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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.

The carbon deposition profile on a Directional Material Probe (DMP) installed in the inboard side of the torus in the Large Helical Device (LHD) is investigated using the ERO2.0 code. The experimental result of the carbon deposition profile with short and wide shadows (lower deposition density areas) on the DMP is reasonably explained by the carbons sputtered from the carbon divertor plates installed in the inboard side. The simulation proves that the short and wide shadows are produced by carbons sputtered from the right and left divertor plate arrays, respectively. The experimental carbon deposition profile accumulated in the previous experimental campaign (FY2010) was successfully reproduced by the simulation, which provides detailed understanding of material (carbon) migration in the divertor region in the LHD.

Based on the conventional model of hydrogen retention in plasma-facing components, the question of hydrogen outgassing during and after plasma exposure is addressed in relation to mass spectrometry and laser-induced breakdown sprectroscopy (LIBS) measurements. Fundamental differences in retention and release data acquired by LIBS and by mass spectrometry are described analytically and by modelling. Reaction-diffusion simulations are presented that demonstrate possible thermal outgassing effects caused by LIBS. Advantages and limitations of LIBS as a tool for analysis of short term retention are discussed.

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

Controlled radiative power exhaust with impurity seeding has been demonstrated in island divertor configurations at the stellarator Wendelstein 7-X. Detachment has been established with neon (Ne) and nitrogen (N2) seeding for the price of a loss of ΔWp>-15% in the stored energy [1]. Ne seeding allows for sustained enhancement of edge radiation with (frad>80%) a prolonged decay of line emission of several tens of seconds after the end of the injection indicating high recycling of this noble gas at the carbon main plasma facing components (PFCs). In N2-seeded discharges, it is shown that a response of line emission and plasma parameters is in correlation to the puff duration, which indicates a higher level of absorption of this seeding gas in the PFCs. Continuous N2 seeding results in global cooling of the scrape-off layer (SOL) and decay of radiation over several seconds after the injection. These results have been obtained by local gas injection into one of the five independent island chains.Zeff increases from 1.2 to 2.1-2.2 at maximum radiation with both seeding gases and reduces after the injection only in the case of N2 seeding. Damping of the counter-streaming SOL island flows is in correlation with a reduction of the divertor particle fluxes in response to impurity seeding. A correlated reduction of density has been compensated by feedback-controlled divertor gas fueling. The controlled reduction of heat fluxes within this complex 3D edge island geometry is a very promising finding concerning detachment optimization in future scenarios with steady-state operation and metallic wall upgrades [2]. The significant features of the 3D distribution of radiative power losses and the effects on heat and particle fluxes have been recovered with EMC3-EIRENE simulations. Simulations with different island geometries and parameter scans have been investigated to resolve the dependencies of the radiative cooling effects and detachment transitions for Ne and N2 seeding. [1] F. Effenberg et al 2019 Nucl. Fusion 59 106020 [2] Y. Feng et al 2016 Nucl. Fusion 56 126011 Acknowledgments: This work was supported in part by the U.S. Department of Energy (DoE) under grants DE-AC02-09CH11466 and DE-SC0014210. This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014-2018 and 2019-2020 under grant agreement No 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

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

EUROfusion Consortium 633053

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 ...

ASDEX Upgrade (AUG) has been converted to all W plasma facing components (PFCs) in 2007 and JET has implemented the ITER like wall (ILW) project (2011) using the same PFC configuration as ITER during its active phase, namely Be in the main chamber and tungsten in the divertor. As a result of the all metal PFCs in both devices much less surface conditioning is needed to arrive at reproducible wall conditions. Specifically, the Be PFCs of JET led to a very small low- Z content (reduction of C and O by at least a factor of 10), reducing the edge radiation in steady-state operation as well as during disruptions. Both devices successfully employ massive gas injection to mitigate disruption forces and power loads to PFCs by radiating up to 100% of the available energy. Hydrogen retention is strongly reduced (AUG: factor 5, JET: factor 10) and the remaining retention is still dominated by codeposition with residual C in AUG and intrinsic Be in JET. The very low edge and divertor radiation could be compensated by impurity seeding either by a single gas species (N2) (AUG and JET) or by combining N2 and Ar (AUG) injection for divertor and main chamber radiation, respectively. The W sputtering in the divertor increases when seeding small amounts of N2, but decreases for higher fluxes due to the plasma cooling provided by the nitrogen radiation. The tungsten content is controlled by the source as well as by its edge and central transport. It could be kept sufficiently small by using a minimum gas fueling to reduce the W erosion and to diminish the W penetration. The control of the central W transport by central (wave) heating had been well established in AUG, however, in both devices the W content is increased during ICRH operation most probably due to increased W sputtering caused by rectified sheaths. The H-Mode threshold is reduced by 20%-30% in AUG and JET, but on average the confinement is lower in JET-ILW than with C PFCs. To date it is not yet clear, whether the reduced H-Mode confinement has to be attributed to the use of W PFCs, since such a clear trend as in JET was not found in AUG. The increase of confinement with N2 seeding observed in both devices hints to the fact, that low- Z impurities like carbon or nitrogen play a beneficial role for the pedestal confinement.

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 W45+ and W46+ 3p-4d inner shell excitation lines in addition to Mo32+ 2p-3s lines have been identified from the spectrum taken by an upgraded high-resolution x-ray spectrometer. It is found from analysis of the absolute intensities of the W46+ and Mo32+ lines that W and Mo concentrations are in the range of similar to 10(-5) and similar to 10(-6), respectively, with a ratio of similar to 5% in JET with the ITER-like wall configuration for ELMy H-mode plasmas with a plasma current of 2.0-2.5 MA, a toroidal magnetic field of 2.7 T and a neutral beam injection power of 14-18 MW. For the purpose of checking self-consistency, it is confirmed that the W concentration determined from the W45+ line is in agreement with that from the W46+ line within 20% and that the plasma effective charge determined from the continuum of the first order reflection spectrum is also in agreement with that from the second order within 50%. Further, the determined plasma effective charge is in agreement with that determined from a visible spectroscopy, confirming that the sensitivity of the x-ray spectrometer is valid and that the W and the Mo concentrations are also likely to be valid.

JET is used as a test bed for ITER, to investigate beryllium migration which connects the lifetime of first-wall components under erosion with tokamak safety, in relation to long-term fuel retention. The (i) limiter and the (ii) divertor configurations have been studied in JET-ILW (JET with a Be first wall and W divertor), and compared with those for the former JET-C (JET with carbon-based plasma-facing components (PFCs)). (i) For the limiter configuration, the Be gross erosion at the contact point was determined in situ by spectroscopy as between 4% (E-in = 35 eV) and more than 100%, caused by Be self-sputtering (E-in = 200 eV). Chemically assisted physical sputtering via BeD release has been identified to contribute to the effective Be sputtering yield, i.e. at E-in = 75 eV, erosion was enhanced by about 1/3 with respect to the bare physical sputtering case. An effective gross yield of 10% is on average representative for limiter plasma conditions, whereas a factor of 2 difference between the gross erosion and net erosion, determined by post-mortem analysis, was found. The primary impurity source in the limiter configuration in JET-ILW is only 25% higher (in weight) than that for the JET-C case. The main fraction of eroded Be stays within the main chamber. (ii) For the divertor configuration, neutral Be and BeD from physically and chemically assisted physical sputtering by charge exchange neutrals and residual ion flux at the recessed wall enter the plasma, ionize and are transported by scrape-off layer flows towards the inner divertor where significant net deposition takes place. The amount of Be eroded at the first wall (21 g) and the Be amount deposited in the inner divertor (28 g) are in fair agreement, though the balancing is as yet incomplete due to the limited analysis of PFCs. The primary impurity source in the JET-ILW is a factor of 5.3 less in comparison with that for JET-C, resulting in lower divertor material deposition, by more than one order of magnitude. Within the divertor, Be performs far fewer re-erosion and transport steps than C due to an energetic threshold for Be sputtering, and inhibits as a result of this the transport to the divertor floor and the pump duct entrance. The target plates in the JET-ILW inner divertor represent at the strike line a permanent net erosion zone, in contrast to the net deposition zone in JET-C with thick carbon deposits on the CFC (carbon-fibre composite) plates. The Be migration identified is consistent with the observed low long-term fuel retention and dust production with the JET-ILW. EURATOM 633053