Tritium decay helium-3 effects in tungsten

Tritium (T) implanted by plasmas diffuses into bulk material, especially rapidly at elevated temperatures, and becomes trapped in neutron radiation-induced defects in materials that act as trapping sites for the tritium. The trapped tritium atoms will decay to produce helium-3 (3He) atoms at a half-life of 12.3 years. 3He has a large cross section for absorbing thermal neutrons, which after absorbing a neutron produces hydrogen (H) and tritium ions with a combined kinetic energy of 0.76 MeV through the 3He(n,H)T nuclear reaction. The purpose of this paper is to quantify the 3He produced in tungsten by tritium decay compared to the neutron-induced helium-4 (4He) produced in tungsten. This is important given the fact that helium in materials not only creates microstructural damage in the bulk of the material but alters surface morphology of the material effecting plasma-surface interaction process (e.g. material evolution, erosion and tritium behavior) of plasma-facing component materials. Effects of tritium decay 3He in tungsten are investigated here with a simple model that predicts quantity of 3He produced in a fusion DEMO FW based on a neutron energy spectrum found in literature. This study reveals that: (1) helium-3 concentration was equilibrated to ∼6% of initial/trapped tritium concentration, (2) tritium concentration remained approximately constant (94% of initial tritium concentration), and (3) displacement damage from 3He(n,H)T nuclear reaction became >1 dpa/year in DEMO FW. Keywords: Tritium, Helium-3, Neutron-irradiation, Plasma facing-component

Deuterium retention and blistering in tungsten foils

To investigate deuterium retention and the onset of blistering, deuterium was implanted in cold rolled tungsten foils at fluences ranging from 3 ×1020 to 3 ×1022D/m2. Ion energies were 300eV and 2000eV in order to be below and above the tungsten theoretical damage energy threshold. While energy dependent phenomena were observed, blistering occurs regardless of ion energy. Both plastically and elastically deformed blisters were found, as manifest in before and after micrographs. The fraction of plastically deformed blisters did not saturate at the fluences used in these studies. However, the size of the largest blister that relaxed during TDS does saturate at ∼7µm. A simple conceptual model is presented, which proposes that the deuterium released from elastically deformed blisters appears at ∼600K in the thermal desorption spectra, which is consistent with large vacancy clusters. Keywords: Tungsten, Deuterium, Retention, Blistering, Thermal desorption spectroscop