Le Mardi 15 Janvier 2019 à 14h00 en salle René Gravier bât.506 RdC
On Tuesday January 15, 2019, at 2:00 p.m., in René Gravier room, Bldg.506 groundfloor
Aura lieu le Séminaire / The following Seminar will be held:
Modeling Helium Plasma Surface Interactions in a Tungsten Divertor
Présenté par / Presented by :
Brian Wirth, on behalf of the SciDAC PSI project team
UTK, University of Tenessee, Knoxville
Nous comptons sur votre présence !
We are looking forward your attending !
Résumé / Abstract :
A new, integrated simulation capability for modeling wall erosion, impurity transport, surface and sub-surface changes, and gas recycling due to plasma-material interactions in the ITER tungsten (W) divertor is being developed through the PSI-SciDAC project, aiming to predict the long-term performance of plasma facing components (PFCs). A scale-bridging approach is being applied to develop surface response models, building on atomistic/microstructural models towards continuum-based simulations capable of simulating large time- and length-scale evolution. The surface models have been integrated with boundary plasma models to predict long-term surface evolution in the face of plasma exposure, and future plans include exploring the coupling between the plasma and wall during transient phenomena, with an emphasis on the dynamic recycling process. Simulations have been benchmarked against PISCES-A experiments of helium, in addition to mixed deuterium-helium, exposure on tungsten, which have successfully reproduced tungsten transport through the plasma and the sub-surface gas concentrations. Subsequently, we will describe modeling predictions of the plasma surface interactions expected in the ITER tungsten divertor for plasma discharge conditions of either helium (He) or D-T burning-plasma, with our current simulation results encompassing on the order of 10 second long discharges. Our results, focused on the analysis of the burning plasma case, demonstrate that neon (Ne) is the main radiative species and main contributor to wall erosion (over light impurities and low-energetic W ions). Most of the eroded W (>90%) re-deposits locally and leads to net deposition where the plasma temperature is low (R~Rsep<0.15m), and to net erosion where the plasma temperature is high (R-Rsep>0.2m). The depth profiles of gases implanted in the W divertor show little effect from impurities such as He, due to the low concentration of the latter. However, heat fluxes greatly affect the sub-surface tritium profiles, and increases in substrate temperature (from 343K to 525K at the peak heat flux location) during steady-state operation lead to faster gas diffusion, both into the bulk and outgassing. These simulations assumed an initial crystalline, clean W substrate. In comparison, pre-exposure of the W substrate to He plasma (e.g., from the He-operation) leads to a higher concentration of He and vacancy clusters near the surface, which locally increases the tritium concetration (relative to initial crystalline W) and reduces the permeation of hydrogenic species. Our work continues, mainly focused on reaching longer time-scales and seeking opportunities for experimental validation.