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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.

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At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Design, fabrication, and testing of a water-cooled tungsten-foam high-heat flux plasma-facing module

Here, we present a comprehensive study detailing the design, fabrication and testing a new modular “finger” component, featuring a hexagonal tungsten foam armor tile bonded to the top surface of the water-cooled finger unit. The mechanical, thermal, and particle transport properties of the tungsten foam armor were determined for a wide range of porosity and cell size, and then combined in an equivalent solid material with “effective” properties suitable for large-scale simulations. A series of 3-D multiphysics simulations for coupled thermal and mechanical behavior of the “finger” module are used to optimize the design before fabrication. The simulation results show that, for the expected experimental conditions, a safe range of heat flux and foam porosity is possible for both water-cooling and helium-cooling. Based on these multiphysics simulations, an optimum finger cooling module has been fabricated at Ultramet, Inc. A series of experiments were designed to study the thermomechanical performance of the “finger” unit under steady-state and transient thermal loading at surface heat flux up to 16 MW/m 2 in the High Energy Flux Test facility (HEFTY) at UCLA. Testing conditions included low-power, steady-state plasma exposure at 6 MW/m 2 , followed by severe high heat flux cycles at 16 MW/m 2 . Results of these tests are explained through comparisons with multiphysics simulations. To eliminate the effects of thermomechanical damage carried out on a module on its thermo-mechanical response on a subsequent test, three different yet identically designed and fabricated finger modules were tested. The results show that the current design is capable of cooling the walls of fusion devices at surface heat flux up to 6 MW/m 2 without failure, and that the failure heat flux limit is below 16 MW/m 2 .

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Chelation-directed interface engineering of in-place self-cleaning membranes

Water–energy sustainability will depend upon the rapid development of advanced pressure-driven separation membranes. Although energy-efficient, water-treatment membranes are constrained by ubiquitous fouling, which may be alleviated by engineering self-cleaning membrane interfaces. In this study, a metal-polyphenol network was designed to direct the armorization of catalytic nanofilms (ca. 18 nm) on inert polymeric membranes. The chelation-directed mineralized coating exhibits high polarity, superhydrophilicity, and ultralow adhesion to crude oil, enabling cyclable crude oil-in-water emulsion separation. The in-place flux recovery rate exceeded 99.9%, alleviating the need for traditional ex situ cleaning. The chelation-directed nanoarmored membrane exhibited 48-fold and 6.8-fold figures of merit for in-place self-cleaning regeneration compared to the control membrane and simple hydraulic cleaning, respectively. Precursor interaction mechanisms were identified by density functional theory calculations. Here, chelation-directed armorization offers promise for sustainable applications in catalysis, biomedicine, environmental remediation, and beyond.

42 ENGINEERING↗
DOE OSTI2024

The Maintainable Fusion Pilot Plant

The US fusion community has coalesced around the goal of building an FPP as described by the National Academies of Science, Engineering, and Medicine (NASEM). In addition to demonstrating the viability of the technologies necessary to operate such a plant, including demonstration of net energy and electricity production, NASEM found that a “fusion pilot plant will need to demonstrate the ability to efficiently perform remote maintenance and replacement in support of the design of a power plant, taking into account details of the consequences of the fusion environment, such as material activation and tritium retention in components.” Current designs of fusion demonstration reactors do usually foresee a regular exchange of their first wall modules, including the tritium breeding blankets. In the European Power Plant Conceptual Studies, it is assumed that a fusion reactor will need to change its divertor every 2 years and its first wall blanket module every 5 to 6 years to reach acceptable availability. Underlying this capability are remote-handling technologies to keep the outage for the exchange of these components short. There are many uncertainties in the remote-handling schemes, and most schemes are at a preconceptual level at best. In addition, the exchange of these components would either produce an enormous rad-waste stream or would require an enormous refurbishment activity with huge cost-prohibitive hot-cells. Past Fusion Nuclear Science Facility (FNSF) preconceptual studies have led to hot cell dimensions of an unbelievable size, likely costing tens of billions of dollars. Already at The Way (previously International Thermonuclear Experimental Reactor, ITER), hot-cells have become cost-prohibitive, demanding redesigns of the ITER first wall to reduce the toxic rad-waste/inventory. In this in-situ PFC repair project, a concept for a long-life, maintainable first wall module concept is developed and tested. This first wall concept relies on innovative remote handling to repair the first wall modules in-situ, avoiding costly refurbishments outside of the tokamak vessel. This approach was highlighted in the Fusion Energy Sciences Advisory Committee (FESAC) report on Transformative Enabling Capabilities for Efficient Advance Toward Fusion Energy. In general, the damage of the first wall armor is due to particle and radiation exposures. Load conditions vary from one fusion reactor design to another. In tokamaks, first wall Plasma Facing Components (PFCs) are exposed to far-Scrape-Off-Layer plasma fluxes, electromagnetic radiation, energetic CX neutrals, and potentially runaway electron beams. Protecting the first wall to the worst-case load conditions would require the design of a very thick first wall armor. Transient heat and particle fluxes due to disruptions or edge localized modes can lead to excessive heat loads resulting potentially in melting PFCs down to the cooling channel. Catastrophic events like these need to be avoided by appropriate disruption mitigation systems. However, failure of these systems will still put a first wall at an unacceptable risk. Hence, a first wall design needs to accommodate the occasional transient heat loads by introducing sacrificial limiters, which will absorb these transients.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Dimensionless parameters for ballistic performance evaluation of ceramic-faced bicomponent targets against sharp-nosed projectiles

Two dimensionless parameters are proposed to give first-order approximations for the ballistic performance of ceramic-faced bicomponent armor plates impacted by conical-nosed steel projectiles. The current work collates expansive experimental data from Wilkins and Mayseless covering a broad range of material and mechanical properties, target construction, and projectile dimensions, the data is collapsed using the Hugoniot elastic limit of the ceramic strike face as a strength metric. We find that within datasets where experimental conditions are kept constant, a singular cubic curve can be generated, but variations in experimental conditions for the ballistic limit test between datasets can result in differences in the predictive curve coefficients.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Employee Spotlight: Christopher Graham

Christopher Graham of Special Security IT and Cyber Operations (SSO-3) makes medieval armor using a combination of old-school techniques and tools with modern accouterments, such as a blacksmith’s gas-powered forge and a power hammer.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2021

Process Robustness of Friction Stir Dovetailing of AA7099 to Steel with In Situ AA6061 Interlayer Linking

Rolled homogeneous armor (RHA) steel was joined to aluminum alloy AA7099-T7451 with an interlayer of AA6061-T651 by single-pass friction stir dovetailing (FSD). The evolution of the AA7099/AA6061 and AA6061/steel interfacial microstructures and mechanical properties of the joints were examined. The AA6061 interlayer prevented Zn migration from the AA7099 to the steel during FSD and inhibited formation of brittle, Zn-rich, Fe-Al intermetallic compounds within the joint. The flow of the dissimilar aluminum alloys at the interface was controlled with a partially threaded friction stir welding pin and a dovetail geometric configuration optimized in previous work. The process robustness was explored to establish parameter windows by elucidating parametric effects on the relationship between process conditions and the joint’s microstructure and mechanical properties. Ranges of process forge force (50–70 kN) and interface temperature (475°C–515°C) consistently yielded joints with adequate performance. The load-carrying capacity in lap shear tensile tests was 2025 ? 100 N/mm.

Friction Stir Dovetailing, FSW, AA7099, AA6061, Ro↗
DOE OSTI2026

Mechanical characterization of fine-grain dispersion-strengthened tungsten as a plasma facing material

Field-Assisted Sintering Technology (FAST) was used to produce fine-grained, dispersion-strengthened tungsten (W) materials. Investigated materials 4138, 4353, and 4355 composed of 3 wt% ZrC sintered at 1800 °C, 5 wt% ZrC sintered at 1800 °C, and 3 wt% ZrC sintered at 2000 °C, respectively. They were compared against ITER-grade W. A series of mechanical and thermal property testing and microstructure studies were conducted to study them as a potential plasma facing material (PFM) for fusion reactors. Hardness testing showed that manufacturing conditions substantially altered hardness. Material 4355 had an average HV10 value of 497.2 ± 16.8, slightly higher than ITER-grade at 378.5 ± 40.3. However, material 4353 was substantially higher with an HV10 value of 738.9 ± 31.7 over the investigated temperature range. Electron Backscatter Diffraction (EBSD) analysis showed that FAST produced substantially smaller grains than the hot-rolled ITER-grade W material, offering notable control over grain size. Materials 4353 and 4355 had grain sizes of 0.44 ± 0.20 µm and 3.67 ± 0.89 µm, respectively, whereas ITER-grade 27.14 ± 19.76 µm at room temperature. The fine grain structures showed no net coarsening after 1 hr. anneals up to 1800 °C, several hundred degrees above the 1100 – 1500 °C recrystallization range reported for conventional W. Inverse application of the Zener pinning relationship to the measured grain sizes indicates that these two FAST sintering conditions produce markedly different effective dispersoid populations, with effective particle diameters of approximately 90 nm at a peak sintering temperature of 1800 °C and approximately 460 nm at 2000 °C, respectively. This result demonstrates that the FAST thermal condition itself, and not the nominal ZrC content alone, governs the pinning effectiveness of the dispersion. Thermal diffusivity measurements support this finding independently. Materials of identical composition sintered at different temperatures differ by approximately 19% in measured thermal diffusivity with statistically indistinguishable density, while materials of different composition and sintering temperature converge to within approximately 2%. At a representative divertor heat flux of 10 MW/m², the lower thermal conductivity of the fine-grained materials corresponds to approximately 28 to 33 °C per millimeter of armor thickness relative to ITER-grade W, traded against a substantially larger margin to recrystallization-driven degradation. While high temperature tensile testing revealed likely contamination that motivates refinement of the manufacturing process, FAST-produced, fine-grained, dispersion-strengthened W offers process-controlled microstructural stability well above the operating temperatures of conventional W and supports its continued development as a PFM for economically viable commercial fusion power.

Parker, Gabe [ORNL] (ORCID:0000000190394100)↗
DOE OSTI2024

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. Here, this study serves as a preliminary examination of W-SiC compositionally graded composites as a potential candidate divertor material in future tokamak devices.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Comparative neutron-irradiation effects on thermal conductivity degradation and dimensional stability of TiC, TiB 2 , and ZrB 2 at 200–1000 °C

Ultra-high-temperature ceramics (UHTCs), including TiC, Ti 11 B₂, and Zr 11 B₂, show great potential for plasma-facing components due to their excellent high-temperature properties prior to irradiation. However, their response to neutron irradiation remains insufficiently understood, limiting robust assessment of their viability for fusion energy applications. Here, this study examines the thermal conductivity, dimensional stability and microstructure of TiC, TiB₂, and ZrB₂ following neutron irradiation at temperatures of 200–1000 °C and fast neutron fluences of 2.0 × 10 25 to 1.1 × 10 26 n/m 2 (E > 0.1 MeV). Lattice swelling measured by synchrotron X-ray diffraction in all three UHTCs was maximized at 200 °C and decreased with increasing irradiation temperature, with no evidence of amorphization observed at 200 °C. Above 600 °C, significant macroscopic volume swelling was observed in irradiated Ti 11 B₂ and Zr 11 B₂, but not in TiC, likely due to cavity formation in the diborides. The post-irradiation thermal conductivity, measured at the irradiation temperature, ranged from 28 to 45 W/m·K, representing a 34–45% reduction relative to the unirradiated material. Notably, neutron-irradiated UHTCs exhibit recoverable thermal conductivity at elevated temperatures, comparable to ferritic–martensitic steels and potentially superior to W when transmutation effects are considered, highlighting promise for shielding or armor plasma-facing components. At 600 °C, both thermal conductivity degradation and lattice swelling saturated at doses exceeding 2–4 dpa.

fusion materials↗
DOE OSTI2025

Impact of mixed grain size sediment input on braided river Morphodynamics: Insights from experimental and numerical modeling

The rate of sediment supply has significant impacts on river morphology, making it crucial to understand the geomorphic changes and grain size distribution dynamics in rivers. However, the effects of varying grain size sediment input on morphological changes in braided channels remain poorly understood. This study is the first to investigate the bar development and sediment sorting processes in braided channels with non-uniform sediment inputs using both numerical and experimental approaches. We applied a two-dimensional numerical model, Nays2DH to confirm and generalize experimental results. The model reproduced key experiment results, including 1) stream elevation changes, and 2) grain size distribution. Using this validated model, we explored the morphological changes and sorting process in a braided river with sediment inputs. The numerical experiments demonstrate that sediment input controls the elevation of the stream bed and the grain size distribution. Notably, both the elevation and grain-size distribution become relatively stable in downstream of the channel. Additionally, the simulation results suggest that an increased sediment supply leads to greater channel complexity, with bed surface armoring decreasing.

Nays2DH↗
DOE OSTI2025

Packing fraction control during additive manufacturing of powder green bodies

The demand for high performance ceramic and metal components with complex geometry necessitates developments in powder handling and green body shaping. Here, in this study, vibrational powder deposition is adapted to three-dimensional printing with the ability to modulate packing fraction during printing. Boron carbide powder with sub-micron primary particle size is printed over a packing fraction range of 25.0 % to 46.5 % (below the poured density to above the tapped density). Solid metal powders are printed at 67.5 %, which enables novel freestanding, vertical-walled features without binder. These capabilities introduce opportunities for more complex geometries in high-performance materials, e.g. large-scale uniaxial hot-pressing of ceramic ballistic armor components with graded thickness and three-dimensional curved or stepped features (by compensating for displacement differences during compaction). This technology also enables multi-material patterning of additive manufacturing powder beds with reduced feedstock quantity requirements and wastage.

36 MATERIALS SCIENCE↗
DOE OSTI2024

The influence of microbial colonization on inflammatory versus pro-healing trajectories in combat extremity wounds

A combination of improved body armor, medical transportation, and treatment has led to the increased survival of warfighters from combat extremity injuries predominantly caused by blasts in modern conflicts. Despite advances, a high rate of complications such as wound infections, wound failure, amputations, and a decreased quality of life exist. To study the molecular underpinnings of wound failure, wound tissue biopsies from combat extremity injuries had RNA extracted and sequenced. Wounds were classified by colonization (colonized vs. non-colonized) and outcome (healed vs. failed) status. Differences in gene expression were investigated between timepoints at a gene level, and longitudinally by multi-gene networks, inferred proportions of immune cells, and expression of healing-related functions. Differences between wound outcomes in colonized wounds were more apparent than in non-colonized wounds. Colonized/healed wounds appeared able to mount an adaptive immune response to infection and progress beyond the inflammatory stage of healing, while colonized/failed wounds did not. Although, both colonized and non-colonized failed wounds showed increasing inferred immune and inflammatory programs, non-colonized/failed wounds progressed beyond the inflammatory stage, suggesting different mechanisms of failure dependent on colonization status. Overall, these data reveal gene expression profile differences in healing wounds that may be utilized to improve clinical treatment paradigms.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2021

High resolution >40 keV x-ray radiography using an edge-on micro-flag backlighter at NIF-ARC

Radiography of low-contrast features in high-density materials evolving on a nanosecond timescale requires a bright photon source in the tens of keV range with high temporal and spatial resolution. One application for sources in this category is the study of dynamic material strength in samples compressed to Mbar pressures at the National Ignition Facility, high-resolution measurements of plastic deformation under conditions relevant to meteor impacts, geophysics, armor development, and inertial confinement fusion. Here, we present radiographic data and the modulation transfer function (MTF) analysis of a multi-component test object probed at ~100 keV effective backlighter energy using a 5 μm-thin dysprosium foil driven by the NIF Advanced Radiographic Capability (ARC) short-pulse laser (~2 kJ, 10 ps). The thin edge of the foil acts as a bright line-projection source of hard x rays, which images the test object at 13.2× magnification into a filtered and shielded image plate detector stack. The system demonstrates a superior contrast of shallow (5 μm amplitude) sinusoidal ripples on gold samples up to 90 μm thick as well as enhanced spatial and temporal resolution using only a small fraction of the laser energy compared to an existing long-pulse-driven backlighter used routinely at the NIF for dynamic strength experiments.

47 OTHER INSTRUMENTATION↗
DOE OSTI2022

Characterization of a 1D-imaging high-energy x-ray backlighter driven by the National Ignition Facility Advanced Radiographic Capability laser

Plastic deformation of samples compressed to Mbar pressures at high strain rates at the National Ignition Facility (NIF) forms the basis of ongoing material strength experiments in conditions relevant to meteor impacts, geophysics, armor development, and inertial confinement fusion. Hard x-ray radiography is the primary means of measuring the evolution of these samples, typically employing a slit-collimated high-Z microdot driven by the NIF laser to generate >40 keV x rays [E. Gumbrell et al., Rev. Sci. Instrum. 89, 10G118 (2018) and C. M. Huntington et al., Rev. Sci. Instrum. 89, 10G121 (2018)]. Alternatively, a dysprosium “micro-flag” target driven by the Advanced Radiographic Capability laser (~2 kJ, 10 ps) can deliver significantly higher spatiotemporal resolution [M. P. Hill et al., Rev. Sci. Instrum. 92, 033535 (2021)], especially in high-opacity samples. Here, initial experiments revealed problematic brightness and spectral gradients from this source, but by radiographing a set of diamond-turned, 105 µm-thick Pb test objects and supported by simulations using the 3D Monte Carlo code GEANT4, these geometry-dependent gradients across the field of view are quantified and mitigation strategies are assessed. In addition to significantly enhancing the modulation transfer function compared to the existing system, image stacking from multiple layers of image plate is shown to almost double the signal to noise ratio that will reduce uncertainties in future dynamic strength experiments.

47 OTHER INSTRUMENTATION↗
DOE OSTI2025

Temperature effects of ion irradiation on the nanostructural features in ductile-phase-toughened tungsten composites

Ductile-phase toughened tungsten (DPT W) composites have emerged as promising candidates for load-bearing components behind the plasma-facing tungsten armor in fusion reactors due to their enhanced thermomechanical properties. This study focuses on a composite consisting of W particles embedded in a ductile NiFeW solution matrix, hot-rolled to a thickness reduction of 87% (87R DPT W). Sequential irradiations with Ni2+ and He+ ions were performed to identical doses and helium concentrations at room temperature (RT) and 1273 K. Irradiation at RT produced no discernible nanostructural features due to the immobility of mono-vacancies, whereas cavity formation was observed at 973 K. At 1273 K, the W phase exhibited larger cavities, reduced cavity number density, and lower volumetric swelling compared to 973 K. Notably, nanosized NiFeW precipitates formed within the W phase at 1273 K, a phenomenon absent at 973 K. A new phase of cubic (NiFe)6W6C was also observed at the interphase boundary. In contrast, the NiFeW matrix showed no nanostructural changes at 1273 K, likely due to cavity dissociation. Separate irradiations at 1273 K indicated that Ni2+ ions induced precipitate formation in the W phase, while He+ ions exclusively caused cavity formation. The microstructure of 87R DPT W irradiated at RT and subsequently annealed at 1273 K closely resembled that of material irradiated directly at 1273 K. Like oxide-dispersion-strengthened steels, the observed nanoparticle-embedded W can inhibit dislocation propagation, potentially delaying the ductile-to-brittle transition temperature. These findings highlight the potential of NiFeW nanoparticle-reinforced W composites as irradiation-resistant materials for fusion reactors.

Jiang, Weilin (ORCID:0000000183028313)↗
DOE OSTI2024

Heat affected zone liquation cracking evaluation on FeMnAl alloys

FeMnAl steels are currently generating a lot of interest with potential applications for structural parts in armored vehicles due to their lower density and outstanding mechanical properties. Despite the extensive mechanical performance and heat treatment exploration of this alloy class, further weldability investigation is required for future large-scale deployment. In the present study, the liquation cracking of four heats of cast FeMnAl alloys was investigated by the spot-Varestraint technique. The study focuses primarily on the effect of the major elements of the FeMnAl system: C, Mn and Al. Optical and electron microscopy were employed to investigate the microstructural features, and CALPHAD was employed to aid the discussion regarding the alloy’s composition differences and their effect on the liquation cracking susceptibility. The study was able to identify that compositions with the higher Mn, C, and lower Al presented the highest liquation cracking susceptibility. Conversely, composition presenting lower Mn, C, and Al showed the most resistant behavior. Furthermore, lower Al content promoted a fully-γ microstructure at low temperatures, which encouraged the appearance of longer cracks as a γ-matrix is more susceptible to HAZ cracking than a fully ferritic (α) or duplex (α + γ) microstructure.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Design and Analysis of an Advanced Three-Point Bend Test Approach for Miniature Irradiated Disk Specimens

Candidate tungsten armor materials in a magnetic confinement fusion device must be able to withstand thermal variation that leads to internal stresses caused by the impinging heat load. In addition, the thermomechanical properties of these materials are degraded by irradiation-induced defect accumulation. Fission reactor–based irradiation data are used to predict the fusion neutron damage and property change. This study examines the motivation and design of a custom-designed three-point bend test for neutron-irradiated disk specimens that are 3 mm in diameter to be able to define the flexural strength of advanced tungsten materials, alloys, and composites—and to the extent that embrittlement occurs after neutron irradiation. The theory provided shows a calculation for the flexural deflection and shear deflection due to the small-geometry constraints. A finite element deformation analysis is performed to evaluate the mechanical stress field of disk bend specimens. The stress values above 80% of the maximum stress are concentrated in 2.4 mm of the 3.0-mm length of the centerline across the tungsten disk diameter. A bend test fixture has been designed and fabricated to enable testing of these specimens with precisely engineered tolerance and minimal machine compliance. This fixture will be able to be placed inside a universal testing frame at elevated temperatures for the mechanical property evaluation of future neutron-irradiated disk specimens.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Developing solid-surface plasma facing components for pilot plants and reactors with replenishable wall claddings and continuous surface conditioning. Part B: required research in present tokamaks

Abstract The companion part A paper (Stangeby et al 2022) reports a number of independent estimates indicating that high-duty-cycle DT tokamaks starting with pilot plants will likely experience rates of net erosion and deposition of solid PFC, plasma facing component, material in the range of 10 3 to 10 4 kg yr −1 , regardless of the material used. The subsequent redeposition of such large quantities of material has the potential for major interference with tokamak operation. Similar levels and issues will be involved if ∼continuous low-Z powder dropping is used for surface conditioning of DT tokamaks, independent of the material used for the PFC armor. In Stangeby et al (2022) (part A) it is proposed that for high-duty-cycle DT tokamaks, non-metallic low-Z refractory materials such as ceramics (graphite, SiC, etc) used as in situ replenishable, relatively thin—of order mm—claddings on a substrate which is resistant to neutron damage could provide a potential solution for protecting the main walls, while reducing the risk of degrading the confined plasma. Assessment of whether such an approach is viable will require information, much of which is not available today. Section 6 of part A identifies a partial list of major physics questions that will need to be answered in order to make an informed assessment. This part B report describes R&D needed to be done in present tokamaks in order to answer many of these questions. Most of the required R&D is to establish better understanding of low-Z slag generation and to identify means to safely manage it. Powder droppers provide a unique opportunity to carry out controlled studies on the management of low-Z slag in current tokamaks, independent of whether their protection tiles use low-Z or high-Z material.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗