Review of indirect-drive ignition design options for the National Ignition Facility

Abstract

Several inertial confinement fusion (ICF) capsule designs have been proposed as possible candidates for achieving ignition by indirect drive on the National Ignition Facility (NIF) laser [Paisner et al., Laser Focus World 30, 75 (1994)]. This article reviews these designs, their predicted performance using one-, two-, and three-dimensional numerical simulations, and their fabricability. Recent design work at a peak x-ray drive temperature of 250 eV with either 900 or 1300 kJ total laser energy confirms earlier capsule performance estimates [Lindl, Phys. Plasmas 2, 3933 (1995)] that were based on hydrodynamic stability arguments. These simulations at 250 eV and others at the nominal 300 eV drive show that capsules having either copper doped beryllium (Be+Cu) or polyimide (C22H10N2O4) ablators have favorable implosion stability and material fabrication properties. Prototypes of capsules using these ablator materials are being constructed using several techniques: brazing together machined hemishells (Be+Cu), sputter deposition (Be+Cu), and monomer deposition followed by thermal processing (polyimide).

Short-pulse laser-driven x-ray radiography

“…The development of high-energy lasers like NIF and LMJ opens new opportunities for experiments and makes the study of these extreme states of matter with this type of diagnostic even more critical. The transient character of all these experiments requires x-ray radiography with high temporal resolution, which allows measuring density distributions, direct density measurements [1] , instabilities growth [2] and plasma shapes [3] . Depending on the time evolution of the plasma to be investigated, this diagnostic requires an x-ray source with a short duration (≈1 ps), photon energies in the range of a few keV to > 100 keV and a sufficient large photon number to obtain a radiograph.…”

Alternative Fuels for ICF Targets

“…Either methane or diborane would act as an intermediate density layer between an outer container shell of C-H polymer and the inner pure DT layer. The reduced Atwood number at the interfaces should lessen the shock reflection and the destabilizing effects of the "feed-in feed-out" processes 5 , thus promoting ignition. Can the amount of excess pure DT be too great?…”

“…However, for a polyimide ablator target, and especially for a C-H plastic ablator, the DT surface may need to be smoothed by an enhancement technique such as infrared illumination 3 or Joule heating 4 . The sensitivity of fusion yield to the internal DT ice roughness is a function of many factors, including feed-in/feed-out instabilities 5 . But the relatively low density of solid DT, 0.25 g/cm 3 , results in a high Atwood number for the ablator/fuel interface and is certainly a contributor 6 .…”

Dynamics of converging laser-created plasmas in semicylindrical cavities studied using soft x-ray laser interferometry

“…Laser heating of cylindrical hohlraums [1,2] creates pressure gradients near the walls that radially accelerate the plasma distributed along the surface towards the axis of the cavity, where it collides. This interaction of converging plasmas depends on their degree of collisionality and can range from stagnation to extended interpenetration, as determined by the geometry and irradiation conditions [3].…”