Plasma Material Interactions Research Articles

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.

Fascinating physics at the edge of magnetic fusion devices

The physics of the processes at the edge of magnetic fusion devices is multifaceted and exhibits complex, nonlinear synergistic effects. Even though this region occupies only a small portion of the whole device, it plays a crucial role in overall plasma confinement, heat exhaust, and plasma–wall interactions. The latter affects not only the performance but also the lifetime of plasma-facing components and therefore remains an outstanding challenge for future fusion reactors. At the edge of fusion devices, researchers are dealing with phenomena including classical and anomalous plasma transport, atomic physics effects, and physics of plasma-facing material under strong irradiation by particle and energy fluxes. The diversity of edge physics makes it particularly attractive for young scientists. Working in this field, they can find endless possibilities to demonstrate their talents and creativity. This short review describes just some of the basic scrape-off layer and divertor plasma phenomena including divertor plasma detachment, intermittent bursts of anomalous cross-field plasma transport, plasma–material interactions, and dust in fusion plasmas, which are of particular interest in relation to fusion reactors.

HPIC2: A hardware-accelerated, hybrid particle-in-cell code for dynamic plasma-material interactions

The exascale era of high performance computing promises to bring the field of computational plasma physics ever closer to the goal of accurate multiscale modeling. Such computers will rely on hardware acceleration to offload work to dedicated components, notably general-purpose graphics processing units (GPUs). However, devices from different manufacturers require software to be written with different parallel programming models, greatly increasing the code maintenance burden of applications designed to perform on more than one such device. hPIC2 is a hybrid plasma simulation code developed with the Kokkos performance portability framework to target the architectures that will drive exascale computing for the foreseeable future. As a hybrid simulation code, hPIC2 investigates the simultaneous use of various plasma models on the same domain, at the same time. hPIC2 also optionally couples to RustBCA, which accurately models ion-material interactions using the binary collision approximation (BCA) method [1]. hPIC2 therefore achieves scalable performance on a variety of computing architectures when simulating complex and diverse plasmas, particularly near plasma-material interfaces.

Recent EAST Experimental Results and Systems Upgrade in Support of Long-Pulse Steady-State Plasma Operation

In recent years, EAST has achieved a 60-s scale steady-state plasma operation (SSO), which is electron dominant heated by the electron cyclotron wave and low hybrid wave. Substantial progresses have been made in the development of long-pulse operation, such as active control of radiative divertor, edge localized mode (ELM) suppression, and He plasma experiment. To further develop the steady-state operation scenario and study the related critical physical issues, the upgrade of EAST has been completed in 2020. The auxiliary heating systems were rearranged and upgraded with a total heating power capability of 33 MW. The 2.45-GHz lower hybrid current drive (LHCD) launcher has been updated from full active multijunction (FAM) to passive active multijunction (PAM) to avoid damage caused by plasma–material interaction. The injection power capability of electron cyclotron resonance heating (ECRH) system is upgraded to 1.4 MW. The upgraded ion cyclotron resonance frequency (ICRF) antenna injects low parallel wavenumber ( <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$k_{\vert \vert }$ </tex-math></inline-formula> ) spectra with good single pass absorption. The neutral beam injection (NBI) system is moved from F-port to D-port and is reoriented from counterclockwise to clockwise injection. The armor material of lower divertor has been upgraded from graphite to tungsten improving the steady-state heat handling capability from 2 to 10 MW/m2. Based on the upgraded pumping system and improved conductance, the effective pumping speed for <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$D_{2}$ </tex-math></inline-formula> of the EAST lower divertor increased twice. With all these upgrades, 100-s scale steady-state operation ( <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$T_{e0} &gt;10$ </tex-math></inline-formula> keV) and 1056-s high-performance operation with 1.73-GJ injection energy were achieved in 2021 campaigns. The experience in high-performance long-pulse operation obtained on EAST will lay a foundation for the ITER and Chinese Fusion Engineering Testing Reactor (CFETR) operation.

Simulation of plasma transport in the linear plasma device MPS-LD by SOLPS-ITER

• The direct imposing particle source is the suitable method to model the plasma source for the SOLPS-ITER modeling of plasma transport in linear device. • The MPS-LD can achieve plasma detachment by increasing the equivalent H + injection rate (source strength). • The combination of high heating power and short device length significantly raise the plasma parameters near the target. Linear plasma devices are the ideal devices to simulate the divertor conditions of tokamak experimentally. The Multiple Plasma Simulation Linear Device (MPS-LD) is under construction at Dalian University of Technology with a focus of studying the plasma-material interactions (PMIs) and edge plasma transport. Before the experiment, numerical modeling plays a crucial role in predicting the main plasma parameters and discharge performance. To this end, simulations are carried out by using scrape-off layer plasma simulation code SOLPS-ITER for the designing studies of MPS-LD in the present work. The effects of particle source simulation method, heating power, and device chamber length on the plasma are investigated systemically. The simulation demonstrates that the direct imposing particle source in the source region is suitable for handling the plasma source. The radiation region, recombination front region and recombination region are formed during plasma transport. Enhancing the plasma source strength can promote the achievement of plasma detachment. The helicon source with a 6kW radio frequency power source can only obtain the maximum electron density of 5 × 10 18 m -3 and electron temperature of 2 eV in the vicinity of the target. Raising the heating power and shortening the distance from the plasma source to the target can significantly increase the plasma parameters at the target.

Analysis of Azimuthal Doppler Shift of Anisotropically Absorbed Laguerre-Gaussian Beam Propagating in Transverse Flow

The particle flux onto a material is an important parameter in the study of the plasma-material interaction. With conventional Doppler spectroscopy, it is difficult to measure the flow velocity perpendicular to the material because only the velocity component projected on the wave number vector can be measured. To overcome the limitation, we are developing a transverse flow measurement method using the azimuthal Doppler shift of a Laguerre-Gaussian (LG) beam absorption spectroscopy. In this paper, the feasibility of this spectroscopy method has been examined by numerical analysis. The LG beam is anisotropically absorbed in the transverse flow due to the azimuthal Doppler shift. Since the anisotropic LG beam rotates with propagation, the spatial structure of resonance absorption in plasma and the intensity structure of the LG beam that has propagated through the plasma are inevitably different in the LG beam absorption spectroscopy. It was shown that the deviation from the original azimuthal Doppler shift is reduced to several percent at the position where the intensity distribution of the LG beam reaches its maximum value. Therefore, the transverse flow can be measured with sufficient accuracy by properly selecting the position on the beam cross-section used to evaluate the azimuthal Doppler shift.

First lithium experiments in HIDRA and evidence of helium retention during quasi-steady-state stellarator plasma operations

Recent experiments in Hybrid Illinois Device for Research and Applications (HIDRA) have had operational discharges between t discharge = 60 and 1000 s using electron cyclotron resonant heating (ECRH) of the plasma. This means that quasi-steady-state plasma discharges reach conditions to study long-pulse plasma material interactions (PMIs). The newly commissioned HIDRA-Material Analysis Test-stand PMI diagnostic is used to place a drop of lithium onto a heated tungsten surface, transfer the sample in-vacuo and expose it in a helium plasma. Helium is of interest as there is an open question to whether lithium will be able to remove helium ash in real fusion devices. The introduction of the W-Li sample in HIDRA resulted in evaporation of lithium into the helium plasma during a 600 s pulse and caused a reduction of over 90% in neutral pressure during the discharge. It was also observed that the plasma density and temperature increased by over 2.5 times. Using spectroscopy and a helium collisional radiative model, the peak temperature and density of the helium plasma can be monitored during the discharge. During lithium evaporation, as significant lithium ionization occurs, there is a 85% drop in the HIDRA vessel neutral pressure, despite a constant flow rate of He gas. This reduction in neutral pressure is supported by spectroscopy data with corresponding reductions in He I line intensities (587 nm, 667 nm, 706 nm, and 728 nm), as well as those of other impurities. At one point in the discharge a lithium plasma is created, as indicated by an increase in Li+ emission and a complete reduction in He+ emission, but the electron density jumps from n e = 3 × 1018 m−3 to over n e = 8 × 1018 m−3 while the core temperature stays relatively constant between T e = 16 eV and 20 eV. Once lithium has completely evaporated from the sample and the majority of the ionized lithium has diffused from the plasma to the vessel walls, pressure and spectroscopy data paired with He collisional radiative model calculations shows a re-establishment of a helium plasma in a low recycling regime. In this regime, the density drops down to n e = 2 × 1018 m−3 and the electron temperature increases from T e = 20 eV to over T e = 50 eV indicating an increase in helium heating efficiency. This is also indicated by the He+ emission re-establishing and having a higher intensity. In this paper we show the results from the first lithium campaign in HIDRA. In the presence of lithium, and in particular when lithium ions are present, the helium disappears from the plasma via an as of yet unknown complex relationship that needs to be further studied. The most likely explanation is that the lithium ions are distributed around the vessel and able to trap helium to the surface turning HIDRA into a large gettering surface. These results have potential implications on future plasma facing component design using liquid lithium for impurity and recycling control using limiters and divertors.

HIDRA-MAT liquid metal droplet injector for liquid metal applications in HIDRA

A liquid metal droplet injector was explicitly designed for the Hybrid Illinois Device for Research and Applications Material Analysis Test-stand (HIDRA-MAT) at the University of Illinois Urbana-Champaign to prepare liquid metal plasma-facing components (PFCs) for plasma exposure. The design goals were to create a compact, reliable, and robust design that could apply liquid metal droplets in-vacuo to a variety of samples. The injector was designed for liquid metal use, and results discussed pertain to lithium. The injector can produce lithium droplets of consistent size by utilizing programmable piston movement. Droplet formation data is presented for four different piston step sizes (0.125 mm, 0.25 mm, 0.5 mm, and 1 mm) at three different nozzle temperatures (185 °C, 230 °C, and 270 °C). Droplets are formed on the injector nozzle tip and a linear shift mechanism vertically translates the droplet to the substrate for application. This design prevents the droplet from inadvertently detaching off the nozzle and falling to the sample. Application of the liquid metal droplet ensures placement consistency on the substrate and helps avoid damage to components in HIDRA-MAT that should not interact with liquid metals. A description of the liquid lithium droplet creation and application results is given and provides additional insight into the cooling and oxidation of lithium droplets under vacuum conditions. A second nozzle was fabricated and demonstrated repeatability in droplet diameter creation having a variance of ±0.14 mm for droplets created from the same number of piston steps. The end-use for the injector on HIDRA-MAT is to apply liquid metals to PFCs and expose them to HIDRA's plasma with subsequent intershot in-vacuo surface analysis in HIDRA-MAT for plasma-material interaction studies.

Microstructure deformation and near-pore environment of resolidified tungsten in high heat flux conditions

Tungsten is a candidate divertor material for many tokamak reactors and has potential to be durable to high heat flux conditions. However, plasma-materials interaction and edge localized modes (1–10 GW/m2) can cause significant hardening and melting damage. Tungsten in a high heat flux (46.3 GW/m2) and helium plasma environment was investigated. We studied possible residual helium and microstructure deformation in resolidified tungsten. Following this, a 1–30 keV Ga focused ion beam was used for TEM sample milling. In as-received tungsten, dark spots of possible lattice strain and defects were in the grains. After high heat exposure under plasma pressure, intergranular features, dark spots in TEM, diffraction patterns, and elongated FIB induced pore-artifact structures (second phase) formed. Helium appears to be absent from the resolidified tungsten matrix due to erosive melting damage and high temperature conditions. A simple four-step microstructure deformation process from the elastic bulk side to the near-pore environment is proposed. The presence of defects in TEM images, grain size reduction, and microstructure deformation affected hardening. The near-pore environment likely experienced grain refinement, and further disordered nanophases are possible under severe deformation conditions.

Determination of the characteristic magnetic pre-sheath length at divertor surfaces using micro-engineered targets on DiMES at DIII-D

The magnetic pre-sheath (MPS) length, L MPS, is a critical parameter to define the sheath potential, which controls the ion trajectory of low-Z species (D, T, He, and C), as well as the prompt re-deposition of high-Z species. To determine L MPS, we fabricated micro-trenches (30 × 30 × 4 μm) via focused ion beam milling on a silicon surface and exposed them to L-mode deuterium plasmas in DIII-D via the divertor material evaluation system (DiMES) removable sample exposure probe. The areal distribution of impurity depositions, mainly consisting of carbon, was measured by energy-dispersive x-ray spectroscopy (EDS) to reveal the deuterium ion shadowing effect on the trench floors. The carbon deposition profiles showed that the erosion was maximized for the azimuthal direction of φ = −40° (referenced to the toroidal magnetic field direction) as well as the polar angle of θ = 80°. A Monte Carlo equation-of-motion (EOM) model, based on a collisionless MPS, was used to calculate the azimuthal and polar deuterium ion angle distributions (IADs) at the surface for a range of L MPS = k × ρ i, where ρ i is the ion gyro radius and k = 0.5–4. Then, gross erosion profiles were calculated by a Monte Carlo micro-patterning and roughness (MPR) code for ion sputtering using as input the calculated azimuthal and polar IADs for each value of k. Good agreement with the experimental C deposition profiles was obtained for the case k = 2.5–3.5. This result is consistent with a previous kinetic modeling prediction of k ∼ 3, as well as previous analytical investigations that predicted the L MPS to be several ion gyro radii. A validation of theoretical sheath models supports its applicability to ITER and pilot plant divertors to successfully predict plasma–materials interactions.

NSTX-U theory, modeling and analysis results

The mission of the low aspect ratio spherical tokamak NSTX-U is to advance the physics basis and technical solutions required for optimizing the configuration of next-step steady-state tokamak fusion devices. NSTX-U will ultimately operate at up to 2 MA of plasma current and 1 T toroidal field on axis for 5 s, and has available up to 15 MW of neutral beam injection power at different tangency radii and 6 MW of high harmonic fast wave heating. With these capabilities NSTX-U will develop the physics understanding and control tools to ramp-up and sustain high performance fully non-inductive plasmas with large bootstrap fraction and enhanced confinement enabled via the low aspect ratio, high beta configuration. With its unique capabilities, NSTX-U research also supports ITER and other critical fusion development needs. Super-Alfvénic ions in beam-heated NSTX-U plasmas access energetic particle (EP) parameter space that is relevant for both α-heated conventional and low aspect ratio burning plasmas. NSTX-U can also generate very large target heat fluxes to test conventional and innovative plasma exhaust and plasma facing component solutions. This paper summarizes recent analysis, theory and modelling progress to advance the tokamak physics basis in the areas of macrostability and 3D fields, EP stability and fast ion transport, thermal transport and pedestal structure, boundary and plasma material interaction, RF heating, scenario optimization and real-time control.

Simulations of fusion edge plasmas by linear plasma devices: physics and plasma–material interactions

Simulations of fusion edge plasmas by linear plasma devices: physics and plasma–material interactions

Modelling of edge plasma dynamics with active wall boundary conditions

AbstractA self‐consistent 2D model is presented for transport in boundary plasma and plasma‐facing material walls. Plasma dynamics in the domain is represented by a 2D collisional plasma fluid model in the edge‐plasma code UEDGE (Rognlien et al., J. Nucl. Mater. 196–198 (1992) 347), and transport of hydrogen and heat in the wall is represented by a system of reaction–diffusion equations in the 1D wall code FACE (Smirnov et al., Fusion Sci. Technol. 71 (2017) 75). To account for variation of parameters along the wall, in the coupled model multiple instances of the FACE code run in parallel. The coupled model provides a tool for investigating a range of dynamic plasma–material interactions phenomena in 2D. For demonstration of its capability, one application of particular interest is the role of active wall in tokamak strike point sweeping proposed for mitigation of divertor heat loads. In the present study, the coupled calculations are applied to investigation of the impact of heat and hydrogen transport in the material wall on the divertor plasma and target heat load during sweeping of the target strike point for parameters of a high‐power tokamak.

NSTX-U theory, modeling and analysis results

Abstract The mission of the low aspect ratio spherical tokamak NSTX-U is to advance the physics basis and technical solutions required for optimizing the configuration of next-step steady-state tokamak fusion devices. NSTX-U will ultimately operate at up to 2 MA of plasma current and 1 T toroidal field on axis for 5 seconds, and has available up to 15 MW of Neutral Beam Injection (NBI) power at different tangency radii and 6 MW of High Harmonic Fast Wave (HHFW) heating. With these capabilities NSTX-U will develop the physics understanding and control tools to ramp-up and sustain high performance fully non-inductive plasmas with large bootstrap fraction and enhanced confinement enabled via the low aspect ratio, high beta configuration. With its unique capabilities, NSTX-U research also supports ITER and other critical fusion development needs. Super-Alfvénic ions in beam-heated NSTX-U plasmas access energetic particle parameter space that is relevant for both -heated conventional and low aspect ratio burning plasmas. NSTX-U can also generate very large target heat fluxes to test conventional and innovative plasma exhaust and plasma facing component (PFC) solutions. This paper summarizes recent analysis, theory and modelling progress to advance the tokamak physics basis in the areas of macrostability and 3D fields, energetic particle stability and fast ion transport, thermal transport and pedestal structure, boundary and plasma material interaction, RF heating, scenario optimization and real-time control.

Deciphering the influence of high-flux helium plasma exposure on tungsten nanostructure growth using real-time, in situ spectroscopic ellipsometry

In situ analysis of surfaces during high-flux plasma exposure represents a long-standing challenge in the study of plasma–material interactions. While post-mortem microscopy can provide a detailed picture of structural and compositional changes, in situ techniques can capture the dynamic evolution of the surface. In this study, we demonstrate how spectroscopic ellipsometry can be applied to the real-time characterization of W nanostructure (also known as “fuzz”) growth during exposure to low temperature, high-flux He plasmas. Strikingly, over a wide range of sample temperatures and helium fluences, the measured ellipsometric parameters (Ψ,Δ) collapse onto a single curve that can be directly correlated with surface morphologies characterized by ex situ helium ion microscopy. The initial variation in the (Ψ,Δ) parameters appears to be governed by small changes in surface roughness (&amp;lt;50 nm) produced by helium bubble nucleation and growth, followed by the emergence of 50 nm diameter W tendrils. This basic behavior appears to be reproducible over a wide parameter space, indicating that the spectroscopic ellipsometry may be of general practical use as a diagnostic to study surface morphologies produced by high-flux He implantation in refractory metals. An advantage of the methods outlined here is that they are applicable at low incident ion energies, even below the sputtering threshold. As an example of this application, we apply in situ ellipsometry to examine how W fuzz growth is affected both by varying ion energy and the temperature of the surface.

Development of plasma sources and diagnostics for the simulation of fusion edge plasmas

Although the research on divertors and scrape-off layers (SOLs) has been not as focused on as the recent success of the Korean fusion program, a few linear plasma devices have been developed for simulating divertor and SOL plasmas: (1) diversified plasma simulator (DiPS), a versatile linear machine, has been developed for simulations of divertor and space plasmas with various electric probes, such as single, triple, and Mach Probes and gridded energy analyzer. DiPS consists of two major parts: a divertor plasma simulator with a \(\hbox {LaB}_6\) DC plasma source and a space plasma simulator with a helicon RF plasma source, (2) divertor plasma simulator-1 (DiPS-1) is a part of DiPS with only a \(\hbox {LaB}_6\) cathode, where a low-power laser-induced fluorescence (LIF) is added and more electric probe diagnostics are augmented; it is dedicated only for fusion edge and divertor plasmas, (3) Divertor Plasma Simulator-2 (DiPS-2) has been modified from the DiPS-1 by adding a magnetic nozzle with a limiter structure and by removing the helicon source and space chamber. DiPS-2 is a linear plasma device with a 4-inch \(\hbox {LaB}_6\) cathode, the same as DiPS-1, and it is focused on the development of various diagnostics, such as those used for LIF and laser Thomson scattering (LTS) along with various electric probes, on the divertor and scrape-off plasmas and on the plasma-material interaction (PMI) research, such as that of tungsten and graphite as plasma-facing components (PFCs), (4) A Multi-Purpose Plasma (\(\hbox {MP}^2\)) device is a renovation of the Hanbit mirror device [Kwon et al., Nucl. Fusion 43, 686 (2003)] with the installation of two plasma sources: \(\hbox {LaB}_6\) (DC) and helicon (RF) plasma sources. A honeycomb-like large-area \(\hbox {LaB}_6\) (HLA-\(\hbox {LaB}_6\)) cathode has been developed for the divertor plasma simulation to improve the resistance against the thermal shock fragility for large (8-inch) and high density plasma generation, (5) DiPS-2 has been augmented by adding another cylindrical device, called the Dust interaction with Surfaces Chamber (DiSC) for the generation and diagnostics of dusts. This combined system (DiPS-2+DiSC) has added two more diagnostics: Laser Photo-Detachment (LPD) for dust density and laser Mie Scattering (LMS) for dust size. Moreover, dusts or negative ions have been analyzed by using electric probes and capacitive diagram gauges in Transport and Removal of Dusts (TReD) device.

Generation of ion cyclotron instability by parametric coupling of gyrating ion beam with lower hybrid wave in a complex plasma

In a two-ion species tokamak plasma, with the help of gyrating ion beam was to investigate the impact of dust charge fluctuations on the parametric up-conversion of lower hybrid wave (LHW) into an upper side band wave and an ion cyclotron wave. On the existing electrons, through an upper side band wave and lower hybrid pump, a ponderomotive force has been exerted which in turn drives the ion cyclotron decay mode. The growth rate of instability is calculated by considering typical existing D-T (Deuterium-Tritium) dusty plasma parameters and observed that with an increment in amplitude of pump waves, wave number of ion cyclotron mode, number density of dust grains and relative density of dust grains, an enhanced growth rate is reported. However, a reduction in the growth rate has been detected with enhancing dust grain size and electron cyclotron frequency. The findings of this work can play an imperative role in magnetized plasma sources used in plasma material interaction / surface processing. The tokamak operation is severely affected by the various sizes and density of dust grains, which in turn, considerably affect the choice of material requisite in tokamak.

Electronegative warm plasma and material interaction under secondary electron emission

Plasma-material interaction is examined in an electronegative warm plasma under secondary electron emission from the wall. The sheath formed during the interaction is evaluated and its thickness is compared for three different plasmas (C60plasma, Oxygenplasma and CF4,plasma) and the sheath thickness is found to be maximum in C60plasma and minimum in Oxygenplasma. The sheath potential, floating potential, and sheath thickness decrease when the secondary electrons are present in larger number. The density of secondary electrons increases with higher number of the negative ions at the sheath edge as well as inside the sheath for various values of density of multi-charged negative ions.

The role of fluorination during the physicochemical erosion of yttria in fluorine-based etching plasmas

A physicochemical mechanism acting between the reactive plasma and the material surface controls the erosion of polycrystalline ceramics in fluorine containing etching plasmas. In this study, a Y2O3/YOF composite was exposed to a fluorine etching plasma. Relocalization enables the direct correlation of crystalline orientation with material response. Our study reveals an orientation dependent surface fluorination of Y2O3, which controls the etching resistance and morphology formation. Orientations near the low index planes (001), (010) and (100) exhibit the lowest stability due to a homogeneous surface reaction. The presented results help to extend the mechanistic understanding of the plasma-material interaction of Y2O3.

Integrated model predictions on the impact of substrate damage on gas dynamics during ITER burning-plasma operations

Divertor design and choice of plasma-facing materials (PFM) will be essential to the success of next-generation fusion reactors as they operate under more powerful scenarios. Understanding and controlling interactions between the plasma and PFM is essential to making these choices. Within these plasma–material interactions and especially in tungsten (W), the interplay between the most abundant plasma species (hydrogen isotopes and helium, He) with the wall material alters fuel retention. However, this interplay is yet to be sufficiently understood to confidently project fuel retention levels to future fusion devices. The paper presents a series of integrated simulations of fusion plasmas and their interaction with tungsten. Specifically, this study assesses the impact of He plasma pre-exposure on hydrogenic species retention during 100 s of burning plasma operations (BPO) in ITER. Multiple pre-exposure scenarios are considered, including sub-surface damage resulting from exposures in the linear device PISCES and from early ITER He-operation. The predictions from these consecutive He-BPO exposures show that fuel content and spatial distribution in the material are largely determined by the He-induced damage, as manifest in: (i) changes in surface temperature expected during BPO have little effect on fuel retention in the presence of He-induced damage; (ii) gas content stabilizes quickly in substrates pre-exposed in PISCES, at levels set by the concentration of pre-existing vacancies, while it continues to increase in substrates initially pristine or pre-exposed to ITER He plasmas; (iii) the presence of He and He–V clusters in the near-surface region locally increases hydrogenic retention, but decreases its permeation; this results in hydrogenic species that remain closer to the surface in pre-damaged substrates, while the bulk content is higher for initially pristine cases. In summary, the interaction and binding of D and T with the pre-existing He–V clusters modifies retention and permeation of hydrogen species during ITER BPO.