Main Chamber Wall Research Articles

Effect of carbon and tungsten plasma-facing materials on the divertor and pedestal plasma in EAST

The edge plasma simulation code SOLPS-ITER is applied to study the effects of plasma-facing materials (PFMs) on the plasma in the EAST device, whose lower carbon (C) divertor will be replaced by tungsten (W) material. The significant difference in particle and energy reflection coefficients between the C and W materials may influence the production and dynamics of the neutral particles in the divertor region, which is characterized by the collisions between the plasma and the recycled neutrals. To separate other effects on the plasma, only the particle and energy recycling of different materials are considered, and no impurity is included. Three PFMs cases, (i) full C PFMs, (ii) W divertor and C main chamber wall (MCW), and (iii) C divertor and W MCW, are simulated. The simulation results show that the PFMs have a remarkable impact on the neutral particles’ recycling, and thus the divertor and pedestal plasma. The C divertor is more capable of enhancing the neutral density in the divertor, and promoting the achievement of the plasma detachment than the W divertor is. However, the PFMs of the MCW have little influence on the divertor plasma. The simulation demonstrates that the C divertor is better than the W divertor from the plasma detachment point of view. In addition, both the divertor and the MCW PFMs have significant roles in the pedestal plasma.

First mirror test in JET for ITER: Complete overview after three ILW campaigns

The First Mirror Test for ITER has been carried out in JET with mirrors exposed during: (i) the third ILW campaign (ILW-3, 2015–2016, 23.33 h plasma) and (ii) all three campaigns, i.e. ILW-1 to ILW-3: 2011–2016, 63,52 h in total. All mirrors from main chamber wall show no significant changes of the total reflectivity from the initial value and the diffuse reflectivity does not exceed 3% in the spectral range above 500 nm. The modified layer on surface has very small amount of impurities such as D, Be, C, N, O and Ni. All mirrors from the divertor (inner, outer, base under the bulk W tile) lost reflectivity by 20–80% due to the beryllium-rich deposition also containing D, C, N, O, Ni and W. In the inner divertor N reaches 5 × 1017 cm−2, W is up to 4.3 × 1017 cm−2, while the content of Ni is the greatest in the outer divertor: 3.8 × 1017 cm−2. Oxygen-18 used as the tracer in experiments at the end of ILW-3 has been detected at the level of 1.1 × 1016 cm−2. The thickness of deposited layer is in the range of 90 nm to 900 nm. The layer growth rate in the base (2.7 pm s − 1) and inner divertor is proportional to the exposure time when a single campaign and all three are compared. In a few cases, on mirrors located at the cassette mouth, flaking of deposits and erosion occurred.

EMC3-EIRENE modeling of edge plasma to improve the ICRF coupling with local gas puffing in DEMO

We report the first 3D EMC3-EIRENE simulations of scrape-off layer (SOL) plasma in DEMO. EMC3-EIRENE is a 3D edge plasma fluid and neutral particle transport code. Effects of local gas puffing on the SOL density and ion cyclotron range of frequencies (ICRF) power coupling have been studied. A pure deuterium plasma is simulated and the puffing/fueling gas is deuterium gas. The gas puffing cases investigated include divertor, top, midplane and antenna gas puffing. The ICRF antenna is distributed through 360° toroidally, and poloidally located in the outer top of the vessel. The results show that toroidally distributed but poloidally localized antenna gas puffing increases the density in front of the antenna most significantly, while top or midplane gas puffing increases this antenna density to a moderate level. Influences of gas puff rate and particle transport parameters on the SOL density are investigated. The parameter scans indicate that the shift of the fast wave cut-off density position to the antenna depends almost linearly on the total gas puff rate. To achieve a significant antenna density increase, the antenna gas puff rate should be at the level of 1.8 × 1023 el s−1 (i.e. 1.0 × 1022 el s−1 per tokamak segment). As the total gas puff rate increases to 4.0 × 1023 el s−1, the evanescent layer of the fast wave (with = 2 m−1) almost vanishes. Moreover, it is found that antenna gas puffing reduces the local power flux to the main chamber wall near the gas valve due to a local decrease of the SOL temperature.

Micro ion beam analysis for the erosion of beryllium marker tiles in a tokamak limiter

Beryllium limiter marker tiles were exposed to plasma in the Joint European Torus to diagnose the erosion of main chamber wall materials. A limiter marker tile consists of a beryllium coating layer (7–9 μm) on the top of bulk beryllium, with a nickel interlayer (2–3 μm) between them. The thickness variation of the beryllium coating layer, after exposure to plasma, could indicate the erosion measured by ion beam analysis with backscattering spectrometry. However, interpretations from broad beam backscattering spectra were limited by the non-uniform surface structures. Therefore, micro-ion beam analysis (μ-IBA) with 3 MeV proton beam for Elastic backscattering spectrometry (EBS) and PIXE was used to scan samples. The spot size was in the range of 3–10 μm. Scanned areas were analysed with scanning electron microscopy (SEM) as well. Combining results from μ-IBA and SEM, we obtained local spectra from carefully chosen areas on which the surface structures were relatively uniform. Local spectra suggested that the scanned area (≈600 μm × 1200 μm) contained regions with serious erosion with only 2–3 μm coating beryllium left, regions with intact marker tile, and droplets with 90% beryllium. The nonuniform erosion, droplets mainly formed by beryllium, and the possible mixture of beryllium and nickel were the major reasons that confused interpretation from broad beam EBS.

Real-time protection of the JET ITER-like wall based on near infrared imaging diagnostic systems

In JET with ITER-like wall (JET-ILW), the first wall was changed to metallic materials (tungsten and beryllium) [] which require a reliable protection system to avoid damage of the plasma-facing components (PFCs) due to beryllium melting or cracking of tungsten owing to thermal fatigue. To address this issue, a protection system with real time control, based on imaging diagnostics, has been implemented on JET-ILW in 2011.This paper describes the design, implementation, and operation of the near infrared imaging diagnostic system of the JET-ILW plasma experiment and its integration into the existing JET-ILW protection architecture. The imaging system comprises eleven analogue CCD cameras which demonstrate a high robustness against changes of system parameters like the emissivity. The system covers about two thirds of the main chamber wall and almost half of the divertor. A real-time imaging processing unit is used to convert the raw data into surface temperatures taking into account the different emissivity for the various materials and correcting for artefacts resulting e.g. from neutron impact. Regions of interest (ROI) on the selected PFCs are analysed in real time and the maximum temperature measured for each ROI is sent to other real time systems to trigger an appropriate response of the plasma control system, depending on the location of a hot spot.A hot spot validation algorithm was successfully integrated into the real-time system and is now used to avoid false alarms caused by neutrons and dust. The design choices made for the video imaging system, the implications for the hardware components and the calibration procedure are discussed. It will be demonstrated that the video imaging protection system can work properly under harsh electromagnetic conditions as well as under neutron and gamma radiation. Examples will be shown of instances of hot spot detection that abort the plasma discharge. The limits of the protection system and the associated constraints on plasma operation are also presented.The real-time protection system has been operating routinely since 2011. During this period, less than 0.5% of the terminated discharges were aborted by a malfunction of the system. About 2%–3% of the discharges were terminated due to the detection of actual hot spots.

Relationship between frequency power spectra and intermittent, large-amplitude bursts in the Alcator C-Mod scrape-off layer

Fluctuations in the boundary region of the Alcator C-Mod tokamak have been analyzed using gas puff imaging data from a set of Ohmically heated plasma density scan experiments. It is found that the relative fluctuation amplitudes are modest and close to normally distributed at the separatrix but become increasingly larger and skewed towards the main chamber wall. The frequency power spectra are nevertheless similar for all radial positions and line-averaged densities. Predictions of a stochastic model, describing the plasma fluctuations as a super-position of uncorrelated pulses, are shown to be in excellent agreement with the measurements. This implies that the pulse duration is the same, while the degree of pulse overlap decreases radially outwards in the scrape-off layer. The universal frequency power spectral density is thus determined by the shape and duration of the large-amplitude bursts associated with blob-like structures. The model also describes the rate of threshold level crossings, for which the exponential tails underline the intermittency of the fluctuations in the far scarpe-off layer.

Overview of fuel inventory in JET with the ITER-like wall

Post mortem analyses of JET ITER-Like-Wall tiles and passive diagnostics have been completed after each of the first two campaigns (ILW-1 and ILW-2). They show that the global fuel inventory is still dominated by co-deposition; hence plasma parameters and sputtering processes affecting material migration influence the distribution of retained fuel. In particular, differences between results from the two campaigns may be attributed to a greater proportion of pulses run with strike points in the divertor corners, and having about 300 discharges in hydrogen at the end of ILW-2. Recessed and remote areas can contribute to fuel retention due to the larger areas involved, e.g. recessed main chamber walls, gaps in castellated Be main chamber tiles and material migration to remote divertor areas. The fuel retention and material migration due to the bulk W Tile 5 during ILW-1 are presented. Overall these tiles account for only a small percentage of the global accountancy for ILW-1.

Type I ELM filament heat fluxes on the KSTAR main chamber wall

Heat loads deposited on the first wall by mitigated Type I ELMs are expected to be the dominant contributor to the total thermal plasma wall load of the International Thermonuclear Experimental Reactor (ITER), particularly in the upper main chamber regions during the baseline H-mode magnetic equilibrium, due to the fast radial convective heat propagation of ELM filaments before complete loss to the divertor. Specific Type I ELMing H-mode discharges have been performed with a lower single null magnetic geometry, where the outboard separatrix position is slowly (∼7s) scanned over a radial distance of 7cm, reducing the wall probe–separatrix distance to a minimum of ∼9cm, and allowing the ELM filament heat loss to the wall to be analyzed as a function of radial propagation distance. A fast reciprocating probe (FRP) head is separately held at fixed position toroidally close and 4.7cm radially in front of the wall probe. This FRP monitors the ELM ion fluxes, allowing an average filament radial propagation speed, found to be independent of ELM energy, of 80–100ms−1 to be extracted. Radial dependence of the peak filament wall parallel heat flux is observed to be exponential, with the decay length of λq, ELM ∼25 ± 4mm and with the heat flux of q∥, ELM= 0.05MWm−2 at the wall, corresponding to q∥ ∼ 7.5MWm−2 at the second separatrix. Along with the measured radial propagation speed and the calculated radial profile of the magnetic connection lengths across the SOL, these data could be utilized to analyze filament energy loss model for the future machines.

Recent progress towards a quantitative description of filamentary SOL transport

A summary of recent results on filamentary transport, mostly obtained with the ASDEX-Upgrade tokamak (AUG), is presented and discussed in an attempt to produce a coherent picture of scrape-off layer (SOL) filamentary transport. A clear correlation is found between L-mode density shoulder formation in the outer midplane and a transition between the sheath-limited and the inertial filamentary regimes. Divertor collisionality is found to be the parameter triggering the transition. A clear reduction of the ion temperature takes place in the far SOL after the transition, both for the background and the filaments. This coincides with a strong variation of the ion temperature distribution, which deviates from Gaussianity and becomes dominated by a strong peak below 5 eV. The filament transition mechanism triggered by a critical value of collisionality seems to be generally applicable to inter-ELM H-mode plasmas, although a secondary threshold related to deuterium fueling is observed. EMC3-EIRENE simulations of neutral dynamics show that an ionization front near the main chamber wall is formed after the shoulder formation. Finally, a clear increase of SOL opacity to neutrals is observed, associated with the shoulder formation. A common SOL transport framework is proposed to account for all these results, and their potential implications for future generation devices are discussed.

Main chamber wall plasma loads in JET-ITER-like wall at high radiated fraction

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

SOL width and intermittent fluctuations in KSTAR

Radial profiles of the ion saturation current and its fluctuation statistics are presented from probe measurements in L-mode, neutral beam heated plasmas at the outboard mid-plane region of KSTAR. The results are consistent with the familiar two-layer structure, seen elsewhere in tokamak L-mode discharges, with a steep near-SOL profile and a broad far-SOL profile. The profile scale length in the far-SOL increases drastically with line-averaged density, thereby enhancing plasma interactions with the main chamber walls. Time series from the far-SOL region are characterised by large-amplitude bursts attributed to the radial motion of blob-like plasma filaments. Analysis of a data time series of several seconds duration under stationary plasma conditions reveals the statistical properties of these fluctuations, including the rate of level crossings and the average duration of periods spent above a given threshold level. This is shown to be in excellent agreement with predictions of a stochastic model, giving novel predictions of plasma–wall interactions due to transient transport events.

Surface modification of He pre-exposed tungsten samples by He plasma impact in the divertor manipulator of ASDEX Upgrade

Tungsten (W) will be used as material for plasma-facing components (PFCs) in the divertor of ITER and interact with Helium (He) ions either from initial He plasma operation or from Deuterium-Tritium (DT) fusion reactions in the active operation phase. Laboratory experiments reported that in a specific operational window of impact energy, ion fluence, and surface temperature (Ein ≥ 20 eV, ϕ ≥ 1 × 1024 He+mBt=2.5T,Ip=0.8MA,Paux≃8.0MWTsurf ≥ 1000 K) a modification of W surfaces occurs resulting in the formation of He-induced W nanostructures. Experiments in ASDEX Upgrade H-mode plasmas (Ein=37keV T, ϕ≃0.75×1024He0m−2 MA, Paux ≃ 8.0 MW) in He have been carried out to investigate in detail (a) the potential growth of W nanostructures on pre-damaged W samples incorporating He nanobubbles, and (b) the potential ELM-induced erosion of W nanostructure. Both W surface modifications were generated artificially in the GLADIS facility by He bombardment of W samples at ϕ≃1×1024He0m−2 keV (a) to ϕ ≃ 0.75 × 1024 He0mϕ≃1.6×1024He+m−2 at Tsurf ≃ 1800 K and (b) ϕ ≃ 1 × 1024 He0m−2 at Tsurf ≃ 2300 K prior to exposure in the divertor manipulator of ASDEX Upgrade. Though in part (a) conditions of W nanostructure growth with a total He ion fluence of ϕ ≃ 1.6 × 1024 He+m−2 and peak He ion impact energies above 150 eV were met, no growth could be detected. In part (b) lower density plasmas with more pronounced type I ELMs, carrying energetic He ions in the keV range, were executed with the strike-line positioned on 2 µm thick W nanostructure accumulating a fluence of ϕ ≃ 0.8 × 1024 He+m−2. Post-mortem analysis revealed that co-deposition by predominantly W, and Boron (B), eroded at the main chamber wall and transported into the divertor, took place on all W samples. Erosion of W nanostructure or its formation was hindered by the fact that the outer divertor at the location of the samples was turned under these He plasma conditions into a net deposition zone by W, B and Carbon (C) ions. The surface morphology with large roughness and effective surface area act as a catcher for the impinging impurities. Thus, apart from operation in the existence diagram of W nanostructure with respect to Tsurf, ϕ, and Ein, also the impinging impurity flux contribution needs to be considered in predictions concerning the formation of W nanostructures.

Investigation of deuterium retention in/desorption from beryllium-containing mixed layers

This article reports experimental results regarding the deuterium (D) retention and release characteristics of beryllium-containing mixed material layers. The main aim is an assessment of the efficiency of the tritium removal procedure currently suggested for ITER, i.e., wall baking at 513 K (240 °C) for the main chamber walls and 623 K (350 °C) for the divertor.It is observed that the tritium retention as well as the removal efficiency by a following baking procedure will strongly depend on the composition of the deposited layer and the wall temperature during plasma operation. In the case that D is implanted at moderate temperatures (300 to 400 K), which would correspond to the “cool divertor” scenario in ITER, a large fraction of retained D is trapped in states having a low D binding energy and correspondingly low release temperature. Therefore, in this case baking can release a significant amount of retained D. On the other hand, if D is implanted at temperatures above 520 K, which would correspond to the “hot divertor” scenario in ITER, the amount of retained D will be significantly lower, but the retained D is predominantly trapped in high-energy binding states. In such a case, even baking at 623 K might not efficiently remove the retained tritium.

Improved EDGE2D-EIRENE simulations of JET ITER-like wall L-mode discharges utilising poloidal VUV/visible spectral emission profiles

A discrepancy in the divertor radiated powers between EDGE2D-EIRENE simulations, both with and without drifts, and JET-ILW experiments employing a set of NBI-heated L-mode discharges with step-wise density variation is investigated. Results from a VUV/visible poloidally scanning spectrometer are used together with bolometric measurements to determine the radiated power and its composition. The analysis shows the importance of D line radiation in contributing to the divertor radiated power, while contributions from D radiative recombination are smaller than expected.Simulations with W divertor plates underestimate the Be content in the divertor, since no allowance is made for Be previously deposited on the plates being re-eroded. An improved version of EDGE2D-EIRENE is used to test the importance of the deposited layer in which the sputtering yield from supposed pure Be divertor plates is reduced to match the spectroscopic signals, while keeping the sputtering yield for the Be main chamber walls unchanged.

Ion cyclotron resonance frequency heating in JET during initial operations with the ITER-like wall

In 2011/12, JET started operation with its new ITER-Like Wall (ILW) made of a tungsten (W) divertor and a beryllium (Be) main chamber wall. The impact of the new wall materials on the JET Ion Cyclotron Resonance Frequency (ICRF) operation is assessed and some important properties of JET plasmas heated with ICRF are highlighted. A ∼ 20% reduction of the antenna coupling resistance is observed with the ILW as compared with the JET carbon (JET-C) wall. Heat-fluxes on the protecting limiters close the antennas, quantified using Infra-Red thermography (maximum 4.5 MW/m2 in current drive phasing), are within the wall power load handling capabilities. A simple RF sheath rectification model using the antenna near-fields calculated with the TOPICA code can reproduce the heat-flux pattern around the antennas. ICRF heating results in larger tungsten and nickel (Ni) contents in the plasma and in a larger core radiation when compared to Neutral Beam Injection (NBI) heating. The location of the tungsten ICRF specific source could not be identified but some experimental observations indicate that main-chamber W components could be an important impurity source: for example, the divertor W influx deduced from spectroscopy is comparable when using RF or NBI at same power and comparable divertor conditions, and Be evaporation in the main chamber results in a strong reduction of the impurity level. In L-mode plasmas, the ICRF specific high-Z impurity content decreased when operating at higher plasma density and when increasing the hydrogen concentration from 5% to 15%. Despite the higher plasma bulk radiation, ICRF exhibited overall good plasma heating performance; the power is typically deposited at the plasma centre while the radiation is mainly from the outer part of the plasma bulk. Application of ICRF heating in H-mode plasmas has started, and the beneficial effect of ICRF central electron heating to prevent W accumulation in the plasma core has been observed.

Implications of the Tore-Supra WEST-Project on Radio-Frequency Additionnal Heating Systems

This year Tore-Supra celebrated its 25 years of operation. During this long time, a number of technologies have been developed. First of all, it was mandatory to develop reliable superconducting magnets at ~ 1.8 K, with superfluid helium as an efficient coolant. For the production of steady state discharge, three types of radio frequency (RF) additional heating systems have been developed: 1) lower hybrid current drive; 2) ion cyclotron resonance heating; and 3) electron cyclotron resonance heating. To cope with long lasting discharges (up to 380 s ×2.8 MW) and large RF additional heating power (12.3 MW ×3 s), actively cooled (AC) plasma facing components were deployed in Tore-Supra for the first time in a tokamak environment. Tore-Supra is now being modified into a D-shape axisymmetric tokamak with AC tungsten main chamber walls and a divertor, the WEST project (W-for tungsten-environment in steady-state tokamak). This new facility has the objective to offer ITER a test bed for validating the relevant AC metallic technologies in D-shape H-mode plasmas. In contrast to other metallic devices, such as JET and ASDEX Upgrade, WEST will rely only on the RF additional power systems. A set of plasma scenarios have been identified, ranging from a high total RF power scenario up to 15 MW in 30 s, to a high fluence scenario of 1000 s with up to 10 MW of injected RF power. These scenarios are able to reproduce ITER relevant conditions of steady state heat loads of 10-20 MW/m2, to test tungsten AC divertor technologies with relevant power heat fluxes and particle fluence.

Characterisation of plasma breakdown at JET with a carbon and ITER-like wall

The recent installation of a full metal, ITER-like, first wall provided the opportunity to study the impact of the plasma-facing materials on plasma initiation or breakdown. This study for the first time presents a full experimental characterisation of tokamak breakdown at JET, using all discharges since 2008, covering both operations with a main chamber carbon and a beryllium ITER-like main chamber wall. It was found that the avalanche phase was unaffected by the change in wall material. However, changes in out-gassing by the wall and lower carbon levels resulted in better controlled density and significantly lower radiation during the burn-through phase with the ITER-like wall. Breakdown failures, that usually developed with a carbon wall during the burn-through phase (especially after disruptions) were absent with the ITER-like wall. These observations match with the results obtained from a new model of plasma burn-through that includes plasma–surface interactions (Kim et al 2012 Nucl. Fusion 52 103016). This shows that chemical sputtering of carbon is the determining factor for the impurity content, and hence also radiation, during the burn-through phase for operations with a carbon wall. As seen experimentally, with a beryllium main wall, the plasma surface effects predicted by the model do not raise the radiation levels much above those expected for pure deuterium plasmas. With the ITER-like wall, operation with higher pre-fill pressures, and thus higher breakdown densities, was possible, which helped maintaining the density after breakdown.

Near wall plasma simulation using penalization technique with the transport code SolEdge2D-Eirene

In order to model plasma interaction with full complexity of the geometry of PFCs, an immersed method to impose boundary conditions has been implemented in the transport code SolEdge2D. This penalization technique has proven to properly recover Bohm boundary conditions, including supersonic solutions at the sheath entrance. The role of the ionization source has been taken into account by coupling SolEdge2D with the 3D kinetic neutral code Eirene. Density regimes are properly recovered and major differences between core located and divertor located ionization source regimes are emphasized. One also presents main chamber wall recycling regime simulations that can now be addressed with simulations of the entire edge plasma up to the wall.

Advanced spatial discretizations in the B2.5 plasma fluid code

Abstract Plasma wall interaction modeling for fusion devices requires consistent plasma backgrounds, which are typically provided by SOL codes like SOLPS 5 [1] . However, the computational domain of these codes typically does not cover the far scrape-off layer, resulting in large uncertainties in the modeling of the interaction between the plasma and the main chamber wall. The work presented here addresses this issue by extending the B2.5 plasma fluid code to provide plasma backgrounds up to the device wall.

Beryllium migration and evolution of first wall surface composition in the JET ILW configuration

Material migration and the resulting evolution of plasma facing surfaces were studied at the beginning of the JET ILW campaign using the singular opportunity of well-defined initial conditions with virgin Be and W wall components. In a sequence of identical Ohmically heated discharges the evolution of wall material sources as well as that of residual impurity sources were studied by spectroscopic detection of suitable emission lines of corresponding neutral atom and singly charged ion species in the visible spectral range. The evolution of divertor surface composition resulting from wall material migration occurred at a similar time scale as previously observed in Be migration experiments in the JET carbon wall configuration. In contrast to these experiments with initial Be evaporation on the carbon main chamber wall, the JET ILW migration experiment is characterised by a continuous Be wall source because the main chamber wall now consists of bulk Be components. The experiment further reveals unexpectedly high Be deposition at W divertor surfaces already during preceding limiter discharges for system commissioning, which has implications for predictive modelling of the expected fuel retention in ITER.