Main Chamber Plasma Facing Components Research Articles

Fuel inventory and material migration of JET main chamber plasma facing components compared over three operational periods

Fuel retention and material migration results from JET ITER-like wall beryllium limiter tiles are presented for three operating periods. Ion beam analysis results support the general picture of erosion during limiter configurations with local deposition on tile ends far into the scrape off layer. Similar trends of fuel concentrations are observed in all JET operating periods; (i) low on surfaces exposed to high heat flux and erosion and (ii) higher in deposits. The pattern of fuel retention and deposition correlates with heat flux and distribution of limiter plasmas touching inner and outer limiters. The D/Be ratio in the thickest deposit is ∼0.01. Global fuel retention attributed to limiters is <0.01% of injected fuel. Marker coatings reveal a rapid transition from strong erosion to deposition within 15 mm. This work highlights the need to carefully consider the requirements of marker coatings for long exposure studies in high erosion areas.

Scrape-off layer ion temperature measurements at the divertor target during type III ELMs in MAST measured by RFEA

Edge-localised modes (ELMs) can carry significant fractions of their energy as far as main chamber plasma-facing components in divertor tokamaks. Since in future devices (e.g. ITER, DEMO) these energies could cause issues for material lifetime and impurity production, the energy and temperature of ions in ELMs needs to be investigated. In MAST, novel divertor measurements of Ti during ELMs have been made using the divertor retarding field energy analyser (RFEA) probe. These measurements have shown instantaneous ion energy distributions corresponding to an effective Ti at 5cm from the strike point at the target that can be as high as 60eV and that this decreases with time after the ELM start. This is consistent with the hottest, fastest ions arriving at the target first by parallel transport, followed by the lower end of the ion energy distribution. This analysis will form a basis for future data analysis of fast swept measurements of ion distributions in ELMs.

EMC3-Eirene simulations of particle- and energy fluxes to main chamber- and divertor plasma facing components in ASDEX Upgrade compared to experiments

We report on first EMC3-Eirene simulations with an extended computational grid including both divertor- and main chamber (MC) 3D wall plasma-facing components (PFCs). In a first step we compare the simulations to low-power L-mode discharges systematically analyzed by Carralero et al., who observed a transition from a low- to a high density regime (Carralero et al., 2014). Case A is a configuration at low density and high clearance, while B is a high density regime configuration at medium clearance. In order to explain the upstream far-SOL ne profiles of B the MC PFCs and an enhanced transport region at ρ=1.01,…,1.03 need to be included in the simulations. In a second step we compute the particle-, and power fluxes to the limiter for the realistic geometry, for limiters displaced radially inward and for a toroidally symmetric limiter. Almost the same fraction of power as that to the divertor is absorbed by the MC PFCs in the simulation of case B.

A protection system for the JET ITER-like wall based on imaging diagnostics

The new JET ITER-like wall (made of beryllium and tungsten) is more fragile than the former carbon fiber composite wall and requires active protection to prevent excessive heat loads on the plasma facing components (PFC). Analog CCD cameras operating in the near infrared wavelength are used to measure surface temperature of the PFCs. Region of interest (ROI) analysis is performed in real time and the maximum temperature measured in each ROI is sent to the vessel thermal map. The protection of the ITER-like wall system started in October 2011 and has already successfully led to a safe landing of the plasma when hot spots were observed on the Be main chamber PFCs. Divertor protection is more of a challenge due to dust deposits that often generate false hot spots. In this contribution we describe the camera, data capture and real time processing systems. We discuss the calibration strategy for the temperature measurements with cross validation with thermal IR cameras and bi-color pyrometers. Most importantly, we demonstrate that a protection system based on CCD cameras can work and show examples of hot spot detections that stop the plasma pulse. The limits of such a design and the associated constraints on the operations are also presented.

Preparing the scientific basis for an all metal ITER

The use of beryllium, carbon and tungsten as plasma-facing materials (PFMs) in ITER calls for dedicated investigations on their behaviour under the expected particle and power loads and neutron irradiation. Their simultaneous use implies the formation of mixed materials during plasma operation, which can have significantly different properties compared with the initial materials concerning thermo-mechanical behaviour and T retention. This contribution presents the latest results on these issues mainly achieved under the umbrella of the European taskforce on plasma wall interaction. While laboratory results provide the basis for the understanding of the basic properties and the behaviour of PFMs, experiments in fusion devices are indispensable in order to test their integral performance, incorporating also the internal feedback on the plasma properties. For this reason ASDEX Upgrade has been converted to a full W device, while JET is beginning operation with its ITER-like wall consisting of Be main chamber plasma-facing components and a W divertor. In parallel, dedicated W experiments are performed in several devices in order to investigate specific uncertainties such as the characteristics of melt-layer movement in the presence of a strong magnetic field. Based on the results on ASDEX Upgrade, which reveal a narrower operational space as compared with the operation with graphite plasma-facing components but also provide tools for a successful operation with tungsten, the experiments at JET will provide a unique opportunity to address specific issues related to the parallel use of Be and W as PFMs with plasma parameters closest to those of ITER. Amongst these are the anticipated but still to be demonstrated reduction in hydrogen retention compared with a carbon device, the test of conditioning procedures, mixed materials effects, erosion and transport, the effect of edge localized modes (ELMs) and ELM mitigation methods as well as the behaviour of melt layers and their influence on plasma operation under steady state and transient heat loads.

Modelling the erosion of beryllium carbide surfaces

Redeposition of beryllium eroded from main chamber plasma facing components of ITER onto the divertor material carbon creates a mixed material, beryllium carbide Be 2C, whose interaction with the plasma is not well known. In this study, we have investigated the erosion of Be 2C by deuterium using molecular dynamics simulations and ERO impurity modelling. We found that beryllium sputters preferentially over carbon and identified the sputtering mechanism in the ion energy range 10–100 eV to be both physical and swift chemical sputtering. In addition to single atoms, different types of small molecules/clusters were sputtered, the most frequently occurring molecules being BeD, Be 2D, and CD. The sputtering threshold was found to lie between 10 and 15 eV. The MD sputtering yields were used in plasma impurity simulations, serving as a replacement for input data obtained with TRIM. This changes the accumulation rate of impurity Be in the divertor region compared to previous estimates.

Heat loads on plasma facing components during disruptions on JET

For the first time, fast measurements of heat loads on the main chamber plasma facing components (about 1 ms time resolution) during disruptions are taken on JET. The timescale of energy deposition during the thermal quench is estimated and compared with the timescale of the core plasma collapse measured with soft x-ray diagnostic. The energy deposition time is 3–8 times longer than the plasma energy collapse during density limit disruptions or radiative limit disruptions. This factor is rather in the range 1.5–4 for vertical displacement events. The heat load profiles measured during the thermal quench show substantial broadening of the power footprint on the upper dump plate. The scrape-off layer power width is increased by a factor of 3 for the density limit disruptions. The far scrape-off layer is characterized by a steeper gradient which could be explained by shadowing of the dump plate by other main chamber plasma facing components such as the outer limiter.

Plasma–surface interaction, scrape-off layer and divertor physics: implications for ITER

Recent research in scrape-off layer (SOL) and divertor physics is reviewed; new and existing data from a variety of experiments have been used to make cross-experiment comparisons with implications for further research and ITER. Studies of the region near the separatrix have addressed the relationship of profiles to turbulence as well as the scaling of the parallel power flow. Enhanced low-field side radial transport is implicated as driving parallel flows to the inboard side. The medium-n nature of edge localized modes (ELMs) has been elucidated and new measurements have determined that they carry ∼10–20% of the ELM energy to the far SOL with implications for ITER limiters and the upper divertor. The predicted divertor power loads for ITER disruptions are reduced while those to main chamber plasma facing components (PFCs) increase. Disruption mitigation through massive gas puffing is successful at reducing PFC heat loads. New estimates of ITER tritium retention have shown tile sides to play a significant role; tritium cleanup may be necessary every few days to weeks. ITER's use of mixed materials gives rise to a reduction of surface melting temperatures and chemical sputtering. Advances in modelling of the ITER divertor and flows have enhanced the capability to match experimental data and predict ITER performance.

Tungsten: an option for divertor and main chamber plasma facing components in future fusion devices

The tungsten programme in ASDEX Upgrade is pursued towards a full high-Z device. The spectroscopic diagnostic of W has been extended and refined and the cooling factor of W has been re-evaluated. The W coated surfaces now represent a fraction of 65% of all plasma facing components (24.8 m2). The only two major components that are not yet coated are the strikepoint region of the lower divertor as well as the limiters at the low field side. While extending the W surfaces, the W concentration and the discharge behaviour have changed gradually pointing to critical issues when operating with a W wall: anomalous transport in the plasma centre should not be too low, otherwise neoclassical accumulation can occur. One very successful remedy is the addition of central RF heating at the 20–30% level. Regimes with low ELM activity show increased impurity concentration over the whole plasma radius. These discharges can be cured by increasing the ELM frequency through pellet ELM pacemaking or by higher heating power. Moderate gas puffing also mitigates the impurity influx and penetration, however, at the expense of lower confinement. The erosion yield at the low field side guard limiter can be as high as 10−3 and fast particle losses from NBI were identified to contribute a significant part to the W sputtering. Discharges run in the upper W coated divertor do not show higher W concentrations than comparable discharges in the lower C based divertor. According to impurity transport calculations no strong high-Z accumulation is expected for the ITER standard scenario as long as the anomalous transport is at least as high as the neoclassical one.

Erosion and deposition in the ASDEX Upgrade tungsten divertor experiment

Tungsten coated graphite tiles were mounted in the divertor of the ASDEX Upgrade tokamak and exposed to approximately 800 plasma discharges. After the experimental campaign, a poloidally complete set of samples was removed. The poloidal tungsten distribution pattern is presented and compared to those of other metallic impurities. The amount of tungsten detected on main chamber plasma facing components is found to be nearly unaffected by the operation of a fully tungsten covered divertor. Modifications of the tungsten surfaces are analyzed. Deposition is observed to dominate in the inner divertor while erosion prevails in the outer divertor. The influence of this result on the target plate deuterium inventories is demonstrated. Finally, a model description for the deuterium inventories observed in the main chamber is presented.