Speaker
Description
The fuel in thermonuclear and hybrid fission/fusion reactors will be a deuterium-tritium (DT) mixture. Plasma-facing materials will be exposed to both stationary and high heat pulsed plasma loads, as well as helium and neutrons, arising from the DT reaction. Such extreme operating conditions can lead to degradation of materials due to erosion, formation of plasma- and radiation- induced defects, and hydrogen isotope embrittlement. To ensure the safe operation of a fusion reactor, it is necessary to be able to predict a stability of material properties under irradiation and an amount of radioactive tritium retained in materials.
Tungsten (W) is envisaged as reference plasma-facing material for ITER. Because tungsten oxidation and helium embrittlement are suppressed in W-Cr-Y alloy, this alloy can be an alternative choice for the first wall in future reactors. In this paper, the effect of dynamic change of surface impurities (B and O), temperature and development of plasma- and radiation- induced damage on D retention in W-based materials is discussed. In order to study stability of W and W-Cr-Y alloy under irradiation, radiation- and plasma- induced defects and surface impurities were characterized by positron annihilation spectroscopy, scanning and transmission electron microscopy, atomic probe tomography, and energy dispersive X-ray spectroscopy. D accumulation was studied by thermal desorption spectroscopy.
It is shown that plasma-induced damage, produced by high fluxes of particles and heat, leads to similar increase in the D concentration as radiation-induced damage. However, total D retention is determined by D retention in radiation defects because they distributed over all material thickness while plasma-induced damage occurs only near the irradiation surface. Coating of boron (B) increases D retention in the material. The binding energy of D to defects in B coating is higher than in W and W-Cr-Y alloy, and also higher than the binding energy of D to defects caused by high-temperature plasma. Although, the D concentration in radiation-induced damage is comparable with the D concentration in layers deposited together with boron and D, it is expected that the dominant contribution to the total D retention will be the D retention in radiation-induced defects due to the larger surface of the first wall and bulk retention. Radiation damage has less influence on the D retention in W-Cr-Y alloy compared to pure W. Moreover, the Cr and Y alloying elements suppress the formation of dislocation loops and pores. Modelling was used for predictions of the D retention for ITER.