Search NASA⌕ Search

SEARCH · Search NASA

Results for “plasma facing component”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

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.

Divertor↗
DOE OSTI2026

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.

Parsons, Matthew S. [Princeton Plasma Physics Labo↗
DOE OSTI2024

Integrated modeling of boron powder injection for real-time plasma-facing component conditioning

An integrated modeling framework for investigating the application of solid boron (B) powder injection for real-time surface conditioning of plasma-facing components (PFCs) in tokamak environments is presented. Utilizing the DIII-D impurity powder dropper (IPD) setup, this study simulates B powder injection scenarios ranging from milligrams to tens of milligrams per second, corresponding to boron flux rates of 10 20 –10 21 B/s in standard L-mode conditions. The comprehensive modeling approach combines EMC3-EIRENE for simulating the deuterium plasma background and the Dust Injection Simulator (DIS) for the ablation and transport of the boron powder particles. EMC3 trace impurity fluid modeling results show substantial boron transport to the inboard lower divertor, predominantly influenced by the main ion plasma flow. The dependency on powder particle size (5-250 µm) was found to be insignificant for the scenario considered. The effects of erosion and redeposition were considered to reconcile the discrepancies with experimental observations, which saw substantial deposition on the outer divertor plasma-facing components. For this purpose, the WallDYN3D code was updated to include boron sources within the plasma domain and integrated into the modeling framework. The mixed-material migration modeling shows evolving boron deposition patterns, suggesting the formation of mixed B-C layers or predominantly B coverage depending on the powder mass flow rate. While the modeling outcomes at lower B injection rates tend to align with DIII-D experimental observations, the prediction of near-pure boron layers at higher rates has yet to be experimentally verified in the carbon environment of the DIII-D tokamak. The extensive reach of boron layers found in the modeling suggests the need for modeling that encompasses the entire wall geometry for more accurate experimental correlations. This integrated approach sets a precedent for analyzing and applying real-time in-situ boron coating techniques in advanced tokamak scenarios, potentially extendable to ITER.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Runaway electron-induced plasma facing component damagein tokamaks

This Roadmap article addresses the critical and multifaceted challenge of plasma facing component (PFC) damage caused by runaway electrons (REs) in tokamaks, a phenomenon that poses a significant threat to the viability and longevity of future fusion reactors such as ITER and DEMO. The dramatically increased RE production expected in future high-current tokamaks makes it very difficult to avoid or mitigate REs in such devices when a plasma discharge terminates abnormally. Preventing damage from the intense localised heat loads they can cause requires a holistic approach that considers plasma, REs and PFC damage. Despite decades of progress in understanding the physics of REs and the thermomechanical response of PFCs separately, their complex interplay remains poorly understood. This document aims to initiate a coordinated, interdisciplinary approach to bridge this gap by reviewing experimental evidence, advancing diagnostic capabilities, and improving modelling tools across different scales, dimensionalities, and fidelities. Key topics include RE beam formation and transport, damage mechanisms in both brittle and metallic PFCs, and observed effects in major facilities such as JET, DIII-D, WEST and EAST. The Roadmap emphasises the urgency of predictive, high-fidelity modelling validated against well-diagnosed controlled experiments, particularly in the light of recent changes in ITER’s wall material strategy and the growing importance of private sector fusion initiatives. Each section of the Roadmap article is written to provide a concise overview of one area of this multidisciplinary subject, with an assessment of the status, a look at current and future challenges, and a brief summary. The ultimate goal of this initiative is to guide future mitigation strategies and design resilient components that can withstand the intense localised loads imposed by REs, thus ensuring the safe and sustainable operation of the next generation of fusion power plants.

Ratynskaia, Svetlana [KTH Royal Inst. of Technolog↗
DOE OSTI2024

Data and figures for "Integrated modeling of boron powder injection for real-time plasma-facing component conditioning"

This dataset contains raw and processed data, as well as supplementary figures used in the paper titled "Integrated modeling of boron powder injection for real-time plasma-facing component conditioning." The data includes simulation results for boron transport and deposition in DIII-D tokamak scenarios, and processed plots. It provides insights into the effects of boron powder injection on plasma-facing component conditioning and surface composition.

ablative particle injection↗
DOE OSTI2026

Surface Segregation of Liquid Metal Plasma-Facing Component Alloys: A ReaxFF Investigation

Engineering liquid metal alloys offers a transformative pathway for plasma-facing components by enabling chemically tailored surfaces that can simultaneously optimize plasma-material interactions, reduce divertor heat flux, and enhance core plasma confinement, thereby advancing the commercial viability of nuclear fusion power plants. This study, employing an atomistic simulation approach, provides direct evidence that incorporating nonmetal surface-active agents (such as O and H, or their combination) enables strong surface segregation. This capability makes tin−aluminum (Sn−Al) and tin−lithium (Sn−Li) alloys, with suitable compositions, good candidates for PFC applications. Specifically, the presence of low-Z solutes (Li, Al) leads to preferential surface enrichment, which imparts low-Z sputtering characteristics, while the Sn solvent maintains thermophysical stability. To systematically examine this behavior, we developed a ReaxFF force field spanning the full Sn/Al/Li/O/H chemistry, validated it against formation energies and elastic constants, and applied it in reactive molecular dynamics simulations at fusion-relevant temperatures. We also introduced an overlapbased segregation index that captures interfacial compositional separation directly from atomistic density distributions. Here, this metric reveals a clear hierarchy of segregation regimes and provides a unified view across all systems studied. Together, these findings establish a mechanistic link between nonmetal chemistry and interfacial structure, providing a predictive framework for designing self-adaptive, low-sputtering liquid metal alloys for fusion applications.

Alloys↗
DOE OSTI2026

Assessing plasma face component thermal response to rotating 3D magnetic fields for SPARC tokamak

Thermal response simulations of plasma-facing components (PFCs) in the SPARC tokamak, performed with the HEAT code, show that three-dimensional (3D) heat loads resulting from stationary n=1 perturbations require highly radiative scenarios, with up to 95% of the power crossing the separatrix (P SOL ) being radiated, to maintain PFC temperatures within acceptable operational limits, whereas the application of slowly rotating 3D fields substantially reduces the thermal loads. The HEAT module, developed to predict heat loads from non-axisymmetric plasmas, is extended to model time-dependent heat flux patterns generated by rotating 3D fields, and a comprehensive thermal analysis is performed on PFCs subjected to both the maximum and minimum power loads, as well as to rotating heat flux distributions, to evaluate the temperature evolution for varying perturbation amplitudes and rotation frequencies. The extension of this analysis to 3D fields with toroidal mode number n=2 shows that this configuration leads to weaker localized heat flux peaks relative to the n=1 case, enabling safe operation with less than 80% of the power radiated when static 3D fields of low amplitude are applied, while using slowly rotating fields at higher amplitudes. These results indicate that n=2 perturbations are generally less detrimental to divertor power exhaust, emphasizing the strong dependence of divertor power exhaust on the characteristics of the applied 3D fields.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

CFD Simulations of Molten Salts Within Plasma-Facing Components with Twisted Tape Inserts Under One-Sided–Heated Conditions

For liquid immersion blankets (LIBs), the heat removal within different plasma-facing components (PFCs) requires substantial heat removal capabilities that could benefit from the usage of heat transfer enhancements. Of consideration are twisted tape insert heat transfer enhancements that are ribbon-like metal pieces inserted within cooling channels that increase fluid mixing, resulting in higher heat transfer and pressure drop. Unlike traditional applications, PFCs such as divertor coolant tubes experience a significantly higher heat flux on one side versus the other or entirely one-sided heating. These conditions are not well studied for molten salts where the higher Prandtl numbers characteristic of the fluids are significantly higher than water, gases, and liquid metals. As an effort to study this knowledge gap, large eddy simulations were done with Prandtl numbers representative of FLiBe at proposed operating conditions. In this study, the effects of applying a surface heat flux to one side of a tube versus the entire tube with a twisted tape insert were investigated for the heat transfer coefficient and effects on the temperature profile. The findings of the study suggest that for Prandtl number fluids above unity, the Nusselt number was largely unaffected by the change in boundary condition. Both boundary conditions showed good agreement with the Manglik and Bergles Nusselt number correlation for tight-fitting twisted tapes. Finally, the outcomes of this study will be useful for future design studies of LIBs to provide fundamental guidance implementing twisted tape inserts and their performance in one-sided–heated conditions.

molten salts↗
DOE OSTI2025

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.

Hatler, Caleb [University of Wisconsin-Madison]↗
DOE OSTI2025

Shadow masks predictions in SPARC tokamak plasma-facing components using HEAT code and machine learning methods

Here, this work uses machine learning (ML) to complement HEAT (Heat flux Engineering Analysis Toolkit) by developing 3-D footprint surrogate models for fast and accurate heat load calculations in the divertor of the SPARC tokamak. The focus is on shadowed regions, or magnetic shadows, caused by the 3-D geometry of plasma-facing components (PFCs). ML classifiers are employed to create a surrogate model for HEAT generated shadow masks, predicting these shadow masks and divertor heat flux profiles based on a diverse range of equilibria and only the plasma current, safety factor(q95) at the edge, and magnetic flux angles as input parameters. The ultimate goal is to integrate the model for real-time control and future operational decisions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Retention and surface morphology evaluation of fine-grain dispersion-strengthened tungsten for plasma-facing component applications

This study exposed novel fine-grain dispersion-strengthened tungsten (W) to high fluence, low energy deuterium (D) and helium (He) plasmas to evaluate how material microstructure and composition affect hydrogen retention and surface morphology. Tested materials included fine-grain dispersion-strengthened tungsten (DSW) with 3 wt% zirconium carbide (ZrC) dispersoids, fine-grain dense W without any dispersoids (FGW), and coarse-grained polycrystalline ‘ITER-grade’ W. Samples were exposed to D 2 + and He + plasmas at fusion-relevant fluences (∼10 25 m -2 ) and ion energies (75 eV) over a range of temperatures (200 °C, 300 °C, 450 °C for D, 850 °C for He). Helium ion microscopy was performed on the exposed samples to evaluate surface morphology changes and material integrity. After D plasma exposure, the ZrC dispersoids showed near-surface degradation at exposure temperatures above 300 °C, but no detrimental morphology changes were observed for the adjacent W grains. After He plasma-exposure, nano-structured fuzz formation was observed in the tungsten matrix of all samples. The ZrC dispersoids maintained their integrity despite the surrounding fuzz growth, with clear delineation between the W fuzz and dispersoid regions. Thermal desorption spectroscopy showed that ZrC DSW consistently retained more D than the FGW by about a factor of 2 across all temperatures. At 200 °C and 300 °C, the ITER-W displayed lower D retention than both the DSW and FGW, however at 450 °C ITER-W showed the highest retention, about 50% more than DSW. He retention was comparable across all samples, with the highest retention observed in the fine-grain W, only 26% higher than in ITER-W. These insights on retention behavior will inform further optimization of these novel fine-grained tungsten materials with and without dispersoid additives.

Dispersion-strengthened tungsten↗
DOE OSTI2026

Surface normal emissivity measurements in uncoated and oxidized B-coated plasma facing component materials in the long wavelength infrared spectrum

Accurate surface emissivity measurements are integral to the fidelity of the infrared thermography diagnostic evaluation in fusion reactors. The emissivities of ATJ TM graphite, Sigrafine® (R6510) graphite, and TZM alloy were measured as a function of temperature in accordance with the Contact Thermometer Method in the ASTM E1933 standard. Samples were heated resistively in a high-vacuum chamber, and the surface temperature was monitored using a surface thermocouple and a Telops long-wavelength infrared camera. The surfaces of Sigrafine® graphite and TZM alloy samples were coated with 10 nm and 20 nm layers of oxidized boron and were also measured in the uncoated condition. The thickness was assessed from measurement by a quartz crystal microbalance (QCM). ATJ TM graphite was uncoated. The measured emissivity of ATJ TM Graphite ranged from 0.82 to 0.83, uncoated Sigrafine® graphite ranged from 0.76 to 0.98, and uncoated TZM alloy ranged from 0.11 to 0.13. Oxidized boron coatings increased the emissivity of the TZM alloy to 0.18–0.23 but decreased the emissivity of Sigrafine® graphite to 0.56–0.66 for a 10 nm coating and to 0.65–0.74 for a 20 nm coating. In contrast to the expected blackbody radiance, the emissivity of uncoated Sigrafine® graphite and TZM alloy did not monotonically increase with temperature from 100 to 500 ℃.

Emissivity↗
DOE OSTI2025

Comparative Mechanical Properties Analysis of Triple Ion-Beam Irradiated and Neutron Irradiated Potential Plasma Facing Components

Abstract Several classes of materials are being proposed for use in fusion reactors including oxide dispersion strengthened (ODS) and reduced activation ferritic-martensitic (RAF/M) steels to withstand the severe and harsh conditions. In this work, the mechanical properties of a Fe-16Cr-4Al-2W-0.3Ti-0.3Y 2 O 3 (K3) (ODS) ferritic steel and a Fe-8.9Cr-1.1W-0.47Mn-0.2V-0.14Ta-0.11C (Eurofer 97) (RAF/M) steel) after triple ion beam irradiation were locally evaluated utilizing in-situ micro-pillar compression tests, and continuous stiffness/quasi-static nanoindentation. No change in mechanical properties was observed in the K3 ODS steel. However, the Eurofer 97 RAF/M steel exhibited radiation-induced effects via increases in yield strength. Micro-pillar techniques were expanded to neutron-irradiated materials via an in-situ testing technique employing lift-out methods on Fe-14Cr-0.9Ti-0.3Mo-0.25Y 2 O 3 (MA957) ODS ferritic steel. Both the non-irradiated and irradiated compressive yield stresses of the MA 957 micro-pillars were in good agreement with bulk yield stress values reported in the literature, suggesting that the lift-out micro-pillar compression testing technique is a promising method. The demonstration of these techniques on ion beam and neutron irradiated ODS steels and ion beam RAF/M steels gives information to inform models of the material degradation during use in a fusion device.

alloys↗
DOE OSTI2025

Self-consistent modeling of tokamak edge plasma transport with lithium sources

Magnetic confinement fusion devices require effective heat and particle exhaust solutions on the divertor plates to operate sustainably, especially under reactor-relevant conditions. Liquid lithium divertors have been proposed to address two major challenges: control of excessive heat flux to plasma-facing components through vapor shielding and minimization of core plasma contamination from impurities. The National Spherical Torus Experiment-Upgrade (NSTX-U) will explore lithium as a divertor material due to its potential to meet both objectives. We present a self-consistent coupling framework between the plasma boundary transport code UEDGE and the lithium wall transport code Wall–Li to evaluate the feasibility and operational limits of lithium-based divertors. The model aims to optimize lithium sourcing levels to prevent core plasma contamination via fuel dilution while ensuring divertor protection through vapor shielding. This integrated framework, applicable to any tokamak with lithium sources, dynamically adjusts lithium sourcing based on local plasma conditions and surface temperature. The coupled model is tested using NSTX-like geometry and plasma conditions to assess its performance and reliability. Wall–Li calculates lithium fluxes from plasma-facing components, incorporating physical sputtering, thermally enhanced sputtering, and evaporation driven by surface temperature and ion flux. These fluxes are reintroduced into UEDGE as neutral lithium atoms, enabling simulation of their transport and distribution within the plasma. UEDGE computes plasma and neutral transport, surface heat flux, and iteratively feeds this information back to Wall–Li. A small time step is employed to ensure numerical stability and convergence, enabling accurate simulations over typical tokamak discharge durations. This integrated modeling approach provides a robust tool for identifying operational regimes that balance effective lithium sourcing with minimal core plasma contamination, offering critical insights for optimizing lithium-based divertor systems in current and future fusion devices.

Magnetic confinement fusion↗
DOE OSTI2025

The effect of ICRH and NBI on tungsten transport in neon edge radiation cooled discharges in TEXTOR

Tungsten is planned to be the plasma-facing material for the main chamber and divertor in future devices like ITER, SPARC, and DTT. To address risks associated with tungsten, R&D is being carried out on various toroidal confinement devices. One particular research question is related to the optimum heating mix for ITER. To investigate the influence of the heating scheme on the release and transport of tungsten, a comparison of neutral beam injection (NBI) and ion cyclotron resonance frequency (ICRF) heating was carried out in TEXTOR tungsten test limiter experiments. The experiments were performed under standard L-mode conditions and in radiative improved-mode operation with neon seeding and boronized walls covering the graphite plasma-facing components. The plasma was heated with hydrogen or deuterium NBI alone or with deuterium NBI in combination with H-minority, ion cyclotron resonance heating. A movable solid tungsten limiter was inserted through a limiter lock system into the edge plasma. The impurity release from this limiter was evaluated from visible spectroscopy. The tungsten concentrations in the plasma core were determined by extreme ultraviolet spectroscopy and bolometry. With deuterium NBI alone, strong central radiation and accumulation of W was observed. This can, however, be avoided by adding ICRF heating.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Kinetic Plasma Simulation in the MOOSE Framework: Verification of Electrostatic Particle In Cell Capabilities

In magnetic confinement nuclear fusion reactors, the interaction between the plasma edge and plasma facing components is extremely important. At the plasma edge, a kinetic representation such as particle-in-cell (rather than a fluid representation) is required to accurately capture the plasma behavior. General purpose particle-in-cell plasma simulation capabilities have been developed in the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework. This new capability is a part of the development of a new MOOSE-based framework for modeling plasma facing components, the Fusion ENergy Integrated multiphys-X (FENIX) framework. In this work, the verification of foundational particle-in-cell capabilities in FENIX is presented. This new plasma simulation capability has three main components: moving particles in discrete steps on the finite element mesh, mapping charge density from the particle's location to the finite element mesh, and solving for the electrostatic potential based on the charge density mapped from particles to the mesh. In this paper, simple verification problems demonstrating each of these new capabilities are presented, and future work includes electromagnetic capabilities and Monte Carlo collisions with neutral gas particles.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

FreeMHD: Validation and verification of the open-source, multi-domain, multi-phase solver for electrically conductive flows

The extreme heat fluxes in the divertor region of tokamaks may require an alternative to solid plasma-facing components, for the extraction of heat and the protection of the surrounding walls. Flowing liquid metals are proposed as an alternative, but raise additional challenges that require investigation and numerical simulations. Free surface designs are desirable for plasma-facing components, but steady flow profiles and surface stability must be ensured to limit undesirable interactions with the plasma. Previous studies have mainly used steady-state, 2D, or simplified models for internal flows and have not been able to adequately model free-surface liquid metal (LM) experiments. Therefore, FreeMHD has been recently developed as an open-source magnetohydrodynamics (MHD) solver for free-surface electrically conductive flows subject to a strong external magnetic field. The FreeMHD solver computes incompressible free-surface flows with multi-region coupling for the investigation of MHD phenomena involving fluid and solid domains. The model utilizes the finite-volume OpenFOAM framework under the low magnetic Reynolds number approximation. FreeMHD is validated using analytical solutions for the velocity profiles of closed channel flows with various Hartmann numbers and wall conductance ratios. Next, experimental measurements are then used to verify FreeMHD, through a series of cases involving dam breaking, 3D magnetic fields, and free-surface LM flows. These results demonstrate that FreeMHD is a reliable tool for the design of LM systems under free surface conditions at the reactor scale. Furthermore, it is flexible, computationally inexpensive, and can be used to solve fully 3D transient MHD flows.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗