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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 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

First graded metal pushered single shell capsule implosions on the National Ignition Facility

Graded metal pushered single shell (PSS) capsules are predicted to be a viable alternative to low-Z capsule indirect drive inertial confinement fusion (ICF) implosions for achieving high fusion yields [MacLaren et al., Phys. Plasmas 28, 122710 (2021)]. The first experiments with Be/Cr-graded metal PSS capsules indicate that the implementation of the principle design feature, the graded density inner metal layer, has succeeded in producing a stable implosion with performance in agreement with predictions. With 50% Cr concentration in the pusher, PSS capsules have greater than ~2× higher shell densities during stagnation for enhanced core confinement and radiation trapping at ~35% lower shell implosion velocities than low-Z capsules. High-energy >30 keV inflight shell radiography recorded 215 km/s implosion velocities and show that implosion Legendre mode P 2 asymmetry can be tuned via inner-to-outer beam wavelength separation, similar to other implosions. Shell radiographs and neutron core images show similar P 2 asymmetry, suggesting no symmetry swings between peak implosion velocity and stagnation times. Despite the modest implosion velocities, gas-filled deuterium–tritium capsule implosions generate 10 15 neutron yields at relatively modest core ion temperatures of 2.75 keV, indicating that in spite of the high-density inner layer, the implosions have been stabilized by the design density gradient. When compared with hydrodynamic simulations, the measured yield-over-simulated is 35% due to fuel–pusher mix and other perturbations such as the capsule fill tube. Finally, simple analytical scalings of hot spot pressure and neutron yield show that PSS implosions reach similar performance at lower implosion velocities and higher shell densities to low-Z ICF capsules.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Inferences of hot electron preheat and its spatial distribution in OMEGA direct drive implosions

Hot electrons generated from laser plasma instabilities degrade performance of direct drive implosions by preheating the deuterium and tritium (DT) fuel resulting in early decompression and lower areal densities at stagnation. A technique to quantify the hot electron preheat of the dense DT fuel and connect it to the degradation in areal density is described in detail. Hot electrons are measured primarily from the hard x-rays they emit as they slow down in the target. The DT preheat is inferred from a comparison of the hard x-ray signals between a DT-layered implosion and its mass equivalent ablator only implosion. The preheat energy spatial distribution within the imploding shell is inferred from experiments using high Z payloads of varying thicknesses. It is found that the electrons deposit their energy uniformly throughout the shell material. For typical direct-drive OMEGA implosions driven with an overlapped intensity of ∼9·1014 W/cm2, approximately ∼0.02%–0.03% of the laser energy is converted into preheat of the stagnated fuel which corresponds to areal density degradations of 10%–20%. The degradations in areal density explain some of the observed discrepancies between the simulated and measured areal densities.

Christopherson, A. R. (ORCID:000000033578635X)↗
DOE OSTI2024

Needed computations and computational capabilities for stellarators

Stellarator plasmas are externally controlled to a degree unparalleled by any other fusion concept, magnetic or inertial. This control is largely through the magnetic fields produced by external coils. The development of fusion energy could be expedited by carrying out remarkably straight-forward computations to define strategies for exploiting this external control. In addition to these computations, which have a reliability limited only by competence, certain physics areas that affect the development of stellarator power plants should have a more intense study. The low cost and speed with which computations can be carried out relative to experiments have implications for the development of fusion. Computations should be used to develop a strategy that to the extent possible allows major issues to be circumvented. Required computations for this strategy are the subject of this paper.

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 OSTI2021

Total fusion yield measurements using deuterium–tritium gamma rays

The deuterium–tritium (DT) fusion reaction, D ( T , γ ) 5 He , has a rare (4.2 × 10 – 5 ) branching ratio that releases a ~16 MeV gamma ray instead of the more common 14.1 MeV fusion neutron. These fusion gamma rays can be used as a complementary source of measuring the net fusion yield of experiments. Fusion gamma rays have the advantage of having negligible Doppler broadening, being isotropic and less down-scattered compared to fusion neutrons. At the National Ignition Facility, we measured the DT fusion gamma rays by the Gamma Reaction History diagnostic that thresholds to only measure >10 MeV gamma rays. After removing a ~12% contribution from neutron capture interactions in deuterium, D ( n , γ ) T , and carbon, 12 C ( n , γ ) 13 C , the deviation between GRH measurements and neutron-based yield measurements has been reduced to 10%. Further improvements are needed for detailed physics comparison.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Investigating the impact of intermediate-mode perturbations on diagnosing plasma conditions in DT cryogenic implosions via synthetic x-ray Thomson scattering

The pursuit of inertial confinement fusion ignition target designs requires precise experimental validation of the conditions within imploding capsules, in particular the density and temperature of the compressed shell. Previous work has identified x-ray Thomson scattering (XRTS) as a viable diagnostic tool for inferring the in-flight compressed deuterium-tritium shell conditions during capsule implosions (Poole et al 2022 Phys. Plasmas 29 072703). However, this study focused on one-dimensional simulations, which do not account for the growth of hydrodynamic instabilities. In this work, two-dimensional DRACO simulations incorporating intermediate-mode perturbations up to Legendre mode l = 50 were used to generate synthetic XRTS spectra with the SPECT3D code. The analysis employed Markov-Chain Monte Carlo techniques to infer plasma conditions from these spectra. The results demonstrate that the XRTS diagnostic platform can effectively discern the in-flight compressed shell conditions for targets with varying adiabats, even in the presence of intermediate-mode perturbations. This work underscores the potential of XRTS for realistic inertial confinement fusion experiments, providing a robust method for probing the complex dynamics of fusion implosions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Self-consistent investigation of density fueling needs on ITER and CFETR utilizing the new Pellet Ablation Module

Abstract Self-consistent modeling using the stability, transport, equilibrium, and pedestal (STEP) workflow in the OMFIT integrated modeling framework (predicting pedestal with EPED, core profiles with TGYRO, current profile with ONETWO, and EFIT for equilibrium) suggests ITER and future devices such as China Fusion Engineering Test Reactor (CFETR) Zhuang et al (2019 Nucl. Fusion 59 112010) will benefit from high-density operation (Greenwald limit fraction f g w ≈ 0.7−1.3). Regimes with an operational density near the Greenwald limit will likely need peaked density profiles so that the pedestal density remains below the Greenwald limit. Peaked density profiles can be achieved with the help of pellet injection. A flexible Pellet Ablation Module (PAM), which predicts the density source based on a comprehensive analytical pellet ablation model, has been developed for predicting pellet fueling for transport studies, and has been incorporated into the STEP workflow for predictive modeling. This workflow is applied to DIII-D and finds good agreement with experiments. On ITER the effect of pellet fueling is examined in an advanced inductive scenario, where a fusion gain of up to Q = 9 is predicted with strong central pellet fueling. On CFETR, with a mid-radius density source, an average of 1.5 × 10 22 electrons s −1 are required to achieve the density and temperature profiles necessary for the 1000 MW advanced scenario with a tritium burn-up fraction of ∼ 3 % .

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

LLE Diagnostic Resource Team for Innovative Fusion Concepts (Final Report)

The LLE Diagnostic Resource Team for Innovative Fusion Concepts provided travelling neutron diagnostics, consultancy on neutron diagnostics, and consultancy on tritium handling to fusion companies. An existing neutron detector was recalibrated for low yields and two detectors were built based on designs proven on the Omega Laser Facility. The three detectors allow measurements over a wide range of yields (from 10 to 1E7 incident neutrons). The detectors were calibrated for DD neutrons on a dedicated OMEGA shot day. Neutron detectors from the company MIFTI were also calibrated on OMEGA. MIFTI is currently using their detectors and two of our detectors on staged Z-pinch experiments at UCSD. The Team acted as consultants for the development of a liquid scintillator array to detect neutrons from muon induced fusion by NK Labs. Segments of NK Labs’ detector were tested at the LLE using radioactive sources and a DD neutron generator. The Team also acted as consultants for the design of a tritium handling system for NK Labs. A new type of neutron spectrometer was designed and built, and has completed preliminary testing.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Development of an ab initio learned model of electron deposition range in deuterium-tritium plasmas through time-dependent density functional theory calculations and machine learning

Accurate hydrodynamic modeling for laser-direct-drive (LDD) inertial-confinement-fusion (ICF) relies on precise calculations of the electron thermal conduction in all target materials. The nonlocal stopping range of electrons in ICF plasmas directly influences thermal conduction; yet, no first principles model exists for the electron mean free path in the conduction-zone regime. This work utilized time-dependent stochastic density-functional theory (TD-sDFT) to calculate the electron stopping power in deuterium-tritium (DT) plasmas at (ρ, T) conditions relevant to the conduction zone and the compressed shell in ICF. Using a combination of our TD-sDFT data and already established analytical models, we developed and trained an artificial neural network to create a global model for the nonlocal electron deposition range, λ E . We compared our machine-learning (ML) based model for λ E to the currently-used modified-Lee-More model in LDD radiation-hydrodynamic codes, such as lilac, and saw an overall decrease in the deposition range. To understand the effects of λ E on LDD ICF implosion dynamics, we implemented the ML-based model into lilac; specifically, we looked at designs consistent with a current experiment on the OMEGA laser and for a newly designed LDD-ICF target for the future OMEGA-Next facility. In both cases, we saw an overall drop in predicted ablation pressure, peak areal density, and neutron yield due to the reduced thermal conduction (smaller λ E ) in DT plasmas. Comparisons with the experiment on OMEGA are also made.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Improvements to the Faraday cup fast ion loss detector and magnetohydrodynamic induced fast ion loss measurements in Joint European Torus plasmas

Upgrades to electronic hardware and detector design have been made to the JET thin-foil Faraday cup fast ion loss detector in anticipation of the upcoming deuterium–tritium (DT) campaign. An improved foil stack design has been implemented, which greatly reduces the number of foil-to-foil shorts, and triaxial cabling has mitigated ambient noise pickup. Initial tests of 200 kHz digitizers, as opposed to the original 5 kHz digitizers, have provided enhanced analysis techniques and direct coherence measurements of fast ion losses with magnetohydrodynamic activity. Here, we present recent loss measurements in JET deuterium plasmas correlated with kink modes, fishbone modes, edge-localized modes, and sawteeth. Sources of systematic noise are discussed with emphasis on capacitive plasma pickup. Overall, the system upgrades have established a diagnostic capable of recording alpha particle losses due to a wide variety of resonant fast ion transport mechanisms to be used in future DT-experiments and modeling efforts.

47 OTHER INSTRUMENTATION↗
DOE OSTI2024

A Physical Metric for Inertial Confinement Fusion Capsules

The performance of fusion capsules on the National Ignition Facility (NIF) is strongly affected by the physical properties of the hot deuterium–tritium (DT) fuel, such as the mass, areal density, and pressure of the hot spot at the stagnation time. All of these critical quantities depend on one measured quantity, which is the ratio of the specific peak implosion energy to the specific internal energy of the hot spot. This unique physical quantity not only can measure the incremental progress of the inertial confinement fusion capsules towards ignition but also measures the conversion of the peak implosion kinetic energy of the pusher shell into the internal energy of the hot fuel in a capsule. Analysis of existing NIF shots to date are performed. The ratio metric is compared quantitatively with the ignition criterion. Results provide new perspectives on the NIF experiments by which the performance of the burning plasma can be determined and controlled through the fine tune of the implosion parameters, which improves future designs and predictions of the ignition capsules.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

High-volume and -adiabat capsule (“HVAC”) ignition: Lowered fuel compression requirements using advanced Hohlraums

Lower-than-expected deuterium–tritium fuel areal densities have been experimentally inferred across a variety of high-convergence, nominally low-adiabat implosion campaigns at the National Ignition Facility (NIF) using cylinder-shaped Hohlraums [Hurricane et al., Phys. Plasmas 26, 052704 (2019)]. A leading candidate explanation is the presence of atomic mix between the fuel and ablator from hydrodynamic instability growth [Clark et al., Phys. Plasmas 26, 050601 (2019)], leading to reduced fuel compressibility and an effectively higher (in-flight) fuel adiabat α. Tolerating a high-α implosion can be obtained with significantly higher capsule absorbed energy Ecap according to the one-dimensional (1-D) ignition-threshold-factor analytic scaling [S. Atzeni and J. Meyer-ter-Vehn, Nucl. Fusion 41, 465 (2001)], ITF∼Ecap·α−1.8. Recent experiments with large Al shells in rugby-shaped Hohlraums have established high laser-capsule coupling efficiencies of ≽ 30% [Ping et al., Nat. Phys. 15, 138 (2019)], enabling a path to Ecap≽ 0.5 MJ at the NIF and increased performance margin M ≡ ITF − 1. The ability to operate at high adiabat with large capsules using nonstandard Hohlraums leads to the predicted onset of a volume-ignition mode, defined as when both the entire fuel is the “hot spot” and inertial confinement is principally provided by the ablator compared with the compressed fuel. Such an ignition mode, normally reserved for high-Z targets, e.g., double shells [Amendt et al., Phys. Plasmas 14, 056312 (2007)], is predicted to lead to lower fuel convergence and less exposure to mix due to the intended high adiabat—but at the expense of ∼3–4 × reduced (1-D) yield compared with conventional central hot-spot ignition designs.

Amendt, Peter (ORCID:0000000174021777)↗
DOE OSTI2024

Use of tritium-rich fuel to improve the yield of layered deuterium/tritium inertial fusion capsules

In deuterium–tritium (DT) ice layered implosions, nearly all hot spot mass at peak burn comes from the dense fuel. Accurate prediction of the fuel mass ablation, including the enthalpy associated with mass inflow into the hot spot from the dense fuel, is essential to understanding the energetics and ignition of the hot spot in layered implosions. A recently published boundary layer analysis (Daughton et al., 2023) indicates a faster mass ablation rate than in previous analyses of layered implosions. Inclusion of this effect provides a better match to simulations and leads to a new ignition threshold where the temperature of the dense fuel plays a critical role. This analysis motivates possible new directions for improved capsule performance. Here, the authors present evidence in support of one such approach: the use of tritium-rich ice to decrease 14 MeV neutron scattering and heating of the dense fuel, resulting in less mass ablation and more robust burn of the hot spot. It is found from numerical simulations that despite a less favorable D:T ratio in the ice, the use of a 40:60 D:T ratio leads to an increase in capsule yield of 17% percent compared with that of a 50:50 D:T ratio fuel for capsules resembling those of the recent N210808 ignition experiment on the NIF (Abu-Shawareb et al., 2022) and an increase of 74% compared with that of a 60:40 D:T ratio fuel capsule. Finally, these results are potentially important for modeling all layered implosions, since some degree of DT fractionization may arise naturally during the beta layering process. In addition, this physics is important for the feasibility of high-gain capsule designs that seek to minimize tritium usage, as in some inertial fusion energy concepts.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Alfvén eigenmode stability in a JET afterglow deuterium plasma and projections to deuterium–tritium plasmas

Abstract The performance of fusion devices relies strongly on the good confinement of energetic particles (EPs). Therefore, the investigation of EP transport by magnetohydrodynamic instabilities is one of the key aspects in the development of plasma scenarios. Alfvénic instabilities in particular can lead to significant losses of alpha particles that are essential for plasma self-heating. A so-called afterglow scheme has been developed to study the destabilization of Alfvén eigenmodes (AEs) by alpha particles and associated EP transport in the JET tokamak. In this work, the linear stability of AEs is discussed for the partial afterglow phase in a JET deuterium plasma discharge and for the full afterglow phase in a projected deuterium–tritium (DT) plasma. Thanks to recent upgrades in the tokamak transport code TRANSP, one can account for the contributions of different EP species to mode stability. Analysis of deuterium plasmas shows that AE growth rates are extremely sensitive to the energy and distribution of fast ions. An increase in fast ion energy can lead to more unstable AEs. In the afterglow phase of projected DT plasmas, it is EPs that mostly drive the AEs. However, the drive by alpha particles is comparable to that by beam ions and their contribution to the net growth rate might be hard to separate. According to the discussed projections, the destabilization of AEs might be ineffective because the background plasma damping significantly exceeds the EP drive. In this case, the development of an alternative plasma scenario that allows us to overcome such damping would be required in future experiments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2026

Stellarator Theory and Non-Axisymmetric Shaping (Final Technical Report)

Divertors are a critical concept for all magnetically confined fusion systems. The W7-X stellarator has an island divertor as do many of the privately funded stellarator reactor designs, but non-resonant divertors have many advantages, which are defined in the paper OSTI ID: 3021689, “Stellarators with enhanced tritium confinement and edge radiation control” and discussed in OSTI ID: 3021687, 3021688, 3021692, and 3021695. The three basic advantages are: (1) Resilience, the location at which the diverted plasma reaches the walls is insensitive to plasma conditions, unlike island divertors. (2) The width of the region at the plasma edge that flows into the divertor has a controllable width. It can be made sufficiently broad to avoid neutrals impending on the plasma from charge exchanging with high energy plasma particles. The resulting high energy neutrals erode the walls. (3) The controllable width and confinement of non-resonant divertors allow high-Z impurities to be added. The shortness of their confinement time limits their diffusion into the plasma interior. High-Z impurity radiation at the plasma edge seems the only way to sufficiently spread the outcoming power over the walls. The lack of experiments is a major impediment to application of non-resonant divertors, although the small experiment STAR_LITE is being built at Hampton University to study them.

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↗
DOE OSTI2021

Fusion gain from cylindrical liner-driven implosions of field reversed configurations

MagLIF experiments [M.R. Gomez et al., Phys. Plasmas 22, 056306 (2015)] on Z have demonstrated the basic principles of Magneto-Inertial Fusion (MIF) for wall confined plasmas. Other MIF schemes have been proposed based on the liner implosion of closed field magnetically confined plasmas such as Field Reversed Configurations (FRCs) [T. P. Intrator et al., Phys. Plasmas 15, 042505 (2008)]. We present a semi-analytical model of liner driven FRC implosions that predicts the fusion gain of such systems. The model predicts a fusion gain near unity for an FRC imploded by a liner driven by the Z Machine. We show that FRCs could be formed and imploded at the Z facility using the AutoMag liner concept [S. A. Slutz et al., Phys. Plasmas 24, 012704 (2017)]. Furthermore, an initial bias magnetic field can be supplied by the external magnets used in MagLIF experiments. The reverse field is then supplied by an AutoMag liner, which has helical conducting paths imbedded in an insulating substance. Experiments [Shipley et al., Phys. Plasmas 26, 052705 (2019)] have demonstrated that AutoMag can generate magnetic fields greater than 30 Tesla inside of the liner. We have performed 2D Radiation MHD simulations of the formation and implosion of an FRC, which are in good agreement with the analytical model. The FRC formation process could be studied on small pulsed power machines delivering about 1 MA.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Optimal Chloride Salt Mixture for a Fusion Blanket

Deuterium-tritium fusion reactors cannot operate for a significant period without a closed tritium fuel cycle, according to a recent National Academies of Sciences (NAS) report on bringing fusion reactors to the US electrical grid. This fact places breeder blankets as one of the foundational systems for self-sustained fusion reactor operation. The typical functional requirements for a breeder blanket system include producing tritium, absorbing kinetic energy, transporting thermal energy, and being environmentally attractive. State-of-the-art research on liquid blankets has converged to primarily focus on (LiF) 2 and BeF 2 (FLiBe) molten salts and a metallic eutectic of lead and lithium (PbLi), but “virtually all of the technologies related to the tritium fuel cycle are at a low technological readiness level”. This work sought to explore optimum blanket configurations as it aligned with the Oak Ridge National Laboratory (ORNL) FY 2023 Laboratory Directed Research and Development Program’s research priority of developing and expanding the current understanding of fusion blanket science and technology. This purpose of this work was to address ORNL research priorities and NAS recommendations by investigating novel liquid blanket materials that could provide self-sustaining operation and draw on experience from research on molten salts used for advanced fission reactors, concentrated solar, and thermal energy storage. The hypothesis when proposing this research was that there could be chloride-based blanket designs that can exceed the tritium breeding ratios of (FLiBe) molten salt blankets while reducing the use of Be (FLiBe), avoiding the generation of HF (FLiBe), and minimizing magnetohydrodynamic (MHD)-perturbed flow fields (PbLi). The fastest and most cost-effective path to deploying liquid fusion breeder blankets could be from maximizing the synergistic technological overlap between fusion, fission, concentrated solar, and thermal energy storage industries.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗