Liquid lead and Lithium. A Comprehensive molecular dynamics study

Pb17Li is today a reference breeder material in diverse fusion R&D programs worldwide. One of the main issues in these programs is the problem of liquid metals breeder blanket behavior. Structural material of the blanket should meet high requirements because of extreme operating conditions. Therefore the knowledge of eutectic properties like optimal composition, physical and thermodynamic behavior or diffusion coefficients of Tritium are extremely necessary for current designs. In particular, the knowledge of the function linking the tritium concentration dissolved in liquid materials with the tritium partial pressure at a liquid/gas interface in equilibrium, CT=f(PT), is of basic importance because it directly impacts all functional properties of a blanket determining: tritium inventory, tritium permeation rate and tritium extraction efficiency. Nowadays, understanding the structure and behavior of this compound is a real goal in fusion engineering and materials science. Simulations of liquids can provide much information to the community; not only supplementing experimental data, but providing new tests of theories and ideas, making specific predictions that require experimental tests, and ultimately helping to lead to the deeper understanding and better predictive behavior.

New simulations to qualify eutectic lithium-lead as breeder material

Pb17Li is today a reference breeder material in diverse fusion R&D programs worldwide. One of the main issues is the problem of liquid metals breeder blanket behavior. The knowledge of eutectic properties like optimal composition, physical and thermodynamic behavior or diffusion coefficients of Tritium are extremely necessary for current designs. In particular, the knowledge of the function linking the tritium concentration dissolved in liquid materials with the tritium partial pressure at a liquid/gas interface in equilibrium, CT =f(PT ), is of basic importance because it directly impacts all functional properties of a blanket determining: tritium inventory, tritium permeation rate and tritium extraction efficiency. Nowadays, understanding the structure and behavior of this compound is a real goal in fusion engineering and materials science. Atomistic simulations of liquids can provide much information; not only supplementing experimental data, but providing new tests of theories and ideas, making specific predictions that require experimental tests, and ultimately helping to a deeper understanding

Ion Beam Analysis of He-implanted fusion solid breedes

Introduction Lithium-based ceramics (silicates, titanates, ?) possess a series of advantages as alternative over liquid lithium and lithium-lead alloys for fusion breeders. They have a sufficient lithium atomic density (up to 540 kg*m-3), high temperature stability (up to 1300 K), and good chemical compatibility with structural materials. Nevertheless, few research is made on the diffusion behavior of He and H isotopes through polycrystalline structures of porous ceramics which is crucial in order to understand the mobility of gas coolants as well as, the release of tritium. Moreover, in the operating conditions of actual breeder blanket concepts, the extraction rate of the helium produced during lithium transmutation can be affected by the composition and the structure of the near surface region modifying the performance of BB materials

Thermal diffusion profile by NRA of Deuterium implanted in solid breeder blanket materials

This paper studies the relationship between aging, physical changes and the results of non-destructive testing of plywood. 176 pieces of plywood were tested to analyze their actual and estimated density using non-destructive methods (screw withdrawal force and ultrasound wave velocity) during a laboratory aging test. From the results of statistical analysis it can be concluded that there is a strong relationship between the non-destructive measurements carried out, and the decline in the physical properties of the panels due to aging. The authors propose several models to estimate board density. The best results are obtained with ultrasound. A reliable prediction of the degree of deterioration (aging) of board is presented. Breeder blanket materials have to produce tritium from lithium while fulfilling several strict conditions. In particular, when dealing with materials to be applied in fusion reactors, one of the key questions is the study of light ions retention, which can be produced by transmutation reactions and/or introduced by interaction with the plasma. In ceramic breeders the understanding of the hydrogen isotopes behaviour and specially the diffusion of tritium to the surface is crucial. Moreover the evolution of the microstructure during irradiation with energetic ions, neutrons and electrons is complex because of the interaction of a high number of processes.

Systematic Analysis of Direct-Drive Baseline Designs for Shock-Ignition with the Laser Megajoule

Direct-drive inertial confinement thermonuclear fusion consists in illuminating a shell of cryogenic Deuterium and Tritium (DT) mixture with many intense beams of laser light. Capsule is composed of DT gassurrounded by cryogenic DT as combustible fuel. Basic rules are used to define shell geometry from aspect ratio, fuel mass and layers densities. We define baseline designs using two aspect ratio (A=3 and A=5) who complete HiPER baseline design (A=7.7). Aspect ratio is defined as the ratio of ice DT shell inner radius over DT shell thickness. Low aspect ratio improves hydrodynamics stabilities of imploding shell. Laser impulsion shape and ablator thickness are initially defined by using Lindl (1995) pressure ablation and mass ablation formulae for direct-drive using CH layer as ablator. In flight adiabat parameter is close to one during implosion. Velocitie simplosions chosen are between 260 km/s and 365 km/s. More than thousand calculations are realized for each aspect ratio in order to optimize the laser pulse shape. Calculations are performed using the one-dimensional version of the Lagrangian radiation hydrodynamics FCI2. We choose implosion velocities for each initial aspect ratio, and we compute scaled-target family curves for each one to find self-ignition threshold. Then, we pick points on each curves that potentially product high thermonuclear gain and compute shock ignition in the context of Laser MegaJoule. This systematic analyze reveals many working points which complete previous studies ´allowing to highlight baseline designs, according to laser intensity and energy, combustible mass and initial aspect ratio to be relevant for Laser MegaJoule.

Symmetry Issues in Directly Irradiated Targets

In direct drive Inertial Confinement Fusion (ICF), the typical laser beam to laser beam angle is around 30o. This fact makes the study of the irradiation symmetry agenuine 3D problem. In this paper we use the three dimensional version of the MULTI hydrocode to assess the symmetry of such ICF implosions. More specifically, we study a shock-ignition proposal for the Laser-M´egajoule facility (LMJ) in which two of the equatorial beam cones are used to implode and pre compress a spherical capsule (the “reference” capsule of HiPER project) made of 0.59 mg of pure Deuterium-Tritium mixture. The symmetry of this scheme is analysed and optimized to get a design inside the operating limits of LMJ. The studied configuration has been found essentially axial-symmetric, so that the use of 2D hydrocodes would be appropriate for this specific situation.

Differences into HT and HTO concentrations in air into the Western Mediterranean Basin and Continental Europe and Safety Related Issues.

Real time Tritium concentrations in air in two chemical forms, HT and HTO, coming from an ITER-like fusion reactor as source were coupled the European Centre Medium Range Weather Forecast (ECMWF) numerical model with the Lagrangian Atmospheric-particle dispersion model FLEXPART. This tool was analyzed in nominal tritium discharge operational reference and selected incidental conditions affecting the Western Mediterranean Basin during 45 days during summer 2010 together with surface “wind observations” or weather data based in real hourly observations of wind direction and velocity providing a real approximation of the tritium behavior after the release to the atmosphere from a fusion reactor. From comparison with NORMTRI - a code using climatologically sequences as input - over the same area, the real time results have demonstrated an apparent overestimation of the corresponding climatologically sequence of Tritium concentrations in air outputs, at several distances from the reactor. For this purpose two development patterns were established. The first one was following a cyclonic circulation over the Mediterranean Sea and the second one was based on the plume delivered over the Interior of the Iberian Peninsula and Continental Europe by another stabilized circulation corresponding to a High Pressure System. One of the important remaining activities defined then, was the qualification tool. In order to validate the model of ECMWF/FLEXPART we have developed of a new complete data base of tritium concentrations for the months from November 2010 to March 2011 and defined a new set of four patterns of HT transport in air, in each case using real boundary conditions: stationary to the North, stationary to the South, fast and very fast displacement. Finally the differences corresponding to those four early patterns (each one in assessments 1 and 2) has been analyzed in terms of the tuning of safety related issues and taking into account the primary phase o- - f tritium modeling, from its discharge to the atmosphere to the deposition on the ground, will affect to the complete tritium environmental pathway altering the chronic dose by absorption, reemission and ingestion both from elemental tritium, HT and from the oxide of tritium, HTO

Hydrodynamic stability theory of double ablation front structures in inertial confinement fusion.

The aim of inertial confinement fusion is the production of energy by the fusion of thermonuclear fuel (deuterium-tritium) enclosed in a spherical target due to its implosion. In the direct-drive approach, the energy needed to spark fusion reactions is delivered by the irradiation of laser beams that leads to the ablation of the outer shell of the target (the so-called ablator). As a reaction to this ablation process, the target is accelerated inwards, and, provided that this implosion is sufficiently strong a symmetric, the requirements of temperature and pressure in the center of the target are achieved leading to the ignition of the target (fusion). One of the obstacles capable to prevent appropriate target implosions takes place in the ablation region where any perturbation can grow even causing the ablator shell break, due to the ablative Rayleigh-Taylor instability. The ablative Rayleigh-Taylor instability has been extensively studied throughout the last 40 years in the case where the density/temperature profiles in the ablation region present a single front (the ablation front). Single ablation fronts appear when the ablator material has a low atomic number (deuterium/tritium ice, plastic). In this case, the main mechanism of energy transport from the laser energy absorption region (low density plasma) to the ablation region is the electron thermal conduction. However, recently, the use of materials with a moderate atomic number (silica, doped plastic) as ablators, with the aim of reducing the target pre-heating caused by suprathermal electrons generated by the laser-plasma interaction, has demonstrated an ablation region composed of two ablation fronts. This fact appears due to increasing importance of radiative effects in the energy transport. The linear theory describing the Rayleigh-Taylor instability for single ablation fronts cannot be applied for the stability analysis of double ablation front structures. Therefore, the aim of this thesis is to develop, for the first time, a linear stability theory for this type of hydrodynamic structures.

Ignition conditions for inertial confinement fusion targets with a nuclear spin-polarized DT fuel

The nuclear fusion cross-section is modified when the spins of the interacting nuclei are polarized. In the case of deuterium?tritium it has been theoretically predicted that the nuclear fusion cross-section could be increased by a factor d = 1.5 if all the nuclei were polarized. In inertial confinement fusion this would result in a modification of the required ignition conditions. Using numerical simulations it is found that the required hot-spot temperature and areal density can both be reduced by about 15% for a fully polarized nuclear fuel. Moreover, numerical simulations of a directly driven capsule show that the required laser power and energy to achieve a high gain scale as d-0.6 and d-0.4 respectively, while the maximum achievable energy gain scales as d0.9.

Quantum molecular dynamics and ab initio studies of the crystal structure of hydrogen and deuterium

Hydrogen isotopes play a critical role both in inertial and magnetic confinemen Nuclear Fusion. Since the preferent fuel needed for this technology is a mixture of deuterium and tritium. The study of these isotopes particularly at very low temperatures carries a technological interest in other applications. The present line promotes a deep study on the structural configuration that hydrogen and deuterium adopt at cryogenic temperatures and at high pressures. Typical conditions occurring in present Inertial Fusion target designs. Our approach is aims to determine the crystal structure characteristics, phase transitions and other parameters strongly correlated to variations of temperature and pressure.

Ab initio Methodology for Calculating Elastic Constant and Sound Velocity for Hydrogen and Deuterium in Molecular Solid Phase

Hydrogen isotopes play a critical role both in inertial and magnetic confinement Nuclear Fusion. Since the preferent fuel needed for this technology is a mixture of deuterium and tritium. The study of these isotopes particularly at very low temperatures carries a technological interest in other applications. The present line promotes a deep study on the structural configuration that hydrogen and deuterium adopt at cryogenic temperatures and at high pressures. Typical conditions occurring in present Inertial Fusion target designs. Our approach is aims to determine the crystal structure characteristics, phase transitions and other parameters strongly correlated to variations of temperature and pressure. With this results is possible calculated the elastic constant and sound velocity for hydrogen and deuterium in molecular solid phase.

Nanocrystals in the manufacture of target for inertial confinement fusion

Systems inertial confinement fusion (ICF) need of a manufacturing process targets very accurate and efficient (Fig. A). Due to the frequency needed for energy production techniques are necessary to achieve high repetition rates, however it is also necessary to increase or maintain the quality and efficiency of these targets. In order to observe more resolution possible problems in the target manufacture (B), we propose the following theoretical methodology, by means of which analyze different phenomena present in the conditions which are fabrication and handled deuterium tritium target spheres (DT ice). Recent experiments show that addition of instabilities caused by the geometry of the solid layer of DT ice (C), and the cover (ablator), one can relate the loss of power delivery in the implosion due to different conformations of the solid layers with regarding handling conditions.

Nanostructured tungsten as a first wall material for the future nuclear fusion reactors

The lack of materials able to withstand the severe radiation conditions (high thermal loads and atomistic damage) expected in fusion reactors is the actual bottle neck for fusion to become a reality. The main requisite for plasma facing materials (PFM) is to have excellent structural stability since severe cracking or mass loss would hamper their protection role which turns out to be unacceptable. Additional practical requirements for plasma facing materials are among others: (i) high thermal shock resistance, (ii) high thermal conductivity (iii) high melting point (iv) low physical and chemical sputtering, and (v) low tritium retention.