Inertial Confinement Fusion Pellet Research Articles

Parametric design study of a nuclear-pumped laser-driven inertial confinement fusion power plant

In an earlier preliminary design study, we proposed a novel nuclear-pumped laser-driven (NPL) inertial confinement fusion (ICF) power reactor that represents an important variation on the “neutron feedback” concept for ICF. This NPL-driven ICF concept also included an advanced, DT-seeded, D3He-fueled pellet and magnetic protection of the first wall of the reactor chamber. Advantages that were demonstrated for this approach included increased efficiency for laser-to-target energy coupling, increased efficiency for thermalto-electric energy conversion, and reduced neutron activation and waste. The coupling efficiency is enhanced because a nuclear-pumped flashlamp is directly pumped by fission fragments from uranium micropellets within the lamp medium. The thermal conversion efficiency is greater because a large fraction of the ICF pellet&s fusion yield is in charged Finally, the fraction of the fusion yield carried by neutrons is significantly reduced in comparison with pure D-T-fueled pellets; thus, neutron-induced activity in the first wall is decreased and safety is increased. The initial study indicated these factors could result in a required driver energy of 5 MJ (vice 10 MJ currently projected) and a pellet gain of only 50 (vice 100 currently projected) for a feasible l,000-MWe power reactor operating with approximately six pellets per second. The current study includes a refined analysis of an NPL-driven ICF power reactor of this type. A cylindrical design for the fission/NPL blanket is selected as a “natural” geometry for pumping the NPL. Required enrichments and criticalities are then predicted for the multiplication of the fusion neutron yield needed to pump the NPL. Based upon these results, we report a more detailed parametric study of the efficiencies for converting neutron, X-ray, and plasma yields from advanced ICF pellets into electrical and optical energy flows required in this concept. We also examined breeding tritium in a lithium blanket layer. Results from these studies help define topics and parameter spaces for further research on this unique reactor concept.

Calculational Method for Neutron Heating in Dense Plasma Systems, (I). Model and Formulation.

Abstract The effect of neutron heating on the burn characteristics of inertial confinement fusion pellets is investigated by applying the calcuiational method developed in an earlier paper (Part I). The basic equations are time-dependent transport equations for fusion neutrons and recoil ions, being written in terms of the modified Eulerian coordinates originally proposed by Wienke (1974). After incorporating these equations into the one-dimensional hydrodynamics code MEDUSA, burn simulations are made for isobaric D-T pellets models compressed to 1,000 times the liquid density. It is found that in reactor-grade pellets, the inclusion of neutron heating decreases the fuel gain from the values obtained by neglecting the neutron heating. Calculations neglecting the energy transport by recoil ions overestimate the neutron energy deposition to plasma ions. The energy spectrum of neutrons emitted out of a typical D-T burning pellet is also shown. These neutrons contain fast components whose energies reach more ...

Ion cluster interaction with cold targets for ICF: Fragmentation and stopping

Heavy ion clusters are considered for driving an inertial confinement fusion (ICF) pellet. They are shown to be effective at much lower energy than required for atomic ions. Multifragmentation is discussed with a maximum entropy hypothesis. Direct pressure imparted to a DT pellet is estimated. The stopping of strongly correlated ion cluster debris flowing in a degenerate electron jellium target is calculated for polarized and unpolarized (with respect to velocity) projectiles. A reduced stopping occur for elongated configurations at low velocities (υ ⋍ 1 a.u.) and very dense targets. Transverse geometries experience a very large and enhanced stopping. Systematic trends are stressed out. Largest stopping is achieved when debris correlation lengths are smaller than the target jellium screening length.

Initial Investigation of the Feasibility of Antiproton-Catalyzed Inertial Confinement Fusion

The deposition of antiprotons in and subsequent fission of uranium or plutonium has been proposed as a method to assist the driver of an inertial confinement fusion (ICF) pellet and as a spark init...

Coded imaging of thermonuclear neutrons (invited)

Imaging of thermonuclear neutrons from inertial confinement fusion pellets is reviewed. The most promising technique, penumbral imaging, is discussed in detail and the design criteria are outlined. Simulations are presented. More advanced aspects are also briefly discussed, such as nonlinear image reconstruction algorithms and more complex apertures.

Penumbral neutron imaging—optimization and simulation

This paper describes the optimization of the penumbral neutron imaging camera. The optimum parameters are qualified by practical considerations and the resulting camera is tested using a full Monte Carlo simulation. It is shown that an object emitting only 1.3×1010 neutrons/resolution element should be imageable. This result is confirmed by simulations of three different object shapes which indicate that the shape of the core of a compressed inertial confinement fusion pellet may be determined by this technique. It is further shown that these results correspond to the lowest possible imageable yield for any coded-aperture technique using the parameters assumed here. Finally, an alternative aperture and detector arrangement is briefly described which may lower the imageable yield by more than an order of magnitude.

Simultaneous Analysis of Neutron Damage, Tritium Generation and Energy Deposition in Different Cavity Designs for ICF Systems

This article discusses the effect of different potential spectra of neutrons leaking from an inertial confinement fusion (ICF) pellet on integral parameters characterizing the blanket performance. We have compared time dependent and steady state neutron calculations to determine the neutron and gamma spectra from the target. Results show, that assuming a fuel burnup-average ρR parameter,<ρR>b, in the steady state method, realistic spectra can be obtained. The influence of the fuel <ρR>b, ρΔR pusher parameter as well as that of the gammas exploding from the target, on the chamber response has been analyzed.

Repetitive Pulse Accelerator Technology for Light Ion Inertial Confinement Fusion

Successful ignition of an inertial confinement fusion (ICF) pellet is calculated to require that several megajoules of energy be deposited in the pellet's centimeter-sized shell within 10 ns. This implies a driver power of several hundreds of terawatts and power density around 100 TW/cm2 . The Sandia ICF approach is to deposit the energy with beams of 30 MV lithium ions. The first accelerator capable of producing these beams (PBFA II, 100 TW) will be used to study beam formation and target physics on a single pulse basis. To utilize this technology for power production, repetitive pulsing at rates that may be as high as 10 Hz will be required. This paper will overview the technologies being studied for a repetitively pulsed ICF accelerator. As presently conceived, power is supplied by rotating machinery providing 16 MJ in 1 ms. The generator output is transformed to 3 MV, then switched into a pulse compression system using laser triggered spark gaps. These must be synchronized to about 1 ns. Pulse compression is performed with saturable inductor switches, the output being 40 ns, 1.5 MV pulses. These are transformed to 30 MV in a self-magnetically insulated cavity adder structure. Space charge limited ion beams are drawn from anode plasmas with electron counter streaming being magnetically inhibited. The ions are ballistically focused into the entrances of guiding discharge channels for transport to the pellet. The status of component development from the prime power to the ion source will be reviewed.

Numerical modelling of radiation Marshak Waves

In this paper we discuss the importance of radiation transport in inertial confinement fusion (ICF) target design. It is shown that a self similar solution of non-linear heat conduction can be used to estimate the penetration depth of radiation thermal waves (Marshak Waves) in ion-beam ICF pellets. An improved numerical treatment of non-linear heat conduction has been incorporated into the hydrodynamic code MEDUSA-KA to simulate radiation transport in ICF target design studies. The numerical results have been checked against self-similar solutions and a comparison between the two is presented in this paper. We find good agreement between the two. The necessity of using a high-z radiation shield to protect the fuel from radiative preheat is also discussed.

Neutronics and photonics of inertial confinement fusion pellets with internal breeding

The neutronics and photonics performance of a pellet with a small DT core spark trigger, surrounded by a large volume of D to enable tritium and He-3 breeding, is examined. The response to a 70% DD and 30% DT composite neutron spectrum is calculated using either W, Be, or Pb as structural materials at core density radius products ranging from 9.42 to 94.2 kg/m2. At a core density-product of 94.2, the DT neutron source leads to an excess particle multiplication of 0.43 neutrons per source neutron. The percentage of energy leakage from the pellet in the form of escaped neutrons is 42.3% of the source energy for the DT source, and 28.8% for the DD source. The gamma-ray energy percentage deposited in the pellet is 26.7% for the DT source and 106.6% for the DD source. For the pellet with the composite source, the energy multiplication factor is 1.27. Thus the large DD contribution to the composite neutron source results in the pellet performing many of the functions normally reserved for the blanket such as spectral softening, breeding, and neutron and energy multiplication. The neutron energy leakage is 38.4% of the source energy for the composite source. It is estimated that the neutron energy leakage amounts to 10% of the fusion energy, compared with 70% as neutron energy in a DT pellet. These results are significantly different from those encountered in conventional DT inertial confinement designs, and thus lower tritium inventories, higher power densities, reduced radiation damage, and materials activation of the reactor coolant and structure may be achievable.

Neutron Moderation in Inertial Confinement Fusion Pellets and Effects on Damage and Radioactive Inventory

A standard time-dependent neutron transport computer code, TDA, is modified to allow time-varying material density to calculate the neutron spectrum from exploding inertial confinement fusion pellets with ρR values of 0 to 6 g/cm2. Softening of the spectra due to neutron-fuel interactions causes a time-of-flight broadening of the neutron arrival time distribution at the chamber wall. It is found that the total number of displacements per atom (dpa) produced in a graphite first wall increases with the ρR of the pellet over the ρR range investigated because the dpa cross section is larger at lower neutron energy. However, the total helium production decreases with increasing ρR, as does the peak damage rate. Neutron-induced radioactivity generated in a 10-mg iron tamper is of the same magnitude as that produced in the rest of an entire reactor system.

Measurement of core 〈ϱ R〉 and compression in inertial confinement fusion targets by activation technique

The density in the fuel of an inertial confinement fusion pellet may be measured by neutron activation of krypton or, preferably, xenon admixtures.

Inertial Confinement Fusion and Long-Term Nuclear Waste Management

Inertial confinement fusion (ICF) pellet center burnup of fission reactor waste, namely, 14-MeV neutron fission of the very long-lived actinides that post storage problems, is calculated for realistic target designs. A major advantage of pellet center burnup is safety: Only milligram quantities of highly toxic and active material need to be present in the fusion chamber, whereas blanket burnup requires the continued presence of tons of actinides in a small volume. One ICF plant can transmute the actinide waste of up to ten power equivalent fission reactors, i.e., large-scale development appears to provide a foreseeable-future technology that greatly reduces the necessity of high integrity waste storage (burial) time: from 10/sup 7/ to just over 10/sup 2/ yr.