One-point calibration laser-induced breakdown spectroscopy for the quantitative analysis of EAST-like plasma-facing materials

Quantitative in-situ elemental analyses of the impure deposited layer on the plasma-facing components (PFCs) of magnetically confined fusion devices such as EAST are crucial and challenging. Laser-induced breakdown spectroscopy (LIBS)...

Reduced defect recovery in self-ion damaged W due to simultaneous deuterium exposure during annealing

Deuterium (D) plasma exposure during annealing of self-ion damaged tungsten (W) is shown to exhibit reduced defect recovery when compared to annealing without D plasma exposure. In these experiments, samples were first damaged with 20 MeV W ions. Next, samples were annealed either with or without simultaneous D2 plasma exposure. The simultaneous annealed samples were first decorated by D2 plasma at 383 K prior to ramping up to an annealing temperature of 473, 573, 673, or 773 K and held for 1 hour with concurrent plasma exposure. The vacuum annealed samples each had a corresponding temperature history but without D$_2$ plasma treatment. Finally, all samples were exposed to D2 plasma at 383 K to decorate any remaining defects. Nuclear reaction analysis (NRA) and thermal desorption spectroscopy (TDS) shows that the simultaneous plasma-exposed and annealed samples exhibited virtually no defect recovery at annealing temperatures of up to 673 K, and had higher D retention than found in the vacuum annealed samples. TDS results indicate that only the lowest detrapping energy defects recover at an 773~K anneal for the simultaneous plasma annealed samples, while the vacuum annealed samples showed defect recovery at all anneal temperatures. This experiment clearly demonstrates that D occupied defects can significantly reduce or eliminate defect annealing in W, and is consistent with the existence of synergistic plasma exposure/displacement damage effects in fusion-energy relevant plasma facing materials.

Comparative Analysis of Lithium First Wall Concepts for Tokamak with Reactor Technologies

When developing the stationary fusion reactor, an unresolved issue is the design of its intra-chamber plasma-facing elements. It has now become obvious that among the materials conventionally used for intra-chamber elements, there are no solid structural materials that would meet the requirements for the long-term operation under the effect of the flux of fusion neutrons (14 MeV) with a density of ~1014 cm–2 s–1 and the heat flux with a power density of 10–20 MW/m2. An alternative solution to this problem is the use of liquid metals as a plasma-facing materials, and, first of all, the use of lithium, which has a low atomic number (low charge number Z). Other easily-melting metals are also considered, which have higher Z number, but lower saturation vapor pressure than lithium. This will make it possible to create the long-lived, heavy-to-damage and self-renewing surface of the intra-chamber elements, which will not contaminate the plasma. The main ideas of the alternative concept of the intra-chamber elements can be formulated based on the comprehensive analysis of the problems and requirements arising during the development of intra-chamber elements of the stationary reactor, for example, the DEMO-type reactor. The article presents the analysis of the possible design of the lithium-coated intra-chamber elements and discusses the main ideas of the lithium first wall concept for the tokamak with reactor technologies.