H. Iwakiri

Microstructure evolution in tungsten during low-energy helium ion irradiation

In situ transmission electron microscopy (TEM) study was performed to investigate the microstructural changes in tungsten during low-energy He+ ion irradiations in an electron microscope linked with an ion accelerator. The irradiations were carried out with 8 and 0.25 keV He+ ions at 293, 873 and 1073 K. In the case of the 8 keV irradiation, irradiation-induced vacancies act as nucleation sites for dislocation loops and helium (He) bubbles. Accordingly, such defects were formed even at the higher temperatures. With increasing irradiation temperature, the growth rates of dislocation loops and He bubbles rise remarkably. Although no vacancies are produced during 0.25 keV irradiation, He platelets, interstitial loops and He bubbles were formed. Impurity atoms may act as trapping centers for He atoms, which form bubbles by ejecting W atoms from their lattice site.

Sub-ms laser pulse irradiation on tungsten target damaged by exposure to helium plasma

The effects of a transient heat load on tungsten damaged by helium plasma irradiation have been investigated using a ruby laser with long pulse duration in the divertor simulator NAGDIS-II (Takamura et al 2002 Plasma Sources Sci. Technol. 11 A42). The pulse width of the ruby laser was ∼0.6 ms, which is close to that of the expected heat load accompanied by type-I edge localized modes (ELMs) in ITER operation. Helium holes/bubbles, which were formed in the surface region of powder metallurgy tungsten due to the exposure to the helium plasma, disappeared after the laser pulse irradiation to the tungsten surface with sufficient pulse energy. The results indicated that the transient heat loads similar to those expected by ELMs will mitigate damages such as bubbles and holes produced by helium irradiation. When a vacuum plasma sprayed tungsten coating on graphite was exposed to the helium plasma, the surface was covered with arborescent nanostructured tungsten containing many helium bubbles inside the structure. Melting traces were found on the surface after the laser pulses irradiated the surface even though the pulse energy was lower than that for melting bulk tungsten. A numerical temperature calculation of the sample suggested that the effective thermal conductivity near the surface dramatically decreased by several orders of magnitude due to the formation of nanostructured tungsten.

Effects of helium bombardment on the deuterium behavior in tungsten

The effects of pre-irradiation with helium ions of fusion relevant energy on trapping of injected deuterium in W was studied by thermal desorption spectrometry technique using high-resolution quadrupole mass spectrometer. Pre-irradiation with He ions caused remarkable effects on the trapping of injected deuterium. Most of the injected deuterium is desorbed between 400 and 600 K for the case without helium pre-irradiation, while additional desorption occurs between 600 and 800 K for the helium pre-irradiation case. Total amount of the trapped deuterium for irradiations of 2.0×1021 He/m2 and 1.0×1022 D2/m2 is 6.2×1020 D2/m2, which is more than three times higher than that in the case of no helium pre-irradiation. The present result indicates that irradiation effects of He bombardment must be taken into account to understand and evaluate the behavior of hydrogen isotopes in fusion environment.

Suppression of blister formation and deuterium retention on tungsten surface due to mechanical polishing and helium pre-exposure

Low-energy deuterium (D) plasma exposure on tungsten (W), which is an important material for ITER, results in blister formation on the surface. Blister formation increases both micron-sized dust production and D retention. Blister formation depends greatly on surface pre-treatment. Deuterium plasma exposure on mirror-finished powder metallurgy W at 500 K for 3 h forms a blister with a diameter of a few hundred micrometres on the surface. Blister formations on the mechanically-polished and helium-pre-exposed surfaces are drastically suppressed. Deuterium retention is also reduced on both the mechanically-polished surface and the helium-pre-exposed surface compared with that on the mirror-finished surface. The suppressive effect of blister formation on the mechanically-polished surface is maintained for 50 h. The size of blisters and D retention on mirror-finished surface exposed for 50 h increases by some degrees (500 µm, 7 × 1020 m−2) compared with that on mirror-finished surface exposed for 3 h (200 µm, 5 × 1020 m−2), but is not proportional to the exposure time. The saturation level of D retention even on blister-rich surfaces seems to be lower than an order of 1021 D2 m−2.

Effects of dislocation on thermal helium desorption from iron and ferritic steel

Abstract Thermal desorption measurements were performed to investigate helium trapping in α-iron and a reduced activation martensitic steel (RAMS) bombarded at room temperature with mono-energetic He + ions. Incident energies were both 8 keV and 150 eV. Prior to helium implantation, the samples of α-iron were plastically deformed by rolling at room temperature followed by annealing at 673, 873 and 1073 K for 2, 12 and 2 h, respectively. These samples are hereafter called PR1, PR2 and FA, respectively, and the as-rolled sample is called CW. The dislocation densities of the samples decrease in the order of CW, PR1, PR2 and FA. Thermal desorption spectra show a clear peak at around 800 K, only for the CW, PR1, PR2 and RAMS samples, which may correspond to desorption of helium atoms trapped by dislocations. The effects of dislocations on micro-structural evolution in iron and the steel during helium implantation are discussed.

Microscopic damage of metals exposed to the helium discharges in TRIAM-1M tokamak and its impact on hydrogen recycling process

High-energy charge-exchange particles bombarding the plasma-facing wall may cause not only surface sputtering but also damage the materials inside due to their rather high energy. In the case of burning plasma, we should take into account the effects of helium because it is well known that helium atoms have much stronger effects on material damage than hydrogen atoms. In this paper, therefore, microscopic damage of metals exposed to long pulse discharges of helium plasma in TRIAM-1M was studied and the impact on the hydrogen recycling process was discussed by comparing it with helium ion irradiation experiments. Considerably larger amounts of dislocation loops and very dense fine bubbles were formed by the irradiation of rather low-energy charge-exchange neutrals of helium. According to the irradiation experiments with low-energy helium ions, formation of dense helium bubbles drastically enhances hydrogen trapping and makes the desorption difficult. These results indicate that the hydrogen recycling phenomenon during the burning plasma discharge must be quite different from that of the hydrogen plasma discharge experiments.

Effects of helium irradiation on chemical behavior of energetic deuterium in SiC

Structural changes of SiC caused by deuterium (D2+) and helium ion irradiation were studied with X-ray photoelectron spectroscopy (XPS) for ion energies in the range from 0.5 to 3.0 keV. Thermal desorption spectroscopy (TDS) measurements were performed to study the desorption processes from SiC surfaces pre-irradiated with deuterium and helium of 10 and 8.0 keV ion energy, respectively. For both ion species the range is approximately the same. The XPS experiments showed that the carbon concentration in the vicinity of the SiC surface was reduced owing to the helium irradiation. This suggests that carbon was selectively sputtered by the helium irradiation. In the TDS experiments, the D2 desorption was found to consist of three processes. The activation energy and the released amounts of deuterium for each peak were affected by helium pre-irradiation.

Studies on retention of tritium implanted into tungsten by β-ray-induced X-ray spectrometry

Trapping and diffusion of tritium implanted into tungsten at room temperature were examined by β-ray-induced X-ray spectrometry. One of two thin tungsten plates used was previously irradiated with He ions. After the tritium irradiation, X-ray spectra from a W sample were measured in an argon atmosphere, and changes in the X-ray spectra with time were followed for five months at room temperature. The observed X-ray spectra consisted of three characteristic X-ray peaks and a bremsstrahlung X-ray peak. For the W sample without He pre-irradiation all characteristic X-ray intensities decreased with time, while those for the sample with He pre-irradiation were almost constant, indicating that the He irradiation strongly affects the migration of tritium. Furthermore, the X-ray spectra were analyzed by computer simulation to estimate a tritium depth profile. From those results, the diffusion coefficient of tritium at room temperature and effects of He pre-irradiation on trapping of tritium were evaluated.

Studies of reflectivity degradation of retroreflectors in LHD and mitigation of impurity deposition using shaped diagnostic ducts and protective windows

Maintaining the reflectivity of first mirrors is indispensable in future fusion devices. While a retroreflector (corner cube mirror) is useful for laser diagnostics, impurities tend to accumulate and form a thick deposition layer in the central region, which causes degradation of reflectivity, due to the hollow shape of the retroreflector. Two mirror structures are tested to retain the reflectivity in the Large Helical Device (LHD). One is a bending mirror structure with a protective cylinder with fins and it could maintain the reflectivity over a three-month experimental campaign. The other is a cover window just in front of the reflector. Candidates of the window materials were exposed to the LHD plasmas and the degradation of the transmissivity of ZnSe and silicon, which are used for infrared and far infrared laser light, respectively, were small.

Super-saturated hydrogen effects on radiation damages in tungsten under the high-flux divertor plasma irradiation

Tungsten is a prime candidate as the divertor material of the ITER and DEMO reactors, which would be exposed to unprecedentedly high-flux plasmas as well as neutrons. For a better characterization of radiation damages in the tungsten under the divertor condition, we examine influences of super-saturated hydrogen on vacancies in the tungsten. The present calculations based on density functional theory (DFT) reveal unusual phenomena predicted at a super-saturated hydrogen concentration: (1) strongly enhanced vacancy concentration with the super-saturated hydrogen concentration is predicted by a thermodynamics model assuming multiple-hydrogen trapping, i.e. hydrogen clusters formation, in the vacancies; and (2) DFT molecular dynamics revealed that hydrogen clusters can prevent a vacancy from recombining with the neighboring crowdion-type self-interstitial-atom. This suggests that neutron damage effects will be increased in the presence of the hydrogen clusters.