Preliminary assessment of possible carbide formation on Be and T contaminated CFC tiles from JET (original) (raw)
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Nuclear Fusion
Samples extracted from several divertor tiles following the 2011-2012 operation of JET with the ITER-Like wall were analyzed using ion beam analysis methods, x-ray fluorescence spectroscopy, scanning electron microscopy with energy dispersive spectroscopy analysis and x-ray diffraction. The emphasis was on the determination of light species and on material mixing including compound formation on the bottom and the outer divertor tiles. Deposition of deuterium, beryllium, carbon, nitrogen, oxygen, iron, chromium, nickel and molybdenum has been detected on all studied tiles. The thickest deposition, of around 4 µm, was measured on the bottom of the outer divertor, whereas the other surfaces (inner bottom and vertical outer) the co-deposits were around 1 µm. x-ray diffraction measurements have revealed the formation of the compound W 2 C on all specimens.
Investigation of tungsten-coated divertor tiles at the JET tokamak with the ITER‐Like Wall
HNPS Proceedings
Materials migration in fusion plasma devices and fuel retention in plasma-facing components are issues of great importance for the safe operation of fusion devices. The underlying mechanisms require a good understanding in order to make predictions regarding the lifetime of wall components and to assess the amount of fuel retained in the machine mainly in co-deposited layers. To reduce fuel inventory and to investigate plasma-wall interactions a large-scale experiment at the JET (Joint European Torus) tokamak is realized: operation with the ITER-Like Wall (JET-ILW) which comprises beryllium and tungsten. The current work reports on the post-mortem analysis of W/CFC tiles retrieved after the first deuterium-deuterium campaign at JET-ILW. Specimens from different areas of the divertor have been analyzed by means of several techniques including nuclear reaction analysis and Rutherford backscattering employing a deuterium beam. In addition, X-ray fluorescence spectroscopy and scanning e...
Spectrometric analysis of inner divertor materials of JET carbon and ITER-like walls
Fusion Engineering and Design, 2018
One of main reasons of the Joint European torus (JET) transformation from the carbon (JET-C) to ITER-like (JET-ILW) wall was high tritium retention of carbon. In order to compare the tritium retention, samples of analogous positions of the plasma-facing side of vertical tiles No. 3 of two campaigns: JET-C (2008-2009) and JET-ILW (2011-2012) were cut out. Temperature-programmed tritium desorption spectrometry in He + 0.1% H 2 gas flow showed that JET-C sample without a tungsten coating had by a factor of > 20 higher surface concentration of tritium than JET-ILW tungsten-coated sample: 4.9 × 10 13 and 1.7-2.2 × 10 12 T atoms/cm 2 respectively. Installation of metallic plasma facing wall in the JET was a unique possibility to collect from the vacuum vessel the first wall erosion products (EP)dust and flakes. Selected EP were investigated by means of energy dispersion X-ray (EDX), electron spin resonance (ESR), infrared and Raman spectrometry. EDX analysis shows presence of metallic impurities and carbon as a main component. Investigations with ESR spectrometry allows to estimate presence of two main paramagnetic centresg = 2.002 and g = 2.12. Infrared spectra show presence of inorganic oxides. The obtained results supplement the information about composition of the EP from fusion devices.
Tungsten and beryllium armour development for the JET ITER-like wall project
Nuclear Fusion, 2007
The operational behaviour and the interplay of the ITER plasma facing material choice has never been investigated in a tokamak experiment. This motivated the ITER-like Wall Project at JET in which the present main chamber CFC tiles will be exchanged with Be tiles and in parallel a fully tungsten-clad divertor will be prepared. Among the scientific objectives of the ITER-like Wall project are general questions of plasma operation with a low melting Be wall, compatibility of all envisaged ITER scenarios with a W divertor, tritium retention and removal and mixed materials effects, erosion behaviour and lifetime investigations. Three R&D programs were initiated: Be coatings on inconel as well as Be erosion markers are developed for the first wall of the main chamber. This work is done by the Romanian Euratom association under the coordination of the Swedish Royal Institute of Technology. Forschungszentrum Jülich, Germany, developed a conceptual design for a bulk W horizontal target plate, based on an assembly of tungsten lamellae. For all other divertor parts five Euratom Fusion Associations performed R&D to provide the technology to coat the 2-directional CFC material used at JET with thin tungsten coatings. High heat flux screening and cyclic loading tests carried out on the Be coatings on Inconel showed excellent performance, above the required power and energy density. For the bulk W, a design was developed to minimise electromagnetic forces. The design consists of stacks of W lamellae of 6 mm width that are insulated in toroidal direction. High heat flux tests of a test module were performed on the electron beam facility JUDITH at Forschungszentrum Jülich at an absorbed power density up to 9 MW/m² for more than 150 pulses and finally with increasing power loads leading to surface temperatures in excess of 3000°C. No macroscopic failure occurred during the test while SEM showed the development of microcracks on the loaded surface. The W coated CFC tiles were subjected to heat loads with power densities ranging up to 23.5 MW/m². In a second step, a selection of coatings was exposed to cyclic heat loading for 200 pulses at 10.5 MW/m². All coatings developed cracks perpendicular to the CFC fibres due to the stronger contraction of the coating upon cool-down after the heat pulses.
Surface composition and morphology changes of JET tiles under plasma interactions
Fusion Engineering and Design, 2011
Plasma interactions with the main chamber of magnetic fusion devices result in net erosion from some areas and net deposition at other locations. However, high energy particle irradiation means that there are continuous erosion and re-deposition processes involved, creating new surface structures. Although the net deposition can be readily assessed, net erosion and material mixing is difficult to determine. In 2005 marker tiles were mounted in the JET vessel which have a thin tungsten (W) layer deposited on the CFC substrate with a ∼10 m carbon layer on top. This layered structure was designed to determine the areas where some erosion had occurred during JET plasma operations, when the tiles were removed for analysis in 2007. This paper describes the results from a set of tiles mounted in a poloidal limiter (in Octant 8) at the outer wall of the main chamber; a comparison is made between the data from tiles near the top, middle and bottom of the limiter. A set of ion beam techniques together with electron microscopy were used to provide a detailed analysis of the tiles. In general, since plasma interaction is strongest near the centre of the limiter where the tile is closest to the boundary of the confined plasma, erosion dominated on the central tiles, with deposition further from the plasma boundary. Also the amount of retained deuterium is higher in the tiles located in the upper and lower regions of the plasma chamber.
Microanalysis of deposited layers in the inner divertor of JET with ITER-like wall
Nuclear Materials and Energy
In JET with ITER-like wall, beryllium eroded in the main chamber is transported to the divertor and deposited mainly at the horizontal surfaces of tiles 1 and 0 (high field gap closure, HFGC). These surfaces are tungsten coated carbon fibre composite (CFC). Surface sampleswere collected following the plasma operations in 2011-2012 and 2013-2014 respectively. The surfaces, as well as polished cross sections of the deposited layers at the surfaces have been studied with micro ion beam analysis methods (μ-IBA).Deposition of Beand other impurities, and retention of D is microscopically inhomogeneous. Impurities and trapped deuterium accumulate preferentially in cracks, pits and depressed regions, and at the sides of large pits in the substrate (e.g. arc tracks where the W coating has been removed). With careful overlaying of μ-NRA elemental maps with optical microscopy images, it is possible to separate surface roughness effects from depth profiles at microscopically flat surface regions.
Impact of carbon and tungsten as divertor materials on the scrape-off layer conditions in JET
Nuclear Fusion, 2013
The impact of carbon and beryllium/tungsten as plasma-facing components on plasma radiation, divertor power and particle fluxes, and plasma and neutral conditions in the divertors has been assessed in JET both experimentally and by edge fluid code simulations for plasmas in low confinement mode. In high-recycling conditions the studies show a 30% reduction in total radiation in the scrape-off layer when replacing carbon (JET-C) with beryllium in the main chamber and tungsten in the divertor (JET-ILW). Correspondingly, at the low field side divertor plate a twofold increase in power conducted to the plate and a twofold increase in electron temperature at the strike point were measured. In low-recycling conditions the SOL was found to be nearly identical for both materials configurations. Saturation and rollover of the ion currents to both plates was measured to occur at 30% higher upstream densities and radiated power fraction in JET-ILW. Past saturation, it was possible to reduce the ion currents to the low field side targets by a factor of 2 and to continue operating in stable, detached conditions in JET-ILW; in JET-C the reduction was limited to 50%. These observations are in qualitative agreement with predictions from the fluid edge code package EDGE2D/EIRENE, for which a 30% reduction of the total radiated power is also yielded when switching from C to Be/W. For matching upstream parameters the magnitude of predicted radiation is, however, 50 to 100% lower then measured, independent of the materials configuration. Inclusion of deuterium molecules and molecular ions, and temperature and density dependent rates in EIRENE reproduced the experimentally observed rollover of the ion current to the low field side plate, via reducing the electron temperature at the plate.
Influence of the Carbidized Tungsten Surface on the Processes of Interaction with Helium Plasma
Materials
This paper presents the results of experimental studies of the interaction of helium plasma with a near-surface tungsten carbide layer. The experiments were implemented at the plasma-beam installation. The helium plasma loading conditions were close to those expected in the ITER divertor. The technology of the plasma irradiation was applied in a stationary type linear accelerator. The impact of the helium plasma was realized in the course of the experiment with the temperatures of ~905 °C and ~1750 °C, which were calculated by simulating heat loading on a tungsten monoblock of the ITER divertor under the plasma irradiation at the load of 10 MW/m2 and 20 MW/m2, respectively. The structure was investigated with scanning microscopy, transmitting electron microscopy and X-ray analysis. The data were obtained showing that the surface morphology changed due to the erosion. It was found that the carbidization extremely impacted the plasma–tungsten interaction, as the plasma–tungsten intera...
Journal of Nuclear Materials
Inconel-600 blocks and stainless steel covers for quartz microbalance crystals from remote corners in the JET-ILW divertor were studied with time-of-flight elastic recoil detection analysis and nuclear reaction analysis to obtain information about the areal densities and depth profiles of elements present in deposited material layers. Surface morphology and the composition of dust particles were examined with scanning electron microscopy and energy-dispersive X-ray spectroscopy. The analysed components were present in JET during three ITER-like wall campaigns between 2010 and 2017. Deposited layers had a stratified structure, primarily made up of beryllium, carbon and oxygen with varying atomic fractions of deuterium, up to more than 20%. The range of carbon transport from the ribs of the divertor carrier was limited to a few centimeters, and carbon/deuterium co-deposition was indicated on the Inconel blocks. High atomic fractions of deuterium were also found in almost carbon-free layers on the quartz microbalance covers. Layer thicknesses up to more than 1 mm were indicated, but typical values were on the order of a few hundred nm. Chromium, iron and nickel fractions were less than or around 1% at layer surfaces while increasing close to the layer-substrate interface. The tungsten fraction depended on the proximity of the plasma strike point to the divertor corners. Particles of tungsten, molybdenum and copper with sizes less than or around 1 mm were found. Nitrogen, argon and neon were present after plasma edge cooling and disruption mitigation. Oxygen-18 was found on component surfaces after injection, indicating in-vessel oxidation. Compensation of elastic recoil detection data for detection efficiency and ion-induced release of deuterium during the measurement gave quantitative agreement with nuclear reaction analysis, which strengthens the validity of the results.
Atomic Energy, 1997
The use of carbon materials as the facing for the liner, limiters, and divertor plates in contact with the plasma in large tokamaks has made it possible to approach very closely the conditions for thermonuclear fusion in a D-T mixture. New experiments have raised many questions as regards optimizing the choice of the carbon materials, developing methods for testing them, and determining the possible maximum content of hydrogen isotopes in graphite under conditions of neutron irradiation. The task of the present work is, first, to propose the methodology of experiments enabling the maximum possible concentration of hydrogen isotopes in the graphite elements of a liner structure to be predicted by means of relatively simple tests with molecular hydrogen, without performing plasma experiments on tokamaks, and second to measure the sorption capacity of the graphite recommended for use in tokamak reactors for the determined hydrogen pressure range and sample temperature. NEUTRAL GAS PRESSURE IN A TOKAMAK WALL PLASMA The lifetime of hydrogen particles in the central plasma of contemporary tokamaks (TFTR, JET, DIII-D, Tore-Supra, JT-60U) is several tenths of a second [1], while the duration of the operating pulse exceeds 10 see. During a pulse (with a current of 1-3 MA) a hydrogen particle can leave and enter the plasma tens of times before it enters the divertor volume and becomes a component of the divertor neutral gas with an increased density. On the average, there are (32-64) x 1019 hydrogen molecules in the plasma column during a pulse (DIII-D [2, 3]) and (32-320) x 1020 molecules in the liner walls and divertor plates. The rate at which they enter the plasma is (32-96) x 1019 see-1 and the pumping rate of the liner during a pulse is (3.2-32) x 1019 see-1. A large fraction of the injected particles is absorbed by the liner wall and the divertor. A steady-state plasma with a constant particle density of (6-9) x 1019 m-3 and temperature of 2-3 keV is maintained due to the injection of neutral particles and the arrival of hydrogen from the liner walls and the limiter/divertor plates (recycling). Using a phenomenological approach to describing the particle flux, one can consider the central and peripheral (divertor) plasma as a source of ions and neutral particles and the walls and divertor plates as a source of hydrogen atoms and molecules, with the plasma as a rather perfect sink. During the steady-state part of the operating pulse these exchange fluxes between the plasma and the liner are-(32-320) x 1020 sec-1 and far exceed the particle injection and hydrogen pumping rates [4]. Normally the total flux of particles from the plasma to the divertor plates is roughly a factor of 5-10 higher than the particle flux to the liner walls, being-3.2 x 102 2 sec-1 [5]. To a first approximation one can assume that the exchange fluxes are stabilized and create a def'mite pressure near the surfaces of the liner walls and divertor. The neutral gas pressure in the divertor space has been measured in several tokamaks (DIII-D [6, 7], ASDEX-UP [8], JT-60U [9]) and found to be 0.3-3 Pa.