DT Gas Research Articles

Fuel-shell interface instability growth effects on the performance of room temperature direct-drive implosions

Performance degradation in direct-drive inertial confinement fusion implosions is caused by several effects, one of which is Rayleigh–Taylor (RT) instability growth during the deceleration phase. In room-temperature plastic target implosions, deceleration-phase RT growth is enhanced by the density discontinuity and finite Atwood number at the fuel–shell interface. In this paper, the Atwood number of the interface is systematically varied by altering the ratio of deuterium to tritium (D:T) within the DT gas fill. It is shown that the stability of the interface is best characterized by the effective Atwood number, which is primarily determined by radiation heating of the shell and not by the composition of the fuel. Both simulation and experimental data show that yield performance scales with the fraction of D and T present in the fuel and that the observed inferred ion temperature asymmetry (ΔTi=Timax−Timin), which indicates the presence of long-wavelength modes, has a small sensitivity to the different D:T ratios.

Partitioning of tritium between surface and bulk of 316 stainless steel at room temperature

The distribution of tritium between the near surface and the bulk of 316 stainless steel has been measured using two independent techniques: pulsed-plasma exposures and a zinc-chloride wash. Between 17% and 20% of the total inventory absorbed into a stainless-steel sample during a 24-h exposure to DT gas at room temperature resides in the water layers present on the metal surface. Redistribution of tritium between the surface and the bulk of stainless steel, if it occurs, is very slow. Tritium does not appear to enter into the bulk at a rate defined solely by lattice diffusivity.

Development of H, D, T Simultaneous TDS Measurement System and H, D, T Retention Behavior for DT Gas Exposed Tungsten Installed in LHD Plasma Campaign

All the hydrogen isotope (H, D, T) simultaneous TDS (Thermal desorption spectroscopy) measurement system (HI-TDS system) was newly designed to evaluate all hydrogen isotope desorption behavior in materials. The present HI-TDS system was operated under Ar purge gas and the H and D desorptions were observed by a quadruple mass spectrometer equipped with an enclosed ion source, although T desorption was evaluated by an ionization chamber or proportional counters. Most of the same TDS spectra for D and T were derived by optimizing the heating rate of 0.5 K s−1 with Ar flow rate of 13.3 sccm.Using this HI-TDS system, D and T desorption behaviors for implanted or DT gas exposed tungsten samples installed in LHD (Large Helical Device) at NIFS (National Institute for Fusion Science) was evaluated. It was found that major hydrogen desorption stages consisted of two temperature regions, namely 700 K and 900 K, which was consistent with the previous hydrogen plasma campaign and most of hydrogen would be trapped by the carbon-dominated mixed-material layer. By implantation, major D desorption was found at ~900 K with a narrow peak due to energetic ion implantation. For gas exposure, H was preferentially replaced by D and T with a lower trapping energy. In addition, T replacement rate by additional H2 gas exposure was evaluated. This fact indicates that the hydrogen replacement mechanism would be clearly changed by exposure methods.

Effect of spatial nonuniformity of heating on compression and burning of a thermonuclear target under direct multibeam irradiation by a megajoule laser pulse

Direct-drive fusion targets are considered at present as an alternative to targets of indirect compression at a laser energy level of about 2 MJ. In this approach, the symmetry of compression and ignition of thermonuclear fuel play the major role. We report on the results of theoretical investigation of compression and burning of spherical direct-drive targets in the conditions of spatial nonuniformity of heating associated with a shift of the target from the beam center of focusing and possible laser radiation energy disbalance in the beams. The investigation involves numerous calculations based on a complex of 1D and 2D codes RAPID, SEND (for determining the target illumination and the dynamics of absorption), DIANA, and NUT (1D and multidimensional hydrodynamics of compression and burning of targets). The target under investigation had the form of a two-layer shell (ablator made of inertial material CH and DT ice) filled with DT gas. We have determined the range of admissible variation of compression and combustion parameters of the target depending on the variation of the spatial nonuniformity of its heating by a multibeam laser system. It has been shown that low-mode (long-wavelength) perturbations deteriorate the characteristics of the central region due to less effective conversion of the kinetic energy of the target shell into the internal energy of the center. Local initiation of burning is also observed in off-center regions of the target in the case of substantial asymmetry of irradiation. In this case, burning is not spread over the entire volume of the DT fuel as a rule, which considerably reduces the thermonuclear yield as compared to that in the case of spherical symmetry and central ignition.

Tritium Retention in Reduced-Activation Ferritic/Martensitic Steels

The effects of displacement damage, plasma exposure and heat loads on T retention in reduced-activation ferritic/martensitic (RAFM) steels were investigated by exposing the steels to DT gas at 473 K. Despite enormous change in surface morphology, T retention in the heat-loaded specimen was comparable with that in the unloaded specimen. The exposure to plasma resulted in a drastic increase in T retention at the surface and/or sub surface. However, the T trapped at the surface/subsurface was easily removed by maintaining the specimens in air at ~300 K. Formation of radiation-induced defects led to a significant increase in T retention, and T trapped in the defects was not removed at ~300 K. These observations suggest that displacement damages have the largest effects on T retention at ~473 K.

Simulations of indirectly driven gas-filled capsules at the National Ignition Facility

Gas-filled capsules imploded with indirect drive on the National Ignition Facility have been employed as symmetry surrogates for cryogenic-layered ignition capsules and to explore interfacial mix. Plastic capsules containing deuterated layers and filled with tritium gas provide a direct measure of mix of ablator into the gas fuel. Other plastic capsules have employed DT or D3He gas fill. We present the results of two-dimensional simulations of gas-filled capsule implosions with known degradation sources represented as in modeling of inertial confinement fusion ignition designs; these are time-dependent drive asymmetry, the capsule support tent, roughness at material interfaces, and prescribed gas-ablator interface mix. Unlike the case of cryogenic-layered implosions, many observables of gas-filled implosions are in reasonable agreement with predictions of these simulations. Yields of TT and DT neutrons as well as other x-ray and nuclear diagnostics are matched for CD-layered implosions. Yields of DT-filled capsules are over-predicted by factors of 1.4–2, while D3He capsule yields are matched, as well as other metrics for both capsule types.

Studying Validity of Single-Fluid Model in Inertial Confinement Fusion

The validity of single-fluid model in inertial confinement fusion simulations is studied by comparing the results of the multi- and single-fluid models. The multi-fluid model includes the effects of collision and interpenetration between fluid species. By simulating the collision of fluid species, steady-state shock propagation into the thin DT gas and expansion of hohlraum Au wall heated by lasers, the results show that the validity of single-fluid model is strongly dependent on the ratio of the characteristic length of the simulated system to the particle mean free path. When the characteristic length L is one order larger than the mean free path λ, the single-fluid model's results are found to be in good agreement with the multi-fluid model's simulations, and the modeling of single-fluid remains valid. If the value of L/λ is lower than 10, the interpenetration between fluid species is significant, and the single-fluid simulations show some unphysical results; while the multi-fluid model can describe well the interpenetration and mix phenomena, and give more reasonable results.

Real viscosity effects in inertial confinement fusion target deuterium–tritium micro-implosions

We report on numerical studies of real viscous effects on the implosion characteristics of imploded DT micro-targets. We use the implicit ePLAS code to perform 2D simulations of spherical and slightly ellipsoidal DT shells on DT gas filled ∼40 μm diameter voids. Before their final implosions the shells have been nearly adiabatically compressed up to 102 or 103 g/cm3 densities. While the use of conventional artificial viscosity can lead to high central densities for initially spherical shells, we find that a real physical viscosity from ion-ion collisions can give a high (>20 keV) central temperature but severely reduced central density (<200 g/cm3), while the elliptical shells evidence p = 2 distortion of the heated central fuel region. These results suggest that the general use of artificial viscosities in Inertial Confinement Fusion (ICF) modeling may have lead to overly optimistic yields for current NIF targets and that polar direct drive with more energy for the imploding capsule may be needed for ultimate ICF success.

Progress toward ignition at the National Ignition Facility

Progress toward ignition at the National Ignition Facility (NIF) has been focused on furthering the understanding of implosion performance. Implosion performance depends on the capsule fuel shape, on higher mode asymmetries that may cause hydrodynamic instabilities to quench ignition, on time-dependent asymmetries introduced by the hohlraum target, and on ablator performance. Significant findings in each of these four areas is reported. These investigations have led to improved in-flight capsule shape, a demonstration that a capsule robust to mix can generate high levels of neutrons (7.7 × 1014), hohlraum modifications that should ultimately provide improved beam propagation and better laser coupling, and fielding of capsules with high-density carbon (HDC) ablators. A capsule just fielded with a HDC ablator and filled with DT gas generated a preliminary record level of neutrons at 1.6 × 1015, or 5 kJ of energy. Future plans include further improvements to fuel shape and hohlraum performance, fielding robust capsules at higher laser power and energy, and tuning the HDC capsule. A capsule with a nanocrystalline diamond (HDC) ablator on a DT ice layer will be fielded at NIF later this year.

Initial magnetic field compression studies using gas-puff Z-pinches and thin liners on COBRA

This magnetic compression of cylindrical liners filled with DT gas has promise as an efficient way to achieve fusion burn using pulsed-power machines. However, to avoid rapid cooling of the fuel by transfer of heat to the liner an axial magnetic field is required. This field has to be compressed during the implosion since the thermal insulation is more demanding as the compressed DT plasma becomes hotter and its volume smaller. This compression of the magnetic field is driven both by the imploding liner and plasma. To highlight how this magnetic field compression by the plasma and liner evolves we have separately studied Z-pinch implosions generated by gas puff and liner loads. The masses of the gas puff and liner loads were adjusted to match COBRA's current rise times. Our results have shown that Ne gas-puff implosions are well described by a snowplow model where electrical currents are predominately localized to the outer surface of the imploding plasma and the magnetic field is external to the imploding plasma. Liner implosions are dominated by the plasma ablation process on the inside surface of the liner and the electrical currents and magnetic fields are advected into the inner plasma volume; the sharp radial gradient associated with the snowplow process is not present.

Analytical estimation of neutron yield in a micro gas-puff X pinch

In this paper, we present the basic concepts for developing a micro x pinch as a small-scale neutron source. For compact sources, these concepts offer repetitive function at higher yields and pulsing rates than competing methods. The uniqueness of these concepts arises from the use of microelectronic technology to reduce the size of the target plasma and to efficiently heat the target gas. The use of repetitive microelectromechanical systems (MEMs) gas puff technology, as compared to cryogenic wires or solid targets (for the beam-target alternatives), has the potential to be robust and have a long lifetime because the plasma is not created from solid surfaces. The modeling suggests that a 50 J at the wall plug pulse could provide &amp;gt;105 tritium (DT) neutrons and 103 deuterium (DD) neutrons at temperatures of a few keV. At 1 kHz, this would be &amp;gt;108 and 106 neutrons per second, DT and DD, respectively, with a 250 μm anode-cathode gap. DT gas puff devices may provide &amp;gt;1012 neutrons/s operating at 1 kHz and requiring 100 kW. The MEMs approach offers potentially high pulse rates and yields.

Knudsen Layer Reduction of Fusion Reactivity

Knudsen layer losses of tail fuel ions can significantly reduce the fusion reactivity of multi-keV DT in capsules with small fuel ρr; sizable yield reduction can result for small inertial confinement fusion (ICF) capsules. This effect is most pronounced when the distance from a burning DT gas region to a nonreacting or cold wall is comparable to the mean free path of reacting fuel ions. A simplified asymptotic theory of Knudsen layer tail depletion is presented and a nonlocal reduced fusion reactivity model is obtained. Application of the model in simulations of ICF capsule implosion experiments gives calculated yields and ion temperatures that are in much closer agreement with observations than are the results of "nominal" or mixed simulations omitting the model.

The Radiochemical Analysis of Gaseous Samples (RAGS) apparatus for nuclear diagnostics at the National Ignition Facility (invited)

The Radiochemical Analysis of Gaseous Samples (RAGS) diagnostic apparatus was recently installed at the National Ignition Facility (NIF). Following a NIF shot, RAGS is used to pump the gas load from the NIF chamber for purification and isolation of the noble gases. After collection, the activated gaseous species are counted via gamma spectroscopy for measurement of the capsule areal density and fuel-ablator mix. Collection efficiency was determined by injecting a known amount of (135)Xe into the NIF chamber, which was then collected with RAGS. Commissioning was performed with an exploding pusher capsule filled with isotopically enriched (124)Xe and (126)Xe added to the DT gas fill. Activated xenon species were recovered post-shot and counted via gamma spectroscopy. Results from the collection and commissioning tests are presented. The performance of RAGS allows us to establish a noble gas collection method for measurement of noble gas species produced via neutron and charged particle reactions in a NIF capsule.

Drive Asymmetry and the Origin of Turbulence in an ICF Implosion

2D and 3D numerical simulations with the adaptive mesh refinement Eulerian radiation-hydrocode RAGE at unprecedented spatial resolution are used to investigate the connection between drive asymmetry and the generation of turbulence in the DT fuel in a simplified inertial-confinement fusion (ICF) implosion. Long-wavelength deviations from spherical symmetry in the pressure drive lead to the generation of coherent vortical structures in the DT gas and it is the three-dimensional instability of these structures that in turn leads to turbulence and mix. The simulations suggest that this mechanism may be an additional important source of mix in ICF implosions. Applications to target ignition at the National Ignition Facility are briefly discussed.

Analysis of the conditions of experiments on the compression of capsules with a foam absorber at Iskra V facility

The compression conditions are analyzed and the capsules are discussed for the energy of ∼2 kJ and pulse duration of ∼0.5 ns at the second harmonic of an iodine laser (λ = 0.66 μm) under irradiation by a small number of beams at Iskra V facility (the facility has 12 beams, which are focused in a certain way onto the capsule). The authors consider different capsules filled with DT gas: the glass capsules (the inner radius is 260 μm; the thickness, 1.2-1.3 μm; the aspect ratio A = 200) experimentally examined with Iskra V; the capsules made of polystyrene (R = 140-170 μm; δ = 5-10 μm; A = 14-34), and the capsules made of polystyrene and covered by a foam absorber (R = 140-170 μm; δ = 5-10 μm; A = 14-34, foam absorber δ = 150 μm) - ‘Laser Greenhouse’ [4].

Studies in capsule design for mid-Z ion-driven fast ignition

Ion fast ignition (FI) is an approach to fast ignition inertial confinement fusion (ICF) where an energetic ion beam is used to heat a hot spot in a compressed ICF capsule to ignition conditions. Recent work at LANL and elsewhere suggests that ion beams with Z > 1 may be produced with characteristics suitable for FI. Indeed, fast ignition using mid-Z ions may have advantages over protons FI in terms of number of ions required for ignition, robustness of the ion beam to beam-plasma instabilities, sharpness of Bragg peak, and ability to locate the ion source far from the ignition capsule. In this paper, simple preliminary capsule designs for mid-Z ion fast ignition are assessed. These designs comprise a DT gas pocket, a DT ice fuel layer, and a low-Z ablator subject to a drive achievable on facilities such as the NIF and LMJ. Dependence of capsule gain with various beam parameters are obtained.

Multiframe detection of X-rays with a 100-ps frame duration on the Iskra-5 facility

Three techniques for multiframe detection of X rays from laser targets of different types in experiments on the Iskra-5 facility are described. The frame duration is 100 ps, the number of frames is ten, and the spatial resolution is up to 10 μm. The developed techniques allow determination of the degree of uniformity of the emission from the end of a cylindrical target; the heating time of foils made from different materials; the collapse time of spherical glass targets filled with the DT gas; and the shape, size, and time of existence of a compressed core.

Calculations of a Fill Hole/Tube in an Ignition Capsule

Filling an ignition capsule through a drilled hole in the ablator is the current approach to fielding an ignition capsule. But it adds an initial defect to the capsule which might grow large enough to affect or even prevent ignition. We present here calculations of the effects of fill tubes and holes for the 1.4 MJ 300 eV BeCu NIF capsule. The code used is the 3D AMR code written by Los Alamos and SAIC, “RAGE”. Several fill tube/hole sizes were tried. Most calculations were made in a planar 2D geometry, providing reliable information on the first part of the implosion before convergence effects become important. A 5 μm diameter hole generates a 25 by 30 μm jet when the main shock breaks out into the DT gas. The mass involved in the jet is insignificant, less than 1/1000 of the hot spot mass. There is no large difference between the jets formed by a plug and a fill tube, before they break out into DT gas. High resolution spherical calculations are still in progress to understand the end of the implosion. Experiments are planned as a support to this study.

Tritium Operations at the Laboratory for Laser Energetics

The Laboratory for Laser Energetics (LLE) at the University of Rochester has conducted inertial confinement fusion experiments since the early 1970s. Beginning in 1996, LLE filled and fielded targets containing DT gas with pressures as high as 30 atm. Facilities are being upgraded to prepare, characterize, and field targets with DT ice on their inner surface. To this end, process loops that can pressurize DT gas to 1200 bar and operate at 17 K are in the final stages of commissioning. To preclude both accidental and chronic tritium releases and to minimize the potential for exposures to personnel, both metal hydride-based and oxidation drier-based cleanup systems have been installed and commissioned with hydrogen. Cryogenic DT targets will be fielded in 2006.

Technical and Scientific Aspects of the JET Trace-Tritium Experimental Campaign

The JET Trace Tritium (TTE) programme marked the first use of tritium in experiments under the managerial control of UKAEA, which operates the JET Facility on behalf of EFDA. The introduction of tritium into the plasma by gas fuelling and neutral beam injection, even in trace quantities, required the mobilisation of gram-quantities of tritium gas from the Active Gas Handling System (AGHS) product storage units into the supply lines connected to the torus gas valve and the neutral beam injectors. All systems for DT gas handling, recovery and reprocessing were therefore recommissioned and operating procedures re-established, involving extensive operations staff training. The validation of Key Safety Related Equipment (KSRE) is described with reference to specific examples. The differences between requirements for TTE and full DT operations are shown to be relatively small. The scientific motivation for TTE, such as the possibility to obtain high-quality measurements in key areas such as fuel-ion transport and fast ion dynamics, is described, and the re-establishment and development of JET’s 14MeV neutron diagnostic capability for TTE and future DT campaigns are outlined. Some scientific highlights from the TTE campaign are presented.