Session BO4: ICF: Compression and Burn I

Evolution of the Hot Spot Density and Temperature on the National Ignition Facility

The electron density and temperature and their evolution in the hot spot of a Kr-doped, big-foot implosion target were measured for the first time using an absolutely calibrated, streaked, high-resolution x-ray spectrometer named dHIRES [1] on the National Ignition Facility (NIF). The electron density was inferred through Stark-broadened line shapes and the temperature was derived from the relative intensities of dielectronic satellites to the resonance line. The data show that the hot spot density peaks after the x-ray bang time while its temperature peaks before the x-ray bang time. Such trend is compared with a 1-D calculation of the Symcap implosion using a self-similar temperature profile [2], as well as collisional-radiative calculations for line intensities and shapes [3]. [1] L. Gao \textit{et al.}, Rev. Sci. Inst. \textbf{89}, 10F125 (2018). [2] R. Betti \textit{et al.}, Phys. Plasmas \textbf{8}, 5257 (2001). [3] H. A. Scott, JQSRT \textbf{71}, 689--701 (2001).   

Diagnosing 3D asymmetries in indirect drive implosions at the National Ignition Facility

To achieve hotspot ignition, inertial confinement fusion (ICF) implosions must achieve high hotspot pressures that are inertially confined by a dense shell of DT fuel. This requires high inflight shell velocity, good energy coupling between the hotspot and imploding shell, and high areal-density at peak compression. Three-dimensional (3D) asymmetries seeded by the drive and/or target can grow during an implosion and damage both the coupling of energy to the hotspot and confinement of that energy. Low mode (l $\le $2) 3D asymmetries have been present to some extent in nearly every campaign and target design type fielded at the National Ignition Facility (NIF). To demonstrate that these asymmetries are real and significant we have cross-compared independent experimental measurements. To better understand their cause and impact, we are combining the available experimental data, simplified models, and 3D radiation hydrodynamic simulations, while also proposing new efforts. This presentation will briefly overview some of these efforts and present the status of our understanding of the origin and impact of 3D asymmetry in ignition experiments at the NIF as well as plans for how they may be mitigated.   

\textbf{Using X-Ray Images and Secondary DT Neutrons to Diagnose Convergence and hot-spot asymmetries in implosions at the NIF }

An important figure of merit for the performance of an ICF (Inertial Confinement Fusion) implosion is the implosion convergence; the ratio of the final and initial capsule radii. At the NIF, this is routinely inferred from images of self-emitted x-rays during peak compression. Convergence is also inferred from the yield ratio of secondary DT neutrons to primary DD neutrons. While these independent analysis methods track one another, there are clear differences dictated by the underlying physics. Understanding the nature of these differences could potentially offer interesting insights on the sensitivities of both methods to effects of high-Z mix and hot-spot asymmetries. Spectral measurements of the secondary DT neutron spectra, along different lines of sight, provide additional information about the impact of these physics phenomena. Here, we present our experiments at the NIF, supported by a large set of hydro simulations. This work was supported in part by the DOE and LLNL.   

Hot-Spot Flow Dynamics and Residual Kinetic Energy in NIF ICF Implosions

Inertial confinement fusion relies on converting kinetic energy imparted during an implosion into thermal energy of the deuterium-tritium fuel. Non-radial motion in the fusing volume (``hot-spot'') indicates residual kinetic energy and hence inefficient conversion. We present velocities internal to the hot-spot measured by 2D particle velocimetry of x-ray images. These are quantitatively compared to 2D hydrodynamic simulations for several magnitudes of low-mode asymmetric drive. Effects on burn-averaged ion temperatures are shown in observations, simulations and theory. Measurements of burn-averaged electron temperatures appear independent of residual motion and remain several hundred keV below minimum T$_{\mathrm{ion}}$ for all cases of asymmetric drive. This discrepancy will be discussed in terms of fuel velocity-variance artificially increasing apparent ion temperatures. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.   

\textbf{First experiments of hohlraum asymmetries mitigation in implosions using capsule shims on the National Ignition Facility}

Low mode asymmetries are suspected to be among the main causes for ignition performance degradation in indirect drive capsule implosions on the National Ignition Facility. Radiation hydrodynamic simulations show that in implosion designs using plastic ablators, the hohlraum driven inflight capsule P4 asymmetry can be tuned significantly with a modest shim of the outer capsule surface (\textasciitilde 7 \textmu m inflight P4 per - 1 \textmu m P4 shim amplitude). Recent 2D inflight shell radiography experiments on the NIF employed the first P4 shimmed capsules to validate the inflight shape tuning vs shim amplitude. This work will summarize the data and uses simulations to extrapolate the results to future DT fuel implosions for which a significant improvement in fusion neutron yield via capsule shim is expected.   

HybridE design, status, and look ahead

Inertial Confinement Fusion (ICF) implosion performance has been shown to be sensitive to implosion velocity, adiabat, inflight ablation pressure, and capsule size [1]. Recently the hybridB campaign has fielded the largest high-density carbon (HDC or diamond) ablator implosions at NIF while maintaining symmetry using an experimental database to inform hohlraum design choices [2]. HybridE further increases the implosion velocity beyond hybridE and capsule scale (by $+$5{\%} compared to hybridB) in a smaller hohlraum that reaches higher radiation temperatures. Time-dependent drive symmetry is controlled using cross beam energy transfer (CBET) between laser beams in low gas filled hohlraums (0.3 mg/cc He) with high laser coupling (low backscatter). This talk outlies the hybridE design, preliminary comparisons to experimental data (including symmetry experiments and layered DT implosions), and look ahead to future hybridE experiments. This work was performed under the auspices of the U.S. Department of Energy by LLNS, LLC, under Contract No. DE-AC52- 07NA27344. LLNL-ABS-780157 [1] O A Hurricane \textit{et al}, Plasma Phys. Control. Fusion \textbf{61, }014033 (2019). [2] D. A. Callahan \textit{et al,} Physics of Plasmas \textbf{25}, 056305 (2018).   

Driving larger NIF implosions with smaller CCR designs

The expected fusion performance of an ICF implosion is strongly dependent on the capsule scale, roughly as the 4$^{\mathrm{th}}$ power. The program at NIF is pursuing several avenues towards driving larger capsules within the constraints of the existing laser system. We present new results for a design with a case-to-capsule ratio (CCR) of \textasciitilde 2.7, significantly smaller than other modern low-gas-fill hohlraum designs which have operated at CCR \textgreater 3. Small CCR increases the coupling efficiency to the capsule, at a cost of more challenging Legendre mode 2 symmetry, which we compensate using wavelength tuning to empirically adjust the cross-beam energy transfer between the inner and outer beams. Results from shock timing, in-flight and stagnation symmetry of gas-filled implosions, and DT layered experiments will be presented.   

Assessing options for improved implosion performance on the National Ignition Facility.

Indirect drive implosion experiments at the National Ignition Facility (NIF) continue to shed light on the degradation mechanisms that are limiting implosion yields. As understanding of these degradations and their relative importance becomes clearer, new designs can be proposed that address the current performance limiters. Based on our current understanding of recent NIF experiments, this talk surveys the design space in the neighborhood of NIF's best performing implosions in search of options for higher yield with minimal compromises to implosion stability or increased demands on the hohlraum. A few near-term options are identified that, when combined with modest upgrades to NIF's delivered power and energy, show promise for reaching yields in the several 100 kJ range.   

An alternative approach to the self-heating regime for inertial confinement fusion

To achieve self-heating and gain in an inertial confinement fusion (ICF) hot spot, the power generated from D-T fusion must exceed the losses due to radiation, conduction and PdV work as the compressed fuel disassembles. In typical ICF designs, this requires efficient conversion of the implosion kinetic energy into ion thermal temperature to ``spark'' the fusion reactions and generate fusion power before the loss terms take over. Thus, a premium is placed on the peak fuel velocity for these designs to the detriment of stability. A recent paper[2] re-formulates the time-dependent power balance of a self-heating implosion into an expression for an enhancement in pressure amplification above that which is achieved by adiabatic compression alone. This re-statement highlights the crucial roles that tamping and Bremsstrahlung losses play in reaching the threshold for self-heating, implying that an approach that focuses on these two, i.e. increasing confinement and reducing radiative losses, may point to a more readily achievable route to ignition. In this context we discuss a capsule design using a graded Be to high-Z shell to enhance tamping and trap Bremsstrahlung losses while minimizing the hydrodynamic instability from the increase in density with radius[3]. As a result this design achieves robust self-heating at much lower velocity and greater remaining mass than typical low-Z ablator designs. [2] O.A. Hurricane et al., Phys. Plasmas \textbf{26}, 052704 (2019). [3] D. Ho et al., APS DPP PO6.11 (2018)   

High-yield implosions via high rho-R+radiation trapping using Mo doped Be ablators (PSS)

Beryllium ablators with inner layer doped with increasing Mo concentration towards the center can increase rho-R with the benefit of radiation trapping. Configurations of this type with acceptable RT growth were reported.[1] Here we present further improvement against RT growth by placing a thin layer of Be between the DT fuel and the ablator inner surface. 2D simulations show this method is effective in reducing the interface mix during post-ignition expansion. Because of the heavier shell mass, the fall-line behavior is good at both the hotspot surface and the fuel-ablator interface. The RT growth factor at the hotspot surface is only 1/3 that of the same-size HDC capsule. There is essentially no mix at the fuel-ablator interface at ignition and in addition to high rho-R, make the PSS an attractive platform for achieving ignition. The good confinement behavior of the PSS allows the use of wetted-foam for high gain. The yield of DT-gas only capsules can be considerably higher than the same-size DT-gas capsules using conventional low-Z ablators. [1]. D. Ho et al., APS-DPP (2018).   

Alternative approach to ICF ignition and burn propagation

The path to ICF ignition and propagating burn via high convergence ratio (CR) DT ice layer capsules has been challenging. In the present study, we use xRage and Hydra simulations to explore an alternative path to ignition and burn propagation. In this alternative approach, we propose that hot spot formation and burn propagation can be explored in sequential experiments. We begin with DT gas-filled capsules with modest CR to demonstrate a relatively large ignition-level hot spot. Although energy invested in the hot spot increases with size, important tradeoffs are that the stagnation pressure required for self-heating is reduced as hot spot size increases and that hot spot formation has improved robustness to instabilities and asymmetries as CR decreases$^{\mathrm{1}}$. In follow-up experiments, we would add a liquid DT layer$^{\mathrm{2}}$ to explore burn propagation from the previously demonstrated ignition-level hot spot. Designs of both the DT gas capsules and the corresponding DT liquid layer capsules will be discussed. 1. B. M. Haines et al., ``The effects of convergence ratio on the implosion behavior of DT layered ICF capsules,'' Phys. Plasmas 24, 072709 (2017). 2. R. E. Olson et al., ``First Liquid Layer Inertial Confinement Fusion Implosions at the National Ignition Facility,'' Phys. Rev. Lett. 117, 245001 (2016).   

Scaling of Two-Shock NIF Capsule Implosions

There has been recent interest in possible next generation Inertial Confinement Fusion facilities that would have enough driver energy to produce ignition and robust burn ($>100$ MJ). As part of this work, we are taking round, low convergence capsule implosions from the two-shock campaign (Phys. Plasmas, {\bf 23}, 042708, 2016) and increasing their scale size to determine at what size and laser pulse will such a capsule achieve ignition (symcap) and propagating burn (liquid layer capsule, Phys. Plasmas, {\bf 26}, 012707, 2019). We show the results of 1-D xRAGE simulations of capsules scaled up in size and in length of the laser drive pulse up to four times the original size. Once the capsule is scaled up more than 50\% of the original size, the ablator thickness and laser pulse require retuning to achieve optimal performance. We describe the scaling process and present our performance results, including the required capsule absorbed energy and implications for possible driver size.   

Self-generated Magnetic Fields in ICF Implosion due to the Fill-tube

Magnetic fields self-generated during ICF implosions are not typically considered in the design or post-shot analysis of experiments and have the potential to explain discrepancies between simulations and experimental results. During target stagnation, the fields are anticipated to grow to greater than 10,000T [1], magnetizing the plasma at the hot-spot edge. The magnetic field transport and subsequent plasma magnetization is highly dependent on the perturbation type. This talk extends previous work by simulating a fill-tube perturbation, with subsequent modifications to the performance degradation and diagnostic signatures due to magnetic fields studied. Front-to-back simulations using a surrogate fill-tube show magnetic fields self-generated around the perturbation, predominantly in the layer with large temperature/density gradients, which is hard to magnetize. However, the large ablation of plasma from the spike into the hot-spot overcomes the de-magnetizing Nernst process and results in magnetic field loops injected into the hot-spot core. As the hot-spot core is easier to magnetize, the field loop sustains a large region of cooled plasma ahead of the spike. The Righi-Leduc effect deflects heat-flow away from the spike, allowing it to propagate deeper into the core. References [1] C.A.Walsh, et al., Physical Review Letters 118, 155001 (2017)   

Developing stagnating-corona fusion targets as neutron sources

We describe the development of an `inverted-corona' fusion platform for neutron generation. Spherical, low-Z targets with either an inner CD layer, or filled with fusionable gas (D2 or DT) are irradiated on the inside surface via laser beams entering through laser-entrance holes. The resulting ablative flow reaches high velocities, 1000 km/s, before interacting at the target's center. This generates fusion reactions through stagnation and thermalization of the fast ions. This platform has been demonstrated at the kJ-level [1,2], and is expected to scale to intermediate neutron yields in excess of 1e14 at moderate laser energies (hundred-kJ level), while offering advantages over conventional, laser-driven neutron sources. For example, substantial neutron fluences at the target wall make it an interesting platform for basic science applications, while the potential for single-sided drive of the neutron source make it ideal for neutron radiography. We will present results from proof-of-principle experiments on OMEGA and design calculations for NIF-scale targets. Prepared by LLNL under Contract DE-AC52-07NA27344. [1] Ren et al., Phys. Rev. Lett. 118, 165001 (2017) [2] Abe et al., Appl. Phys. Lett. 111, 233506 (2017)   

On energy required for achieving ignition on NIF

Inertial confinement fusion experiments on the National Ignition Facility (NIF) have significantly improved hohlraum behavior, implosion symmetry and capsule yield in the last several years, but the gap between the record capsule yield (1.8x1016 neutrons) performance and ignition is still prominent. Recently, a series of theoretical studies and numerical simulations suggest that the highest performing NIF ignition capsules are close to ignition; in particular, the capsules would be able to achieve ignition with a modest upgrade of laser driver energy. In this presentation, we will present a physics analysis of the NIF ignition capsules and the minimum energy required for achieving ignition on NIF using both forward and inverse models. We find that the required minimum driver energy strongly depends on the adiabat of the main fuel and remaining ablator mass, which determines the energy partition between the main fuel and hot spot. Our analysis is consistent with experimental data, but differs from the recent published theoretical and numerical analysis. We suggest the hot spot areal density may have been overestimated, especially for the record yield capsule, and that the ignition threshold has been underestimated (LA-UR-19-25576).