Yasuhisa Oya

Comparison of deuterium retention for ion-irradiated and neutron-irradiated tungsten

The behavior of D retention for Fe2+-irradiated tungsten with a damage of 0.025–3 dpa was compared with that for neutron-irradiated tungsten with 0.025 dpa. The D2 thermal desorption spectroscopy (TDS) spectra for Fe2+-irradiated tungsten consisted of two desorption stages at 450 and 550 K, while that for neutron-irradiated tungsten was composed of three stages and an addition desorption stage was found at 750 K. The desorption rate of the major desorption stage at 550 K increased as the displacement damage increased due to Fe2+ irradiation increasing. In addition, the first desorption stage at 450 K was found only for damaged samples. Therefore, the second stage would be based on intrinsic defects or vacancy produced by Fe2+ irradiation, and the first stage should be the accumulation of D in mono-vacancy and the activation energy would be relatively reduced, where the dislocation loop and vacancy is produced. The third one was found only for neutron irradiation, showing the D trapping by a void or vacancy cluster, and the diffusion effect is also contributed to by the high full-width at half-maximum of the TDS spectrum. Therefore, it can be said that the D2 TDS spectra for Fe2+-irradiated tungsten cannot represent that for the neutron-irradiated one, indicating that the deuterium trapping and desorption mechanism for neutron-irradiated tungsten is different from that for the ion-irradiated one.

Deuterium trapping at defects created with neutron and ion irradiations in tungsten

The effects of neutron and ion irradiations on deuterium (D) retention in tungsten (W) were investigated. Specimens of pure W were irradiated with neutrons to 0.3 dpa at around 323 K and then exposed to high-flux D plasma at 473 and 773 K. The concentration of D significantly increased by neutron irradiation and reached 0.8 at% at 473 K and 0.4 at% at 773 K. Annealing tests for the specimens irradiated with 20 MeV W ions showed that the defects which play a dominant role in the trapping at high temperature were stable at least up to 973 K, while the density decreased at temperatures equal to or above 1123 K. These observations of the thermal stability of traps and the activation energy for D detrapping examined in a previous study (≈1.8 eV) indicated that the defects which contribute predominantly to trapping at 773 K were small voids. The higher concentration of trapped D at 473 K was explained by additional contributions of weaker traps. The release of trapped D was clearly enhanced by the exposure to atomic hydrogen at 473 K, though higher temperatures are more effective for using this effect for tritium removal in fusion reactors.

The deuterium depth profile in neutron-irradiated tungsten exposed to plasma

The effect of radiation damage has been mainly simulated using high-energy ion bombardment. The ions, however, are limited in range to only a few microns into the surface. Hence, some uncertainty remains about the increase of trapping at radiation damage produced by 14 MeV fusion neutrons, which penetrate much farther into the bulk material. With the Japan-US joint research project: Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), the tungsten samples (99.99 % pure from A.L.M.T., 6mm in diameter, 0.2mm in thickness) were irradiated to high flux neutrons at 50 C and to 0.025 dpa in the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL). Subsequently, the neutron-irradiated tungsten samples were exposed to a high-flux deuterium plasma (ion flux: 1021-1022 m-2s-1, ion fluence: 1025-1026 m-2) in the Tritium Plasma Experiment (TPE) at the Idaho National Laboratory (INL). First results of deuterium retention in neutron-irradiated tungsten exposed in TPE have been reported previously. This paper presents the latest results in our on-going work of deuterium depth profiling in neutron-irradiated tungsten via nuclear reaction analysis. The experimental data is compared with the result from non neutron-irradiated tungsten, and is analyzed with the Tritium Migration Analysis Program (TMAP) to elucidate the hydrogen isotope behavior such as retention and depth distribution in neutron-irradiated and non neutron-irradiated tungsten.

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.

Effect of neutron energy and fluence on deuterium retention behaviour in neutron irradiated tungsten

Deuterium (D) retention behaviours for 14 MeV neutron irradiated tungsten (W) and fission neutron irradiated W were evaluated by thermal desorption spectroscopy (TDS) to elucidate the correlation between D retention and defect formation by different energy distributions of neutrons in W at the initial stage of fusion reactor operation. These results were compared with that for Fe2+ irradiated W with various damage concentrations. Although dense vacancies and voids within the shallow region near the surface were introduced by Fe2+ irradiation, single vacancies with low concentration were distributed throughout the sample for 14 MeV neutron irradiated W. Only the dislocation loops were introduced by fission neutron irradiation at low neutron fluence. The desorption peak of D for fission neutron irradiated W was concentrated at low temperature region less than 550 K, but that for 14 MeV neutron irradiated W was extended toward the higher temperature side due to D trapping by vacancies. It can be said that the neutron energy distribution could have a large impact on irradiation defect formation and the D retention behaviour.

Hydrogen retention of JT-60 open divertor tiles exposed to HH discharges

Hydrogen retention in graphite tiles exposed to hydrogen discharges at the JT-60 open divertor has been investigated by means of thermal desorption spectroscopy (TDS). Most of the plasma facing area was covered with re-deposited layers of maximum thickness of about 70??m appearing at the inner divertor region. Major parts of retained hydrogen were thermally desorbed as hydrogen molecules with a peak temperature of around 970?K. Almost all the hydrogen atoms were retained homogeneously in the re-deposited layers with an averaged hydrogen concentration of ~0.03 in H/C, which is much smaller than the saturated hydrogen concentration (H/C = 0.4?1.0). Since the saturated hydrogen concentration in carbon materials decreases with increasing temperature, the re-deposited carbon layers are very likely subjected to higher temperatures during the discharges, which are supported by the higher release temperature of hydrogen in TDS. This result suggests that hydrogen retention can be significantly reduced with higher wall temperatures.

Hydrogen retention and carbon deposition in plasma facing components and the shadowed area of JT-60U

In JT-60U, erosion/deposition analyses of the plasma facing wall have shown that local carbon transport in the inboard direction was appreciable in addition to long-range transport. The total deposition and erosion rates in the divertor region were ~1 × 1021 C atoms s−1 and ~−6 × 1020 C atoms s−1, respectively. About 40% of the deposition in the divertor region likely originates from the main chamber wall. At the plasma facing surfaces of the divertor region, the highest hydrogen concentration in the (H + D)/C ratio and the retention amount were found to be ~0.13 and ~1 × 1023 atoms m−2, respectively. In the plasma-shadowed area underneath the divertor region with a vacuum vessel baking temperature of 420 K, redeposited layers of ~2 µm thickness were found with a high hydrogen concentration of ~0.75 in (H + D)/C, which was nearly the same level as that observed in JET. Large deuterium retention was also observed at the main chamber wall covered with boron layers. Their H + D retention and (H + D)/C were ~1 × 1023 atoms m−2 and ~0.16, respectively, for the vacuum vessel temperature of 570 K. Such a high deuterium retention is most likely caused by D retained in the boron layers. Nevertheless, the integration of this retention over the whole main chamber wall results in significant inventory and needs further investigation.

Trapping behaviour of deuterium ions implanted into tungsten simultaneously with carbon ions

The trapping behaviour of deuterium ions implanted into tungsten simultaneously with carbon ions was investigated by thermal desorption spectroscopy (TDS) and x-ray photoelectron spectroscopy (XPS). The D2 TDS spectrum consisted of three desorption stages, namely desorption of deuterium trapped by intrinsic defects, ion-induced defects and carbon with the formation of the C–D bond. Although the deuterium retention trapped by intrinsic defects was almost constant, that by ion-induced defects increased as the ion fluence increased. The retention of deuterium with the formation of the C–D bond was saturated at an ion fluence of 0.5×1022 D+ m–2, where the major process was changed from the sputtering of tungsten with the formation of a W–C mixture to the formation of a C–C layer, and deuterium retention as the C–D bond decreased. It was concluded that the C–C layer would enhance the chemical sputtering of carbon with deuterium with the formation of CDx and the chemical state of carbon would control the deuterium retention in tungsten under C+–D2+ implantation.

Function of Water Molecule for Tritium Behavior on the Water-Metal Boundary

Abstract In order to investigate the function of water molecule for tritium transport behavior on the water-metal boundary, a series of experiments of tritium permeation into humid atmosphere was performed through pure iron piping with different surfaces of oxide etc., which contained about 1 kPa of pure tritium gas at 423 K. Chemical forms of tritium permeated into water were monitored continuously under purging outer jacket by <1000ppm of water vapor in Ar. Observation of metal surfaces was also carried out by Secondary Electron Microscope (SEM) and X-Ray Diffraction (XRD) analysis. The results were compared with those permeated into pressurized liquid water at 423 K. The actual tritium permeation rate into Ar with <1000 ppm of water vapor was not clearly changed that into liquid water. In the vapor atmosphere, a magnetite layer did not grow on the surface clearly, and tritium permeation rate and chemical species (∼100% of HTO) through pure iron piping with mechanically polished surface were not changed drastically comparing with data with a magnetite surface. On the other hand, hydrogen gas (HT) fraction of tritium permeated into the outer jacket increased drastically in case of a gold plating surface.

Retention and replacement of hydrogen isotopes and isotope effect in SiC by H+ and D+ ion irradiation

The retention and isotope replacement behavior of hydrogen isotopes in SiC has been studied by means of the elastic recoil detection (ERD) technique. The protium ions or deuterium ions were implanted into SiC at room temperature up to almost the saturation. Successively, the ion source was switched to deuterium ions or protium ions and the samples were irradiated at the same fluence under the same experimental conditions, respectively. The mass balance equations were applied to analyze the retention curves and the curves of decay and uptake in the isotope replacement. The saturation concentrations of H and D have been found to be about 0.70 and 0.75, respectively. The detrapping cross sections of H and D implants by the D + 2 and H + 2 ion bombardments have been determined to be 3.2±0.3 x 10 -22 m 2 / D + and 2.6±0.2 x 10 -22 m 2 /H + , respectively. The magnitude of these detrapping cross sections from SiC is as large as that from graphite. These isotope differences are discussed in terms of the elementary processes included in the mass balance equations.