Plasma Facing Components In Fusion Reactors Research Articles

Influence of temperature gradients on helium diffusion and clustering in tungsten: A molecular dynamics study

Tungsten and its alloys, used as plasma-facing materials in fusion reactors, endure high-flux, low-energy helium ion irradiation and temperature gradients induced by thermal load and shocks. This study utilizes molecular dynamics (MD) simulations to explore the diffusion and clustering behavior of helium (He) in tungsten (W) under temperature gradients. The migration behavior of isolated helium atoms, small helium clusters, and helium self-interstitial atom (He-SIA) clusters within tungsten is analyzed. It is revealed that both helium and He-SIA clusters tend to migrate towards higher temperature regions, exhibiting negative thermophoresis. The nucleation of helium clusters, the formation of Frenkel pairs, and the interactions between self-interstitial clusters and helium clusters are also investigated. The results indicate that helium concentration significantly impacts He nucleation behavior. Directional diffusion may lead to areas of high concentration over time, potentially promoting the formation of He clusters and bubbles. These insights enhance the understanding of tungsten's performance and durability as a plasma-facing component in fusion reactors.

Helium retention feature in the boron deposited layer on tungsten substrate by laser-induced breakdown spectroscopy and machine learning approach

ITER is designed for a burning plasma operation in which Tungsten (W) tiles are used as the first wall and diverter materials. Studying He dynamics in B-coated W wall is essential to understanding the effect of boronized plasma-facing components in fusion reactors, as these post-conditioning materials significantly influence helium retention and release. Plasma-wall interactions (PWI) are an important issue in ITER fusion reactors. PWI would lead to wall erosion and impurity redeposition. As a product of the D-T burning plasma, helium (He) ash would be retained or co-deposited on plasma-facing components (PFCs), affecting the stable operation of burning plasma. Laser-induced breakdown spectroscopy (LIBS) is proposed as a promising in-situ diagnostic approach for monitoring D/T and He retention and impurity deposition on PFCs of tokamak devices. In this study, the LIBS technique was used to investigate the helium retention feature in the boron (B) layer on tungsten substrate in 10−5 mbar. Five He-retention samples on the boron deposition layer on tungsten substrates were prepared by pulsed laser deposition (PLD) method at different ambient pressures in our lab. The investigations indicate that the atomic spectral line of helium (He-I 587.56 nm) was observed in the spectra of the first three laser shots. The depth profiles of He, B, and W in the boron-deposited layer on tungsten substrate were performed by LIBS to determine the co-deposition layer thickness. The concentration of He in the co-deposition layer samples measured by TDS is 7 × 1020 He/m2. The plasma parameters, such as plasma electron temperature and electron number density, were calculated to validate the local thermodynamic equilibrium. Machine learning (ML) algorithms are used to classify the co-deposits and substrates. The first three principal components (PC1, PC2 & PC3) of the unsupervised ML algorithm (PCA) give a classification accuracy of 91.7 %. The supervised ML algorithm neural network achieved training and testing accuracy of 100 % and 96.7 %, respectively.

A tungsten-wall sputtering model for the plasma start-up simulation in tokamaks

Abstract Tungsten (W) is the most probable material for the plasma-facing components of fusion reactors due to its excellent thermal and physical properties. A W-wall sputtering model has been established to simulate the start-up of a tokamak plasma using the 0D simulation code DYON. This model incorporates the revised Bohdansky formula to calculate the sputtering yield and a modified formula for calculating the energy impacting the walls. This formula integrates the temporal behavior of electron and ion temperatures at the plasma edge, which has been partially verified by the Thomson scattering diagnostic data. With the new model in place, predictive simulations were conducted for KSTAR’s Ohmic plasma under two W-wall scenarios: one with the entire wall surface covered by W and the other with 95% coverage of W and 5% coverage of carbon (C). The results indicate that the full-W wall may perform better from the perspective of start-up performance. The disparity can primarily be attributed to impurities generated through sputtering and recycling on the C wall. The validity of this model will be finally confirmed when the Thomson diagnostic system is able to precisely measure the edge electron temperature during plasma start-up.

Effect of W-Cu joining on D transport behavior in plasma-facing components for fusion reactors

In future fusion reactors, W-Cu joining technology will be utilized to fabricate plasma facing components (PFCs). In this study, hydrogen isotope gas-driven permeation (GDP) and thermal desorption spectrometry (TDS) are performed for W-Cu, W and Cu samples to understand the impact of W-Cu joining on fuel transport. GDP results reveal that the diffusion activation energy of the W-Cu sample is 0.289 eV, significantly smaller than that of the reference sample. Compared with the reference sample, the deuterium (D) retention in the W-Cu sample is doubled. High resolution transmission electron microscope (HRTEM) images indicate the presence of a mixed region of about 20 nm at the W-Cu interface, with a dislocation density in this region of 7.38 × 1016 m−2, three orders of magnitude higher than that in W and Cu. Simultaneously, energy dispersive spectrometer (EDS) data show that significant oxygen (O) impurities within the mixed region. Results from first principles simulations suggest that the presence of O at the interface can significantly alter the formation energy of H. The mixed layer not only serves as a diffusion barrier but also has the capacity to capture D. The presence of a high number of dislocations and O impurities in the mixed layer greatly impacts D transport in the W-Cu component.

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.

Processing and Properties of Tungsten-Steel Composites and FGMs Prepared by Spark Plasma Sintering

Tungsten is the prime candidate material for the plasma-facing components of fusion reactors. For the joining of tungsten armor to the cooling system or support structure, composites or graded interlayers can be used to reduce the stress concentration at the interface. These interlayers can be produced by several technologies. Among these, spark plasma sintering appears advantageous because of its ability to fabricate fully dense parts at lower temperatures and in a shorter time than traditional powder metallurgy techniques, thanks to the concurrent application of temperature, pressure, and electrical current. In this work, spark plasma sintering of tungsten-steel composites and functionally graded layers (FGMs) was investigated. As a first step, pure tungsten and steel powders of different sizes were sintered at a range of temperatures to find a suitable temperature window for fully dense compacts. Characterization of the sintered compacts included structure (by SEM); porosity (by the Archimedean method and image analysis); thermal diffusivity (by the flash method) and mechanical properties (microhardness and flexural strength). Compacts with practically full density and fine grains were obtained; while the temperature needed to achieve full sintering decreased with decreasing powder size (down to about 1500 °C for the 0.4 μm powder). For fully sintered compacts, the hardness and thermal diffusivity increased with decreasing powder size. Composites with selected tungsten/steel ratios were produced at several conditions and characterized. At temperatures of 1100 °C or above, intermetallic formation was observed in the composites; nevertheless, without a detrimental effect on the mechanical strength. Finally, the formation of graded layers and tungsten-steel joints in various configurations was demonstrated.

Chromium doped tungsten alloy for plasma-facing components of fusion reactors formed by compression plasma flows

Chromium doped tungsten (Cr-W) alloys are formed by compression plasma flows (CPFs) treatment at different absorbed energy densities (20 – 50 J/cm2), and are characterized by X-ray diffraction (XRD), scanning electron microscopy (SEM), energy dispersive X-ray microanalysis (EDX), and the Vickers hardness test. XRD analysis betokens that the dominant crystal orientation changes from (110) to (200) after CPFs treatment, and the intensity ratio (I200/I110) between the diffraction lines corresponding to (200) and (110) planes increases with an increase in the absorbed energy density. SEM and EDX analyses betoken spatial inhomogeneity at 20 J/cm2, improved uniformity at 30 J/cm2, and cracks at higher energy density. The hardness of the Cr-W alloy varies with absorbed energy density and has a maximum value of 379 HV at 20 J/cm2.

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.

Correlation study on tensile properties of Cu, CuCrZr and W by small punch test and uniaxial tensile test

Cu, CuCrZr alloy and W are three typical materials used in the plasma-facing components of fusion reactors, whose mechanical properties vary significantly. The small punch test (SPT) is used to evaluate the mechanical properties for materials in the nuclear industry due to its advantage of requiring minimal sample volumes. In this paper, the correlation between SPT and uniaxial tensile testing (UTT) is investigated for Cu, CuCrZr alloy and W. From the views of mechanical properties and fracture morphologies, materials demonstrate consistent behaviours between SPT and UTT, which proves that the SPT is an effective method to characterize the mechanical properties and fracture mechanism of Cu, CuCrZr alloy and W. Moreover, the finite element simulation results of SPT for five other copper alloys with different mechanical properties are added to establish the correlation equations of mechanical properties between SPT and UTT. One unified correlation equation between SPT and UTT can be established for the ductile materials, even if with largely different mechanical properties. However, the introduction of a brittle material W invalidates the correlation equation. This work provides a prospect of SPT to study the mechanical properties of materials used in plasma-facing components.

New Method for Calculation of Radiation Defect Dipole Tensor and Its Application to Di-Interstitials in Copper

The effect of external and internal elastic strain fields on the anisotropic diffusion of radiation defects (RDs) can be taken into account if one knows the dipole tensor of saddle-point configurations of the diffusing RDs. It is usually calculated by molecular statics, since the insufficient accuracy of the available experimental techniques makes determining it experimentally difficult. However, for an RD with multiple crystallographically non-equivalent metastable and saddle-point configurations (as in the case of di-interstitials), the problem becomes practically unsolvable due to its complexity. In this paper, we used a different approach to solving this problem. The molecular dynamics (MD) method was used to calculate the strain dependences of the RD diffusion tensor for various types of strain states. These dependences were used to determine the dipole tensor of the effective RD saddle-point configuration, which takes into account the contributions of all real saddle-point configurations. The proposed approach was used for studying the diffusion characteristics of RDs, such as di-interstitials in FCC copper (used in plasma-facing components of fusion reactors under development). The effect of the external elastic field on the MD-calculated normalized diffusion tensor (ratio of the diffusion tensor to a third of its trace) of di-interstitials was fully consistent with analytical predictions based on the kinetic theory, the parameters of which were the components of the dipole tensors, including the range of non-linear dependence of the diffusion tensor on strains. The results obtained allowed for one to simulate the anisotropic diffusion of di-interstitials in external and internal elastic fields, and to take into account the contribution of di-interstitials to the radiation deformation of crystals. This contribution can be significant, as MD data on the primary radiation damage in copper shows that ~20% of self-interstitial atoms produced by cascades of atomic collisions are combined into di-interstitials.

Temperature dependence of deuterium retention at displacement damage in tungsten

Tritium retention at displacement damage will increase the tritium inventory in tungsten plasma-facing components in fusion reactors. However, the quantity retained strongly depends on the temperature of the material during exposure to the plasma. Here we examine how the temperature dependence of the various physical processes affects hydrogen isotope retention at damage in tungsten. Processes considered here are diffusion, precipitation of molecular D2 in internal voids, binding at traps and annealing of displacement damage. A conceptual model is presented which incorporates the effect of precipitation on D permeation and trapping. The model predicts that D retention at damage should be highest in a temperature range from about 200 °C to 400 °C. D retention is lower at T < 200 °C due to slow permeation, and lower at T > 500 °C due to weak binding to the traps.

Studies on accelerated hydrogen isotopes permeation through plasma-facing components of fusion reactors

Experiments on deuterium (D) plasma-driven permeation through plasma-facing components (PFCs) mock-up have been performed in a linear plasma facility PREFACE. The effects of sample temperature, ion incident energy and surface conditions on the permeation behavior have been investigated. An 8 mm thick W armor tile with gaps is installed on membranes made by structural materials to simulate the monoblock type of PFC. CuCrZr alloy and CLF-1 steel membranes are tested at 500 K–750 K. Compared with gas-driven permeation cases, D permeation flux increases dramatically when the plasma is ignited, indicating that the energetic particles could pass through the gaps and result in enhanced permeation flux. The steady state D permeation flux increases with the increase of mock-up temperature, but the sample bias effects (i.e. incident energy effects) are not significant. PICS2 code calculations confirm the thickness of the sheath is comparable with the gap width under the plasma discharge conditions in PREFACE, suggesting that most of the incident ions into the gaps could not reach the bottom of the gaps.

The effects of dilute concentrations of substitutional Re or Os on the thermodynamics and kinetics of oxygen in tungsten

Tungsten is a candidate material for use in plasma-facing components of fusion reactors. Under operating conditions, transmutation occurs through exposure to neutrons giving rise to rhenium and osmium. Since oxidation is a major operational concern leading to material degradation, it is important to study this issue. In this work, the effects of dilute concentrations of rhenium and osmium on oxygen binding and diffusion in tungsten were investigated using multi-scale modeling. It was determined that the addition of dilute levels of Re or Os (less than 1 at%) led to complex, attractive and long range interactions with the interstitial oxygen. The cases where the interstitial oxygen occupied the 1st, 2nd, and 5th nearest neighboring T-sites were associated with the most favorable binding. Oxygen diffusivity in the material was found to decrease with increasing the substitutional defect concentrations. Finally, osmium was found to have a larger impact on the energy landscape.

Deuterium addition to liquid Li–Sn alloys: implications for plasma-facing applications

Liquid metals are being explored actively as candidates for plasma-facing components (PFCs) in fusion reactors. Recently, Li–Sn alloys have appeared as promising alternatives that could overcome some of the challenges faced by the well-studied liquid Li system, namely, a vapor pressure that limits the operating temperature and a high hydrogen isotope retention. However, only scarce data (experimental or theoretical) are available concerning the performance of Li–Sn alloys, specifically only for the compositions of Li30Sn70 and Li20Sn80, related to their bonding and retention of deuterium (D). Here, we present a comprehensive, first-principles molecular-dynamics study of static and dynamic properties of liquid Li30Sn70 at various D concentrations. We observe the formation of D2 gas bubbles for β in Li30Sn70Dβ greater than 22.5 along with Li segregation towards D2 bubbles. To understand the effect of Sn addition on D retention in Li–Sn alloys, we perform a thermodynamic evaluation of maximum D retention in Li-rich Li–Sn alloys. Overall, this work will provide useful data and guidance in the development of Li–Sn PFCs in fusion reactors.

Tensile and impact properties of tungsten-rhenium alloy for plasma-facing components in fusion reactor

Low temperature brittleness and high ductile-to-brittle transition temperature (DBTT) are potential drawbacks related to the mechanical properties of tungsten (W) for divertor application in fusion reactors. In the present study, to improve the mechanical properties of W, rhenium (Re) addition has been applied as a solid solution alloying method. In this paper, the effects of Re addition on the tensile and Charpy impact properties of W thick plate fabricated by powder metallurgy and hot rolling were investigated, which has enough volume as the plasma facing material and the microstructural uniformity for mass-production. At the temperature below 1000 °C, the W-3%Re showed 5–30% higher strength, 10–35% higher total elongation, 100 °C lower DBTT, and 16% higher upper shelf energy than those of the pure W.

A tritium permeation ‘short cut’ for plasma-facing components of fusion reactors

In ITER and other fusion reactors, plasma-facing components (PFCs) will be subjected to intense plasma/neutral fluxes and tritium (T) permeation issue will be raised. In the present work, first-of-a-kind experiments on hydrogen isotope (D) transport through ITER-like PFC mock-ups have been demonstrated in a linear plasma facility. 8 mm thick W tiles with 0.5 mm gaps are installed on structural materials and exposed to D plasmas. The angle between the gaps and the magnetic field line is set to be 5° to ensure the structural materials behind W are not exposed to charged D ions from plasma. When plasma is started up, the sample mock-ups are found to be penetrated by D within tens of seconds, suggesting that D can penetrate through heat sink of PFCs without passing through the thick W armor, i.e. the gaps act as a hydrogen isotope leakage ‘short cut’. To make a straightforward comparison of permeation flux for these materials, direct D plasma-driven permeation (PDP) experiments through W and structural materials are performed as well. The measured permeated D fluxes through structural materials are found to be orders of magnitude higher than that of W, indicating T transport from the gaps of PFCs may contribute a remarkable fraction of total leakage in reactors. In the latter part of this paper, a solution to address the permeation issue is proposed.

Effect of the magnetic field and current orientation on the splashing of liquid metal free surface of fusion reactor PFCs

In future tokamak nuclear fusion devices, the development of suitable plasma facing materials is considered to be one of the great challenges. Liquid metal, instead of conventional solid materials, has been proposed as a potential solution to plasma facing components (PFCs) for future nuclear fusion reactors, and it can solve the problems that most fusion reactors are facing. The splashing of liquid metal into the fusion plasma is a big problem because of the implementation of liquid metal PFCs in fusion reactors. An experiment involving the splashing of liquid metal has been developed under a strong magnetic field and electric field to imitate the interaction between plasma electric current and PFCs of liquid metal in fusion reactors. The current direction is perpendicular or parallel to the magnetic field direction. The influences of Kelvin–Helmholtz (KH) instability are experimentally observed and quantitatively analyzed on the magneto-electricity-driven liquid metal free surface. Processes of liquid metal splashing in the rectangular chamber are recorded by a high-speed camera. When the Lorentz force is upward and the external magnetic field and liquid thickness are constant, the liquid metal splashing from the liquid metal free surface when the external current exceeds the critical value. Critical current presents as almost inverse proportional functions with the magnetic field variation for different liquid thicknesses. In the experiment, the instability is characterized by the dimensionless numbers Bond number, Bd, determined from the importance of gravitational forces compared to surface tension forces, and the Hartmann number, Ha, determined from the intensities of the imposed magnetic field. Multiple linear regression models of the critical current density are summarized which indicates that the magnetic field and current’s influence determine the splashing critical point of liquid metal under the condition of multi-field coupling.

The response of ZrB2 to simulated plasma-facing material conditions of He irradiation at high temperature

Zirconium diboride (ZrB2) has many potentially beneficial properties for fusion plasma-facing component application, but almost no data exist on the response of ZrB2 to ion irradiation. In this work, ZrB2 samples were irradiated with 30 keV He+ to fluences of 8.4 × 1021 and 5.0 × 1022 He/m2 at temperatures of 920, 1020, and 1120 K in the Materials Irradiation Experiment (MITE-E) at the University of Wisconsin Inertial Electrostatic Confinement (UW-IEC) Laboratory to simulate some of the conditions of plasma-facing components in fusion reactors. The samples irradiated to the higher fluence developed surface morphology changes in the ion irradiated zone including rough, porous, and ripple structures. The ZrB2 had similar mass loss as W irradiated to similar conditions. Additionally, the ZrB2 samples did not exhibit flaking as did the SiC samples previously irradiated to similar conditions. This first look at ZrB2 behavior under ion irradiation is promising and justifies further testing of this emerging ultra-high temperature ceramic material for fusion applications.

Characterization of high flux magnetized helium plasma in SCU-PSI linear device

A high-flux linear plasma device in Sichuan University plasma-surface interaction (SCU-PSI) based on a cascaded arc source has been established to simulate the interactions between helium and hydrogen plasma with the plasma-facing components in fusion reactors. In this paper, the helium plasma has been characterized by a double-pin Langmuir probe. The results show that the stable helium plasma beam with a diameter of 26 mm was constrained very well at a magnetic field strength of 0.3 T. The core density and ion flux of helium plasma have a strong dependence on the applied current, magnetic field strength and gas flow rate. It could reach an electron density of 1.2 × 1019 m−3 and helium ion flux of 3.2 × 1022 m−2 s−1, with a gas flow rate of 4 standard liter per minute, magnetic field strength of 0.2 T and input power of 11 kW. With the addition of −80 V applied to the target to increase the helium ion energy and the exposure time of 2 h, the flat top temperature reached about 530 °C. The different sizes of nanostructured fuzz on irradiated tungsten and molybdenum samples surfaces under the bombardment of helium ions were observed by scanning electron microscopy. These results measured in the SCU-PSI linear device provide a reference for International Thermonuclear Experimental Reactor related PSI research.

Characterization of the liquid Li-solid Mo (1 1 0) interface from classical molecular dynamics for plasma-facing applications

An understanding of the wetting properties and a characterization of the interface between liquid lithium (Li) and solid molybdenum (Mo) are relevant to assessing the efficacy of Li as a plasma-facing component in fusion reactors. In this work, a new second-nearest neighbor modified embedded-atom method (2NN MEAM) force field is parameterized to describe the interactions between Li and Mo. The new force field reproduces several benchmark properties obtained from first-principles quantum mechanics simulations, including binding curves for Li at three different adsorption sites and the corresponding forces on Li atoms adsorbed on the Mo (1 1 0) surface. This force field is then used to study the wetting of liquid Li on the (1 1 0) surface of Mo and to examine the Li–Mo interface using molecular dynamics simulations. From droplet simulations, we find that liquid Li tends to completely wet the perfect Mo (1 1 0) surface, in contradiction with previous experimental measurements that found non-zero contact angles for liquid Li on a Mo substrate. However, these experiments were not carried out under ultra-high vacuum conditions or with a perfect (1 1 0) Mo surface, suggesting that the presence of impurities, such as oxygen, and surface structure play a crucial role in this wetting process. From thin-film simulations, it is observed that the first layer of Li on the Mo (1 1 0) surface has many solid-like properties such as a low mobility and a larger degree of ordering when compared to layers further away from the surface, even at temperatures well above the bulk melting temperature of Li. These findings are consistent with temperature-programmed desorption experiments.