James K. Hoffer

The development and advantages of beryllium capsules for the National Ignition Facility

Capsules with beryllium ablators have long been considered as alternatives to plastic for the National Ignition Facility laser ; now the superior performance of beryllium is becoming well substantiated. Beryllium capsules have the advantages of relative insensitivity to instability growth, low opacity, high tensile strength, and high thermal conductivity. 3-D calculation with the HYDRA code NTIS Document No. DE-96004569 (M. M. Marinak et.al. in UCRL-LR-105821-95-3) confirm 2-D LASNEX U. B. Zimmerman and W. L. Kruer, Comments Plasmas Phys. Controlled Thermonucl. Fusion, 2, 51(2975) results that particular beryllium capsule designs are several times less sensitive than the CH point design to instability growth from DT ice roughness. These capsule designs contain more ablator mass and leave some beryllium unablated at ignition. By adjusting the level of copper dopant, the unablated mass can increase or decrease, with a corresponding decrease or increase in sensitivity to perturbations. A plastic capsule with the same ablator mass as the beryllium and leaving the same unablated mass also shows this reduced perturbation sensitivity. Beryllium`s low opacity permits the creation of 250 eV capsule designs. Its high tensile strength allows it to contain DT fuel at room temperature. Its high thermal conductivity simplifies cryogenic fielding.

The Science and Technologies for Fusion Energy With Lasers and Direct-Drive Targets

We are carrying out a multidisciplinary multi-institutional program to develop the scientific and technical basis for inertial fusion energy (IFE) based on laser drivers and direct-drive targets. The key components are developed as an integrated system, linking the science, technology, and final application of a 1000-MWe pure-fusion power plant. The science and technologies developed here are flexible enough to be applied to other size systems. The scientific justification for this work is a family of target designs (simulations) that show that direct drive has the potential to provide the high gains needed for a pure-fusion power plant. Two competing lasers are under development: the diode-pumped solid-state laser (DPPSL) and the electron-beam-pumped krypton fluoride (KrF) gas laser. This paper will present the current state of the art in the target designs and lasers, as well as the other IFE technologies required for energy, including final optics (grazing incidence and dielectrics), chambers, and target fabrication, injection, and tracking technologies. All of these are applicable to both laser systems and to other laser IFE-based concepts. However, in some of the higher performance target designs, the DPPSL will require more energy to reach the same yield as with the KrF laser.

Demonstrating a Target Supply for Inertial Fusion Energy

Abstract A central feature of an Inertial Fusion Energy (IFE) power plant is a target that has been compressed and heated to fusion conditions by the energy input of the driver. The technology to economically manufacture and then position cryogenic targets at chamber center is at the heart of future IFE power plants. For direct drive IFE (laser fusion), energy is applied directly to the surface of a spherical CH polymer capsule containing the deuterium-tritium (DT) fusion fuel at approximately 18K. For indirect drive (heavy ion fusion, HIF), the target consists of a similar fuel capsule within a cylindrical metal container or ’’hohlraum’’ which converts the incident driver energy into x-rays to implode the capsule. For either target, it must be accurately delivered to the target chamber center at a rate of about 5-10Hz, with a precisely predicted target location. Future successful fabrication and injection systems must operate at the low cost required for energy production (about

Beta-layering in foam-lined surrogate IFE targets

0.25/target, about 104 less than current costs). Z-pinch driven IFE (ZFE) utilizes high current pulses to compress plasma to produce x-rays that indirectly heat a fusion capsule. ZFE target technologies utilize a repetition rate of about 0.1 Hz with a higher yield. This paper provides an overview of the proposed target methodologies for laser fusion, HIF, and ZFE, and summarizes advances in the unique materials science and technology development programs.

Ultrasonic Characterization of Inertial Confinement Fusion Targets

Abstract Solid deuterium-tritium (the symbol DT is used here to represent the equilibrium mixture of 50% deuterium and 50% tritium, having the molecular composition: 25% D2, 50% deuterium tritide molecules, and 25% T2) (DT) is nucleated from DT-wetted foam and subsequently forms a uniform layer by the beta-layering phenomenon. Compared to DT frozen on smooth metal surfaces, the surface roughness of the inner-lying pure DT solid-vapor interface is substantially lower at all modal values higher than ~10, possibly due to the small-grain-size polycrystalline nature of the solid. For thick layers, deleterious effects are observed, notably the formation of DT-rich vapor voids in the foam matrix and the subsequent propagation of these voids into the pure solid DT layer.

Beta-Layering of Solid Deuterium-Tritium in a Spherical Polycarbonate Shell

Inertial confinement fusion (ICF) targets designed to achieve ignition must meet strict surface smoothness and sphericity requirements. One potentially valuable method for evaluating the quality of these targets is resonant ultrasound spectroscopy (RUS). When applied to simple geometries, such as layered spheres or rectangular parallelepipeds, RUS may yield significant information about alloy homogeneity, elastic constants, cavity geometry the presence of gross defects such as cracking or hemishell bonding problems, and properties of interior fluids. The strengths of RUS techniques for ICF target characterization include applicability at all temperatures of interest with a single apparatus, high sensitivity in frequency spectral measurements, and the inherent acoustic indifference to optically opaque samples. Possible applicational and the limitations of RUS methods for examining layer geometry and material properties are addressed. Preliminary room temperature experiments with a deuterium-filled aluminum shell are used to evaluate the utility of many of the described applications. The frequency spectrum compares favorably with theory and displays measurable mode splitting, acoustic-mode resonance widths indicative of cavity. boundary dissipative mechanisms, and low-Q elastic modes. The acoustic cavity resonance structure confirms the internal gas density and is used to calculate the two lowest even-order cavity boundary perturbation amplitudes.

High-resolution optical measurements of surface roughness for beta-layered deuterium-tritium solid inside a re-entrant copper cylinder

In this paper, the authors examine two of the variables that affect the beta-layering process in which nonuniform layers of solid deuterium-tritium (DT) are driven toward uniformity by beta-decay induced sublimation. For these experiments, a 9 mm diameter polycarbonate sphere was partially filled with a 50-50 mix of DT liquid, frozen, and then held at 17 K. The authors measured the equilibration time constant r as functions of solid layer thickness, He exchange gas pressure, and age. Solid layer thicknesses ranged from 200 {mu}m to 650 {mu}, exchange gas pressures from 0 to 600 torr, and age from 0 to 104 days. Results show a significant final solid layer anisotropy with exchange gas pressures above 5 torr, and r values that increased with age by 0.01 min/day for 200 {mu}m-thick layers, and by 0.5 min/day for 650 {mu}m-thick layers. The time constant is shown to be a weak function of exchange gas pressure.

Development of the Los Alamos National Laboratory Cryogenic Pressure Loader

A high-resolution optical imaging system and custom-designed image analysis software are used to make surface roughness measurements for deuterium-tritium (D-T) solid layers, equilibrated inside a 2-mm-inside-diameter re-entrant copper cylinder. Several experiments are performed that yield D-T layer thicknesses of between 75 and 139 {mu}m, with equilibration temperatures between 17.4 and 18.8 K. A 1024- x 1024-pixel charge-coupled-device imaging camera, coupled with a Maksutov-Cassegrain long-range microscope, produces a 2.5-{mu}m (single-pixel) image resolution. The error function fitting of the image analysis data produces submicron resolution of the layer interior surface finish. The length scale for the cylinder inner bore is just over 6 mm, and the final layer surface roughness for this length ranges from 3- to 1.7-{mu}m root-mean-square. The feasibility is being explored of using these highly uniform and smooth D-T solid layers inside future targets for inertial confinement fusion reactors to produce surface finishes that will meet target design requirements for the National Ignition Facility. Techniques for improving the D-T solid layer surface finish are examined, limitations of the current D-T cell configuration and fuel mix are discussed, and cell configurations for future experiments are described. 10 refs., 8 figs.

Overview of recent tritium target filling, layering, and material testing at Los Alamos national laboratory in support of inertial fusion experiments

Targets for inertial fusion research and ignition at OMEGA, the National Ignition Facility, LMJ, and future facilities rely on beta-radiation-driven layering of spherical cryogenic DT ice layers contained within plastic or metal shells. Plastic shells will be permeation filled at room temperature then cooled to cryogenic temperatures before removal of the overpressure. The cryogenic pressure loader (CPL) was recently developed at Los Alamos National Laboratory as a testbed for studying the filling and layering of plastic target shells with DT. A technical description of the CPL is provided. The CPL consists of a cryostat, which contains a high-pressure permeation cell, and has optical access for investigating beta layering. The cryostat is housed within a tritium glovebox that contains manifolds for supplying high-pressure DT. The CPL shares some design elements with the cryogenic target handling system at the OMEGA facility to allow testing of tritium issues related to that system. The CPL has the capability to fill plastic targets by permeation to pressures up to 100 MPa and to cool them to 15 K. The CPL will accommodate a range of targets and may be modified for future experiments.

Deuterium-tritium beta-layering within a national ignition facility scale polymer target in the lanl cryogenic pressure loader

Abstract The Tritium Science and Engineering (AET-3) Group at Los Alamos National Laboratory (LANL) performs a variety of activities to support Inertial Fusion (IF) research - both to further fundamental fusion science and to develop technologies in support of Inertial Fusion Energy (IFE) power generation. Inertial fusion ignition target designs have a smooth spherical shell of cryogenic Deuterium-Tritium (DT) solid contained within a metal or plastic shell that is a few mm in diameter. Fusion is attained by imploding these shells under the symmetric application of energy beams. For IFE targets the DT solid must also survive the process of injecting it into the power plant reactor. Non-ignition IF targets often require a non-cryogenic DT gas fill of a glass or polymeric shell. In this paper an overview will be given of recent LANL activities to study cryogenic DT layering, observe tritium exposure effects on IF relevant materials, and fill targets in support of IF implosion experiments.