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At least 1,729 records · Page 96
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Experimental quantification of the impact of heterogeneous mix on thermonuclear burn

In inertial confinement fusion, deuterium–tritium (DT) fuel is brought to densities and temperatures where fusion ignition occurs. However, mixing of the ablator material into the fuel may prevent ignition by diluting and cooling the fuel. MARBLE experiments at the National Ignition Facility provide new insight into how mixing affects thermonuclear burn. These experiments use laser-driven capsules containing deuterated plastic foam and tritium gas. Embedded within the foam are voids of known sizes and locations, which control the degree of heterogeneity of the fuel. Initially, the reactants are separated, with tritium concentrated in the voids and deuterium in the foam. During the implosion, mixing occurs between the foam and gas materials, leading to DT fusion reactions in the mixed region. Here, it is shown that by measuring the ratios of DT and deuterium–deuterium neutron yields for different macropore sizes and gas compositions, the effects of mix heterogeneity on thermonuclear burn may be quantified, supporting an improved understanding of these effects.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Divertor shaping with neutral baffling as a solution to the tokamak power exhaust challenge

Exhausting power from the hot fusion core to the plasma-facing components is one fusion energy’s biggest challenges. The MAST Upgrade tokamak uniquely integrates strong containment of neutrals within the exhaust area (divertor) with extreme divertor shaping capability. By systematically altering the divertor shape, this study shows the strongest evidence to date to our knowledge that long-legged divertors with a high magnetic field gradient (total flux expansion) deliver key power exhaust benefits without adversely impacting the hot fusion core. These benefits are already achieved with relatively modest geometry adjustments that are more feasible to integrate in reactor designs. Benefits include reduced target heat loads and improved access to, and stability of, a neutral gas buffer that ‘shields’ the target and enhances power exhaust (detachment). Analysis and model comparisons shows these benefits are obtained by combining multiple shaping aspects: long-legged divertors have expanded plasma-neutral interaction volume that drive reductions in particle and power loads, while total flux expansion enhances detachment access and stability. Containing the neutrals in the exhaust area with physical structures further augments these shaping benefits. These results demonstrate strategic variation in the divertor geometry and magnetic topology is a potential solution to one of fusion’s power exhaust challenge.

magnetically confined plasmas↗
DOE OSTI2022

Phased plan for the implementation of the time-resolving magnetic recoil spectrometer on the National Ignition Facility (NIF)

The time-resolving magnetic recoil spectrometer (MRSt) is a transformative diagnostic that will be used to measure the time-resolved neutron spectrum from an inertial confinement fusion implosion at the National Ignition Facility (NIF). It uses a CD foil on the outside of the hohlraum to convert fusion neutrons to recoil deuterons. An ion-optical system positioned outside the NIF target chamber energy-disperses and focuses forward-scattered deuterons. A pulse-dilation drift tube (PDDT) subsequently dilates, un-skews, and detects the signal. While the foil and ion-optical system have been designed, the PDDT requires more development before it can be implemented. Therefore, a phased plan is presented that first uses the foil and ion-optical systems with detectors that can be implemented immediately—namely CR-39 and hDISC streak cameras. These detectors will allow the MRSt to be commissioned in an intermediate stage and begin collecting data on a reduced timescale, while the PDDT is developed in parallel. A CR-39 detector will be used in phase 1 for the measurement of the time-integrated neutron spectra with excellent energy-resolution, necessary for the energy calibration of the system. Streak cameras will be used in phase 2 for measurement of the time-resolved spectrum with limited spectral coverage, which is sufficient to diagnose the time-resolved ion temperature. Simulations are presented that predict the performance of the streak camera detector, indicating that it will achieve excellent burn history measurements at current yields, and good time-resolved ion-temperature measurements at yields above 3 × 10 17 . The PDDT will be used for optimal efficiency and resolution in phase 3.

47 OTHER INSTRUMENTATION↗
DOE OSTI2025

Reaction-in-flight neutrons as a diagnostic for hydrodynamical mixing in double shell inertial confinement fusion capsules

We examine reaction-in-flight (RIF) neutrons as diagnostics for hydrodynamical mixing of high-Z shell material into the hotspot of double shell capsules that are designed for the National Ignition Facility. In particular, we consider the effects of different levels of mixing of tungsten shell material into the DT gas on RIF spectra. Using a set of 1D simulations from the radiation hydrodynamic code xRAGE to determine the temperature and density profiles of the mixed W-DT gas, we find that increasing the mass of mix systematically reduces the ratio of RIF neutrons to primary 14 MeV neutrons. The shape of the RIF spectrum also changes with mix, with the predicted spectrum softening in energy with increasing mix.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Experimental characterization of hot-electron emission and shock dynamics in the context of the shock ignition approach to inertial confinement fusion

We report on planar target experiments conducted on the OMEGA-EP laser facility performed in the context of the shock ignition (SI) approach to inertial confinement fusion. The experiment aimed at characterizing the propagation of strong shock in matter and the generation of hot electrons (HEs), with laser parameters relevant to SI (1-ns UV laser beams with I ~10 16 W/cm 2 ). Time-resolved radiographs of the propagating shock front were performed in order to study the hydrodynamic evolution. The hot-electron source was characterized in terms of Maxwellian temperature, T h , and laser to hot-electron energy conversion efficiency η using data from different x-ray spectrometers. The post-processing of these data gives a range of the possible values for T h and η [i.e., T h [keV] ϵ (20, 50) and η ϵ (2%, 13%)]. These values are used as input in hydrodynamic simulations to reproduce the results obtained in radiographs, thus constraining the range for the HE measurements. According to this procedure, we found that the laser converts ~10% ± 4% of energy into hot electrons with T h = 27 ± 8 keV. The paper shows how the coupling of different diagnostics and numerical tools is required to sufficiently constrain the problem, solving the large ambiguity coming from the post-processing of spectrometers data. The effect of the hot electrons on the shock dynamics is then discussed, showing an increase in the pressure around the shock front. Furthermore, the low temperature found in this experiment without pre-compression laser pulses could be advantageous for the SI scheme, but the high conversion efficiency may lead to an increase in the shell adiabat, with detrimental effects on the implosion.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Pushered single shell implosions for mix and radiation trapping studies using high-Z layers on National Ignition Facility

Pushered Single Shells (PSSs) are another approach to Inertial Confinement Fusion implosions that employ high-Z materials in the innermost capsule layer (pusher) as a means to enhance radiation trapping and lower core ignition requirements. However, adding high-Z materials can also increase losses due to mix, provide extra tamping, and make the capsule emission opaque to x-ray diagnostics. The first PSS implosions performed on the National Ignition Facility use plastic ablators with a germanium (Ge) dopant as a high-Z surrogate in the pusher to isolate the effects of high-Z mix and radiation trapping without changing tamping. Using a 2-shock laser pulse, the PSS implosions are designed and symmetrized to reach 3.7 keV core temperatures. A low concentration (2.8%) Ge dopant is added to the innermost layer, and the resulting effects on mix and x-ray opacity are observed. The method of separated reactants is used to infer information about mixing between the deuterated plastic pusher and the capsule fill gas (25% tritium) from the resulting nuclear DT reactions. Radiation transport is studied via capsule emission x-ray spectroscopy and imaging. Both nuclear and x-ray data corroborate the hypothesis that the addition of Ge strongly affects the mix region through radiation losses but has a minimal effect on the core and the warm, unmixed regions. Simulations using diffusive and turbulent mix models agree qualitatively with data, but quantitative agreement may require hybrid mix models that can model the transitional regime between turbulence and diffusion. Simulations matching the observables show increased core radiation trapping when Ge is added.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

A “polar contact” tent for reduced perturbation and improved performance of NIF ignition capsules

In indirectly driven Inertial Confinement Fusion implosions conducted on the National Ignition Facility (NIF), the imploding capsule is supported in a laser-heated radiation enclosure (called a “hohlraum”) by a pair of very thin (~15–45 nm) plastic films (referred to as a “tent”). Even though the thickness of these tents is a small fraction of that of the spherical capsule ablator (~165 μm), both numerical simulations as well as experiments indicate that this capsule support mechanism results in a large areal density (ρR) perturbation on the capsule surface at the contact point where the tent departs from the capsule. As a result, during deceleration of the deuterium-tritium (DT) fuel layer, a jet of the dense ablator material penetrates and cools the fuel hot spot, significantly degrading the neutron yield (resulting in only ~10%–20% of the unperturbed 1-D yield). In this article, we present a hypothesis and supporting design simulations of a new “polar contact” tent support system, which reduces the contact area between the tent and the capsule and results in a significant improvement in the capsule performance. Simulations predict a ~70% increase in neutron yield over that for an implosion with a traditional tent support. Overall, an initial demonstration experiment was conducted on the NIF and produced highest ever recorded primary DT neutron yield among all layered DT implosions with plastic ablators on the NIF, though more experiments are needed to comprehensively study the effect of the polar tent on implosion performance.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Simulated impact of fill tube geometry on recent high-yield implosions at the National Ignition Facility

Inertial confinement fusion capsules fielded at the National Ignition Facility are filled with deuterium and tritium fuel by means of a fill tube. The fill tube introduces a low-density pathway into the fuel region of the capsule that allows high Z contaminant to invade the hot spot during the course of the implosion. A recent series of nominally identical high-yield implosions on the NIF has exhibited significant variability in performance. We evaluate the impact of the fill tube in these implosions computationally to determine whether variations in fill tube geometry could have contributed to this variability. The main contrast between the fill tube geometry in the six shots was the outer diameter of the capsule bore hole, a conical hole into which the fill tube is inserted. In our simulations, the geometry of the bore hole can play a significant role in the development of nonlinear flows seeded by the fill tube. Additionally, we find that the amount of space between the bore hole and the fill tube is the primary factor that determines the amount of contaminant jetted into the hot spot by the fill tube and, in turn, the level of yield reduction due to the fill tube in our simulations. As a consequence, some capsules with 5 μm fill tubes are predicted to outperform capsules with 2 μm fill tubes. We also find that micrometer-scale changes to bore hole size can impact fusion yields by up to four times near the ignition threshold. Nevertheless, simulation trends do not reproduce experimental yield trends, suggesting that the fill tube geometry was not the primary factor contributing to the observed variability in performance and that the fill tube could be masking sensitivity to other asymmetries such as other micrometer-scale capsule defects like voids that were not included in our simulations.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

The effects of convergence ratio on the implosion behavior of DT layered inertial confinement fusion capsules

The wetted foam capsule design for inertial confinement fusion capsules, which includes a foam layer wetted with deuterium-tritium liquid, enables layered capsule implosions with a wide range of hot-spot convergence ratios (CR) on the National Ignition Facility. In this paper, we present a full-scale wetted foam capsule design that demonstrates high gain in one-dimensional simulations. In these simulations, increasing the convergence ratio leads to an improved capsule yield due to higher hot-spot temperatures and increased fuel areal density. High-resolution two-dimensional simulations of this design are presented with detailed and well resolved models for the capsule fill tube, support tent, surface roughness, and predicted asymmetries in the x-ray drive. Our modeling of these asymmetries is validated by comparisons with available experimental data. In 2D simulations of the full-scale wetted foam capsule design, jetting caused by the fill tube is prevented by the expansion of the tungsten-doped shell layer due to preheat. While the impacts of surface roughness and predicted asymmetries in the x-ray drive are enhanced by convergence effects, likely underpredicted in 2D at high CR, simulations predict that the capsule is robust to these features. Nevertheless, the design is highly susceptible to the effects of the capsule support tent, which negates all of the one-dimensional benefits of increasing the convergence ratio. Indeed, when the support tent is included in simulations, the yield decreases as the convergence ratio is increased for CR > 20. Finally and nevertheless, the results suggest that the full-scale wetted foam design has the potential to outperform ice layer capsules given currently achievable levels of asymmetries when fielded at low convergence ratios (CR < 20).

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Impact of imposed mode 2 laser drive asymmetry on inertial confinement fusion implosions

Low-mode asymmetries have emerged as one of the primary challenges to achieving high-performing inertial confinement fusion implosions. These asymmetries seed flows in the implosions, which will manifest as modifications to the measured ion temperature (T ion ) as inferred from the broadening of primary neutron spectra. The effects are important to understand (i) to learn to control and mitigate low-mode asymmetries and (ii) to experimentally more closely capture thermal T ion used as input in implosion performance metric calculations. Here, in this paper, results from and simulations of a set of experiments with a seeded mode 2 in the laser drive are described. The goal of this intentionally asymmetrically driven experiment was to test our capability to predict and measure the signatures of flows seeded by the low-mode asymmetry. The results from these experiments [first discussed in M. Gatu Johnson et al., Phys. Rev. E 98, 051201(R) (2018)] demonstrate the importance of interplay of flows seeded by various asymmetry seeds. In particular, measured T ion and self-emission x-ray asymmetries are expected to be well captured by interplay between flows seeded by the imposed mode 2 and the capsule stalk mount. Measurements of areal density asymmetry also indicate the importance of the stalk mount as an asymmetry seed in these implosions. The simulations brought to bear on the problem (1D LILAC, 2D xRAGE, 3D ASTER, and 3D Chimera) show how thermal T ion is expected to be significantly lower than T ion as inferred from the broadening of measured neutron spectra. They also show that the electron temperature is not expected to be the same as T ion for these implosions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Demonstration of low-mode shape control in indirect-drive double shell implosions at the NIF

Double shell inertial confinement fusion is a concept for achieving robust thermonuclear burn that uses dense metal shells to compress deuterium-tritium (DT) fuel to fusion conditions. Double shell implosions are typically indirectly driven and involve a target that consists of a low-Z ablator, a foam layer, and a high-Z pusher surrounding the DT fuel. The goal of the campaign is to achieve a volumetric burn as radiation losses from the DT fuel are trapped by the opaque high-Z shell. The overall performance of double shell implosions relies on the efficient collisional transfer of kinetic energy between layers. The efficiency of this transfer (and therefore the overall performance of a given implosion) is degraded by the presence of low-mode asymmetries. P2 asymmetries are often observed in spatially resolved 2D radiographs of nominal double shell implosions. This work discusses three such experiments: one with an oblate P2 asymmetry, one with a prolate P2 asymmetry, and one with an approximate spherical symmetry. After performing a shape analysis of the oblate and prolate implosions to quantify asymmetries, these experimental results were compared with the results of hydrodynamic simulations for the two experiments. Differences between the experiment and simulation were then used to design an approximately spherical implosion by altering the incident laser cone fraction. Radiographs from the experiment that implemented the modified cone fraction show evidence of an implosion that is approximately spherical until bang time. This design is intended to serve as a point design for future studies that will seek to optimize various aspects of the double shell target.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Demonstration of Super-X divertor exhaust control for transient heat load management in compact fusion reactors

Nuclear fusion could offer clean, abundant energy. However, managing the power exhausted from the core fusion plasma towards the reactor wall remains a major challenge. This is compounded in emerging compact reactor designs promising more cost-effective pathways towards commercial fusion energy. Alternative Divertor Configurations (ADCs) are a potential solution. In this work, we demonstrate exhaust control in ADCs, employing a novel method to diagnose the neutral gas buffer, which shields the target. Our work on the Mega Ampere Spherical Tokamak Upgrade shows that ADCs tackle key risks and uncertainties for fusion energy. Their highly reduced sensitivity to perturbations enables active exhaust control in otherwise unfeasible situations and facilitates an increased passive absorption of transients, which would otherwise damage the divertor. We observe a strong decoupling of each divertor from other reactor regions, enabling near-independent control of the divertors and core plasma. Our work showcases the real-world benefits of ADCs for effective heat load management in fusion power reactors.

Imaging techniques↗
DOE OSTI2024

Effects of drive pulse shape on graded metal pushered single shell capsule implosions on the National Ignition Facility

Graded metal pushered single shells (PSS) are a viable alternative to low-Z capsules (Z is the atomic number) for indirect drive inertial confinement fusion implosions due to enhanced core tamping and radiation trapping, but they can be compromised by the pusher mixing with the fuel. We compare 2-shock and 3-shock laser pulses for Be/Cr PSS capsules filled with deuterium–tritium gas fuel at 6 mg/cc density. 1D radiation-hydrodynamic simulations predict higher core compression and, hence, ∼2× higher fusion yield for the 3-shock drive than for 2-shock. Nevertheless, we observe similar core ion temperatures and fusion yields for both drives. The implosion burn duration is 25% shorter and the core volume is ∼2.5× smaller for the 3-shock drive than for 2-shock, consistent with a higher compression. 1D LASNEX mix simulations using a buoyancy-drag model matching the measured yields also agree with the observed core sizes and burn durations and suggest ∼40% and ∼70% yield degradations for 2-shock and 3-shock drives due to hydrodynamic instabilities and atomic mix at the pusher–fuel interface. At the same time, 2D HYDRA simulations show that mid-mode (2–250) instability degradations are negligible for the 2-shock implosion (9%) and significant (45%) for 3-shock. Subtracting these from the 1D mix simulations, we infer similar degradations from high-mode instabilities and atomic mix for both drives. Due to its robustness to mid-mode instabilities, future pusher–gas mix studies will use the 2-shock drive.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Hot-spot model for inertial confinement fusion implosions with an applied magnetic field

Imposing a magnetic field on inertial confinement fusion implosions magnetizes the electrons in the compressed fuel; this suppresses thermal losses, which increases temperature and fusion yield. Indirect-drive experiments at the National Ignition Facility with 12 and 26 T applied magnetic fields demonstrate up to 40% increase in temperature, 3× increase in fusion yield, and indicate that magnetization alters the radial temperature profile [Moody et al., Phys. Rev. Lett. 129, 195002 (2022); Lahmann et al., APS DPP (2022)]. In this work, we develop a semi-analytic hot-spot model, which accounts for the two-dimensional (2D) Braginskii anisotropic heat flow due to an applied axial magnetic field. First, we show that hot-spot magnetization alters the radial temperature profile, increasing the central peakedness, which is most pronounced for moderately magnetized implosions (with 8–14 T applied field), compared to both unmagnetized (with no applied field) and highly magnetized (with 26 T or higher applied field) implosions. This model explains the trend in the experimental data, which finds a similarly altered temperature profile in the 12 T experiment. Next, we derive the hot-spot model for gas-filled (Symcap) implosions, accounting for the effects of magnetization on the thermal conduction and in changing the radial temperature (and density) profiles. Using this model, we compute predicted central temperature amplification and yield enhancement scaling with the applied magnetic field. The central temperature fits the experimental data accurately, and the discrepancy in the yield suggests a systematic (independent of applied field) degradation, such as mix, and additional degradation in the reference unmagnetized shot, such as reduced laser drive, increased implosion asymmetry, or the magnetic field suppressing ablator mixing into the hot-spot.

Alpha particles↗
DOE OSTI2010

Neutral particle measurements of fusion tritons in JET

A neutral particle analyzer [A.D. Izvozchikov et al., JET Report No. JET-R(91)-12, 1991] operating in the MeV energy range was used to measure the flux of neutralized d-d fusion tritons emitted from the hot-ion H-mode deuterium plasma heated by deuterium neutral beams. It was found that tritons in the energy range of 0.3-1.1 MeV were largely neutralized by the beam atoms and the beam halo atoms. This enabled us to find the localized energy distribution function of the fusion tritons in the central plasma region. Simulation of the triton energy distribution function shows that MeV ions in the JET hot-ion H-mode plasma behave classically.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2010

Analysis of Recurrent Patterns in Toroidal Magnetic Fields

In the development of magnetic confinement fusion which will potentially be a future source for low cost power, physicists must be able to analyze the magnetic field that confines the burning plasma. While the magnetic field can be described as a vector field, traditional techniques for analyzing the field's topology cannot be used because of its Hamiltonian nature. In this paper we describe a technique developed as a collaboration between physicists and computer scientists that determines the topology of a toroidal magnetic field using fieldlines with near minimal lengths. More specifically, we analyze the Poincare map of the sampled fieldlines in a Poincare section including identifying critical points and other topological features of interest to physicists. The technique has been deployed into an interactive parallel visualization tool which physicists are using to gain new insight into simulations of magnetically confined burning plasmas.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2018

Plasma Facing Components and Advanced System Design:A technical review of FY14-17 work

Fusion energy holds the promise of being a major source of world energy in this century and beyond. One of the two major approaches to controlled fusion is confinement of hot plasma gas by strong magnetic fields. Many countries have experimental devices, and ITER is a large international collaboration device being built in France with U.S. participation. Because fusion gain of the deuterium/tritium (D/T) plasma fuel increases strongly with temperature, increasing the core-plasma energy confinement time, tE, has been a key requirement. With excellent progress on increasing tE, considerable attention has turned to other requirements for an economical power plant. Because core plasma tE will always be finite, plasma is exhausted from the core and strikes surrounding material surfaces with peak heat fluxes that challenge the structural integrity of present-day materials. Furthermore, sputtering of such material can introduce impurities into the core that radiate energy or dilute the D/T fuel, thereby reducing fusion gain. Simulationbased assessments of such plasma/wall interactions is the goal of the two projects described here.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Role of poloidal E × B drift in divertor heat transport in DIII‐D

Abstract Simulations for DIII‐D high confinement mode plasmas with the multifluid code UEDGE show a strong role of poloidal E × B drifts on divertor heat transport, challenging the paradigm of conduction‐limited scrape‐off layer (SOL) transport. While simulations with reduced drift magnitude are well aligned with the assumption that electron heat conduction dominates the SOL heat transport, simulations with drifts predict that the poloidal convective E × B heat transport dominates over electron heat conduction in both attached and detached conditions. As poloidal E × B flow propagates across magnetic field lines, poloidal transport with shallow magnetic pitch angles can reach values that are of the same order as would be provided by sonic flows parallel to the field lines. These flows can lead to strong convection‐dominated divertor heat transport, increasing the poloidal volume of radiative power front, consistent with previous measurements at DIII‐D. Due to these convective flows, the Lengyel integral approach, assuming zero convective fraction, is expected to provide a pessimistic estimate for the radiative capability of impurities in the divertor. For the DIII‐D simulations shown here, the Lengyel integral approach underestimates the radiated power by a factor of 6, indicating that, for reliable DIII‐D divertor power exhaust predictions, full two‐dimensional (2D) calculations, including drifts, would be necessary.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗