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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 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Patterns of Care, Tolerability, and Safety of the First Cohort of Patients Treated on a Novel High-Field MR-Linac Within the MOMENTUM Study: Initial Results From a Prospective Multi-Institutional Registry

High-field magnetic resonance-linear accelerators (MR-Linacs), linear accelerators combined with a diagnostic magnetic resonance imaging (MRI) scanner and online adaptive workflow, potentially give rise to novel online anatomic and response adaptive radiation therapy paradigms. The first high-field (1.5T) MR-Linac received regulatory approval in late 2018, and little is known about clinical use, patient tolerability of daily high-field MRI, and toxicity of treatments. Herein we report the initial experience within the MOMENTUM Study (NCT04075305), a prospective international registry of the MR-Linac Consortium.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2020

Magnetic Detection of Chemical Threats

Los Alamos researchers have developed a technology to quickly detect chemical threats using a fieldable detector that provides real-time detection to ensure safety for warfighters, agricultural workers, and the general public. Traditional laboratory-based detection does not allow for agile response to threats on the battlefield or to our food supply chain. Our portable detector brings the laboratory to the field for real-time detection of chemical agents in samples without needing to send the samples to a remote location for analysis. This technology fits a market need in public safety, military sensor design, and pesticide detection. We have proven that this technology works on the benchtop, and are working to miniaturize the system to enable deployment. We are seeking qualified licensing or CRADA partners to finalize development of this system.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Conceptual Design for a Blanket Tritium Extraction Test Stand

Future fusion reactors must be able to breed the tritium they will consume. Several breeding and tritium extraction technologies are under investigation internationally. PbLi is of particular interest as a breeder material. A new forced convection PbLi loop, the Tritium Extraction eXperiment (TEX), is being designed and constructed to investigate tritium extraction from PbLi. Specifically, TEX will serve to investigate the extraction efficiency of various vacuum permeator membranes and configurations. The major components of the loop include a moving magnet pump, reverse permeator, furnace test section, analysis chamber, supply tank, and plenum. As significant hazards are present in such experiments, safety is an integral focus of the experimental design.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI1998

Fusion Safety Program Peer Review (August 31 and September 1, 1999) [Slides]

The Fusion Safety Program Peer Review conference, held on August 31 and September 1, 1999, focuses on advancing the understanding and management of radioactive and hazardous materials within deuterium-tritium (D-T) fusion machines. The conference aims to investigate the behavior of significant sources of radioactive materials, such as activation products, dust, tritium, and beryllium. Additionally, it seeks to comprehend how various energy sources in fusion facilities—such as magnets, plasma, decay heat, and chemical reactions—can mobilize these materials. A key objective of the conference is to develop integrated, state-of-the-art analytic tools to demonstrate the safety and environmental potential of fusion technology. Furthermore, the conference assesses and evaluates safety and environmental issues associated with emerging fusion concepts, including those in the ARIES, ALPS, APEX, and IFE projects. This comprehensive approach aims to ensure the safe and sustainable advancement of fusion energy.

70 - PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Accuracy of kinetic equilibrium reconstruction of NSTX and NSTX-U plasmas and its impact on the transport and stability analysis

An accurate magnetohydrodynamic (MHD) equilibrium reconstruction is an essential starting point for stability and transport plasma analysis. Herein this work describes an approach for obtaining kinetic equilibrium reconstructions using the OMFIT framework, which has been applied for the first time to spherical tokamak data from NSTX and NSTX-U. The EFIT equilibrium solver is integrated with experimental data analysis procedures and subsequent TRANSP transport simulations to enhance the accuracy of the reconstruction, in particular, at the edge region, by adding constraints on the total pressure and current density profiles, based on the transport code solution. The accuracy of the equilibrium reconstruction depends on the uncertainty and number of constraints, as well as the choice of basis functions to represent the pressure and current density profiles. Improved fidelity of the equilibrium reconstruction is demonstrated by reducing the variability of the magnetic axis and boundary locations from several centimeters, for reconstructions based on magnetic and experimental pressure constraints, to only several millimeters, for kinetic reconstructions based on transport code constraints, when different representations of basis functions were tested. The variability of the safety factor on axis was reduced ten times in the same sensitivity study. The accuracy of the equilibrium reconstruction and subsequent mapping of the experimental kinetic profile data have a significant impact on the trapped gyro Landau fluid and linear CGYRO turbulence simulations, which predict different spectra of unstable modes and turbulent fluxes for cases with different numbers of constraints in the equilibrium reconstruction. Conversely, the stability analysis performed using the GATO code shows plasmas that are stable to n = 1 MHD modes in both equilibria using magnetic and experimental pressure constraints as well as the transport code constrained equilibrium. However, a scan of parameters away from these conditions shows considerable deviation in the threshold of unstable modes between these reconstructions. Therefore, for reliable plasma analysis and use in turbulence and stability calculations, a high-fidelity equilibrium reconstruction with accurate kinetic constraints based on transport code solutions is necessary.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

A curvature-based mechanism for the spontaneous emission of electromagnetic radiation at curved plasma–vacuum interfaces

In this work, we derive from first principles the dispersion relation for electromagnetic radiation that is spontaneously emitted at a 2D curved plasma–vacuum interface in the absence of both background and external electric and magnetic fields. This plasma geometry is motivated by plasmas generated through hypervelocity impact projectiles—an area that is of importance to satellite safety in the aerospace community, and may be of importance for the developing study of projectile impact-driven inertial confinement fusion. We find spontaneous radiation occurs when thermal surface waves propagate along the edge of a circular plasma, with a frequency lower than the plasma frequency and spatial decay proportional to 1/r. This is distinct from more well-known mechanisms of radiation, in which the radiation frequency is greater than or equal to the plasma frequency. This theory is validated against 2D, electromagnetic particle-in-cell simulations and shows reasonable agreement. Finally, we discuss the effects of plasma geometry and attribute the radiation generation mechanism to the acceleration of charged surface waves as they traverse azimuthally along a curved path.

Lau, Raymond (ORCID:0000000299318101)↗
DOE OSTI2025

Future Circular Collider Feasibility Study Report

In response to the 2020 Update of the European Strategy for Particle Physics , the Future Circular Collider (FCC) Feasibility Study was launched as an international collaboration hosted by CERN. This report describes the FCC integrated programme , which consists of two stages: an electron-positron collider (FCC-ee) in the first phase, serving as a high-luminosity Higgs, top, and electroweak factory; followed by a proton-proton collider (FCC-hh) at the energy frontier in the second phase. The FCC-ee is designed to operate at four key centre-of-mass energies: the Z pole, the WW pair production threshold, the ZH production peak, and the top/anti-top production threshold—each delivering the highest possible luminosities to four experiments. Over 15 years of operation, FCC-ee will produce more than 6 trillion Z bosons, 200 million WW pairs, nearly 3 million Higgs bosons, and 2 million top anti-top pairs. Precise energy calibration at the Z pole and WW threshold will be achieved through frequent resonant depolarisation of pilot bunches. The sequence of operation modes between the Z, WW, and ZH substages remains flexible. The FCC-hh will operate at a centre-of-mass energy of approximately 85 TeV—nearly an order of magnitude higher than the LHC—and is designed to deliver 5 to 10 times the integrated luminosity of the upcoming High-Luminosity LHC. Its mass reach for direct discovery extends to several tens of TeV. In addition to proton-proton collisions, the FCC-hh is capable of supporting ion-ion, ion-proton, and lepton-hadron collision modes. This second volume of the Feasibility Study Report presents the complete design of the FCC-ee collider, its operation and staging strategy, the full-energy booster and injector complex, required accelerator technologies, safety concepts, and technical infrastructure. It also includes the design of the FCC-hh hadron collider, development of high-field magnets, hadron injector options, and key technical systems for FCC-hh.

Benedikt, M. [European Organization for Nuclear Re↗
DOE OSTI2021

A software package for plasma facing component analysis and design: the Heat flux Engineering Analysis Toolkit (HEAT)

The engineering limits of plasma facing components (PFCs) constrain the allowable operational space of tokamaks. Poorly managed heat fluxes that push the PFCs beyond their limits not only degrade core plasma performance via elevated impurities, but can also result in PFC failure due to thermal stresses or melting. Simple axisymmetric assumptions fail to capture the complex interaction between 3D PFC geometry and 2D or 3D plasmas. This results in fusion systems that must either operate with increased risk or reduce PFC loads, potentially through lower core plasma performance, to maintain a nominal safety factor. High precision 3D heat flux predictions are necessary to accurately ascertain the state of a PFC given the evolution of the magnetic equilibrium. A new code, the Heat flux Engineering Analysis Toolkit (HEAT), has been developed to provide high precision 3D predictions and analysis for PFCs. HEAT couples many otherwise disparate computational tools together into a single open source python package. Magnetic equilibrium, engineering CAD, finite volume solvers, scrape off layer plasma physics, visualization, high performace computing, and more, are connected in a single web-based user interface. Linux users may use HEAT without any software prerequisites via an appImage. This manuscript introduces HEAT, discusses the software architecture, presents first HEAT results, and outlines physics modules in development.

divertor physics↗
DOE OSTI2020

ELM control optimization for various ITER scenarios based on linear and quasi-linear figures of merit

For the purpose of better understanding type-I edge localized mode (ELM) control in ITER with resonant magnetic perturbation (RMP) fields, the plasma response to RMP is computed by a resistive full magneto-hydrodynamic model in toroidal geometry. Five scenarios designed for ITER are considered, ranging from the pre-nuclear to nuclear phases. The plasma response to RMP is quantified by the plasma surface displacement near the X-point of the divertor plasma and at the outboard mid-plane. The optimal coil configurations between two high- Q deuterium-tritium (DT) scenarios (at the same plasma current of 15 MA and the same magnetic field of 5.3 T but different fusion gains, Q = 5 and 10) are predicted to be similar. For the other ITER scenarios with similar edge safety factor q 95 ~ 3 to that of the baseline scenario, the optimal coil phasing is also similar. The optimization results are different for a half-current full-field (7.5 MA/5.3 T) scenario, largely due to the difference in q 95 . The RMP coil currents are also optimized to tailor the core vs edge toroidal torques exerted by the 3D RMP fields on the plasma column. Torque optimization, with various objective functions proposed in the study, is useful for minimizing the side effects of RMP on the plasma core flow in ITER, while still maintaining the ELM control capability. Full utilization of three rows of ELM control coils in ITER is found to be essential to ensure both flexibility and robustness of ELM control, in terms of both linear and quasilinear plasma responses.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Magnetohydrodynamic Control of Interfacial Degradation in Lithium-Ion Batteries for Fast Charging Applications

Interfacial anodic degradation in graphitic materials under fast charging conditions causes severe performance loss and safety hazard in lithium ion batteries. We present a novel method for minimizing the growth of these aging mechanism by application of an external magnetic field. Under magnetic field, paramagnetic lithium ions experience a magnetohydrodynamic force, which rotates the perpendicularly diffusing species and homogenizes the ionic transport. This phenomenon minimizes the overpotential hotspots at the anode/separator interface, consequently reducing SEI growth, lithium plating, and interfacial fracture. In situ electrochemical measurements indicate an improvement in capacity for lithium cobalt oxide/graphite pouch cell (20 mAh) charged from 1–5 C under an applied field of 1.8 kG, with a maximum capacity gain of 22% at 5C. Post-mortem FE-SEM and EDS mapping shows that samples charged with magnetic field have a reduced lithium deposition at 3C and a complete suppression of interfacial fracture at 5C. At 5C, a 24% reduction in the lithium content is observed by performing XPS on the anodic interfacial film. Finally, fast charging performance under variable magnetic field strengths indicate a saturation behavior in capacity at high fields (>2 kG), thereby limiting the field and consequent energy requirements to obtain maximum capacity gain under extreme conditions.

electromagnetism↗
DOE OSTI2024

Avoiding fusion plasma tearing instability with deep reinforcement learning

For stable and efficient fusion energy production using a tokamak reactor, it is essential to maintain a high-pressure hydrogenic plasma without plasma disruption. Therefore, it is necessary to actively control the tokamak based on the observed plasma state, to manoeuvre high-pressure plasma while avoiding tearing instability, the leading cause of disruptions. This presents an obstacle-avoidance problem for which artificial intelligence based on reinforcement learning has recently shown remarkable performance. However, the obstacle here, the tearing instability, is difficult to forecast and is highly prone to terminating plasma operations, especially in the ITER baseline scenario. Previously, we developed a multimodal dynamic model that estimates the likelihood of future tearing instability based on signals from multiple diagnostics and actuators. Here we harness this dynamic model as a training environment for reinforcement-learning artificial intelligence, facilitating automated instability prevention. We demonstrate artificial intelligence control to lower the possibility of disruptive tearing instabilities in DIII-D, the largest magnetic fusion facility in the United States. The controller maintained the tearing likelihood under a given threshold, even under relatively unfavourable conditions of low safety factor and low torque. In particular, it allowed the plasma to actively track the stable path within the time-varying operational space while maintaining H-mode performance, which was challenging with traditional preprogrammed control. This controller paves the path to developing stable high-performance operational scenarios for future use in ITER.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Hydrogen isotope analysis in W-tiles using fs-LIBS

Abstract Laser-Induced Breakdown Spectroscopy (LIBS) is a promising technology for in-situ analysis of Plasma-Facing Components in magnetic confinement fusion facilities. It is of major interest to monitor the hydrogen isotope retention i.e. tritium and deuterium over many operation hours to guarantee safety and availability of the future reactor. In our studies we use ultraviolet femtosecond laser pulses to analyze tungsten (W) tiles that were exposed to a deuterium plasma in the linear plasma device PSI-2, which mimics conditions at the first wall. A high-resolution spectrometer is used to detect the Balmer- $$\alpha$$ α transition of the surface from implanted hydrogen isotopes (H and D). We use Calibration Free CF-LIBS to quantify the amount of deuterium stored in W. This proof-of-principle study shows the applicability of femtosecond lasers for the detection of low deuterium concentration as present in first wall material of prevailing fusion experiments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Quasi-linear toroidal simulations of resonant magnetic perturbations in eight ITER H-mode scenarios

Abstract Both linear and quasi-linear aspects of the plasma response to the resonant magnetic perturbation (RMP) field are numerically investigated for various H-mode scenarios in ITER, covering the pre-fusion power operation and the fusion power operation phases. Linear response computations for eight ITER scenarios, with varying plasma current and toroidal magnetic field, reveal that the best coil current phasing for controlling the type-I edge localized modes (ELMs) scales roughly linearly with the edge safety factor. The coil phasing is defined as the relative toroidal phase of the coil currents between different rows, for a given toroidal harmonic. Quasi-linear initial value simulation, which is the focus of the present study, shows that application of the n = 3 ( n is the toroidal mode number) RMP field has a minimum side effect on the plasma core momentum confinement but potentially a large effect on the global particle transport. Generally, the RMP field with the best (worst) coil phasing for ELM control produces the strongest (weakest) effect on the plasma edge flow and the overall density. This robustly holds for all eight ITER scenarios. Consequently, in order to minimize the RMP induced side effects while achieving ELM control (suppression) in ITER, a compromise is necessary in choosing the coil current configuration.

Physics↗
DOE OSTI2025

Edge-localized mode mitigation enabled by active control of pedestal density gradient with new EAST tokamak divertor

Mitigation of large edge-localized modes (ELMs) has been achieved by actively reducing the pedestal density gradient with the EAST new right-angled lower divertor through changing the strike point position from the vertical target to the horizontal target. A series of dedicated experiments in the 2021–2024 EAST campaigns demonstrate that this ELM control solution is highly reproducible in a broad parameter space of edge safety factor q 95 = 4.7–7.1, heating power P total = 2.3–5 MW, and pedestal collisionality $ν_{e,ped}^{*}$ = 1–6, under both favorable and unfavorable magnetic configurations. Higher plasma density could facilitate the achievement of this ELM control solution. Statistical results indicate that the ELM mitigation effect can be observed at relatively larger Greenwald density fraction of f GW > 0.47. In addition, this ELM mitigation effect can be achieved with both lithium-coated and boronized metal walls. The pedestal density gradient is systematically lower in the horizontal target case than that of the vertical target case when the ELM mitigation effect can be observed. SOLPS-ITER simulation results indicate that the pedestal fueling from divertor recycling is significantly lower in the horizontal target case. This could contribute to the formation of a flattened pedestal density profile with small ELMs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

A Nuclear Magnetic Resonance Study of Cation and Anion Dynamics in Polymer–Ceramic Composite Solid Electrolytes

Polymer–ceramic composite electrolytes are promising for the application of all-solid-state lithium-ion batteries with high energy density and improved safety. In this work, we employ temperature-dependent 7 Li and 19 F NMR to examine the chemical environment and the dynamics of Li + cation and triflate anion in poly(ethylene oxide) (PEO)–lithium triflate (LiTf) polymer electrolyte and a composite electrolyte containing 55 vol % doped lithium aluminum titanium phosphate (LICGC). The line shape analysis and spin–lattice relaxation (T 1 ) measurements suggest both the Li + cations and the Tf – anions reside in two distinct environment: a mobile environment and an immobile environment. In the semicrystalline state, a large ratio (>70%) of immobile Li + and Tf – is observed. In the melt state, the ratio of mobile cations and anions significantly increases. With the presence of LICGC ceramic, an increased ratio of immobile Li + and Tf – and reduced mobility of the mobile components are observed, indicating that the ceramic negatively impacts the transport of both the cations and the anions. NMR characterizations are corroborated by conductivity results, which reveal that the intrinsic ionic conductivity of the polymer phase of the composite electrolyte was only 0.15 of the ceramic-free polymer electrolyte at room temperature. These results shed light on the challenges inherent to composite electrolytes with the goal of achieving both enhanced Li + transference number and ionic conductivity.

25 ENERGY STORAGE↗
DOE OSTI2026

First observation of currents induced by Alfven eigenmodes in a magnetic confinement device

Instabilities driven by energetic particles normally reduce fusion power but recent studies find Alfven eigenmode instabilities can drive ``zonal" flows and currents that reduce thermal transport, improving overall performance. Motional Stark effect (MSE) data from the DIII-D tokamak reveal changes in magnetic pitch angle during the phase when Alfven-eigenmode nonlinearity drives zonal flows, suppresses microturbulence, and increases the electron and ion temperatures. Finally, measured changes in safety factor of ∆q/q ≃ 5% occur in roughly 20 ms, consistent with a rough estimate based on the measured zonal flow.

Heidbrink, W. W. [Univ. of California, Irvine, CA ↗
DOE OSTI2023

Tailoring resonant magnetic perturbation to optimize fast-ion confinement during ELM control in KSTAR

Abstract 3D resonant magnetic perturbation (RMP) is one promising way to control edge localized modes that can cause excessive material erosion of tokamak first walls. However, RMP can lead to undesired degradation of plasma confinement, including fast-particle losses, which can impact the performance and safety of the reactor. This work investigates the optimization of the poloidal spectrum of the 3D field to optimize fast ion confinement during edge localized mode (ELM) suppression. In the initial step, the validity of the modeling framework is tested against experimental data. Simulations successfully replicate an increase in poloidal limiter temperature with different poloidal spectra. Then, the simulation shows improvement of fast ion confinement with a reduction of core resonant response, while edge resonant magnetic fields are maintained above the threshold to sustain the ELM suppression. Reduction of the core resonant fields keeps the Kolmogorov–Arnold–Moser surface and reduces the fast particle losses due to the stochastic magnetic field lines. The results highlight the potential of edge localization of the resonant fields to enhance the performance of fusion reactors, but further investigation is needed to improve the validation of this approach.

3D field↗
DOE OSTI2022

Prediction of divertor heat flux width for ITER pre-fusion power operation using BOUT++ transport code

Prediction of divertor heat flux width is performed for the first and the second pre-fusion power operation (PFPO) phases specified in the new ITER research plan using BOUT++ transport code (Li et al 2018 Comput. Phys. Commun.228 69–82). Here the initial plasma profiles inside the separatrix are taken from CORSICA scenario studies. Transport coefficients in transport code are calculated by inverting the plasma profiles inside the separatrix and are assumed to be constants in the scrape-off-layer. An anomalous thermal diffusivity scan is performed with E × B and magnetic drifts. The results in two scenarios identifying two distinct regimes: a drift-dominant regime when diffusivity is smaller than the respective critical thermal diffusivity χ c and a turbulence-dominant regime when diffusivity is larger than it. The Goldston heuristic drift model and the ITPA multi-machine experimental scaling yield a lower limit of the width λ q . From transport simulations, we obtain the critical thermal diffusivity χ c = 0.5 m 2 s –1 for the PFPO-1 scenario with toroidal magnetic field B = 1.77 T and plasma current I p = 5 MA, and χ c = 0.3 m 2 s –1 for the PFPO-2 scenario with toroidal magnetic field B = 2.65 T and plasma current I p = 7.5 MA. Separatrix temperature and collisionality also have a significant impact on the heat flux width in the drift-dominant regime. The investigation clearly yields a scaling for critical thermal diffusivity ${\chi }_{\text{c}}\propto {A}^{1/2}/(Z{\left(1+Z\right)}^{\frac{1}{2}}{B}_{\text{p}}^{2})$ using ITER scenarios with fixed safety factor q 95 , major radius R, aspect ratio R/a, and the separatrix temperature T sep , and establishes the connection with CFETR and C-Mod discharges. This scaling implies that for a given tokamak device with q 95 , R, R/a, and T sep fixed, a reduction of poloidal magnetic field by a factor of 3 leads to a 9 times higher critical value of thermal diffusivity χ c , possibly yielding a transition from turbulence- to drift-dominant regime.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗