W-coated Tiles Research Articles

Modeling of ExB effects on tungsten re-deposition and transport in the DIII-D divertor

Mixed-material DIVIMP–WallDYN modeling, now incorporating ExB drifts, is presented that simultaneously reproduces tungsten (W) erosion and deposition patterns observed during the DIII-D metal rings campaign, in which a toroidally symmetric set of W-coated tiles were installed in the carbon (C) DIII-D divertor. Since most reactor plasma facing component (PFC) designs call for mixed-material environments, including ITER’s W/Be environment, the divertor targets will quickly evolve into reconstituted surfaces of multiple elements. This work identifies controlling physics that affects material migration patterns in the divertor, which impact PFC lifetimes and impurity leakage from the divertor to the core. These simulations indicate that radial and poloidal ExB transport dominates over parallel force balance for high-Z impurities such as W in the divertor region of DIII-D. It is demonstrated that ExB drifts are required to reproduce the experimental observation of non-local W and C co-accumulation in a band ∼7–9 cm outboard of the outer-strike-point (OSP) W source, for attached L-mode conditions in the unfavorable ion grad-B drift direction. In addition, W gross erosion is localized to the region outboard of the OSP, as the formation of C co-deposits suppresses W erosion at the strike point. Time-dependent simulations with scaled ExB impurity drifts (60% of the OEDGE-calculated drift velocity) and W re-erosion quantitatively reproduce these features, including depth-resolved W/C ratios, within a factor of 2 over ∼115 s of accumulated plasma exposure. The location of co-deposition regions is shown to be well-represented by an analytic leakage model, driven largely by poloidal ExB drifts. Qualitative agreement is also found between campaign-integrated W deposition measurements and simulations for the favorable ion grad-B drift direction, the standard mode of operation for most tokamaks. These results imply that a long-term inward radial migration of material from the outer divertor through the private flux region may occur in future devices.

Global distribution of tritium in JET with the ITER-like wall

• Tritium (T) distributions in Be, W and W-coated CFC tiles from JET were examined. • The center of Be outer limiter showed the highest T retention in the main chamber. • The co-deposits in the divertor with high C fraction showed high T content. • The C fraction and T content after 3rd campaign were low compared to 1st campaign. • T retention in the bulk W tile was far smaller than that in the W-coated CFC tiles. Nondestructive analysis of tritium (T) distribution was performed by means of imaging plate technique on specimens cut from the Be limiters, W-coated carbon tiles and bulk W lamellae retrieved from the JET tokamak after the first and third experimental campaigns with the ITER-like wall. Afterwards, analyses were continued using X-ray photoelectron spectroscopy, microscopy techniques and thermal desorption spectroscopy. Co-deposits formed on the W-coated tiles in the 1st campaign showed large T retention because of high carbon content reaching up to 50 atomic %, while the carbon fraction in co-deposits after the 3rd campaign was distinctly lower. The T retention of the plasma-facing surface of the bulk W tile was smaller than that of the W-coated tiles by a factor of 20, while deposition of small amount of T was found at the side surfaces facing to the gaps in a lamella structure. The correlation of T distributions with surface morphology and the discharge conditions is discussed.

Validation of EDGE2D-EIRENE and DIVIMP for W SOL transport in JET

Abstract Tungsten sputtering rates and density profiles predicted using the edge plasma codes EDGE2D-EIRENE and DIVIMP are found to agree within a factor of 4 with measurements of neutral and singly-ionized W spectral line emission in the JET low-field side (LFS) divertor, and within a factor of 2 with SXR, VUV, and bolometric calculations of the W density in the main plasma. The edge plasma W predictions are extended to the core plasma using JINTRAC integrated core-edge modelling. Prompt redeposition of W is identified as the primary reason for the discrepancy between predicted and measured W emission in the divertor. The studied plasmas include attached divertor conditions in L-mode and type-I ELMy H-mode plasmas typical for JET. To more accurately reproduce the spectroscopically inferred W sputtering rates in EDGE2D-EIRENE, imposing the experimentally observed Be concentration of order 0.5% in the divertor is necessary. However, the W density in the main plasma is predicted to be insensitive to whether or not W is sputtered by Be at the divertor targets. Instead, the majority of the predicted core W originated in L-mode from sputtering due to fast D charge-exchange atoms at the W-coated tiles above the LFS divertor, and in H-mode due to D and W ions at the targets during ELMs.

Dust generation in tokamaks: Overview of beryllium and tungsten dust characterisation in JET with the ITER-like wall

Operation of the JET tokamak with beryllium and tungsten ITER-like wall provides unique opportunity for detailed studies on dust generation: quantity, morphology, location, etc. The programme carried out in response to ITER needs for safety assessment comprises: (i) remotely controlled vacuum cleaning of the divertor; (ii) local sampling of loosely bound matter from plasma-facing components (PFC); (iii) collection of mobilized dust on various erosion-deposition probes located in the divertor and in the main chamber. Results of comprehensive analyses performed by a number of complementary techniques, e.g. a range of microscopy methods, electron and ion spectroscopy, liquid scintillography and thermal desorption, are summarized by following points:(a) Total amount of dust collected by vacuum cleaning after three campaigns is about 1–1.4 g per campaign (19.1–23.5 h plasma operation), i.e. over 100 times smaller than in JET operated with carbon walls (i.e. in JET-C).(b) Two major categories of Be dust are identified: flakes of co-deposits formed on PFC and droplets (2–10 μm in diameter). Small quantifies, below 1 g, of Be droplets and splashes are associated mainly with melting of beryllium limiters.(c) Tungsten dust occurs mainly as partly molten flakes originating from the W-coated tiles.

Tritium retention measurements by accelerator mass spectrometry and full combustion of W-coated and uncoated CFC tiles from the JET divertor

Accelerator mass spectrometry (AMS) and the full combustion method (FCM) followed by liquid scintillation counting were applied to quantitatively determine the tritium retention in the tungsten-coated carbon fibre composites (CFC), in comparison to uncoated CFC tiles from the JET divertor. The tiles were adjacent and exposed to plasma operations between 2007 and 2009. The tritium depth profiles are showing that the tritium retention on the W-coated tile was reduced by a factor of 13.5 in comparison to the uncoated tile whereas the bulk tritium concentration is approximately the same for both tiles.

Global erosion and deposition patterns in JET with the ITER-like wall

A set of Be and W tiles removed after the first ITER-like wall campaigns (JET-ILW) from 2011 to 2012 has been analysed. The results indicate that the primary erosion site is in the main chamber (Be) as in previous carbon campaigns (JET-C). In particular the limiters tiles near the mid-plane are eroded probably during the limiter phases of discharges. W is found at low concentrations on all plasma-facing surfaces of the vessel indicating deposition via plasma transport initially from the W divertor and from main chamber W-coated tiles; there are also traces of Mo (used as an interlayer for these coatings). Deposited films in the inner divertor have a layered structure, and every layer is dominated by Be with some W and O content.

Experimental determination of the transient heat absorption of W divertor materials

Fast infrared (IR) thermography resolves the transient edge localized mode (ELM) induced heat fluxes on divertor components on time scales of a few hundred microseconds. These heat loads range from 10 to several 100 MW m−2 and energy densities of 15–200 kJ m−2. The calculation of the local ELM heat flux depends on the so-called surface heat transfer coefficient very sensitively. Therefore we performed dedicated experiments in the high heat flux test facility GLADIS with well-defined temporal and spatial shape of heat fluxes to reduce the uncertainties of the ELM heat flux calculations in JET. We have experimentally determined the surface heat transfer coefficient for the W components used as divertor components of the JET ILW project. Based on the results of the measured transient heat absorption, the coefficient was deduced in a temperature range from 400 to 1200 °C for the bulk W lamella and for 10 and 20 μm W-coated carbon fibre reinforced carbon tiles, respectively. The measurements allow an improved estimation of ELM heat loads in JET on W and W-coated tiles and an error estimate of the absorbed heat flux.

Migration and deposition of 13C in the full-tungsten ASDEX Upgrade tokamak

The migration of carbon in low-density, low-confinement plasmas of ASDEX Upgrade was studied by injecting 13C into the main chamber of the torus at the end of the 2007 experimental campaign. A selection of standard tungsten-coated lower-divertor and main-chamber tiles as well as a complete set of lower-divertor tiles with an uncoated poloidal marker stripe were removed from one poloidal cross section and analysed using secondary ion mass spectrometry. The poloidal deposition profiles of 13C on both the tungsten-coated tiles and on the uncoated graphite areas of the marker tiles were measured and compared. For the W-coated lower-divertor tiles, 13C was deposited mainly on the high-field side tiles, while barely detectable amounts of 13C were observed on low-field side samples. In contrast, on the uncoated marker stripes the deposition was equally pronounced in the high-field and low-field side divertor. The marker-tile results are in agreement with those obtained from graphite tiles after the 2003 and 2005 13C experiments in ASDEX Upgrade. In the case of W-coated tiles, the 13C measurements were complemented by determining the total amount of deposited carbon (12C) on the tiles, which also shows strong deposition at the inner parts of the lower divertor. The estimated deposition of 13C on W at the divertor areas was less than 1.5% of the injected amount of 13C atoms. The 13C analyses of the main-chamber tiles and small silicon samples mounted in remote areas revealed significant deposition in the upper divertor, in upper parts of the heat shield, in the limiter region close to the injection valve, and below the roof baffle. Approximately 8% of the injected 13C is estimated to have accumulated in these regions. Possible reasons for the different deposition patterns on W and on graphite in different regions of the torus are discussed.

First demonstration of non-destructive tests on tungsten-coated JET divertor CFC tiles in the electron beam facility JUDITH-2

In the ITER-like wall project, the JET tokamak will employ tungsten-coated carbon fibre reinforced composite (CFCs) tiles for the outer and inner divertor tile rows. In order to assure the performance under thermal loads at the JET divertor, high heat flux (HHF) tests are performed for the qualification during the series production of the W-coated tiles, some of which might be carried out in the JUDITH-2 facility of FZJ. The electron beam (EB) facility JUDITH-2 consists of an EB-gun, a process chamber with a vacuum system (vacuum chamber extension) and different cooling circuits. This paper presents a summary of preparation steps and the first demonstration of cyclic EB loading experiments with the JUDITH-2 facility. One W-coated CFC tile was loaded at an absorbed average power density of 2.86 MW m−2. The cyclic EB loading time was 10.2 s.

Performance of tungsten coatings as plasma facing components used in ASDEX Upgrade

Tungsten coated graphite tiles were mounted in the divertor of the ASDEX Upgrade tokamak and exposed to approximately 800 plasma discharges with varying average heat fluxes of up to 6 MW/m 2. After the experimental campaign, cracks were detected on almost all tiles in accordance with results of electron beam tests. The W-coated tiles in the inner divertor were found to be covered by deposited layers of low-Z material with thicknesses in the μm range and only a few percent tungsten remaining on the surface. In contrast the outer divertor is dominated by erosion. Here, the tungsten concentration at the surface was found to be close to the initial level. The global W-migration pattern was determined by analyzing the deposition of tungsten on a complete poloidal set of plasma facing wall components. Both divertor regions display broad deposition peaks with deposited layers at least one order of magnitude thicker than the main chamber values.