H.T. Lee

Recent progress of tungsten R&D for fusion application in Japan

The status of ongoing research projects of tungsten R&D in Japan is summarized in this paper. For tungsten material development, a new improved fabrication technique, the so-called superplasticity-based microstructural modification, is described. This technique successfully improved fracture strength and ductility at room temperature. Recent results on vacuum plasma spray W coating and W brazing on ferritic steels and vanadium alloys are explained. Feasibility of these techniques for the manufacture of the blanket is successfully demonstrated. The latest findings on the effect of neutron damage in tungsten on T retention and on the change in mechanical and electrical properties are described. Retention characteristics for neutron-damaged W were different compared to those for ion-damaged W. Upon neutron irradiation, tungsten alloys containing transmutation elements of W (Re and Os) show changes in properties that are different compared with those shown by pure W. The effects of mixed plasma exposure (D/He/C) are described. Both D/He and D/C mixed ion irradiations significantly affect ion-driven permeation in W. He bubble dynamics play a key role in nano-structure formation on the W surface.

Critical concentration for hydrogen bubble formation in metals.

Employing a thermodynamic model with previously calculated first-principle energetics as inputs, we determined the hydrogen (H) concentration at the interstitial and monovacancy as well as its dependence on temperature and pressure in tungsten and molybdenum. Based on this, we predicted the critical H concentration for H bubble formation at different temperatures. The critical concentration, defined as the value when the concentration of H at a certain mH-vacancy complex first became equal to that of H at the interstitial, was 24 ppm/7.3 GPa and 410 ppm/4.7 GPa at 600 K in tungsten and molybdenum in the case of a monovacancy. Beyond the critical H concentration, numerous H atoms accumulated in the monovacancy, leading to the formation and rapid growth of H-vacancy complexes, which was considered the preliminary stage of H bubble formation. We expect that the proposed approach will be generally used to determine the critical H concentration for H bubble formation in metals.

Ion-driven permeation of deuterium in tungsten by deuterium and carbon-mixed ion irradiation

Ion-driven permeation of deuterium (D) through tungsten (W) by D+C ion irradiation at 1?keV was studied. The changes in permeation flux as a function of temperature (Tw=550?1050?K) and carbon (C) fraction in the incident flux (fC~0.1?3%) are discussed. A temperature-dependent increase in the permeation flux under D?C ion irradiation was observed in comparison to D-only ion irradiation, with a value 200 times larger at 700?800?K. However, the measured C/W surface composition was temperature independent, meaning that the temperature-dependent increase in permeation flux was not directly correlated to the C on the W surface. The increase in permeation flux can be qualitatively interpreted as a reduction in the combination coefficient or diffusivity owing to tungsten carbide formation at the incident surface, but the temperature dependence lacks an adequate explanation. The increase in permeation flux was correlated to the increase in blister size observed on the W surface.

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.

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.

Comparison between helium plasma induced surface structures in group 5 (Nb, Ta) and group 6 elements (Mo, W)

Group 5 elements (niobium and tantalum) and group 6 elements (molybdenum and tungsten) were exposed to helium plasma, and the resulting surface structures were observed by electron microscopy. Group 5 elements showed hole structures, where the size of the holes ranged from several tens of nm to a few hundred nm in diameter, while group 6 elements showed fiber-like structures. As a first step in understanding such differences, the difference in helium agglomeration energies and changes in the stress tensor as a function of the number of He atoms at interstitial sites were investigated for each element using density functional theory. The calculations revealed that helium atoms prefer to agglomerate in both of these groups. However, helium in group 6 elements can agglomerate more easily than group 5 elements due to higher binding energy. These results indicate a possible correlation between the shape of helium plasma induced surface nanostructures and the atomic level properties due to helium agglomeration.Group 5 elements (niobium and tantalum) and group 6 elements (molybdenum and tungsten) were exposed to helium plasma, and the resulting surface structures were observed by electron microscopy. Group 5 elements showed hole structures, where the size of the holes ranged from several tens of nm to a few hundred nm in diameter, while group 6 elements showed fiber-like structures. As a first step in understanding such differences, the difference in helium agglomeration energies and changes in the stress tensor as a function of the number of He atoms at interstitial sites were investigated for each element using density functional theory. The calculations revealed that helium atoms prefer to agglomerate in both of these groups. However, helium in group 6 elements can agglomerate more easily than group 5 elements due to higher binding energy. These results indicate a possible correlation between the shape of helium plasma induced surface nanostructures and the atomic level properties due to helium agglomeration.

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.

Influence of helium on deuterium retention in reduced activation ferritic martensitic steel (F82H) under simultaneous deuterium and helium irradiation

Deuterium and helium retention in Japanese reduced activation ferritic martensitic (RAFM) steel (F82H) under simultaneous D–He irradiation at 500, 625, 750, and 818 K was studied. This study aims to clarify tritium retention behavior in RAFM steels to assess their use as plasma facing materials. The irradiation fluence was kept constant at 1 × 1024 D m−2. Four He desorption peaks were observed with He retention greatest at 625 K. At T > 625 K a monotonic decrease in He retention was observed. At all temperatures a systematic reduction in D retention was observed for the simultaneous D–He case in comparison to D-only case. This suggests that He implanted at the near surface in RAFM steels may reduce the inward penetration of tritium in RAFM steels that would result in lower tritium inventory for a given fluence.

Reflection properties of hydrogen ions at helium irradiated tungsten surfaces

Nanostructured W surfaces prepared by He bombardment exhibit characteristic angular distributions of hydrogen ion reflection upon injection of 1 keV H+ beam. A magnetic momentum analyzer that can move in the vacuum chamber has measured the angular dependence of the intensity and the energy of reflected ions. Broader angular distributions were observed for He-irradiated tungsten samples compared with that of the intrinsic polycrystalline W. Both intensity and energy of reflected ions decreased in the following order: the polycrystalline W, the He-bubble containing W, and the fuzz W. Classical trajectory Monte Carlo simulations based on Atomic Collision in Amorphous Target code suggests that lower atom density near the surface can make the reflection coefficients lower due to increasing number of collisions.

Deuterium Retention in Damaged Tungsten

Abstract In this research, thermal desorption characteristics of deuterium retained at trap sites of W created by irradiation of 300 keV hydrogen ions have been studied. With 10 hours of annealing, about 85% of deuterium was desorbed at temperatures of 300 °C and 350 °C, while deuterium desorption at 250 °C was about 60%. To estimate trapping energy of trap sites in this damaged W, TMAP simulation was performed. The result shows that the trapping energy of 1.29eV well accounted for the result of 250 °C annealing. In view that in the literature the vacancy trapping energy of hydrogen in tungsten was estimated to be close to 1.43 eV and the sensitivity analysis has given an uncertainty for the trapping energy of the order of 0.1 eV, it appears that the dominant trapping site type in the investigated damaged tungsten consists of vacancies.