Jiming Chen

Erosion and re-deposition behavior of plasma facing materials due to tokamak plasma disruption

Abstract Plasma irradiation experiments of several kinds of carbon materials and molybdenum were performed in the HL-1M tokamak. The samples emerged into plasma by a depth of 1.5 cm with an angle of 45° between sample surface and magnetic field. After ≈300 plasma discharges with ohmic and NBI heating, melting, microstructure of thin flakes, gas holes and bubbles were found on the surface of molybdenum. No obvious damages were found on the surface of carbon materials, only carbon fibers of 3D-C/C composite were preferentially eroded. Weight changes indicated that most of the samples had little weight loss but few of the samples had large weight increments in which some droplet-shape deposits were found on their surface. Most of the erosion and re-depositions occurred at random on the surface of samples implied that the dominant factor influenced the erosion and re-deposition of plasma facing materials was tokamak plasma disruption.

R&D of the In-Vessel Materials for Future Magnetic Confinement Fusion Reactors at SWIP

Materials are the foundation for realizing fusion reactors, in which the in-vessel materials, such as plasma-facing materials, structural materials, and functional materials, are more critical since they will be exposed to high flux of edge plasma, high heat load, and 14 MeV neutron irradiations. Therefore, good performances at high temperatures and low induced activation by neutrons are essential requirements for in-vessel materials. The existing industrial-grade materials can hardly satisfy all the requirements, and thus fine-grained tungsten alloys by oxides or carbides dispersion strengthening are developed as plasma-facing materials (PFM). Reduced activation ferritic/martensitic (RAFM) steel, named CLF-1, and its oxides dispersion-strengthened (ODS) alloy are fabricated for the purpose of application in ITER-TBM (Tritium Breeder Modules) and possible application in CFETR (China Fusion Engineering Testing Reactor). Additionally, advanced structural material, V–4Cr–4Ti alloy, and its dispersion-strengthened alloys, such as V–4Cr–4Ti–1.5Y–0.3Ti3SiC2, are also developed. Besides plasma-facing materials and structural materials, another key in-vessel material is functional material, such as beryllium pebbles as neutron multiplier and Li4SiO4 pebbles as tritium breeder, and they are intensively studied and prepared. In this paper, the progress of W alloys as PFMs, CLF-1 (including ODS-CLF-1), and V–4Cr–4Ti alloy (including dispersion-strengthened V alloys) as structural materials, Be and Li4SiO4 pebbles as functional materials is overviewed by focusing on the manufacturing technology, characterization, service evaluation, for example, neutron irradiation data, high heat load performances, plasma irradiation effects, etc. Through this research it is hoped to guide the development of fusion reactor materials.