma. The modelled local deposition efficiencies are about 50% for Mo and 60% for W assuming typical plasma parameters for the experimental conditions used. To reproduce the measured deposition efficiencies an enhancement factor for the erosion of deposited Mo and W has to be assumed (∼10 for Mo and ∼25 for W). Due to the rather low electron temperature Te of these plasma conditions (Te∼15eV at the location of injection), Mo and W are mostly sputtered by impurities whereas sputtering due to deuterium is negligible. A parameter study applying larger electron temperature leads to increased sputtering and thus to reduced local deposition efficiencies of about 30% for Mo and 5% for W. Though, even under these conditions enhanced erosion, albeit with reduced enhancement factors, is needed in the modelling to obtain the small measured deposition efficiencies. keywords: Erosion & Deposition, Sputtering, Impurity transport, TEXTOR, PACS: 52.40.Hf, 52.77.Dq, 52.65.Pp

As High-Z materials will likely be used as plasma-facing components (PFCs) in future fusion devices, the erosion of high-Z materials is a key issue for high-power, long pulse operation. High-Z material erosion and redeposition have been studied using tungsten and molybdenum coated samples exposed in well-diagnosed DIII-D divertor plasma discharges. By coupling dedicated experiments and modelling using the 3D Monte Carlo code ERO, the roles of sheath potential and background carbon impurities in determining high-Z material erosion are identified. Different methods suggested by modelling have been investigated to control high-Z material erosion in DIII-D experiments. The erosion of Mo and W is found to be strongly suppressed by local injection of methane and deuterium gases. The 13C deposition resulting from local 13CH4 injection also provides information on radial transport due to E ×B drifts and cross field diffusion. Finally, D2 gas puffing is found to cause local plasma perturbation, suppressing W erosion because of the lower effective sputtering yield of W at lower plasma temperature and for higher carbon concentration in the mixed surface layer.