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

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

A model of the W/Cu interface in the ITER cooling monoblocks from Density Functional Theory

International audienceITER divertor is built with tungsten monoblocks that contain a tunsgten-copper interface. There, hydrogen isotopes could possibly accumulate leading to safety and mechanical issues. As a consequence, the tunsgten-copper interface has to be modeled and characterized, which is here performed at the atomic-scale by means of density functional theory calculations. In order to build the model, we selected the tungsten and copper orientations that minimizes the mismatch between both networks; this results in the W(001)/Cu$^{bcc}$(001) and W(001)/Cu$^{fcc}$(001)R45° interfaces. After relaxation, both systems converge to the same W$^{bcc}$(001)/Cu$^{hcp}$(11-20) structure, which is consistent with previous experimental observations. Such reconstruction of the copper network has the effect of changing the charge density in the copper part of the interface, with possible effects on hydrogen interaction

A model of the W/Cu interface in the ITER cooling monoblocks from Density Functional Theory

International audienceITER divertor is built with tungsten monoblocks that contain a tunsgten-copper interface. There, hydrogen isotopes could possibly accumulate leading to safety and mechanical issues. As a consequence, the tunsgten-copper interface has to be modeled and characterized, which is here performed at the atomic-scale by means of density functional theory calculations. In order to build the model, we selected the tungsten and copper orientations that minimizes the mismatch between both networks; this results in the W(001)/Cu$^{bcc}$(001) and W(001)/Cu$^{fcc}$(001)R45° interfaces. After relaxation, both systems converge to the same W$^{bcc}$(001)/Cu$^{hcp}$(11-20) structure, which is consistent with previous experimental observations. Such reconstruction of the copper network has the effect of changing the charge density in the copper part of the interface, with possible effects on hydrogen interaction

Multi-physics modelling of nuclear fusion device sub-components: The tritium breeding blanket

International audienceThe DEMOnstration fusion reactor is planned to be a first-of-a-kind nuclear fusion power plant to be built as the successor to ITER. The objectives of the design are the production of net electricity and operation with a closed fuel cycle. Replenishment of tritium fuel on-site will be imperative for the functioning of future commercial nuclear fusion reactors. Thus, accurate modelling of hydrogen transport and inventories within the reactor will be essential for safety issues and economic sustainability.The open-source FEniCS-based hydrogen transport code FESTIM [1] is used to perform multi-material, multi-dimensional and multi-physics simulations of a reactor sub-component: the breeding blanket. The design showcased is the Water-Cooled-Lithium-Lead (WCLL) concept [2], which utilises a liquid metal (LiPb) tritium breeding material. A companion fluid dynamics solver has been developed with FEniCSx to model the flow of the breeding material, accounting for MHD effects. The solver was verified using published analytical solutions [3, 4]. The resulting velocity field was coupled with FESTIM to accurately simulate hydrogen transport in both the liquid and structural domains of the component.[1] R.Delaporte-Mathurin, et al. NME, 21, 2019[2] A.Del Nevo, et al. Fusion Eng. Des., 124, 2017[3] J.Shercliff. Math. Proc. Camb. Philos. Soc, 49, 1953[4] J.Hunt. J. Fluid Mech, 21, 196

Multi-physics modelling of nuclear fusion device sub-components: The tritium breeding blanket

International audienceThe DEMOnstration fusion reactor is planned to be a first-of-a-kind nuclear fusion power plant to be built as the successor to ITER. The objectives of the design are the production of net electricity and operation with a closed fuel cycle. Replenishment of tritium fuel on-site will be imperative for the functioning of future commercial nuclear fusion reactors. Thus, accurate modelling of hydrogen transport and inventories within the reactor will be essential for safety issues and economic sustainability.The open-source FEniCS-based hydrogen transport code FESTIM [1] is used to perform multi-material, multi-dimensional and multi-physics simulations of a reactor sub-component: the breeding blanket. The design showcased is the Water-Cooled-Lithium-Lead (WCLL) concept [2], which utilises a liquid metal (LiPb) tritium breeding material. A companion fluid dynamics solver has been developed with FEniCSx to model the flow of the breeding material, accounting for MHD effects. The solver was verified using published analytical solutions [3, 4]. The resulting velocity field was coupled with FESTIM to accurately simulate hydrogen transport in both the liquid and structural domains of the component.[1] R.Delaporte-Mathurin, et al. NME, 21, 2019[2] A.Del Nevo, et al. Fusion Eng. Des., 124, 2017[3] J.Shercliff. Math. Proc. Camb. Philos. Soc, 49, 1953[4] J.Hunt. J. Fluid Mech, 21, 196

3D effects on hydrogen transport in divertor monoblocks: influence of thickness and recombination on poloidal side

International audienceBombardment by high energy hydrogen particles (deuterium and tritium) of the tungsten divertor surfaces will lead to a build-up of hydrogen inventory, which can induce material embrittlement and so reduce the lifetime of plasma-facing components. Because tritium can be retained in the material and permeate to the coolant, it also represents radioactive issue.Hydrogen transport in ITER monoblocks has already been modelled numerically with 1D and 2D simulations for large sets of irradiation conditions, assuming there is no effect of the monoblock axial thickness (poloidal direction) due to its large size defined for ITER design (12 mm). Since the conceptual design for DEMO monoblock can still change, the aim for this study is to explore the impact of the monoblock axial thickness on the retention and permeation during plasma operation. Desorption from both, toroidal and poloidal gaps, is also studied during the baking phase. A 3D FESTIM [1] model is first built and transient simulations up to 1e6 s of continuous exposure are run with or without instantaneous recombination on poloidal side surfaces.In the case of instantaneous recombination, the poloidal gaps act as a strong sink for hydrogen leading to a decrease of the monoblock inventory. The total desorption flux on poloidal surfaces is greater than on poroidal surfaces but remains orders of magnitude lower than the retro-desorbed flux at the plasma-facing surface. For a monoblock thickness of 4 mm, the relative difference in the hydrogen inventory per unit thickness between the two cases (with and without recombination on poloidal sides) is ~200%. As the thickness of the monoblock increases, this difference decreases (~30% at 14 mm). The monoblock’s response to baking is then studied at different baking temperatures. For example, at 600 K almost all the hydrogen content in the monoblock is removed after 15 days of baking (mostly outgassing ~70% from poloidal side surfaces). Finally, it is shown that assuming a non-instantaneous recombination on the tungsten surfaces according to the literature data [2] would not have a major impact for baking temperatures above 600 K.[1] R. Delaporte-Mathurin et al., Nuclear Materials and Energy 21, p. 100709 (2019) [2] D. F. Anderl et al., Fusion Technology 21(2P2), pp. 745–752 (1992

3D effects on hydrogen transport in divertor monoblocks: influence of thickness and recombination on poloidal side

International audienceBombardment by high energy hydrogen particles (deuterium and tritium) of the tungsten divertor surfaces will lead to a build-up of hydrogen inventory, which can induce material embrittlement and so reduce the lifetime of plasma-facing components. Because tritium can be retained in the material and permeate to the coolant, it also represents radioactive issue.Hydrogen transport in ITER monoblocks has already been modelled numerically with 1D and 2D simulations for large sets of irradiation conditions, assuming there is no effect of the monoblock axial thickness (poloidal direction) due to its large size defined for ITER design (12 mm). Since the conceptual design for DEMO monoblock can still change, the aim for this study is to explore the impact of the monoblock axial thickness on the retention and permeation during plasma operation. Desorption from both, toroidal and poloidal gaps, is also studied during the baking phase. A 3D FESTIM [1] model is first built and transient simulations up to 1e6 s of continuous exposure are run with or without instantaneous recombination on poloidal side surfaces.In the case of instantaneous recombination, the poloidal gaps act as a strong sink for hydrogen leading to a decrease of the monoblock inventory. The total desorption flux on poloidal surfaces is greater than on poroidal surfaces but remains orders of magnitude lower than the retro-desorbed flux at the plasma-facing surface. For a monoblock thickness of 4 mm, the relative difference in the hydrogen inventory per unit thickness between the two cases (with and without recombination on poloidal sides) is ~200%. As the thickness of the monoblock increases, this difference decreases (~30% at 14 mm). The monoblock’s response to baking is then studied at different baking temperatures. For example, at 600 K almost all the hydrogen content in the monoblock is removed after 15 days of baking (mostly outgassing ~70% from poloidal side surfaces). Finally, it is shown that assuming a non-instantaneous recombination on the tungsten surfaces according to the literature data [2] would not have a major impact for baking temperatures above 600 K.[1] R. Delaporte-Mathurin et al., Nuclear Materials and Energy 21, p. 100709 (2019) [2] D. F. Anderl et al., Fusion Technology 21(2P2), pp. 745–752 (1992

2D Tritium retention prediction in the ITER/DEMO actively cooled divertor in the presence of neutron induced defects

International audienc