Name: Chao Yin
Date: November 20, 2020 16h00 - 19h00
This is an online event.
Assessment of mechanical properties of neutron irradiated tungsten and its alloys
Tungsten (W) is the main candidate material for divertor component and plasma-facing components (PFCs) armor in the nuclear fusion reactor. In the magnetic confinement reactor called “Tokamak”, the confined hot plasma is directed by the magnetic field to the divertor component in order to remove the impurities and He ash. Moreover, the surrounding components in the reactor vacuum chamber are irradiated by the fast neutron generated from D-T reaction. These high heat flux and neutron flux will degrade the mechanical performance of the divertor component. Therefore, the investigation of mechanical properties after high-temperature neutron irradiation is essential for designing a divertor that can endure this harsh environment.
Twelve different W grades were investigated to explore the effect of initial microstructure, presence of strengthening particles, fabrication route applied, potential of additive manufacturing to deliver the materials with sufficient resistance to the neutron irradiation. In order to reach these targets, the studies have been performed across three directions: (i) miniaturized mechanical testing methods were developed, adapted, or modified in terms of specimen design, loading condition, loading fixtures, and data reduction schemes; (ii) the mechanical tests on non-irradiated W products were analyzed and interpreted concerning the microstructure and fracture surfaces as characterized by SEM, EBSD, and EDS; (iii) both procedures were applied to neutron irradiated materials, performing tests in Hot Cells of Belgian Nuclear Research Center (SCK CEN). The neutron irradiation was performed in the Belgian materials testing reactor (BR2) in Mol, Belgium. For the first time, the irradiation temperature applied was as high as 1200°C and the exposure fluence corresponded to the end-of-life of ITER (~1 displacement per atom).
SCK CEN mentor:
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