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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 145 records · Page 8
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DOE OSTI2022

Foreword to special issue: Papers from the 63rd annual meeting of the APS Division of Plasma Physics, November 8–12, 2021

The 63rd annual meeting of the APS Division of Plasma Physics (DPP) was held on November 8–12, 2021 in Pittsburgh at the David Lawrence Convention Center with both a live (in person) component and a virtual component. Following guidance from an APS COVID task force, all in-person attendees were fully vaccinated and masked. More than 800 physicists attended, safely, in-person. With both virtual and on-site participants, discussions were lively, and the research presentations showed unmatched mastery in the modern observation, theory, simulation, and manipulation of plasma. The presentations included four invited review talks, 97 invited talks, four tutorials, and four presentations from this year's prize and award recipients. There were more than 1200 contributed poster presentations and 725 contributed oral presentations. Including both in-person and remote attendees, DPP 2021 had a record of 2232 participants. As a hybrid meeting, in-person presentations of all invited presentations were broadcast live and were accompanied by a Q&A discussion. Contributed oral and poster presentations were prerecorded along with options to schedule in-person discussions on demand. Five mini-conferences were held: “Gatekeeper Workshop: Creating a Diverse, Equitable, and Inclusive Pipeline,” “Collisionless Shocks in Laboratory and Space Plasmas,” “The High Repetition Rate Frontier in High-Energy-Density Physics,” “Measuring and Modeling Plasma Surface Interactions,” and “The Second Mini-Conference on Machine Learning, Data Science and Artificial Intelligence in Plasma Research.” Finally, on the day before the official start of the meeting, an afternoon “for students, by students” included lightning talks, plasma trivia, and an informal occasion to connect with other students, learn how to get the most from the DPP Annual Meeting, and share successful ways to connect with colleagues and advance their professional careers.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Thermal expansion of plasma-exposed tungsten

We report results from a systematic analysis of thermal expansion of plasma-exposed tungsten based on molecular-dynamics simulations using models of tungsten with distributions of helium (He) bubbles in the tungsten matrix. We distinguish between two approaches of filling the bubbles with He, where the amount of He in the bubble can or cannot vary with temperature. In the former case, the thermal expansion coefficient decreases monotonically with the porosity and He content of the tungsten matrix, while in the latter case, the thermal expansivity increases monotonically with increasing porosity and He content. The latter condition, where the He content in the bubble is determined at the implantation temperature and remains constant with varying temperature in the tungsten matrix, is consistent with He species transport in tungsten used as a plasma-facing component (PFC) in nuclear fusion reactors and implies the development of biaxial compressive thermal strains in the PFC material that contribute to accelerating the growth of a nanostructure on PFC tungsten surfaces. Furthermore our analysis advances the fundamental understanding of thermal expansion in PFC tungsten and contributes to the development of a thermophysical property database for properly incorporating effects of realistic heat loads into modeling the dynamical response of PFC tungsten under fusion reactor operating conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Different roles of elongation and triangularity in the H-mode pedestal height enhancement through strong plasma shaping

While the peeling–ballooning mode is known to be stabilized by strong plasma shaping in a high-confinement mode (H-mode), a significant difference is found in the roles of elongation and triangularity. The elongation mainly stabilizes the high-n ballooning branch, while the triangularity of the low-n peeling one with some destabilization of the high-n ballooning one, where n is the toroidal mode number. This difference is conjectured mainly due to the magnetic shear effect which is substantially enhanced in a high triangularity regime, giving a destabilization on the ballooning branch, while a stabilization on the peeling one. As shown here, an important consequence of these different roles of the two shape parameters is that they should work together to stabilize simultaneously both branches of the PBM, implying the necessity of synergy between them for the large pedestal height enhancement.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

An investigation of shock formation vs shock mitigation of colliding plasma jets

Here, this work studies the interaction between colliding plasma jets to understand regimes in which jet merging results in shock formation vs regimes in which the shock formation is mitigated due to the collisionless interpenetration of the jets. A kinetic model is required for this study because fluid models will always produce a shock upon the collision of plasma jets. The continuum-kinetic, Vlasov–Maxwell–Dougherty model with one velocity dimension is used to accurately capture shock heating, along with a novel coupling with a moment equation to evolve perpendicular temperature for computational efficiency. As a result, this relatively inexpensive simulation can be used for detailed scans of the parameter space toward predictions of shocked vs shock-mitigated regimes, which is of interest for several fusion concepts such as plasma-jet-driven magneto-inertial fusion, high-energy-density plasmas, astrophysical phenomena, and other laboratory plasmas. The initial results obtained using this approach are in agreement with the preliminary outcomes of the plasma liner experiment.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Use of intrinsic hysteresis for the active control of internal transport barriers in magnetically confined fusion plasmas

In magnetically confined fusion devices, control of internal transport barriers (ITBs) is important both to enhance and suppress the turbulent transport to improve confinement control. Barrier control should allow for the improvement of confinement to aid in achieving the needed fusion criteria while also permitting the degradation of confinement to control profiles and clean the device by moving out the impurities accumulated near the core. In this work, we present a novel control scenario that takes advantage of the hysteresis intrinsic to transport barriers to easily cycle through enhanced and degraded confinement regimes. The control scenario is illustrated using a five-field simplified transport model for an ITB using typical parameters of a neutral beam injection-heated DIII-D tokamak discharge. Pellets and ion cyclotron resonance frequency power are used as control knobs for this active control scenario. These knobs adequately modify at will the local gradients and, therefore, the growth rates and shearing rates, allowing for easy and efficient control of the barrier by taking advantage of the barrier hysteresis. The result is a control cycle that could be operated with a relatively small amount of power in high performance regimes which, nowadays, typically require large power to control. It may also have advantages to avoid, or at least ameliorate, the appearance of magnetohydrodynamic instabilities in the barrier region.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Close coupling of the peeling ballooning and external kink modes in high poloidal-beta discharges with strong shape and large pedestal width

In plasmas with strong shape and large pedestal width, like DIII-D high poloidal-beta discharges, the eigenvalue spectrum of the peeling ballooning mode (PBM) is shown to shift to the very low-n regime near one, where n is the toroidal mode number. This peeling-type eigenvalue spectrum has a further shift to n = 1, thus being smoothly connected to the n = 1 external kink mode (EKM), as the normalized beta (β N ) increases. Once this connection occurs, the mode takes a mixture form of the PBM and EKM with its mode structure varying from the PBM-like to the EKM-like one as β N increases. The mode stability also becomes sensitive to both the local pedestal gradient and global β N , thus allowing an anti-correlation between the two driving forces. Further, these results appear to provide a qualitative explanation of the two unexpected features observed in the DIII-D high poloidal-beta discharges, that is, the dominance of the n = 1 mode in the edge-localized-modes and the negative correlation between the pedestal height and the internal transport barrier strength.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Performance scaling with an applied magnetic field in indirect-drive inertial confinement fusion implosions

Magnetizing a cryogenic deuterium–tritium (DT)-layered inertial confinement fusion (ICF) implosion can improve performance by reducing thermal conduction and improving DT-alpha confinement in the hot spot. A room-temperature, magnetized indirect-drive ICF platform at the National Ignition Facility has been developed, using a high-Z, high-resistivity AuTa 4 alloy as the hohlraum wall material. Experiments show a 2.5× increase in deuterium–deuterium (DD) neutron yield and a 0.8-keV increase in hot-spot temperature with the application of a 12-T B-field. Further, for an initial 26-T B-field, we observed a 2.9× yield increase and a 1.1-keV temperature increase, with the inferred burn-averaged B-field in the compressed hot spot estimated to be 7.1 ± 1.8 kT using measured primary DD-n and secondary DT-n neutron yields.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Exploration of cross-beam energy transfer mitigation constraints for designing an ignition-scale direct-drive inertial confinement fusion driver

The compression of direct-drive inertial confinement fusion (ICF) targets is strongly impacted by cross-beam energy transfer (CBET), a laser-plasma instability that limits ablation pressure by redirecting laser energy outward and that is projected to be mitigated by laser bandwidth. Here, we explore various CBET mitigation constraints to guide the design of future ICF facilities. First, we find that the flat, Gaussian, and Lorentzian spectral shapes have similar CBET mitigation properties, and a flat shape with nine spectral lines is a good surrogate for what can be obtained with other spectral shapes. Then, we conduct a comprehensive study across energy scales and ignition designs. 3D hydrodynamic simulations are used to derive an analytical model for the expected CBET mitigation as a function of laser and plasma parameters. From this model, we study the bandwidth requirements of conventional and shock ignition designs across four different energy scales and find that they require between 0.5 and 3±0.2% relative bandwidth. Further, best mitigation is achieved when the beam radius over critical radius R b /R c is kept low during the drive while the plasma temperature is kept high. In a steady state, we find that the bandwidth required to mitigate 85% of CBET scales as (R b /R c ) 2.15 L n –0.58 I 0.7 , where L n is the density scale length, and I the laser intensity. Finally, we find that the chamber beam port layout does not influence CBET mitigation. In the case of a driver using many monochromatic beamlets, we find that ~10 beamlets per port is required, with diminishing returns above ~20.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Resonance cones in cold plasma: Origin, singularities, and power flow

In magnetized tenuous plasma, typical at the plasma edge of fusion devices, a nearly electrostatic wave mode with relatively enhanced electric field can propagate along a specific angle with the magnetic field. For this characteristic, it is known as a “resonance cone.” For instance, these waves can be excited by radio frequency antennas in the ion-cyclotron and lower-hybrid range of frequencies. We consider the resonance cones emitted by idealized spatially extended sources. In 2D, we use a novel geometric construction which generalizes the d'Alembert solution to curved boundaries/moving sources, and show, for the first time, that singular electric fields arise under these conditions, thereby bringing the resonance cones in line with the other resonances of the cold plasma theory. Still in 2D, we give an expression for the amount of power radiated by resonance cones in terms of surface quantities on the source, which is finite despite the singular electric field. In conclusion, we generalize the conclusions regarding the presence and location of singular electric fields to the 3D electromagnetic case.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Flexible, integrated modeling of tokamak stability, transport, equilibrium, and pedestal physics

The STEP (Stability, Transport, Equilibrium, and Pedestal) integrated-modeling tool has been developed in OMFIT to predict stable, tokamak equilibria self-consistently with core-transport and pedestal calculations. STEP couples theory-based codes to integrate a variety of physics, including magnetohydrodynamic stability, transport, equilibrium, pedestal formation, and current-drive, heating, and fueling. The input/output of each code is interfaced with a centralized ITER-Integrated Modelling & Analysis Suite data structure, allowing codes to be run in any order and enabling open-loop, feedback, and optimization workflows. This paradigm simplifies the integration of new codes, making STEP highly extensible. STEP has been verified against a published benchmark of six different integrated models. Core-pedestal calculations with STEP have been successfully validated against individual DIII-D H-mode discharges and across more than 500 discharges of the H98,y2 database, with a mean error in confinement time from experiment less than 19%. STEP has also reproduced results in less conventional DIII-D scenarios, including negative-central-shear and negative-triangularity plasmas. Predictive STEP modeling has been used to assess performance in several tokamak reactors. Simulations of a high-field, large-aspect-ratio reactor show significantly lower fusion power than predicted by a zero-dimensional study, demonstrating the limitations of scaling-law extrapolations. STEP predictions have found promising scenarios for an EXhaust and Confinement Integration Tokamak Experiment, including a high-pressure, 80%-bootstrap-fraction plasma. ITER modeling with STEP has shown that pellet fueling enhances fusion gain in both the baseline and advanced-inductive scenarios. Finally, STEP predictions for the SPARC baseline scenario are in good agreement with published results from the physics basis.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Kinetic-ballooning-bifurcation in tokamak pedestals across shaping and aspect-ratio

We use a new gyrokinetic threshold model to predict a bifurcation in tokamak pedestal width-height scalings that depends strongly on plasma shaping and aspect-ratio. The bifurcation arises from the first and second stability properties of kinetic-ballooning-modes that yields wide and narrow pedestal branches, expanding the space of accessible pedestal widths and heights. The wide branch offers potential for edge-localized-mode-free pedestals with high core pressure. For negative triangularity, low-aspect-ratio configurations are predicted to give steeper pedestals than conventional-aspect-ratio. Both wide and narrow branches have been attained in tokamak experiments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Millimeter-wave high-wavenumber scattering diagnostic developments on EAST and NSTX-U

A pioneering 4-channel, high-k poloidal, millimeter-wave collective scattering system has been successfully developed for the Experimental Advanced Superconducting Tokamak (EAST). Engineered to explore high-k electron density fluctuations, this innovative system deploys a 270 GHz mm-wave probe beam launched from Port K and directed toward Port P (both ports lie on the midplane and are 110° part), where large aperture optics capture radiation across four simultaneous scattering angles. Tailored to measure density fluctuations with a poloidal wavenumber of up to 20 cm -1 , this high-k scattering system underwent rigorous laboratory testing in 2023, and the installation is currently being carried out on EAST. Its primary purpose lies in scrutinizing ion and electron-scale instabilities, such as the electron temperature gradient (ETG) mode, by furnishing measurements of the k θ (poloidal wavenumber) spectrum. This advancement significantly bolsters the capacity to probe high-k electron density fluctuations within the framework of EAST. Finally, beam tracing and data interpretation modules developed for both EAST and NSTX-U high-k scattering diagnostics are described.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Hybrid direct drive with a two-sided ultraviolet laser

This paper presents a “hybrid” approach to direct drive inertial confinement fusion that can exploit a high-energy gas laser with two opposed beams. The target and driver are asymmetric, much like experiments performed on the National Ignition Facility, but have been designed to benefit from scale and their particular compatibility with a fusion power plant. The imploded masses (and areal densities) are increased by a factor of 12 (3) and provide a path to high-gain implosions that robustly ignite. The design also mitigates common concerns such as laser imprint and cross-beam energy transfer. We discuss the rationales for a hybrid target, the methods used to control implosion symmetry, and the implication(s) for inertial fusion energy.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Exploring scenarios for enhanced fuel compression and performance on the National Ignition Facility with machine-learning-aided design techniques

Recent fusion experiments on the National Ignition Facility (NIF) have achieved ignition, producing multi-MJ fusion yields for input laser energies of roughly 2 MJ [Abu-Shawareb et al., Phys. Rev. Lett. 132, 065102 (2024)]. Building on the success of the target designs that have achieved ignition, we explore new implosion scenarios predicted to generate significantly more compression of the dense DT ice layer and correspondingly higher yields while preserving many of the key physics characteristics of present-day ignition designs. Our main result is a novel 3-shock implosion scheme that effectively minimizes the shock-induced entropy in the dense, accelerating DT shell and maximizes the resulting fuel compression subject to a fixed leading shock strength consistent with present-day ignition experiments, which is necessary to melt the crystalline high-density carbon ablator. Compared to the first NIF experiment to fulfill Lawson's ignition criterion, shot N210808 [Abu-Shawareb et al., Phys. Rev. Lett. 129, 075001 (2022)], our design exhibits a 40% increase in simulated peak areal density (ρR) and a 5× increase in 1D fusion yield using a 4% lighter ablator and identical DT payloads. We also present a complete integrated 2D hohlraum design and laser pulse specifications capable of generating the desired 3-shock drive and maintaining control of the low-mode capsule implosion symmetry, where the increase in simulated 2D yield relative to N210808 is > 10×. This new implosion regime was discovered with help from a machine-learning-enabled capsule design optimization framework. We outline the workflow this automated tool uses to identify improved design candidates by running several rounds of capsule simulations, constructing a surrogate model mapping input variations to key physics output quantities, and querying the resulting statistical model to propose adjustments to the x-ray drive and capsule to reach a set of physics objectives prescribed by the designer.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

First-wall fluxes from large and small ELMs

Experimental analysis at DIII-D shows that small edge localized modes (ELMs) deposit a larger fraction of their energy to the first wall, compared to type-I ELMs in similar magnetic configuration and input power. The energy (λQ) and particle (λΓ) flux decay lengths in the scrape-off layer (SOL) are up to 3 and 5 times larger, respectively, for small ELMs than for larger type-I ELMs. Transport dynamics of ELM filaments in the SOL are found to be related to divertor conditions, where high divertor collisionality, typical for partially detached plasmas, is associated with increased cross field ELM radial fluxes. Results show that a sufficiently large outer wall gap and/or limiters might be needed in future scenarios to protect the first wall, if operating with small ELMs and a cold divertor. This might also have implications for RF heating in future devices, where the coupling efficiency is dependent to some degree to the outer-wall gap.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

The LIBRA Experiment: Investigating Robust Tritium Accountancy in Molten FLiBe Exposed to a D-T Fusion Neutron Spectrum

The Liquid Immersion Blanket: Robust Accountancy (LIBRA) experiment will be a first-of-a-kind experiment to explore and develop the liquid immersion blanket (LIB) concept. The LIB is a radically simple molten–LiF-BeF2 (FLiBe)–salt tritium breeding blanket for deuterium-tritium (D-T)–fueled fusion power plants (FPPs) achieving a high tritium breeding ratio (TBR) in neutronics models. However, tritium breeding in FLiBe is inherently difficult to study experimentally. As a result, the coupled issues of FLiBe radiochemistry and tritium (T) transport are poorly understood. LIBRA approaches this challenge by simulating an FPP blanket environment using a D-T neutron generator and ≈ 1000 kg of FLiBe. LIBRA will investigate T breeding, containment, and extraction, coupled with FLiBe redox control and radiochemistry. The primary goal of LIBRA is to demonstrate robust T accountancy in blanket prototypical conditions. Here, T accountancy encompasses accurate predictions of T breeding in the FLiBe; detection and measurement of all T bred in LIBRA; and speciation of the T extracted from the FLiBe. Initial neutronics simulations of LIBRA indicate that a global TBR of ~ 1 is possible, where the TBR is defined as the number of tritons bred and extracted from FLiBe relative to the number of neutrons produced by D-T fusion reactions in the neutron generator. In this paper, we present the LIBRA concept and its scientific goals in the context of T breeding experiments. We also consider the potential impact of the LIB on the future fusion power industry, motivating further development of FLiBe-based T breeding research activities such as LIBRA.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Radiation Damage Analysis of FNSF Components Using McCad and MCNP

The Fusion Energy System Studies Fusion Nuclear Science Facility (FESS-FNSF) concept represents a transitional step between ITER and a commercial fusion power plant. The FNSF is a conceptualized D-T fueled tokamak with 518 MW of fusion power that has been extensively used to explore and optimize design features. The energetic 14.1-MeV neutrons can produce significant localized heating and activations, and can cause damage to plasma-facing components, which can determine maintenance/outage scheduling needs and also impact the lifetime of the device as a whole. This study illustrates a neutronics analysis that was conducted on a 22.5-degree symmetric sector of the FNSF with the goal of understanding the neutron heating and radiation damage that can be characterized by quantifying the displacements per atom (dpa). Concurrently, this study also focused on the development of analysis capabilities by converting a three-dimensional computer-aided design model of the FNSF into MCNP6.2 input using the McCad code. Accordingly, some confirmatory results on tritium production and the tritium breeding ratio (TBR) are provided to support model validation. The results produced by MCNP6.2 simulations showed that the highest heating and damage occurred in the outboard region, which concentrated approximately 290 MW of the total nuclear heating, in contrast to 97 MW within the inboard region. These results are consistent with previous studies that employed earlier versions of the FNSF concept and different modeling approaches. Furthermore, this study also provides additional details on neutron wall loading, as well as total heating from neutrons and gammas, results which show the total heating of the device (16 sectors) is approximately 477.83 ± 0.80% MW, indicating a neutron energy multiplication factor of 1.15. Additionally, the capability to calculate hydrogen and helium production, as well as dpa, is illustrated. Finally, the neutronics effects of using alternative materials to tungsten carbide were evaluated for the vacuum vessel, low-temperature shield, and structural ring components, which showed that compounds like YH 2 , Mg(BH 4 ) 2 , and ZrH 2 could reduce the total heating on the magnet and also reduce the TBR.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗