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At least 199 records · Page 11
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Thermal and structural analysis of a new WEST reciprocating probe head with Titanium-Zirconium-Molybdenum armour material

The Tore Supra tokamak was transformed into an X-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support of the ITER tungsten divertor strategy. WEST began operation in 2017. A new reciprocating probe head, equipped with a Titanium-Zirconium-Molybdenum (TZM) heat shield, was recently built and operated on WEST during the C9 Campaign in early 2024. TZM was chosen as armour material for its very good mechanical properties between 700 °C and 1400 °C. This material was already extensively used for probe heads and other plasma facing components on the Alcator C-Mod tokamak. Here, this paper presents the design of the new WEST probe head. It is equipped with tungsten filaments that can be heated to emit electrons and provide a direct measurement of the plasma potential in the scrape-off layer, and it is also equipped with a pair of Langmuir probes. The probe plunges vertically into the plasma, reaches its maximum dive depth, then returns to its protected resting position in about 0.25 s. Thermal loads are calculated using as input the thermal properties of TZM and radial profiles of the heat flux measured in the past by other reciprocating probes, and scaled upwards to the expected heat loads in the WEST scrape-off layer. An analytical estimation of Electromagnetic (EM) loads will also be presented, as well as a thermal and structural analysis of the probe for 15 cycles at maximum plunge depth. The probe is observed by an infrared camera. Comparisons between the heat flux deduced by the Langmuir probe and the infrared radiance of the probe housing during plasma exposure are finally presented.

Diagnostic↗
DOE OSTI2025

High-Power Impulse Magnetron Sputter Deposition of Boron Carbide with Full-Face Erosion Magnetron and Mixed Ar-Ne Plasma

Boron carbide (B 4 C) is an attractive inertial confinement fusion ablator material. The fabrication of B4C ablators by magnetron sputtering requires process optimization. To increase process flexibility, here we explore high-power impulse magnetron sputter (HiPIMS) deposition of B 4 C in pure Ar and mixed Ar-Ne plasmas. Here, the results show that higher plasma discharge currents can be reached with a mixed Ar-Ne plasma in the entire working pressure range studied (5 𝑡𝑜 50 mTorr). At 45 mTorr with 10% of Ne in the Ar-Ne mix, high peak target current densities of ~1 A cm −2 were demonstrated. Films deposited with such a mixed Ar-Ne plasma with a full-face erosion magnetron source on substrates biased at −25 V exhibited higher density and improved mechanical properties, albeit with higher compressive residual stresses compared to the case of HiPIMS deposition in a pure Ar plasma. This work demonstrates additional process flexibility of the HiPIMS discharge mode for the deposition of B 4 C coatings.

Ablator capsule↗
DOE OSTI2022

A process optimization framework for laser direct energy deposition: Densification, microstructure, and mechanical properties of an Fe-Cr alloy

Laser Direct Energy Deposition (DED) is a metal additive manufacturing technique with the ability to fabricate large and complex parts through deposition of metal powders. However, achieving high-density parts and targeted build heights using DED can be challenging due to the large number of highly sensitive process variables. This work proposes a robust fabrication parameter optimization framework to generate process maps for primary parameters in DED, including laser power, scan speed, mass flow rate, hatch spacing, and layer height. Simple single-track experiments were utilized to map out the parameter space, and a combination of geometric criteria for hatch spacing and layer height were proposed to determine parameter sets that achieve both targeted build heights and mitigate porosity formation. Using this framework, specimens with >99 % density and consistent mechanical properties were successfully fabricated over a wide range of process parameters for an Fe-9wt.%Cr (Fe9Cr) alloy, a surrogate for radiation damage-resistant reduced activation ferritic/martensitic (RAFM) steels. Processing these materials using DED is of particular interest in the development of plasma facing components for nuclear fusion applications. The microstructure and mechanical properties of as-printed Fe9Cr were characterized using optical and electron microscopy, X-ray diffraction, and uniaxial tensile tests. As-printed Fe9Cr displayed ~25 % elongation and ultimate tensile strengths of up to 475 MPa which is comparable to similar wrought alloys. Finally, the proposed framework will allow for accelerated DED parameter optimization for novel alloy systems, as well as open the possibility for local microstructure control while simultaneously mitigating defect formation.

42 ENGINEERING↗
DOE OSTI2023

The comparative strength and fracture toughness properties of commercial 95W-3.5Ni1.5Fe and 95W-3.5Ni1.5Cu tungsten heavy alloys

Due to the sharp surface cracks resulting from high and cyclic thermal loads, fracture toughness is a crucial property for fusion divertor materials. Tungsten heavy alloys (WHA) have emerged as a leading candidate for plasma facing component applications. Both the WNiFe and WNiCu WHAs have been considered. However, there is no fracture toughness data in the literature on the WNiCu WHA. Therefore, maximum load elastic–plastic fracture toughness (K Jm ) of a commercial 95W-3.5Ni1.5Cu (NiCu) WHA plate is reported here for the first time. The NiCu WHA K Jm is compared to a corresponding K Jm for commercial 95W-3.5Ni1.5Fe (NiFe) WHA, along with comparisons of their respective microstructures, microhardness, and tensile properties. Although their room temperature yield strengths are similar (≈ 600 MPa), the NiFe WHA has higher ultimate stress (818 MPa vs. 642 MPa) and total elongation (8% vs. 1.2%) compared to the NiCu WHA. Most notably, the room temperature elastic–plastic K Jm for the NiFe WHA of ≈ 80 ± 8 MPa$\sqrt{m}$, is nearly twice that of the NiCu WHA at ≈ 47 ± 4 MPa$\sqrt{m}$. Further, while the NiFe WHA experienced stable ductile tearing, the crack propagated unstably (approximately elastically) after reaching the maximum load in the NiCu WHA. The higher K Jm of the NiFe WHA is attributed to its higher flow stress and ductility, along with the energy dissipation and microcrack dilatational stress shielding, provided by a wide distribution of arrested process-zone microcracks, which are largely absent in the NiCu WHA. Since, the 95W-NiFe WHA has distinctly better mechanical properties than 95W-NiCu WHA, it is a more suitable candidate for the fusion reactor application.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Upgrade of the Materials Analysis Particle Probe (MAPP-U) to decipher the impact of lithium-based surfaces on NSTX-U plasma behavior

Understanding the plasma wall interaction (PWI) remains a critical issue for the feasibility of thermonuclear magnetic fusion energy solutions. Key issues with PWI mechanisms in fusion tokamak reactors include: evolution of surface chemistry and its role on hydrogen retention. In particular how low-Z coatings such as lithium can impact the behavior of plasma at the edge and in the core. PMI (plasma-material interactions) are particularly important for strategies that involve low-recycling regimes and the use of lithium PFS (plasma-facing surfaces) to attain them, as in the case of NSTX-U. Recent reports have indicated the importance of access to the evolving plasma-facing surface during and in-between plasma discharges. Changes in surface chemistry and morphology due to ion bombardment and the difficulty of diagnosing plasma-facing surfaces, especially reactive surfaces, complicate the development of a predictive understanding of the wall and its interaction with the plasma. Consequently, this impairs the ability to design advanced PFC materials for future plasma-burning fusion reactors and appropriate PMI code validation. The Materials Analysis Particle Probe (MAPP) is an established and on-going PMI probe diagnostic system compatible with the highly chemically reactive system of lithium and boron coatings adopted by the NSTX-U research program. MAPP is the first PMI diagnostic to capture the surface physics and chemistry in-vacuo in a fusion tokamak system and correlate this data to controlled plasma shots. Currently MAPP captures this information at a fixed radial location at the NSTX-U outboard divertor region. The MAPP diagnostic has enabled understanding of the near-surface and surface chemistry of complex evolving lithiated and borated carbon-based PFC surfaces retention and transport of hydrogen. Coupled to atomistic simulations in collaboration with P. Krstic of Stony Brook U. MAPP has been very successful in achieving high-impact scientific research in its current grant period evidenced by two invited review articles and over 20 peer-reviewed manuscripts and over 40 contributed and invited presentations at both national and international conferences.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
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 OSTI2024

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.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
NASA NTRS2001

An Experimental Investigation of the Effects of Charging on the International Space Station

An experimental investigation has been undertaken to determine the effects associated with operating the International Space Station (ISS) without a Plasma Contactor Unit (PCU). The role of the PCU is to maintain the potential of the ISS to within 40 volts of the ambient plasma potential. In the event of a PCU failure, ISS structure may charge to - 160 volts with respect to the Low Earth Orbit (LEO) plasma, due to the use of high voltage photovoltaic solar arrays. Operation without a PCU will result in the charging of the oxide layers on anodized aluminum components facing into the RAM direction. In this investigation, arcs were observed as a result of anodized materials charging in a plasma environment. Actual ISS materials were used in the investigation. The materials included meteoroid and debris shield samples, as well as components from the Extravehicular Mobility Unit (EMU). Results show that the occurrence of arcs was dependent on several factors including material composition and applied voltage. The breakdown thresholds for the materials tested will be shown, as well as the voltage and current waveforms during an arc event. The damage resulting from an arc will also be described. In addition, a description of associated operational hazards resulting from arcing on the spacecraft will be provided.

Schneider, Todd↗
DOE OSTI2025

The interaction of edge dislocations with hydrogen-helium bubbles in tungsten

In fusion reactors, plasmas-facing components undergo degradation due to displacement damage and injection of gas impurities. Significant amounts of hydrogen (H) and helium (He) impurities can be introduced into materials through plasma exposure and nuclear transmutation, and their synergistic effects lead to the formation of mixed He/H bubbles under irradiation, changing their hardening mechanisms. In this study, molecular dynamics (MD) models were developed to study the interaction between ½<111> edge dislocations with mixed He/H-bubbles in W. The critical resolved shear stress (CRSS) that is required for dislocation breakaway increases with He/H-to-vacancy ratio at 600 K, indicating strong pinning effects of mixed He/H bubbles. However, as the temperature increases up to 1400 K, the mixed He/H bubbles become unstable as the H atoms are increasingly emitted from the mixed He/H bubbles and migrate into W matrix, which lowers bubble pressure and CRSS. Overall, these results highlight the synergistic effects of He and H on dislocation behavior in irradiated W during deformation.

36 - MATERIALS SCIENCE↗
DOE OSTI2020

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.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Influence of the inverse sheath on divertor plasma performance in tokamak edge plasma simulations

Recently, it was shown that strong electron thermionic emission from material walls could result in the formation of an “inverse sheath”, which prevents the flow of cold ions to the wall. Such regimes look very favorably from the point of view of plasma-material interactions at the edge of magnetic fusion devices, where the problem of the erosion of plasma facing components under ion irradiation is one of the key issues for developing of future magnetic fusion reactors. However, it is not clear whether such regimes are compatible with edge plasma parameters and heat removal requirements in fusion reactors.To address the issue of practicality of the “inverse sheath” regime for edge tokamak plasma conditions, we perform a set of numerical simulations with 2D edge plasma transport code UEDGE for a DIII-D-like geometry and magnetic configuration.To describe both “standard” and “inverse sheath” conditions within the framework of the UEDGE code (which does not consider the sheath region per se), at the material surfaces, we apply effective boundary conditions which emulate both “standard” and“inverse sheath” regimes. We demonstrate that for the same input parameters, spatial distributions of edge plasma parameters corresponding to detached divertor and“inverse sheath” regimes are similar, with only a few minor differences. We discuss the compatibility of “inverse sheath” regimes with core plasma parameters.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Dependence of Heat Removal Rate of Pebble-Based Rods on Inter-Pebble Matrix Fill Fraction

Carbon pebble rods are a promising candidate for use in high heat flux regions of magnetic fusion energy reactor walls. Under high (10 – 50 MW/m 2 ) heat loads, carbon pebble rods release hot pebbles from the exposed surface, carrying away heat as the pebble rod surface recedes. In this work, we show that the surface recession rate during heating can be adjusted by changing the mechanical strength of the extruded rods, modifying the heat removal rate; this is accomplished here by varying the fill fraction of the inter-pebble matrix. A three-dimensional finite element model is presented that captures many experimental observations, including the sphere temperature and the surface recession rate. The model predicts that pebble release is caused by thermally driven crack propagation through the matrix and that the matrix strength against breaking is the single most important material parameter setting the pebble release rate; this prediction is supported by experimental results.

36 MATERIALS SCIENCE↗
NASA NTRS2001

Large-Scale Mini-Magnetosphere Plasma Propulsion (M2P2) Experiments

Mini-Magnetosphere Plasma Propulsion (M2P2) is an innovative plasma propulsion system that has the potential to propel spacecraft at unprecedented speeds of 50 to 80 km/s, with a low power requirement of approx. 1 kW per 100 kg of payload and -1 kg of neutral gas [fuel] consumption per day of acceleration. Acceleration periods from several days to a few months are envisioned. High specific impulse and efficiency are achieved through coupling of the spacecraft to the 400 km/s. solar wind through an artificial magnetosphere. The mini-magnetosphere or inflated magnetic bubble is produced by the injection of cold dense plasma into a spacecraft-generated magnetic field envelope. Magnetic bubble inflation is driven by electromagnetic processes thereby avoiding the material and deployment problems faced by mechanical solar sail designs. Here, we present the theoretical design of M2P2 as well as initial results from experimental testing of an M2P2 prototype demonstrating: 1) inflation of the dipole magnetic field geometry through the internal injection of cold plasma; and 2) deflection of and artificial solar wind by the prototype M2P2 system. In addition, we present plans for direct laboratory measurement of thrust imparted to a prototype M2P2 by an artificial solar wind during the summer of 2001.

Winglee, R. M.↗
NASA NTRS2001

Large-Scale Mini-Magnetosphere Plasma Propulsion (M2P2) Experiments

Mini-Magnetosphere Plasma Propulsion (M2P2) is an innovative plasma propulsion system that has the potential to propel spacecraft at unprecedented speeds of 50 to 80 km per second with a low-power requirement of approx. 1 kW per 100 kg of payload and approx. 1 kg of neutral gas [fuel] consumption per day of acceleration. Acceleration periods from several days to a few months are envisioned. High specific impulse and efficiency are achieved through coupling of the spacecraft to the 400 km per second solar wind through an artificial magnetosphere. The mini-magnetosphere or inflated magnetic bubble is produced by the injection of cold dense plasma into a spacecraft-generated magnetic field envelope. Magnetic bubble inflation is driven by electromagnetic processes thereby avoiding the material and deployment problems faced by mechanical solar sail designs, Here, we present the theoretical design of M2P2 as well as initial results from experimental testing of an M2P2 prototype demonstrating: 1) inflation of the dipole magnetic field geometry through the internal injection of cold plasma; and 2) deflection of and artificial solar wind by the prototype M2P2 system. In addition, we present plans for direct laboratory measurement of thrust imparted to a prototype M2P2 by an artificial solar wind during the summer of 2001.

Winglee, R. M.↗
DOE OSTI2020

Perspectives on the FESAC transformative enabling capabilities: Priorities, plans, and Status

In early 2017, the Fusion Energy Sciences Advisory Committee (FESAC), an advisory committee to the United States Department of Energy, was charged with identifying transformative enabling capabilities (TECs) “that could promote efficient advance toward fusion energy, building on burning plasma science and technology.” A subcommittee with broad expertise was formed and sought feedback from scientific experts, including experts from outside the fusion community. Three workshops were conducted, and a report was approved by FESAC in 2018 that identified four of the “most promising” TECs: advanced algorithms, high-critical-temperature superconductors, advanced materials and manufacturing, and novel technologies for tritium fuel cycle control. In addition, one “promising” TEC was identified: fast flowing liquid metal plasma-facing components. This paper will give details on the promising TECs and an overview on considerations of these TECs in the United States since the publication of the report.

Lumsdaine, Arnold↗
DOE OSTI2025

Improved liquid lithium surfaces in the Lithium Tokamak Experiment-β

Advances in vacuum, surface, and lithium conditioning techniques throughout five years of continuous operations in LTX-β have produced mirror-like liquid lithium surfaces and demonstrated the feasibility of high-performance tokamak discharges fully surrounded by liquid metal without significant operational problems. Improvements in conditioning techniques and procedures, including many weeks of baking and accumulation of 70 g of Li, led to reduced residual gasses and clean Li surfaces - all while still maintaining enough operational flexibility for multiple in-vacuum diagnostic upgrades and calibrations. Coatings had a visibly clean appearance, with reflective liquid metal demonstrating good wetting and surface adhesion with films that were now macroscopically thick. Solidified Li showed large crystal grains, while surface science measurements observed reduced impurities in the lithium. Steadily improved plasma performance was achieved with liquid lithium, with discharges able to match solid Li in terms of evolution of I p and n e , including rapid density pumping indicating low recycling. There were indications of moderately increased Li impurity influx, though few significant disturbances by the large liquid surfaces on tokamak operations over hundreds of discharges. Liquid metal plasma facing components are a potential solution to the extreme heat and particle fluxes that could cause unacceptable damage to solid materials, while liquid lithium also has the potential for greatly increased confinement in the low-recycling regime. While many liquid metal approaches are possible, and numerous experiments have been conducted in test stands and small modules in fusion devices, LTX-β is the only tokamak operated while fully surrounded by liquid metal.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Recrystallization suppression through dispersion-strengthening of tungsten

Tungsten is the material of choice for the divertor region of future nuclear fusion reactors, an environment that will expose plasma-facing components (e.g. divertor, etc...) to high temperatures and transient high heat flux events. Under these conditions, recrystallization and grain growth of tungsten can occur, leading to undesirable microstructural and mechanical property changes. Therefore, there is a need to raise the recrystallization temperature of tungsten and limit the kinetics of the recrystallization and grain growth processes. In this work, we examine the effect of different types (TiC vs. TaC vs. ZrC) and different concentrations (1.1 vs. 5 vs. 10 wt.%) of dispersed second phase particles in a tungsten matrix on the high temperature performance. Additionally, the addition of second-phase particles effectively increases the temperature of and time for recrystallization and slow grain growth; however, the addition of a high weight fraction of particles alters the surface chemistry, which may impact subsequent plasma-surface interactions. These results show that the addition of small concentrations of dispersed particles can be effectively employed in tungsten to raise the upper operating temperature limit for tungsten in a fusion reactor.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Reconfiguration of an Electrothermal-Arc Plasma Source for In Situ PMI Studies

An electrothermal-arc plasma source (ET-Arc) has been developed to produce transient plasma heat and particle fluxes similar to those produced by edge localized modes onto divertor plasma-facing components in tokamaks. The ET-Arc utilizes a capacitive discharge to send current through a 4-mm-diameter, 9-cm-long capillary source liner. The liner material is ablated to form a high-velocity plasma jet that impacts the target downstream. With the current discharge circuit configuration, pulse lengths are 1 to 2 ms in duration and deliver heat fluxes of 0.25 to 2.1 GW m-2. The plasma was previously characterized with optical emission spectroscopy (OES) on helium emission lines. The He I line ratios were interpreted with collisional radiative analysis to calculate ne and Te. The electron temperature and electron density ranged from Te = 1 to 5 eV and ne = 1022 to 1028 electrons/m3, respectively.

Lindquist, Elizabeth G.↗