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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 631 records · Page 35
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

The rate of development of atomic mixing and temperature equilibration in inertial confinement fusion implosions

The MARBLE project is a novel inertial confinement fusion platform for studying the development of atomic mixing and temperature equilibration in inertial confinement fusion implosions and their impact on thermonuclear burn. Experiments involve the laser-driven implosion of capsules filled with deuterated engineered foams whose pores are filled with a gaseous mixture of hydrogen and tritium. By varying the size of the foam pores, we can study the timescale of the development of atomic mix relative to the development of thermal equilibrium between species. In contrast, previous separated reactant experiments have only provided information on the total amount of mix mass. Additionally, we report on the series of MARBLE experiments [first reported in Haines et al., Nat. Commun. 11, 544 (2020)] performed on the University of Rochester's OMEGA laser facility and detailed and highly resolved three-dimensional radiation-hydrodynamic simulations of the implosions. In both the experimental and simulation results, we observe that the reactants do not achieve thermal equilibrium during the course of the implosion except in atomically mixed regions—i.e., that atomic mixing develops faster than thermal equilibration between species. The results suggest that ion temperature variations in the mixture are at least as important as reactant concentration variations for determining the fusion reaction rates.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

High Energy Density Physics of Inertial Confinement Fusion Ablator Materials (Final Technical Report)

The goal of this project was to conduct dynamic compression experiments and predictive simulations to reveal the fundamental high-energy-density (HED) physics of amorphous carbon. These results are essential for assessing amorphous carbon as a potential ablator material for next generation of inertial confinement fusion (ICF) capsules. We made significant progress in exploring the HED properties of amorphous carbon through experiments at Omega EP Laser and the European XFEL, in addition to billion-atom, quantum-accurate molecular dynamics (MD) simulations. Through our joint experimental and simulation program, we mapped the phase diagram of amorphous carbon, uncovering its range of metastability and identifying phase transitions to diamond and liquid carbon along the Hugoniot and at higher pressures using double shock compression pathways. Our findings indicate that amorphous carbon melts at significantly lower shock pressures than high-density carbon (diamond). However, nanocrystalline diamond nucleates across a broad range of pressures and temperatures. This emergence of the nanocrystalline microstructure during compression can negatively impact the planarity of the shock front and potentially trigger ablator/fuel mixing during Inertial Fusion Energy (IFE) applications.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

NIF-TAS Miniature Calibration Station

“On Aug. 8, 2021, an experiment at Lawrence Livermore National Laboratory’s (LLNL’s) National Ignition Facility (NIF) made a significant step toward ignition, achieving a yield of more than 1.3 megajoules (MJ). This advancement puts researchers at the threshold of fusion ignition, an important goal of the NIF, and opens access to a new experimental regime.” NIF experiments are conducted by symmetrically focusing 192 laser beams from NIF (three football fields in length) into the Target Chamber (10-meter diameter sphere) onto a target 2mm in size. The reaction produces a hot spot the diameter of a human hair, generating more than 10 quadrillion watts of fusion power for 100 trillionths of a second. These high pressures and temperatures created by the NIF have only been seen on the surface of the sun. To achieve fusion ignition, the target must be compressed to high pressures and temperatures, which requires symmetric compression. This symmetry requires precise alignment of the laser beams to the target. To achieve this precise symmetry a Target Alignment System (TAS) was developed to aid in aligning the beams to the target. The accomplishment on Aug 8 th could not have been accomplished without the TAS. TAS has proven to be a critical contributor to the success of NIF experiments. Each TAS is composed of several Charge Couple Device (CCD) sensors and various support electronics. Unfortunately, these CCDs are no longer available in industry as they have become outdated and obsolete. There is a limited amount of spare CCDs in the facility. As the NIF continues to increase its energy output of its fusion ignition shot experiments (heigh yield shots), more neutrons are being emitted from the Target Chamber. These neutrons are very damaging to the TAS electronics and its sensors, thus degrading and reducing its life expectancy dramatically. The increase of neutrons exceeds the damage threshold of the CCDs. This would increase the demand for more frequent TAS refurbishments and calibrations. The increase in resources to perform these refurbishments and calibrations would not be sustainable. In efforts to prolong the life of each TAS system it was determined that the TAS should be removed for every High Yield Shot and then reinstalled. The high precision fully calibrated TAS can easily be knocked out of alignment during these installs and reinstall exchanges and being transported in and out of the NIF. The Miniature Calibration Station System was developed to verify that the critical calibration parameters have been preserved during every TAS exchange. The Miniature Calibration Station System (Mini-Cal) would need to be a scaled down (in size and scope), portable version of the offline TAS calibration station to verify the critical TAS parameters are still within tolerances.

42 ENGINEERING↗
DOE OSTI2026

Lessons Learned from Open-Source Software Training Toward Expanding the Fusion Workforce

Within the United States (U.S.), the Fusion Innovation Research Engine (FIRE) Collaboratives seek to accelerate fusion technology development through wide-ranging community-driven research activities that bring together industry, laboratories, research institutes, and academia. Increasing the maturity of key components and systems for future fusion power plants (FPPs) toward commercialization, will, by necessity, increase the need for knowledgeable engineers, designers, and researchers in industry capable of integrating and further improving these technologies within industry FPP concepts. Further, various modeling and simulation packages support these programs and are under-development within them to drive design iteration and the development of FPP digital twins. To derive the greatest benefit from model output and capabilities, training these same specialists will be vital. Thus, the development of effective and accessible software onboarding and professional development opportunities focused on fusion will be key to keeping the pace of growth high in the coming decade and beyond. A similar need exists within the advanced fission reactor community, driven by an ever-growing need for carbon-free baseload power for industrial and data center applications. Within the U.S. and around the world, several open-source packages and frameworks exist to support this endeavor, and two we will highlight here are the Multiphysics Object Oriented Simulation Environment (MOOSE) framework and OpenMC. These packages, notably, are also being utilized in the FIRE Collaboratives program. In this presentation, we will highlight the lessons learned from over 30 years of combined software development and training experience, focused on developing strong technical software foundations within the nuclear workforce, from students to professional engineers & scientists. We will connect this experience to present and emerging needs in the fusion energy community, and, finally, will outline possible paths forward to develop a large, robust, global fusion workforce.

70 - PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Record Energetics for an Inertial Fusion Implosion at NIF

Inertial confinement fusion seeks to create burning plasma conditions in a spherical capsule implosion, which requires efficiently absorbing the driver energy in the capsule, transferring that energy into kinetic energy of the imploding DT fuel and then into internal energy of the fuel at stagnation. In this work, we report new implosions conducted on the National Ignition Facility (NIF) with several improvements on recent work [Phys. Rev. Lett. 120, 245003 (2018); Phys. Rev. E 102, 023210 (2020)]: larger capsules, thicker fuel layers to mitigate fuel-ablator mix, and new symmetry control via cross-beam energy transfer; at modest velocities, these experiments achieve record values for the implosion energetics figures of merit as well as fusion yield for a NIF experiment.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Liquid metal walls

Here, the plasma performance of fusion devices depends strongly on the chosen wall materials. Solid plasma-facing components (PFCs) are predominantly used in present devices, and are the most investigated candidates for fusion designs. High-Z materials such as tungsten (W) are the leading solid PFC material candidates. To date, a material choice that scales to steady-state reactor conditions has not been identified. Moreover, if plasma transient events such as edge-localized modes and disruptions cannot be altogether avoided or sufficiently mitigated, the projected peak heat and particle loads far exceed the power exhaust capabilities of solids. Liquid metal (LM) PFCs represent an intrinsic self-healing boundary that are both resilient to surface damage from transients, and that could handle high steady-state heat and particle fluxes. Flowing LM PFCs can be designed to remove “slag,” the buildup of material erosion due to plasma-material interactions, including charge exchange sputtering in the main chamber. Further, LM offer the prospect to manage hydrogenic species otherwise retained in the PFCs, which is important from a safety and inventory standpoint. The two most promising LMs are lithium (Li) and tin (Sn), although Sn–Li eutectics may be considered. While Sn offers a higher temperature window with low vapor pressure and low hydrogen retention, Li offers the prospect of enhanced energy confinement and higher acceptable core contamination limits, and this section focuses on Li PFCs. An LM PFC development research program developed LM PFC concepts for a nuclear fusion device via engineering design calculations, single-effect experiments, and staged prototypical experiments. A self-consistent design window was identified with liquid Li flow speeds ~5–10 m/s; plasma contamination was negligible for predicted Li evolution rates. While these preconceptual designs hold promise, there is substantial R&D needed to advance the technical readiness levels of LM PFCs for application to fusion power plants.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Using distributions to understand neutron and x-ray production in ICF ignition capsules and other high energy density plasmas

Here this paper describes how x-ray and neutron distribution functions can be useful tools to visualize the conditions measured in many types of plasma physics experiments. In particular, we model a standard inertial confinement fusion ignition capsule that consists of a Si doped plastic ablator surrounding a layer of deuterium–tritium (DT) ice as the yield varies from 18 kJ to 16.7 MJ and use the distribution functions to show that neutrons and high energy x rays (15 keV) are produced under similar conditions when the yield is low. However, as the capsule starts to support a propagating burn due to alpha heating, the x rays and neutrons are produced under somewhat different conditions in different parts of the plasma. In particular, the x-ray production takes place mainly in the hot plastic ablator for the full yield ignition capsule under quite different plasma conditions from the DT region producing the 14 MeV neutrons, which results in x-ray images with larger radii than the corresponding neutron images. These same distribution functions can be applied to many other plasma physics experiments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Stable Deuterium-Tritium plasmas with improved confinement in the presence of energetic-ion instabilities

Providing stable and clean energy sources is a necessity for the increasing demands of humanity. Energy produced by Deuterium (D) and Tritium (T) fusion reactions, in particular in tokamaks, is a promising path towards that goal. However, there is little experience with plasmas formed by D-T mixtures, since most of the experiments are currently performed in pure D. After more than 20 years, the Joint European Torus (JET) has carried out new D-T experiments with the aim of exploring some of the unique characteristics expected in future fusion reactors, such as the presence of highly energetic ions in low plasma rotation conditions. A new stable, high confinement and impurity-free D-T regime, with reduction of energy losses with respect to D, has been found. Multiscale physics mechanisms critically determine the thermal confinement. These crucial achievements importantly contribute to the establishment of fusion energy generation as an alternative to fossil fuels.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Numerical modeling of impurity powder injection in W7-X

In this work, we present numerical simulation results of powder injection in W7-X using the EMC3-EIRENE and DIS codes. First, we model powder injection experiments performed in W7-X with the Probe Mounted Powder Injector. The simulation results qualitatively agree with visible imaging measurements. Secondly, we perform predictive simulations to guide the installation of an Impurity Powder Dropper in W7-X, allowing to choose in between several available non-vertical ports to maximize the amount of powder penetrating into the plasma, as well as the verticality of the port, to minimize sticking of the powders in the in-vessel stainless steel guiding tube. Port AEM41 is selected as the best candidate for IPD installation. The robustness of the simulation results has been verified for different plasma densities, powder materials and sizes, powder friction coefficient and changes in the plasma flow.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

The development of a high-resolution Eulerian radiation-hydrodynamics simulation capability for laser-driven Hohlraums

Hohlraums are hollow cylindrical cavities with high-Z material walls used to convert laser energy into uniform x-ray radiation drives for inertial confinement fusion capsule implosions and high energy density physics experiments. Credible computational modeling of hohlraums requires detailed modeling and coupling of laser physics, hydrodynamics, radiation transport, heat transport, and atomic physics. We report on improvements to Los Alamos National Laboratory's xRAGE radiation-hydrodynamics code in order to enable hohlraum modeling. xRAGE's Eulerian hydrodynamics and adaptive mesh refinement make it uniquely well suited to study the impacts of multiscale features in hohlraums. In order to provide confidence in this new modeling capability, we demonstrate xRAGE's ability to produce reasonable agreement with data from several benchmark hohlraum experiments. We also use xRAGE to perform integrated simulations of a recent layered high density carbon capsule implosion on the National Ignition Facility in order to evaluate the potential impacts of the capsule support tent, mixed cell conductivity methodologies, plasma transport, and cross-beam energy transfer (XBT). We find that XBT, seeded by plasma flows in the laser entrance hole (LEH), causes a slight decrease in energy coupling to the capsule and that all of these impact the symmetry of the x-ray drive such that they have an appreciable impact on the capsule implosion shape.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Description of FY25 Theory and Simulation Performance Target: Development of an integrated modeling framework for fusion reactor design and assessment

The urgency to deliver fusion power is growing now more than ever, with increasing pressure for both public programs and private companies to meet milestones timelines and overcome significant remaining technical challenges to ensure growth of a nascent fusion industry in time to meet rapidly growing clean energy demands. With incredible advancements in computation and years of investment in fusion model development and validation, integrated modeling is poised to fill a key role in accelerating the timeline to a fusion pilot plant (FPP). Future fusion pilot plants will operate in regimes far beyond current experience, and device design will rely on physics-based prediction and extrapolation. Many concepts will also rely on simulation to assess safety (shielding, tritium management, materials activation and lifetimes), economics and scalability before the decision to build. Importantly, integrated simulation can be used to reveal and solve the complexities of system integration that may otherwise not be apparent in physical components or models developed in isolation. New experimental test facilities that produce relevant conditions to validate and resolve key technical challenges for various subsystems (materials, blankets, fuel cycle, etc.) have been repeatedly called for by the fusion community but are not yet realized. Integrated modeling has an important role in identifying realistic load conditions (thermal, electromagnetic, plasma, neutron and photon loads, etc.) and defining the components and experiments for these test facilities in order to ensure meaningful validation that sufficiently reduces modeling uncertainties and technical risk for the full integrated reactor. The Fusion REactor Design and Assessment (FREDA) SciDAC project is building a component-based integrated modeling framework & data structure to enable self-consistent, multi-fidelity, iterative optimization workflows for the fusion reactor design process. FREDA aims to shorten the time to viable designs by providing a set of flexible workflows to support the various stages of the design process using an integrated model hierarchy, ranging from the simple analytic descriptions to the highest fidelity, theory-based plasma and engineering modeling developed by the fusion and fission communities. These tools are expected to be needed for timely support of FPP design in the milestone program and in the FIRE collaboratives. The plasma simulation backbone of FREDA is IPS-FASTRAN with newly developed coupled Core-Edge Pedestal-SOL (CESOL) workflows, which is being extended to the far-SOL region up to the plasma facing components. FREDA incorporates the FERMI engineering modeling suite and will enable self-consistent evaluation of the thermal shields, limiters, blanket, magnets, and other surrounding structures with predictions of temperatures, erosion, dpa, activation, tritium generation and transport, creep, corrosion, material degradation, etc. Parametric generation of 3D CAD enables rapid iteration of component geometry in response to plasma and loading specifications.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

DOE SCGSR Final Report - Project Accomplishments and Additional Materials

The major accomplishments of my project were successfully developing and testing the laser interferometry diagnostic on the Plasma Liner Experiment (PLX) at Los Alamos National Laboratory (LANL). This diagnostic helped advance my research by providing a simpler way to obtain electron density measurements of magnetized plasmas compared to other electron density diagnostics such as triple Langmuir probes. The laser used was a 561 nm continuous wave (CW) diode-pumped solid state (DPSS) laser. The PLX laser interferometry diagnostic consists of two parts. These include the launching and receiving sides of the laser diagnostic. The launching side consists of the main laser beam from the DPSS laser being split into five chords(probe beams) and a reference beam which are then directed with fiber optic couplers into fiber optic cables which transmit the probe beamsto the PLX vacuum chamber. The probe beams then pass through the plasma in the chamber and into the receiving side optics where they are again directed through fiber optic cables to be combined with the reference beam. The combined beams are transmitted via multi-mode fiber optic cables to photodiodes to convert the light signals into electrical signals. Before the electrical signals get digitized, they pass through bandpass and low pass filters to eliminate electromagnetic noise and unwanted frequencies. The electrical signals are then processed by IQ demodulators to determine phase angle difference between the probe and reference beams for each chord in order to calculate line-integrated electron density.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Low mode implosion symmetry sensitivity in low gas-fill NIF cylindrical hohlraums

Achieving an efficient capsule implosion in National Ignition Facility indirect-drive target experiments requires symmetric hohlraum x-ray drive for the duration of the laser pulse. This is commonly achieved using two-sided two-cone laser irradiation of cylindrical hohlraums that, in principle, can zero the time average of all spherical harmonic asymmetry modes <6 as well as the time dependence of the usually dominant mode 2. In practice, experimental evidence indicates that maintaining symmetric drive becomes limited late in the pulse due to the inward expansion of the hohlraum wall and outward expansion of the capsule ablator plasmas impairing the propagation of the inner-cone laser beams. This effect is enhanced in hohlraums employing low gas-fill, now used almost exclusively as these provide the highest performing implosions and reduce Stimulated Brillouin and Raman backscatter losses, since the gas plasma provides less back pressure to limit blow-in of the hohlraum wall and capsule ablator plasmas. In order to understand this dynamic behavior, we combined multi-keV X-ray imaging of the wall and imploded fuel plasmas as we changed a single parameter at a time: hohlraum gas-fill, laser outer cone picket energy, radius of high density carbon capsules used, and laser beam polar and azimuthal pointing geometry. In this work, we developed a physics-based multi-parameter experimental scaling to explain the results that extend prior scalings and compare those to radiation hydrodynamic simulations to develop a more complete picture of how hohlraum, capsule, and laser parameters affect pole vs equator drive symmetry.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Counter-propagating toroidal Alfvén eigenmodes in tokamaks

Mechanisms of destabilization of toroidal Alfvén eigenmodes (TAEs) in tokamaks are analyzed with the aim to reveal those leading to modes propagating in the direction opposite to plasma current, i.e. counter-propagating modes (ctr-TAE). Plasmas with fast-ions sources [such as neutral beam injection (NBI), ion cyclotron resonance heating, fusion reactions] and without them (Ohmic discharges) are considered. A particular NSTX-U experiment with NBI, where co- and counter-propagating TAEs were observed simultaneously (Podestà et al 2018 Nucl. Fusion 58 082023), is considered. It is concluded that both types of TAEs occurred because their destabilization was caused by the velocity anisotropy of beam ions, which overrode effects of spatial inhomogeneity of these ions.

Alfvénic instabilities↗
DOE OSTI2021

Data on erosion and hydrogen fuel retention in Beryllium plasma-facing materials

ITER will use beryllium as a plasma-facing material in the main chamber, covering a total surface area of about 620 m 2 . Given the importance of beryllium erosion and co-deposition for tritium retention in ITER, significant efforts have been made to understand the behaviour of beryllium under fusion-relevant conditions with high particle and heat loads. This paper provides a comprehensive report on the state of knowledge of beryllium behaviour under fusion-relevant conditions: the erosion mechanisms and their consequences, beryllium migration in JET, fuel retention and dust generation. The paper reviews basic laboratory studies, advanced computer simulations and experience from laboratory plasma experiments in linear simulators of plasma–wall interactions and in controlled fusion devices using beryllium plasma-facing components. A critical assessment of analytical methods and simulation codes used in beryllium studies is given. The overall objective is to review the existing set of data with a broad literature survey and to identify gaps and research needs to broaden the database for ITER.

Erosion–deposition↗
DOE OSTI2024

Self-driven ion deflectometry measurements using MeV fusion-driven protons and accelerated deuterons in the deuterated hybrid x-pinch on the MAIZE LTD generator

Abstract We report on the results of point-projection ion deflectometry measurements from a mid-size university z-pinch experiment. A 1 MA 8 kJ LTD generator at the University of Michigan (called MAIZE) drove a hybrid x-pinch (HXP) with a deuterated polyethylene fiber load to produce a point-like source of MeV ions for backlighting. In these experiments, 2.7 MeV protons were generated by DD beam-target fusion reactions. Due to the kinematics of beam-target fusion, the proton energies were down-shifted from the more standard 3.02 MeV proton energy that is released from the center-of-mass rest frame of a DD reaction. In addition to the 2.7 MeV protons, strongly anisotropic beams of 3 MeV accelerated deuterons were detected by ion diagnostics placed at a radial distance of 90 mm from the x-pinch. Numerical reconstruction of experimental data generated by deflected hydrogen ion trajectories evaluated the total current in the vacuum load region. Numerical ion-tracking simulations show that accelerated deuteron beams exited the ion source region at large angles with respect to the pinch current direction.

Physics↗
DOE OSTI2022

An assessment of generating quasi-static magnetic fields using laser-driven “capacitor” coils

Over the previous decade, numerous experiments have been performed using a laser to drive a strong, quasi-static magnetic field. Field strength and energy density measurements of these experiments have varied by many orders of magnitude, painting a confusing picture of the effectiveness of these laser-driven coils (LDCs) as tools for generating consistent fields. At the higher end of the field energy spectrum, kilotesla field measurements have been used to justify future experimental platforms, theoretical work, and inertial confinement fusion concepts. In this paper, we present the results from our own experiments designed to measure magnetic fields from LDCs as well as a review of the body of experiments that have been undertaken in this field. Here, we demonstrate how problems with prior diagnostic analyses have led to overestimations of the magnetic fields generated from LDCs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Parametric Study of the Vacuum Permeator for the Tritium Extraction eXperiment

Tritium breeding is a critical component of any self-sustaining future fusion reactor. The liquid metal eutectic, PbLi, is of particular interest as a tritium breeder material due to its favorable thermophysical and neutronic properties. One of the several remaining challenges facing PbLi breeder blankets is the need to design and validate a highly efficient tritium extraction system. The vacuum permeator is a promising extraction concept that utilizes tritium permeation through a highly permeable metal membrane. The Tritium Extraction eXperiment (TEX) is a forced-convection PbLi loop constructed to investigate tritium extraction from PbLi with vacuum permeators. Accurate thermal-hydraulic and tritium transport models are required to establish appropriate test matrices, predict experiment outcomes, and analyze data. However, the hydrogen transport properties of PbLi and permeator materials have large uncertainties. A database is collected, and a parametric analysis is conducted on the effect of hydrogen transport material properties: diffusivity of H in PbLi and permeator, solubility of H in PbLi and permeator, and permeator surface recombination constant on the expected tritium extraction efficiency for a vacuum permeator installed in TEX. As a result, we observe solubility of H in PbLi and the permeator and the recombination constant of the permeator have the largest effect on the extraction efficiency.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗