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In this report we describe a preliminary concept for a Tritium Removal System (TRS) to remove tritium that is generated in the ⁹⁹Mo production process. Preliminary calculations have been performed to evaluate an approximate size for the system. The concept described utilizes well-established detritiation technology based on catalytic oxidation of tritium and tritiated hydrocarbons to water in a high temperature (400 °C) reactor and capture of water in a molecular sieve bed. The TRS concept involves use of a single system that would cycle through each of the seven online target systems and remove tritium that has been accumulated after one week’s run time. The TRS would perform cleanup operations on each target system for a period of approximately 24 hours. This would occur while the system is still online and just prior to target replacement, so tritium levels would at their minimum values for target replacement. In the concept, during normal operation a small fraction (1%) of the helium recirculating in the system would be diverted through the TRS and returned to the flow loop. With this approach sufficient levels of detritiation can be accomplished in a 24 hour period. In the study it was found that because ofmore » the need to maintain low oxygen levels in the system (

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. (authors)

A conceptual design is proposed to support the recovery of un-expended fuel, ash, and associated post-detonation products resident in plasma exhaust from a {approx}2 GWIFE direct drive power reactor. The design includes systems for the safe and efficient collection, processing, and purification of plasma exhaust fuel components. The system has been conceptually designed and sized such that tritium bred within blankets, lining the reactor target chamber, can also be collected, processed, and introduced into the fuel cycle. The system will nominally be sized to process {approx}2 kg of tritium per day and is designed to link directly to the target chamber vacuum pumping system. An effort to model the fuel recovery system (FRS) using the Aspen Plus engineering code has commenced. The system design supports processing effluent gases from the reactor directly from the exhaust of the vacuum pumping system or in batch mode, via a buffer vessel in the Receiving and Analysis System. Emphasis is on nuclear safety, reliability, and redundancy as to maximize availability. The primary goal of the fuel recovery system design is to economically recycle components of direct drive IFE fuel. The FRS design is presented as a facility sub-system in the context of supporting themore » larger goal of producing safe and economical IFE power. (authors)« less

A system containing a combination of getters (Zr-Mn-Fe, SAES St909; and Zr 2 Fe, SAES St198) was used to process the nitrogen-hydrogen-helium atmosphere in a glovebox used for handling metal tritide samples. During routine operations, the glovebox atmosphere is recirculated and hydrogenous impurities (i.e. CQ 4 , Q 2 O, and NQ 3 , where Q =H, D, T) are decomposed (cracked) and removed by Zr-Mn-Fe without absorbing elemental hydrogen isotopes. If the tritium content of the glovebox atmosphere becomes unacceptably high, the getter system can rapidly strip the glovebox atmosphere of all hydrogen isotopes by absorption on the Zr 2 Fe, thus lessening the burden on the facility waste gas treatment system. The getter system was designed for high flowrate (> 100 l/min), which is achieved by using a honeycomb support for the getter pellets and 1.27-cm diameter tubing throughout the system for reduced pressure drop. The novel getter bed design also includes an integral preheater and copper liner to accommodate swelling of the getter pellets, which occurs during loading with oxygen and carbon impurities. Non-tritium functional tests were conducted to determine the gettering efficiencies at different getter bed temperatures and flowrates by recirculating gas through the system from a 6-m 3 glovebox containing known concentrations of impurities.

A set of laser implosion experiments were conducted at the OMEGA laser at the University of Rochester, Laboratory for Laser Energetics (LLE) to study the effect of 3 He concentration in DT-filled target shells on fusion yield in ICF implosions.. Eleven laser fusion shells consisting of 1100-μm diameter, hollow, fused silica spheres with 4.6 to 4.7-μm-thick walls were loaded with 520 kPa of deuterium-tritium (DT) and then with 3 He (101.3 or 520 kPa). The 3 He permeabilities of the shells were determined by measuring the pressure rate of rise into a system with known volume. A mathematical method was developed that relied on the experimental fill pressure and time, and the rate of rise data to solve differential equations using MathCAD to simultaneously calculate 3 He permeability and initial 3 He partial pressure inside the shell. Because of the high permeation rate for 3 He out of the shells compared to that for DT gas, shells had to be recharged with 3 He immediately before being laser imploded or "shot" at LLE. The 3 He partial pressure in each individual shell at shot time was calculated from the measured 3 He permeability. Two different partial pressures of 3 He inside the shell were shown to reduce neutron and gamma yields during implosion.

Neutron imaging diagnostics are needed for understanding the principles of fusion ignition. Current experiments on the University of Rochester OMEGA laser facility and future experiments at the NIF...

An important component of the U.S. effort to achieve thermonuclear ignition in 2010 on the National Ignition Facility is the development of high quality 2 mm diameter spherical capsules to function as the ablator and contain the cryogenic DT fuel. Three ignition capsule designs have been developed, and detailed fabrication specifications for each design have been established and placed under change control. A research program with activities coordinated mainly between Lawrence Livermore, General Atomics and Los Alamos is underway to demonstrate fabrication of capsules meeting specifications. The point design for ignition campaigns beginning in 2010 is a Cu-doped Be capsule that has a radial gradient in Cu dopant level in the capsule wall. This capsule is being produced by sputter deposition of Be and Cu onto either a hollow glow discharge polymer (GDP) spherical mandrel or a solid spherical mandrel, followed by removal of the mandrel and polishing of the capsule. A key goal in the U.S. is to demonstrate fabrication of this capsule by the end of 2006. Two other ignition capsule designs are also being developed as contingencies to the point design. One contingency design is a GDP capsule that has a radial Ge dopant level in its wall. This capsule is produced by co-deposition of Ge and GDP onto a PAMS mandrel followed by thermal removal of the mandrel. The second contingency design is a uniform Cu-doped Be capsule that is fabricated from high purity fine grain Be0.3at.%Cu alloy using a precision machining route followed by polishing. Ignition targets to be fielded in 2010 will be filled with DT fuel through a small fill hole. Laser drilling capability has been developed and used to drill approximately 5 μm diameter holes through capsule walls for DT filling. Characterization methods necessary for characterizing capsules are being developed.

Low density polymer foams have been an integral part of targets used in Inertial Confinement Fusion (ICF) experiments. In order to accomplish this the formulation of the High Internal Phase Emulsion (HIPE) foam had to be optimized. Along with a change in density and the incorporation of fillers into a foam, there is also a change in the pore size and the distribution of pore sizes. The emulsion technique used to produce these foams is amenable to modification providing foams with a wide range of densities, 15 mg/cm3 to 500 mg/cm3, and metal loadings up to 90 percent by weight. A qualitative distribution of pore size can be seen visually with scanning electron microscopy (SEM). At low metal loadings SEM did not observe the submicron metal particles. The distribution of these particles in the foam matrix was seen with near-infrared imaging.

Continued advances in the design of ignition targets have stimulating new development paths for target fabrication, with potentially important simplifications for fielding cryogenic ignition targets for the National Ignition Facility. Including graded dopants in ablators as well as optimizing capsule and fuel layer dimensions increase implosion stability. This has led to developments of micron-scale fill tubes to fill and field the targets. Rapid progress has been made in development of the graded dopant layers in capsules as well as their characterization, in fabrication methods for micro-fill-tubes, and in fuel fill control with these fill tubes. Phase-contrast x-ray radiography has allowed characterization of fuel layers in beryllium targets. This target development program includes participation from General Atomics, Lawrence Livermore National Laboratory, and Los Alamos National Laboratory.

Planar samples of varying thicknesses of both CH and CD glow discharge polymer have been measured with Fourier transform infrared (FTIR) spectroscopy before and after exposure to deuterium-tritium (DT) gas at elevated temperature and pressure. Planar samples of polyimide films made from both hydrogenated and deuterated precursors have also been examined by FTIR before and after DT exposure. The post-exposure FTIR spectra demonstrated no measurable exchange of hydrogen with deuterium or tritium for either polymer. Evidence for oxidation of the glow discharge polymer due to atmospheric oxygen was the only chemical change indicated by the FTIR data.

Work is underway at Los Alamos National Laboratory to fabricate machined-and-bonded target capsules of Be-6 wt% Cu for the National Ignition Facility. Significant progress has been made in producing material with the desired composition, purity, and homogeneity of composition, by arc melting. This material is thermomechanically processed by equal channel angular extrusion, to break down the coarse ascast structure and refine the grain size, to about 20 μm. Machining with diamond tooling results in a significant improvement of the as-machined roughness, that also results in improved bond strengths. Bonding with a sputtered layer of Al can achieve high strengths with a bond 1.2 μm thick, and thinner bonds are being investigated. Laser-drilled holes and fill-tube counterbores produced by electrodischarge machining appear to be feasible, but will require improvements in specimen positioning.

Inertial Confinement Fusion (ICF) energy hohlraums are composed of a high-Z material filled with foam. Because of the small pore size and transparency, silica aerogels are used in some ICF targets. The traditional synthesis of silica aerogels require sol-gel polymerization of silicon alkoxide followed by supercritical drying. Some constituents in sol-gel polymerization have been found to contribute to leaching of certain metals at the silica/metal interface. Since the hohlraums are composed of metals, possible chemical reactivity at the silica aerogel and metal hohlraum interface was investigated. The hohlraums studied are aluminum lined with either copper or copper/chromium. Upon initial inspection, the aerogel appeared transparent and uniform, however, closer inspection of the copper wall suggested possible leaching. Alternatively the quality of the aerogel in the copper-chromium hohlraum was very poor with the chromium layer of the hohlraum and some copper completely etched. Control experiments were used to determine the cause of the leaching. When copper is in the presence of sol-gel constituents, Cu 2+ ion formed, thus leaching copper from the hohlraum walls. In the presence of chromium, Cr 2 O 7 2- or CrO 4 2- was identified in solution with the Cu 2+ , these anions are believed to form copper chromite under the aerogel synthesis procedures utilized.

The purpose of the experiments described in this paper was to expose samples of polymeric materials to a mixture of deuterium-tritium (DT) gas at elevated temperature and pressure to investigate the effects (i.e. damage) on the materials. The materials and exposure parameters were chosen with to be relevant to proposed uses of similar materials in inertial fusion ignition experiments at the National Ignition Facility. Two types of samples were exposed and tested. The first type consisted of 10 4-lead ribbon cables of fine manganin wire insulated with polyimide. Wires of this type are proposed for use in thermal shimming of hohlraums and the goal of this experiment was to measure the change in electrical resistance of the insulation due to tritium exposure. The second type of sample consisted of 20 planar polymer samples that may be used as ignition capsule materials. The exposure was at 34.5 GPa (5010 psia) and 70 C for 48 hours. The change in electrical resistance of the wire insulation will be presented. The results for capsule materials will be presented in a separate paper in this issue.

Beta-layering, the process of beta-decay heat-driven mass redistribution, has been demonstrated in a deuterium-tritium (D-T)-filled polymer sphere of the type required for fusion ignition experiments at the National Ignition Facility. This is the first report, to the best of the authors' knowledge, of a D-T layer formed in a permeation-filled sphere. The 2-mm-diam sphere was filled with D-T by permeation; cooled to cryogenic temperatures while in the high-pressure permeation vessel; and, while cold, removed to an optical axis where the D-T was frozen, melted, and beta-layered in a series of experiments over several weeks' time. This work was performed in the Los Alamos National Laboratory cryogenic pressure loader system. The beta-layering time constant was 24.0 {+-} 2.5 min, less than the theoretical value of 26.8 min, and not showing the significant increase due to build-up of {sup 3}He often observed in beta-layered samples. Supercooling of the liquid D-T was observed. Neither the polymer target nor its tenting material showed visual signs of degradation after 5 weeks of exposure to D-T. Small external thermal gradients were used to shift the D-T material back and forth within the sphere.

Success in inertial confinement fusion (ICF) requires sophisticated, characterized targets. The increasing fidelity of three-dimensional (3D), radiation hydrodynamic computer codes has made it possible to design targets for ICF which can compensate for limitations in the existing single shot laser and Z pinch ICF drivers. Developments in ICF target fabrication technology allow more esoteric target designs to be fabricated. At present, requirements require new deterministic nano-material fabrication on micro scale.

High-pressure laser chemical vapor deposition (HP-LCVD) is a powerful tool for growing complex microstructures at rapid rates. Not only is it possible to deposit functionally graded materials, but new metastable phases, alloys, and composite materials may be realized. In this paper, the diversity of microstructures that may be obtained through HP-LCVD is demonstrated, including the growth of metastable materials, e.g., diamond-like carbon (DLC). For the first time, a pressure-temperature (P-T) phase diagram has been created for HP-LCVD, identifying nine distinct material phases of carbon from ethene. Regions of high sp 3 content are identified via Raman spectroscopy. The kinetics, rate limitations, and thermodynamics of the process are also characterized at hyperbaric pressures, creating a first-ever process-rate map-covering the entire useful pressure range for ethene. Thermodynamically enhanced growth is also documented for the first time, where the contribution of the heat of reaction is much greater than the incident laser power-demonstrating a quasi-self-sustaining reaction. Finally, sufficient information is provided to reconstruct specific fiber geometries, structures, and growth rates for potential industrial production of carbon fibers from the gas phase.

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 economical...

The target for an Inertial Fusion Energy (IFE) power plant is compressed and heated to fusion conditions by the driver beams. The “Target Fabrication Facility” (TFF) of a 1000 MW(e) IFE power plant must supply over 500,000 targets per day. The target is then injected into the target chamber at a rate of 5–10 Hz and tracked precisely so the driver beams can be directed to the target. The feasibility of developing successful fabrication and injection methodologies at the low cost required for energy production (about $0.25/target, about 104 less than current costs) is a critical issue for inertial fusion. The technologies for producing Heavy Ion Fusion (HIF) targets have significant overlaps and synergisms with current-day inertial fusion experimental targets and with laser fusion (direct drive) IFE targets. Capsule formation and characterization, permeation filling, and layering of the DT using a cryogenic fluidized bed are common methodologies shared between laser fusion and HIF. Specific to HIF targets are the techniques for fabricating and assembling the hohlraum components. We will report on experimental progress with the Laser-assisted Chemical Vapor Deposition (LCVD) technique to produce “micro-engineered” low-density metallic foams for the hohlraum, and calculations of hohlraums materials performance during handling. Fiber growth by LCVD in arrays has been demonstrated for the first time, important to achieve the volume production needed for IFE. We have also evaluated a variety of hohlraum material selections, with consideration of target physics, cost, ES&H, activation, and compatibility with the molten salt coolant. These materials include selections for once-through and for recycle scenarios. We have performed a cost analysis for an “nth-of-a-kind” Target Fabrication Facility using our current assumptions about the production processes. Some of these scenarios result in future target manufacturing costs consistent with economical electricity production.

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. This is true whether the driver is a laser system, heavy ion beams or Z-pinch system. The IFE target fabrication, injection and tracking programmes are focusing on methods that will scale to mass production. We are working closely with target designers, and power plant systems specialists, to make specifications and material selections that will satisfy a wide range of required and desirable target characteristics. One-of-a-kind capsules produced for today’s inertial confinement fusion experiments are estimated to cost about US$2500 each. Design studies of cost-effective power production from laser and heavy-ion driven IFE have suggested a cost goal of about $0.25–0.30 for each injected target (corresponding to ∼10% of the ‘electricity value’ in a target). While a four orders of magnitude cost reduction may seem at first to be nearly impossible, there are many factors that suggest this is achievable. This paper summarizes the design, specifications, requirements and proposed manufacturing processes for the future for laser fusion, heavy ion fusion and Z-pinch driven targets. These target manufacturing processes have been developed—and are proposed—based on the unique materials science and technology programmes that are ongoing for each of the target concepts. We describe the paradigm shifts in target manufacturing methodologies that will be needed to achieve orders of magnitude reductions in target costs, and summarize the results of ‘nth-of-a-kind’ plant layouts and cost estimates for future IFE power plant fuelling. These engineering studies estimate the cost of the target supply in a fusion economy, and show that costs are within the range of commercial feasibility for electricity production.

The velocities and temperatures of shock waves generated by laser-driven hohlraum radiation fields have been measured in indirect-drive inertial confinement fusion (ICF) capsule ablator materials. Time-resolved measurements of the preheat temperature ahead of the shock front have been performed and included in the analysis. Measurements of the x-ray burnthrough of the ablation front and the ablator x-ray re-emission have also been made in the Cu-doped beryllium, polyimide, and Ge-doped CH ablator samples. The experiments utilize 15 beams of the University of Rochester Omega Laser [Soures et al., Phys. Plasmas 3, 2108 (1996)] to heat hohlraums to radiation temperatures of ∼120–200 eV. In the experiments, planar samples of ablator material are exposed to the hohlraum radiation field, generating shocks in the range of 10–50 Mbars. The experimental results are compared to integrated two-dimensional Lasnex [G. B. Zimmerman and W. L. Kruer, Comments Plasma Phys. Control. Fusion 2, 51 (1975)] calculations, in wh...

Beryllium doped with 6 weight % copper is the material of choice for fabrication of target capsules for the National Ignition Facility because of its combination of attractive neutronic, electronic, physical, and mechanical properties. The target capsules are 2 mm in diameter and thin-walled (150 microns) and must meet demanding dimensional specifications. The material must be fine-grained and of low inclusion content. Arc-melted Be-Cu is being produced to eliminate the oxide content that is inevitably present in conventional powder-metallurgy materials. Equal channel angular extrusion (ECAE) is being used to refine the as-cast grain structure. Be-Cu rods produced by the arc-melting process (5 mm in diameter by 30 mm in length) are enclosed in nickel cans with electron-beam welded plugs. The Be-in-Ni billets (9.5 mm in diameter by 45 mm in length) have been processed by ECAE at temperatures from 500 to 750°C in tooling with a 120° angle. Selected samples have been annealed for 1 hour at temperatures from 700 to 775°C. The ECAE processing creates a heavily deformed and finely subdivided structure, and the annealing can produce an equiaxed microstructure with a grain size of approximately 20 μm.

High Internal Phase Emulsion (HIPE) polystyrene foams have been made at LANL for the past decade. It is a robust system that offers flexibility in tailoring density and the incorporation of halogens and metals. As target designs become more complex the demands placed on the foams are more stringent. Parts are machined from 30 mg/cm{sup 3} foams to thicknesses of 50 {mu}m. At three percent of full density these foams are to withstand extraction with ethanol to remove the wax utilized as a machining aid and not allow shrinkage or warpage. In order to accomplish this the formulation of the HIPE foam had to be modified. Recently some new processing issues have arisen. At low densities voids have become a problem. To determine a formulation that reduces void content and allows minimum shrinkage, experimental design was utilized. We also developed image analysis techniques that allow us to quantify the amount of voids in the system. These techniques also allow us to evaluate the surface finish of the foam. In order to machine these low density foams to the tolerance required with an optimum surface finish the foams are backfilled with Brij 78, an alcohol soluble wax. After the part ismore » machined, the Brij is leached out. Recent batches of Brij have exhibited high shrinkage, which in turn affects the surface finish of the foam.« less

Experimental campaigns are being conducted at the 60 beam OMEGA laser at the University of Rochester's Laboratory for Laser Energetics to acquire data to validate hydrodynamic models in the high energy-density regime. This paper describes targets that have been developed and constructed for these experimental campaigns. Targets are 860 {mu}m inner diameter by 2.2 mm length cylinders with 70 {mu}m thick polymer ablator. On the ablator inner surface and located halfway along the axis of the cylinder is a 500 {mu}m wide Al marker band. Band thicknesses in the range 8-16 microns are used. CH foam with densities in the range 30-90 mg/cc fills the inside of the cylinder. While these targets have been fabricated for years, several new improvements and features have recently been developed. Improvements include the use of epoxy instead of polystyrene for the ablator, and the use of electrodeposited Al for the marker band. A critical feature of the target is the surface feature that is placed on the marker band. Experiments are aimed at understanding the hydrodynamic behavior of imploding cylinders as a function of this surface feature. Recent development work has focused on production of engineered surface features on the target marker band. Usingmore » a fast tool servo on a diamond turning lathe, a wide range of specified surface features have been produced. This paper will address improvements to the cylinder targets as well as current development efforts.« less

Beryllium-copper alloy (Be0.9%Cu) ICF capsules are being developed for the pursuit of thermonuclear ignition at the National Ignition Facility (NIF). Success of this capsule material requires that its shock propagation and radiation burnthrough characteristics be accurately understood. To this end, experiments are being conducted to measure the shock propagation and radiation burnthrough properties of Be0.9%Cu alloy. These experiments involve measurements on small Be0.9%Cu wedge, step and flat samples. Samples are mounted on 1.6-mm-diameter x 1.2-mm-length hohlraums that are illuminated by the OMEGA laser at the University of Rochester. X-rays produced by the hohlraum drive the sample. A streaked optical pyrometer detects breakout of the shock produced by the X-ray pulse. In this paper we describe synthesis of the alloy material, fabrication and characterization of samples, and assembly of the targets. Samples were produced from Be0.9%Cu alloy that was synthesized by hot isostatic pressing of Be powder and copper flake. Samples were 850 {mu}m diameter disks with varying thickness in the case of wedge and step samples, and uniform thickness in the case of flat samples. Sample thickness varied in the range 10-90 {mu}m. Samples were prepared by precision lathe machining and electric discharge machining. The samples were characterized by amore » Veeco white light interferometer and an optical thickness measurement device that simultaneously measured the upper and lower surface contours of samples using two confocal laser probes. Several campaigns with these samples have been conducted over the past two years.« less

The Target Fabrication Facility (TFF) of an IFE power plant must supply about 500,000 targets per day. The targets are injected into the target chamber at a rate of 5-10 Hz and tracked precisely so the driver beams can be directed to the target. The feasibility of developing successful fabrication and injection methodologies at the low cost required for energy production (about $0.25/target, about 10 4 less than current costs) is a critical issue for inertial fusion. To help identify major cost factors and technology development needs, we have utilized a classic chemical engineering approach to the TFF. The analyses assume an nth-of-a-kind TFF and utilize standard industrial engineering cost factors. The results indicate that the direct drive target can be produced for about $0.16 each. Iterations are still underway for the indirect drive target. These cost analyses assume that the process development is accomplished to allow scaling of current laboratory methods to larger sizes, while still meeting target specifications. A development program is underway at various laboratories to support this scale-up.

Addressing the issues associated with target fabrication and injection is a major part of an international program to establish the feasibility of inertial fusion energy (IFE), both for laser-driven and heavy-ion driven concepts. A summary of the unique materials science and chemistry research programs associated with supplying targets for an IFE power plant is presented. The cost of manufacturing targets for commercial power applications is a significant perceived feasibility issue for IFE, and preliminary estimates of Target Fabrication Facility costs are discussed for both direct and indirect drive systems.

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 (

Low-density polymer foams have been an integral part of targets used in inertial confinement fusion (ICF) experiments. Target designs are unique in the ICF program, and targets are made on an individual basis. Costs for these targets are high due to the time required to machine, assemble, and characterize each target. To produce targets in high volume and at low cost, a polymer system is required that is amenable to scale up. High internal phase emulsion (HIPE) polystyrene is a robust system that offers great flexibility in terms of tailoring the density and incorporating metal dopants. Emulsions used to fabricate HIPE foams currently are made in a batch process. With the use of metering pumps for both the water and oil phases, emulsions can be produced in a continuous process. This not only makes these foams potential candidates for direct-drive capsules, but high-Z dopants can be metered in making these foams attractive for hohlraum components in indirect-drive systems. Preparation of HIPE foams are discussed for both direct-drive and indirect-drive systems.

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 15K. The CPL will accommodate a range of targets and may be modified for future experiments.

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 {approx}200 mg/cm{sup 3} 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 seemmore » acceptable for future IFE target DT filling facilities.« less

The Target Fabrication Facility (TFF) of an inertial fusion energy (IFE) power plant must supply about 500,000 targets per day. The target is injected into the target chamber at a rate of 5–10 Hz and tracked precisely so the heavy ion driver beams can be directed to the target. The feasibility of developing successful fabrication and injection methodologies at the low cost required for energy production (about $0.25/target, approximately 104 times less than current costs) is a critical issue for inertial fusion energy. A significant program is underway to develop the high-volume methods to supply economical IFE targets. This article reviews the requirements for heavy ion driven IFE target fabrication and injection, and presents the current status of and results from the development program. For the first time, an entire pathway from beginning to end is outlined for fabrication of a high-gain, distributed radiator target. A significant development and scale-up program will be necessary to implement this pathway for mass production of IFE targets.

Fueling of a commercial Inertial Fusion Energy (IFE) power plant consists of supplying about 500,000 fusion targets each day. The most challenging type of target in this regard is for laser-driven, direct drive IFE. Spherical capsules with cryogenic DT fuel must be injected into the center of a reaction chamber operating at temperatures as high as 1500 °C and possibly containing as much as 0.5 Torr of xenon fill gas. The DT layer must remain highly symmetric, have a smooth inner ice surface finish, and reach the chamber center at a temperature of about 18.5 K. This target must be positioned at the center of the chamber with a placement accuracy of ±5 mm. The accuracy of alignment of the laser driver beams and the target in its final position must be within ±20 μm. All this must be repeated six times per second. The method proposed to meet these requirements is injecting the targets into the reaction chamber at high speed (∼400 m/s), tracking them, and hitting them on the fly with steerable driver beams. The challenging scientific and technological issues associated with this task are being addressed through a combination of analyses, modeling, materials property measurements, and demonstration tests with representative injection equipment. Measurements of relevant DT properties are planned at Los Alamos National Laboratory. An experimental target injection and tracking system is now being designed to support the development of survivable targets and demonstrate successful injection scenarios. Analyses of target heating are underway. Calculations have shown that the direct drive target must have a highly reflective outer surface to prevent excess heating by thermal radiation. In addition, heating by hot chamber fill gas during injection far outweighs the thermal radiation. It is concluded that the dry-wall, gas-filled reaction chambers must have gas pressures less than previously assumed in order to prevent excessive heating in the current direct drive target designs. An integrated power plant systems study to address this issue has been initiated.

At the heart of an inertial fusion energy (IFE) power plant is a target that has been compressed and heated to fusion conditions by the incident driver energy beams. The “Target Factory” at an inertial fusion power plant must produce about 500,000 targets per day, fill them with deuterium–tritium fuel, cool them to cryogenic temperature, and layer the solid fuel into a symmetric and smooth shell inside the capsule. The target must then accurately be delivered to the target chamber center at a rate of about 5 Hz, with a precisely predicted target location. These fragile targets must survive injection into the target chamber without damage. While IFE power plant design studies have presented plausible scenarios for IFE target fabrication and injection, these issues have become “believability” issues for IFE. A credible pathway for development of accurate, economic and reliable IFE target fabrication and injection must be demonstrated before we can proceed with the next major step in the IFE Program, the construction of an IFE Integrated Research Experiment. General Atomics is designing, constructing, and testing an experimental Target Injection and Tracking System to develop the scientific understanding necessary for injection of IFE targets into a high temperature reaction chamber. This paper summarizes the requirements for IFE target fabrication and injection, reviews the results from the studies that predict success, discusses the development program now underway, and presents the current status of and results from that program.

Fuelling of a commercial inertial fusion energy (IFE) power plant consists of supplying about 500 000 fusion targets each day. The most challenging type of target in this regard is that for laser driven direct drive IFE power plants. Spherical capsules with cryogenic DT fuel must be injected into the centre of a reaction chamber operating at temperatures as high as 1500° C and possibly containing as much as 0.5 torr of xenon fill gas. The DT layer must remain highly symmetric, have a smooth inner ice surface finish and reach the chamber centre (CC) at a temperature of about 18.5 K. This target must be positioned at the centre of the chamber with a placement accuracy of ±5 mm. The accuracy of alignment of the laser driver beams and the target in its final position must be within ±20 μm. All this must be repeated six times per second. The method proposed to meet these requirements is to inject the targets into the reaction chamber at high speed ( ≈ 400 m/s), track them, and hit them in flight with steerable driver beams. The challenging scientific and technological issues associated with this task are being addressed through a combination of analyses, modelling, materials property measurements and demonstration tests with representative injection equipment. Measurements of relevant DT properties are planned at Los Alamos National Laboratory. An experimental target injection and tracking system is now being designed to support the development of survivable targets and demonstrate successful injection scenarios. Analyses of target heating are under way. Calculations have shown that a direct drive target must have a highly reflective outer surface to prevent excess heating by thermal radiation. In addition, heating by hot chamber fill gas during injection far outweighs that by the thermal radiation. It is concluded that the dry wall, gas filled reaction chambers must have gas pressures and wall temperatures less than previously assumed in order to prevent excessive heating in current direct drive target designs. An integrated power plant systems study to address this issue has been initiated.

Capability to fill inertial confinement fusion (ICF) targets with DT has recently been established at the Weapons Engineering Tritium Facility (WETF) at Los Alamos National Laboratory (LANL). The target filling system provides DT-filled glass targets for the U.S. National ICF Program. Tritium storage, purification, mixing, analysis, and high pressure capabilities at WETF are used to provide DT at pressures up to 400 atm to a target filling cell that can operate at temperatures to 400{degree}C. Isotopically pure tritium is obtained from the Tritium Systems Test Assembly at LANL, and typically has purities of 99% tritium or better. At WETF, a palladium-silver diffuser is used for removal of decay {sup 3}He from tritium prior to mixing with deuterium. After preparation, DT mixtures are stored in a passivated volume to minimize impurity accumulation from stainless steel. Analysis of tritium and DT mixtures is performed with a quadrupole mass spectrometer/beta scintillation detector system that utilizes an analytical technique previously developed at LANL to provide hydrogen isotope, helium, and impurity analysis. Glass targets are filled in aluminum eggcrates. The target filling cell has been designed to contain two eggcrates while maintaining isothermal conditions across the eggcrates during diffusion filling of targets. Results from amore » cryogenic condensation technique performed at Lawrence Livermore National Laboratory have confirmed the fill pressures. 3 refs., 5 figs., 1 tab.« less

A reaction engineering approach was used to design a SAES St 198 metal getter reactor for a glovebox detritiation system. The detritiation system will be used to decontaminate and decommission an Li(D, T)-contaminated glovebox previously used in the U.S. nuclear weapons program. The approach involved development of a model that calculates reactor breakthrough curves as a function of various reactor physical parameters. Experiments involving flow of deuterium in nitrogen through a small metal getter reactor validated the model. The model was then used to investigate the effects of temperature, getter pellet size, reactor diameter, and reactor volume on the reactor performance. The resulting design was a 7 cm diam. by 40 cm long cylindrical reactor that operates at 250 {degree}C, and is filled with 5 kg of as-received SAES St 198 getter pellets. The reactor handles a flow rate of 100 L/min. An St 909 getter reactor was used upstream of the St 198 reactor for impurity removal and water decomposition. The glovebox cleanup system design and getter reactor mechanical design are discussed. 14 refs., 6 figs.

A first set of shock timing, laser-plasma interaction, hohlraum energetics and hydrodynamic experiments have been performed using the first 4 beams of the National Ignition Facility (NIF), in support of indirect drive Inertial Confinement Fusion (ICF) and High Energy Density Physics (HEDP). In parallel, a robust set of optical and X-ray spectrometers, interferometer, calorimeters and imagers have been activated. The experiments have been undertaken with laser powers and energies of up to 8 TW and 17 kJ in flattop and shaped 1–9 ns pulses focused with various beam smoothing options. The experiments have demonstrated excellent agreement between measured and predicted laser-target coupling in foils and hohlraums, even when extended to a longer pulse regime unattainable at previous laser facilities, validated the predicted effects of beam smoothing on intense laser beam propagation in long scale-length plasmas and begun to test 3D codes by extending the study of laser driven hydrodynamic jets to 3D geometries.

A conceptual design has been developed for the recovery of un-expended fuel, ash, and associated post-detonation products from a ~ 2 GW IFE power reactor. The conceptual design incorporates systems for the safe, efficient collection, processing, and purification of IFE plasma exhaust fuel components. The system has been designed and sized such that tritium bred within blankets can also be collected, processed, and introduced into the fuel cycle. The system is nominally sized to process ~2 kg of tritium per day and is designed to link directly to the target chamber mechanical pumping system. The plasma exhaust can be directly processed from the exhaust of the vacuum pumping system or can be processed in batch mode from buffer vessels in the receiving and analysis system (RAS). Systems for the accurate measurement of material in-process (MIP) have been included. Design emphasis is on safety, reliability, redundancy, and efficiency in order to maximize availability. The primary goal of the fuel recovery system (FRS) design is to economically recycle components of IFE fuel back to the target manufacturers in a fashion by which fuel components are rapidly made available for re-use thus lowering the total active inventory. The FRS design is presented as a facility sub-system in the context of supporting the safe and efficient operation of the IFE target chamber.

This report including appendices provides information to complete this deliverable. It summarizes the important features of each ablator material, with particular focus to its usefulness for ignition capsules. More detailed discussions of each ablator type are in the Appendix. Included at the end of each separate discussion in the Appendix is a list of all published work with an ICF focus on that ablator type. This report is organized into Be based and polymer (C) based ablators. We summarize status, outstanding issues, and how we plan to address them. Details are in the Appendix. For Be there are two fabrication routes, one by machining bulk pieces into hemi-shells which are then bonded together, and the other by sputtering Be with Cu dopant onto spherical plastic mandrels to build up a wall. This method allows for radial variation in the Cu dopant concentration, while the machining approach is best suited to a uniform doping level. For plastic, we have already made a down select, eliminating polyimide because its performance as an ablator has been seen to be significantly different from that predicted by simulations. The other polymer, GDP (glow discharge polymer or sometimes called plasma polymer) comes in both a normalmore » (hydrogenated) and deuterated form. There are differences between them (besides the H or D) and these will be detailed. The choice between them will be determined in part by cryogenic measurement of the IR absorption spectrum of DT scheduled to occur in the next few months. An initial list of specifications for ignition targets exists. However these specifications are continuing to evolve. This is due to evolving plans for NIF's deliverable energy and to more refined design simulations. Many requirements are not well specified due to lack of knowledge of the effect on the implosion. These requirements include: grain size and texture, fill hole size, fill tube size, bond joint thickness, allowable porosity (size and number), diameter and wall thickness. Experiments are currently underway to assess the effect of grain size and texture, fill hole size, fill tube size, and bond joint thickness on the implosion. These parameters are being specified by simulation where possible. In cases such as bond joint thickness and fill hole size, the calculation methods are still under development. Completion of ignition capsule development will require final specifications.« less

General Atomics is designing and building the OMEGA Cryogenic Target System (Fig. 1) for the University of Rochester's Laboratory for Laser Energetics. The purpose of this system is to deliver millimeter sized polymer shell targets to the center of the target chamber for inertial confinement fusion experiments. Prior to insertion, these targets are filled to pressures as high as 1500 atm with hydrogen isotopes (DT), the gas is cryogenically condensed, and the condensed material is layered to form a uniform inner shell. GA has demonstrated the successful filling with D/sub 2/ of plastic targets to 1100 atm, cooling, and cold transport utilizing prototype equipment. In addition to proving the viability of the proposed fill process, the prototypes have led to significant equipment simplifications and process improvements for the University of Rochester system.

The OMEGA Laser Facility at the University of Rochester Laboratory for Laser Energetics (UR/LLE) will begin laser illumination campaigns on cryogenic DT ICF targets in the year 2000. The OMEGA Cryogenic Target System (OCTS) will fill plastic ICF targets to high pressure, cool them to cryogenic temperature, layer and characterize targets, then transport them to the center of the OMEGA Target Chamber where they will be illuminated by a 30 kJ laser. The OCTS is being designed and constructed by General Atomics. Although the quantities of tritium being handled are relatively small (

Glass capsules were imploded in direct drive on the OMEGA laser [Boehly et al., Opt. Commun. 133, 495 (1997)] to look for anomalous degradation in deuterium/tritium (DT) yield and changes in reaction history with H3e addition. Such anomalies have previously been reported for D/H3e plasmas but had not yet been investigated for DT/H3e. Anomalies such as these provide fertile ground for furthering our physics understanding of inertial confinement fusion implosions and capsule performance. Anomalous degradation in the compression component of yield was observed, consistent with the “factor of 2” degradation previously reported by Massachusetts Institute of Technology (MIT) at a 50% H3e atom fraction in D2 using plastic capsules [Rygg, Phys. Plasmas 13, 052702 (2006)]. However, clean calculations (i.e., no fuel-shell mixing) predict the shock component of yield quite well, contrary to the result reported by MIT but consistent with Los Alamos National Laboratory results in D2/H3e [Wilson et al., J. Phys.: Conf....