Deuterium-tritium Ice Research Articles

Target design for shock ignition

The conventional approach of laser driven inertial fusion involves the implosion of cryogenic shells of deuterium-tritium ice. At sufficiently high implosion velocities, the fuel ignites by itself from a central hot spot. In order to reduce the risks of hydrodynamic instabilities inherent to large implosion velocities, it was proposed to compress the fuel at low velocity, and ignite the compressed fuel by means of a convergent shock wave driven by an intense spike at the end of the laser pulse. This scheme, known as shock ignition, reduces the risks of shell break-up during the acceleration phase, but it may be impeded by a low coupling efficiency of the laser pulse with plasma at high intensities. This work provides a relationship between the implosion velocity and the laser intensity required to ignite the target by a shock. The operating domain of shock ignition at different energies is described.

Targets for the National Ignition Campaign

The National Ignition Facility (NIF) is a 192 beam Nd-glass laser facility presently under construction at Lawrence Livermore National Laboratory (LLNL) for performing inertial confinement fusion (ICF) and experiments studying high energy density (HED) science. When completed in 2009, NIF will be able to produce 1.8 MJ, 500 TW of ultraviolet light for target experiments that will create conditions of extreme temperatures (>108 K), pressures (10 GBar) and matter densities (>100 g/cm3). A detailed program called the National Ignition Campaign (NIC) has been developed to enable ignition experiments in 2010, with the goal of producing fusion ignition and burn of a deuterium-tritium (DT) fuel mixture in millimeter-scale target capsules. The first of the target experiments leading up to these ignition shots will begin in 2008. The targets for the NIC are both complex and precise, and are extraordinarily demanding in materials fabrication, machining, assembly, cryogenics and characterization.The DT fuel is contained in a 2-millimeter-diameter graded copper/beryllium or CH shell. The 75-μm-thick cryogenic ice DT fuel layer is formed to sub-micron uniformity at a temperature of approximately 18 Kelvin. The capsule and its fuel layer sit at the center of a gold/depleted uranium ‘cocktail’ hohlraum. Researchers at LLNL have teamed with colleagues at General Atomics to lead the development of the technologies, engineering design and manufacturing infrastructure necessary to produce these demanding targets. We are also collaborating with colleagues at the Laboratory for Laser Energetics (LLE) at the University of Rochester in DT layering, and at Fraunhofer in Germany in nano-crystalline diamond as an alternate ablator to Beryllium and CH.The Beryllium capsules and cocktail hohlraums are made by physical vapor deposition onto sacrificial mandrels. These coatings must have high density (low porosity), uniform microstructure, low oxygen content and low permeability. The ablator capsule has a 5-μm-diameter hole laser drilled to permit removal of the mandrel and introduction of the DT fuel. A 10-μm-diameter fill tube is bonded to the capsule to enable filling with the DT gas. These components must then be assembled to tolerances of approximately 5—10 microns, with comprehensive characterization and metrology. The DT ice is formed through controlled seeding, aided by beta decay of the tritium to help smooth the layer, and differential heating of the hohlraum to counteract the effects of natural convection.We present an overview of the technologies for target fabrication, assembly and metrology and advances in growth and imaging of DT ice layers. The sum of these efforts represents a quantum leap in target precision, characterization, manufacturing rate and flexibility over current state-of-the-art.This work performed under the auspices of the U.S. Department of Energy by the University of California, Lawrence Livermore National Laboratory under Contract No. W-7405-ENG-48.

Initial experiments on the shock-ignition inertial confinement fusion concept

Shock ignition is a two-step inertial confinement fusion concept where a strong shock wave is launched at the end of the laser pulse to ignite the compressed core of a low-velocity implosion. Initial shock-ignition technique experiments were performed at the OMEGA Laser Facility [T. R. Boehly et al., Opt. Commun. 133, 495 (1997)] using 40-μm-thick, 0.9-mm-diam, warm surrogate plastic shells filled with deuterium gas. The experiments showed a significant improvement in the performance of low-adiabat, low-velocity implosions compared to conventional “hot-spot” implosions. High areal densities with average values exceeding ∼0.2g∕cm2 and peak areal densities above 0.3g∕cm2 were measured, which is in good agreement with one-dimensional hydrodynamical simulation predictions. Shock-ignition technique implosions with cryogenic deuterium and deuterium-tritium ice shells produced areal densities close to the 1D prediction and achieved up to 12% of the predicted 1D fusion yield.

Sensitivity of Laser Mégajoule Ignition Targets to Technological Defects

The first ignition experiments on the Laser Mégajoule facility will use an indirect drive scheme. Our A1040 point design target is a graded doped plastic capsule filled by permeation within a gold cylinder. The deuterium-tritium ice layer may be formed either by classical slow cooling at 1.5 K below triple point, or by rapid cooling at 2.3 K below triple point. To complete the specifications, we first studied the robustness to all technological defects with the current CEA capabilities for these two options of ice formation. The technological imperfections taken into account are regrouped into 1D errors, which keep the implosion spherical, and 3D errors, which induce a deformation of the shell. The 3D robustness is expressed in terms of deformation at peak velocity and compared to the deformation threshold obtained with 2D simulations. The 1D robustness is given by the probability of exceeding 50% of nominal yield. We have taken into account 22 1D parameters and the fusion energy is approximated by a neural network based on 2000 simulations. Although the studies are not finished yet, the first results show that the A1040 design with rapid cooling has sufficient margins with respect to technological defects.

Synthesis and Characterization of Fully Deuterated Upilex Type Polyimides

Polyimide films are materials of choice as membrane shutters of the hohlraum that contains a plastic microshell in which nuclear products are located to carry out Inertial Confinement Fusion experiments. In addition to high mechanical properties, polymer membranes must exhibit a high transparency in the infrared (IR) wavelengths range between 2.86 and 4 μm (3500 and 2500 cm-1) to allow the IR assisted deuterium-tritium ice layer redistribution. Upilex™ type polyimides exhibit the desired mechanical properties but are not transparent in this wavelengths range due to the aromatic C-H stretching bands. In order to provide the required optical properties, the hydrogen atoms must be substituted by deuterium atoms.In the present contribution, we wish to report the first synthesis of a fully deuterated Upilex™ type polyimide. Optimized ways of synthesis and purification for the two deuterated monomers 3,3′,4,4′-biphenyltetracarboxylic dianhydride (BPDA-d6) and p-phenylene diamine (PDA-d4) have been developed. These monomers have been used to prepare deuterated poly(amic-acid) solutions in NMP. Thermal treatment of films obtained from these solutions gives rise to deuterated Upilex type polyimide films. These films show a high transparency in the 2.86-4 μm region. The synthesis and the characterization of this new deuterated polymer will be discussed.

Design of the OMEGA Laser Target Chamber Tritium Removal System

Preparations are currently underway at the OMEGA laser at the University of Rochester Laboratory for Laser Energetics (UR/LLE) to conduct direct drive laser implosion campaigns with inertial confinement fusion targets containing deuterium-tritium (DT) cryogenic ice layers. The OMEGA Cryogenic Target Handling System will fill plastic targets with high-pressure DT (150 MPa) at 300 to 500 K, cool them down to cryogenic temperature (<25 K), form the DT ice layer, and transport the targets to the OMEGA laser target chamber. Targets will then be shot with the 60-beam 30-kJ OMEGA laser. A tritium removal system has been designed to remove tritium from effluents associated with operation of the target chamber and its associated diagnostic antechambers, vacuum pumping systems, and target insertion systems. The design of the target chamber tritium removal system (TCTRS) is based on catalytic oxidation of DT and tritiated methane to tritiated water (DTO), followed by immobilization of DTO on molecular sieves. The design of the TCTRS presented a challenge due to the low tritium release limits dictated by the tritium license at UR/LLE. Aspen Plus, a commercial software package intended for the simulation and design of chemical processing systems operating at steady state, was used to simulate and design the TCTRS. A second commercial software package, Aspen ADSIM, was used to simulate and design the TCTRS molecular sieve beds, which operate at unsteady state. In this paper, we describe the design of the TCTRS and the benefits that were realized by use of the Aspen Plus and Aspen ADSIM software packages.

Tritium Inventory of Inertial Fusion Energy Target Fabrication Facilities: Effect of Foam Density and Consideration of Target Yield of Direct Drive Targets

The tritium inventory of direct drive inertial fusion energy (IFE) target filling facilities is examined in the interest of minimizing the tritium inventory. A model is described that has been developed to evaluate the tritium inventory of the target filling process as a function of filling and layering parameters, as well as target design parameters. Previous studies by A. Nobile et al. showed that the temperature and the fill system void fraction have a significant effect on the tritium inventory. The current study uses the model to examine the effect of deuterium-tritium (DT) ice layering time and density of the CH foam in the target on the tritium inventory. The study shows that increasing the foam density and decreasing the DT ice layering time significantly reduce the tritium inventory. Fortunately, one-dimensional target design calculations indicate that the foam density in the direct drive target can be increased to ~200 mg/cm3 without significant degradation of the target yield. Having evaluated and minimized the theoretical tritium inventory, calculations were performed with more realistic batch filling scenarios. The inventories associated with “real” filling scenarios approach the theoretical minimum inventory as the number of batches is increased, resulting in tritium inventories that seem acceptable for future IFE target DT filling facilities.

Numerical Raytrace Verification of Optical Diagnostics of Ice Surface Roughness for Inertial Confinement Fusion Experiments

Targets for future laser-fusion ignition experiments will consist of a frozen deuterium-tritium ice layer adhering to the inner surface of a spherical shell, and the specifications for the inner surface quality of this ice layer are extremely demanding. We have developed a numerical raytrace model in order to validate backlit optical shadowgraphy as an ice-surface diagnostic, and we have used the code to simulate shadowgraph data obtained from mathematical ice layers having known modal imperfections. We find that backlit optical shadowgraphy is a valid diagnostic of the mode spectrum of ice-surface imperfections for mode numbers as high as 80 provided the experimental data are analyzed appropriately. We also describe alternative measurement techniques, which may be more sensitive than conventional backlit shadowgraphy.

Imprint reduction using an intensity spike in OMEGA cryogenic targets

A novel technique for reducing laser imprint in OMEGA [T. R. Boehly et al., Opt. Commun. 133, 495 (1997)] cryogenic targets has been developed. A typical direct-drive inertial confinement fusion target consists of a shell of deuterium–tritium ice with a thin outer layer of polystyrene (CH). The presence of the CH layer gives rise to a brief period of early-time Rayleigh–Taylor instability. This effectively increases the amount of laser imprint by about a factor of 2 for typical target configurations, and especially for the nonuniformity modes considered most dangerous for target performance. Two-dimensional hydrodynamic simulations show that by introducing a short, high-intensity spike at the start of the implosion, this early-time growth can be significantly reduced with only a modest decrease in the calculated one-dimensional neutron yield.

Smoothing of deuterium-tritium ice by electrical heating of the saturated vapor

High gain targets for inertial confinement fusion (ICF) are spherical shells which contain a layer of deuterium-tritium (DT) ice which surrounds a volume of DT gas in thermal equilibrium with the solid. The roughness of the inner surface of the cryogenic fuel layer inside of these targets is one of the sources of imperfections which cause implosions to deviate from perfect one-dimensional performance. Reductions in the surface roughness of this fuel layer improve confidence in the ability of the ICF program to achieve ignition in the National Ignition Facility (NIF), and increase the relevance of cryogenic experiments on the omega laser. We have developed a technique to generate a heat flux across this surface by applying an electric field to the DT vapor in the center of these shells. This vapor has a small but significant conductivity due to ionization caused by {beta} decay of tritium in the vapor and the solid. We describe here experiments using a 1.15 GHz cavity to apply an electric field to frozen DT inside of a sapphire test cell. The cell and cavity geometry allows visual observation of the frozen layers. We find that the resulting heat flux reduces the roughness of the icemore » surface. {copyright} {ital 1997} {ital The American Physical Society}« less

Beta heating driven deuterium–tritium ice redistribution, modeling studies

We have previously demonstrated, based upon a one-dimensional analysis, that beta heating could cause a redistribution of nonuniform deuterium–tritium (DT) ice within a cryogenic target at an ideal rate that would decrease nonuniformity by a factor of 10 every hour. Our analysis was extended to address two pertinent questions: for a spherical target, where will the DT ice finally go and how quickly will it get there? This analysis is a two-dimensional heat and mass transfer problem. We have used PLTMG, a state-of-the-art multigrid computer code for the analysis. For the ‘‘tent’’ mount method of positioning a target in a cryogenic enclosure, an ice layer nonuniformity of &amp;lt;1% can be achieved. We have evaluated the rate of ice redistribution and report its variation as a function of 3He content within the target, the target temperature, target geometry, and thermal conductivity of the shell wall.

Beta energy driven uniform deuterium–tritium ice layer in reactor-size cryogenic inertial fusion targets

We have studied the effectiveness of beta heating on driving a deuterium–tritium (DT) ice layer toward uniformity within spherical inertial fusion targets. Beta heating results from the deposition of energy from the beta particles from the tritium decay in these targets. These targets are enclosed in a constant temperature environment. This study first considered a one-dimensional heat transfer model of an idealized target, geometry, and environment. We concluded that beta energy deposition can drive a nonuniform DT ice layer towards uniformity. The maximum rate at which this redistribution can occur is given by dδ/dt=δ̇=−δ/θTC, where δ=DT ice thickness in excess of that for a uniform layer and θTC=a time constant ≊1500 s. This translates to a factor of 10 reduction in nonuniformity every hour. We also evaluated, and continue to do so, the consequences of deviations from the idealized case assumptions and the selection of certain experimental parameters, on the actual behavior of DT ice in such cryogenic targets. The status of this work is such that we have confidence in the use of beta heating to drive the DT ice toward uniformity for reactor-size targets (∼1 mm and larger in diameter).

Experimental verification of beta-decay-driven sublimation in deuterium-tritium ice held in spherical fusion targets

A nonuniform layer of deuterium-tritium (DT) ice inside a spherical inertial confinement fusion (ICF) target held in an isothermal cryogenic environment should be driven toward uniformity by the beta-decay heat of the tritium. Experiments have been performed at KMS fusion to verify this hypothesis. Two major conclusions may be drawn from the initial results: (1) the beta decay of the tritium does deposit energy in the target, as evidenced by melting of DT ice when the target is well insulated from its surroundings, and (2) solid layers of DT ice sublime because of beta-decay heat. Both conclusions are reinforced by companion studies with nonradioactive hydrogen-deuterium (HD) ice in similar targets held under similar experimental conditions.