CFC Materials Research Articles

Properties test of 1D carbon-carbon composite tiles for the diagnostic calorimeter of CRAFT NNBI test platform

An important aspect of the Comprehensive Research Facility for Fusion Technology (CRAFT) project is to conduct research on neutral beam injection based on accelerated negative ions, which is one of the main auxiliary heating methods. To evaluate beam characteristics, a diagnostic calorimeter based on a unidirectional carbon fiber-carbon matrix (CFC) composite has been developed in the negative ion source test facility. The CFC calorimeter can diagnose key features of a particle beam, including its divergence and uniformity. The CFC calorimeter has been developed by a worldwide research team of neutral beam injection (NBI) due to its high spatial resolution, which is determined by the thermal properties of CFC materials. This paper aims to verify the thermal properties of CFC materials from TianCe Technology Co., Ltd. by comparing data acquired by infrared (IR) camera with simulations. Experimental and simulation results show that the CFC material provided by the company satisfies the requirements of experiment.

Microstructural stability of tungsten coated CFC plates aiming to metal Wall experiments at JT-60SA

Abstract Tungsten “Metal wall” experiments are planned at JT-60SA to contribute to ITER. Because tungsten coated CFC materials are attractive for the “metal wall” experiments, their fabrication techniques need to be developed. Aiming to the application of divertor, a sinter bonding method is being developed to join thick tungsten plates and CFC blocks. Present research reports about the fabrication of W/CFC joints, basic microstructural and mechanical characterizations, and results of preliminary high temperature exposure tests at 1073 K and 1173K for 1000 h in medium vacuum.

Retention and release of hydrogen isotopes in carbon materials priorly charged in gas phase

Thermal desorption of deuterium from isotropic graphites and a CFC material priorly charged with deuterium gas has been investigated to obtain quantitative information on hydrogen recycling and tritium inventory in fusion experimental devices and future reactors. The spectra of thermal desorption spectrometry (TDS) appear to have five peaks, which are defined as peaks 1–5 in order of the increase of temperature. Diffusion coefficients of deuterium for the absorption process are evaluated for the samples charged with deuterium gas for various absorption times. Absorption pressure dependence of TDS spectra shows a unique change in the temperatures of peak 2. As for the absorption temperature dependence of TDS spectra, peak 4 has a drastic change in the shape of spectra and also in the peak temperatures. The mechanism of desorption for peak 5 will be associated to a detrapping process with a dispersion of energies in the range of 4.0–4.9 eV.

Investigation on the erosion/deposition processes in the ITER-like wall divertor at JET using glow discharge optical emission spectrometry technique

As a complementary method to Rutherford back scattering (RBS), glow discharge optical emission spectrometry (GDOES) was used to investigate the depth profiles of W, Mo, Be, O and C concentrations into marker coatings (CFC/Mo/W/Mo/W) and the substrate of divertor tiles up to a depth of about 100 μm. A number of 10 samples cored from particular areas of the divertor tiles were analyzed. The results presented in this paper are valid only for those areas and they cannot be extrapolated to the entire tile. Significant deposition of Be was measured on Tile 3 (near to the top), Tile 6 (at about 40 mm from the innermost edge) and especially on Tile 0 (HFGC). Preliminary experiments seem to indicate a penetration of Be through the pores and imperfections of CFC material up to a depth of 100 μm in some cases. No erosion and a thin layer of Be (<1 μm) was detected on Tiles 4, 7 and 8. On Tile 1 no erosion was found at about 1/3 from bottom.

Trapping State of Hydrogen Isotopes in Carbon and Graphite Investigated by Thermal Desorption Spectrometry

Thermal desorption spectrometry (TDS) has been investigated to obtain fundamental information of tritium behavior in graphite and carbon materials especially at high temperatures. 29 brands of graphite, HOPG, glassy carbon and CFC materials charged with deuterium gas are tested up to the temperature of 1735 K with a heating rate of 0.1 K/s. TDS spectra have five peaks at 600–700 K, around 900 K, 1200 K, 1300–1450 K and 1600–1650 K. The amounts of released deuterium have been compared with crystallographic parameters derived from XRD analysis. Reduction of tritium retention and inventories are discussed.

Thermal desorption of hydrogen from carbon and graphite at elevated temperatures

Abstract Thermal desorption spectrometry (TDS) has been investigated to obtain fundamental information of hydrogen behavior in graphite and CFC especially at high temperatures. Thirteen brands of graphite and CFC materials charged with deuterium gas are tested up to the temperature of 1720 K with a heating rate of 0.1 K/s. TDS spectra have at least four peaks at 600–700 K, around 900 K, 1300–1450 K and 1600–1650 K. The change of TDS spectra is measured for the samples, which are charged with deuterium at 1273 K under a different pressure in the range of 83 Pa to 79 kPa. Physical and chemical states of deuterium in graphite and mechanisms of desorption are discussed.

X-ray micro-tomography studies on carbon based composite materials for porosity network characterization

Quantitative characterization of the porosity structure of the carbon reinforced carbon fibre (CFC) materials is carried out by high resolution cone-beam X-ray micro-tomography (CBμCT). It is shown that CBμCT provides useful information pertaining to the in-depth fuel migration into carbon tiles function of the CFC material structure. The investigated materials comprised two types of non-irradiated CFC samples (former ITER reference CFC NB31 and JET CFC DMS780) and a series of CFC N11 samples in the frame of the Deuterium Inventory in Tore Supra (DITS) post-mortem analysis. A procedure for the quantitative evaluation of the CFC porosity factor has been introduced and tested. Useful information about the pores connectivity as well as the metal impregnation inside the CFC macroscopic pores, in case of heat sink region of the TS CFC, can be retrieved. The method can be used for the quality control monitoring of the new CFC ITER reference materials.

Preparation of Functionally Graded W/Cu Interlayers and Brazing to CuCrZr and CFC for Actively Cooled Flat Tile Divertor Mock-Ups

The designed plasma facing materials for the divertor-components in ITER are up to now carbon (graphite or CFC) and tungsten. The heat flux loading can result in temperatures of up to 550°C at the interface between the plasma facing material and the CuCrZr heat sink, an operation temperature which is too high for CuCrZr. Additionally the temperature gradient and the mismatch in thermal expansion (CTE) of both parts of the divertor result in high stresses at the interface. A tungsten-copper composite material with a gradation from pure copper on the one side to pure tungsten on the opposite side is supposed to support the stress reduction. The tungsten contributes to the strength of the composite, whereas the copper provides the required thermal conductivity (TC) of at least 200 W/mK. The graded W/Cu layers were produced by the “chemical mixing” method and subsequent liquid phase sintering. Here the use of sub-µm W-particle sizes offers an opportunity to gain well dispersed W-particles with high Cu-containing composites resulting in lower CTE values and higher thermal conductivities compared to coarser W-powders. Brazing of the functionally graded W/Cu interlayers to CuCrZr and CFC materials resulted in adjoined mock-ups.

Deuterium retention in different carbon materials exposed in TEXTOR

CFC NB31, CFC DMS780 and fine-grain graphite EK98 were simultaneously exposed in the SOL of TEXTOR in order to measure the retention of deuterium in the material bulk. After exposure, the samples were analysed by thermal desorption spectrometry and nuclear reaction analysis with a conventional beam as well as with a microbeam. The deuterium retention amounts (2–4)·1021 D/m2 for the incident fluence of ~2·1025 D/m2. The retention values are similar for both CFC materials and lower by ~20–40% for EK98. The retention in all three materials scales roughly with a square root of incident fluence without saturation for the range of fluences obtained. The majority of deuterium is stored in a surface layer of <8 μm. However, in NB31 deuterium is detectable as deep as 80 μm. The in-bulk retention estimated for a TEXTOR experimental campaign of ≍7500 s of plasma has a contribution of ≍10% to the total retention, which is dominated by deuterium-carbon co-deposition.

Hydrogen isotope retention in plasma-facing materials: review of recent experimental results

Recent data on deuterium retention in carbon fibre composites and tungsten both irradiated with D ions and exposed to D plasmas are presented. Deuterium depth profiles measured up to depths of 7–14 μm allow understanding of the mechanism which is responsible for the hydrogen isotope trapping in these materials. In the CFC materials the amount of retained deuterium increases with the ion fluence at all irradiation temperatures in the range from 323 to 723 K. No saturation is reached as observed in pyrolytic graphite. Depth profiles show that saturation occurs only within a near surface layer corresponding to the ion range. The increase in total retention at near-room temperature is accompanied by an increasing of the long profile tail extending beyond 14 μm with the D concentration of about 10−1 at.% at a depth of 10 μm for fluences above 1024 D m−2. The depth at which deuterium is retained in tungsten (W) can be divided into three zones: (i) the near-surface layer (up to a depth of 0.2–0.5 μm depending on ion energy), (ii) the sub-surface layer (from ∼0.5 to ∼2 μm), and (iii) the bulk (>5 μm). Low-energy D ion irradiation modifies the W structure to depths of up to about 5 μm, both for W single crystals and polycrystalline W. The high D concentration (0.1–0.3 at.%) at depths of 1–3 μm relates to accumulation of D2 molecules in vacancy clusters and voids. These defects are supposed to be generated due to plastic deformation of the W surface caused by deuterium supersaturation within the near-surface layer.

Tungsten and beryllium armour development for the JET ITER-like wall project

For the ITER-like wall project at JET the present main chamber CFC tiles will be exchanged with Be tiles and in parallel a fully tungsten-clad divertor will be prepared. Therefore three R&D programmes were initiated: Be coatings on Inconel as well as Be erosion markers were developed for the first wall of the main chamber. 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 divertor a conceptual design for a bulk W horizontal target plate was investigated, with the emphasis on minimizing electromagnetic forces. The design consisted of stacks of W lamellae of 6 mm width that were insulated in the toroidal direction. High heat flux tests of a test module were performed with an electron beam at an absorbed power density up to 9 MW m−2 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 micro-cracks on the loaded surface. For all other divertor parts R&D was performed to provide the technology to coat the 2-directional CFC material used at JET with thin tungsten coatings. The W-coated CFC tiles were subjected to heat loads with power densities ranging up to 23.5 MW m−2 and exposed to cyclic heat loading for 200 pulses at 10.5 MW m−2. All coatings developed cracks perpendicular to the CFC fibres due to the stronger contraction of the coating upon cool-down after the heat pulses.

Development of Carbon Materials and Plasma Facing Components for ITER

A significant R&D effort was carried out in EU to develop suitable CFC materials with a three-directional fibre structure. The main goal was to obtain a high thermal conductivity, which is required to remove the expected heat load with a sacrificial thickness of at least 15–20 mm, with a proper balance of the mechanical properties. Two CFC materials have been developed and are both suitable for the ITER divertor construction, namely NB31 by the French company Snecma Propulsion Solide and Concept 2 by the British company Dunlop Aerospace. The main R&D effort is now focussed in developing cheaper grades and optimising the manufacturing process.The development of suitable technologies for the manufacturing of carbon armoured high heat flux components resulted in the construction of a medium-scale divertor prototype, which was successfully tested well above the ITER design requirements. This component was followed by the manufacturing of a full-scale prototype, which is now being high heat flux tested. In parallel to the technology development, suitable non-destructive methods have been developed to inspect the delicate CFC/Cu joint. These techniques include: infrared examinations, lock-in thermography and ultrasonic examinations.An important topic that will be addressed in the near future is the definition of suitable acceptance criteria for carbon armoured components in view of the series production for ITER.

Detection of Heat Transfer Properties of CFC-Plasma Facing Components by IR-Observation and High Heat Flux Testing

Infrared inspections of a CFC mock-up designed for ITER were performed using the IRINA test facility (Infrared Inspection for Nondestructive Analysis). The impact of local differences in emissivity of the CFC material on temperature measurements and the detection limit of the presented analysis was studied. The tested CFC mock-up showed an area with poor heat transfer properties. The same CFC mock-up was loaded under high heat flux in JUDITH. Here, overheating was observed in the area with poor heat transfer properties. Thus the infrared inspection can be seen as successful tool to verify the quality of CFC mock-ups. However, further studies are necessary to define the types of defects.

A New Inspection Method for Plasma Facing Components

Transient thermography testing is an established method for testing the joint quality of CFC tiles in metallic heat sink systems. With this method hot pressurized water is introduced into the cooling channels of the elements and the transient heating of the CFC surface is recorded by an infrared camera. CFC tiles with a lack of bonding can be detected since they heat more slowly than well joined tiles. To perform this inspection a water flow in the tubes must be established and therefore the connection of the component to a hot water circuit is necessary. A new approach to this inspection problem is the application of lock-in thermography. With this method periodic illumination with powerful lamps of the CFC surface causes periodic heating of the CFC material. The phase shift between the periodic surface temperature evolution and the periodically operating heat source is recorded and evaluated. The lock-in thermography can be applied immediately after joining of the tiles to the component because the presence of a cooling channel system is not required for the application of this method. The method is suited to test the bonding of CFC tiles onto the component and the integrity of CFC tiles regarding possible internal cracks.

3-D simulation of macroscopic erosion of CFC under ITER off-normal heat loads

Carbon fiber composites, CFC, newly developed for the divertor armor of the future tokamak ITER, have a rather complex structure of fiber framework and carbon matrix, allowing CFC to match the thermal conductivity requirements for the tokamak stationary regimes. But for the ITER off-normal events, their behavior is still not investigated in detail. Recent experiments on electron beam facilities simulating the slow transient loads of 20 MW/m 2 have shown unusually high erosion rates of CFC in the temperature range of 3000–4500 K. To understand the mechanism of the enhanced erosion, numerical simulations for the new CFC are carried out using the Pegasus code that has been generalized to describe three-dimensional effects of brittle destruction and heat transport in the carbon composites. A new erosion mechanism is discovered. The enhanced erosion is caused by local overheating in CFC materials due to preferential cracking on the fiber surface and thermal isolation of the fibers from matrix. A large difference of fiber and matrix coefficients of thermal expansion is essential for the mechanism.

Study of brittle destruction and erosion mechanisms of carbon-based materials during plasma instabilities

Erosion damage due to plasma instabilities such as hard disruptions, edge-localized modes, and vertical displacement events remains a major obstacle to successful realization of the tokamak-reactor concept. As a result of these plasma instabilities, intense plasma energy that is deposited during short periods can cause severe erosion, structural damage, and surface modifications of the plasma-facing materials. Experimental work is being carried out at the high-power VIKA-93 plasma-gun facility in the Efremov Institute, Russia. Interesting results were obtained during preliminary heating of the samples (to 1200°C) and use of maximum plasma gun parameters, i.e., E in =30 MJ/m 2, τ=360 μ s. In all samples, a large increase in weight loss (up to 80%) was observed during plasma bombardment when preheating was used. Scanning electron microscope investigations have demonstrated a considerable evolution of surface recrystallization processes, especially for preheated CFC materials. Significant differences among various carbon materials are found for specimens with and without preliminary heating.

Hydrogen retention and erosion of carbon–tungsten mixed material

Upon machine operation, the wall surface in the divertor region of ITER results to be covered by a mixture of carbon and tungsten, due to erosion and re-deposition of CFC and tungsten wall materials. Thus, it is necessary to evaluate the fuel hydrogen retention and the erosion of such mixed materials. In the present study, the deuterium desorption spectra were examined upon deuterium ion irradiation for the materials with tungsten concentrations of 6, 13 and 32 at.%. Two desorption peaks appeared at approximately 150 and 700°C. The peaks at 150 and 700°C correspond to the desorptions from tungsten and carbon, respectively. The peak height at 150°C increased and that at 700°C decreased as the tungsten concentration increased. The total amount of deuterium retained was reduced by the addition of tungsten, as compared with a case of graphite. The desorption of CD 4 largely decreased with the tungsten concentration increase. The present data show that both the hydrogen retention and the chemical erosion of carbon–tungsten mixed material were reduced, as compared to graphite.

Measurements of Chemical Erosion in the Plasmagenerator PSI-1

The stationary plasma of the PSI-1 facility was used for studying the ion energy dependence of chemical erosion of different CFC materials in deuterium plasmas. A weak dependence on the bombarding energy was found having a maximum between 150 and 200 eV. Furthermore, long term exposure experiments with two different CFC materials were carried out showing that the reduction of chemical erosion can be explained by covering the surface with impurities. A study of the contribution of energetic neutrals to erosion revealed effects of the order of 50%; their influence, however, depends on the plasma conditions.

Thermal Fatigue Testing of Actively Cooled Divertor Mock-Ups after Neutron Irradiation

In order to study degradation effects of neutrons on plasma-facing materials and joints, actively-cooled beryllium and CFC samples were irradiated in the High Flux Reactor in Petten up to 0.35 dpa at 350 and 700°C. Later, these samples were tested by means of an electron beam facility under static and cyclic heating conditions. The heat removal efficiency and the thermal fatigue behavior of these samples were compared to those of corresponding non-irradiated samples. A significant increase of surface temperature was observed for all samples, due to a reduced thermal conductivity of the CFC materials after neutron irradiation. This effect is less distinctive for samples irradiated at the higher temperature. Long term fatigue tests with 1000 heating cycles at 15 MW/m2 did not create any failure of the plasma-facing material or the bond layer of the tested mock-ups. Similar experiments have been performed with brazed beryllium-copper mock-ups. Flat tile mock-ups with an S65 C armor on a CuCrZr heat sink were loaded up to 1000 cycles at a power density of 7 MW/m2 without detachment of tiles.

Erosion of carbon-based materials in a steady-state deuterium plasma

Experimental results on plasma induced erosion of graphites and CFC materials are presented. Experiments were performed in the LENTA linear simulator allowing the ion fluence corresponding to the ITER divertor conditions for one operation cycle (∼10 22 D +/cm 2) to be achieved in a few hours. RG–Ti–91 titanium doped and RG–Ti–91 with 0.1% boron addition graphites (Russia), American POCO-AXF-5Q and European composites SEP N112-A and DUNLOP-exp B were taken for the study to compare their erosion properties under similar conditions. Erosion yield has been evaluated using weight loss measurements. Plasma contamination is monitored by optical spectroscopy and gas composition is analyzed by mass-spectrometric method during exposure. Both methods have shown intensive deuterocarbon formation for the ITER relevant working temperatures of the samples, 750°C and 1150°C.