Materials For Plasma Facing Components Research Articles

Helium irradiation behavior of tungsten-niobium alloys under different ion energies

Tungsten is a primary material in plasma-facing components (PFCs), but several disadvantages limit its application in fusion reactors. This work examined tungsten–niobium (W–Nb) alloys as potential PFC materials and investigated their response to low-energy helium ion irradiation. W–Nb powders were milled for 25 and 36 h, and the W–Nb samples were exposed to 50 and 80 eV helium ion irradiation with a flux of 1.5 × 1022 ions/m2 s at 1230 °C. The sample with a dense structure and obtained after a 25-h milling exhibited enhanced anti-irradiation performance. Sufficient energy can be provided to promote helium ion migration in the matrix, resulting in the formation of nanotendrils. Furthermore, in W–Nb alloys, the Nb-rich regions are more attractive to helium ions and thus provide protection to the tungsten matrix.

Effects of Welding Parameters on Mechanical Properties in Electron Beam Welded CuCrZr Alloy Plates

CuCrZr alloys are attractive structural materials for plasma-facing components (PFC) and heat sink element in the International Thermonuclear Experimental Reactor (ITER) fusion reactors. This material has gained so much attention because of its high thermal conductivity and fracture toughness, high resistance to radiation damage and stability at elevated temperatures. The objective of this work is to study the effects of electron beam welding parameters on the mechanical strength of the butt welded CuCrZr joint. Taguchi method is used as the design of experiments to optimize the input parameters, such as accelerating voltage, beam current, welding speed, oscillation amplitude and frequency. The joint strength and ductility are the desired responses, which are measured through ultimate tensile strength and percent elongation, respectively. Accelerating voltage and welding speed are found to have significant influence on the strength. A combination of low amplitude and high-frequency oscillation is suggested for the higher joint strength and ductility. There is a close agreement between Taguchi predicted results and experimental ones. Fractographic analysis of joint and weld zone analysis are carried out to study the failure behaviour and microstructural variation in the weld zone, respectively.

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.

Effects of in situ dual ion beam (He+ and D+) irradiation with simultaneous pulsed heat loading on surface morphology evolution of tungsten–tantalum alloys

The strong thermal and mechanical properties of tungsten (W) are well suited for the harsh fusion environment. However, increasing interest in using tungsten as plasma-facing components (PFCs) has revealed several key issues. These potential roadblocks necessitate more investigation of W and other alternative W based materials exposed to realistic fusion conditions. In this work, W and tungsten–tantalum (W–Ta) alloys were exposed to single (He+) and dual (He+ + D+) ion irradiations with simultaneous pulsed heat loading to elucidate PFCs response under more realistic conditions. Laser only exposer revealed significantly more damage in W–Ta samples as compared to pure W samples. This was due to the difference in the mechanical properties of the two different materials. Further erosion studies were conducted to evaluate the material degradation due to transient heat loading in both the presence and absence of He+ and/or D+ ions. We concluded that erosion of PFC materials was significantly enhanced due to the presence of ion irradiation. This is important as it demonstrates that there are key synergistic effects resulting from more realistic fusion loading conditions that need to be considered when evaluating the response of plasma facing materials.

Deuterium trapping at vacancy clusters in electron/neutron-irradiated tungsten studied by positron annihilation spectroscopy

Deuterium trapping at irradiation-induced defects in tungsten, a candidate material for plasma facing components in fusion reactors, was revealed by positron annihilation spectroscopy. Pure tungsten was electron-irradiated (8.5 MeV at ∼373 K and to a dose of ∼1 × 10−3 dpa) or neutron-irradiated (at 573 K to a dose of ∼0.3 dpa), followed by post-irradiation annealing at 573 K for 100 h in deuterium gas of ∼0.1 MPa. In both cases of electron- or neutron-irradiation, vacancy clusters were found by positron lifetime measurements. In addition, positron annihilation with deuterium electrons was demonstrated by coincidence Doppler broadening measurements, directly indicating deuterium trapping at vacancy-type defects. This is expected to cause significant increase in deuterium retention in irradiated-tungsten.

Melt layer erosion during ELM-like heat loading on molybdenum as an alternative plasma-facing material

Transient events that occur during plasma instabilities in fusion reactors impart large heat fluxes onto the surrounding plasma-facing components (PFCs). Erosion and splashing of PFCs can contaminate the plasma and shorten material lifetime. Although tungsten is currently considered the most promising candidate material for future PFCs, concerns over the thermal shock performance during type-I ELMs (transient events expected in fusion devices) necessitate the study of other comparable materials. ELM-like heat loading was applied via a pulsed Nd:YAG millisecond laser on a pristine molybdenum (Mo) surface to measure surface melting and mass loss. One potential advantage of Mo is its higher specific heat of vaporization, which could lead to reduced particle emission. Imaging of the surface after loading revealed that complete surface melting began at 1.0 MJ m−2 (heat load parameter of 31.62 MJ m−2 s−1/2). Photon excitation also increased significantly above 1.0 MJ m−2, indicating possible phase change. At 1.4 MJ m−2 (44.27 MJ m−2 s−1/2), in situ mass loss measurements found an exponential increase in particle emission, indicating the presence of droplet formation and boiling. Direct comparisons of erosion during pulsed heat loading between PFC candidate materials will ensure that future fusion devices design components with optimal thermal strength.

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.

Tungsten coating by ATC plasma spraying on CFC for WEST tokamak

In the field of fusion experiments using a tokamak, the plasma facing components (PFC) are the closest object to the hot plasma. Due to the plasma-wall interaction, the material composing the PFC may enter the plasma and disturb the experiments. In the past, the main material for PFC was carbon (CFC, graphite), while the future reactors like ITER will be fully metallic, in particular tungsten. The Tore Supra tokamak has been transformed in an x-point divertor fusion device within the frame of the WEST (W (tungsten) Environment in Steady-state Tokamak) project in order to have plasma conditions close to those expected in ITER. The PFC other than the divertor has been coated with W to transform Tore Supra into a fully metallic environment. Different coating techniques have been selected for different kind of PFC. This paper gives an overview on the coating process used for the antennae protection limiter, the associated validation programme and concludes on the adequacy of the W coating with the WEST experimental programme requirements and gives perspectives on the development to be pursued.

Tungsten ion source under double-pulse laser ablation system

New tungsten ion source is produced by using single and double-pulse laser ablation system. Combined collinear Nd:YAG laser beams (266+1064 nm) are optimized to focus on the sample in air. Optimization of the experimental parameters is achieved to enhance the signal-to-noise ratio of the emission spectra. The velocity distribution of the emitted plasma cloud is carefully measured. The influences of the potential difference between the bias electrodes, laser wavelength and intensity on the current signal are also studied. The results show that the increase in the tungsten ion velocity under the double-pulse lasers causes the output current signal to increase by about three folds. The electron density and temperature are calculated by using the Stark-broadened line profile of tungsten line and Boltzmann plot method of the upper energy levels, respectively. The signal intensity dependence of the tungsten ion angular distribution is also analyzed. The results indicate that the double-pulse laser ablation configuration is more potent technique for producing more metal ion source deposition, thin film formation, and activated plasma-facing component material.

Current Status of Korea Heat Load Test Facility KoHLT-EB for Fusion Reactor Materials

The Korea heat load test facility using electron beam (KoHLT-EB) was constructed and commissioned to perform high heat flux testing on plasma facing components (PFCs), such as ITER blanket first wall, the tungsten divertor, and heat transfer experiments on other cooling devices. Also it will be used for the thermo-hydraulic performance test of ITER helium cooled ceramic reflector test blanket module in Korea. Qualification tests will be performed to evaluate the use of the high heat flux test facility on PFC and fusion reactor materials. For the thermal fatigue test of PFCs, two types of tungsten mockups were fabricated. One is a coated tungsten on the ferritic–martensitic steel (FMS). The other is a hot isostatic pressing bonded tungsten with FMS. Also, the integrity and cooling performance should be tested under the same or similar operation conditions of the ITER PFCs. After the completion of the preliminary mockup test and the facility qualification, the high heat flux test facility, KoHLT-EB, will provide the performance test for the various PFCs in fusion reactor materials and PFCs.

The effect of low energy helium ion irradiation on tungsten-tantalum (W-Ta) alloys under fusion relevant conditions

Currently, tungsten remains the best candidate for plasma-facing components (PFCs) for future fusion devices because of its high melting point, low erosion, and strong mechanical properties. However, continued investigation has shown tungsten to undergo severe morphology changes under fusion-like conditions. These results motivate the study of innovative PFC materials which are resistant to surface morphology evolution. The goal of this work is to examine tungsten-tantalum (W-Ta) alloys, a potential PFC material, and their response to low energy helium ion irradiation. Specifically, W-Ta samples are exposed to 100 eV helium irradiations with a flux of 1.15 × 1021 ions m−2 s−1, at 873 K, 1023 K, and 1173 K for 1 h duration. Scanning electron microscopy (SEM) reveals significant changes in surface deterioration due to helium ion irradiation as a function of both temperature and tantalum concentration in W-Ta samples. X-Ray Diffraction (XRD) studies show a slight lattice parameter expansion in W-Ta alloy samples compared to pure W samples. The observed lattice parameter expansion in W-Ta alloy samples (proportional to increasing Ta wt.% concentrations) reflect significant differences observed in the evolution of surface morphology, i.e., fuzz development processes for both increasing Ta wt.% concentration and target temperature. These results suggest a correlation between the observed morphology differences and the induced crystal structure change caused by the presence of tantalum. Shifts in the XRD peaks before and after 100 eV helium irradiation with a flux of 1.15 × 1021 ions m−2 s−1, 1023 K, for 1 h showed a significant difference in the magnitude of the shift. This has suggested a possible link between the atomic spacing of the material and the accumulated damage. Ongoing research is needed on W-Ta alloys and other innovative materials for their application as irradiation resistant materials in future fusion or irradiation environments.

Influence of the first wall material on the particle fuelling in ASDEX Upgrade

In the period from 2002 to 2007 the material of the plasma facing components (PFCs) of ASDEX Upgrade (AUG) was changed from carbon (C) to tungsten (W). Comparing the measured density profiles of low-density L-mode discharges with little or no gas puff before and after this modification, a significantly higher pedestal-top density was found for W PFCs together with a steeper gradient and a lower pedestal temperature. This change can be explained by larger particle- and energy reflection coefficients for D on W compared to D on C, as shown by EMC3-EIRENE simulations of AUG discharges in similar conditions on a computational grid extending to the main chamber first wall. In the simulations, a change of the wall material at fixed separatrix density indeed shows that for W PFCs more neutrals cross the separatrix, resulting in a steeper density gradient. Analysis of the source resolved and poloidally resolved neutral flux densities across the separatrix show a dominant contribution of the divertor targets to the fuelling profile in the simulation of the low density case. Increasing the density decreases the electron temperature at the target and therefore the potential drop in the electrostatic sheath as well as the energy of the ions impinging on the surface. Neutrals with ∼eV energies, able to reach the separatrix, are then only produced via molecular dissociation processes in the plasma volume independently of the PFC material. Also the contribution of the main chamber PFCs to the fuelling is observed to increase at higher densities.

Secondary radiation damage and gas production in plasma facing materials under fusion neutron irradiation

Primary radiation damage (atomic displacements) and Helium and Hydrogen production rates in plasma facing components (PFCs) of a fusion system are usually determined by the high energy part of neutron spectra formed in its plasma chamber from the initial fusion neutron source. The energetic alphas and protons, appearing in PFC materials in the (n,α) and (n,p) threshold reactions, per se may cause an additional, secondary material damage. In most cases this additional damage is non-essentially small. Ex altera parte an extra Helium production may become apparent in water cooled systems especially. Material inventory considerations show that Os, Ni and Zn as transmutation products generated in W, Cu-alloys and steels under long term neutron irradiation account for the secondary sources of intensive Helium production in (n,α) reactions with these transmutants. These even if preliminary evaluated responses should be taken into account in new material developments for long term fusion concepts.

Melt infiltrated tungsten–copper composites as advanced heat sink materials for plasma facing components of future nuclear fusion devices

The exhaust of power and particles is regarded as a major challenge in view of the design of a magnetic confinement nuclear fusion demonstration power plant (DEMO). In such a reactor, highly loaded plasma facing components (PFCs), like the divertor targets, have to withstand both severe heat flux loads and considerable neutron irradiation. Existing divertor target designs make use of monolithic tungsten (W) and copper (Cu) material grades that are combined in a PFC. Such an approach, however, bears engineering difficulties as W and Cu are materials with inherently different thermomechanical properties and their optimum operating temperature windows do not overlap. Against this background, W–Cu composite materials are promising candidates regarding the application to the heat sink of highly loaded PFCs. The present contribution summarises recent results regarding the manufacturing and characterisation of such W–Cu composite materials produced by means of liquid Cu melt infiltration of open porous W preforms. On the one hand, this includes composites manufactured by infiltrating powder metallurgically produced W skeletons. On the other hand, W–Cu composites based on textile technologically produced fibrous reinforcement preforms are discussed.

Topographical, structural and hardness changes in surface layer of stainless steel-AISI 304 irradiated by fusion-relevant high energy deuterium ions and neutrons in a low energy plasma focus device

This paper aims to investigate the effects of fusion relevant intense pulse of energetic and high fluence deuterium ions and neutrons generated in a low energy (3.0kJ) plasma focus device (UNU/ICTP PFF) on stainless steel-AISI 304 samples. The irradiation was performed using different number of deuterium plasma focus shots (1, 3, 5 and 10) at a fixed axial distance of 8cm from the anode tip. X-ray diffraction spectra of irradiated samples indicated the structural phase transition from mixed ferittic-austenitic form in virgin AISI 304 SS sample to only austenitic form indicated by the presence of only γ-Fe diffraction peaks and increase in corresponding peak intensities. A small shift in all austenitic diffraction peaks towards the higher diffraction angles was observed indicating the decrease in lattice spacing by vacancy defects most probably due to energetic fusion neutrons. Field emission scanning electron microscopy and atomic force microscopy images showed the formation of nanometer sized particles on irradiated samples' surfaces. The size of nanoparticles initially decreased as the number of focus shots was increased from 1 to 3 and to 5, and later as the number of focus shots was increased to 10, the aggregations of the particles occurred to form bigger clusters. The sample surface hardness was found to decrease with increasing number of irradiation shots which can be attributed to the induction of vacancy defects by 2.45MeV fusion neutron irradiations and the increasing deposition of the copper ablated from the anode-top on the irradiated samples. The study clearly established that the plasma focus device provides intense fusion relevant pulses suitable for causing the topographical, structural and hardness changes on the surface of the irradiated SS AISI 304 which is one of the candidate materials for plasma facing components in fusion reactors.

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.

Surface modification of He pre-exposed tungsten samples by He plasma impact in the divertor manipulator of ASDEX Upgrade

Tungsten (W) will be used as material for plasma-facing components (PFCs) in the divertor of ITER and interact with Helium (He) ions either from initial He plasma operation or from Deuterium-Tritium (DT) fusion reactions in the active operation phase. Laboratory experiments reported that in a specific operational window of impact energy, ion fluence, and surface temperature (Ein ≥ 20 eV, ϕ ≥ 1 × 1024 He+mBt=2.5T,Ip=0.8MA,Paux≃8.0MWTsurf ≥ 1000 K) a modification of W surfaces occurs resulting in the formation of He-induced W nanostructures. Experiments in ASDEX Upgrade H-mode plasmas (Ein=37keV T, ϕ≃0.75×1024He0m−2 MA, Paux ≃ 8.0 MW) in He have been carried out to investigate in detail (a) the potential growth of W nanostructures on pre-damaged W samples incorporating He nanobubbles, and (b) the potential ELM-induced erosion of W nanostructure. Both W surface modifications were generated artificially in the GLADIS facility by He bombardment of W samples at ϕ≃1×1024He0m−2 keV (a) to ϕ ≃ 0.75 × 1024 He0mϕ≃1.6×1024He+m−2 at Tsurf ≃ 1800 K and (b) ϕ ≃ 1 × 1024 He0m−2 at Tsurf ≃ 2300 K prior to exposure in the divertor manipulator of ASDEX Upgrade. Though in part (a) conditions of W nanostructure growth with a total He ion fluence of ϕ ≃ 1.6 × 1024 He+m−2 and peak He ion impact energies above 150 eV were met, no growth could be detected. In part (b) lower density plasmas with more pronounced type I ELMs, carrying energetic He ions in the keV range, were executed with the strike-line positioned on 2 µm thick W nanostructure accumulating a fluence of ϕ ≃ 0.8 × 1024 He+m−2. Post-mortem analysis revealed that co-deposition by predominantly W, and Boron (B), eroded at the main chamber wall and transported into the divertor, took place on all W samples. Erosion of W nanostructure or its formation was hindered by the fact that the outer divertor at the location of the samples was turned under these He plasma conditions into a net deposition zone by W, B and Carbon (C) ions. The surface morphology with large roughness and effective surface area act as a catcher for the impinging impurities. Thus, apart from operation in the existence diagram of W nanostructure with respect to Tsurf, ϕ, and Ein, also the impinging impurity flux contribution needs to be considered in predictions concerning the formation of W nanostructures.

Tritium decay helium-3 effects in tungsten

Tritium (T) implanted by plasmas diffuses into bulk material, especially rapidly at elevated temperatures, and becomes trapped in neutron radiation-induced defects in materials that act as trapping sites for the tritium. The trapped tritium atoms will decay to produce helium-3 (3He) atoms at a half-life of 12.3 years. 3He has a large cross section for absorbing thermal neutrons, which after absorbing a neutron produces hydrogen (H) and tritium ions with a combined kinetic energy of 0.76 MeV through the 3He(n,H)T nuclear reaction. The purpose of this paper is to quantify the 3He produced in tungsten by tritium decay compared to the neutron-induced helium-4 (4He) produced in tungsten. This is important given the fact that helium in materials not only creates microstructural damage in the bulk of the material but alters surface morphology of the material effecting plasma-surface interaction process (e.g. material evolution, erosion and tritium behavior) of plasma-facing component materials. Effects of tritium decay 3He in tungsten are investigated here with a simple model that predicts quantity of 3He produced in a fusion DEMO FW based on a neutron energy spectrum found in literature. This study reveals that: (1) helium-3 concentration was equilibrated to ∼6% of initial/trapped tritium concentration, (2) tritium concentration remained approximately constant (94% of initial tritium concentration), and (3) displacement damage from 3He(n,H)T nuclear reaction became >1 dpa/year in DEMO FW.

Experimental mechanistic investigation of the nanostructuring of tungsten with low energy helium plasmas

Helium ion bombardment of tungsten at temperatures between approximately one third and one half of its melting point has shown growth of nanostructures colloquially referred to as “fuzz”. The nanostructures take the form of thin tendrils of diameter about 30 nm and grow out of the bulk material. Tungsten will and does compose one of the key materials for plasma facing components (PFCs) in fusion reactors. The formation of nanostructured fuzz layers on PFCs would be detrimental to the performance of the reactor, and must therefore be avoided. Previous experiments have shown evidence that tungsten fuzz is initially grown by loop punching of helium bubbles created in the bulk. However, once the tendrils grow to sufficient length, the tendrils should intercept the entire helium flux, halting the production of fuzz. Fuzz continues to grow though. To increase the understanding of the mechanisms of tungsten fuzz formation, and thereby aid the avoidance of its production, a series of tests were performed to examine the validity of several theories regarding later stage tungsten fuzz growth. Tests showed that the fuzz formation was dependent solely on the bombardment of helium ions, and not on electric fields, or adatom diffusion. Experiments employing a tungsten coated molybdenum sample indicate the presence of a strong mixing layer and strongly suggest that tungsten fuzz growth continues to occur from the bottom up even as the tendrils grow in size. Tests also show a similarity between different metals exposed to helium ion fluxes where the ratio of bubble diameter to tendril diameter is constant.

Tungsten coatings for application in WEST project

The WEST (Tungsten Environment in Steady-state Tokamak) project is aimed to implement and test the ITER W divertor technology in relevant plasma operation conditions. In addition to the ITER-like W monoblock divertor target, the WEST project involves W coated CuCrZr target and W coated Fine Grain Graphite (FGG) and CFC-N11 (Carbon Fiber Composite) tiles.Combined Magnetron Sputtering and Ion Implantation (CMSII) technology was developed and applied in the framework of ITER-like wall (ILW) project at JET for W coating of CFC Dunlop 780 DMS tiles Ruset et al. (2007) [1]. The coating thickness was in the range of 10–25 µm depending on the position of the tiles in the wall. The same technology was also used for W coating of more than 1300 FGG tiles for ASDEX Upgrade tokamak.In the present paper W coatings of 12.8 µm and 20 µm deposited on PFC (Plasma Facing Components) materials used in WEST project will be analyzed. As substrate materials CFC N11, FGG samples were used. These substrate materials have different properties compared to materials used for ILW project. GDOES (Glow Discharge Optical Emission Spectrometry) and XRD (X-Ray Diffraction) measurements were used for assessment of chemical composition and structure of the coatings.The behavior of the coatings to the High Heat Flux tests (HHFT) has been investigated by using an electron beam test facility. The testing program consisted of ∼300 pulses with duration of 420 s each. The surface temperature during the tests was 1200 °C whereas the power density was about 3.5 MW/m2.