Divertor Materials Research Articles

Research on the characteristics of the Helium plasma beam in HIT-PSI

As a high heat flux linear plasma device designed for studying divertor materials in future fusion reactors, HIT-PSI(Plasma Surface Interaction device at Harbin Institute of Technology) has been successfully constructed and has maintained stable operation since its completion. The characteristics of He plasma beams in HIT-PSI are investigated by emission spectroscopy and an infrared camera, with preliminary irradiation experiments conducted by bombarding tungsten with the beam. For relatively conservative discharge parameters, HIT-PSI achieved a steady-state heat flux capacity of ∼40 MW/m2 using infrared measurements, with the full width at half maximum (FWHM) of the heat flux beam reaching 4 mm. These characteristics make HIT-PSI an advanced platform for testing divertor materials and plasma-facing components, providing essential experimental supports for research and development of high-performance divertor materials.

Optimized collection optic design for divertor Thomson scattering diagnostics in KSTAR.

Last year, the KSTAR divertor material was changed from carbon to tungsten tiles. An optimized collection optic design for divertor Thomson scattering diagnostics in KSTAR was conducted for electron temperature (1-100eV) and electron density (1 × 1018-1 × 1019m-3) profile diagnostics. This diagnostic system will utilize a 1064nm Nd:YAG laser directly from the K-top port toward the beam dump located at K-bottom, while collecting scattered light from five spatial points in the divertor area via collection optics situated in the j-middle port. Given spatial limitations, the solid angle of measurement points is limited, and the collection optic design facing the tungsten divertor is susceptible to stray light. So, the design of the collection optic is important for divertor Thomson scattering diagnostics. For optimal performance, we performed two types of collection optic designs: Cooke-triplet and double-Gaussian. We present performance ray tracing analysis results for both designs and derive the optimal design.

Novel data interpretation method for DIII-D divertor retarding field energy analyzer with 3-D particle-in-cell simulations.

A novel data interpretation process that utilizes comprehensive particle-in-cell (PIC) simulations is developed for the new retarding field energy analyzer (RFEA) currently being constructed at DIII-D for the lower divertor using the Divertor Material Evaluation System. This probe is expected to survive a heat load of up to 100 MW/m2 for up to 5s and reliably measure the main ion temperature (Ti) on the divertor target ranging from 10 to 200 eV. These extreme conditions posed significant engineering limitations on the probe geometry, thus extensive validation work has been performed. The conventional fitting method for the RFEA I-V characteristics is based on a simplified 1-D model without considering the ion space charge inside the probe cavity and may not be sufficient for probes designed for the DIII-D divertor environment. In this article, a more realistic description of the particle propagation process within the RFEA cavity is achieved by including both 3-D geometric effects and ion space charge in the PIC simulations, and the capability to reconstruct the ion energy distribution functions is demonstrated with reasonable consistency.

Re-evaluation of shielding ability and induced activity of KSTAR with increased neutron yields

To determine the viability of the 2-fold neutron yield increase of the Korea Superconducting Tokamak Advanced Research (KSTAR) owing to the installation of a second neutral beam injection system to increase the plasma density and a divertor material change from graphite to tungsten to reduce the heat load, we performed a radiological evaluation and shielding analysis of an existing building structure and the KSTAR with an increased neutron yield per shot. According to the shielding analysis results, the weekly and annual doses outside the building did not exceed the criteria, and the fluxes and doses in the plasma experimental room were unaffected by the divertor material change. The maximum dose rate in the working area in the tokamak was 2.16E+03 μSv/h at the surface-plasma-facing components. However, the dose rate was reduced to the background level one month after a 1-shot operation of 300 s. Ar-41 was the major nuclide for the air activation calculations; however, the effluent level exceeded the derived air concentration after the 1-shot operation ended. The residual activity concentration of Ar-41 was reduced below the derived air concentration after 243 min. The annual operating condition of the KSTAR is suggested based on the evaluation results.

Tungsten sputtering coefficients by light impurities of plasma

Calculations of the tungsten sputtering coefficients (the divertor material in the ITER tokamak) by He, Be, N, O – impurity atoms in the plasma – were carried out at collision energy of 0.010–100 keV using the Monte–Carlo method. To calculate the trajectory of the incident particle, pair potentials obtained within the framework of density functional theory were used. These potentials were corrected for the parameters of the potential well obtained from spectroscopic measurements. The target consisted of tungsten randomly oriented crystals the size of one lattice constant. Next, the trajectories of the recoil particles were calculated using many-particle potentials calculated using density functional theory. Thermal vibrations of target atoms were taken into account. The vibration amplitude was taken to be 0.05 Å, which corresponded to room temperature. The strong dependence of the results on the shape of the surface potential barrier is shown and the results are presented for two limiting cases of the surface state: a flat surface, when a planar surface potential barrier is realized, and a surface consisting of cones, when a spherical potential barrier occurs. In the experiment, the surface has some roughness, which depends on the experimental conditions. It is shown that the experimental results lie between the limiting cases we considered. Information was obtained on the average energy of sputtered atoms and angular distributions, necessary for calculating the entry of impurities into the tokamak plasma.

Investigation of W-SiC compositionally graded films as a divertor material

W-SiC composite material is a promising plasma-facing material candidate alternative to pure W due to the low neutron activation, low impurity radiation, and low tritium diffusivity of SiC while leveraging the high erosion resistance of the W armor. Additionally, W and SiC have high thermomechanical compatibility given their similar thermal expansion rates. The present study addresses the synthesis and performance of compositionally graded W-SiC films fabricated by pulsed-DC magnetron sputtering. Compositional gradients were characterized using transmission electron microscopy (TEM) and energy-dispersive X-ray spectroscopy (EDS), and crystallographic information was obtained using electron diffraction and X-ray diffraction (XRD). Samples were exposed to L-mode deuterium plasma discharges in the DIII-D tokamak using the Divertor Material Evaluation System (DiMES). Post-mortem characterizations were performed using scanning electron microscopy (SEM) and XRD. Electron diffraction and XRD showed that the compositionally graded W-SiC films were composed of polycrystalline W and amorphous SiC with amorphous W+SiC interlayers. No macroscopic delamination or microstructural changes were observed under mild exposure conditions. This study serves as a preliminary examination of W-SiC compositionally graded composites as a potential candidate divertor material in future tokamak devices.

Model validation of tungsten erosion and redeposition properties using biased tungsten samples on DiMES

An experiment was performed in the DIII-D lower divertor to validate numerical SOL tungsten (W) impurity erosion and redeposition simulations against experimental data. The net and gross erosion of W were calculated as a function of the voltage (or bias) applied to the exposed material. Five samples were inserted into the DIII-D lower divertor using the Divertor Material Evaluation System (DiMES) manipulator and exposed to constant L-mode attached plasma conditions. Each sample was partially coated with W. During plasma shots, samples were biased with respect to the machine vessel ground, ranging from −60 V to 25 V. The ERO2.0 code was used to numerically simulate the experiment aiming to compare the numerical results with experimental measures. A good agreement is found between estimated and measured tungsten erosion at least for negative biases.

Measurements of heat flux components due to charged and non-charged particles in DIII-D divertor near and under detachment

Surfacing Eroding Thermocouples (SETC) have been used to provide high spatial and temporal resolution heat flux measurement in DIII-D. SETCs were first tested in the lower divertor of DIII-D using the Divertor Material Evaluation System (DiMES), then an array of SETCS was permanently installed in the upper SAS-VW divertor for studying the heat flux mitigation in closed divertor geometry. SETCs proved that inducing divertor detachment with various methods is an effective way to reduce the peak heat flux at the divertor targets. In the investigated discharges the heat flux is reduced to less than 30% under detachment compared to attached conditions. A new method of using the combination of recessed SETC and flush SETC was demonstrated to be able to measure the heat flux from volumetric radiation and the charge exchange neutrals. It was observed that the magnitude of the heat flux from radiation and charge exchange neutrals increased during the process of detachment. While 30% of the total incident heat flux is attributable to the volumetric radiation and the charge exchange neutrals in the fully detached divertor condition with B×∇B drift into the divertor, it could reach more than 60% with B×∇B drift away from the divertor.

In-situ coating of silicon-rich films on tokamak plasma-facing components with real-time Si material injection

Abstract Experiments have been conducted in the DIII-D tokamak to explore the in-situ growth of silicon-rich layers as a potential technique for real-time replenishment of surface coatings on plasma-facing components (PFCs) during steady-state long-pulse reactor operation. Silicon (Si) pellets of 1 mm diameter were injected into low- and high-confinement (L-mode and H-mode) plasma discharges with densities ranging from 3.9– 7.5 × 10 19 m−3 and input powers ranging from 5.5 to 9 MW. The small Si pellets were delivered with the impurity granule injector at frequencies ranging from 4 to 16 Hz corresponding to mass flow rates of 5–19 mg s−1 (1– 4.2 × 10 20 Si s−1) at cumulative amounts of up to 34 mg of Si per five-second discharge. Graphite samples were exposed to the scrape-off layer and private flux region plasmas through the divertor material evaluation system to evaluate the Si deposition on the divertor targets. The Si II emission at the sample correlates with silicon injection and suggests net surface Si-deposition in measurable amounts. Post-mortem analysis showed Si-rich coatings containing silicon oxides, of which SiO2 is the dominant component. No evidence of SiC was found, which is attributed to low divertor surface temperatures. The in-situ and ex-situ analysis found that Si-rich coatings of at least 0.4–1.2 nm thickness have been deposited at 0.4–0.7 nm s−1. The technique is estimated to coat a surface area of at least 0.94 m2 on the outer divertor. These results demonstrate the potential of using real-time material injection to form Si-enriched layers on divertor PFCs during reactor operation.

Unusual ablation behaviors of tungsten graded network reinforced copper composites synthesized by 3D printing and infiltration sintering at ultra-high temperatures

Tungsten network reinforced functionally graded copper composites were prepared using an integrated 3D printing and infiltration technology to improve high temperature ablation properties of nuclear fusion divertor materials. The composites' ablation behavior was systematically investigated using an argon arc welding torch. Results showed that mass loss of composites was increased firstly but then decreased with the increase of ablation time. During ablation, evaporation of the melted copper reduces the local temperature of tungsten skeleton, whose high melting point also restrict the further evaporation of copper. These mechanisms significantly improve the ablative resistance of CuW gradient composite. Tungsten twinning structures were identified to form during ultra-high temperature (at 6273.15 K) ablation process in a cold environment in air. The formation of distinctive W twinning structures and gradient distribution of copper and tungsten skeleton were identified as the key mechanisms for the significantly enhanced ablation performance of the graded composite.

SOLPS-ITER modeling of deuterium throughput impact on the ITER SOL plasma

The impact of the ionization source in the ITER far SOL associated with main chamber deuterium puffing on the overall SOL performance is studied numerically using SOLPS-ITER modeling. It is found out that this source can have significant influence on the retention of the seeded impurity radiator (Ne) in the divertor and on plasma temperature in far SOL. The former is responsible for typical separatrix concentration of seeded impurity compatible with the required power handling limits on the divertor targets, while the latter is responsible for sputtering of divertor material. Both factors indicate the beneficial effect of high fuel throughput on separatrix and pedestal plasma composition.

Physical and technical basis of Materials Plasma Exposure eXperiment from modeling and Proto-MPEX results *

The Materials Plasma Exposure eXperiment (MPEX) is a steady-state linear plasma device that will address plasma-material interaction (PMI) science and enable testing of fusion reactor-relevant divertor plasma-facing materials. The MPEX source concept consists of a helicon plasma source to generate the plasma, electron cyclotron heating (ECH) for electron heating, and ion cyclotron heating (ICH) for ion heating. The MPEX source plasma is then transported axially to the PMI material target region to test material samples in fusion reactor-relevant divertor conditions. This paper will summarize the physical and technical basis of MPEX. The paper will first define the MPEX parameters and scenarios at the target relevant to PMI science for various fusion reactor-relevant divertor conditions and show plasma transport modeling results to set the MPEX source parameters. Recent experimental and modeling results from Proto-MPEX, a short-pulse experiment to develop the plasma production, heating, and transport physics for MPEX, will be shown. From these results, it will be shown that MPEX can reach its desired scenarios. The MPEX physical and technical basis will also determine important functional requirements for magnetic field, radiofrequency (RF) power, RF frequency, and neutral pressure in the helicon, ECH, ICH, and PMI regions that are required to achieve the desired MPEX scenarios. The necessity for key in-vessel components such as skimmers, limiters, and microwave absorbers will also be highlighted.

Machine learned interatomic potential for dispersion strengthened plasma facing components.

Tungsten (W) is a material of choice for the divertor material due to its high melting temperature, thermal conductivity, and sputtering threshold. However, W has a very high brittle-to-ductile transition temperature, and at fusion reactor temperatures (≥1000K), it may undergo recrystallization and grain growth. Dispersion-strengthening W with zirconium carbide (ZrC) can improve ductility and limit grain growth, but much of the effects of the dispersoids on microstructural evolution and thermomechanical properties at high temperatures are still unknown. We present a machine learned Spectral Neighbor Analysis Potential for W-ZrC that can now be used to study these materials. In order to construct a potential suitable for large-scale atomistic simulations at fusion reactor temperatures, it is necessary to train on ab initio data generated for a diverse set of structures, chemical environments, and temperatures. Further accuracy and stability tests of the potential were achieved using objective functions for both material properties and high temperature stability. Validation of lattice parameters, surface energies, bulk moduli, and thermal expansion is confirmed on the optimized potential. Tensile tests of W/ZrC bicrystals show that although the W(110)-ZrC(111) C-terminated bicrystal has the highest ultimate tensile strength (UTS) at room temperature, observed strength decreases with increasing temperature. At 2500K, the terminating C layer diffuses into the W, resulting in a weaker W-Zr interface. Meanwhile, the W(110)-ZrC(111) Zr-terminated bicrystal has the highest UTS at 2500K.

MPEX High Heat Flux Plasma Dump Design, Manufacture, and Articles Test

The Material Plasma Exposure eXperiment (MPEX) device is a steady-state linear plasma device currently in the final design phase at the Oak Ridge National Laboratory. This device will reach ion fluences up to <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$10^{31}\,\,\text{m}^{-2}$ </tex-math></inline-formula> and will be used to expose neutron-irradiated materials to divertor-relevant plasmas and to study the effects of plasma-material interactions. These studies will elucidate the complex effects of plasmas with divertor candidate materials capable of withstanding high heat flux and high fluences for next-generation fusion devices. Bidirectional plasma will be generated using a high-power (200 kW) helicon source. Plasma will be confined by superconducting magnets. The last plasma-facing component on the upstream side of the MPEX device is the dump, which has been designed to intercept plasma and energetic particles. The dump will have a total heat load of 9.2 kW. A copper alloy (Glidcop AL-15) was selected for use in the water-cooled flange design because of its high thermal conductivity, its retaining strength at elevated temperatures, and its ability to be used in the high-temperature braze joints used in this application. Titanium-zirconium-molybdenum (TZM) tiles are brazed to the Glidcop AL-15 flange using high-temperature braze alloy. External water-cooling channels are used on the dump flange to prevent water leakage inside the vacuum space. This article discusses the details of the high heat flux dump design, including the computational fluid dynamics (CFD) and structural analyses performed to validate the design to meet the operational requirements of the MPEX device.

Liquid lithium as divertor material to mitigate severe damage of nearby components during plasma transients

The successful operation of thermonuclear fusion reactors such as ITER, DEMO, and future commercial plants is mainly determined by the optimum choice of materials for various components. The objective of this work is to accurately and comprehensively simulate the entire device in 3D to predict pros and cons of various materials, e.g., liquid lithium in comparison to tungsten and carbon to predict future ITER-like and DEMO divertor performances. We used our comprehensive HEIGHTS simulation package to investigate ITER-like components response during transient events in exact 3D geometry. Starting from the lost hot core plasma particles through SOL, deposition on the divertor surface, and the generation of secondary plasma of divertor materials. Our simulations predicted significant reduction in the heat loading and damage to the divertor nearby and internal components in the case when lithium is used on the divertor plates. While if tungsten or carbon are used on the divertor plate, significant melting areas and vaporization spots can occur (less for carbon) on the reflector, dome, and stainless steel tubes, and even parts of the first walls can melt due to the high radiation power of the secondary divertor plasma. Lithium photon radiation deposition into the divertor and nearby surfaces was decreased by two orders of magnitude compared to tungsten and by one order of magnitude compared to carbon. This analysis showed that using liquid lithium for ITER-like surfaces and future DEMO can lead to significant enhancement in components lifetime.

Investigation of dust formation during changes in the structural and surface properties of plasma-irradiated materials

• The contribution of surface defects (cracks, craters, bubbles) and structural changes (recrystallization) exposed by the plasma beam candidate materials to dust formation is studied. • The scanning electron microscopy (SEM), and elemental and X-ray diffraction (XRD) analysis were used to study the surface morphology, chemical composition, and crystal structure of plasma beam irradiated materials. • On the surface of tungsten target the well-developed micro-cracks are formed, and defects such as melting and flaking of the surface layer, drop-shaped structures formation are also observed. The increasing of the pulse dose led to an increase in the width and depth of these microcracks. • Tungsten target erosion (mass loss) occurs due to the ejection of particles from the tungsten surfaces, both as droplets and as solid dust during crack development. • The measured peak shift in carbon was 0.03 degrees after x5 plasma load. Thereby, carbon is generally more resistant to plasma beam impact than tungsten. This paper presents the results of studying the impact of a plasma beam during the interaction of divertor candidate materials in a coaxial plasma accelerator, where graphite and tungsten plates were used as candidate materials. The authors consider the contribution of surface defects (cracks, craters, bubbles) and structural changes (recrystallization) in candidate materials to dust formation under the action of a plasma beam. The other aspect studied in the paper is the erosion of copper substrates (nearby components) under the action of a hot target dust cloud with account for the kinetic influence of hot dust particles on nearby components. It should be noted that these particles are formed by the erosion of the material by the plasma beam. Scanning electron microscopy (SEM), elemental and X-ray diffraction analysis (XRD) were used to study the material surface morphology, chemical composition, and crystal structure. The authors also consider some undesirable effects revealed after irradiation of divertor candidate materials. The obtained experimental results will be useful for evaluating the prospects for further application of these materials in existing and future fusion reactors (e.g., in the international fusion reactor ITER).

Measurements of multiple heat flux components at the divertor target by using surface eroding thermocouples (invited).

The Surface Eroding Thermocouple (SETC) is a robust diagnostic utilized in DIII-D to provide fast, edge-localized modes (ELMs) resolved heat flux measurements, in particular in geometric regions that are too shadowed for traditional infrared thermography. In order to further investigate the power dissipation in the divertor region, a combination of flush-mounted and recessed SETCs was developed to assess the effect on surface heating from non-charged particles at the divertor target. Utilizing the Divertor Materials Evaluation System sample exposure platform, the first demonstration of the feasibility of using this new method to distinguish between the heat flux from charged particles and that from neutrals and radiative heating was achieved. This paper details the process of using the combination of flush SETCs and recessed SETCs to measure the multiple heat flux components at the divertor target and further discusses how to determine two important ratios, α (ratio of heat flux from charged particles deposit on recessed SETC to that deposit on flush SETC) and β (ratio of heat flux from non-charged particles deposit on recessed SETC to that deposit on flush SETC), in the estimation of the heat flux from non-charged particle sources. Using a time dependent ratio α, it was found that ∼50% of the total incident heat flux is attributable to the non-charged particles in the fully detached open divertor in DIII-D. Finally, the new application of similar SETC diagnostics in the Small Angle Slot divertor with a V-like configuration and partial tungsten coated surface (SAS-VW) is also introduced.

Comparison between SOLPS-ITER and B2.5-Eunomia for simulating Magnum-PSI

The interaction between plasma, neutrals and surfaces in fusion reactors is of immense importance for heat and particle control, especially for the next generation of devices. Heat loads of are expected for steady state operation at ITER and up to in slow transient situations. To study the complex physics appearing between the plasma and the divertor material, as well as techniques for heat flux mitigation, plasma linear devices are employed. Magnum-PSI, located at DIFFER, can reproduce the heat and particle loads expected at ITER. However, due to the complexity of the plasma–wall interaction, numerical models are required to better understand the experiments and to extrapolate the results to a tokamak divertor configuration. For tokamak geometries, SOLPS-ITER (formerly known as B2.5-Eirene) is employed to solve the plasma and neutral distribution in a coupled way. However, the utilization of this code for linear devices is not straightforward. Thus, a neutral module was developed with linear devices in mind, named Eunomia. Nevertheless, there is still a relevant interest in using SOLPS-ITER with linear devices, as it allows to easily transfer knowledge about relevant atomic and molecular processes close to the surface and the effect of different mitigation techniques. This work presents a systematic comparison between the two neutral modules, Eirene and Eunomia, in stand-alone and coupled runs. Special attention is paid to the implementation of plasma-neutral interactions, in which both codes diverge significantly. The sources of particles and energy that are used by B2.5 in a coupled run are analysed. Significant differences in the implementation of electron impact ionization, molecular assisted recombination and proton–molecule elastic collisions lead to disparate sources of particles and energy and, in some cases, differences in the distribution of neutrals achieved by each code. Moreover, a double counting in proton–atom collisions was identified in Eunomia as a result of this analysis, artificially increasing the plasma-neutral sink of energy. This would lead to different plasma evolutions in coupled runs. Nevertheless, additional-free parameters in both coupled code suites leave sufficient freedom to match experimental data. Additional data would be required to further constrain these parameters and the coupled solution.

Monitoring of two-dimensional tungsten concentration profiles on the HL-2A tokamak

In this article, we demonstrate the inference of two-dimensional tungsten concentration profiles in tokamak plasmas, using Gaussian process tomography applied to bolometry and assuming a specific model for the tungsten cooling factor. In ITER, tungsten has been selected for divertor material due to its low tritium retention and ability to handle large heat and particle flux loads. On the other hand, this will cause tungsten impurities to enter the bulk plasma through various plasma-wall interaction processes. Therefore, a detailed understanding of tungsten impurity transport and active control of the impurity transport in tokamaks is crucial. The computational complexity of the method described in their article compares favorably to a simple least-squares approach , n represents the number of pixels and m the number of measurement channels, hence bringing real-time tungsten profile monitoring within reach (<10 ms repetition time). The feasibility study of this method has been demonstrated here to a discharge in HL-2A for observing the entire process of tungsten impurities entering the bulk plasma from the scrape-off layer area, tungsten pump-out by edge-localized modes, as well as the formation of a poloidally asymmetric tungsten distribution.

Effect of divertor plasma-facing materials and drift on the power decay width in the scrape-off layer and divertor on EAST

The narrow power decay width λ q may lead to high heat flux density to the divertor target, which is one of the most severe challenges for future fusion devices. EAST has upgraded its lower graphite (C) divertor to a tungsten (W) one, and it is important to reveal the dependence of λ q on the divertor condition. In the present work, SOLPS-ITER is applied to estimate the effect of various conditions [C/W plasma facing material (PFM), with/without drift] on λ q of EAST in attached divertor regime. The simulation results confirm that the power decay width of the upstream λ q,SOL, divertor λ q,OT, and private-flux region (PFR) S OT all depend on the PFMs and drifts. However, the effect of PFM on λ q,SOL is slight, while it has a significant influence on λ q,OT and S OT. In the upstream, positive correlation between λ q,SOL and is obtained, and the simulation is in qualitative agreement with the two-point model. Both λ q,SOL and λ q,OT in without drift case are larger than that of with drift case. The regression of simulation results is obtained, showing a strong positive correlation between SOL collisionality , power radiation friction in divertor region f rad,OD and λ q,OT. Thus, the C PFM with intrinsic impurity has the largest λ q,OT. In the PFR, S OT with drift is smaller than that of without drift. Moreover, the combination of more D atoms accumulation and stronger E p × B drift lead to larger S OT in W PFM case than that of C PFM case.