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All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

NSTX-U liquid metal core-edge facility (LMCE)

NSTX-U/LMCE will provide a unique and world-leading research facility to address the primary challenge to delivering economic and timely magnetic fusion energy, namely the need to develop a power and particle exhaust and first-wall system that can withstand very high edge heat fluxes, maximize energy confinement, and avoid the production of large masses of solid eroded first-wall material. The NSTX-U/LMCE facility will assess the ability of liquid metals (LMs) – especially liquid lithium – to provide a new boundary condition for magnetic fusion systems, to extend the lifetime of the plasma facing components (PFCs) and improve core plasma confinement. Such capability is needed to establish the basis for next-step fusion facilities including fusion pilot plants, and to maintain U.S. world leadership in core-edge integration research. NSTX-U/LMCE will leverage the ability to generate very high divertor perpendicular heat flux q⊥ ~ 100MW/m 2 , extensive diagnostics, and liquid-metal-applicable infrastructure of NSTX-U. NSTX-U/LMCE will provide access to a high-confinement plasma core with majority self-driven plasma current, the flexibility to test a range of liquid metal divertor concepts, access to a range of separatrix collisionalities (from high to very low), and the ability to controllably vary the first-wall temperature to vary the plasma- wall interaction physics on liquid lithium components. Further, NSTX-U/LMCE will utilize more reactor-relevant high-Z refractory-metal PFC substrates. With these capabilities the NSTX-U/LMCE facility will explore the full continuum of core-edge solutions ranging from high core radiated power, to conditions with radiative losses concentrated in the scrape-off layer (SOL), and ultimately low recycling conditions. The low collisionality SOL that may be accessible in the low recycling regime is relatively unexplored and will require a kinetic treatment of the edge, which can be addressed theoretically, and with experiments in LTX-β. Additional smaller-scale preparatory R&D facilities will be required to reduce the risk of premature technical/engineering failure of liquid metal systems implemented in NSTX-U. The NSTX-U/LMCE facility aligns very well with recommendations in the FESAC Long-Range Plan and NASEM Pilot Plant reports and the Bold Decadal Vision, will be unique in the world program throughout the next decade, and is garnering private company interest in utilizing NSTX-U/LMCE for development of LM PFCs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

On a High-Mirror Stellarator Reactor Exploratory Concept With Neutrons Concentrated on Centrifuge Liquids

In the framework of fusion energy research based on magnetic confinement, stellarators allow numerous degrees of freedom for the design of the magnetic trap and plasma shape. Taking advantage of these features, some plasma shapes might benefit several of the many integrated elements involved in commercial fusion reactors, like, e.g., decreasing the number, mass and complexity of replaced activated in-vessel components (IVC) (i.e., by using liquids), extraction of large power, tritium generation, and remote maintenance. Certainly, free-surface liquid materials were proposed for tokamaks and field-reversed configuration (FRC) to try to improve some of such elements, i.e., in advanced power extraction (APEX) studies. Some reactor-relevant quasi-isodynamic (QI) magnetic configurations exhibit a relatively straight sector of plasma and high magnetic mirror. The combination of those elements and possibilities in a single stellarator reactor concept might have some advantages, in spite of the uncertainties due to the current low technological readiness level (TRL). The proposed and studied reactor concept is based on a vacuum vessel having short curved sectors and longer wide cylindrical sectors, which encloses a high-mirror low-vertical-excursion magnetic configuration, and swirling liquids or rotating cylinders, which centrifuge molten Li salts located at the low field region. Thus, the molten salts (if possible covered by a thin layer of liquid lithium) would be located on the internal perimeter of the cylinder, to act as particle exhaust (except for helium), neutron power extractor, and tritium breeder. The high-mirror feature tries to concentrate the neutron power at the cylindrical sectors, which might avoid using breeding materials at the curved sectors. The different elements of the concept are exploratorily studied and defined, and the difficulties assessed.

centrifuge↗
DOE OSTI2024

Controlled large-area lithium deposition to reduce swelling of high-energy lithium metal pouch cells in liquid electrolytes

Lithium (Li) metal battery technology, renowned for its high energy density, faces practical challenges, particularly concerning large volume change and cell swelling. Despite the profound impact of external pressure on cell performance, there is a notable gap in research regarding the interplay between external pressure and the electroplating behaviours of Li+ in large-format pouch cells. Here we delve into the impact of externally applied pressure on electroplating and stripping of Li in 350 Wh kg –1 pouch cells. Employing a hybrid design, we monitor and quantify self-generated pressures, correlating them with observed charge–discharge processes. A two-stage cycling process is proposed, revealing controlled pouch cell swelling of less than 10%, comparable to state-of-the-art Li-ion batteries. Finally, the pressure distribution across the cell surface unveils a complex Li+ detour behaviour during electroplating, highlighting the need for innovative strategies to address uneven Li plating and enhance Li metal battery technology.

25 ENERGY STORAGE↗
DOE OSTI2023

Active Lithium Injection for a Real Time Control of the Divertor Heat Flux for Fusion Devices

When the local heat flux exceeds specified flux limit, tungsten PFC surfaces can be damaged, which is not acceptable for a reliable reactor operations. The divertor PFCs are typically designed for a specific heat flux limit usually assuming an average steady-state heat flux which is typically 5–10 MW/m 2 . However, in addition to steady-state heat flux, fusion reactor divertor PFCs could experience transient heat fluxes such as ELMs and/or other magnetic reconnection events which can deposit large transient heat fluxes onto the divertor PFCs. The transient divertor heat flux could be significantly larger than the steady-state heat flux which could damage the solid PFC surfaces. The divertor heat flux can be subjected to additional complications such as the uncertainties in the the divertor strike point heat flux projection. Moreover, there are additional experimental observations of non-axisymmetric power flux which can occur under non-axisymmetric magnetic perturbations. The liquid lithium (LL) PFCs is more resilient against such transient heat fluxes as they could evaporate LL as needed and the lost LL can be then replenished afterward. In this paper, we analyze a case for a transient divertor heat pulse of 1 MJ in 10 ms for a ITER-size reactor. This is a small perturbation (~ 0.1%) to the expected plasma stored energy compared to the previously analyzed case of 20 MJ heat pulse. Even with this relatively modest heat pulse, the LL surface undergoes ~ 100 °C temperature rise. However, the resulting LL surface heating without rapid cooldown mechanism could lead to excessive LL evaporation continuing well after the transient heat flux resulting in a significant Li injection of ~ 0.6 mol in about a 200 ms period. This amount of Li injection could cause plasma dilution and performance degradation. On the other hand, an active Li injection capability if optimized could prevent the LL surface temperature rise and thus reducing subsequent Li evaporation into the plasma by a factor of 7 compared to the passive LL PFC case. A crucial tool of active Li injection is a rapid response pellet injector which could inject light impurity pellets before the excessive heat flux could reach the divertor plate causing serious damage. A simple pellet ablation model suggests a favorable pellet deposition profile for smaller ~ 0.1 mm radius pellet with ~ 10–20 m/s velocity. Moreover, if it is possible to inject from the private flux region, the pellet injection efficiency into the high heat flux strike point region can be as high as 80% compared to ~ 50% for the injection from outer radius region. The pellet deposition efficiency can be further improved by designing a shell-pellet which can burst when a certain ablation fraction is reached. Finally, a possible implementation technique using an inductive pellet injector with a rapid time response of a few msec is proposed here which can be tested in NSTX-U.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

INFUSE:FLARED (Final Technical Report)

This report summarizes the final outcomes of the INFUSE:FLARED collaboration between Tokamak Energy (TE), the University of Illinois Urbana-Champaign (UIUC), and the Department of Energy (DOE). The project aimed to experimentally determine and model the pumping, absorption, and separation behavior of deuterium in flowing liquid lithium under gas and plasma exposure conditions. The work supports the development of lithium-based plasma-facing components (PFCs) and tritium management strategies for next-generation fusion plants.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Optimization of lithium vapor box divertor evaporator location on NSTX-U using SOLPS-ITER

Commercial fusion reactors will be faced with extremely high divertor target heat fluxes that will require mitigation. Simulations of detachment in an NSTX-U scenario projected to have 92 MW/m 2 unmitigated peak target heat flux are presented, which reaches sub-10 MW/m 2 target heat flux using a highly dissipating lithium vapor box divertor design. The lithium vapor box is a detached divertor design which employs lithium vapor evaporation and condensation to contain lithium below the X-point. Previous SOLPS modeling has indicated a lithium vapor box can reduce the heat flux down to 10 MW/m 2 via simultaneous evaporation from the Private Flux Region (PFR) and the Common Flux Region (CFR) sides of the vapor box. It is found here that PFR evaporation has improved access to the separatrix leading to significantly more efficient power dissipation than CFR evaporation. Simulations of target evaporation with an evaporation distribution that is self-consistent with the temperature of a Capillary Porous System with Fast flowing liquid lithium (CPSF) could reach n Li / n e ~ 0.025-0.030 at the Last Closed Flux Surface (LCFS) depending on the liquid metal flow speeds and lithium sputtering yield, while PFR-side evaporation can reach acceptable heat fluxes with n Li / n e ~ 0.038 at the LCFS. However, PFR evaporator performance can be improved if the target is allowed to be hot enough such that it reflects lithium, reaching n Li / n e ~ 0.028 and reducing required lithium evaporation. Ultimately PFR evaporation and target evaporation are found to have similar ability to produce acceptable heat flux solutions with minimal upstream concentration.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Effects of negative triangularity on microinstabilities in a low-recycling lithium-wall spherical tokamak

In this work, we present a linear gyrokinetic study of the impact of negative triangularity (NT) on microinstabilities in the Lithium Tokamak eXperiment-β (LTX-β), a low-recycling spherical tokamak with liquid lithium plasma-facing components that produce flat electron-temperature profiles [Elliott et al., IEEE Trans. Plasma Sci. 48, 1382 (2020)]. While NT is widely recognized as a stabilizing mechanism and often associated with improved confinement in conventional tokamaks, this study reveals that its effect is not universally stabilizing in the parameter regime of LTX-β and is shown to be highly sensitive to local equilibria. Using local linear simulations with the GS2 code [Kotschenreuther et al., Comput. Phys. Commun. 88, 128 (1995)] at ρ=0.3, 0.5, and 0.8 for two representative discharges (#103955 and #109355), and employing the Miller equilibrium model to isolate shaping effects, we find that NT can transition from stabilizing to destabilizing depending on magnetic shear, safety factor, and electron-temperature gradient. In shot #103955, NT reduces growth rates across radii, with strongest stabilization at the edge, whereas in shot #109355, it is stabilizing in the core but destabilizing at mid-radius and edge under experimental conditions. Parametric scans show that flattening the electron-temperature profile, increasing magnetic shear, and reducing the safety factor recover NT stabilization. These results demonstrate that NT stabilization is tunable rather than intrinsic and requires coordinated control of magnetic geometry and gradient drive.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Savannah River National Laboratory – General Fusion 2023 INFUSE Report (Rev.1)

This report describes the results from an INFUSE research project, where Savannah River National Laboratory (SRNL) in collaboration with General Fusion (GF) used process modeling to understand and optimize commercial power plant (CPP) fuel cycle designs based on parameters provided by GF. The study primarily focused on two candidate fuel cycles with different blanket materials, one with a lead lithium eutectic (LLE) blanket and the other with a pure lithium (Li) blanket. LLE benefits from a low melting point, favorable neutronics, and lower reactivity, but liquid lithium has the potential for higher tritium breeding ratios (TBR) and does not poison the plasma as a high Z contaminant. It was found that the main differences between LLE and Li designs are the extraction technologies required to remove tritium from the blanket and the amount of tritium and its distribution within the facility. More than 80% of the in-process tritium inventory for the LLE design is contained in the isotope separation system, while for the Li design, over 60% of the in-process tritium inventory is contained within the blanket material. This is due to significant tritium retention by Li. For the Li blanket, the burden of tritium processing rests on the blanket extraction technology rather than the traditional exhaust processing route. Thus, the blanket extraction technology is a main driver of tritium inventory in the Li system and determines the subsequent interface with the tritium processing plant.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
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 OSTI2026

Impact of nitrogen on lithium-induced corrosion and cracking of RAFM steel

The emergence of private fusion enterprises, combined with advancements in U.S. and EU ITER-TBM as well as EU DEMO liquid breeder blanket concepts, has renewed focus on the materials challenges. This study examines the compatibility of reduced activation ferritic/martensitic (RAFM) steel in liquid Lithium (Li) containing various Nitrogen (N) concentrations, tested at 600 and 700 °C for 500 and 1000 h under static conditions. Microstructural characterization revealed the corrosion-induced dissolution and precipitation in Li containing 0.082 and 0.17 wt.% N at 600 °C, while limited dissolution and precipitation were observed in N-gettered Li using zirconium (Zr) and titanium (Ti) foils at 600 °C. Tensile testing at room temperature after Li exposure revealed a decrease in strength and ductility of the RAFM steel as a function of N content in Li. Severe embrittlement was observed after 500 h in Li containing 0.17 wt.% N at 600 °C and after 500 h in hot-gettered Li at 700 °C. Chromium (Cr) depletion from the steel matrix was observed, leading to the formation of M6C-type carbides on the surface identified by electron microscopy (SEM), X-ray diffraction (XRD), in agreement with thermodynamic calculations. These findings highlight the critical materials challenges posed by liquid Li breeder blanket concepts, emphasizing the need for improved strategies to address Li-induced dissolution, precipitation, and steel embrittlement due to N contamination of the coolant.

Romedenne, Marie [ORNL] (ORCID:0000000317936561)↗
DOE OSTI2023

Molecular Simulation of Lithium Carbonate Reactive Vapor–Liquid Equilibria Using a Deep Potential Model

We developed a first-principles machine learning model for the reactive vapor–liquid phase behavior of molten Li 2 CO 3 . The model was trained on ab initio electronic density functional theory data using the Deep Potential (DP) methodology, and its accuracy was evaluated by comparing model predictions of density and viscosity to experimental measurements. Direct coexistence simulations with the DP model over time scales of tens of nanoseconds were used to observe equilibrium dissociation of Li 2 CO 3 into CO 2 residing primarily in the vapor phase and Li 2 O which remains dissolved in the liquid. The simulations covered a range of temperatures, overall system sizes, and vapor-to-liquid volume ratios. Results were analyzed in terms of the observed chemical composition of the liquid and vapor phases, product structure, and CO 2 partial pressures. In addition, we calculated equilibrium constants for the dissociation reaction by assuming ideal-solution behavior for the liquid. As expected on the basis of thermodynamic arguments and prior experiments for this system, the observed partial pressure of CO 2 in the gas phase depends on both the temperature and the ratio of vapor to liquid volumes, while the calculated equilibrium constants only depend on temperature. DP model predictions for the equilibrium constant of the reaction are generally consistent with the available experimental measurements. Furthermore, the present study establishes the validity of the DP methodology for the description of reactive, multiphase equilibria from first principles, with possible applications to many other systems of scientific and technological interest even in the absence of relevant experimental measurements.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Recycling Measurements for the Lithium Tokamak Experiment

The lithium tokamak experiment (LTX) is a spherical torus designed to investigate a low-recycling wall regime for magnetically confined plasmas. Traditional measurements of recycling using Dα emission are not feasible in LTX due to reflections off the plasma-facing-components: stainless-steel shell, liquid lithium layer, and molybdenum limiters. Since lithium has low VUV reflectivity, LTX will use Lyman-α emission to measure particle recycling. LTX will use 10- and 20-channel AXUV diode arrays with Lyman-α filters to view recycling off the inboard and outboard shells. A single detector views recycling off a limiter. Measurements will be used with neutral modeling to calculate recycling and neutral deposition. To investigate the performance of an AXUV diode with directly-deposited filter, measurements were obtained with this detector and a visible detector viewing a calibration plasma then compared with emission calculated using a collisional-radiative model.

Physics - Plasma physics↗
DOE OSTI2026

Understanding NBI heating and fueling in LTX-β tokamak

My work on the contract DE-SC=0023274 (July, 2022 - July, 2025) was focused on understanding the plasma fueling by NBI. On Dec. 1998 with Sergei Krasheninnikov (UCSD, San Diego, CA) we initiated the Li Wall Fusion (LiWFusion) as a new concept of magnetic fusion [1]. The new 1 2 plasma 3 4 5 6 2.1 Overview of rate coefficients, mean-free paths λH0 , and diffusions coefficients 6 2. GSV code for ⟨σv⟩ analysis of NBI fueling concept was a reaction of the lack of luck of tokamak fusion with QDT = 1 on TFTR and JET. We recognized the edge plasma cooling by recycling to be the route reason of the tokamak problems on the way to burning plasma. LiWFusion relies on plasma pumping by a lithium layer on the inner walls of the plasma chamber, combined with the plasma heating and fueling by the Neutral Beam Injection (NBI). These two innovations eliminate the route problem of the tokmak fusion. The concept became theoretically mature in 2006. In 2012, the technology of continuously Flowing Liquid Lithium (24/7-FLiLi) was invented by me for the future implementations of the LWFusion concept. At the same time, my every presentation on LiWFusion to International Symposium on Lithium Applications since 2010 was objected by some person with words “Everybody since the 1970s knows that plasma fueling by NBI is impossible”. I dropped the name of the author of this objection and ignored his views. My work on the current grant on NBI fueling of LTX-β not only clarified the issue but resolved it in an astonishing way. The NBI fueling was fully understood (thus humiliating the current dogmatic fusion community). The new future of LTX-β dedicated for decades to Li in tokamaks, as well as of the entire magnetic fusion program was envisioned, in sharp contrast with the fallure of OFES in the post-TFTR era of 21st century.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

A transferable classical force field to describe glyme based lithium solvate ionic liquids

A non-polarizable force field for lithium (Li + ) and bis(trifluoromethanesulfonyl)imide (TFSI – ) ions solvated in diglyme at around 0.2 mol fraction salt concentration was developed based on ab initio molecular dynamics (AIMD) simulations and a modified polymer consistent force field model. A force–torque matching based scheme, in conjunction with a genetic algorithm, was used to determine the Lennard-Jones (LJ) parameters of the ion–ion and ion–solvent interactions. This force field includes a partial charge scaling factor and a scaling factor for the 1–4 interactions. The resulting force field successfully reproduces the radial distribution function of the AIMD simulations and shows better agreement compared to the unmodified force field. The new force field was then used to simulate salt solutions with glymes of increasing chain lengths and different salt concentrations. The comparison of the MD simulations, using the new force field, with experimental data at different salt concentrations and AIMD simulations on equimolar concentrations of the triglyme system demonstrates the transferability of the force field parameters to longer glymes and higher salt concentrations. Furthermore, the force field appears to reproduce the features of the experimental x-ray structure factors, suggesting accuracy beyond the first solvation shell, for equimolar salt solutions using both triglyme and tetraglyme as the solvent. Altogether, the new force field was found to accurately reproduce the molecular descriptions of LiTFSI-glyme systems not only at various salt concentrations but also with glymes of different chain lengths. Thus, the new force field provides a useful and accurate tool to perform in silico studies of this family of systems at the atomistic level.

25 ENERGY STORAGE↗
DOE OSTI2026

Upgrades to the Thomson Scattering Diagnostic on the Lithium Tokamak eXperiment-Beta (LTX-$β$)

We describe the upgrades to the Thomson scattering (TS) diagnostic on the Lithium Tokamak eXperiment-beta (LTX- β ) to provide complete radial coverage of the plasma. LTX- β is a spherical tokamak with plasma limited at a major radius of R = 14 cm and R = 66 cm. Flat temperature profiles with peaked density were observed in LTX- β using an existing multipoint TS diagnostic with radial coverage of the plasma core. There is a need for more comprehensive core and edge TS measurements to thoroughly examine beam fueling, beam heating, and transport under low recycling conditions with liquid lithium walls and a lower collisionality scrape-off layer (SOL). The core TS set up provides plasma profiles from R = 40–60 cm. To extend the TS diagnostic coverage: 1) we added a TS viewing optics to cover the SOL region at five radial points from R = 57–64 cm, and 2) another TS viewing optics in the HFS region from R = 15–45 cm to obtain TS data from ten radial positions while making use of the original laser beam. The main focus of this article is on the new HFS TS setup that is equipped with better-matched collection optics located in the horizontal midplane where the laser polarization lies and views from a scattering angle of 130°–150° with respect to the laser path. A set of unequal-length Y-branched fiber bundles transports the TS light from two radial points to each polychromator. The new HFS view is better matched in spot size, scattering angle, and throughput to the laser and polychromators and has new V-grooved style viewing dumps. The new HFS TS view, along with the core and SOL TS views, will provide a full radial measurement coverage from R = 15–64 cm to explore the flat temperature profiles.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Design and Analysis of the Open-Surface Slow Li Flow Divertor and Comparison to the Fast Li Flow Divertor

In the ongoing U.S. project, “Liquid Metal Plasma Facing Components,” sponsored by the U.S. Department of Energy, efforts have been taken to develop two open-surface divertor designs for the Fusion Nuclear Science Facility using liquid lithium (Li) as a heat and particle flux removal media. The main focus of this study is the design and analysis of a slow (~1 mm/s) and thin (<1 mm) open-surface Li flow divertor with a Li-cooled substrate, which is then compared with an earlier design of a fast (up to 10 m/s) and thick (~0.5 cm) Li flow divertor with the substrate cooled with helium. The slow Li flow divertor design is based on the original LiWall concept developed at the Princeton Plasma Physics Laboratory. Such a thin and slow Li layer can remove the particle flux by reducing the recycling flux, while the heat flux is removed mainly through the heat sink located beneath. In the present study, the heat sink is provided through a Li cooling flow inside the substrate of reduced activation ferritic/martensitic steel. By performing a multiphysics analysis with COMSOL that included liquid-metal magnetohydrodynamics (MHD), heat transfer, and structural mechanics, the impact of various factors on the divertor heat removal capability, such as Li flow velocity, MHD effects, and inlet velocity boundary condition, were examined. Based on comparisons of the two divertor designs, it was shown that the fast-flow divertor significantly outperformed the slow-flow design, whose heat removal capability was limited to ~1 to 2 MW/m2.

Jiang, Yuchen [ORNL] (ORCID:0000000283280284)↗
DOE OSTI2023

Understanding Biases in Sample Preparation Techniques for Coupled Scanning Electron Microscopy and MAMA PuO 2 Morphological Analysis

In this project, the scanning electron microscopy (SEM) sampling method used during the statistical design study (SDS) was investigated to determine if any sampling biases were present in the analyzed data. Using standard particle size distribution powders from the National Institute of Standards and Technology (NIST 1984 standard reference material) with the origin wet dispersion method, it was determined that a bias to smaller particles was present. This was supported by theoretical calculations using Stokes’ law to determine the settling rate of spherical particles of roughly the same size and mass as those found in the SDS. Based on the theoretical calculations, it was determined that the settling rate for each of the 76 powder sets in the SDS could be unique based on specific particle shape and mass distributions, making a universal correction factor/formula not applicable. Therefore, priority shifted to developing an improved wet dispersion method that significantly reduced the particle settling rate for all particle size and shapes. This was achieved by replacing the original solvent (isopropyl alcohol) with a heavy liquid (lithium heteropolytungstates), which dramatically slowed the settling rate and allowed for the capture of a suitable homogeneous aliquot. SEM imaging and Morphological Analysis for Material Attribution (MAMA) software analysis were conducted on the NIST standard, and the SEM/MAMA data were compared to data captured by a dynamic image analysis particle size analyzer. The resulting data confirmed that the new wet dispersion method does indeed deliver an improved representative aliquot to the SEM stub. For instance, in the NIST certificate, the average particle size is ~17.1 µm ± 2.2 µm with a normal distribution. The initial wet dispersion method resulted in a drastically reduced average particle size of 6.1 µm in addition to a non-representative heavy bi-modal distribution whereas the improved LST wet dispersion method resulting in an average particle size that was much closer to the NIST certificate (12.7 µm) with a similar normal distribution. Although the improved method was still short of the NIST certificate average, atomic force microscopy analysis determined that the resulting ~20-25% reduction in size was due to particles sinking into the carbon sticky tape used for SEM imaging. It is believed that that this bias can be calibrated in a much more predicable manner than the original settling rate bias. In addition, the matching normal distribution curves between the NIST certificate and the heavy liquid method indicate a much-improved representative aliquot has been sampled and imaged. A surrogate CeO 2 powder was used to reflect PuO 2 more accurately and to aid in implementing radiological controls and shielding. The resulting data sets from the SEM/MAMA method and the particle size analyzer give almost identical average particle sizes and particle distribution statistics. Future work will re-analyze several select runs from the SDS to determine if morphological signatures can be found with the improved sampling method.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

Deuterium retention in pre-lithiated samples and Li–D co-deposits in the DIII-D tokamak

Divertor designs involving liquid lithium have been proposed as an alternative to solid designs and wall conditioning techniques. However, Li affinity with tritium poses a risk for the fuel cycle. This study investigates deuterium retention in pre-lithiated samples and Li–D co-deposits in the DIII-D tokamak, making for the first time a direct comparison between Li–D co-deposits and pre-deposited Li films. Samples were exposed to H-mode plasmas in the far scrape-off layer (SOL), and Li powder was injected in-situ with the impurity powder dropper to study the uniformity of Li coatings, and the dependence of fuel retention on Li thickness. The results show that at temperatures below the melting point of lithium, deuterium retention is independent of the thickness of pre-deposited Li layers, with Li–D co-deposits being the primary factor for fuel retention. Both pre-deposited and in-situ deposited Li showed lower erosion than predicted by sputtering yield calculations. These results suggest that fuel retention in fusion reactors using lithium in the divertor will likely be dominated by co-deposits rather than in the divertor itself. If one desires to use Li to achieve flatter temperature profiles, operando Li injection is advantageous over pre-deposited Li films, at least at temperatures below the melting point of lithium.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗