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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 379 records · Page 21
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Estimating cross-field particle transport at the outer midplane of TCV by tracking filaments with machine learning

Abstract Cross-field transport of particles in the boundary region of magnetically confined fusion plasmas is dominated by turbulence. Blobs, intermittent turbulent structures with large amplitude and a filamentary shape appearing in the scrape-off layer (SOL), are known from theoretical and experimental studies to be the main contributor to the cross-field particle transport. The dynamics of blobs differs depending on various plasma conditions, including triangularity ( δ ). In this work, we analyze triangularity dependence of the cross-field particle transport at the outer midplane of plasmas with δ = + 0.38 , +0.15, −0.14, and −0.26 on the Tokamak à Configuration Variable, using our novel machine learning (ML) blob-tracking approach applied to gas puff imaging data. The cross-field particle flux determined in this way is of the same order as the overall transport inferred from KN1D, GBS, and SOLPS-ITER simulations, suggesting that the blobs identified by the ML blob-tracking account for most of the cross-field particle transport in the SOL. Also, the ML blob-tracking and KN1D show a decrease in the cross-field particle transport as δ becomes more negative. The blob-by-blob analysis of the result from the tracking reveals that the decrease of cross-field particle transport with decreasing δ is accompanied by a decrease in the number of blobs in a fixed time, which tend to have larger area and lower radial speed. Also, the blobs in these plasmas are in the connected sheath regime, and show a velocity scaling consistent with the two-region model.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

The role of plasma–atom and molecule interactions on power & particle balance during detachment on the MAST Upgrade Super-X divertor

This paper shows first quantitative analysis of the detachment processes in the MAST Upgrade Super-X divertor (SXD). We identify an unprecedented impact of plasma-molecular interactions involving molecular ions (likely $D^+_2$), resulting in strong ion sinks (Molecular Activated Recombination—MAR), leading to a reduction of ion target flux. The MAR ion sinks exceed the divertor ion sources before electron-ion recombination (EIR) starts to occur, suggesting that significant ionisation occurs outside of the divertor chamber. In the EIR region, $T_e \ll 0.2$ eV is observed and MAR remains significant in these deep detached phases. The total ion sink strength demonstrates the capability for particle (ion) exhaust in the Super-X Configuration. Molecular Activated Dissociation is the dominant volumetric neutral atom creation process can lead to an electron cooling of 20% of $P_\mathrm{SOL}$. The measured total radiative power losses in the divertor chamber are consistent with inferred hydrogenic radiative power losses. This suggests that intrinsic divertor impurity radiation, despite the carbon walls, is minor in the divertor chamber. This contrasts previous TCV results, which may be associated with enhanced plasma-neutral interactions and reduced chemical erosion in the detached, tightly baffled SXD. The above observations have also been observed in higher heat flux (narrower SOL width) type I ELMy H-mode discharges. This provides evidence that the characterisation in this paper may be general.

plasma-molecular interactions↗
DOE OSTI2023

Exploration of non-resonant divertor features on the Compact Toroidal Hybrid

Abstract Non-resonant divertors (NRDs) separate the confined plasma from the surrounding plasma facing components (PFCs). The resulting striking field line intersection pattern on these PFCs is insensitive to plasma equilibrium effects. However, a complex scrape-off layer (SOL), created by chaotic magnetic topology in the plasma edge, connects the core plasma to the PFCs through varying magnetic flux tubes. The Compact Toroidal Hybrid (CTH) serves as a test-bed to study this by scanning across its inductive current. Simulations observe a significant change of the chaotic edge structure and an effective distance between the confined plasma and the instrumented wall targets. The intersection pattern is observed to be a narrow helical band, which we claim is a resilient strike line pattern. However, signatures of finger-like structures, defined as heteroclinic tangles in chaotic domains, within the plasma edge connect the island chains to this resilient pattern. The dominant connection length field lines intersecting the targets are observed via heat flux modeling with EMC3-EIRENE. At low inductive current levels, the excursion of the field lines resembles a limited plasma wall scenario. At high currents, a private flux region is created in the area where the helical strike line pattern splits into two bands. These bands are divertor legs with distinct SOL parallel particle flow channels. The results demonstrate the NRD strike line pattern resiliency within CTH, but also show the underlying chaotic edge structure determining if the configuration is diverted or limited. This work supports future design efforts for a mechanical structure for the NRD.

Physics↗
DOE OSTI2024

Interpretive modeling of tungsten divertor leakage during experiments with neon gas seeding

Abstract Many existing and future tokamaks with tungsten divertors operate, or will operate, with low- Z impurity seeding, but the direct effect of these seeded impurities on tungsten Scrape-off-Layer (SOL) transport has not been explored in detail. This paper reports on a DIII-D experiment designed to test how tungsten divertor leakage from the Small-Angle Slot V-Shaped, tungsten-coated divertor is impacted by neon seeding at a variety of injection rates and poloidal injection locations. Measurements from the experiment show an inverse relationship between the neon injection rate and the tungsten core penetration factor. Interpretive modeling is performed with a combination of the SOLPS-ITER and DIVIMP codes to assess the underlying tungsten behavior. The modeling results show that the reduction in tungsten divertor leakage is driven by both an increase in the divertor collisionality as well as a reduction in the ion temperature gradient near the divertor target. Collisions between low- Z impurities and tungsten impurities are found to have a significant impact on the tungsten SOL transport, such that ignoring the low- Z impurity collisional effects on the tungsten transport can result in an overestimate of the divertor leakage by an order-of-magnitude. Given the importance of these localized interactions, neon seeding from the closed, slot-like divertor has a clear advantage in being able to reduce tungsten divertor leakage without the high levels of neon core contamination that occur when seeding from other poloidal locations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Density fluctuation statistics and turbulence spreading at the edge of L–mode plasmas

Long-wavelength density fluctuations ($k{\rho _i}$ <1) are studied using beam emission spectroscopy (BES) at the edge of DIII-D L-mode plasmas (ρ = 0.88–1.1) in scenarios with electron cyclotron heating (ECH) power ramp (P ECH up to 1.5 MW), neutral beam injection (NBI) power ramp (P NBI up to 2.5 MW), and injected torque scan (-1 < T inj <0.6 Nm). We find that broadband turbulent density fluctuations (ƒ ~ 20–120 kHz) have a non-Gaussian distribution. The skewness of $\delta n/n$ changes sign from negative at ρ < 0.95–0.97 to positive at ρ > 0.97, indicating the prevalence of density 'voids' at inner radii and density 'blobs' at outer radii and outside of the separatrix. The turbulence intensity flux $\left\langle {{{\tilde v}_{\text{r}}}{{\tilde n}^2}} \right\rangle$ is calculated to characterize turbulence spreading at the plasma edge. During ECH/NBI power ramps and at counter-I p injected torque, $\left\langle {{{\tilde v}_{\text{r}}}{{\tilde n}^2}} \right\rangle$ is directed inward inside the separatrix, which is evidence of inward spreading of turbulence intensity from the edge gradient region caused by the inner propagation of density 'voids'. Significantly weaker $\left\langle {{{\tilde v}_{\text{r}}}{{\tilde n}^2}} \right\rangle$ is observed with co-I p torque. A correlation between co-I p torque, turbulence intensity $\delta n/n$ at ρ = 0.97, and increased srape-off layer (SOL) heat flux decay length ${\lambda _q}$ is found in the torque scan scenario, showing that edge turbulence plays a material role in determining the SOL conditions and heat flux width.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Impact of ionization peak location on measured opaqueness in DIII-D H-mode plasmas

This study investigates the relationship between electron pedestal density and the location of the ionization peak on neutral penetration in DIII-D H-mode plasmas, utilizing a database of Lyman-α emission measurements. The high electron density leads to neutrals being ‘screened’ and the ionization front being pushed out into the Scrape-Off Layer (SOL). This is also referred to as the neutral opaqueness, which is heuristically expected to scale with edge plasma density and machine size. However, at lower electron pedestal density, the penetration depth of the neutrals varies, and measured opaqueness deviates from the heuristic scaling. The database reveals that at low density, when the ionization peak is located in SOL region, the linear relationship between the electron density and neutral penetration holds. However, when the peak is located inside the separatrix, the penetration of the neutrals (λ n 0 ) is much wider ~3.0–3.5 cm, breaking the heuristic opaqueness approximation. These findings provide valuable insights into fueling efficiency and plasma behavior, with implications for Fusion Pilot Plants where high pedestal densities are anticipated and where the neutral opaqueness behaves like its heuristic approximation. This analysis offers a framework to refine neutral opaqueness approximations, enhancing the predictive capability for advanced tokamak operations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Disentangling core and edge mechanisms of the density limit in DIII-D negative triangularity plasmas

The density limit is investigated in the DIII-D negative triangularity plasmas which lack a standard H-mode edge. We find the limit may not be a singular disruptive boundary but a multifaceted density saturation phenomenon governed by distinct core and edge transport mechanisms. Sustained, non-disruptive operation is achieved at densities up to 1.8 times the Greenwald limit ($n_\mathrm{G}$) until the termination of auxiliary heating. Systematic power scans show distinct power scalings for the core ($n_e \propto P_\mathrm{SOL}^{0.27\pm0.03}$) and edge ($n_e \propto P_\mathrm{SOL}^{0.42\pm0.04}$) density limits. The edge density saturation is triggered by the onset of a non-disruptive, high-field side radiation front and the associated cooling, which clamps the edge density below . In contrast, the core density continues to rise until it saturates, a state characterized by enhanced core turbulence. Core transport evolves from a diffusive to an intermittent, avalanche-like state, as indicated by heavy-tailed probability density functions (kurtosis ≈ 6), increased Hurst exponents, and a $1/f$-type power spectrum. These findings suggest that the density limit in the low-confinement regime is determined by a combination of edge radiative cooling and core turbulent transport. This distinction provides separate targets for control strategies aimed at extending the operational space of future fusion devices.

density limit↗
DOE OSTI2026

Autoregressive long-horizon prediction of plasma edge dynamics *

Accurate modeling of scrape-off layer (SOL) and divertor-edge dynamics is vital for designing plasma-facing components in fusion devices. High-fidelity edge fluid/neutral codes such as SOLPS-ITER capture SOL physics with high accuracy, but their computational cost limits broad parameter scans and long transient studies. We present transformer-based, autoregressive surrogates for efficient prediction of 2D, time-dependent plasma edge state fields. Trained on SOLPS-ITER spatiotemporal data for the KSTAR tokamak, the surrogates forecast electron temperature, electron density, and radiated power over extended horizons. We evaluate model variants trained with increasing autoregressive horizons (1–100 steps) on short- and long-horizon prediction tasks. Longer-horizon training systematically improves rollout stability and mitigates error accumulation, enabling stable predictions over hundreds to thousands of steps and reproducing key dynamical features such as the motion of high-radiation regions. Measured end-to-end wall-clock times show the surrogate is orders of magnitude faster than SOLPS-ITER, enabling rapid parameter exploration. Prediction accuracy degrades when the surrogate enters physical regimes not represented in the training dataset, motivating future work on data enrichment and physics-informed constraints. Overall, this approach provides a fast, accurate surrogate for computationally intensive plasma edge simulations, supporting rapid scenario exploration, control-oriented studies, and progress toward real-time applications in fusion devices.

autoregressive deep learning↗
DOE OSTI2026

Power and particle exhaust in the ST-E1 fusion power plant

Power exhaust challenges and potential solutions for a 5 m major radius, low-aspect ratio burning tokamak have been explored. 1D edge plasma models have been used to screen for access to detachment using short and long outer divertor legs in double and single null configurations, using Ar as the primary impurity and assuming tungsten plasma-facing components (PFCs). These show that detachment access can be accessed for all but the most conservative assumptions on scrape-off layer (SOL) width and power, but that trade-offs will be required between magnet engineering and the size of the acceptable window of as-yet uncertain plasma parameters. SOLPS-ITER was used to further model selected plasma scenarios, confirming that Ar seeding can be used to achieve dissipative divertor scenarios with peak deposited heat fluxes below 15 MWm -2 . Initial scoping of first wall loads and positioning of limiters has been carried out, showing the feasibility of protecting the breeding blanket wall during steady state without impeding tritium breeding. Initial PFC technology selection is also presented, identifying this as a critical area where further work is needed to find an attractive solution for helium-cooled PFCs that can handle high heat fluxes without excessive power requirements. Key questions and trade-offs for concept development have been identified, including: how to achieve high radiation for reduction of SOL power without core performance degradation; whether power exhaust can be well-controlled in a double null plasma; mechanical design and materials challenges of high-heat flux PFCs; and control of material erosion, redeposition and tritium retention.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Modelling the limiter ramp-up of WEST for addressing the future challenges of ITER

This paper presents a joint experimental and numerical investigation into the physics of long limited plasma ramp-up in tokamaks with tungsten (W) first walls, a critical phase for ITER operation. The comparison between the average plasma quantities simulated using the SolEdge-HDG code and the measurements taken during three successive WEST discharges after boronisation shows how challenging it is to predict this phase. While simulations reproduce the general trends at the midplane, with reasonable match in density profiles, they consistently underestimate core temperatures possibly due to too large perpendicular heat conductivity. On the contrary, at the high field side (HFS) limiter, simulations overestimate the measured quantities, and highlights the limitation of using Bohm boundary conditions at grazing magnetic angles. Experimental measurements reveal that the boron layer is rapidly eroded, on a timescale comparable to a single ITER discharge. The subsequent transition from a boron-coated to a tungsten wall increases recycling and significantly degrades the core plasma, reducing the electron temperature by nearly half due to W contamination, despite wall parameters remaining stable. Furthermore, comparisons with Langmuir probes, bolometry, reflectometry, and spectroscopy indicate that the experimental far scrape-Off layer (SOL) is significantly wider than simulated. This wide SOL implies that boron erosion extends along the entire HFS limiter rather than being confined to the contact point. This work highlights some characteristics of the plasma during this phase of the discharge and emphasizes the current modelling issues that need to be resolved in order to obtain reliable predictions concerning the ITER ramp-up.

ITER↗
DOE OSTI2020

Sr 1-x Ca x Fe 1-y Co y O 3-δ as facile and tunable oxygen sorbents for chemical looping air separation

Abstract Chemical looping air separation (CLAS) is a promising technology for oxygen generation with high efficiency. The key challenge for CLAS is to design robust oxygen sorbents with suitable redox properties and fast redox kinetics. In this work, perovskite-structured Sr 1-x Ca x Fe 1-y Co y O 3 oxygen sorbents were investigated and demonstrated for oxygen production with tunable redox properties, high redox rate, and excellent thermal/steam stability. Cobalt doping at B site was found to be highly effective, 33% improvement in oxygen productivity was observed at 500 °C. Moreover, it stabilizes the perovskite structure and prevents phase segregation under pressure swing conditions in the presence of steam. Scalable synthesis of Sr 0.8 Ca 0.2 Fe 0.4 Co 0.6 O 3 oxygen sorbents was carried out through solid state reaction, co-precipitation, and sol-gel methods. Both co-precipitation and sol-gel methods are capable of producing Sr 0.8 Ca 0.2 Fe 0.4 Co 0.6 O 3 sorbents with satisfactory phase purity, high oxygen capacity, and fast redox kinetics. Large scale evaluation of Sr 0.8 Ca 0.2 Fe 0.4 Co 0.6 O 3 , using an automated CLAS testbed with over 300 g sorbent loading, further demonstrated the effectiveness of the oxygen sorbent to produce 95% pure O 2 with a satisfactory productivity of 0.04 g O2 g sorbent −1 h −1 at 600 °C.

25 ENERGY STORAGE↗
DOE OSTI2020

Photonic Curing Enabling High-Speed Processing for Perovskite Solar Cells

The thermal annealing step is the bottleneck to high-throughput manufacturing of perovskite solar cells. In this work, we investigate photonic curing to replace thermal annealing for the sol gel NiO hole transport layer. Using solar cell device current density-voltage performance as a metric, we compare the effects of thermal annealing and photonic curing on the conversion of sol gel NiO films deposited on top of indium tin oxide coated Willow® glass substrates. The significant reduction in processing time validates photonic curing as the enabling tool for high-speed manufacturing.

Xu, Weijie↗
DOE OSTI2020

Plasma Facing Component Characterization and Correlation With Plasma Conditions in Lithium Tokamak Experiment-β

Lithium coatings in the Lithium Tokamak eXperiment (LTX) led to flat temperature profiles. The flat temperature profiles were observed along with a hot, low density edge, implying a broad, collisionless scrape-off layer (SOL). Additionally, in vacuo X-ray photoelectron spectroscopy (XPS) measurements established that lithium coatings evaporatively deposited onto high-Z plasma facing components (PFCs) became oxidized while retaining the ability to achieve good plasma performance long after lithium was applied to the PFCs. Longstanding theory predicted flat temperature profiles with low recycling walls, which was presumed to be due to hydrogen binding with elemental lithium to form lithium hydride. The presence of oxidized lithium, however, raised questions regarding the exact mechanism of hydrogen retention in LTX. To investigate these questions, the upgraded facility LTX- β includes a new sample exposure probe (SEP) for more detailed in vacuo analysis of PFC samples. The SEP is equipped with a vacuum suitcase capable of transporting samples representative of the LTX- β outer midplane PFCs to a stand-alone XPS system while maintaining pressures lower than the LTX- β base vacuum to limit the contamination between sample exposure and analysis. The low-energy resolution XPS system used in past experiments could only enable the determination of elemental percentages on the PFC sample surfaces. Because the new XPS system has higher energy resolution, it is more direct to assign chemical compounds to the measured binding energies. This capability has been confirmed by comparing XPS data from PFC test samples with measurements using a commercial high-resolution XPS system. Quartz crystal microbalances (QCMs) were used to quantify the thickness of the deposited lithium on the LTX- β PFCs. This article describes the application of the SEP to characterize the PFC surfaces using XPS and their relationship to plasma conditions.

Maan, Anurag↗
DOE OSTI2022

Development of an Integrated Multidiagnostic to Assess the High-Z Impurity Fluxes in the Metallic Environment of WEST Using IMAS

WEST is an actively cooled, long-pulse tokamak with nearly all plasma-facing components (PFC) made of tungsten. One of the aims of WEST is to study plasma operations with tungsten PFCs in preparation for long-pulse operations on high-Z divertor devices, such as ITER. For long-pulse operation, the high-Z impurity content and transport to the core plasma are critical concerns that require further measurement and interpretation to improve plasma performance and PFC durability. This work details the impurity influxes in WEST during a series of discharges in which the lower hybrid (LH) injected power was incrementally increased. An analysis has been performed of measurements collected from an array of edge diagnostics. Visible spectroscopy was utilized to measure the spectral radiances generated by fuel particles (D) and impurities (W, O, and C) at the divertor and at the antennas with a newly developed spectral peak-fitting tool used to analyze the data in WEST. The scrape-off layer (SOL) plasma conditions (n e and T e ) measured at the divertor target with flush Langmuir probes and near the outer mid-plane (OMP) using reciprocating Langmuir probes (RCPs) are used to evaluate the number of ionizations per photons (S/XB) coefficients required to estimate the impurity fluxes obtained with the collisional-radiative model ColRadPy. Here, the array of edge diagnostics discussed in this work, coupled with SOL plasma modeling tools, represents a multidiagnostic interpretative modeling workflow that will continue to be applied to upcoming experimental campaigns on the WEST experiment to assess high-Z impurity transport.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

First Wall Design of a Tokamak Pilot Plant Using a Monte Carlo Model for 3-D Heat Flux Deposition

We present a method for calculating the heat fluxes deposited on nonaxisymmetric tokamak first wall components, allowing for a first-of-its-kind model for power handling in the tokamak far scrape-off layer (SOL). The DIV3D Monte Carlo model features strict global power conservation and can calculate the finite cross-field plasma transport into magnetically-shadowed regions, which is significant when dealing with meter-scale shadows introduced by components such as poloidal limiters or antennas. As a case study, we apply the DIV3D model to inform the distribution of first wall poloidal limiters in an ARC-class reactor device. We demonstrate that discrete protection limiters can efficiently reduce peak heat fluxes on recessed breeder wall components in the presence of significant far-SOL plasma fluxes. By varying the toroidal periodicity and radial standoff depth of the limiters, we demonstrate one of the tradeoffs that must be considered in first wall design: more limiters provide greater protection, but at the cost of reduced breeding performance. We also present the impact that radial misalignments between limiters would have on first wall power loading.

Monte Carlo methods↗
DOE OSTI2022

Mitigation of Leakage and Water Ingress for a Promising Phase Change Material (PCM) Containing Polyethylene Glycol (PEG)

A promising and cost-effective phase change material (PCM) containing polyethylene glycol (PEG) 600, 800, and/or 1000 has been synthesized. It is shape stabilized using a sol-gel method and displays promising thermal properties. These properties make the PCMs good candidates for thermal batteries and other thermal energy storage applications in the building sector. However, PEG leakage is observed at elevated temperatures, and due to the hygroscopic nature of PEG, moisture ingress is also observed, which can have a significant effect on performance and durability. Thus, we have developed hydrophobic coatings to prevent PEG leakage and mitigate moisture ingress. The selected approach involves a secondary sol-gel process during which a thin layer formulated with hydrophobic precursors is deposited onto the surface of the PCM samples. Several coating formulations have been investigated, including formulations composed of tetraethyl orthosilicate (TEOS), diethoxy(dimethyl)silane (DEDMS), triethoxymethylsilane (TEMS), and n-octyltriethoxysilane (OTES). We find that this is effective at increasing hydrophobicity of the PCM substrates. It also decreases substrate leakage.

ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATION↗
DOE OSTI2020

S/U Comparison Study with a Focus on USLs

Under a DOE Nuclear Criticality Safety Program (NCSP) task involving Analytical Methods, three Laboratories collaborated in a comparison of results obtained from Sensitivity/Uncertainty (S/U) packages relevant to validation of transport codes. The task involves Institut de Radioprotection et de Sûreté Nucléaire (IRSN), Los Alamos National Laboratory (LANL), and Oak Ridge National Laboratory (ORNL) comparing results of MORET 5/MACSENS V3.0, MCNP6.2/Whisper-1.1, and SCALE 6.2.3/TSUNAMI/USLSTATS respectively. All Monte Carlo transport code results utilize ENDF/B-VII.1. Four cases from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) were selected as application models: HEU-MET-FAST-013-001, HEU-SOL-THERM 001-008, PU-MET-FAST-022-001, and PU-SOL-THERM 001-001. Ultimately, comparison is made between Upper Subcritical Limits (USLs) obtained using each code package for each application case. Since differences exist in whether packages take into account margin of subcriticality (MOS), the USL may be computed using bias and bias uncertainty, also known as the calculational margin (CM) in ANSI/ANS 8.24. Application of portions of MOS to the USL for nuclear data uncertainty of and potential code margin is referred to as USL herein. In either case, additional MOS is considered for actual application cases.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Scrape-off layer stability, turbulence and transport in MAST-U (Final Report)

U.S. Department of Energy (DOE) supported work at Lodestar Research Corporation under grant DE-SC0019270 is summarized in this final report. The research conducted under this grant explored the properties of the edge plasma in spherical torus magnetic fusion devices. Harnessing the energy that is released in fusion reactions would provide a safe and abundant source of power to meet the growing energy needs of the world population. The edge and scrape-off layer (SOL) region of fusion plasmas is the region where the hot core plasma transitions to a cooler surrounding boundary plasma. The properties of this boundary plasma determine how interactions with material surfaces will impact both the overall device performance and the lifetime of crucial components such as the walls of the device chamber and other hardware. The research carried out studied filamentary plasma instabilities and turbulent transport in the edge and SOL of the MAST U spherical torus.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗