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Plasma Facing Component Characterization and Correlation With Plasma Conditions in Lithium Tokamak Experiment-β
Lithium coatings in the Lithium Tokamak eXperiment (LTX) led to flat temperature profiles. The flat temperature profiles were observed along with a hot, low density edge, implying a broad, collisionless scrape-off layer (SOL). Additionally, in vacuo X-ray photoelectron spectroscopy (XPS) measurements established that lithium coatings evaporatively deposited onto high-Z plasma facing components (PFCs) became oxidized while retaining the ability to achieve good plasma performance long after lithium was applied to the PFCs. Longstanding theory predicted flat temperature profiles with low recycling walls, which was presumed to be due to hydrogen binding with elemental lithium to form lithium hydride. The presence of oxidized lithium, however, raised questions regarding the exact mechanism of hydrogen retention in LTX. To investigate these questions, the upgraded facility LTX- β includes a new sample exposure probe (SEP) for more detailed in vacuo analysis of PFC samples. The SEP is equipped with a vacuum suitcase capable of transporting samples representative of the LTX- β outer midplane PFCs to a stand-alone XPS system while maintaining pressures lower than the LTX- β base vacuum to limit the contamination between sample exposure and analysis. The low-energy resolution XPS system used in past experiments could only enable the determination of elemental percentages on the PFC sample surfaces. Because the new XPS system has higher energy resolution, it is more direct to assign chemical compounds to the measured binding energies. This capability has been confirmed by comparing XPS data from PFC test samples with measurements using a commercial high-resolution XPS system. Quartz crystal microbalances (QCMs) were used to quantify the thickness of the deposited lithium on the LTX- β PFCs. This article describes the application of the SEP to characterize the PFC surfaces using XPS and their relationship to plasma conditions.
Design and Analysis of an Integrated Additively Manufactured Test Article for Plasma-Facing Components
Current plasma-facing components (PFCs) used in helium-cooled divertor modules are complex structures with tungsten tile, steel sleeve components, and cartridges, all assembled in a helium-cooled multiple jet (HEMJ) structure. The goal of this project is to simplify the complex PFC design using additive manufacturing techniques to create a single integrated tungsten test article. Apart from the flexibility this opens up in exploring a wide array of geometries for the article, having a single integrated article significantly reduces the number of joints and parts in the article, thus reducing chances of leaks. A process called electron beam melting has shown to produce very high-density samples and unique geometries, enabling HEMJ or similar designs. To validate and optimize this novel design, the model underwent a series of computational fluid dynamics and finite element analysis simulations to replicate steady-state heat flux in the divertors. The simulations presented in this study consider a steady-state base heat flux of 5 MW/m 2 , with water serving as the coolant. Future research will explore the use of helium as a coolant, simulate edge-localized-mode conditions, and include experimental validation. Since 3D-printed tungsten is anisotropic, the build direction versus build plane of the article are taken into consideration for the test article strength. Because of the high operating temperatures and low ductility of tungsten, thermal creep and brittle fracture are important failure mechanisms to consider. In conclusion, the cap is evaluated with various flow velocities and nozzle diameters, and an optimal design choice is made for which this cap will survive the divertor conditions with a conservative safety margin.
Lithium sputtering from lithium-coated plasma facing components in the NSTX divertor
Not Available
The Path towards Plasma Facing Components: A Review of State-of-the-art in W-Based Refractory High-Entropy Alloys
Developing advanced materials for plasma-facing components (PFCs) in fusion reactors is a crucial aspect for achieving sustained energy production. Tungsten (W) - based refractory high-entropy alloys (RHEAs) have emerged as promising candidates due to their superior radiation tolerance and high-temperature strength. This review paper will focus on recent advancements in W-based RHEA research, particularly emphasizing the key role of modelling using machine learning (ML) in the stage of discovery by predicting properties for each composition and expediting the identification of optimal RHEA compositions with desired properties. Additionally, the application of additive manufacturing (AM) techniques for fabricating W-based RHEAs is explored, highlighting their advantages for rapid prototyping and multi-compositional sample production in a high throughput manner. The review critically evaluates the current understanding of mechanical properties relevant to PFC applications, including high-temperature strength and ductility. Furthermore, the radiation tolerance of W-based RHEAs under irradiated conditions is discussed. Finally, the validity of current AM-manufactured W-based RHEAs as PFC materials is assessed, and key challenges and opportunities for future research are identified. This review aims to provide a comprehensive overview of W-based RHEAs for fusion applications and their potential to guide the development and validation of advanced refractory high entropy alloys.
Initial Assessment of 3D-Printed Molybdenum Lattices as a Substrate for Liquid Metal Plasma-Facing Components
Many liquid metal plasma-facing component concepts for fusion reactors require a capillary porous material to wick the liquid metal to the surface of the component. In this work, samples of 3D-printed molybdenum lattices designed to have pore sizes ranging from 100 to 1000 µm were fabricated by two different suppliers. The dimensions of the lattice structures were characterized using an optical microscope, revealing that the lattice walls were almost systematically thicker than they were designed to be and in turn the pores are all smaller than they were designed to be. In many cases the pores were completely clogged. Several spectroscopic techniques were used to characterize the chemical composition of the materials, collectively indicating the presence of a thick layer of oxygen-rich material extending more than 1 µm below the surface. Compression testing revealed significant asymmetry in the mechanical behavior depending on which direction the force was applied relative to the print direction. Lastly, lithium wetting tests of several lattice samples were performed, and a sample with ~250-µm pores was found to soak up lithium quickly after initial wetting of the surface occurred. The implications of these results for future liquid metal plasma-facing component design are discussed.
Differentiating the role of lithium and oxygen in retaining deuterium on lithiated graphite plasma-facing components
Laboratory experiments have been used to investigate the fundamental interactions responsible for deuterium retention in lithiated graphite. Oxygen was found to be present and play a key role in experiments that simulated NSTX lithium conditioning, where the atomic surface concentration can increase to >40% when deuterium retention chemistry is observed. Quantum-classical molecular dynamic simulations elucidated this oxygen-deuterium effect and showed that oxygen retains significantly more deuterium than lithium in a simulated matrix with 20% lithium, 20% oxygen, and 60% carbon. Simulations further show that deuterium retention is even higher when lithium is removed from the matrix. Experiments artificially increased the oxygen content in graphite to ∼16% and then bombarded with deuterium. X-ray photoelectron spectroscopy showed depletion of the oxygen and no enhanced deuterium retention, thus demonstrating that lithium is essential in retaining the oxygen that thereby retains deuterium.
Plasma exposures of a high-conductivity graphitic foam for plasma facing components
The plasma-surface interactions from samples of high-conductivity graphitic foam biased to 120 V and placed in 6–8 eV deuterium plasmas with densities as high as 10 19 m -3 were investigated at the PSI-2 linear plasma device in Jülich. Graphitic foam-plasma interactions were also studied at the Wendelstein 7-X (W7-X) stellarator in Greifswald by exposure to hydrogen and helium plasmas using the Jülich multi-purpose manipulator. The purpose was to explore the possibility of using the material in a plasma facing component, and initial results were encouraging. In W7-X, no measurable erosion or cracking was observed. The PSI-2 samples received a deuterium fluence of 5 × 10 25 m -2 resulting in an average erosion of 43 µm or about 5 mg per sample. Residual gas analysis (RGA) data were acquired to monitor sample outgassing. Laser-induced Breakdown Spectroscopy (LIBS) was used to measure deuterium retention in the porous foam. After exposure, the surfaces were characterized with scanning electron microscopy, energy dispersive x-ray analysis and time-of-flight secondary ion mass spectrometry (ToF-SIMS). The graphitic foam has a thermal conductivity as high as 287 W/mK and is considered as a replacement to more exotic carbon fiber composites such as SepCarb NB31 or isotropic graphites like ATJ that are no longer produced but used in present-day tokamak experiments. Actively cooled monoblocks were made from the foam and underwent extensive materials characterization including infrared response studies at Oak Ridge National Laboratory. This material is under consideration for the proposed actively-cooled W7-X divertor scraper element.
Oxidation of lithium plasma facing components and its effect on plasma performance in the lithium tokamak experiment-β
The characteristics of lithium-coated plasma-facing components (PFCs) have been correlated with plasma performance on the lithium tokamak experiment-β (LTX-β). Previous experiments on LTX showed that the application of lithium to PFCs was needed to achieve higher performance discharges with flat electron temperature profiles and high edge temperatures. Samples that match the LTX-β PFCs were exposed to plasmas before and after PFCs were coated with lithium and transferred under vacuum to a surface analysis station. Measurements using x-ray photoelectron spectroscopy (XPS) revealed that the primary surface constituent was lithium oxide. Earlier XPS analysis of lithium-coated PFCs on LTX was only able to show the presence of surface oxygen. The new XPS data from LTX-β have sufficient resolution to clearly identify lithium compounds for the first time, and enable them to be correlated with how lithium-coated PFCs can reduce impurities and retain hydrogen to reduce recycling.
FENIX: Towards a Fully Integrated Multiphysics Framework for Plasma Facing Component Modeling
Computational tools have a crucial role to play in accelerating the deployment of fusion as a clean, reliable, abundant, and sustainable energy source. Multiphysics, high-fidelity simulation capabilities can help model, study, and predict intricate interactions between materials performance, plasma exposure, neutron irradiation, and engineering processes. As such, they can assist in the resolution of scientific and engineering challenges underpinning design, construction, and commission of fusion power plants. To address these needs, ongoing efforts are leveraging the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework and delivering new computational tools for the fusion community. These tools inherit crucial attributes from MOOSE. They are open-source, modular, integrated with nuclear industry-standard software quality assurance processes, and enable multiphysics, multi-fidelity, fully integrated, zero- to three-dimensional, and massively parallel simulations. After a short overview of these capabilities, we will present the development of Fusion ENergy Integrated multiphys-X (FENIX), a MOOSE-based application designed to enable plasma facing component design and performance evaluation. Throughout their lifetime, plasma facing components are exposed to extreme thermal loads, repeated thermal shocks, and irradiation by plasma ions, neutral particles, and high-energy neutrons. Consequently, designing a plasma facing component with acceptable lifetime degradation is extremely challenging. FENIX aims to model the multiphysics environment in which plasma facing components evolve to accelerate their design studies. To that end, FENIX couples existing MOOSE capabilities such as heat transfer, thermomechanics, and thermal hydraulics, with tritium transport via the MOOSE-based Tritium Migration Analysis Program, Version 8 (TMAP8), with neutronics via the MOOSE-based high-fidelity neutron-photon transport and fluid dynamics code Cardinal, and finally with Particle-in-Cell plasma simulation capabilities being developed in this project. In this study, we present the current FENIX capabilities and preliminary results of its application to model the Tritium Plasma Experiment set up at Idaho National Laboratory.
Magnetohydrodynamics in free surface liquid metal flow relevant to plasma-facing components
Abstract While flowing Liquid Metal (LM) Plasma-Facing Components (PFCs) represent a potentially transformative technology to enable long-pulse operation with high-power exhaust for fusion reactors, Magnetohydrodynamic (MHD) drag in the conducting LM will reduce the flow speed. Experiments have been completed in the linear open-channel LMX-U device [Hvasta et al 2018 Nucl. Fusion 58 01602] for validation of MHD drag calculations with either insulating or conducting walls, with codes similar to those used to design flowing LM PFCs for a Fusion Nuclear Science Facility [Kessel et al 2019 Fusion Sci. Technol . 75 886]. We observe that the average channel flow speed decreased with the use of conducting walls and the strength of the applied transverse magnetic field. The MHD drag from the retarding Lorentz force resulted in an increase of the LM depth in the channel that ‘piled up’ near the inlet, but not the outlet. As reproduced by OpenFOAM and ANSYS CFX calculations, the magnitude and characteristics of the pileup in the flow direction increased with the applied traverse magnetic field by up to 120%, as compared to the case without an applied magnetic field, corresponding to an average velocity reduction of ∼45%. Particle tracking measurements confirmed a predicted shear in the flow speed, with the surface velocity increasing by 300%, despite the 45% drop in the average bulk speed. The MHD effect makes the bulk flow laminarized but keeps surface waves aligned along the magnetic field lines due to the anisotropy of MHD drag. The 3D fringe field and high surface velocity generate ripples around the outlet region. It was also confirmed that the MHD drag strongly depends on the conductivity of the channel walls, magnetic field, and volumetric flow rate, in agreement with the simulations and a developed analytical model. These validated models are now available to begin to determine the conditions under which the ideal LM channel design of a constant flow speed and fluid depth could be attained.
Design and Analysis of Liquid Lithium Plasma Facing Components
In this study, liquid metal (LM) on plasma-facing components (PFCs) creates a renewable protective cover enhancing power exhaust and protecting the solid substrate. Additionally, energy confinement improvement through particle pumping can be achieved in the case of liquid lithium (LL). A new variant of PFCs was recently introduced at Princeton Plasma Physics Laboratory (PPPL) where a porous wall is used to stabilize the LM surface, while using magnetohydrodynamic (MHD) drive to push the LM flow inside the component. This arrangement allows efficient heat exhaust, and enhanced control of the LM surface temperature, leading to spatial control of evaporation and condensation of LL on the plasma interface. This feature is particularly attractive when vapor shielding is introduced to allow heat flux redistribution. This system has the advantage that as the heat flux increases, the evaporation rate will increase, while decreasing heat flux will decrease the evaporation rate of lithium, ideally creating a feedback effect which could self-regulate the amount of lithium evaporated. Analytical and numerical models for LL PFCs were developed in PPPL. To calculate the target temperature distribution for the case of evaporation from the divertor, we apply an iterative process allowing two-way coupling between the fluid-kinetic analysis of plasma using SOLPS-ITER code and the flow and heat transfer analysis of the PFC using an analytical model. In the last stage, the results are validated using computational fluid dynamics (CFD) analysis with customized version of the CFX code from ANSYS. CFX was modified at PPPL to allow MHD analysis at the high magnetic field typical for fusion applications. Results of the analysis for NSTX-U tokamak conditions will be presented.
Thermal stability of oxidized ultrathin Li films on TZM for plasma facing components
For improved understanding of the behavior of Li plasma facing components (PFCs) in the presence of oxygen impurities, we report on the thermal stability of ultrathin (up to 1.0 nm) LiO x films on a titanium-zirconium-molybdenum (TZM) alloy substrate over the temperature range of 310–1400 K. LiO x films were prepared by either post-oxidation of Li films or Li films deposited on an oxygen-precovered TZM surface. Thermal stability of these films was studied using a combination of temperature programmed desorption (TPD), X-ray photoelectron spectroscopy (XPS), and low-energy ion scattering (LEIS). For post-oxidized Li films, TPD showed that no Li desorption occurred until temperatures above 620 K, and then Li desorbed from the surface via at least three desorption peaks. The O 1s XPS spectra on the post-oxidized films at 310 K indicate the formation of lithium oxide (Li2O) and peroxide/hydroxide (Li 2 O 2 /LiOH). The peroxide/hydroxide converts to oxide after heating to 680 K with no desorption of Li or O 2 , and then this film decomposes to liberate Li into the gas phase while leaving oxygen at the TZM surface. Heating the LiO x films to 1070 K led to a solid-state reaction with surface Mo atoms of the substrate to form a condensed binary lithium molybdenum oxide (LixMoOy) phase, which upon further heating decomposed to liberate gas phase binary lithium molybdenum oxide species. Li deposition on an oxygen-precovered TZM surface formed a complex, highly stable, oxygen rich, Li-O-Mo interfacial oxide. Li desorption from multilayer Li films on oxygen-precovered TZM surfaces occurred in a metallic Li multilayer peak and three other oxide-derived peaks. These results support the improved thermal stability of oxidized lithium as compared to metallic Li and have implications for Mo impurity transport into the plasma at high temperatures.
Plasma–wall interaction studies within the EUROfusion consortium: progress on plasma-facing components development and qualification
The provision of a particle and power exhaust solution which is compatible with first-wall components and edge-plasma conditions is a key area of present-day fusion research and mandatory for a successful operation of ITER and DEMO. The work package plasma-facing components (WP PFC) within the European fusion programme complements with laboratory experiments, i.e. in linear plasma devices, electron and ion beam loading facilities, the studies performed in toroidally confined magnetic devices, such as JET, ASDEX Upgrade, WEST etc. The connection of both groups is done via common physics and engineering studies, including the qualification and specification of plasma-facing components, and by modelling codes that simulate edge-plasma conditions and the plasma–material interaction as well as the study of fundamental processes. WP PFC addresses these critical points in order to ensure reliable and efficient use of conventional, solid PFCs in ITER (Be and W) and DEMO (W and steel) with respect to heat-load capabilities (transient and steady-state heat and particle loads), lifetime estimates (erosion, material mixing and surface morphology), and safety aspects (fuel retention, fuel removal, material migration and dust formation) particularly for quasi-steady-state conditions. Alternative scenarios and concepts (liquid Sn or Li as PFCs) for DEMO are developed and tested in the event that the conventional solution turns out to not be functional. Here, we present an overview of the activities with an emphasis on a few key results: (i) the observed synergistic effects in particle and heat loading of ITER-grade W with the available set of exposition devices on material properties such as roughness, ductility and microstructure; (ii) the progress in understanding of fuel retention, diffusion and outgassing in different W-based materials, including the impact of damage and impurities like N; and (iii), the preferential sputtering of Fe in EUROFER steel providing an in situ W surface and a potential first-wall solution for DEMO.
Upgrades toward high-heat flux, liquid lithium plasma-facing components in the NSTX-U
Liquid metal plasma-facing components (PFCs) provide numerous potential advantages over solid-material components. One critique of the approach is the relatively less developed technologies associated with deploying these components in a fusion plasma-experiment. Exploration of the temperature limits of liquid lithium PFCs in a tokamak divertor and the corresponding consequences on core operation are a high priority informing the possibilities for future liquid lithium PFCs. An all-metal NSTX-U is envisioned to make direct comparison between all high-Z wall operation and liquid lithium PFCs in a single device. By executing the all-metal upgrades incrementally, scientific productivity will be maintained while enabling physics and engineering-science studies to further develop the solid- and liquid-metal components. Six major elements of a flowing liquid-metal divertor system are described and a three-step program for implementing this system is laid out. The upgrade steps involve the first high-Z divertor target upgrade in NSTX-U, pre-filled liquid metal targets and finally, an integrated, flowing liquid metal divertor target. As a result, two example issues are described where the engineering and physics experiments are shown to be closely related in examining the prospects for future liquid metal PFCs.
Digital Holography for in Situ Real-Time Measurement of Plasma-Facing-Component Erosion
In situ, real time measurement of net plasma-facing-component (PFC) erosion/deposition in a real plasma device is challenging due to the need for good spatial and temporal resolution, sufficient sensitivity, and immunity to fringe-jump errors. Design of a high-sensitivity, potentially high-speed, dual-wavelength CO2 laser digital holography system (nominally immune to fringe jumps) for PFC erosion measurement is discussed.
Plasma–surface interaction in the stellarator W7-X: conclusions drawn from operation with graphite plasma-facing components
W7-X completed its plasma operation in hydrogen with island divertor and inertially cooled test divertor unit (TDU) made of graphite. A substantial set of plasma-facing components (PFCs), including in particular marker target elements, were extracted from the W7-X vessel and analysed post-mortem. The analysis provided key information about underlying plasma–surface interactions (PSI) processes, namely erosion, transport, and deposition as well as fuel retention in the graphite components. The net carbon (C) erosion and deposition distribution on the horizontal target (HT) and vertical target (VT) plates were quantified and related to the plasma time in standard divertor configuration with edge transform ι = 5/5, the dominant magnetic configuration of the two operational phases (OP) with TDU. The operation resulted in integrated high net C erosion rate of 2.8 mg s -1 in OP1.2B over 4809 plasma seconds. Boronisations reduced the net erosion on the HT by about a factor 5.4 with respect to OP1.2A owing to the suppression of oxygen (O). In the case of the VT, high peak net C erosion of 11μm at the strike line was measured during OP1.2B which converts to 2.5 nm s -1 or 1.4 mg s -1 when related to the exposed area of the target plate and the operational time in standard divertor configuration. PSI modelling with ERO2.0 and WallDYN-3D is applied in an interpretative manner and reproduces the net C erosion and deposition pattern at the target plates determined by different post-mortem analysis techniques. This includes also the 13 C tracer deposition from the last experiment of OP1.2B with local 13 CH 4 injection through a magnetic island in one half module. The experimental findings are used to predict the C erosion, transport, and deposition in the next campaigns aiming in long-pulse operation up to 1800 s and utilising the actively cooled carbon-fibre composite (CFC) divertor currently being installed. The CFC divertor has the same geometrical design as the TDU and extrapolation depends mainly on the applied plasma boundary. Extrapolation from campaign averaged information obtained in OP1.2B reveals a net erosion of 7.6 g per 1800 s for a typical W7-X attached divertor plasma in hydrogen.
Critical Exploration of Liquid Metal Plasma-Facing Components in a Fusion Nuclear Science Facility
Liquid metal (LM) plasma-facing components (PFCs) may provide a resolution to the challenging fusion environment, particularly the first wall and divertor surfaces. Transforming these concepts into viable technologies will require considerable research and development. Here, with the fusion nuclear regime in mind, the Fusion Energy System Studies group examined LM PFCs in order to identify needed research thrusts that could accelerate their development and assess their viability. Liquid metal behavior, solid substrate aspects, and fusion facility integration aspects are examined, with concepts as the research focusing element. The concepts applied to a fusion nuclear device are the primary definer of the LM parameters, environmental conditions, and operational aspects. This forms the research strategy recommended for these complex systems.