Sputtering of beryllium oxide by deuterium at various temperatures simulated with molecular dynamics

The sputtering yield of beryllium oxide (BeO) by incident deuterium (D) ions, for energies from 10 eV to 200 eV, has been calculated for temperatures between 300 K and 800 K using classical molecular dynamics. First, cumulative irradiations are carried out to build up a concentration of D in the material, equal to the experimentally measured concentration, that varies from an atomic fraction of 0.12 (300 K-500 K) to 0.02 ( 800 K). After building up the concentration of D, noncumulative irradiations are carried out to estimate the sputtering yields of BeO. For all incident energies, the sputtering yield peaks at 500 K, being closely related to the decrease of the concentration of D above this temperature. At 10 eV, the concentration of D on the surface drives the temperature dependence, while above 30 eV, it is the amount of surface damage created during the cumulative irradiation.Peer reviewe

Surface coverage dependent mechanisms for the absorption and desorption of hydrogen from the W(110) and W(100) surfaces : a density functional theory investigation

Herein we investigate absorption and desorption of hydrogen in the sub-surface of tungsten via Density Functional Theory. Both the near-surface diffusion and recombination of a bulk hydrogen atom with a hydrogen atom adsorbed upon the W(110) and W(100) surfaces are investigated at various surface adsorption coverage ratios. This study intends to model the desorption processes occurring during Thermal-Desorption Spectroscopy experiments and the absorption of hydrogen during gaseous or low energy atomic exposure. Since the diffusion and recombination processes are expected to change as the hydrogen coverage of the surface varies, different coverage ratios were investigated on both surfaces. We found that at saturation coverage of hydrogen on both surfaces, the activation barriers for the recombination of molecular hydrogen are below 0.8 eV. On the contrary, below saturation, the activation barriers for recombination rise to 1.35 eV and 1.51 eV depending on the coverage and on the orientation of the surface. Regarding the absorption of atomic hydrogen from the surface into the bulk, the activation barrier raises from less than 1.0 eV at saturation to around 1.7 eV below saturation on both surfaces. These results indicate that surface mechanisms certainly play a significant role in the kinetics of desorption of hydrogen from tungsten; it is also expected that surface mechanisms affect the total amount on hydrogen absorbed in tungsten during implantation.Peer reviewe

Analytical bond order potential for simulations of BeO 1D and 2D nanostructures and plasma-surface interactions

An analytical interatomic bond order potential for the Be–O system is presented. The potential is fitted and compared to a large database of bulk BeO and point defect properties obtained using density functional theory. Its main applications include simulations of plasma-surface interactions involving oxygen or oxide layers on beryllium, as well as simulations of BeO nanotubes and nanosheets. We apply the potential in a study of oxygen irradiation of Be surfaces, and observe the early stages of an oxide layer forming on the Be surface. Predicted thermal and elastic properties of BeO nanotubes and nanosheets are simulated and compared with published ab initio data.Peer reviewe

New rate equation model to describe the stabilization of displacement damage by hydrogen atoms during ion irradiation in tungsten

The effect of deuterium (D) presence on the amount of displacement damage created in tungsten (W) during high-energy W-ion irradiation is investigated. For this purpose, we have performed modelling of experimental results where W was sequentially or simultaneously irradiated by 10.8 MeV W ions and exposed to 300 eV D ions. A novel displacement damage creation and stabilization model was newly developed and introduced into the MHIMS-Reservoir (Migration of Hydrogen Isotopes in MaterialS) code. It employs macroscopic rate equations (MREs) for solving the evolution of solute and trapped D concentrations in the material. The new displacement damage creation and stabilization model is based on spontaneous recombination of Frenkel pairs and stabilization of defects that are occupied by D atoms. By using the new model, we could successfully replicate the measured D depth profiles and D thermal desorption data, where a higher defect concentration was observed when D was present during W irradiation as compared to when no D was present. For this we utilized parameters, which include the number of distinct defect types, the de-trapping energies of their fill-levels, their saturation concentrations and their probability for stabilization if they contain a D during the W-ion irradiation. To successfully replicate the experimental results three distinct defect types were needed with several fill-levels. By comparing the de-trapping energies of the defect fill-levels with data available from the literature, the defect types were identified as single-vacancies, small vacancy clusters and large vacancy clusters. The effect of D presence was found to be largest in single vacancies as its concentration increased by about a factor of three, while the concentration of small vacancy clusters increased by about a factor of two. Large vacancy clusters were found to be largely unaffected as they showed very little increase in concentration when D was presentPeer reviewe

Finite element analysis of hydrogen retention in ITER plasma facing components using FESTIM

The behaviour of hydrogen isotopes in ITER monoblocks was studied using the code FESTIM (Finite Element Simulation of Tritium In Materials) which is introduced in this publication. FESTIM has been validated by reproducing experimental data and the Method of Manufactured Solutions was used for analytical verification. Following relevant plasma scenarios, both transient heat transfer and hydrogen isotopes (HIs) diffusion have been simulated in order to assess HIs retention in monoblocks. Relevant materials properties have been used. Each plasma cycle is composed of a current ramp up, a current plateau, a current ramp down and a resting phase before the following shot. 100 cycles are simulated. The total HIs inventory in the tokamak during resting phases reaches 1.8 x 10(-3) mgwhereas during the implantation phases it keeps increasing as a power law of time. Particle flux on the cooling channel of the monoblock is also computed. The breakthrough time is estimated to be t = 1 x 10(5) s which corresponds to 24 cycles. Relevance of 2D modelling has been demonstrated by comparing the total HIs inventory obtained by 2D and 1D simulations. Using 1D simulations, a relative error is observed compared to 2D simulations which can reach -25% during the resting phase. The error during implantation phases keeps increasing.Peer reviewe

Stabilization of defects by the presence of hydrogen in tungsten : simultaneous W-ion damaging and D-atom exposure

The possible mutual influence and synergistic effect between defect production and the presence of hydrogen isotopes in the crystal lattice of tungsten is studied. For this purpose, we perform modeling of experimental data where polycrystalline tungsten samples were in one case sequentially irradiated by 10.8 MeV tungsten ions followed by low-energy deuterium exposure and in the other case simultaneously irradiated by tungsten ions, while exposed to deuterium atoms. Modeling of the measured deuterium depth profiles and thermal-desorption spectra for different irradiation temperatures is performed by the MHIMS (migration of hydrogen isotopes in materials) code. A model of trap creation due to tungsten ion irradiation during the deuterium atom exposures is implemented. In both experimental series, the deuterium desorption peaks corresponding to defects induced by tungsten irradiation are described by the same two de-trapping energies of 1.83 and 2.10 eV. The experiments give unambiguous proof that the presence of deuterium increases the overall trap density. The modeling reveals that the two trap concentrations are affected differently by the temperature and presence of deuterium: the concentration of the low-energy trap is significantly higher in the case of simultaneous exposure compared to sequential exposure, especially at high temperature (2.2 times higher at 1000 K). The concentration of the high-energy trap is only weakly affected by the presence of hydrogen.Peer reviewe

Study and modeling of the deuterium trapping in ITER relevant materials

Lors de l’opération d’ITER, des flux importants d’isotopes d’hydrogène (HI) constituant le fuel interagissent avec les composants face au plasma (CFP) de la machine. Dans le cas du Tungstène (W) composant le divertor qui est la zone la plus exposée aux interactions plasma paroi, le flux incident est implanté et diffuse ensuite dans le corps du matériau entrainant un piégeage du fuel. Pour des raisons de sureté, l’inventaire de Tritium retenu dans les parois d’ITER est limité. De plus, le dégazage du fuel depuis les parois vers le plasma, lors des opérations plasma peut avoir un impact sur le contrôle global du plasma. Le but de cette thèse est d’abord de déterminer les paramètres de piégeages du fuel dans le W (énergies/températures de dépiégeage, concentrations de pièges) grâce à la modélisation de résultats expérimentaux. Ces simulations de résultats expérimentaux montrent que l’implantation d’HIs dans le W peut induire, sous certaines conditions, la formation de lacunes contenant des impuretés. En plus de ce piège induit par l’implantation d’ions, 2 pièges intrinsèques sont présents dans le W. Ces 3 pièges retiennent les HIs jusqu’à 700 K. Enfin, il est montré que le W endommagé par des ions lourds ou des neutrons contient des dislocations, des boucles de dislocations et des cavités retenant les HIs jusqu’à 1000 K.Après avoir déterminé ces paramètres de piégeages des HIs dans le W, la rétention des HIs durant l’opération d’ITER est estimée. Lors de cette opération, la température des CFP W atteint environ 1000 K. Les simulations montrent donc que la rétention dans les CFPs non endommagé est bien plus faible que dans le cas d’un CFP endommagé.During ITER operation, important flux of Hydrogen Isotopes (HIs) constituting the fuel interact with the plasma facing components (PFC) of the machine. In the case of tungsten (W) making the divertor which is the most exposed area to the plasma wall interaction, the incident flux can be implanted and diffuse inside the bulk material inducing a trapping of the fuel. To safety issue, the tritium inventory retained in ITER’s PFC is limited. In addition, the outgassing of the fuel during plasma operation can impact the edge plasma control.The aim of this PhD project is first to determined relevant trapping parameters of the fuel in W (detrapping energies/temperatures and trap concentrations) by modelling experimental results. The simulations of experimental results shows that under specific condition, the HI implantation can induce the formation of mono-vacancies containing impurities. In addition to this induced trap, 2 intrinsic traps are present in W. This 3 traps retain HIs up to 700 K. Finally, it has been shown that the damaged W by heavy ions or neutrons contains dislocations, dislocation loops and cavities that can trap HIs up to 1000 K.After determining the fuel retention properties of W, the HIs retention during ITER operation is estimated. During this operation, the PFC temperature reaches around 1000 K so the simulations show that the damaged W retains much more HIs than the undamaged W

Etude de l’implantation du deutérium dans les composés face au plasma constituants du tokamak ITER

During ITER operation, important flux of Hydrogen Isotopes (HIs) constituting the fuel interact with the plasma facing components (PFC) of the machine. In the case of tungsten (W) making the divertor which is the most exposed area to the plasma wall interaction, the incident flux can be implanted and diffuse inside the bulk material inducing a trapping of the fuel. To safety issue, the tritium inventory retained in ITER’s PFC is limited. In addition, the outgassing of the fuel during plasma operation can impact the edge plasma control.The aim of this PhD project is first to determined relevant trapping parameters of the fuel in W (detrapping energies/temperatures and trap concentrations) by modelling experimental results. The simulations of experimental results shows that under specific condition, the HI implantation can induce the formation of mono-vacancies containing impurities. In addition to this induced trap, 2 intrinsic traps are present in W. This 3 traps retain HIs up to 700 K. Finally, it has been shown that the damaged W by heavy ions or neutrons contains dislocations, dislocation loops and cavities that can trap HIs up to 1000 K.After determining the fuel retention properties of W, the HIs retention during ITER operation is estimated. During this operation, the PFC temperature reaches around 1000 K so the simulations show that the damaged W retains much more HIs than the undamaged W.Lors de l’opération d’ITER, des flux importants d’isotopes d’hydrogène (HI) constituant le fuel interagissent avec les composants face au plasma (CFP) de la machine. Dans le cas du Tungstène (W) composant le divertor qui est la zone la plus exposée aux interactions plasma paroi, le flux incident est implanté et diffuse ensuite dans le corps du matériau entrainant un piégeage du fuel. Pour des raisons de sureté, l’inventaire de Tritium retenu dans les parois d’ITER est limité. De plus, le dégazage du fuel depuis les parois vers le plasma, lors des opérations plasma peut avoir un impact sur le contrôle global du plasma.Le but de cette thèse est d’abord de déterminer les paramètres de piégeages du fuel dans le W (énergies/températures de dépiégeage, concentrations de pièges) grâce à la modélisation de résultats expérimentaux. Ces simulations de résultats expérimentaux montrent que l’implantation d’HIs dans le W peut induire, sous certaines conditions, la formation de lacunes contenant des impuretés. En plus de ce piège induit par l’implantation d’ions, 2 pièges intrinsèques sont présents dans le W. Ces 3 pièges retiennent les HIs jusqu’à 700 K. Enfin, il est montré que le W endommagé par des ions lourds ou des neutrons contient des dislocations, des boucles de dislocations et des cavités retenant les HIs jusqu’à 1000 K.Après avoir déterminé ces paramètres de piégeages des HIs dans le W, la rétention des HIs durant l’opération d’ITER est estimée. Lors de cette opération, la température des CFP W atteint environ 1000 K. Les simulations montrent donc que la rétention dans les CFPs non endommagé est bien plus faible que dans le cas d’un CFP endommagé

Multiphysics tritium transport modelling in WCLL breeding blankets: Influence of MHD effects and neutron damage

International audienceThe replenishment of tritium fuel in a breeding blanket is fundamental for the operation of commercial DT fusion reactors. Accurate modelling of hydrogen transport and inventories within the breeding blanket will be essential for safety issues and economic sustainability. One of the proposed breeding blanket concepts for DEMO, the Water Cooled Lithium-Lead (WCLL) concept is modelled using The open-source hydrogen transport code FESTIM. Multi-material and multi-physics 3D simulations of the WCLL design have been conducted to investigate the significance of tritium inventories and permeation into cooling channels. Trapping effects are considered in the solid domains, in addition to how trapping properties alter as a results of neutron damage over time, subsequently affecting tritium inventories. A fluid dynamics model is implemented to simulate the flow of the liquid metal LiPb in the blanket, accounting for MHD effects. The resulting velocity field was coupled with FESTIM to accurately simulate hydrogen transport in both the liquid and structural domains of the model. A novel method for modelling permeation barriers is presented by modification of the conditions at the discontinuous boundaries between material domains. Simulations have been conducted assuming DEMO operates at steady-state, for a full power year. The presence of a magnetic field is shown to severely disrupts the flow regime of the liquid metal breeder. The inclusion of trapping mechanisms has been shown to increase tritium inventories by 15%. However, when considering the neutron damage effects, inventories increase by several orders of magnitude and result in localised build ups closest to the sources of tritium

Multiphysics tritium transport modelling in WCLL breeding blankets: Influence of MHD effects and neutron damage

International audienceThe replenishment of tritium fuel in a breeding blanket is fundamental for the operation of commercial DT fusion reactors. Accurate modelling of hydrogen transport and inventories within the breeding blanket will be essential for safety issues and economic sustainability. One of the proposed breeding blanket concepts for DEMO, the Water Cooled Lithium-Lead (WCLL) concept is modelled using The open-source hydrogen transport code FESTIM. Multi-material and multi-physics 3D simulations of the WCLL design have been conducted to investigate the significance of tritium inventories and permeation into cooling channels. Trapping effects are considered in the solid domains, in addition to how trapping properties alter as a results of neutron damage over time, subsequently affecting tritium inventories. A fluid dynamics model is implemented to simulate the flow of the liquid metal LiPb in the blanket, accounting for MHD effects. The resulting velocity field was coupled with FESTIM to accurately simulate hydrogen transport in both the liquid and structural domains of the model. A novel method for modelling permeation barriers is presented by modification of the conditions at the discontinuous boundaries between material domains. Simulations have been conducted assuming DEMO operates at steady-state, for a full power year. The presence of a magnetic field is shown to severely disrupts the flow regime of the liquid metal breeder. The inclusion of trapping mechanisms has been shown to increase tritium inventories by 15%. However, when considering the neutron damage effects, inventories increase by several orders of magnitude and result in localised build ups closest to the sources of tritium