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.     

Hydrogen implantation defects in MgO

Deuterium and hydrogen ions with an energy of 15 keV have been implanted in virgin MgO (1 0 0) single crystals and in single crystals containing helium implantation generated microcavities. Doses were varied from 2 × 10¹⁵ to 2 × 10¹⁶ cm⁻². The samples were annealed from room temperature to 950 K. The defects produced by hydrogen and the trapping of hydrogen at the defects were monitored by photon absorption and positron beam analysis. With this novel technique a depth distribution of defects can be determined for implantation depths from 0 to 2000 nm. The technique is very sensitive for vacancy and vacancy clusters, i.e. sites with low electron density. After 950 K annealing microcavities were observed for the 2 × 10¹⁶ cm⁻² dose but not for the 10 times lower dose. During annealing up to 750 K point defects are mobile but the defect clusters remain small and filled with hydrogen. In samples which contain already microcavities, point defects and deuterium from the deuterium irradiation are accumulated by the microcavities.     

Influence of surface condition on deuterium release from Li2TiO3 pebble

Lithium titanate (Li₂TiO₃) pebbles were irradiated with D₃⁺ ions with energy of 5.0 keV, and the amounts of retained deuterium in the pebbles were measured by thermal desorption spectroscopy. In this research the irradiation/heating cycles were carried out repeatedly in order to investigate the influence of surface condition on deuterium release from Li₂TiO₃. The composition ratio of Li decreased with the increase of the number of the irradiation/heating cycle. Then, the desorption peaks of the gases contained deuterium atoms were shifted to higher temperature region, and the amount of desorbed gases in forms of water tended to increase. In addition, we carried out other experiments for the comparison. Comparing these results, we considered that the increase of the defects created by the irradiation was more responsible for the change in the desorption behavior by the irradiation/heating cycles than the lithium depletion. These results suggest that the tritium recovery efficiency would decrease with the increase of the defects and the damages especially at the low temperature region during the operation.     

Deuterium diffusion in LiOH–water-corroded oxide layer of zirconium alloys

In order to clarify the hydrogen diffusion mechanism in the oxide layer of zirconium alloys, in situ hydrogen isotope diffusion in the oxide layer has been examined. The zirconium alloys used were Zircaloy-2, GNF-Ziron (Zircaloy-2 type alloy with high iron content) and VB (zirconium-based alloy with high iron and chromium contents). They were corroded in 1 or 0.1 M LiOH-containing water at 563 K, producing oxide layers of 1.1–2.1 μm in thickness. The diffusion experiments were carried out in the temperature range from 488 to 633 K by using a combined technique of deuterium plasma exposure and nuclear reaction analysis for D (³He,p)⁴He reaction. From the transient deuterium profiles in the oxide layers, it was concluded the LiOH–water-corroded oxides had a single-layer structure, which was in contrast to the double-layer structure previously observed in steam-corroded oxide layers. The diffusion coefficients in the 1 M LiOH–water-corroded oxides evaluated from the deuterium profiles were smaller in the order of Zircaloy-2 > GNF-Ziron > VB at 573 K. For the 0.1 M LiOH–water-corroded oxide of GNF-Ziron, the diffusivity was lower than that of the 1 M LiOH–water-corroded oxide by a factor of 1/4. The present diffusion coefficients of the 1 M LiOH–water-corroded oxides of GNF-Ziron and VB were approximately 7 times larger than the previous data of the corresponding steam-corroded oxides. The deuterium diffusion properties in the oxides of the three alloys obtained in the in situ experiment were roughly consistent with their hydrogen absorption performances in the LiOH–water-corrosion tests, as well as in the previous steam corrosion tests.     

Spectroscopic studies of homogeneous thin carbon erosion films on mirrors and flakes with a high deuterium content formed in tokamak T-10

Redeposited hydrocarbon films on plasma facing elements in tokamaks accumulate hydrogen isotopes. In the present study such films were made to redeposit on stainless steel mirror substrates as thin films and without any substrate as bare flakes with high deuterium content, under deuterium-plasma discharges inside T-10 tokamak vacuum chamber. These films were subjected to spectral characterizations through Fourier-transform infrared (FT-IR), electron paramagnetic resonance (EPR), and photoluminescence techniques. IR spectra showed the presence of two main deuterium states as observed by the CD_2,3 sp³ stretching modes at 2100–2200 cm⁻¹ and the CD₂ sp³ bending modes at 600–1100 cm⁻¹. Among these, CD₃ stretching mode at 2217 cm⁻¹ may serve as a control for deuterium desorption during the cleanup process of the reactor. As a comparative measure, C60 films were also studied, the luminescence excitation spectrum of which showed similarity in peak positions with tokamak bare flakes pertained to sp² luminescence centers. The observed spectral differences are mainly due to more localized sp² states for C60 and sp³ states for tokamak flakes. EPR spectra of the bare flakes showed the defective states with a high spin density, ∼10¹⁹ cm⁻³ which serve as luminescence quenching centers, and provide a path for hydrogen isotopes adsorption.     

Deuterium trapping in deep traps of differently oriented pyrolytic graphite exposed to D2 gas at 1473 K

Due to their importance for tritium inventories in future DT fueled fusion machines, experimental data on H isotope diffusion, absorption and retention in deep traps (E_b ≅ 4.3 eV) of graphites exposed to hydrogen at elevated temperatures have been reviewed. Deuterium retention was studied in edge- and basal-oriented pyrolytic graphite (PG) and polycrystalline RG-Ti-91 damaged by irradiation with 200 keV carbon ions. Deuterium loading was done by soaking in D₂ gas at 1473 K, and the resulting D retention was measured by nuclear reaction analysis. The microstructure was studied by cross-sectional TEM, SEM and microprofilometry. The concentration of strong traps created by irradiation and estimated by the amount of accumulated deuterium was shown to saturate with the damage above ≈1 dpa at about 1000 appm. In non-damaged and damaged graphites deuterium diffuses via porous grain boundaries and along basal planes within crystallites, while its migration through the graphite lattice along the c direction was found to be negligible. Radiation modifications of PG retard deuterium diffusion and decrease the rate of its chemical erosion by a factor of five. The amount of deuterium accumulated in strong traps in graphites is mainly influenced by their macro- and microstructure, while the degree of graphitization seems to be less important. Derivations are made of the susceptibility of damaged graphites, in particular, CFCs to the retention of hydrogen isotopes in deep traps.     

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.     

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.     

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.     

Dust particle formation due to interaction between graphite and helicon deuterium plasmas

The collection of dust particles using divertor simulation helicon plasmas has been carried out to examine dust formation due to the interaction between a graphite target and deuterium plasmas, which are planned to operate in the large helical device (LHD) at the Japanese National Institute for Fusion Science (NIFS). The collected dust particles are classified into three types: (i) small spherical particles below 400 nm in size, (ii) agglomerates whose primary particles have a size of about 10 nm, and (iii) large flakes above 1 μm in size. These features are quite similar to those obtained through hydrogen plasma operation, indicating that the dust formation mechanisms due to the interaction between a carbon wall and a plasma of deuterium, which is the isotope of hydrogen, is probably similar to those of hydrogen.     

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.     

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.     

Effect of He-pre-injection on dislocation loop formation and irradiation-induced segregation of Fe–12%Cr–15%Mn austenitic steel

The effect of He-injection on irradiation-induced segregation of aging treated Fe–12%Cr–15%Mn austenitic steels, which are candidate materials as the reduced radio-activation of structure material for nuclear and/or fusion reactors was investigated by using the 1250 kV high voltage electron microscope (HVEM) connected with an ion accelerator. The Fe–Mn–Cr steel has been irradiated at 573 K by three irradiation modes of single electron-beam irradiation, electron-beam irradiation after He-injection and electron/He⁺-ion dual-beam irradiation in a HVEM. Irradiation-induced segregation analyses were carried out by an energy dispersive X-ray analyzer (EDX) in a 200 kV FE-TEM with beam diameter of about 0.5 nm. Dislocation loops with strain contrast were formed during irradiation and the loop numbers density increased rapidly with irradiation dose for He-pre-injected specimens. Voids were not observed after irradiations with three irradiation modes up to 5.4 dpa at 573 K. Irradiation-induced segregations of Cr and Mn near grain boundary were observed in each irradiation condition, but the amounts of Mn segregation decreased in the cases of electron/He⁺-ion dual-beam irradiation compared with single electron-beam and electron-beam irradiation after He-injection conditions.     

On the study of catalytic membrane reactor for water detritiation: Membrane characterization

Tritium waste recycling is a real economic and ecological issue. Generally under the non-valuable Q₂O form (Q = H, D or T), waste can be converted into fuel Q₂ for a fusion machine (e.g. JET, ITER) by isotope exchange reaction Q₂O + H₂ = H₂O + Q₂. Such a reaction is carried out over Ni-based catalyst bed packed in a thin wall hydrogen permselective membrane tube. This catalytic membrane reactor can achieve higher conversion ratios than conventional fixed bed reactors by selective removal of reaction product Q₂ by the membrane according to Le Chatelier's Law.This paper presents some preliminary permeation tests performed on a catalytic membrane reactor. Permeabilities of pure hydrogen and deuterium as well as those of binary mixtures of hydrogen, deuterium and nitrogen have been estimated by measuring permeation fluxes at temperatures ranging from 573 to 673 K, and pressure differences up to 1.5 bar. Pure component global fluxes were linked to permeation coefficient by means of Sieverts’ law. The thin membrane (150 μm), made of Pd–Ag alloy (23 wt.%_Ag), showed good permeability and infinite selectivity toward protium and deuterium. Lower permeability values were obtained with mixtures containing non permeable gases highlighting the existence of gas phase resistance. The sensitivity of this concentration polarization phenomenon to the composition and the flow rate of the inlet was evaluated and fitted by a two-dimensional model.     

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.     

Improvement in hydriding property of LaNi4.8Al0.2 alloy encapsulated by SiO2 sol

LaNi_4.8Al_0.2 alloy particles encapsulated by SiO₂ matrix were prepared by the sol gel method. Scanning electron microscope (SEM) imaging was applied to determine the silica network outside the encapsulated alloy. The hydriding kinetics, pulverization and poisoning characteristics of LaNi_4.8Al_0.2 alloy were investigated before and after being encapsulated by silica. The results reveal that the hydriding properties of encapsulated alloy are excellent. The hydrogenation rate of encapsulated alloy is faster than that of the original alloy. The quantities of hydrogen stored by the encapsulated and original alloy are 169.3 Nml/g and 147.1 Nml/g, respectively. The LaNi_4.8Al_0.2 alloy particles are broken up into powder after 10 times hydrogen absorption/desorption cycles, while the encapsulated alloy do not show any breakdown after 30 times hydrogen adsorption/desorption cycles. The quantities of hydrogen absorbed by original alloy particles are less than 8.2 Nml/g in H₂-14.4% CO and 18.6 Nml/g in H₂-12.8% CO₂, while the quantities of hydrogen absorbed by encapsulated alloy agglomerations are 84.5 Nml/g in H₂-14.4% CO and 168.9 Nml/g in H₂-12.8% CO₂. These results clearly indicate that the pulverization and poisoning resistance properties of LaNi_4.8Al_0.2 alloy are evidently enhanced after being encapsulated by silica network.     

Electrical and thermo-mechanical analysis of beam recovery system for megawatt power gyrotron

The paper presents the electrical and thermo-mechanical design of single stage beam recovery system for 120 GHz, 1 MW gyrotron. The electrical study shows that the cylindrical shape single stage beam recovery system enhances the efficiency by 66.26%. The maximum power deposited to collector in depressed collector operation is 0.48 MW for electronic efficiency, 30% and 1.44 MW for DC electron beam. The thermo-mechanical analysis has been performed to evaluate the water cooling system. The cooling system has capability of accommodating a peak wall loading, 0.9 kW/cm² at flow rate of 1500 l/min for safe operating time, 60 ms. Further, a high voltage analysis is also carried out to appraise the electric field distribution in the collector.     

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.     

The impact of thermal fatigue and carbidization on the W coatings deposited on CFC tiles for the ITER-like Wall project at JET

Since August 2011 JET operates with the ITER-like wall comprising bulk Be tiles, bulk W tiles and W coated CFC tiles with a thickness of 10–15 μm and 20–25 μm. In order to evaluate behavior of the W coatings to a cyclic thermal loading relevant to JET operation, high heat flux (HHF) tests have been carried out up to 5100 pulses with an electron beam facility at peak temperatures of 1000 °C, 1250 °C and 1450 °C. The pulse duration was 24 s. Optical inspections of the W layer performed periodically by interrupting the test revealed small delaminations with the size of 50–500 μm. The dependence of the delamination percentage on the number of pulses can be seen as a degradation curve for each particular W coating. In this way the thermo-mechanical properties of the W coatings can be characterized quantitatively. Thermal fatigue and carbidization of the tungsten due to the diffusion of the carbon from the substrate have been recognized as mechanisms for degradation of the coatings. Tungsten carbides have been identified by using TEM (transmission electron microscopy) diffraction analysis on FIB (focused ion beam) prepared cross-section samples subjected to HHF tests. Nano-pores developed at the CFC–Mo and Mo–W interfaces during the tests might be also responsible for the degradation of the coating.     

A new facility for investigation on neutron irradiation effect on superconducting properties of Nb3Sn strand for fusion magnet

To perform post irradiation tests of superconducting strands, a 15.5 T superconducting magnet and a variable temperature insert (VTI) were installed at a radiation control area in Oarai center, Institute for Materials Research, Tohoku University. Both of the 15.5 T magnet and the VTI are conductively cooled with GM refrigeration. The commissioning test of the system is still ongoing because unexpected troubles occurred during the commissioning. Also, a SQUID system with the maximum field of 7 T has been installed at another radiation control area to investigate the magnetic property of uranium and its isotopes. These devices are very useful to study the electro-magnetic properties of the neutron irradiated superconducting strands. This paper will introduce the new superconducting test facility, and some data recently obtained will be presented together with the data of magnetization evaluated with the SQUID, and the discussion on the irradiation effect on superconducting properties will be performed.