Sorption and desorption behavior of hydrogen isotopes from tungsten deposits caused by deuterium gas or deuterium plasma exposure

To understand the behavior of hydrogen isotopes in deposits formed on plasma-facing wall is an important issue for development of a fusion reactor. In this study, sorption/desorption behaviors of hydrogen isotopes when tungsten deposits were exposed to deuterium gas or deuterium plasma at 300 °C were investigated. Samples of tungsten deposits were produced by the sputtering method using hydrogen plasma. After deuterium gas exposure or deuterium plasma exposure, the desorption behavior of hydrogen isotopes from the deposit was observed by the thermal desorption spectroscopy method. It was found that not a small amount of deuterium is retained in tungsten deposit by not only the plasma exposure but also the gas exposure while the amount of hydrogen incorporated in the deposit during sputter-deposition process is reduced. The amount of deuterium retained in the deposit by the plasma exposure was larger than that by the gas exposure in the experimental conditions in this work. The amount of hydrogen left after deuterium plasma exposure was larger than that after deuterium gas exposure.     

Deuterium and helium retentions of V–4Cr–4Ti alloy used as first wall of breeding blanket in a fusion reactor

The deuterium and helium retention properties of V–4Cr–4Ti alloy were investigated by thermal desorption spectroscopy (TDS). Ion energies of deuterium and helium were taken at 1.7 and 5 keV, respectively. The retained amount of deuterium in the sample irradiated at 380 K increased with the ion fluence and was not saturated to fluence of up to 1 × 10²³ D/m². For the irradiation at 773 K, 0.1% of implanted deuterium was retained at the highest fluence. For the helium ion irradiation at room temperature, three groups of desorption peaks appeared at around 500, 850, and 1200 K in the TDS spectrum. In the lower fluence region (<1 × 10²¹ He/m²), the retained helium desorbed mainly at around 1200 K. With increasing fluence, the amount desorbed at 500 K increased. Total amount of retained helium in the samples saturated at fluence up to 5 × 10²¹ He/m² and saturation level was 2.7 × 10²¹ He/m².     

Deuterium trapping by irradiation damage in tungsten induced by different displacement processes

The deuterium trapping behaviors in tungsten damaged by light ions with lower energy (10 keV C⁺ and 3 keV He⁺) or a heavy ion with higher energy (2.8 MeV Fe²⁺) were compared by means of TDS to understand the effects of cascade collisions on deuterium retention in tungsten. By light ion irradiation, most of deuterium was trapped by vacancies, whose retention was almost saturated at the damage level of 0.2 dpa. For the heavy ion irradiation, the deuterium trapping by voids was found, indicating that cascade collisions by the heavy ion irradiation would create the voids in tungsten. Most of deuterium trapped by the voids was desorbed in higher temperature region compared to that trapped by vacancies. It was also found that deuterium could accumulate in the voids, resulting in the formation of blisters in tungsten.     

Deuterium Retention and Physical Sputtering of Low Activation Ferritic Steel

Low activation materials have to be developed toward fusion demonstration reactors. Ferritic steel, vanadium alloy and SiC/SiC composite are candidate materials of the first wall, vacuum vessel and blanket components, respectively. Although changes of mechanical-thermal properties owing to neutron irradiation have been investigated so far, there is little data for the plasma material interactions, such as fuel hydrogen retention and erosion. In the present study, deuterium retention and physical sputtering of low activation ferritic steel, F82H, were investigated by using deuterium ion irradiation apparatus. After a ferritic steel sample was irradiated by 1.7 keV D^ ions, the weight loss was measured to obtain the physical sputtering yield. The sputtering yield was 0.04, comparable to that of stainless steel. In order to obtain the retained amount of deuterium, technique of thermal desorption spectroscopy (TDS) was employed to the irradiated sample. The retained deuterium desorbed at temperature ranging from 450 K to 700 K, in the forms of DHO, D2, D2O and hydrocarbons. Hence, the deuterium retained can be reduced by baking with a relatively low temperature. The fiuence dependence of retained amount of deuterium was measured by changing the ion fiuence. In the ferritic steel without mechanical polish, the retained amount was large even when the fluence was low. In such a case, a large amount of deuterium was trapped in the surface oxide layer containing O and C. When the fluence was large, the thickness of surface oxide layer was reduced by the ion sputtering, and then the retained amount in the oxide layer decreased. In the case of a high fluence, the retained amount of deuterium became comparable to that of ferritic steel with mechanical polish or SS 316 L, and one order of magnitude smaller than that of graphite. When the ferritic steel is used, it is required to remove the surface oxide layer for reduction of fuel hydrogen retention. Ferritic steel sample was exposed to the environment of JFT-2M tokamak in JAERI and after that the deuterium retention was examined. The result was roughly the same as the case of deuterium ion irradiation experiment.     

Hydrogen incorporation in tungsten deposits growing by deuterium plasma sputtering

Tungsten deposits were produced by sputtering method using hydrogen isotope RF plasma, and the density and the incorporated components in the deposits were investigated. The density changed in the range from 14.2 g/cm³ to 6.1 g/cm³, and hydrogen isotope retention changed in the range from 0.25 to 0.05 as (H + D)/W by the difference of deposition conditions. Both the density and hydrogen isotope retention tended to decrease with an increase of pressure. Even though a deuterium gas was used for producing tungsten deposits, not only deuterium but also hydrogen, oxygen and water vapor were incorporated in the deposits. It is considered that the incorporation of these components originated in water vapor unintentionally existing in the vacuum chamber.     

Measurement of deuterium and helium by glow-discharge optical emission spectroscopy for plasma–surface interaction studies

To examine the resolution of isotope analysis of hydrogen with glow-discharge optical emission spectroscopy (GDOES), depth profiles of hydrogen and deuterium in a H-containing Ta/D-containing Ti/Ni layered structure were measured. The depth profiles of deuterium could be measured with sufficient resolution in the presence of relatively large amounts of hydrogen and vice versa. In addition, measurements of depth profiles of He implanted in W at room temperature were also performed with Ne plasma. The intensity of the He emissions was sufficiently high at a fluence of 10²⁰ He m^?2 or higher. The depth profiles of He measured in this manner were in good agreement with the results of cross-sectional observations using a transmission electron microscope. Therefore, it was concluded that GDOES with Ne plasma is a promising technique for the depth profile analysis of plasma-facing materials and deposited layers formed on them.     

Effect of Ne Glow Discharge on Ion Density Control in LHD

Neon glow discharge cleaning was firstly attempted in Large Helical Device (LHD)instead of He glow discharge to remove hydrogen neutrals and to control the ion density n_i. TheNe glow discharge continued for 8 hours overnight after a three-day experiment. At the secondnight H_α emission became weaker than the emission usually observed in the He glow discharge. Aclear reduction of the hydrogen influx was also observed in neutral beam injection (NBI) dischargeswith Ne puff, whereas the neon recycling was strongly enhanced with appearance of a flat densityprofile. As a result, the lowest density limit was further reduced down to 0.2×10～(13) cm～(-3). Theuse of Ar puff formed a peaked density profile with a high T_i of 7 keV.     

Enhancement of hydrogen isotope retention capacity for the impurity deposited tungsten by long-term plasma exposure in LHD

The stress relieved tungsten samples were placed at three positions, PI (sputtering erosion dominated area), DP (deposition dominated area) and HL (Higher heat load area) during 15th plasma experiment campaign in Large Helical Device (LHD) at National Institute for Fusion Science (NIFS), Japan and were exposed to ～ 6700 shots of hydrogen plasma in a 15th long-term experiment campaign in LHD. Thereafter, the additional deuterium ion implantation to these tungsten samples was performed to evaluate the change of hydrogen isotope retention capacity in the samples by long-term plasma exposure. It was found that the carbon-dominant mixed-material layer with more than 100 nm thickness was formed on a wide area of the tungsten surface. The thicker mixed-material layer was formed on the DP sample, where the deuterium retention was about 21 times as high as that for pure W. The major desorption temperature of deuterium was shifted toward higher temperature side, which was comparable to the trapping characteristic of carbon or irradiation damages.     

Effect of induced damage on hydrogen isotope retention of F82H with impurity layer

The effects of induced damage on hydrogen isotope retention in F82H with or without thermal oxidation were investigated using thermal desorption spectroscopy. To induce damage and modify the surface, glow discharge pre-irradiated Ar⁺ ions was examined. In non-oxidized samples, the amount of desorbed deuterium increased with Ar⁺ ion fluence. Oxygen depletion in the surface layer of non-oxidized samples from the Ar⁺ ion irradiation, which resulted in bulk diffusion of deuterium, is responsible for the increase in deuterium retention. A comparison between non-oxidized and oxidized samples clearly indicated that the surface oxide layer greatly influenced deuterium retention/desorption behaviors of F82H.     

Influence of temperature and plasma composition on deuterium retention in refractory metals

Refractory materials are being considered potential candidates to build the first wall of the fusion reactor chamber. This work reports on the results of the study of tungsten and molybdenum metals exposed to high flux densities (～10²⁴ D/m² s) and low temperature (T_e ～ 3 eV) deuterium plasmas in Pilot-PSI irradiation facility.The hydrogenic retention in poly-crystalline W and Mo targets was studied with ³He nuclear reaction analyses (NRA). The NRA results clearly show a two-dimensional radial distribution of the deuterium with a minimum at the center and a maximum close to the edge. These distribution correlates well with the thermal profile of the sample surface, where a maximum of ～1600 K was measured at the center decreasing to ～1000 K in the edges. A maximum deuterium fluence retention of 5 × 10¹⁵ D/cm² was measured. The values of the retained fractions ranging from 10^?5 to 10^?6 D_retained/D_incident were measured with thermal desorption spectroscopy (TDS) and compares well with IBA results. Moreover, the presence of C in the plasma and its co-deposition increases the D retention in the region where a C film is formed. Both NRA and TDS results show no clear dependence of retention on incident fluence suggesting the absence of plasma related traps in W under these conditions.     

Overview of the US–Japan collaborative investigation on hydrogen isotope retention in neutron-irradiated and ion-damaged tungsten

The effect of neutron-irradiation damage has been mainly simulated using high-energy ion bombardment. A recent MIT report (PSFC/RR-10-4, An assessment of the current data affecting tritium retention and its use to project towards T retention in ITER, Lipschultz et al., 2010) summarizes the observations from high-energy ion bombardment studies and illustrates the saturation trend in deuterium concentration due to damage from ion irradiation in tungsten and molybdenum above 1 displacement per atom (dpa). While this prior database of results is quite valuable for understanding the behavior of hydrogen isotopes in plasma facing components (PFCs), it does not encompass the full range of effects that must be considered in a practical fusion environment due to short penetration depth, damage gradient, high damage rate, and high primary knock-on atom (PKA) energy spectrum of the ion bombardment. In addition, neutrons change the elemental composition via transmutations, and create a high radiation environment inside PFCs, which influences the behavior of hydrogen isotope in PFCs, suggesting the utilization of fission reactors is necessary for neutron-irradiation. Under the framework of the US–Japan TITAN program, tungsten samples (99.99 at.% purity from A.L.M.T. Co.) were irradiated by fission neutrons in the High Flux Isotope Reactor (HFIR), Oak Ridge National Laboratory (ORNL), at 50 and 300 °C to 0.025, 0.3, and 2.4 dpa, and the investigation of deuterium retention in neutron-irradiated tungsten was performed in the Tritium Plasma Experiment (TPE), the unique high-flux linear plasma facility that can handle tritium, beryllium and activated materials. This paper reports the recent results from the comparison of ion-damaged tungsten via various ion species (2.8 MeV Fe²⁺, 20 MeV W²⁺, and 700 keV H^?) with that from neutron-irradiated tungsten to identify the similarities and differences among them.     

Hydrogen isotope permeation from cooling water through various metal piping

In order to investigate the behavior of hydrogen isotope on the water–metal boundary, deuterium permeation experiments from heavy water vessel through various metal piping, such as pure iron (Fe), nickel (Ni), stainless steel (SS304), and pure iron with 10 μm gold plating, were performed at 573 K and at 15 MPa. During the experiment, surfaces of metal piping except gold plating one were oxidized at the heavy water boundary and then deuterium would generate by the oxidation reactions. This deuterium could be detected by mass spectrometer, which monitored the inside gases of the piping under continuous evacuation. The result showed clearly that the deuterium permeated through Fe, Ni, and SS304 piping was detected as mainly deuterium gas (D₂) under continuous evacuation, though that through gold plating one could not be detected effectively. The D₂ permeation rate through Fe, Ni, and SS304 piping reached equilibrium conditions with oxide generation at D₂O–metal boundary, although concluding the transfer mechanism will require further testing and modeling activities.     

Hydrogen isotope exchange in tungsten: Discussion as removal method for tritium

Hydrogen isotope exchange in re-crystallized polycrystalline tungsten was investigated at 320 and 450 K. In a first step the tungsten samples were loaded with deuterium to a fluence of 10²⁴ D/m² from a low-temperature plasma at 200 eV/D particle energy. In a second step, H was implanted at the same particle energy and similar target temperature with a mass-separated ion beam at different ion fluences ranging from 2 × 10²⁰ to 7.5 × 10²³ H/m². The analytic methods used were nuclear reaction analysis with D(³He,p)α reaction and elastic recoil detection analysis with ⁴He. In order to determine the D concentration at depths of up to 7.4 μm the ³He energy was varied from 0.5 to 4.5 MeV. It was found that already at an H fluence of 2 × 10²⁰ H/m², i.e. at 1/5000 of the initial D fluence, about 30% of the retained D was released. Depth profiling of D without and with subsequent H implantation shows strong replacement close to the surface at 320 K, but extending to all analyzable depths at 450 K especially at high fluences, leading to higher release efficiency. The reverse sequence of hydrogen isotopes allowed the analysis of the replacing isotope and showed that the release of D is balanced by the uptake of H. It also shows that hydrogen does not diffuse through a region of filled traps into a region were unfilled traps can be encounter but transport is rather a dynamic process of trapping and de-trapping even at 320 K. Initial D retention in H loaded W is an order of magnitude higher than in pristine W, indicating that every H-containing trap is a potential trap for D. In consequence, hydrogen isotope exchange is not a viable method to significantly enhance the operation time before the tritium inventory limit is reached but should be considered an option to reduce the tritium inventory in ITER before major interventions at the end of an operation period.     

Uptake of deuterium in partially D-depleted residual codeposits left on DIII-D tiles after laboratory thermo-oxidation experiments

Recent evidence has shown that tokamak carbon-based codeposits may become partially or fully depleted of hydrogen through thermo-oxidation, as the hydrogen content of the codeposits is removed more rapidly than the carbon content. In this study we examine the ability of such partially-depleted residual DIII-D divertor codeposits to uptake deuterium upon subsequent exposure to deuterium gas or deuterium plasmas. The partially D-depleted specimens used here were obtained from a previous study where DIII-D codeposits were oxidized for 2 h at 623 K (350 °C) and 267 Pa (2 Torr) O₂ [J.W. Davis et al., Thermo-oxidation of DIII-D codeposits on open surfaces and in simulated tile gaps, J. Nucl. Mater. 415 (2011) S789–S792]. In the present study some of these specimens, having undergone prior oxidation, were exposed to D₂ glow discharge plasmas or D₂ gas at 20 kPa (150 Torr) at 300 or 523 K. In the case of plasma exposure, no uptake of D was observed, while an increase in D content was seen following D₂ gas exposures. When the gas exposure took place at 300 K, heating the specimens in vacuum to 623 K for 15 min led to the release of all of the increased D content. For the gas exposure at 523 K, the increase in D content was found to require longer (8 h) vacuum baking to remove. However, in a reference codeposit specimen (from a closeby location on the tile), which had not been previously oxidized, there was a similar increase in D content following D₂ exposure at 523 K, but it could not be released even following 8 h vacuum baking at 623 K.     

Temperature dependence of surface morphology and deuterium retention in polycrystalline ITER-grade tungsten exposed to low-energy,high-flux D plasma

Surface topography and deuterium retention in polycrystalline ITER-grade tungsten have been examined after exposure to a low-energy (38 eV/D), high-flux (10²² D/m² s) deuterium plasma with ion fluences of 10²⁶ and 10²⁷ D/m² at various temperatures. The methods used were scanning electron microscopy equipped with focused ion beam, thermal desorption spectroscopy, and the D(³He,p) ⁴He nuclear reaction at ³He energies varied from 0.69 to 4.0 MeV. During exposure to the D plasma at temperatures in the range from 320 to 815 K, small blisters of size in the range from 0.2 to 5 μm, depending on the exposure temperature and ion fluence, are formed on the W surface. At an ion fluence of 10²⁷ D/m², the deuterium retention increases with the exposure temperature, reaching its maximum value of about 10²² D/m² at 500 K, and then decreases below 10¹⁹ D/m² at 800 K.     

Preparation of Cr2O3 film by MOCVD as hydrogen permeation barrier

Cr₂O₃ film on structural material as hydrogen permeation barrier can be applied in many areas such as hydrogen storage devices, vacuum solar receivers and fusion reactors. In this study, the Cr₂O₃ film was prepared by MOCVD on 316L stainless steel using chromium(III) acetylacetonate as precursor. The film was characterized by X-ray diffraction (XRD), scanning electron microscope (SEM) and X-ray photoelectron spectroscopy (XPS). The hydrogen permeation inhibition performance of films was investigated by deuterium permeation experiment. The 366 nm thick Cr₂O₃ film on 316L could reduce the deuterium permeability by 24–117 times at 823–973 K, revealing efficient inhibition to hydrogen permeation. The Cr₂O₃ film is dense, crack-free and has a corundum structure which possesses a more stable structure than a metastable phase or an amorphous phase. Moreover, the crystalline Cr₂O₃ could be easily obtained by MOCVD at a low temperature, e.g. 773 K.     

Nuclear micro-beam analysis of deuterium distribution in carbon fibre composites for controlled fusion devices

Probes made of carbon fibre composite NB41 were exposed to deuterium plasmas in the TEXTOR tokamak and in a simulator of plasma–wall interactions, PISCES. The aim was to assess the deuterium retention and its lateral and depth distribution. The analysis was performed by means of D(³He, p)⁴He and ¹²C(³He, p)¹⁴N nuclear reactions analysis using a standard (1 mm spot) and micro-beam (20 μm resolution). The measurements have revealed non uniform distribution of deuterium atoms in micro-regions: differences by a factor of 3 between the maximum and minimum deuterium concentrations. The differences were associated with the orientation and type of fibres for samples exposed in PICSES. For surface structure in the erosion zone of samples exposed to a tokamak plasma the micro-regions were more complex. Depth profiling has indicated migration of fuel into the bulk of materials.     

Radiation induced deuterium absorption for RB-SiC,HP-SiC,silicon and graphite loaded during electron irradiation

Absorption, diffusion, and desorption of hydrogen isotopes are expected to occur during operation in future fusion reactors and these processes will strongly depend on the irradiation conditions, neutron flux and purely ionizing radiation. The main aim of the work is to address the electron irradiation induced absorption of hydrogen isotopes in RB-SiC. Deuterium loading was carried out with both the sample and the surrounding deuterium gas exposed to 1.8 MeV electron irradiation in order to evaluate the radiation enhanced deuterium absorption. Thermo stimulated desorption (TSD) measurements were carried out for both electron irradiated and unirradiated samples in order to evaluate the possible radiation enhanced retention of the previously loaded deuterium. The materials subjected to the deuterium loading process were also studied by SIMS. Noticeable radiation enhanced deuterium absorption was observed. Most of the deuterium absorbed during irradiation was thermally released at about 600 °C.     

Low voltage hydrogen plasma interaction with palladium surface

Hydrogen penetration, distribution and content in palladium under low voltage glow discharge plasma, depending on plasma component structure (H₂, He, Ar and their mixtures), irradiation succession and possible surface structure changes have been investigated, using mass-spectrometry, SIMS, proton-proton scattering and electron microscopy techniques. Dependences of hydrogen permeability through palladium membranes upon the plasma component structure and the preliminary irradiation dose in the inert gas atmosphere have been revealed.     

Laboratory studies of H retention and LiH formation in liquid lithium

Laboratory experiments on H/D retention on liquid lithium followed by thermal desorption spectrometry (TDS) have been performed at Ciemat. Two different experimental set ups were used in order to expose liquid Li to hydrogen gas or to hydrogen glow discharge plasmas at temperatures up to 673 K. In the present work the results concerning the gas phase absorption are addressed. Two different kinetics of absorption were identified from the time evolution of the uptake. Alternate exposures to H₂ and D₂ were carried out in order to study the isotope exchange and its possible use for tritium retention control in Fusion Reactor. Although important differences were found in the absorption kinetics of both species, the total retention seems to be governed by the total sum of hydrogenic isotopes, and only small differences were found in the corresponding TDS spectra, on which evidence of some isotope exchange is observed. The results are discussed in relation to the potential use of liquid lithium walls in a Fusion Reactor.