Candidate Material For Plasma Facing Components Research Articles

New insights into microstructure of neutron-irradiated tungsten

The development of appropriate materials for fusion reactors that can sustain high neutron fluence at elevated temperatures remains a great challenge. Tungsten is one of the promising candidate materials for plasma-facing components of future fusion reactors, due to several favorable properties as for example a high melting point, a high sputtering resistivity, and a low coefficient of thermal expansion. The microstructural details of a tungsten sample with a 1.25 dpa (displacements per atom) damage dose after neutron irradiation at 800 °C were examined by transmission electron microscopy. Three types of radiation-induced defects were observed, analyzed and characterized: (1) voids with sizes ranging from 10 to 65 nm, (2) dislocation loops with a size of up to 10 nm and (3) W–Re–Os containing σ- and χ-type precipitates. The distribution of voids as well as the nature of the occurring dislocation loops were studied in detail. In addition, nano-chemical analyses revealed that the σ- and χ-type precipitates, which are sometimes attached to voids, are surrounded by a solid solution cloud enriched with Re. For the first time the crystallographic orientation relationship of the σ- and χ-phases to the W-matrix was specified. Furthermore, electron energy-loss spectroscopy could not unambiguously verify the presence of He within individual voids.

Fracture behavior of random distributed short tungsten fiber-reinforced tungsten composites

In future fusion reactors, tungsten is considered as the main candidate material for plasma-facing components. However, the intrinsic brittleness of tungsten is of great concern during operation. To overcome this drawback, tungsten fiber-reinforced tungsten composites (Wf/W) are being developed relying on extrinsic toughening principles. Tungsten (W) fibers with extremely high tensile strength and ductility are used to reinforce a tungsten matrix.In this work, field assisted sintering technology (FAST) is used to produce Wf/W material. Mechanical characterizations including Charpy impact and 3-point bending tests are performed. Based on the 3-point bending test results, the Wf/W materials can facilitate a promising pseudo-ductile behavior even at room temperature, similar to fiber reinforced ceramic composites. Fracture energy density and fracture toughness together with the crack-resistance curves (R-curves) are measured. Compared to conventional pure tungsten, Wf/W shows significant improvement in fracture toughness.

Molecular dynamics study of trapping and detrapping process of hydrogen in tungsten vacancy

Tungsten (W) alloys and W-based alloys are the primary candidate materials for plasma-facing components in future fusion reactors (e.g. ITER and CFETR). One of the critical issues still to be clarified in the design of the fusion reactor materials is the retention of hydrogen (H) isotopes in W, when the plasma-facing materials are supposed to sustain high-flux plasma and high-energy neutron. The dynamical behaviours of H in W with radiation defects (e.g. vacancy) are of serious concerns for understanding the mechanism of H capture, retention and permeation in W. In this work, a new model to extract the effective capture radius (ECR) and dissociation coefficient simultaneously is presented through coupling the trapping process and detrapping process of H in W vacancy. In the new model, the quantity ratio of vacancy to H atom in vacancy-H complex (VH<sub><i>x</i>+1</sub>) in the molecular dynamics (MD) simulations is described as a function of time, while the exact occurrence time of corresponding event is not required. This new model, combined with extensive MD calculations, enables the simultaneous determining of the ECR and dissociation coefficient of H in W vacancy. It is found that the parameters are dependent not only on the event type but also on temperature. The dissociation energy of H from vacancy-H complex decreases gradually with the increase of the trapped number of H atoms in the vacancy-H complex. It is also found that the common assumption (i.e. the ECR is equal to one lattice constant and the pre-exponential factor is equal to 10<sup>13</sup> s<sup>–1</sup>) in the long-term simulation methods (e.g. kinetic Monte Carlo and rate theory) is not always valid, since these calculated dynamical parameters are dispersive. The new model to obtain more reliable results with lower cost of computing resources can be easily extended into the other similar kinetic processes (e.g. H/He trapping and detrapping processes in other materials systems). These calculated dynamical parameters should be potentially helpful in supplying the initial input parameters for the long-term simulation methods.

Molecular dynamics simulation of hydrogen and helium trapping in tungsten

Tungsten has been chosen as the divertor armour material in ITER and is the main candidate material for plasma-facing components for future fusion reactors. Interaction of plasma components with the material leads to degradation of the performance and thus the lifetime of the in-vessel components. On top of that special attention is drawn to tritium retention in the reactors vessel from a safety point of view, since tritium is radioactive material. In order to gain better understanding of the mechanisms driving accumulation of plasma components in the material and subsequent degradation of the material, atomistic simulations are employed. The focus of this work is on so-called self trapping of H and He atoms or, in other words, Frenkel pair formation in bulk tungsten in the presence of H and He atoms. Two versions of a model embedded atom interatomic potential and a bond order potential were tested by comparing it with ab initio data regarding the binding properties of pure He and He-H-Vacancy clusters and energetics of Frenkel pair formation. As a result of Molecular Dynamics simulations at finite temperature, the values of critical H concentration needed for the generation of a Frenkel pair in the presence of He clusters were obtained. The results show that the critical H concentration decreases with the size of He cluster present in the simulation cell and thus, Frenkel pair formation by H is facilitated in the presence of He clusters in the material.

Influence of the interface strength on the mechanical properties of discontinuous tungsten fiber-reinforced tungsten composites produced by field assisted sintering technology

In future fusion reactors, tungsten is a main candidate material for plasma-facing components. However, the intrinsic brittleness of tungsten is an issue under the extreme fusion environment. To overcome this drawback, tungsten fiber-reinforced tungsten (Wf/W) composites are being developed relying on an extrinsic toughening principle. In this study Wf/W composites are produced by a Field-Assisted Sintering Technology (FAST) process with different fiber–matrix interfaces. The fracture behavior was studied by 3-point bending tests on notched samples. 4-point bending tests and tensile tests are performed to measure the flexural strength and tensile strength, respectively. Wf/W with a weak interface shows a typical pseudo-ductile fracture behavior, similar to ceramic matrix composites. A strong interface is beneficial to achieve higher flexural strength and tensile strength, but in turn, weakens the pseudo-ductile behavior.

In situ investigation of helium fuzz growth on tungsten in relation to ion flux, fluence, surface temperature and ion energy using infrared imaging in PSI-2

Tungsten is a candidate material for plasma-facing components in nuclear fusion reactors. In operation it will face temperatures >800 K together with an influx of helium ions. Previously, the evolution of special surface nanostructures called fuzz was found under these conditions in a limited window of surface temperature, ion flux and ion energy. Fuzz potentially leads to lower heat load tolerances, enhanced erosion and dust formation, hence should be avoided in a fusion reactor. Here the fuzz growth is reinvestigated in situ during its growth by considering its impact on the surfaces infrared emissivity at 4 μm wavelength with an infrared camera in the linear plasma device PSI-2. A hole in the surface serves as an emissivity reference to calibrate fuzz thickness versus infrared emissivity. Among new data on the above mentioned relations, a lower fuzz growth threshold of 815 ± 24 K is found. Fuzz is seen to grow on rough and polished surfaces and even on the hole’s side walls alike. Literature scalings for thickness, flux and time relations of the fuzz growth rate could not be reproduced, but for the temperature scaling a good agreement to the Arrhenius equation was found.

Properties of Vacancy Complexes with Hydrogen and Helium Atoms in Tungsten from First Principles

Tungsten and its alloys are the primary candidate materials for plasma-facing components in fusion reactors. The material is exposed to high-energy neutrons and the high flux of helium and hydrogen atoms. In this work we have studied the properties of vacancy clusters and their interaction with H and He in W using density functional theory. Convergence of calculations with respect to modeling cell size was investigated. It is demonstrated that vacancy cluster formation energy converges with small cells with a size of 6 × 6 × 6 (432 lattice sites) enough to consider a microvoid of up to six vacancies with high accuracy. Most of the vacancy clusters containing fewer than six vacancies are unstable. Introducing He or H atoms increases their binding energy potentially making gas-filled bubbles stable. According to the results of the calculations, the H2 molecule is unstable in clusters containing six or fewer vacancies.

Influence of irradiation with low-energy helium ions on graphite and tungsten for fusion applications

Plasma-facing components (PFCs) in fusion devices are exposed to an irradiation environment of high particle flux. Among various candidate materials for PFCs, graphite is used as a first-wall material of the Korea Superconducting Tokamak Advanced Research (KSTAR) and tungsten is being strongly considered as a divertor material of the international thermonuclear experimental reactor (ITER). In this experiment, graphite and tungsten targets were irradiated by using low-energy helium ions, and the irradiation’s influence on each target was studied with particular emphasis on the surface change. Changes in the surface morphology were observed by using scanning electron microscopy (SEM); induced disorders in the structure were investigated by using X-ray diffraction (XRD), Raman spectroscopy, and electron spin resonance (ESR) measurements.

Neutron-induced transmutation effects in W and W-alloys in a fusion environment

W and W-alloys are among the primary candidate materials for plasma-facing components in the design of fusion reactors, particularly in high-heat-flux regions such as the divertor. Under neutron irradiation W undergoes transmutation to its near-neighbours in the periodic table. Additionally He and H are particles emitted from certain neutron-induced reactions, and this is particularly significant in fusion research since the presence of helium in a material can cause both swelling and a strong increase in brittleness. This paper presents the results of inventory burn-up calculations on pure W and gives quantitative estimates for He production rates in both a fusion-reactor environment and under conditions expected in the ITER experimental device. Transmutation reactions in possible alloying elements (Re, Ta, Ti and V), which could be used to reduce the brittleness of pure W, are also considered. Additionally, for comparison, the transmutation of other fusion-relevant materials, including Fe and SiC, are presented.

Fabrication and Characterization of W-Re-Os Alloys for Studying Transmutation Effects of W in Fusion Reactors

Tungsten (W) is a candidate material for plasma-facing components in fusion reactors. Large amounts of solid transmuted elements of W such as Re and Os will be produced under fusion neutron irradiation. In order to investigate the effects of transmutation products, a series of W-Re-Os alloys were fabricated by an arc melting method. The nominal compositions of these samples were selected based on theoretical predictions of changes in chemical composition for the transmutation of W to Re and Os. Significant hardening, an increase in electrical resistivity and changes in the lattice constant were observed as a function of the Re and Os content. Microstructural observations revealed the existence of the a phase precipitate, which did not significantly affect physical properties and hardness compared with solid solution effect. The main objectives of this work were to demonstrate the processes of material fabrication and to study unirradiated material properties.

Thermal Shock Strength of Carbon-Carbon Composite by Laser Irradiation Technique Based on Shear Strength Criterion

The laminated carbon-carbon (C/C) composite is candidate material for plasma-facing components in fusion reactors. Its thermal shock strength must be evaluated for an application to the components facing the high-temperature plasma field. However, few studies have been done. A laser irradiation method newly proposed was applied effectively to lamination of C/C composite for the evaluation of thermal shock strength by subjecting a heat flux up to 27 J/mm2 on the specimen surface. By introducing the critical power density at which the material fracture occurs, thermal shock strength was obtained. It was concluded that the critical power density could be a new measure to evaluate thermal shock strength for laminated C/C composite.

Review of beryllium and tungsten erosion behavior and universal modeling of plasma impurity effects observed in recent PISCES experiments

Although they have rather drastically different characteristics, currently beryllium and tungsten are both considered as candidate materials for plasma-facing components in the International Thermonuclear Experimental Reactor (ITER). The erosion behavior of these materials under steadystate deuterium plasma bombardment has recently been investigated using the PISCES-B Mod facility. Several effects of plasma impurities on the materials erosion behavior have been observed in these experiments. At elevated temperatures beryllium tends to be contaminated with carboncontaining plasma impurities, resulting in the deposition of carbonaceous films. These films shield beryllium from plasma bombardment and reduce erosion. In contrast, no sign of impurity deposition has been found with tungsten. However, the overall erosion behavior of tungsten is dictated by trace amounts of oxygen-containing impurities at energies below the sputtering threshold by deuterium. A universal materials balance model is presented to explain these impurity effects.

Simulation of Kinetic Electron Emission from Beryllium by keV Ion Impacts

A Monte Carlo simulation model of ion-induced kinetic secondary electron emission from beryllium, a candidate material for plasma-facing components in thermonuclear fusion devices, is developed. In the model, the conduction electron excitation by a projectile ion and the cascade multiplication process of the excited electrons are involved in production of secondary electrons in the solid. Due to increase in the excitation probability, the secondary electron yield increases with increasing impact energy and initial charge state of the projectile, and decreasing mass of the projectile. Nevertheless, the majority of secondary electrons are produced through the cascade process, so that the peak of the energy distribution of secondary electrons depends little on the impact energy and the species of the projectile. Due to increase in the energy transferred from the projectile to the electrons, however, the distribution progressively broadens towards the high-energy side as the impact energy is increased or as the mass of the projectile is reduced.

Erosion and redeposition behavior of selected NET-candidate materials under high-flux hydrogen, deuterium plasma bombardment in pisces

Plasma erosion and redeposition behavior of selected candidate materials for plasma-facing components in the NET-machine have been investigated using the PISCES-A facility. Materials studied include SiC-impregnated graphite, 2D graphite weaves with and without CVD-SiC coatings, and isotropic graphite. These specimens were exposed to continuous hydrogen or deuterium plasmas under the following conditions: electron temperature range from 5 to 35 eV; plasma density range from 5 × 10 11 to 1 × 10 12 cm −3; flux range from 5 × 10 17 to 2 × 10 18 ions cm −2 s −1; fluence of the order from 10 21 to 10 22 ions/cm 2; bombarding energies of 50 and 100 eV; target temperature range from 300 to 1000°C. The erosion yield of SiC-impregnated graphite due to deuterium plasma bombardment is found to be a factor of 2 to 3 less than that of isotropic graphite materials. A further factor of 2–3 reduction in the erosion yield is observed in when redeposition associated with reionization of sputtered particle becomes significant. From post-bombardment surface analysis with AES, the surface composition in terms of the Si/C of SiC-impregnated graphite ratio is found to increase from 0.15 to 0.7 after hydrogen plasma bombardment to a fluence around 4 × 10 21 ions/cm 2 at 350 ° C. However, the final surface composition appears to remain unchanged up to 4 × 10 22 ions/cm 2, the highest fluence in the present study. Significant surface morphological modifications of SiC-impregnated graphite are observed after the high-fluence plasma exposure. Several structural problems such as coating-substrate adhesion have been pointed out for SiC-coated 2D graphite weave.