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Experimental data, together with interpretive modeling tools, are examined to study trends in the tungsten (W) source in the all-W environment of the WEST tokamak, both from the divertor and from the main chamber. In particular, a poloidal limiter protecting an ion cyclotron resonance heating (ICRH) antenna is used as proxy for main chamber sourcing. The key study is carried out by stepping up lower hybrid current drive (LHCD) power, as the only auxiliary power source. Limiter and divertor W sources exhibit a qualitatively similar proportionality to the total power crossing the separatrix, P SEP, although the main chamber source remains substantially lower than the divertor source, for the range of P SEP accessible in the experiments. Intepretive modeling of the limiter source is carried out with a particle-in-cell (PIC) sheath model coupled to a surface sputtering model. Oxygen is used as a proxy for all light impurity species allowing for characterization of the critical W erosion regions. To get a good quantitative match to the data, it is necessary to assume that the oxygen arrives at the surface mostly at high ionization stages (4+ and above). A separate simulation with SOLEDGE-EIRENE, constrained to measured upstream scrape-off-layer plasma profiles, gives oxygen fractional abundances that are compatible with the PIC simulation result. This is understood to arise from transport processes that dominate over recombination. Substituting the LHCD by ICRH, in an equivalent experiment, the local W source exhibits a 3× enhancement. This can be matched by the simulation, by assuming local RF electric field rectification, based on ∼100 eV peak-to-peak, near-antennna electric field. This work has highlighted the particular importance of understanding the ion charge state balance of light impurities as these are most likely the dominant sputtering species in fusion devices with high-Z walls.

Achieving a successful plasma current ramp-up in a full tungsten tokamak can be challenging due to the large core radiation (and resulting low core temperature) that can be faced with this heavy metallic impurity if its relative concentration is too high. Nitrogen injection during the plasma current ramp-up of WEST discharges greatly improves the core temperature and magnetohydrodynamic (MHD) stability. Experimental measurements and integrated simulations with the RAPTOR code, complemented with the QuaLiKiz neural network for computing turbulent transport, allow a detailed understanding of the mechanisms at play. Increased edge radiation during this transient phase is shown to improve confinement properties, driving higher core temperature and better MHD stability. This also leads to increased operation margins with respect to tungsten contamination.

We summarize a series of numerical experiments of collisional dynamics in dense stellar systems such as globular clusters (GCs) and in weakly collisional plasmas using a novel simulation technique, the so-called Multi-particle collision (MPC) method, alternative to Fokker-Planck and Monte Carlo approaches. MPC is related to particle-mesh approaches for the computation of self consistent long-range fields, ensuring that simulation time scales with $N\log N$ in the number of particles, as opposed to $N^2$ for direct $N$-body. The collisional relaxation effects are modelled by computing particle interactions based on a collision operator approach that ensures rigorous conservation of energy and momenta and depends only on particles velocities and cell-based integrated quantities., Comment: 7 pages, 2 figures. To appear in the proceedings of the 362 IAU symposium "Predictive Power of Computational Astrophysics as a Discovery Tool". Chamonix, France 8-12 Nov. 2021

Background Angiography-derived computed (virtual) fractional flow reserve (vFFR) permits the assessment of coronary physiology without the need for a pressure wire or hyperaemia. However, accuracy is limited by assumptions made about coronary microvascular resistance (CMVR) [1]. We hypothesised that machine learning may allow us to “tune” our estimate of the CMVR to increase the accuracy of vFFR. Purpose To determine whether routine clinical data can personalise CMVR and improve the accuracy of vFFR on an individual case basis. Methods Patients with chronic coronary syndromes underwent coronary angiography with FFR assessment. Vessel-specific CMVR was computed using a 3D- computational fluid dynamics simulation with invasively measured proximal and distal pressures. Predictive models were created using non-linear autoregressing moving average with exogenous inputs (NARMAX) modelling with computed CMVR as the dependent variable. vFFR was computed using previously described methods [2]. Three simulations were run, using: 1) a generic CMVR value (Model A); 2) a NARMAX-predicted CMVR based upon a panel of simple clinical data (Model B); and 3) a NARMAX-predicted CMVR incorporating echocardiographic data (Model C). The diagnostic and quantitative accuracy of each model was compared with directly measured FFR. Results Eighty four patients underwent coronary angiography with FFR assessment in 157 vessels. Mean age was 64 (±0.1) years and 64 (76%) were male. Mean FFR was 0.79 (±0.15). Mean CMVR was 1.01e+10 Pa/m3 s–1. vFFR error with Model A was ±0.10, with Model B was ±0.07 (p Conclusion vFFR is dependent upon not only the epicardial stenosis, but also the CMVR, estimation of which can be personalised based upon clinical and echocardiographic data. This can be used to increase the accuracy of vFFR. Funding Acknowledgement Type of funding sources: Other. Main funding source(s): British Heart Foundation Clinial Research Training FellowshipNIHR i4i grant Table 1

International audience; Plasma heating in the full Tungsten (W) Environment in Steady-state Tokamak (WEST) relies on electromagnetic waves in both Lower Hybrid (LH) and Ion Cyclotron Range of Frequencies (ICRF). The present study focuses mostly on the optimization of discharges heated with ICRF, by reporting different methods to first optimize wave coupling, optimize their absorption and reduce the impurity production. It is shown that ICRF coupling can be optimized either by moving the plasma closer to antennas, increasing the plasma density, wave frequency and LH power. We show that the absorption efficiency correlates with the hydrogen concentration with the existence of an optimum between 7 and 10% as expected for a minority heating scenario. Absolutely calibrated visible spectroscopy sightlines were used to monitor ion fluxes in different locations as part of an effort to quantitively estimate the contribution of different impurity sources to the core contamination by tungsten. It is typically found that in discharges with high total RF-power (above 5MW of LH and 3MW of ICRF), divertor and antenna limiter sources can reach similar order of magnitude during the ICRF phase.

This is a PDF file of an article that has undergone enhancements after acceptance, such as the addition of a cover page and metadata, and formatting for readability, but it is not yet the definitive version of record. This version will undergo additional copyediting, typesetting and review before it is published in its final form, but we are providing this version to give early visibility of the article. Please note that, during the production process, errors may be discovered which could affect the content, and all legal disclaimers that apply to the journal pertain.; International audience; WEST is a full metallic tokamak with an extensive set of diagnostics for heat load measurements. In this paper, heat loads on the lower divertor of WEST are investigated using two independent methods. A first method relies on the thermal inversion of temperature measurements from arrays of thermal sensors embedded a few millimeters below the surface, while the second consists in the inversion of black body surface temperatures measured by infra-red (IR) thermography. The challenge of IR based temperature measurements in the full metal environment of WEST is addressed through a simplified model, allowing to correct for global reflections and low surface emissivities of tungsten surfaces. A large database ( L-mode discharges) is investigated. It is found that the energy absorbed by an outer divertor tile during a plasma discharge is closely estimated by the two diagnostics, over a large set of experimental conditions. A similar match is also found for the peak heat flux value on the outer target. The toroidal modulation of target heat loads by magnetic ripple is found to be consistent with the geometrical projection of a parallel heat flux component. Additionally, the heat flux channel width at the target is found to scale linearly with the magnetic flux expansion as expected. These observations give confidence in the robustness of the data from both diagnostics, and confirm the simple geometrical rules at use in the description of heat flux deposition on divertor targets. However, it is shown that the heat flux channel width estimated from infra-red thermography is about three times lower than the width estimated from embedded measurements, which is still under investigation.

Because the gaps between plasma-facing components in fusion devices are comparable in size to the ion Larmor radius (of the order of 1 mm), magnetic field line tracing, the so-called "optical approximation", cannot accurately predict the fine scale heat load distribution around the gap edges. Finite Larmor radius effects dominate ion deposition. The poloidal component of the ion flux striking the surface is always in the diamagnetic/EXB drift direction, meaning that ions preferentially load one side of the gap. Usually electrons can be described optically due to their smaller Larmor radius. Depending on the local inclination of magnetic flux surfaces, it is possible that ions and electrons wet the same side of the gap, or opposite sides. Two-dimensional particle-in-cell simulations and dedicated experiments performed in the COMPASS tokamak are used to better understand processes responsible for plasma deposition on the sides of toroidal gaps between castellated plasma-facing components in tokamaks. The different contributions of the total incoming flux along a toroidal gap have been observed experimentally for the first time in COMPASS. These experimental results confirm the model predictions, demonstrating that in specific cases the heat deposition does not necessarily follow the optical approximation. The role played by electric fields in the deposition pattern is marginal, contrary to local non-ambipolarity that can change the asymmetrical plasma deposition from one side of the toroidal gap to the other.

One of the missions of the WEST tokamak is to test in a realistic tokamak environment, the ITER-like divertor plasma facing components made with tungsten. On exposed leading edges of monoblocks, poloidally-distributed cracks having an average spacing of 0.4 mm, running perpendicular to the cooling tube axis, were observed following the experimental campaigns in 2018. Damage may be induced by different processes which can lead to: brittle fracture below the Ductile to Brittle Transition temperature or ductile failure for which softening (recovery/recrystallization) process plays a major role. To improve our understanding about the leading edge damage process, numerical simulations were run here. The following results are specially studied: (i) the thermal gradients; (ii) the softening fraction gradient and (iii) the stress and strain distributions taking into account the mechanical properties of tungsten (elastic-viscoplastic) and the softening phenomenon. This paper describes the results obtained for a range of WEST steady state parallel heat flux (from 45 MW/ m2 to 70 MW/ m2 ) and disruption (600 MW/ m2 ) heat loading on the leading edge. Estimated results related to the plastic strain accumulation, give an interpretation of the premature cracking of the monoblock leading edge but do not explain the poloidal distribution. According to the numerical results, brittle fracture are expected under disruption as estimated normal stresses are beyond the material yield stress, along 88% of the leading edge.

International audience; During the 2019 experimental campaign, the WEST tokamak was partially equipped with ITER-like plasma facing units (PFUs) made of discrete monoblocks (MBs) in one of the twelve lower divertor sectors. The magnetic field lines can enter the gaps between two MBs and strike their leading edges (LEs) with near normal incidence, leading to high localized heat flux, temperature and thermomechanical stress during both stead-state operation and transients. Exposed leading edges are a particular matter of concern because of potential crack formation, recrystallization or even melting of tungsten. During the 2019 experimental campaign in WEST, five of them were misaligned (vertical misalignment h = 0.30 ± 0.1 mm) with their poloidal leading edges exposed to the plasma heat flux. A medium wavelength IR filter (MWIR: 3.9 ± 0.1 µm) was installed in the very high resolution infrared system, featuring a submillimeter spatial resolution (~0.1mm/pixel). This system has a temperature detection threshold of T threshold, BB ≈ 250°C. In this paper, thermal analysis will be presented with a specific focus on overheating of poloidal and toroidal edges using post-mortem measured emissivity maps. The study reveals an unexpected hot spot in the top LE corner of misaligned PFUs at the outer strike point (OSP). Photonic calculations were performed with different emissivity on top and lateral surfaces, in order to consider the complex reflectance characteristics in the toroidal gaps. During this experiment, the Larmor radius (0.35 mm) is comparable to the gap size. Thus, the helical trajectory of the particles may significantly affect the heat load distribution in the edges vicinity. Heat load simulations were performed with ion orbit modeling that consider the Larmor radius, to study its impact on the thermal distribution of the MB. The photonic simulations showed that it was possible to observe a false hot spot at the top LE corner, with both optical approximation (OA) and ion orbit (IO) modeling, due to reflections and different emissivity values.

We study the non-equilibrium steady-states of a one-dimensional ($1D1V$) fluid in a finite space region of length $L$. Particles interact among themselves by multi-particle collisions and are in contact with two thermal-wall heat reservoirs, located at the boundaries of the region. After an initial ballistic regime, we find a crossover from a normal (kinetic) transport regime to an anomalous (hydrodynamic) one, above a characteristic size $L_*$. We argue that $L_*$ is proportional to the cube of the collision time among particles. Motivated by the physics of emissive divertors in fusion plasma, we consider the more general case of thermal walls injecting particles with given average (non-thermal) velocity. For fast and relatively cold particles, short systems fail to establish local equilibrium and display non-Maxwellian distributions of velocities., Accepted for publication in Physical Review Research

see also: http://west.cea.fr/WESTteam; International audience; In the WEST tokamak, ITER-like divertor targets consisting of tungsten monoblocks bonded via an OHFC-Cu compliance layer to CuCrZr cooling tubes were exposed to plasma during the 2018 experimental campaign in which modest heating power was available. Up to 2.5 MW/m$^2$ divertor surface heat flux was attained. Inspection of the components after the campaign revealed a wide variety of damage at both leading and trailing monoblock edges, and at the optical hot spots which are the projections along magnetic field lines of the toroidal gaps between monoblocks onto the poloidal leading edges. Cracking, deformation, and melting occurred. Consideration of the large body of past work on high heat flux testing, combined with the expected loading conditions in WEST, suggests that fractures form during the first transient events such as disruptions. Deformation occurs under subsequent exposure to steady state heat loads. Nearly identical damage was observed on reciprocating probes made of W-La(10%) alloy under measured irradiation conditions, lending credence to this hypothesis.

The design of the Langmuir probes for the lower divertor of the WEST tokamak is presented, in which uncooled, tungsten-coated graphite targets were installed for the first phase of operation in anticipation of full-tungsten, actively cooled, ITER-like divertor components. Due to the long plasma discharges and high heat loads, the probe tips are flush with the divertor tiles in which they are embedded, as it is foreseen and indeed required in ITER. Tantalum metal was chosen for the probe tips because of its high melting point, and ductility at room temperature which allows simple crimping to electrically connect them to cables, as well as being compatible with the full metal environment of WEST. High heat flux tests of Ta Langmuir probe prototypes confirmed the robustness of the design for energy flux densities up to 8 MW/m2 for 5.5 s, and at lower 3.5 MW/m² energy flux density for 40 s long discharges. The probes were operational from the first discharge in WEST and the good alignment of all probes with the surrounding surface was confirmed by a good match of each adjacent probe's data when the strike points were swept across the target.

International audience; Nitrogen (N2) will be used in ITER to enhance the radiative fraction to ~90%, thereby cooling the edge plasma and preventing damage to the plasma-facing components. However, the reactivity of N2 with hydrogen isotopes can lead to the formation of tritiated ammonia (NT3). This should be considered in terms of the in-vessel tritium inventory, the regeneration of the cryo pumps, and the processes in the ITER de-tritiation plant. In the 'W' Environment in Steady-state Tokamak (WEST), a series of long L-mode discharges (~50 s), with a constant N2 seeding from the outer strike point region has been performed on the upper actively cooled divertor. In the absence of active pumping, the N2 balance shows steady-state retention during plasma discharge, and is partially (~35%) released in between discharges. Although a significant amount of N2( 18.65 Pa m3) has been injected, the wall still exhibited N2 pumping capabilities. Under these conditions, as long as this N2 reservoir is not saturated, there is not enough N available for the detectable threshold of ND3 formation to be reached. In these WEST experiments, no ammonia is detected during the pulse or after the pulse in the outgassing phase. These results are consistent with and complementary to the N2 seeded experiments performed in the Joint European Torus (JET) with its ITER-like wall and in the Axially Symmetric Divertor Experiment (ASDEX) upgrade.

International audience; Twelve ITER-like plasma-facing units made of tungsten were exposed in the WEST tokamak divertor, with three PFUs significantly overexposed to plasma heat flux: one sharp-edged PFU (vertical misalignment h = 0.8 mm) and two chamfered PFUs (h = 0.6 mm and 0.3 mm, respectively). This paper describes the first temperature analysis obtained with a very high spatial resolution infrared camera (pixel size ~ 0.1 mm) on the misaligned PFU edges and shows the consistency obtained on the parallel heat flux derived from these measurements. The analysis is focused on the hottest areas of the PFU misaligned leading edges, since the temperature detection threshold of the VHR camera is high (Tthreshold,BB ≈ 370°C). The heat flux parallel to the magnetic field lines is assessed by matching the toroidal temperature profile in the vicinity of the leading edge with 3D finite element modelling. The tungsten emissivity assumed in this study is 0.6, consistent with laboratory measurements for damaged PFUs. For the three PFUs studied (with different vertical misalignments, incident angles, geometries), the derived parallel heat flux is similar, and consistent with independent measurements by a Fiber Bragg grating embedded in a graphite PFU at another toroidal location, giving confidence in future experiments using the same settings.

The first lower hybrid current drive experiments in the full tungsten WEST tokamak are reported. Good wave coupling is found at rather low plasma current (q95∼4.3) and medium density (ne∼3×1019m-3). Reflection coefficient is in agreement with the expectation from the linear theory of coupling. With low reflection coefficients, 5MW was coupled for 2 seconds. High central electron temperature, up to 5keV, is achieved at ne = 3-4×1019m-3. Flat and even hollow profiles of tungsten density are derived from the bolometry diagnostic. The stored thermal energy follows the H96-P scaling law with very slight degradation with density. The current drive efficiency has been assessed in low loop voltage (VL∼0.15V) discharges. Low plasma current operation and rather high effective charge (Zeff∼3) lead to modest current drive efficiency (ƞ = 0.5-0.65 × 1019 A.W−1m-2). Long pulse operation (∼30s) with high LHCD power (PLH=2.7MW) is achieved with stationary parameters, in particular electron and impurity densities, with the upper water-cooled tungsten divertor.

This study compares experimental observations on the two challenges of Ion Cyclotron Resonance Heating (ICRH), namely the coupling of waves to the plasma and the enhanced impurity sputtering in EAST and WEST medium size tokamaks.In WEST, Lower Hybrid (LH) power helps improving ICRH coupling. In both machines experiments reveal that fueling from the midplane not only helps to couple waves from nearby antennas like in other devices, but also has an impact on the scrape-off layer (SOL) density in regions that are not magnetically connected to the valves. Localized midplane nozzle valves allow similar (in WEST) or better (in EAST) coupling compared to poloidally distributed valves. Core density control requirements for long-pulse operation, in particular in L-mode regime, however limit the amount of gas that can be injected.During ICRH impurities can contaminate the plasma up to a level detrimental for the operation, e.g. up to 100% of ICRH power can be radiated on WEST in certain high power conditions. In WEST, tungsten (W) production measured by visible spectroscopy increases on all the observed objects during ICRH compared to a reference phase without ICRH. On some components (antenna side limiters, baffle, divertor) the rise is larger than with a similar LH power. The relative contribution of each object and physical process (RF-sheaths, fast ion ripple losses) to core contamination yet remains poorly known. Comparing antenna limiters with W-coating vs low-Z materials would help quantifying the role of these components. In EAST, the core W content, measured by EUV spectroscopy in presence of divertor sources only, is correlated with the total injected power, either from ICRH or LH. Since 2018 the LH guard limiter tiles were W-coated. Their contribution to the core W content appears more important than divertor sources when two-strap ICRH antennas magnetically connected to W components at the midplane is powered, compromising high performance operations.

The shaping of the ITER divertor vertical targets has been refined as a consequence of manufacturing and engineering considerations during the prototype manufacturing activities. In this paper, the optimized ITER divertor design is presented together with design validation by 3D field line tracing calculation and thermo-mechanical analysis by finite element calculations. Furthermore, the reduction of W monoblock armour thickness to 6 mm is also discussed.

The consequences of tungsten (W) melting on divertor lifetime and plasma operation are high priority issues for ITER. Sustained and controlled W-melting experiment has been achieved for the first time in WEST on a poloidal sharp leading edge of an actively cooled ITER-like plasma facing unit (PFU). A series of dedicated high power steady state plasma discharges were performed to reach the melting point of tungsten. The leading edge was exposed to a parallel heat flux of about 100 MW.m−2 for up to 5 s providing a melt phase of about 2 s without noticeable impact of melting on plasma operation (radiated power and tungsten impurity content remained stable at constant input power) and no melt ejection were observed. The surface temperature of the MB was monitored by a high spatial resolution (0.1 mm/pixel) infrared camera viewing the melt zone from the top of the machine. The melting discharge was repeated three times resulting in about 6 s accumulated melting duration leading to material displacement from three similar pools. Cumulated on the overall sustained melting periods, this leads to excavation depth of about 230 μm followed by a re-solidified tungsten bump of 200 μm in the JxB direction.

The relation between turbulent transport and scrape off layer width is investigated in circular plasmas toroidally limited on the inner wall. A broad set of experimental observations collected in the Tore Supra scrape off layer is detailed and compared to turbulent interchange models. Blob E × B drift velocities measured in experiments agree reasonably well with an analytical model derived for isolated blobs. Based on a time averaged particle flux balance, it is also shown that the SOL width depends on both the blob drift velocity and a blob intermittency parameter, which is so far not predicted by isolated blob models. A set of 2D isothermal turbulence simulations are used to derive a power law regression of the density width function of global control parameters. Quantitative agreement is found between this regression and experimental density widths measured in Tore Supra, over a large set of plasma conditions. The sensitivity to control parameters (major radius, safety factor and normalized Larmor radius) is roughly explained by the sensitivity of the blob velocity model. The predictions are also extended to power decay length in limited plasma configurations. For ITER start-up phases, the predicted power decay length fall in the range of extrapolations based on multi-machine regressions.

The high heat flux areas on the vertical divertor targets in the ITER tokamak will consist of cuboid tungsten monoblocks bonded to copper cooling tubes. Three-dimensional ion orbit modelling is used to calculate the heating of tungsten monoblocks during ELMs at the inner vertical target, where the highest surface energy densities are expected. The presence of thin gaps between monoblocks results in exposed edges onto which the heat flux can be focused. ELM ions are focused by their gyromotion onto the magnetically shadowed, long toroidal edges of the monoblocks. The risk of monoblock edge melting is greater than the risk of full surface melting on the plasma-wetted zone. Alternative shaping solutions such as edge chamfering, filleting, and poloidal beveling do not show promise; the melt zone simply migrates to other locations on the monoblocks. Without ELM mitigation, there is a marginal risk of edge melting due to uncontrolled ELMs in the pre-nuclear phase of ITER operation, and an absolute certainty of it in the burning nuclear phase. To avoid edge melting altogether, the surface energy density would have to limited to less than 0.15MJ/m2. Keywords: ITER, Divertor, ELM heat loads

We present first results of SOLEDGE-EIRENE modeling a strongly radiating plasma in the WEST tokamak. Using measurements from a reciprocating Langmuir probe we have determined the SOLEDGE input parameters, i.e. separatrix density at outboard midplane and radial transport coefficients typical of L-mode plasma. We have performed deuterium plasma simulations with presence of oxygen, injected from the core boundary into the simulation domain. The comparison between the radiated power measured from the bolometry in the divertor region and the one computed from SOLEDGE simulations seems to indicate a concentration of oxygen of about 2%. Moreover we have obtained good agreement between simulation results and experimental measurements on electron density and electron temperature profiles at the outer strike point. First qualitative analysis of spectroscopic synthetic diagnostic on Dα signals is also presented. Keywords: Edge tokamak plasma, Transport, Fluid modeling

International audience; In 2018, Ion Cyclotron Resonant Frequency (ICRF) waves were for the first time applied to the WEST plasmascenarios. In ICRF-only plasmas at medium density, or on top of a low level of Lower Hybrid (LH) power, the coupledICRF power increases the plasma energy content. However in discharges with large LH power at high core density, nearlyall the applied ICRF power gets radiated, mainly in the plasma bulk. Both the energy content and conducted power decrease.Two peculiarities of WEST, that may combine, are presently invoked to explain this phenomenology:1) Fast ion ripplelosses, evidenced between TF coils on the baffle, likely degrade the ICRF heating efficiency. 2) RF-enhanced W sources,evidenced on several Plasma-Facing Components (PFCs), likely over-contaminate the high-power plasmas.

This study compares experimental observations on plasma surface interactions and heat loads induced by Ion Cyclotron Resonance Heating (ICRH) in WEST medium size tokamak. When powering ICRH, core contamination by metallic impurities systematically results in an increase of the power radiated, often hardening the achievement of long discharges in H-mode at high power. The radiofrequency (RF) sheaths are excited mainly at proximity of ICRF antennas, resulting in an increase of metallic impurity sources. In the presence of RF sheaths, we show that most of the power reaching the wall can become carried by ions. Therefore, both heat loads and material sputtering are the result of ion fluxes on the various plasma facing components, such that some analogy between both phenomena can be made. By combining infrared with spectroscopic measurements, we show experimental evidences from WEST of RF-induced heat fluxes and their correlation with local production of impurity.

Cross-field transport in edge tokamak plasmas is known to be dominated by turbulent transport. A dedicated effort has been made to simulate this turbulent transport from first principle models but the numerical cost to run these simulations on the ITER scale remains prohibitive. Edge plasma transport study relies mostly nowadays on so-called transport codes where the turbulent transport is taken into account using effective ad-hoc diffusion coeffecients. In this contribution, we propose to introduce a transport equation for the turbulence intensity in SOLEDGE2D-EIRENE to describe the interchange turbulence properties. Going beyond the empirical diffusive model, this system automatically generates profiles for the turbulent transport and hence reduces the number of degrees of freedom for edge plasma transport codes. We draw inspiration from the k-epsilon model widely used in the neutral fluid community. ((c) 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

International audience; Predictions for the operation of tokamak divertors are reliant on edge plasma simulations typically utilizing a fluid plasma code in combination with a Monte Carlo code for neutral species. Pilot-PSI is a linear device operating with a cascaded arc plasma source that produces plasmas comparable to those expected in the ITER divertor (Te ∼ 1 eV, ne ∼ 10 21 m −3). In this study, plasma discharges in Pilot-PSI have been modelled using the Soledge2D fluid plasma code [1] coupled to the Eirene neutral Monte Carlo code. The plasma is generated using an external source of plasma density and power. These input parameters are tuned in order to match Thomson scattering (TS) measurements close to the cascaded arc source nozzle. The sensitivity of the simulations to different atomic physics models was explored. It was found that elastic collisions between ions and hydrogen molecules have a strong influence on calculated profiles. Without their inclusion, supersonic flow regimes are obtained with M ∼ 2 close to the target plate. Simulation results have been compared with experimental findings using TS close to the target and in the case of Pilot-PSI, a Langmuir probe embedded in the target. Comparison between experimental trends observed in a background pressure scan [2] and the simulations show that the inclusion of the elastic collision is mandatory for the trends to be reproduced.

Dedicated experiments have been performed in the COMPASS tokamak to thoroughly study the power deposition processes occurring on poloidal and toroidal edges of castellated plasma-facing components in tokamaks during steady-state L-mode conditions. Surface temperatures measured by a high resolution infra-red camera are compared with reconstructed synthetic data from a 2D thermal model using heat flux profiles derived from both the optical approximation and 2D particle-in-cell (PIC) simulations. In the case of poloidal leading edges, when the contribution from local radiation is taken into account, the parallel heat flux deduced from unperturbed, upstream measurements is fully consistent with the observed temperature increase at the leading edges of various heights, respecting power balance assuming simple projection of the parallel flux density. Smoothing of the heat flux deposition profile due to finite ion Larmor radius predicted by the PIC simulations is found to be weak and the power deposition on misaligned poloidal edges is better described by the optical approximation. This is consistent with an electron-dominated regime associated with a non-ambipolar parallel current flow. In the case of toroidal gap edges, the different contributions of the total incoming flux along the gap have been observed experimentally for the first time. They confirm the results of recent numerical studies performed for ITER showing that in specific cases the heat deposition does not necessarily follow the optical approximation. Indeed, ions can spiral onto the magnetically shadowed toroidal edge. Particle-in-cell simulations emphasize again the role played by local non-ambipolarity in the deposition pattern.

The interaction of ELM filaments with the ITER beryllium first wall panels (FWPs) is studied using a simple ad-hoc fluid model of the filament parallel transport, taking into account the full, three-dimensional structure of the FWPs, including magnetic shadowing effects. The calculated ELM surface heat loads are used as input to the RACLETTE heat transfer code to estimate the FWP surface temperature rise. The results indicate that controlled ELMs in ITER during burning plasma operation (Δ W ELM ≈ 0.6 M J) will not lead to melting or significant evaporation of the beryllium surfaces, even in the case of high ELM broadening and the minimum allowable distance between the primary and secondary separatrices. The ELM-averaged steady-state heat load also stays below the maximum power handling capability of the FWPs.

The effect of the magnetised sheath on tungsten migration and its consequences on net erosion of tungsten plasma facing components are investigated. This study points out a strong effect of the E × B drift on the redeposition location as well as a rapid redeposition of the ejected particles that are ionised close to the surface. A plasma parameter study shows that the migration direction is changed at ≈ 10 eV from opposite to E × B drift at low electron temperature to the same direction as E × B at high electron temperature. Different models have been considered in the case of plasma parameter gradients that are also at the source of a component of the electric field parallel to the surface. New features are observed like an enhanced net erosion peak at the strike point if only the component of the electric field towards the surface is taken into account. When the component parallel to the surface that drives the particles away from or towards the surface is considered, a very different net erosion picture is obtained: a net erosion peak exists between two net deposition regions. This points out the importance of incorporating this effect in simulations. Keywords: Redeposition, Magnetised sheath, Net erosion, Impurity migration, Tungsten

During steady state plasma operation in fusion devices, leading edges of the actively cooled plasma-facing components can be impacted by plasma flux with nearly normal angle of incidence, causing local overheating. The overheating can be a critical issue in high-power machines, especially in the presence of mechanical misalignments. Due to heat diffusion through the material, the edge power overload also leads to a local increase of temperature on the top part of the tile that can be detected by the infrared imaging system (viewed from the top of the machine). In the Tore Supra tokamak, heat flux impinging on the top and the leading edge of the carbon fibre composite (CFC) flat tiles are characterized with both an infrared (IR) thermographic system and 2D thermal modelling of the tile. A specific sensor correction based on a laboratory blackbody-slit experiment has been developed to simulate the spatial resolution related effects (necessary here since the temperature gradient near the leading edge is smaller than the pixel size of the IR system). The transfer function of the IR system is characterized by a Gaussian distribution function. The standard deviation is found to be σ = 1.75 mm for a pixel size of 3.1 mm. The heat flux calculation is applied to CFC flat tiles and, after being processed with the transfer function, compared to experimental IR data for two geometrical situations: one with 0.2 mm misalignment between two adjacent tiles and the other without misalignment (well-aligned tiles). The heat flux ratio between the leading edge and top is found to be ~25 in the case of the protruding tile, which is lower than the expected ratio using the guiding-centre ballistic approximation with no cross-field heat flux (57).

International audience; The effect of the magnetised sheath on the gross erosion and redeposition of tungsten is examined with an insight on impurity impact energy and angle. A complete treatment of the impact energy is performed leading to a scaling that differs from the usual 2k B T i +3Zk B T e formula. It is found that even if the energy distribution at the sheath entrance strongly differs from this approximation, the discrepancy remains under 20% for the impact energy. The average impact angle of a set of impurities is calculated. The difference between models with and without sheath electric field depends strongly on the charge state and mass of the impurity considered. This can lead, for example in the case of gross sputtering due to impinging W 12+ , to an underestimation of 3-4 times at T e = 25 eV for very grazing angles. Scalings for high (n e > 10 20 m −3) and low (n e < 10 15 m −3) density cases are deduced and provide a good estimation of the average impact angle of all impurities. Finally the magnetised sheath is shown to have an effect on the redeposition through two contributions: it increases the prompt redeposition contribution to the total redeposition and constitutes a potential barrier for the ions leaving the surface. A discussion is provided about the optimal width of the simulation domain to capture these effects.

The first results of particle-in-cell simulations of the electrostatic sheath and magnetic pre-sheath of thermionically emitting planar tungsten surfaces in fusion plasmas are presented. Plasma con ...

The WEST (W – for tungsten – Environment in Steady-state Tokamak) project is an upgrade of Tore Supra from a limiter based tokamak with carbon PFCs into an X-point divertor tokamak with full-tungsten armour while keeping its long discharge capability. The WEST project will primarily offer the key capability of testing for the first time the ITER technology in real plasma environment. In particular, the main divertor (i.e. the lower divertor) of the WEST project will be based on actively cooled tungsten monoblock components and will follow as closely as possible the design and the assembling technology, foreseen for the ITER divertor units. The current design of WEST ITER-like tungsten divertor has now reached a mature stage following the 2013 WEST Final Design Review. This paper presents the key elements of the design, reports the technological requirements and reviews the main design and integration issues.