Y. Torikai

Current status of nanostructured tungsten-based materials development

Nanostructured tungsten (W)-based materials offer many advantages for use as plasma facing materials and components exposed to heavy thermal loads combined with irradiation with high-energy neutron and low-energy ion. This paper first presents the recent progress in nanostructured toughened, fine grained, recrystallized W materials. Thermal desorption spectrometry apparatus equipped with an ion gun has been installed in the radiation controlled area in our Center at Tohoku University to systematically investigate the effects of displacement damage due to high-energy neutron irradiation on hydrogen isotope retention in connection with the nano- or micro-structures in W-based materials. In this paper, the effects of high-energy heavy ion irradiation on deuterium retention in W with different microstructures are described as a preliminary work with the prospective view of neutron irradiation effects.

Melt-layer ejection and material changes of three different tungsten materials under high heat-flux conditions in the tokamak edge plasma of TEXTOR

The behaviour of tungsten (W) plasma-facing components (PFCs) has been investigated in the plasma edge of the TEXTOR tokamak to study melt-layer ejection, macroscopic tungsten erosion from the melt layer as well as the changes of material properties such as grain-size and abundance of voids or bubbles. The parallel heat flux at the radial position of the exposed tungsten tile in the plasma ranges around q|| ~ 45 MW m−2 causing samples to be exposed at an impact angle of 35° to 20–30 MW m−2. Locally the temperature reached up to 6000 K, high levels of evaporation and boiling are causing significant erosion in the form of continuous fine spray or droplet ejection. The amount of fine-spray tungsten emission depends strongly on the material properties: in the case of the tungsten–tantalum alloy the effect of spraying and droplet emission is significantly higher at even low temperatures when compared with regular tungsten or even ultra-high purity tungsten which shows almost no spraying at all. Differences in the material composition, grain structure and size may be related to the different evolution of macroscopic erosion. In addition the re-solidified material is studied and strong differences in terms of re-crystallized grain size and evolution of the grain structure and grain orientation are observed. The build up of large voids has been observed.

Screen Test of Tritium Recovery from Stainless Steel Type 316

ABSTRACT Various techniques, such as wiping, gas-purge at elevated temperatures, heating by flame simultaneously with air-purge, have been tested for decontamination of stainless steel samples exposed to tritium-containing hydrogen at room and elevated temperatures. Effectiveness of the decontamination techniques is compared depending on the history of the sample loading with tritium, treatment after the loading, the sample thickness, etc.

New technique for non-destructive measurements of tritium in future fusion reactors

To establish a new technique for monitoring and control of high-level tritium in a fuel circulation system of the fusion reactors, the applicability of a new technique to elemental tritium and tritiated water was discussed. The new technique is based on the utilization of x-rays induced by β-rays from tritium, and it is called β-ray-induced x-ray spectrometry. The applicability of the same technique to the tritium species retained in solid materials was also discussed. It was concluded from applications to gaseous and aqueous tritium under the static state that the new technique is applicable as an in-line monitor in the fuel circulation system. In addition, it was also shown that the new technique plays an important role in non-destructive evaluation of tritium retained in solid materials such as plasma-facing materials.

Effect of water vapor on tritium decontamination of stainless steel 316

ABSTRACT To establish efficient decontamination methods for tritium-contaminated stainless steels, the desorption of tritium was studied for SS-316 in dry and wet argon gas carriers at different temperatures. The specimen was exposed to tritium at 523 K for 3 hours. The tritium inventory was in a range from 2 to 12 MBq. The desorption at elevated temperatures was measured by using a liquid scintillation counter. In addition, the tritium depth profiles in the specimen were evaluated by β-ray induced X-ray spectrometry. It was found that the decontamination efficiency was enhanced by the presence of moisture. This effect was ascribed to the isotope exchange reaction on the surface. The depth profile measurements revealed the presence of tritium-rich subsurface layer and the bulk with lower tritium concentration. The majority of tritium was, however, found in the bulk, indicating that extraction of bulk tritium is essential for the decontamination of SS-316 exposed to tritium at high temperatures.

Surface erosion and modification of toughened, fine-grained, recrystallized tungsten exposed to TEXTOR edge plasma

In order to evaluate the applicability of toughened, fine-grained, recrystallized (TFGR)-W to tokamak edge plasma environment, two TFGR-W specimens (TFGR-W 1.1wt%TiC and TFGR-W 3.3wt%TaC) were exposed to 31 identical ohmic discharges in the TEXTOR tokamak by means of a limiter lock system. The highest surface temperature reached was about 1300 °C. Under these temperature conditions, the bulk microstructure and dispersoids distribution of both TFGR-W remained intact, suggesting that these TFGR-W tungsten materials have sufficient stability under these plasma loading conditions. The erosion of TiC dispersoids on the surface was enhanced by plasma exposure above 1150 °C, while such enhanced erosion was not observed for TaC dispersoids probably due to the higher melting temperature of Ta than Ti.

Distribution and Mobility of Tritium in Type 316 Stainless Steel

Abstract Exposure of Type 316 stainless steel to tritium-containing hydrogen at an elevated temperature causes diffusion of the majority into the bulk and trapping of a small fraction in a thin oxide layer on the surface at concentrations far exceeding those in the bulk. The uptake by the bulk and surface layer is temperature and pressure dependent. After chemical erosion of the tritium-rich layer, the concentration of tritium on the topmost surface is slowly and asymptotically restored even at 298 K. Isothermal heating at 373 or 473 K until substantial release of the bulk tritium is associated with a comparatively much smaller liberation from the surface layer suggesting different retention and liberation mechanisms. The tritium inventory and profile evolution of homogeneously loaded Type 316 stainless steel caused by chronic release at the ambient temperature and radioactive decay were followed experimentally over several years and modeled successfully by a diffusion mechanism. The model has been adapted to specimens nonhomogeneously loaded with tritium only up to the subsurface. It simulates profile and inventory changes well even after prolonged aging. Chronic release constitutes an aging loss of tritium comparable to that of radioactive decay that should be taken into account for the assessment of tritium-contaminated stainless steel waste.

Tritium trapping behavior in tungsten pre-irradiated with D, He, Ar and N plasmas

Tritium (T) trapping in tungsten after plasma exposure (deuterium (D), helium (He), nitrogen (N), argon (Ar)) was studied using an imaging plate technique. Specimens were exposed to D and T mixed gas at 77 and 573 K to distinguish T trapped at the outermost surface and several tens of nanometers, respectively. He followed by N, Ar and D plasma exposures resulted in the largest increase in T trapping. This was interpreted to result from He bubble layers that can increase the surface area by formation of pores connected to the surface and/or an increase in surface trapping sites. T exposure at 77 K was found to be a very useful method to observe plasma-induced changes to T trapping at the outermost surface.

Contamination of stainless steel type 316 by tritium

ABSTRACT Tritium distribution in stainless steel type 316 exposed to hydrogen containing 32% of tritium at room and elevated temperatures was studied using thermal desorption, analysis of bremsstrahlung spectrums and acid etching techniques. All samples exhibit a large fraction of the overall tritium inventory concentrated in a thin sub-surface layer of ≈15µm thickness, where tritium concentration is by ≈2 order of magnitude larger than that in the bulk. Observed tritium depth profiles are in contradiction with a classical mechanism of hydrogen penetration to metals by atomic diffusion.

Tritium Interaction with Surface Layer and Bulk of Type 316 Stainless Steel and Consequences of Aging

Abstract Stainless steel exposed to gaseous tritium characteristically shows a firmly trapped fraction of tritium in the surface layer, which is not fully removable by water at ambient temperature. Prolonged thermal treatment of tritium-loaded specimens at <443 K causes substantial depletion of the bulk but almost no depletion of the surface layer. For complete removal of hydrogen isotopes from the bulk and the surface, temperatures exceeding 573 K are necessary. Upon chemical etching virtually all tritium trapped in the surface layer appears in the etching solution as tritiated water. Following removal of the layer by chemical etching, the tritium-rich layer reappears after months of aging at ambient temperature with nearly the original tritium activity. Comparison of chronic tritium release rates into liquid water before and after etching reveals that the surface layer only marginally influences the rate. X-ray photoelectron spectroscopy provides evidence that during prolonged aging the surface layer continues to grow while at the same time trapping a fraction of bulk tritium released at ambient temperature. Experimental results suggest different mechanisms of hydrogen uptake and release by the bulk and surface layers. Inference of tritium activity in the bulk of aged or heat-exposed stainless steel material from surface activity measurements may depart significantly from reality.