Materials For Plasma Facing Components Research Articles

Interaction of iron melt with tungsten and WFe composite structure evolution

Development of new plasma facing components (PFC) draw attention of numerous scientific groups in fusion energy field. Tungsten as a main PFC material main have poor machinability, thus various designs were proposed to overcome this problem. However, such technologies as functional graded layers, sintering and additive technologies are limited in production size and have low cost-efficiency. This research considers the alternative approach of steel melt injection or melt infiltration of tungsten mesh. The results obtained establish the mechanism of phase evolution during interaction of solid W with liquid steel and iron. Energy dispersive spectroscopy, X-ray diffraction and Thermocalc calculations used to analyze the phase composition of tungsten wetted with various steels and pure iron. Results show that interaction rate significantly depends on melt temperature and overheating above its liquidus. Then the overheating exceeds 150 °C erosion of tungsten occur.

Ejection of molten tin in the presence of a hydrogen plasma

One of the significant obstacles left if the development of a fusion power plant is the development of Plasma Facing Components (PFCs) that can withstand the large heat and particle flux incident on the first wall and divertor region. As solid PFCs struggle with microstructure growth, sputtering and melting, liquid metals have been a popular potential replacement. Liquid metal's use as a PFC has been increasing due to its ability to self-repair damage as well as pump lost fuel and waste particles to create a low recycling edge. Currently, tin, lithium and lithium eutectics are the commonly considered liquid metals for use as a fusion PFC. It is therefore important to research the Plasma Material Interaction (PMI) between a hydrogen plasma and the liquid metal PFC candidates. This work, conducted at the Center for Plasma Material Interaction (CPMI), investigated how a molten tin surface reacts when exposed to a hydrogen plasma, building off observations by ASML of particles being ejected from a molten tin surface in the presence of a hydrogen plasma. With the ejection of tin affecting ASML's systems as well, since ejected tin can travel around their systems and cause contamination of equipment or wafers that can be destructive to the lithography process. So, for this work, molten tin was exposed to a hydrogen plasma, of varying electron densities and temperatures, and any ejected particles were collected on a witness plate to determine the particle sizes, angular distribution and mass flux. This work found the ejected tin particles range in size from 10′s ofnms to 10′s of microns and that the mass flux of tin from the molten surface is in the 10′s of mgm2*s, increasing with increased atomic hydrogen flux to the molten tin surface. Showing macroscopic amounts of molten tin are ejected in the presence of a hydrogen plasma, therefore showing tin may not be a suitable fusion PFC material.

Microstructure of additive manufactured materials for plasma-facing components of future fusion reactors

Microstructure of additive manufactured materials for plasma-facing components of future fusion reactors

Synergistical improvement in toughness and strength of multi-doped W-Hf-Y2O3 alloy via formation of complex oxide particles

As the primary candidate material for plasma facing components (PFCs) in fusion reactor, the serious brittleness of tungsten has become an urgent issue to be solved. In this work, a multi-doped W-Hf-Y2O3 alloy was prepared by high-energy ball milling and spark plasma sintering, aiming at improving its toughness through formation of complex oxide particles. Complex Y2Hf2O7 with pyrochlore structure were developed in the multi-doped WHY alloy due to dissolution and precipitation mechanism, which was confirmed by the combined characterization results of X-ray diffraction (XRD), electron probe microanalysis (EPMA) and transmission electron microscope (TEM). Complex oxide particles of Y2Hf2O7 with average size of 38.8 nm were densely distributed in the alloy. The formation of complex Y2Hf2O7 particles with nano-scale size not only consumed the impurity of oxygen, but also played a critical role in grain refinement because of Zener pinning effect. Improvement in mechanical properties of the multi-doped WHY alloy was then achieved, and the fracture toughness and compressive strength reached 19.7 MPa·m1/2 and 2223.4 MPa, respectively.

Divertor heat load estimates on NSTX and DIII-D using new and open-source 2D inversion analysis code

A thermography inversion algorithm has been developed in the open-source Python-based computer code, HYPERION, to calculate the heat flux incident on plasma-facing components (PFCs) in axisymmetric tokamaks. The chosen mesh size at the surface significantly affects the calculated transient heat flux results. The calculated transient heat flux will exceed the real value when the mesh size tends to zero but will underestimate the real value when the mesh size is large. A criterion for determining the appropriate mesh size for the transient heat flux calculation will be discussed. The numerical scheme for HYPERION uses a 2D fully implicit finite-difference approach, allowing temperature-dependent thermal properties of PFC materials. The inversion algorithm is benchmarked against established heat flux calculation codes, TACO and THEODOR, based on thermography data from NSTX and DIII-D respectively. The primary benefits of HYPERION compared to TACO and THEODOR are that it is open-source and it allows for the optimization of mesh thickness along the substrate. The algorithm also accounts for the thermal properties of thin surface layers that characteristically form on PFCs due to plasma-material interactions. The agreement between HYPERION and THEODOR is excellent, as the percent difference between the codes is ∼5% on average in the case of the DIII-D data for moderate to high heat flux. Verification tests with TACO show slightly higher average percent differences of 8% and 12%. In using HYPERION to study filaments in heat flux, the initial results indicate that small ELMs filaments significantly broaden the divertor heat flux, and decrease divertor peak flux. Compared to the inter-ELM, the small ELM filaments decrease the divertor peak surface temperature. With intermittent divertor filaments, the divertor heat flux width is comparable with that found in L-mode.

3D benchmark experiments of tritium in tungsten for tritium measurements by detecting β-ray induced X-rays

Tungsten is one of the most promising materials for plasma facing components (PFCs), and β-ray induced X-ray spectrometry (BIXS) is an important non-destructive method to obtain tritium depth profile and retention information for the development of PFCs, recovery of fuel and control of tritium safety in a fusion reactor. In this article, benchmark experiments to obtain tritium 3D profile in tungsten have been firstly performed using layer-by-layer chemical etching (LLCE) and imaging plate (IP) techniques. BIXS spectra have been detected at each layer. The average layer thickness of each erosion was controlled to be around 150 nm and 8 layers (the total thickness was 1.21 μm) were eroded, and the maximum uncertainty of tritium activity was 7.91 % in the LLCE experiments. Monte-Carlo simulations of BIXS spectra with Geant 4 code and four kinds of physical packages (Penelope, Livermore, EmStd. Opt1 and EmStd. Opt4) have been benchmarked against to the experimental results. Results show that all the four physical packages are not suitable to calculate the intensity of the BIXS spectra, and the current mode or cross sections should be improved to simulate the interaction between low electron/photon and tungsten accurately. However, the relative intensity of W-Mα peak and W-Lα can be simulated well using current physical packages such as EmStd. Opt4 and Livermore, which is a good indicator for tritium depth profile in tungsten.

Tungsten fiber-reinforced tungsten composites and their thermal stability

Tungsten will be used as armor material for plasma-facing components in future fusion reactors, but its propensity to embrittlement by microstructural restoration at high temperatures poses a challenge for its use. Tungsten fiber-reinforced tungsten composites (Wf/W) with drawn tungsten wires embedded in a polycrystalline tungsten matrix remedy the inherent brittleness of tungsten and achieve pseudo-ductile behavior via extrinsic toughening mechanisms. As plasma-facing materials experience high heat fluxes during operation, their thermal stability is important. In Wf/W composites, the restoration processes at high temperatures differ significantly between wires and matrix: initially recrystallization dominates in the wires, as they were plastically deformed during wire drawing, whereas abnormal grain growth occurs in the matrix. Growing grains may obstruct the interface between wire and matrix and deteriorate the otherwise improved fracture properties of Wf/W. An yttria interlayer is introduced to separate wire from matrix, to hinder an interplay between the restoration processes and to impede grains from the wire from growing into the matrix and vice versa. Cylindrical model systems containing a single wire in a chemically vapor-deposited matrix are investigated without any interlayer and with an yttria interlayer of either 1 μm or 3 μm thickness. Isothermal annealing at 1450°C for different times up to 2 weeks, followed by microstructural characterization by means of EBSD are carried out to characterize the microstructural evolution. The role of the interlayer on the microstructural evolution is elucidated to establish if decoupling of the restoration processes is actually achieved.

Annealing of hydrogen trap sites in displacement-damaged EUROFER

For future fusion power plants, the reduced activation ferritic/martensitic (RAFM) steel EUROFER is considered to be the primary choice as a structural material for plasma facing components (PFC) of the first wall, such as the blanket modules. Its exposure to high energy fusion neutrons will lead to creation of radiation-induced defects, which could act as trapping sites for hydrogen isotopes (HI) and consequently, strongly affect HI diffusion and retention. For predictive simulations of HI retention, the concentration of these trap sites needs to be known as a function of the temperature. In this work, EUROFER samples are irradiated by 20 MeV W-ions at 290 K to create displacement damage. Subsequently, they are exposed to deuterium (D) plasma at 370 K to decorate the created defects. Partial thermal desorption spectroscopy (TDS) is conducted in 50 K steps up to 620 K. D depth profiles are determined by nuclear reaction analysis (NRA) after each partial-TDS step. Following this analysis, the defects are again decorated with D and the D depth profiles are measured again by NRA to monitor the evolution of the concentration of trap sites. Most of the D is lost already at 470 K due to thermal de-trapping. Full recovery of the displacement damage is reached at 620 K.

Predictive atmospheric dispersion and deposition characteristics of activated tungsten dust

Tungsten is considered to be the most promising material for plasma-facing components of fusion reactors in addition to other industrial and military uses. Activated tungsten dust produced by plasma-wall interaction and released under accidental conditions would be the emerging environmental pollutant with both chemical and radiation toxicity. For the first time, atmospheric dispersion and deposition characteristics were investigated for tungsten dust with various particle size spectrum similar as generated in fusion reactors. Euler-Lagrange approach was adopted to describe parcels motion and deposition behaviors. Proper numbers of parcels were determined for the atmospheric dispersion and various parcel diameters scale. Near-surface concentration distribution and dry deposition velocity were predicted for tungsten dust with both single size dust and realistic particle size spectrum dust. Influences of chemical forms and release factors on atmospheric dispersion and deposition of tungsten dust were preliminary identified at last. It is indicated that 150 million parcels are needed for 10 μm dust under 5 km scale. Dry deposition velocity of tungsten dust is up to 11.3 cm s−1 and increase along with particle diameter because of its high density and gravitational deposition effect. The critical diameter judging whether gravitational deposition could appear for tungsten dust is 1 µm. Deposited particle size spectrum would be completely different along the downwind distance which would result in differences in subsequent behaviors in other environmental compartments.

Synergistic effects between potassium bubbles and high-energy-rate forging enabled ultrahigh strength and ductility of tungsten alloy

The intrinsic brittleness of tungsten (W) significantly restricts its application as the primary material for plasma-facing components. In this study, we propose a novel approach to improve both the strength and ductility of W-based alloys based on the synergistic effects between potassium (K) bubbles and high-energy-rate forging. K doping was introduced into W, followed by a two-step high-energy-rate forging (HERF) process. This leads to laminated matrix grains with refined sizes and deformation textures, as well as an increase in the density of low-angle grain boundaries and smaller size K bubbles along the grain boundaries. As a result, the prepared W alloys exhibit a substantially reduced ductile-to-brittle transition temperature within the range of room temperature to 50 °C, along with a notable ultimate tensile strength of up to 1393 MPa. Impressively, these alloys also demonstrate remarkable elongation at 200 °C, reaching a value as high as 33.12%. The remarkable mechanical strength and exceptional ductility exhibited by these alloys highlight their considerable potential for commercial application in the field of nuclear fusion engineering.

Effects of grain boundaries on the evolution of radiation-induced bubbles in polycrystalline tungsten: A phase-field simulation

Tungsten (W) is widely recognized as a promising material for plasma-facing components in nuclear fusion reactors, and the evolution of bubbles in polycrystalline tungsten significantly impacts its properties. This study developed a phase-field model considering elastic inhomogeneity to investigate the evolution of radiation-induced bubbles. The Gibbs free energy of helium bubbles was derived using the Van der Waals equation. An explicit nucleation algorithm accounting for the position of grain boundaries (GBs) was coupled to the model. Furthermore, the model also considered the inequality between the internal pressure of the bubble and the surface tension. The growth kinetics of a single bubble with elastic heterogeneity were examined, and based on these findings, the evolution of bubbles under irradiation was simulated, yielding results consistent with experimental observations. Furthermore, the influence of GBs on radiation-induced bubbles was investigated using the developed model. The simulation results demonstrated that both GBs, as preferential nucleation regions and as sinks for vacancy absorption, had an impact on bubble nucleation and growth. The superior radiation resistance observed in nanocrystalline tungsten was dominantly attributed to the absorption of vacancies by GBs.

Measuring subcooled boiling heat transfer on a CuCrZr surface to predict durability of a heat sink material for plasma-facing components

Energy conversion and heat transport are essential technologies for developing a commercial fusion reactor. In particular, plasma-facing components are a key to safety and effective operation. The heat transfer performance of plasma-facing components has been investigated. Boiling heat transfer is a promising method to economically cool a high heat flux in plasma-facing components because the latent heat of vaporization is high, resulting in a high heat transfer coefficient. As a result, less pumping power is needed than in single-phase flow cooling. In the event of a power loss, thermosiphons with pool boiling can be used for emergency cooling. In fusion reactors, copper-chromium-zirconium (CuCrZr) is expected to be used as a heat sink material for plasma-facing components. The surface properties of a heat sink affects boiling heat transfer. Cooling fusion reactors with boiling heat transfer requires investigating boiling characteristics on a CuCrZr surface, including critical heat flux (CHF). To characterize boiling heat transfer with CuCrZr, this study examined subcooled pool boiling with water at atmospheric pressure by varying the liquid subcooling temperature. The experimental results reveal that the CHF does not deteriorate with CuCrZr as it does with copper.

Characterization of boundary precipitation in a heavy ion irradiated tungsten heavy alloy under the simulated fusion environment

In the concerted effort to identify materials capable of surviving the adverse environment of a fusion reactor interior, tungsten heavy alloys have been put forth as candidates. Experimental trials and behavioral studies have yielded positive results for their adoption by taking advantage of the alloy's unique balance of high fracture toughness and high tungsten content; yet due to their relative novelty in the fusion community, there remains a lack of understanding on the response of these materials to the extended high temperature irradiation environment of the reactor interior. To alleviate this issue and provide the necessary data on the behavior of tungsten heavy alloys to the simulated fusion environment, a 90W-7Ni-3Fe alloy has been subjected to elevated temperature sequential Ni+ and He+ ion irradiations to mimic the expected displacement damage and He gas production expected after five years of service as materials for plasma facing components. Atomic-scale structural analyses and nanoscale chemical mapping have identified the formation of two distinct precipitation structures, a surface localized η-carbide and a hexagonal W2C type tungsten carbide, both of which appear to originate at the bi-phase interface between W and the ductile phase. This irradiation enhanced and induced precipitate formation respectively is anticipated to adversely affect these materials by selective embrittlement at bi-phase interfaces leading to a reduction in the material's overall fracture toughness during prolonged high temperature irradiation. It is asserted that any potential exposure to C during fusion reactor operational service should be minimized as to prevent the formation of these phases.

Computer simulation of the sputtering energy thresholds for some plasma-facing component materials irradiated with helium, deuterium, and tritium ions

Simulations for low-energy irradiation of nuclear Fusion plasma’s ion species on several plasma-facing materials were performed. This study is devoted to determining the minimum incident ion energy to cause sputtering or what is known as the “sputtering energy threshold” Eth. The simulations have been performed on various materials such as tungsten, graphite, and beryllium irradiated with various light ionic species such as helium, deuterium, and tritium. Two different Monte-Carlo-based simulation programs were used to perform this task. One of the latest versions of the state-of-the-art SDTrimSP simulation program was used in this study and complemented by the RDS-BASIC simulation program which was specifically designed to perform reasonably accurate and detailed calculations for the low–energy ions irradiation. The later simulation program also adopts a special and more realistic treatment accounting for the reduction in the atomic displacement threshold in the top surface monolayer. The simulation results were also compared with experimental data and theoretical data fittings.

Helium plasma immersion ion implantation studies of tungsten and tungsten heavy alloys for fusion plasma facing components

Plasma fusion devices will require plasma-facing components (PFCs) which can withstand the extreme environment at the edge of a hot fusion plasma. Despite the excellent properties of tungsten(W) as a hard refractory metal, adverse effects such as embrittlement, melting, and morphological evolution have been observed in W when it is bombarded by a high-fluence of low-energy ions including helium. This study investigates the effect of helium ion bombardment on pure tungsten and a tungsten heavy alloy (W-HA)(NAECOMET 1000). Pure tungsten and NAECOMET 1000 samples were implanted with 3 keV helium ions with fluences ranging from to , using Plasma Immersion Ion Implantation (PIII). After PIII treatment, samples were analysed using scanning electron microscopy (SEM) and atomic force microscopy (AFM), which revealed differences in surface morphology and topography. Although the melting and cracking of the Ni/Cu binder phase in the NAECOMET 1000 samples was seen under all implantation conditions, the Ni/Cu presence did somewhat slow the formation of W fuzz in comparison to pure tungsten. X-ray diffraction (XRD) studies showed peak shifts increasing with helium ion fluence for both the pure W and NAECOMET 1000 samples, as well as increase in mean crystallite size, confirming the distortion of the lattice. XPS compositional analysis showed a strong oxidation (>97) near the metal surface, after helium PIII treatment, for both pure W and NAECOMET 1000 W-HA samples. Some conclusions about the potential suitability of W-HA materials for fusion plasma PFCs are drawn.

Thermal expansion of plasma-exposed tungsten

We report results from a systematic analysis of thermal expansion of plasma-exposed tungsten based on molecular-dynamics simulations using models of tungsten with distributions of helium (He) bubbles in the tungsten matrix. We distinguish between two approaches of filling the bubbles with He, where the amount of He in the bubble can or cannot vary with temperature. In the former case, the thermal expansion coefficient decreases monotonically with the porosity and He content of the tungsten matrix, while in the latter case, the thermal expansivity increases monotonically with increasing porosity and He content. The latter condition, where the He content in the bubble is determined at the implantation temperature and remains constant with varying temperature in the tungsten matrix, is consistent with He species transport in tungsten used as a plasma-facing component (PFC) in nuclear fusion reactors and implies the development of biaxial compressive thermal strains in the PFC material that contribute to accelerating the growth of a nanostructure on PFC tungsten surfaces. Our analysis advances the fundamental understanding of thermal expansion in PFC tungsten and contributes to the development of a thermophysical property database for properly incorporating effects of realistic heat loads into modeling the dynamical response of PFC tungsten under fusion reactor operating conditions.

Effect of rare earth oxide doping on microstructure and mechanical property of Mo-30 W solid solution alloy

Oxide dispersion-strengthened (ODS) molybdenum‑tungsten (MoW) alloys are ideal materials for plasma-facing components of fusion reactors. In this paper, the effects of rare earth oxides (RExOy) Y2O3, CeO2 and La2O3 on the sintering behavior, microstructural and mechanical properties of MoW solid solution alloys were investigated. The initial alloy powders were produced by spray drying combined with the two-step reduction method (composed of carbothermal reduction and deep hydrogen reduction steps), and the solid solution alloy with good mechanical properties was prepared by subsequent sintering. The results showed that the doping of rare earth oxides suppressed the grain growth of MoW alloy, and thus improved the mechanical properties. Compared with CeO2 and Y2O3, the doping of La2O3 is more favorable to obtain dense and fine-grained MoW solid-solution alloys. The doping of La2O3 produced a larger Zener force and interfacial drag force, which effectively hindered the grain boundary migration and refined the grains. The microhardness and bending strength of Mo-30 W-1La2O3 alloy reached the maximum values of 557 HV and 771 MPa.

Study on the Heat Absorption Coefficient of Tungsten as the Plasma-Facing Material for the Upgraded EAST Lower Divertor

Divertor is one of the key plasma-facing components (PFCs) for Experimental Advanced Superconducting Tokamak (EAST) since they play important roles in resisting the energetic particle irradiation and huge heat deposition from the core plasma as well as efficaciously dispelling such energy load. Tungsten is used as the plasma-facing material for PFCs due to the excellent properties. During the thermal analysis and the high heat flux test of the PFCs, the heat absorption coefficient is an important parameter that directly affects the availability and reliability of the design structure. Previous studies have shown that there is a large deviation in the measured heat absorption coefficient of tungsten. Therefore, it is very important to accurately measure the heat absorption coefficient of tungsten. In this article, the heat absorption coefficient of tungsten was tested by the e-beam device of the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). In this test, 36 sets of inlet and outlet water temperature data were obtained by changing the water flow rate and the incident heat flux. The heat absorption coefficient of tungsten was calculated by the theoretical calculation using water calorimetry. The results indicate that the heat absorption coefficient of tungsten is stable at ~0.54, which can be used as a reference for the engineering design of PFCs of the upgraded EAST lower divertor. All test parameters for EAST will provide important reference and guidance to the next-generation fusion facility China Fusion Engineering Test Reactor (CFETR).

Improvement of lower hybrid current drive systems for high-power and long-pulse operation on EAST

Aiming at high-power and long-pulse operation up to 1000 s, some improvements have been made for both 2.45 GHz and 4.6 GHz lower hybrid (LH) systems during the recent 5 years. At first, the guard limiters of the LH antennas with graphite tiles were upgraded to tungsten, the most promising material for plasma facing components in nuclear fusion devices. These new guard limiters can operate at a peak power density of 12.9 MW/m2. Strong hot spots were usually observed on the old graphite limiters when 4.6 GHz system operated with power >2.0 MW [B. N. Wan et al., Nucl. Fusion57(2017) 102019], leading to a reduction of the maximum power capability. With the new limiters, 4.6 GHz LH system, the main current drive (CD) and electron heating tool for EAST, can be operated with power >2.5 MW routinely. Long-pulse operation up to 100 s with 4.6 GHz LH power of 2.4 MW was achieved in 2021 and the maximal temperature on the guard limiters measured by an infrared (IR) camera was about 540 °C, much below the permissible value of tungsten material (∼1200 °C). A discharge with a duration of 1056 s was achieved and the 4.6 GHz LH energy injected into the plasma was up to 1.05 GJ. Secondly, the fully-active-multijunction (FAM) launcher of 2.45 GHz system was upgraded to a passive-active-multijunction (PAM), for which the density of optimum coupling was relatively low (below the cut-off value). Good coupling with reflection coefficient ∼3% has been achieved with plasma-antenna distance up to 11 cm for the new PAM. Finally, in order to eliminate the effect of ion cyclotron range of frequencies (ICRF) wave on 4.6 GHz LH wave coupling, the location of the ICRF launcher was changed to a port that is located 157.5° toroidally from the 4.6 GHz LH system and is not magnetically connected.

Developing solid-surface plasma facing components for pilot plants and reactors with replenishable wall claddings and continuous surface conditioning. Part A: concepts and questions

It is estimated that pilot plants and reactors may experience rates of net erosion and deposition of solid plasma facing component (PFC) material of 103–105 kg yr−1. Even if the net erosion (wear) problem can be solved, the redeposition of so much material has the potential for major interference with operation, including disruptions due to so-called ‘unidentified flying objects (UFOs)’ and unsafe dust levels. The potential implications appear to be no less serious than for plasma contact with the divertor target: a dust explosion or a major UFO-disruption could be as damaging for an actively-cooled deuterium-tritium (DT) tokamak as target failure. It will therefore be necessary to manage material deposits to prevent their fouling operation. This situation appears to require a fundamental paradigm shift with regard to meeting the challenge of taming the plasma–material interface: it appears that any acceptable solid PFC material will in effect be flow-through, like liquid–metal PFCs, although at far lower mass flow rates. Solid PFC material will have to be treated as a consumable, like brake pads in cars. ITER will use high-Z (tungsten) armor on the divertor targets and low-Z (beryllium) on the main walls. The ARIES-AT reactor design calls for a similar arrangement, but with SiC cladding on the main walls. Non-metallic low-Z refractory materials such as ceramics (graphite, SiC, etc) used as in situ replenishable, relatively thin—of order mm—claddings on a substrate which is resistant to neutron damage could provide a potential solution for the main walls, while reducing the risk of degrading the confined plasma. Separately, wall conditioning has proven essential for achieving high performance. For DT devices, however, standard methods appear to be unworkable, but recently powder droppers injecting low-Z material ∼continuously into discharges have been quite effective and may be usable in DT devices as well. The resulting massive generation of low-Z debris, however, has the same potential to seriously disrupt operation as noted above. Powder droppers provide a unique opportunity to carry out controlled studies on the management of low-Z slag in all current tokamaks, independent of whether their protection tiles use low-Z or high-Z material.