Silica final lens performance in laser fusion facilities: HiPER and LIFE

Nowadays, the projects LIFE (Laser Inertial Fusion Energy) in USA and HiPER (High Power Laser Energy Research) in Europe are the most advanced ones to demonstrate laser fusion energy viability. One of the main points of concern to properly achieve ignition is the performance of the final optics (lenses) under the severe irradiation conditions that take place in fusion facilities. In this paper, we calculate the radiation fluxes and doses as well as the radiation-induced temperature enhancement and colour centre formation in final lenses assuming realistic geometrical configurations for HiPER and LIFE. On these bases, the mechanical stresses generated by the established temperature gradients are evaluated showing that from a mechanical point of view lenses only fulfil specifications if ions resulting from the imploding target are mitigated. The absorption coefficient of the lenses is calculated during reactor startup and steady-state operation. The obtained results reveal the necessity of new solutions to tackle ignition problems during the startup process for HiPER. Finally, we evaluate the effect of temperature gradients on focal length changes and lens surface deformations. In summary, we discuss the capabilities and weak points of silica lenses and propose alternatives to overcome predictable problems

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

Equation of State and Opacities for Warm Dense Matter.

We will present recent developments in the calculation of opacity and equation of state tables suitable for including in the radiation hydrodynamic code ARWEN [1] to study processes like ICF or X-ray secondary sources. For these calculations we use the code BiG BART to compute opacities in LTE conditions, with self-consistent data generated with the Flexible Atomic Code (FAC) [2]. Non-LTE effects are approximately taken into account by means of the improved RADIOM model [3], which makes use of existing LTE data tables. We use the screened-hydrogenic model [4] to derive the Equation of State using the population and energy of the levels avaliable from the atomic data

Direct-Drive Shock-Ignition for the Laser Megajoule.

We present a review of direct-drive shock ignition studies done as alternative for the Laser Mega-Joule to achieve high thermonuclear gain. One-dimensional analysis of HiPER-like Shock-ignited target designs is presented. It is shown that high gain can be achieved with shock ignition for designs which do not ignite only from the laser compression. Shock ignition is achieved for different targets of the fast ignition family which are driven by an absorbed energy between 100 kJ and 850kJ and deliver thermonuclear energies between 10-130 MJ. Shock-Ignition of Direct-Drive Double-Shell non-cryogenic target is also addressed. 2D results concerning the LMJ irradiation geometry are presented. Few systematic analyses are performed for the fuel assembly irradiation uniformity using the whole LMJ configuration or a part of the facility, and for the ignitor spike uniformity. Solutions for fuel assembly and shock ignition on LMJ using 2D calculations are presented. It is shown that high-gain shock-ignition is possible with intensity of each quad less than 1e15 W/cm2but low modes asymmetries displace the ignitor power in the spike towards higher powers.

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.

Studies of a Self Cooled Lead Lithium blanket for HiPER Reactor

Within the frame of the HiPER reactor, we propose and study a Self Cooled Lead Lithium blanket with two different cooling arrangements of the system First Wall – Blanket for the HiPER reactor: Integrated First Wall Blanket and Separated First Wall Blanket. We compare the two arrangements in terms of power cycle efficiency, operation flexibility in out-off-normal situations and proper cooling and acceptable corrosion. The Separated First Wall Blanket arrangement is superior in all of them, and it is selected as the advantageous proposal for the HiPER reactor blanket. However, it still has to be improved from the standpoint of proper cooling and corrosion rates

Stopping of α particles from the core in corona plasmas

In the laser fusion reactor design, the protection of first wall and the final optics from high energy ions is the key issue. So, it is necessary to predict the precise energy spectra of ions.In the previous reactor designs, the ion energy spectra were provided by the classical ion transport codes. However, this poster shows that the α particle spectrum is significantly modified by the anomalous process in ablated plasmas.

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.

Self Cooled Lead Lithium blanket and reactor for HiPER

Within the frame of the HiPER reactor, we propose and study a Self Cooled Lead Lithium blanket with two different cooling arrangements of the system First Wall – Blanket for the HiPER reactor: Integrated First Wall Blanket and Separated First Wall Blanket. We compare the two arrangements in terms of power cycle efficiency, operation flexibility in out-off-normal situations and proper cooling and acceptable corrosion. The Separated First Wall Blanket arrangement is superior in all of them, and it is selected as the advantageous proposal for the HiPER reactor blanket. However, it still has to be improved from the standpoint of proper cooling and corrosion rates

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.

Effect of ion flux on helium retention in helium-irradiated tungsten

Helium retention in irradiated tungsten leads to swelling, pore formation, sample exfoliation and embrittlement with deleterious consequences in many applications. In particular, the use of tungsten in future nuclear fusion plants is proposed due to its good refractory properties. However, serious concerns about tungsten survivability stems from the fact that it must withstand severe irradiation conditions. In magnetic fusion as well as in inertial fusion (particularly with direct drive targets), tungsten components will be exposed to low and high energy ion (helium) irradiation, respectively. A common feature is that the most detrimental situations will take place in pulsed mode, i.e., high flux irradiation. There is increasing evidence on a correlation between a high helium flux and an enhancement of detrimental effects on tungsten. Nevertheless, the nature of these effects is not well understood due to the subtleties imposed by the exact temperature profile evolution, ion energy, pulse duration, existence of impurities and simultaneous irradiation with other species. Physically based Kinetic Monte Carlo is the technique of choice to simulate the evolution of radiation-induced damage inside solids in large temporal and space scales. We have used the recently developed code MMonCa (Modular Monte Carlo simulator), presented in this conference for the first time, to study He retention (and in general defect evolution) in tungsten samples irradiated with high intensity helium pulses. The code simulates the interactions among a large variety of defects and impurities (He and C) during the irradiation stage and the subsequent annealing steps. In addition, it allows us to vary the sample temperature to follow the severe thermo-mechanical effects of the pulses. In this work we will describe the helium kinetics for different irradiation conditions. A competition is established between fast helium cluster migration and trapping at large defects, being the temperature a determinant factor. In fact, high temperatures (induced by the pulses) are responsible for large vacancy cluster formation and subsequent additional trapping with respect to low flux irradiation.

Effect of ion flux on helium retention in helium-irradiated tungsten

Helium retention in irradiated tungsten leads to swelling, pore formation, sample exfoliation and embrittlement with deleterious consequences in many applications. In particular, the use of tungsten in future nuclear fusion plants is proposed due to its good refractory properties. However, serious concerns about tungsten survivability stems from the fact that it must withstand severe irradiation conditions. In magnetic fusion as well as in inertial fusion (particularly with direct drive targets), tungsten components will be exposed to low and high energy ion irradiation (helium), respectively. A common feature is that the most detrimental situations will take place in pulsed mode, i.e., high flux irradiation. There is increasing evidence of a correlation between a high helium flux and an enhancement of detrimental effects on tungsten. Nevertheless, the nature of these effects is not well understood due to the subtleties imposed by the exact temperature profile evolution, ion energy, pulse duration, existence of impurities and simultaneous irradiation with other species. Object Kinetic Monte Carlo is the technique of choice to simulate the evolution of radiation-induced damage inside solids in large temporal and space scales. We have used the recently developed code MMonCa (Modular Monte Carlo simulator), presented at COSIRES 2012 for the first time, to study He retention (and in general defect evolution) in tungsten samples irradiated with high intensity helium pulses. The code simulates the interactions among a large variety of defects and during the irradiation stage and the subsequent annealing steps. The results show that the pulsed mode leads to significantly higher He retention at temperatures higher than 700 K. In this paper we discuss the process of He retention in terms of trap evolution. In addition, we discuss the implications of these findings for inertial fusion.

Modelización y validación experimental del transporte de tritio en metal líquido y otros materiales para reactores de fusión nuclear magnética e inercial

La presente tesis se centra en el estudio de los fenómenos de transporte de los isótopos de hidrógeno, y más concretamente del tritio, en materiales de interés para los reactores de fusión nuclear. Los futuros reactores de fusión nuclear necesitarán una Planta de Tritio, con una envoltura regeneradora (breeding blanket) y unos sistemas auxiliares claves para su diseño. Por lo tanto su desarrollo y cualificación son cruciales para demostrar que los reactores de fusión son una opción viable como futura fuente de energía. Se han resaltado los diferentes retos de la difusión y retención de estas especies ligeras para cada sistema de la Planta de Tritio, y se han identificado las necesidades experimentales y paramétricas para abordar las simulaciones de difusión, como factores de transporte como la difusividad, absorción/desorción, solubilidad y atrapamiento. Se han estudiado los fenómenos de transporte y parámetros del T en el metal líquido LiPb, componente del breeding blanket tanto para una planta de fusión magnética como inercial. Para ello se han utilizado dos experimentos con características diversas, uno de ellos se ha llevado a cabo en un reactor de alto flujo, y por lo tanto, en condiciones de irradiación, y el otro sin irradiación. Los métodos de simulación numérica aplicados se han adaptado a los experimentos para las mediciones y para estudiar el régimen de transporte. En el estudio de estos experimentos se ha obtenido un valor para algunos de los parámetros claves en el transporte y gestión del tritio en el reactor. Finalmente se realiza un cálculo de la acumulación y difusión de tritio en una primera pared de tungsteno para un reactor de fusión inercial. En concreto para el proyecto de fusión por láser europeo, HiPER (para sus fases 4a y 4b). Se ha estudiado: la implantación de los isótopos de H y He en la pared de W tras una reacción de fusión por iluminación directa con un láser de 48MJ; el efecto en el transporte de T de los picos de temperatura en el W debido a la frecuencia de los eventos de fusión; el régimen de transporte en la primera pared. Se han identificado la naturaleza de las trampas más importantes para el T y se ha propuesto un modelo avanzado para la difusión con trampas. ABSTRACT The present thesis focuses into study the transport phenomenons of hydrogen isotopes, more specifically tritium, in materials of interest for nuclear fusion reactors. The future nuclear reactors will be provided of a Tritium Plant, with its breeding blanket and its auxiliary systems, all of them essential components for the plant. Therefore a reliable development and coalification are key issues to prove the viability of the nuclear fusion reactors as an energy source. The currently challenges for the diffusion and accumulation of these light species for each system of the TP has been studied. Experimental and theoretical needs have been identified and analyzed, specially from the viewpoint of the parameters. To achieve reliable simulations of tritium transport, parameters as diffusivity, absorption/desorption, solubility and trapping must be reliables. Transport phenomenon and parameters of T in liquid metal have been studied. Lead lithium is a key component of the breeding blanket, either in magnetic or inertial fusion confinement. Having this aim in mind, two experiments with different characteristics have been used; one of them has been realized in a high flux reactor, and hence, under irradiation conditions. The other one has been realized without radiation. The mathematical methods for the simulation have been adapted to the experiments, for the measures and also to study the transport behavior. A value for some key parameters for tritium management has been obtained in these studies. Finally, tritium accumulation and diffusion in a W first wall of an inertial nuclear fusion reactor has been assessed. A diffusion model of the implanted H, D, T and He species for the two initial phases of the proposed European laser fusion Project HiPER (namely, phase 4a and phase 4b) has been implemented using Tritium Migration Analysis Program, TMAP7. The effect of the prompt and working temperatures and the operational pulsing modes on the diffusion are studied. The nature of tritium traps in W and their performance has been analyzed and discussed.

Fast ignition driven by quasi-monoenergetic ions: Optimal ion type and reduction of ignition energies with an ion beam array

Fast ignition of inertial fusion targets driven by quasi-monoenergetic ion beams is investigated by means of numerical simulations. Light and intermediate ions such as lithium, carbon, aluminum and vanadium have been considered. Simulations show that the minimum ignition energies of an ideal configuration of compressed Deuterium-Tritium are almost independent on the ion atomic number. However, they are obtained for increasing ion energies, which scale, approximately, as Z2, where Z is the ion atomic number. Assuming that the ion beam can be focused into 10 ?m spots, a new irradiation scheme is proposed to reduce the ignition energies. The combination of intermediate Z ions, such as 5.5 GeV vanadium, and the new irradiation scheme allows a reduction of the number of ions required for ignition by, roughly, three orders of magnitude when compared with the standard proton fast ignition scheme.