M. Oyaidzu

Surface modification and deuterium retention in reduced activation ferritic martensitic steels exposed to low-energy, high flux D plasma and D2 gas

Samples prepared from steels F82H and EUROFER97 were irradiated with 20 MeV W ions at 300 K to 0.54 displacements per atom at the damage peak. Damaged and undamaged samples were exposed at elevated temperatures both to deuterium plasma at ion energies of 60 and 200 eV to a fluence of ≈1026 D m−2 and to D2 gas at a pressure of 100 kPa. The surface modification after plasma exposure was examined by scanning electron microscopy and Rutherford backscattering spectroscopy. Deuterium depth profiles were determined by the D(3He, p)4He nuclear reaction. In damaged steels loaded with deuterium, deuterium decorates the damage profile and the D concentration decreases with increasing temperature. After exposure of the F82H steel to the D plasma W-enriched near-surface layers are formed. The effective concentration of W in the near-surface steel layer depends on plasma exposure conditions.

Deuterium retention in various toughened, fine-grained recrystallized tungsten materials under different irradiation conditions

Deuterium retention in two types of toughened, fine-grained recrystallized W (TFGR W-1.2 wt% titanium carbide (TiC) and TFGR W-3.3 wt% tantalum carbide (TaC)) was studied, compared to pure W. D plasma exposure was performed to a fluence of 1 × 1026 D m−2 at a temperature of 573 K, followed by retention measurement analysis by nuclear reaction analysis and thermal desorption spectroscopy (TDS). It is found that D retention in TFGR W is higher than that in pure W. This is because TFGR W has a high density of trapping sites with low trapping energy and dispersoid (TiC or TaC) may serve as additional trapping sites with high trapping energy. Different irradiation experiments (D ion beam implantation) were also conducted at sample temperatures of 473–873 K, followed by TDS. At higher sample temperature (> 700 K), D retention in TFGR W-3.3 wt% TaC is lower than that in TFGR W-1.2 wt% TiC. This may be due to different types of dispersoids.

Tritium retention in reduced-activation ferritic/martensitic steels

Abstract The effects of displacement damage, plasma exposure and heat loads on T retention in reduced-activation ferritic/martensitic (RAFM) steels were investigated by exposing the steels to DT gas at 473 K. Despite enormous change in surface morphology, T retention in the heat-loaded specimen was comparable with that in the unloaded specimen. The exposure to plasma resulted in a drastic increase in T retention at the surface and/or sub surface. However, the T trapped at the surface/subsurface was easily removed by maintaining the specimens in air at ~300 K. Formation of radiation-induced defects led to a significant increase in T retention, and T trapped in the defects was not removed at ~300 K. These observations suggest that displacement damages have the largest effects on T retention at ~473 K.

Tritium Trapping on the Plasma Irradiated Tungsten Surface

Abstract Tungsten (W) specimens previously exposed to deuterium (D) plasmas both in the TEXTOR tokamak and high flux linear plasma generator (LPG) were subsequently loaded with tritium at 573 K for 3 h. Retention of tritium in the near-surface W layer was examined by imaging plate technique. On the TEXTOR-plasma-exposed W surface, tritium was mainly trapped in carbon deposits. For LPG-plasma-exposed W specimens, tritium was trapped in defects created in the near-surface layer during the course of D plasma exposure.