Deuterium-tritium Mixture Research Articles

Detection of explosives and other illicit materials by a single nanosecond neutron pulses — Monte Carlo simulation of the detection process

Recent progress in the development of a Nanosecond Impulse Neutron Investigation System (NINIS) intended for interrogation of hidden objects (explosives and other illicit materials) by means of measuring elastically and non-elastically scattered neutrons is presented. The method uses very bright neutron pulses having durations of the order of few nanoseconds, generated by a dense plasma focus (DPF) devices filled with pure deuterium or a deuterium-tritium mixture as a working gas. A very short duration of the neutron pulse, as well as its high brightness and mono-chromaticity allows using time-of-flight methods with bases of about few meters to distinguish signals from neutrons scattered by different elements.Results of the Monte Carlo simulations of the scattered neutron field from several compounds (explosives and everyday use materials) are presented. The MCNP5 code has been used to get information on the angular and energy distributions of neutrons scattered by the above mentioned compounds assuming the initial neutron energies to be equal to 2.45 MeV (DD) and 14 MeV (DT). A new input has been elaborated that allows modeling not only a spectrum of the neutrons scattered at different angles but also their time history from the moment of generation up to the detection. Such an approach allows getting approximate signals registered by hypothetic scintillator + photomultipler probes placed at various distances from the scattering object, demonstrating principal capability of the method to identify an elemental content of the inspected objects. The extensive computations reveled also several limitations of the proposed method, namely: low number of neutrons reaching detector system, distortions and interferences of scattered neutron signals etc.Further more, preliminary results of the MCNP modeling of the hidden fissile materials detection process are presented.

Development of a technical scale PERMCAT reactor for processing of highly tritiated water

In view of future fusion rectors fueled by deuterium–tritium mixtures, highly tritiated water (HTW) of up to 5.2·1016Bqkg−1 will be produced, during routine operation and scenarios as an accidental release of tritium into a glove box. Also in the solid breeder blanket concept, a non-negligible fraction of the tritium produced will be in the tritiated water fraction. To decontaminate HTW the PERMCAT using isotope swamping in a Pd/Ag membrane reactor has been identified as a robust and reliable solution. In order to investigate the decontamination of HTW at flow rates relevant for future fusion power plants, a technical scale, fully tritium compatible PERMCAT consisting of a bundle of finger-type membranes inserted in a single catalyst bed was developed. Nevertheless, it represents only one part of a PERMCAT cascade necessary to achieve the required performance to process HTW on technical scale. By improving the existing PERMCAT geometry using experimental data obtained from isotopic exchange between D2O and H2, the performance of the existing PERMCAT reactor was optimised. Based on the optimised geometry a new fully tritium compatible technical scale PERMCAT cascade comprising of two PERMCAT reactors in series was designed, manufactured and commissioned as presented in this paper.

Fabrication of ZrO2 coatings on ferritic steel by wet-chemical methods as a tritium permeation barrier

Layers of tetragonal ZrO2 (180 nm) were prepared on the surfaces of ferritic steel specimens by dip coating and electrolytic deposition techniques. These methods were selected because of their applicability to large and complicated structures. Hydrogen permeation tests were carried out at 300–550 °C, and the results were compared with those obtained in the previous study for thinner ZrO2 coating (100 nm). Although the thickness of the present coatings was less than twice that of the previous coatings, the permeation reduction factor for the former was larger by an order of magnitude than that for the latter at/below 400 °C. Tritium retention in the coatings was also measured after exposure to a deuterium–tritium mixture at 300 °C. The obtained results suggest that the permeation rate was determined by transportation through defects in the coatings, and the density of defects penetrating into the coating/substrate interface was significantly reduced by an increase in coating thickness.

Visco-Plastic Flow Predictions of Solidified Deuterium-Tritium Mixtures

Visco-plastic flow properties of hydrogenic solids are important considerations for the design and operation of continuous hydrogenic pellet extrusion systems. Prior to 2010, the visco-plastic flow behavior of deuterium, tritium, and mixtures of the isotopes was only known at 14 K and no heat transfer studies were available. To address these needs, a Cryogenic Couette Viscometer (CCV) was developed at the University of Wisconsin–Madison. Visco-plastic flow characteristics of solid neon, deuterium, and hydrogen were measured using the CCV from the onset of solidification to sub-cooled solid states over a range of shear rates. This paper discusses the transformation of these measurements, using the Quantum Law of Corresponding States, to predict the visco-plastic flow behavior of solid tritium and deuterium-tritium mixtures. Comparisons of predicted values with experimental measurements are made, where available.

Modeling results for mass production layering in a fluidized bed

A fluidized bed has been proposed as a layering device for the mass production of inertial fusion energy fuel pellets. During this layering process, the frozen deuterium or deuterium–tritium mixture filled into a hollow capsule (about 2–4 mm in diameter) is redistributed leading to a fuel layer of uniform thickness on the inside of the fuel capsule. Several physical processes have been identified to interact with each other to influence the outcome of the layering process in a fluidized bed, which needs to fulfill symmetry requirements of the fuel layer thickness, smoothness and surface damage requirements of the outside target surface and must be able to produce a large number of targets (500 000 per day). This work describes the development and use of numerical tools to conduct a trade-off study focusing on the influence of different flow conditions of the fluidizing gas on the total layering time, final layer uniformity and outer surface damage.

Instrumentation and diagnostics for the ITER Neutral Beam System

The ITER Neutral Beam (NB) System consists of a 100 keV diagnostic injector (DNB) and two 1 MeV heating and current drive injectors (HNB). Injection will be done in the different phases of the experimental programme, ranging from low activation operation with hydrogen and helium plasmas, deuterium plasmas, and finally full performance operation with deuterium–tritium mixtures. The injectors will be operated in a harsh environment with high levels of neutron and gamma radiation from the torus. A high stray magnetic field from the tokamak is reduced to acceptable levels for the region up-stream of the residual ion dump (RID) by a combination of a passive magnetic shield and active correction and compensation coils (ACCC). This broad range of requirements imposes a high flexibility, a high availability and reliability of the instrumentation and on the diagnostics that will be used.

Four Years of Experimental Results on the Prototypes of the Cryogenic Holders of the Laser Mégajoule Facility

The Laser Mégajoule cryotarget positioner, PCC, will be used to set cryogenic targets in appropriate conditions for the laser shot in order to reach ignition. These conditions can be summarized as a few parameters of the deuterium-tritium (DT) solid layer: sphericity, roughness, and density of DT gas. The DT mixture is confined and held in a target assembly that is handled and cooled by the PCC. Thus, the parameters of the DT solid layer are controlled by the PCC. In particular, roughness depends on the control of the target base temperature (±1 mK), on the temperature slope while crossing the DT triple point (0.5 mK/min), and on the thermodynamic way followed to reach gas density conditions expected at the laser shot [slow cooling (0.5 mK/min), quenching (several kelvins per second), or rapid cooling (several kelvins per minute)]. Moreover, the required gas density needs high cryogenic power performances of the PCC to be fulfilled. As the target is gripped at cryogenic temperature by the PCC, thermal contact resistance added to power load problems must be faced.We have investigated all these cryogenic challenges on DEMOCRYTE, the prototype of the cryogenic holder setup. Experimental results obtained between 2006 and 2009 are described in this paper.

Germanium Doped CHx Microshells for LMJ Targets

At the CEA Laser “Megajoule” facility, amorphous hydrogenated carbon (a-C:H or CHx) is the nominal ablator used to achieve inertial confinement fusion experiments. These targets are filled with a fusible mixture of deuterium-tritium in order to perform ignition.Since the achievement of ignition greatly depends on the physical properties of the shell, there must be precise control of thicknesses, doping concentration, and roughness. Experimental devices associated with suitable characterizations are described in this paper. The tolerances and yields for each specification are also presented. Some specifications are largely reached; high-frequency surface roughness due to isolated surface defects appears to be the main yield-limiting factor. A microscopic approach of stress thin film measurement is described to examine oxygen uptake in CHx film.

Multi-dimensional computational model of the movement of the solid–gas interface during the layering process in inertial confinement fusion targets in a non-uniform thermal environment

The redistribution of deuterium (DD) or of a deuterium–tritium mixture (DT) to form a layer on the inside of spherical inertial confinement fusion (ICF) capsules is a challenging problem because of the symmetry requirements of the fuel layer thickness, the smoothness requirement of the inside target surface, and the time restriction on the production process. Heat- and mass-transfer processes have been identified to interact with one another to influence the outcome of the layering process. For example, the mass redistribution speed of the fuel inside the shell towards a uniform layer and the final layer thickness uniformity depend on the variation in local heat transfer coefficient along the outer target surface. The focus of this work was to develop a numerical tool to help understand the physics involved in the layering process to be able to assess the influence of key parameters on the transient layer formation. The coupled mass and heat transfer processes governing target layering have been studied numerically, implementing unique boundary conditions to track the movement of the gas–solid boundary on the inside of the shell. The model was validated through comparison with theoretical results and laboratory-scale experiments. With this model, a window of parameters can by identified, under which layering experiments are likely to be successful.

Viscosity and mutual diffusion of deuterium-tritium mixtures in the warm-dense-matter regime

We have calculated viscosity and mutual diffusion of deuterium-tritium (DT) in the warm, dense matter regime for densities from 5 to 20 g/cm{3} and temperatures from 2 to 10 eV, using both finite-temperature Kohn-Sham density-functional theory molecular dynamics (QMD) and orbital-free molecular dynamics (OFMD). The OFMD simulations are in generally good agreement with the benchmark QMD results, and we conclude that the simpler OFMD method can be used with confidence in this regime. For low temperatures (3 eV and below), one-component plasma (OCP) model simulations for diffusion agree with the QMD and OFMD calculations, but deviate by 30% at 10 eV. In comparison with the QMD and OFMD results, the OCP viscosities are not as good as for diffusion, especially for 5 g/cm{3} where the temperature dependence is significantly different. The QMD and OFMD reduced diffusion and viscosity coefficients are found to depend largely, though not completely, only on the Coulomb coupling parameter Γ , with a minimum in the reduced viscosity at Γ≈25 , approximately the same position found in the OCP simulations. The QMD and OFMD equations of state (pressure) are also compared with the hydrogen two-component plasma model.

Counter-current isotope swamping in a membrane reactor: The PERMCAT process and its applications in fusion technology

A PERMCAT reactor is a catalytic membrane reactor that combines a Pd/Ag membrane and a catalyst bed. It has been developed to ensure very high tritium recovery from the unspent fuel of fusion machines using deuterium tritium mixtures. The PERMCAT process takes advantage of simultaneously unlocking chemically bound tritium via heterogeneously catalysed isotope exchange reactions and removing tritium via its selective permeation through the membrane. The PERMCAT reactor operated in the counter-current isotope swamping mode allows a very low tritium activity at the outlet of the component to be maintained. Two main issues have been solved to achieve efficient and reliable PERMCAT operation. Firstly, the mechanical design has to cope with the elongation and deformation of the membrane resulting from thermal expansion and lattice parameter increase under operation with hydrogen. Secondly, the catalyst material has to be chosen in order to promote isotope exchange reactions while minimising the numerous side reactions that occur especially when the mixture contains carbon oxides. This paper presents a general overview of the R&D performed at the Tritium Laboratory Karlsruhe for PERMCAT technology. Technical solutions to solve both issues together with relevant experimental results including processing tests with tritium are discussed.

Reaction-in-flight neutrons as a signature for shell mixing in National Ignition Facility capsules

Analytic calculations and results from computational simulations are presented that suggest that reaction-in-flight (RIF) neutrons can be used to diagnose mixing of the ablator shell material into the fuel in deuterium-tritium (DT) capsules designed for the National Ignition Facility (NIF) [J. A. Paisner, J. D. Boyes, S. A. Kumpan, W. H. Lowdermilk, and M. S. Sorem, Laser Focus World 30, 75 (1994)]. Such mixing processes in NIF capsules are of fundamental physical interest and can have important effects on capsule performance, quenching the total thermonuclear yield. The sensitivity of RIF neutrons to hydrodynamical mixing arises through the dependence of RIF production on charged-particle stopping lengths in the mixture of DT fuel and ablator material. Since the stopping power in the plasma is a sensitive function of the electron temperature and density, it is also sensitive to mix. RIF production scales approximately inversely with the degree of mixing taking place, and the ratio of RIF to down-scattered neutrons provides a measure of the mix fraction and/or the mixing length. For sufficiently high-yield capsules, where spatially resolved RIF images may be possible, neutron imaging could be used to map RIF images into detailed mix images.

Recent advances in development of materials for laser target

AbstractNearly ten years ago, a research program concerning materials and technologies was defined to develop the very complex cryogenic target for obtaining the combustion of a deuterium-tritium mixture by inertial confinement fusion in the laser megajoule facility. The CEA target fabrication project includes research and development on various organic polymers and materials for the cryogenic laser megajoule target assembly as well as for other targets that can be useful for the fusion program (pre and post-ignition). Recent advances have been accomplished concerning the development of specific organic materials for the fabrication of targets components, including the synthesis of polymers for the laser megajoule microshells and metal-doped organic foams for the elaboration of doped-foam microshells or for the micromachining of components.

Thickness, Doping Accuracy, and Roughness Control in Graded Germanium Doped CHX Microshells for LMJ

In the Commissariat à l’Energie Atomique Laser Megajoule (LMJ) facility, amorphous hydrogenated carbon (a-C:H or CHX) is the nominal ablator used to achieve inertial confinement fusion experiments. These targets are filled with a fusible mixture of deuterium-tritium in order to perform ignition. The a-C:H shell is deposited on a polyalphamethylstyrene (PAMS) mandrel by glow discharge polymerization with trans-2-butene, hydrogen, and helium. Graded germanium doped CHX microshells are supposed to be more stable regarding hydrodynamic instabilities. The shells are composed of four layers, for a total thickness of 180 μm. The germanium gradient is obtained by doping the different a-C:H layers with the addition of tetramethylgermanium in the gas mixture.As the achievement of ignition greatly depends on the physical properties of the shell, the thicknesses, doping concentration, and roughness must be precisely controlled.Quartz microbalances were used to perform an in situ and real-time measurement of the thickness in order to reduce the variations - and so our fabrication tolerances - on each layer thickness. Ex situ control of the thickness of each layer was carried out, with both optical coherent tomography and interferometry (wallmapper).High-quality PAMS and a rolling system have been used to lower the low-mode roughness [root-mean-square (rms) (mode 2) < 70 nm]. High modes were clearly reduced by coating the pan containing the shells with polyvinyl alcohol + CHX instead of polystyrene + CHX resulting in an rms (>mode 10) < 20 nm, which can be <15 nm for the best microshells.The germanium concentration (0.4 and 0.75 at.%) in the a-CH layer is obtained by regulating the tetramethylgermanium flow. Low range mass flow controllers have been used to improve the doping accuracy.

Fusion in a Staged Z-pinch

A Staged Z-pinch (H.U. Rahman, F.J. Wessel, N. Rostoker, Phys. Rev. Lett. 74:714, 1995), configured for a 100 ns, 2 MJ implosion accelerator, is studied using the 2-1/2 D, radiation-MHD code, MACH2. The Z-pinch is configured as a cylindrical, high-atomic number plasma shell that implodes radially onto a co-axial, plasma target, for example: Xenon onto a 50:50 mixture of Deuterium-Tritium. During implosion a shock develops in the plasma liner, producing a conduction channel at the Xe/DT interface as the mass Xe accumulates, and preheating the DT target. During subsequent acceleration and compression the Xe/DT interface remains stable, even as the outer surface of the Xe shell develops RT instabilities. At peak implosion the simulated fusion-energy yield is 7.6 MJ, producing an energy gain of 3.8.

Diagnosing ablator ρR and ρR asymmetries in capsule implosions using charged-particle spectrometry at the National Ignition Facility

By fielding several compact proton spectrometers at various locations around an ignition-capsule implosion at the National Ignition Facility [G. H. Miller, E. I. Moses, and C. R. Wuest, Nucl. Fusion 44, S228 (2004)], ρR and ρR asymmetries of the ablator for a failed implosion can be obtained through absolute measurements of knock-on proton (KO-P) spectra. For ignition capsules with a Cu-doped beryllium (Be) ablator, 50:50 mixture of deuterium-tritium (DT) fuel and ∼1% residual hydrogen (H) by atom, failed implosions can be diagnosed for neutron yields ranging from ∼1011 to ∼6×1015 and local ρRs up to ∼240mg∕cm2. For capsules with an ablator of Ge-doped CH, which contains a large amounts of H, failed implosions can be diagnosed for neutron yields ranging from ∼1010 to ∼6×1015 and local ρRs up to ∼200mg∕cm2. Prior to the first ignition experiments, capsules with a Cu-doped Be ablator (or Ge-doped CH ablator), more deuterium-lean fuel mixture and H-dopant levels up to 25% in the fuel will be imploded to prim...

Overview of the tritium system of Ignitor

Among the recent design activities of the Ignitor program, the analysis of the tritium system has been carried out with the aim to describe the main equipments and the operations needed for supplying the deuterium–tritium mixtures and recovering the plasma exhaust. In fact, the tritium system of Ignitor provides for injecting deuterium–tritium mixtures into the vacuum chamber in order to sustain the fusion reaction: furthermore, it generally manages and controls the tritium and the tritiated materials of the machine fuel cycle. Main functions consist of tritium storage and delivery, tritium injection, tritium recovery from plasma exhaust, treatment of the tritiated wastes, detritiation of the contaminated atmospheres, tritium analysis and accountability. In this work an analysis of the designed tritium system of Ignitor is summarized.

Preparation and Liner Compression of Plasma From an Ultrahigh Speed Flow

Preparation of the target plasma represents a critical issue in liner compression techniques to achieve fusion conditions. We consider the use of an ultrahigh speed plasma flow from a special coaxial-gun arrangement known as the plasma flow switch. Experiments have demonstrated that this arrangement can provide plasma flows with speeds in excess of 2000 km/s. Stagnation of such a plasma flow results in fully stripped aluminum plasma with electron temperatures of 30 keV. Substitution of deuterium or a deuterium-tritium mixture could provide target plasma at kilovolt temperatures within an imploding liner. Such temperatures suggest that, even if substantial heat loss occurred during liner compression, fusion-level temperatures would be possible. The concatenation of events to generate the ultrahigh speed flow, to direct it into the implosion chamber, and to arrange liner dynamics for effective compression demands numerical simulation, which is based on initial analytical estimates. Both types of calculation for exploring this concept are discussed.

Neutron Radiation Shielding for the NIF Streaked X-Ray Detector (SXD) Diagnostic

The National Ignition Facility (NIF) at Lawrence Livermore National Laboratory (LLNL) is preparing for the National Ignition Campaign (NIC) scheduled in 2010. The NIC is comprised of several “tuning” physics sub-campaigns leading up to a demonstration of Inertial Confinement Fusion (ICF) ignition. Some of these experiments requires to use the NIF streak x-ray detector (SXD) to measure fuel capsule trajectory (shock timing) or x-ray “bang-time” from time-resolved x-ray imaging of the imploding capsule fuelled with pure tritium (T) instead of a deuterium-tritium (DT) mixture. The resulting prompt neutron fluence at the planned SXD location (∼1.7 m from the target) would be ∼ 1.4e9/cm2. Previous measurements suggest the onset of significant background at a neutron fluence of ∼ 1e8/cm2 and the radiation damage and operational upsets which start at ∼ 1e8 rad-Si/sec must be factored into an integrated experimental campaign plan. Monte Carlo analyses were performed to predict the neutron and gamma/x-ray fluences and radiation doses for the proposed diagnostic configuration. A possible shielding configuration is proposed to mitigate radiation effects. The primary component of this shielding is an 80 cm thickness of Polyethylene (PE) between target chamber center (TCC) and the SXD diagnostic. Additionally, 6-8 cm of PE around the detector reduces the large number of neutrons that scatter off the inside of the target chamber. This proposed shielding configuration reduces the high-energy neutron fluence at the SXD by approximately a factor of ∼50.

Cryogenic Target Handling, Transfer and Positioning System for the Laser Mégajoule

The Laser MegaJoule (LMJ) program plans to obtain Deuterium-Tritium (DT) mixture ignition leading to a fusion gain of ten. Cryogenic targets are hollow spheres whose interior is covered with a solid cryogenic fuel layer. The success of DT ignition depends on quality of the fuel layer uniformity. These targets must be cooled and kept at temperatures near the triple point (19.8 K) with a very good stability (+/-1 mK) for many hours, in the center of the 5 m radius experimental vacuum chamber with a position accuracy of a few microns. In order to validate our current device concepts, we have manufactured scale one prototypes to confirm all thermal and mechanical challenges, such as sharp thermal regulation, cooling autonomy and cryogenic target transfer.