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Plasma-wall interactions at the first wall in fusion experimental devices are critical for the life time of plasma-facing components and tokamak operation. In-situ measurements of these plasma-wall interaction processes are required in long pulse devices in order to understand better the underlying mechanisms as well as monitoring of the tritium content of the first wall as an immediate safety application. In this work we present measurements of the deuterium retention in the first wall of EAST using laser-induced ablation spectroscopy. The diagnostic is applied during plasma operation and the dynamic retention for graphite and tungsten as plasma-facing materials are compared, showing more than three times higher short-term retention in graphite than in tungsten. Laser-based analysis is applied during a shift of the vertical positions of the plasma column in limited as well as in the start-up phase of discharges in diverted magnetic configuration. Plasma core densities of ne = (1.5–3.2) × 1019/m3 are chosen to vary the ion flux to the first wall and study saturation effects of deuterium on the surface first wall materials during tokamak plasma operation. A linear dependence of the retention from ion fluence is observed for graphite as PFC up to 3 × 1018D+/cm2, followed by a constant retention up to fluences of 3.5 × 1019D+/cm2 for nearby first wall temperatures of TPFC ≈ 350 K.

For simulation of plasma-facing component erosion in fusion experiments, an analytical expression for the ion velocity just before the surface impact including the local electric field and an optional surface biasing effect is suggested. Energy and angular impact distributions and the resulting effective sputtering yields were produced for several experimental scenarios at JET ILW mostly involving PFCs exposed to an oblique magnetic field. The analytic solution has been applied as an improvement to earlier ERO modelling of localized, Be outer limiter, RF-enhanced erosion, modulated by toggling of a remote, however magnetically connected ICRH antenna. The effective W sputtering yields due to D and Be ion impact in Type-I and Type-III ELMs and inter-ELM conditions were also estimated using the analytical approach and benchmarked by spectroscopy. The intra-ELM W sputtering flux increases almost 10 times in comparison to the inter-ELM flux. (C) 2017 Published by Elsevier Ltd. This is an open access article under the CC BY-NC-ND license. For complete list of authors see http://dx.doi.org/10.1016/j.nme.2017.03.031

The first divertor was installed in the JET machine between 1992 and 1994 and was operated with carbon tiles and then beryllium tiles in 1994-5. Post-mortem studies after these first experiments demonstrated that most of the impurities deposited in the divertor originate in the main chamber, and that asymmetric deposition patterns generally favouring the inner divertor region result from drift in the scrape-off layer. A new monolithic divertor structure was installed in 1996 which produced heavy deposition at shadowed areas in the inner divertor corner, which is where the majority of the tritium was trapped by co-deposition during the deuterium-tritium experiment in 1997. Different divertor geometries have been tested since such as the Gas-Box and High-Delta divertors; a principle objective has been to predict plasma behaviour, transport and tritium retention in ITER. Transport modelling experiments were carried out at the end of four campaigns by puffing C-13-labelled methane, and a range of diagnostics such as quartz-microbalance and rotating collectors have been installed to add time resolution to the post-mortem analyses. The study of material migration after D-D and D-T campaigns clearly revealed important consequences of fuel retention in the presence of carbon walls. They gave a strong impulse to make a fundamental change of wall materials. In 2010 the carbon divertor and wall tiles were removed and replaced with tiles with Be or W surfaces for the ITER-Like Wall Project. For complete list of authors see http://dx.doi.org/10.1016/j.fusengdes.2018.10.002

There is experimental evidence that the pedestal dynamics in type-I ELMy H-mode discharges is significantly affected by a change in the recycling conditions at the tungsten plasma-facing components ...

First wall materials in controlled fusion devices undergo serious modification by several physical and chemical processes arising from plasma-wall interactions. Detailed information is required for ...

In the version of this Article originally published, and in the associated Publisher Correction, the members of the W7-X Team were not included. All versions of the Article, and the Publisher Correction, have now been amended to include these team members.

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 lambda(div)(q). Otherwise and further upstream the mid-plane heat decay length lambda(q) parameter is not affected by any change in divertor dissipation. This confirms the assumptions made for the ITER divertor design studies, i.e. that lambda(q) 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 lambda(ne)>lambda(Te)>lambda(q) hold in the near-SOL upstream. Thus for interpretative edge plasma simulations one must take the (experimental) value of lambda(ne) into account, rather than lambda(q), as the former actually defines the required minimum upstream SOL grid extent. For complete list of authors see http://dx.doi.org/10.1016/j.nme.2018.10.013

The JET ITER-Like Wall experiment (JET-ILW) provides an ideal test bed to investigate plasma-surface interaction (PSI) and plasma operation with the ITER plasma-facing material selection employing beryllium in the main chamber and tungsten in the divertor. The main PSI processes: material erosion and migration, (b) fuel recycling and retention, (c) impurity concentration and radiation have be1en studied and compared between JET-C and JET-ILW. The current physics understanding of these key processes in the JET-ILW revealed that both interpretation of previously obtained carbon results (JET-C) and predictions to ITER need to be revisited. The impact of the first-wall material on the plasma was underestimated. Main observations are: (a) low primary erosion source in H-mode plasmas and reduction of the material migration from the main chamber to the divertor ( factor 7 ) as well as within the divertor from plasma-facing to remote areas ( factor 30 - 50 ). The energetic threshold for beryllium sputtering minimises the primary erosion source and inhibits multi-step re-erosion in the divertor. The physical sputtering yield of tungsten is low as 10 - 5 and determined by beryllium ions. (b) Reduction of the long-term fuel retention ( factor 10 - 20 ) in JET-ILW with respect to JET-C. The remaining retention is caused by implantation and co-deposition with beryllium and residual impurities. Outgassing has gained importance and impacts on the recycling properties of beryllium and tungsten. (c) The low effective plasma charge ( Z eff = 1.2 ) and low radiation capability of beryllium reveal the bare deuterium plasma physics. Moderate nitrogen seeding, reaching Z eff = 1.6 , restores in particular the confinement and the L-H threshold behaviour. ITER-compatible divertor conditions with stable semi-detachment were obtained owing to a higher density limit with ILW. Overall JET demonstrated successful plasma operation in the Be/W material combination and confirms its advantageous PSI behaviour and gives strong support to the ITER material selection.

The fuel outgassing efficiency from plasma-facing components exposed in JET-ILW has been studied at ITER-relevant baking temperatures. Samples retrieved from the W divertor and Be main chamber were annealed at 350 and 240 °C, respectively. Annealing was performed with thermal desoprtion spectrometry (TDS) for 0, 5 and 15 h to study the deuterium removal effectiveness at the nominal baking temperatures. The remained fraction was determined by emptying the samples fully of deuterium by heating W and Be samples up to 1000 and 775 °C, respectively. Results showed the deposits in the divertor having an increasing effect to the remaining retention at temperatures above baking. Highest remaining fractions 54 and 87% were observed with deposit thicknesses of 10 and 40 μm, respectively. Substantially high fractions were obtained in the main chamber samples from the deposit-free erosion zone of the limiter midplane, in which the dominant fuel retention mechanism is via implantation: 15 h annealing resulted in retained deuterium higher than 90%. TDS results from the divertor were simulated with TMAP7 calculations. The spectra were modelled with three deuterium activation energies resulting in good agreement with the experiments EURATOM 633053

The divertor solutions obtained with the plasma edge modelling tool SOLPS5.0 are discussed. The code results are benchmarked against carefully analysed L-mode discharges at various density levels with and without impurity seeding in the full-metal tokamaks ASDEX Upgrade and JET. The role of the cross-field drifts and currents in the solutions is analysed in detail, and the improvements achieved by fully activating the drift and current terms in view of matching the experimental signals are addressed. The persisting discrepancies are also discussed.

The JET ITER-Like Wall experiment, with its all-metal plasma-facing components, provides a unique environment for plasma and plasma-wall interaction studies. These studies are of great importance in understanding the underlying phenomena taking place during the operation of a future fusion reactor. Present work summarizes and reports the plasma fuel retention in the divertor resulting from the two first experimental campaigns with the ITER-Like Wall. The deposition pattern in the divertor after the second campaign shows same trend as was observed after the first campaign: highest deposition of 10-15 μm was found on the top part of the inner divertor. Due to the change in plasma magnetic configurations from the first to the second campaign, and the resulted strike point locations, an increase of deposition was observed on the base of the divertor. The deuterium retention was found to be affected by the hydrogen plasma experiments done at the end of second experimental campaign.

The isotope content of the plasma reflects on the dynamics of isotope changeover experiments, efficiency of wall conditioning and the performance of a fusion device in the active phase of operation ...

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, P-heat, 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 Z(eff)(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, D SOL, and the core impurity inward pinch, v(pinch). D-SOL had to be increased with increasing Gamma(Ne) and the level of v(pinch) had to be changed, for any given Gamma(Ne), according to the level of P-heat : it decreases with increasing P-heat. Since the ELM frequency, f(ELM), is experimentally correlated with P-heat, (it increases with P-heat) the impurity inward pinch can be seen as to depend on f(ELM). Therefore, to maintain a low v(pinch) level (i.e. high f(ELM)) Gamma(Ne)/P-heat should not exceed a certain threshold, which slightly increases with the Gamma(D) puffing rate. This might lead to a limitation in the viability of reducing the target heat load by Ne seeding at moderate Gamma(D), while keeping Z(eff) at acceptably low level. (C) 2016 Published by Elsevier Ltd. For complete list of authors see http://dx.doi.org/10.1016/j.nme.2016.10.012

ERO is a Monte-Carlo code for modeling plasma-wall interaction and 3D plasma impurity transport for applications in fusion research. The code has undergone a significant upgrade (ERO2.0) which allows increasing the simulation volume in order to cover the entire plasma edge of a fusion device, allowing a more self-consistent treatment of impurity transport and comparison with a larger number and variety of experimental diagnostics. In this contribution, the physics-relevant technical innovations of the new code version are described and discussed. The new capabilities of the code are demonstrated by modeling of beryllium (Be) erosion of the main wall during JET limiter discharges. Results for erosion patterns along the limiter surfaces and global Be transport including incident particle distributions are presented. A novel synthetic diagnostic, which mimics experimental wide-angle 2D camera images, is presented and used for validating various aspects of the code, including erosion, magnetic shadowing, non-local impurity transport, and light emission simulation. EURATOM 633053

A set of quartz crystal microbalances (QMB) was used at JET with full carbon wall to monitor mass erosion/ deposition rates in the remote areas of the divertor. After introduction of the ITER-like wall (ILW) in JET with beryllium main wall and tungsten divertor, strong reduction of the material deposition and accompanied fuel retention was observed. Therefore the existing QMB electronics have been modified to improve the accuracy of frequency measurements by a factor of ten down to 0.1 Hz which corresponds to 1.4 ng cm(-2). The averaged deposition rates of 1.2-3 ng cm(-2) s(-1) and erosion rates of 5.6-8.1 ng cm(-2) s(-1) were observed in the inner divertor of JET-ILW with the inner strike point positions close to the bottom edge of vertical tile 3 and at the horizontal tile 4 respectively. The erosion with averaged rates of approximate to 2.1 ng cm(-2) s(-1) and approximate to 120 ng cm(-2) s(-1) were observed in the outer divertor for the outer strike point positions at tile 5 and tile 6 respectively. (C) 2017 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY-NC-ND license. For complete list of authors see http://dx.doi.org/10.1016/j.nme.2017.03.007

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 lev ...

Since 2011 the JET tokamak has been operated with a metal ITER-like wall (JET-ILW) including castellated beryllium limiters and lamellae-type bulk tungsten tiles in the divertor. This has allowed for a large scale test of castellated plasma-facing components (PFC). Procedures for sectioning the limiters into single blocks of castellation have been developed. This facilitated morphology studies of morphology of surfaces inside the grooves for limiters after experimental campaigns 2011–2012 and 2013–2014. The deposition in the 0.4–0.5mm wide grooves of the castellation is ‘shallow’. It reaches 1–2mm into the 12mm deep gap. Deuterium concentrations are small (mostly below 1 × 1018 cm−2 ). The estimated total amount of deuterium in all the castellated limiters does not exceed the inventory of the plasma-facing surfaces (PFS) of the limiters. There are only traces of Ni, Cr and Fe deposited in the castellation gaps. The same applies to the carbon content. Also low deposition of D, Be and C has been measured on the sides of the bulk tungsten lamellae pieces. Modelling clearly reflects: (a) a sharp decrease in the measured deposition profiles and (b) an increase in deposition with the gap width. Both experimental and modelling data give a strong indication and information to ITER that narrow gaps in the castellated PFC are essential. X-ray diffraction on PFS has clearly shown two distinct composition patterns: Be with an admixture of Be–W intermetallic compounds (e.g. Be22W) in the deposition zone, whilst only pure Be has been detected in the erosion zone. The lack of compound formation in the erosion zone indicates that no distinct changes in the thermo-mechanical properties of the Be PFC might be expected. EURATOM 633053 Swedish Research Council (VR) 2015–04844

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 P-sep/R > 15 MW/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 P-sep. For complete list of authors see http://dx.doi.org/10.1016/j.nme.2017.03.033

Post-mortem studies with ion beam analysis, thermal desorption, and secondary ion mass spectrometry have been applied for investigating the long-term fuel retention in the JET ITERlike wall components. The retention takes place via implantation and co-deposition, and the highest retention values were found to correlate with the thickness of the deposited impurity layers. From the total amount of retained D fuel over half was detected in the divertor region. The majority of the retained D is on the top surface of the inner divertor, whereas the least retention was measured in the main chamber on the mid-plane of the inner wall limiter. The recessed areas of the inner wall showed significant contribution to the main chamber total retention. Thermal desorption spectroscopy analysis revealed the energetic T from DD reactions being implanted in the divertor. The total T inventory was assessed to be >0.3 mg.

Global gas balance experiments at ASDEX Upgrade (AUG) and JET have shown that a considerable fraction of nitrogen injected for radiative cooling is not recovered as N2 upon regeneration of the liquid helium cryo pump. The most probable loss channels are ion implantation into plasma-facing materials, co-deposition and ammonia formation. These three mechanisms are investigated in laboratory and tokamak experiments and by numerical simulations. Laboratory experiments have shown that implantation of nitrogen ions into beryllium and tungsten leads to the formation of surface nitrides, which may decompose under thermal loads. On beryllium the presence of nitrogen at the surface has been seen to reduce the sputtering yield. On tungsten surfaces it has been observed that the presence of nitrogen can increase hydrogen retention. The global nitrogen retention in AUG by implantation into the tungsten surfaces saturates. At JET the steady state nitrogen retention is increased by co-deposition with beryllium. The tokamak experiments are interpreted in detail by simulations of the global migration with WallDYN. Mass spectrometry of the exhaust gas of AUG and JET has revealed the conversion of nitrogen to ammonia at percent-levels. Conclusions are drawn on the potential implications of nitrogen seeding on the operation of a reactor in a deuterium-tritium mix.

Rotating collectors and quartz microbalances (QMBs) are used in JET to provide time-dependent measurements of erosion and deposition. Rotation of collector discs behind apertures allows recording of the long term evolution of deposition. QMBs measure mass change via the frequency deviations of vibrating quartz crystals. These diagnostics are used to investigate erosion/deposition during JET-C carbon operation and JET-ILW (ITER-like wall) beryllium/ tungsten operation. A simple geometrical model utilising experimental data is used to model the time-dependent collector deposition profiles, demonstrating good qualitative agreement with experimental results. Overall, the JET-ILW collector deposition is reduced by an order of magnitude relative to JET-C, with beryllium replacing carbon as the dominant deposit. However, contrary to JET-C, in JET-ILW there is more deposition on the outer collector than the inner. This reversal of deposition asymmetry is investigated using an analysis of QMB data and is attributed to the different chemical properties of carbon and beryllium. EURATOM 633053

Experiments in the JET tokamak equipped with the ITER-like wall (ILW) revealed that the inner and outer target plate at the location of the strike points represent after one year of operation intact tungsten (W) surfaces without any beryllium (Be) surface coverage. The dynamics of near-surface retention, implantation, desorption and recycling of deuterium (D) in the divertor of plasma discharges are determined by W target plates. As the W plasma-facing components (PFCs) are not actively cooled, the surface temperature (T-surface) is increasing with plasma exposure, varying the balance between these processes in addition to the impinging deuteron fluxes and energies. The dynamic behaviour on a slow time scale of seconds was quantified in a series of identical L-mode discharges (JET Pulse Number (JPN)#81938-73) by intra-shot gas analysis providing the reduction of deuterium retention in W PFCs by 1/3 at a base temperature (T-base) range at the outer target plate between 65 degrees C and 150 degrees C equivalent to a T-surface span of 150 degrees C and 420 degrees C. The associated recycling and molecular D desorption during the discharge varies only at lowest temperatures moderately, whereas desorption between discharges rises significantly with increasing T-base. The retention measurements represent the sum of inner and outer divertor interaction at comparable T-surface. The dynamic behaviour on a fast time scale of ms was studied in a series of identical H-mode discharges (JPN #83623-83974) and coherent edge-localized mode (ELM) averaging. High energetic ELMs of about 3 keV are impacting on the W PFCs with fluxes of 3 x 10(23) D+ s(-1) m(-2) which is about four times higher than inter-ELM ion fluxes with an impact energy of about E-im = 200 eV. This intra-ELM ion flux is associated with a high heat flux of about 60 MW m(-2) to the outer target plate which causes T-surface rise by Delta T = 100 K per ELM covering finally the range between 160 degrees C and 1400 degrees C during the flat-top phase. ELM-induced desorption from saturated near-surface implantation regions as well as deep ELM-induced deuterium implantation areas under varying baseline temperature takes place. Subsequent refuelling by intra-ELM deuteron fluxes occurs and a complex interplay between deuterium fuelling and desorption can be observed in the temporal ELM footprint of the surface temperature (IR thermography), the impinging deuteron flux (Langmuir probes), and the Balmer radiation (emission spectroscopy) as representative for the deuterium recycling flux. In contrast to JET-C, a pronounced second peak, similar or equal to 8 ms delayed with respect to the initial ELM crash, in the D-alpha radiation and the ion flux has been observed. The peak can be related to desorption of implanted energetic intra-ELM D+ diffusing to the W surface, and performing local recycling. For complete list of authors see http://dx.doi.org/10.1088/0031-8949/T167/1/014076

Physical and chemical assisted physical sputtering were characterised by the Be I and Be II line and BeD band emission in the observation chord measuring the sightline integrated emission in front of the inner beryllium limiter at the torus midplane. The 3D local transport and plasma-surface interaction Monte-Carlo modelling (ERO code [18]) is a key for the interpretation of the observations in the vicinity of the shaped solid Be limiter. The plasma parameter variation (density scan) in limiter regime has provided a useful material for the simulation benchmark. The improved background plasma parameters input, the new analytical expression for particle tracking in the sheath region and implementation of the BeD release into ERO has helped to clarify some deviations between modelling and experiments encountered in the previous studies [4,5]. Reproducing the observations provides additional confidence in our 'ERO-min' fit for the physical sputtering yields for the plasma-wetted areas based on simulated data. For complete list of authors see http://dx.doi.org/10.1016/j.nme.2016.08.013

An overview is given on the recent progress on edge modelling activities for the JET ITER-like wall using the computational tools like the SOLPS or EDGE2D-EIRENE code. The validation process of the ...

The W source strength in JET H-mode discharges depends on the W sputtering in the inter and the intra-ELM phase due to impinging hydrogenic ions (D or H) and impurities (mainly Be). The analytical ...

Erosion and deposition were studied in the JET divertor during the first JET ITER-like wall campaign 2011 to 2012 using marker tiles. An almost complete poloidal section consisting of tiles 0, 1, 3, 4, 6, 7, 8 was studied. The data from divertor tile surfaces were completed by the analysis of samples from remote divertor areas and from the inner wall cladding. The total mass of material deposited in the divertor decreased by a factor of 4-9 compared to the deposition of carbon during all-carbon JET operation before 2010. Deposits in 2011 to 2012 consist mainly of beryllium with 5-20 at.% of carbon and oxygen, respectively, and small amounts of Ni, Cr, Fe and W. This decrease of material deposition in the divertor is accompanied by a decrease of total deuterium retention inside the JET vessel by a factor of 10 to 20. The detailed erosion/deposition pattern in the divertor with the ITER-like wall configuration shows rigorous changes compared to the pattern with the all-carbon JET configuration.

The migration of first wall material due to erosion, plasma transport and re-deposition is one of the key challenges in current and future fusion devices. To predict erosion/re-deposition patterns and to understand the underlying principal processes, the global simulation code WallDYN was developed. It couples the evolution of the first wall surface composition to plasma impurity transport. To benchmark the WallDYN model, it was applied to the JET ITER-like wall experiment (JET-ILW), which mimics the ITER first wall material configuration and is thus an ideal environment to validate the predictive significance of WallDYN calculations for ITER application. The WallDYN calculations show good agreement with the Be deposition patterns determined from JET-ILW post-campaign wall tile analysis. The WallDYN results on W erosion and deposition also qualitatively match the post mortem patterns but quantitatively fall short by a factor of 20 which can be consolidated by including the influence of ELMs on the W erosion source. Applying the same model and process physics as for the JET calculations, the impurity migration and resulting fuel species co-deposition in ITER for different wall configurations and background plasmas were calculated. The simulations show that C containing wall configurations lead to unacceptable T retention whereas for the current ITER material choice (Be wall and W divertor) co-deposition will not limit the ITER operation. However the erosion of W by self-sputtering even without the contribution of ELMs could hamper ITER operation under certain plasma scenarios.

The erosion of tungsten (W), induced by the bombardment of plasma and impurity particles, determines the lifetime of plasma-facing components as well as impacting on plasma performance by the influx of W into the confined region. The screening of W by the divertor and the transport of W in the plasma determines largely the W content in the plasma core, but the W source strength itself has a vital impact on this process. The JET tokamak experiment provides access to a large set of W erosion-determining parameters and permits a detailed description of the W source in the divertor closest to the ITER one: (i) effective sputtering yields and fluxes as function of impact energy of intrinsic (Be, C) and extrinsic (Ne, N) impurities as well as hydrogenic isotopes (H, D) are determined and predictions for the tritium (T) isotope are made. This includes the quantification of intra- and inter-edge localised mode (ELM) contributions to the total W source in H-mode plasmas which vary owing to the complex flux compositions and energy distributions in the corresponding phases. The sputtering threshold behaviour and the spectroscopic composition analysis provides an insight in the dominating species and plasma phases causing W erosion. (ii) The interplay between the net and gross W erosion source is discussed considering (prompt) re-deposition, thus, the immediate return of W ions back to the surface due to their large Larmor radius, and surface roughness, thus, the difference between smooth bulk-W and rough W-coating components used in the JET divertor. Both effects impact on the balance equation of local W erosion and deposition. (iii) Post-mortem analysis reveals the net erosion/deposition pattern and the W migration paths over long periods of plasma operation identifying the net W transport to remote areas. This W transport is related to the divertor plasma regime, e.g. attached operation with high impact energies of impinging particles or detached operation, as well as to the applied magnetic configuration in the divertor, e.g. close divertor with good geometrical screening of W or open divertor configuration with poor screening. JET equipped with the ITER-like wall (ILW) provided unique access to the net W erosion rate within a series of 151 subsequent H-mode discharges (magnetic field: T, plasma current: MA, auxiliary power MW) in one magnetic configuration accumulating 900 s of plasma with particle fluences in the range of 5– in the semi-detached inner and attached outer divertor leg. The comparison of W spectroscopy in the intra-ELM and inter-ELM phases with post-mortem analysis of W marker tiles provides a set of gross and net W erosion values at the outer target plate. ERO code simulations are applied to both divertor leg conditions and reproduce the erosion/deposition pattern as well as confirm the high experimentally observed prompt W re-deposition factors of more than 95% in the intra- and inter-ELM phase of the unseeded deuterium H-mode plasma. Conclusions to the expected divertor conditions in ITER as well as to the JET operation in the DT plasma mixture are drawn on basis of this unique benchmark experiment.

Équipe 107 : Physique des plasmas chauds; International audience; The ITER baseline scenario, with 500 MW of DT fusion power and Q = 10, will rely on a Type I ELMy H-mode, with Delta W = 0.7 MJ mitigated edge localized modes (ELMs). Tungsten (W) is the material now decided for the divertor plasma-facing components from the start of plasma operations. W atoms sputtered from divertor targets during ELMs are expected to be the dominant source under the partially detached divertor conditions required for safe ITER operation. W impurity concentration in the plasma core can dramatically degrade its performance and lead to potentially damaging disruptions. Understanding the physics of plasma-wall interaction during ELMs is important and a primary input for this is the energy of incoming ions during an ELM event. In this paper, coupled Infrared thermography and Langmuir Probe (LP) measurements in JET-ITER-Like-Wall unseeded H-mode experiments with ITER relevant ELM energy drop have been used to estimate the impact energy of deuterium ions (D+) on the divertor target. This analysis gives an ion energy of several keV during ELMs, which makes D+ responsible for most of the W sputtering in unseeded H-mode discharges. These LP measurements were possible because of the low electron temperature (T-e) during ELMs which allowed saturation of the ion current. Although at first sight surprising, the observation of low T-e at the divertor target during ELMs is consistent with the `Free-Streaming' kinetic model which predicts a near-complete transfer of parallel energy from electrons to ions in order to maintain quasi-neutrality of the ELM filaments while they are transported to the divertor targets.