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To evaluate the influence of He ions irradiation on the deuterium permeation and retention behavior in RAFM steels, samples made of the CLF-1 steel was irradiated with 3.5 MeV He ions. Gas driven permeation experiments were performed, and the permeability of virgin sample and pre-irradiated sample were obtained and compared. In order to characterize the effect of He ions irradiation on the deuterium retention behavior, deuterium gas exposure was carried out at 623 K, followed by thermal desorption spectra experiments. The total deuterium retention of the CLF-1 steel increased owing to He ions implantation, which could be attributed to the increase in trapping site for deuterium by the He pre-irradiation. [ABSTRACT FROM AUTHOR]

Experiments concerning deuterium gas-driven permeation through rolled tungsten foils in the temperature range of 850–950 K and subsequent deuterium retention have been performed. The steady state permeation flux of deuterium is proportional to the square root of the driving pressure. The permeability of deuterium is in an order of 10 −14 mol m −1 s −1 Pa −1/2 in this temperature range and the activation energy for permeation is 1.21 eV. Measurements of diffusivity are significantly affected by the driving pressure, which can be well explained by a saturable-trap model. Thermal desorption spectra of samples feature a single deuterium release peak at about 873 K. TMAP 4 modeling of this peak gives a detrapping energy of 1.70 eV, which fits the dissociation enthalpy of deuterium desorbing from the inner wall of vacancy clusters or pores in tungsten. [ABSTRACT FROM AUTHOR]

Our previous study has shown that an effective tritium permeation barrier (TPB) with Al 2 O 3 layer found could be obtained by oxidization of a Fe–Cr–Al ferritic steel. In this study, irradiation effects on deuterium permeation behavior through oxidized Fe–Cr–Al ferritic steel (OFFS) have been investigated by Au ions irradiation followed by deuterium gas driven permeation (GDP) experiments. The deuterium permeability of the irradiated and original OFFS samples has been obtained and compared. Oxide layer has been characterized by X-ray photoelectron spectroscopy (XPS) experiment and scanning electron microscopy (SEM) experiment. The defects in the oxide layer with and without Au ions irradiation have been characterized by Doppler broadening spectrometry of positron annihilation (DBS-PA) experiments. The deuterium permeation behavior of OFFS changed owing to the Au ions irradiation, which could be attributed to the increased density of vacancy-type defects. • Oxidized Fe–Cr–Al ferritic steel (OFFS) has been irradiated by Au ions. • The defects in OFFS have been characterized by DBS-PA experiments. • D 2 GDP experiments have been performed for OFFS. • Vacancy-type defects have influence on D 2 permeation behavior of OFFS. [ABSTRACT FROM AUTHOR]

• D was introduced into the CLF-1 steel using D ion irradiation, D plasma exposure in PREFACE and EAST separately. • D retention and release behavior of the CLF-1 steel was examined using TDS. • Different D retention and release behavior of the CLF-1 steel irradiated in three conditions were showed. Deuterium (D) retention behavior in a reduced activation ferritic/martensitic (RAFM) steel (CLF-1) is measured by thermal desorption spectrometry (TDS) after irradiation by D ion in an ion beam facility, exposure to D plasma in a linear ECR plasma generator, and exposure to D plasmas in the EAST tokamak separately. The D release behavior of the CLF-1 steel under the three-different D implantation condition is compared. Results indicate that D release behavior in the CLF-1 steel after the tokamak exposure experiments differs significantly from those after laboratory D ion implantation experiments or D plasma exposure experiments in a linear ECR plasma generator. This demonstrates that using low temperature plasmas in laboratory facilities alone is not sufficient in the study of hydrogen isotope retention behavior in a realistic tokamak environment. In order to simulate the plasma and material interaction in tokamak more felicitously, more work need to do to improve the linear plasma generator at present. Possible explanations have been proposed to explain the different D release behavior in the CLF-1 steel after D implantation in three different environment. [ABSTRACT FROM AUTHOR]

Owing to its excellent properties, high-entropy alloys (HEAs) have been regarded as a kind of promising candidate structural material for nuclear devices. For the safety concern, it is vital to figure out the compatibility of hydrogen isotopes with the HEAs. In this work, the hydrogen isotope permeation and retention behavior of a representative HEA, CoCrFeMnNi, has been mainly checked by a gas-driven permeation device and a thermal desorption spectra device. The hydrogen isotope permeation parameters have been obtained, which are quite similar to those of the 316L austenitic steels. The hydrogen isotope retention behavior has been compared with that for a reduced activation ferritic/martensitic steel, and the relatively high hydrogen isotope inventory in HEA has been explained by crystalline structure and high density of hydrogen trapping site. [ABSTRACT FROM AUTHOR]

Reduced activation ferritic/martensitic (RAFM) steels are the primary candidate structural materials for the blanket in future fusion power plants. In the recent years, a novel 9–12% Cr modified F/M steel named SIMP steel has been developed. The deuterium permeation behavior of the SIMP steel with and without oxide layer has been evaluated using a gas driven permeation (GDP) device. The permeability of original and oxidized SIMP steel has been derived and compared. In order to characterize the oxide layer, X-Ray Diffraction (XRD), Scanning Electron Microscopy (SEM), Energy-Dispersive X-ray Spectroscopy (EDS) experiments have been carried out. GDP results show that the SIMP steel has similar deuterium permeability with the CLF-1 steel. The deuterium permeability of the oxidized SIMP steel is lower than that of the original SIMP steel. Heat treatment during the oxidization process has limited influence on the deuterium permeability of the bare SIMP steel. The cross-section SEM pictures present that with the increase of the oxidation time, the thickness of the oxide layer is increased. The oxide layers are observed to be porous. The results of XRD and EDS show that the oxide layer mainly consists of chromium oxide and accumulation of chromium can be detected in the surface region. The formation of the chromium oxide on the surface of the SIMP steel could be the reason of the decrease of the deuterium permeability of the oxidized SIMP steel. • The deuterium permeation behavior of the SIMP steel with and without oxide layer was first investigated. • The microstructural change has been characterized by SEM, EDS and XRD. • The deuterium permeability of the oxidized SIMP steel was 4–8 times lower than that of original SIMP steel. • The oxide layers were detected to be porous, which might be the reason why the hydrogen permeability is reduced slightly. [ABSTRACT FROM AUTHOR]

Vanadium alloys are attractive candidate structural materials in future fusion reactors. To evaluate the effects of irradiation on the deuterium retention behavior in vanadium alloys, samples made of V-5Cr-5Ti were irradiated by 5.5 MeV carbon ions. Doppler broadening measurements of the positron annihilation radiation tests were carried out to investigate the defect properties in the irradiated samples. Then the irradiated and virgin samples were implanted with deuterium in an ECR (electron-cyclotron resonance) linear plasma device followed by thermal desorption spectroscopy experiments. It was found that the irradiation process introduced large density of vacancy-type defects and subsequently increased the deuterium retention in the V-5Cr-5Ti, which remains as a serious concern for the vanadium alloys in the application as the structural materials of the fusion blanket. [ABSTRACT FROM AUTHOR]

The permeation behavior of creep-resistant, Al 2 O 3 -forming HTUPS austenitic stainless steels was studied using a gas driven permeation (GDP) device. The steel samples were first thermal oxidized at air condition, followed by GDP experiments. The permeability and diffusion coefficients of oxidized samples and bare 316L steels were derived and compared. In order to characterize the oxide layer, X-ray photoelectron spectroscopy was performed. An oxide layer with a thickness of 200 nm which mainly consists of Al 2 O 3 was detected. [ABSTRACT FROM AUTHOR]

Permeation behavior of deuterium (D) through various tungsten (W) materials, including rolled, annealed, recrystallized W and ZrC/La 2 O 3 -dispersion strengthened W, has been studied. Experiments were performed using a gas-driven permeation facility in the temperature range of 700–1000 K. The permeability of D was not sensitive to the dislocations/vacancies or ZrC/La 2 O 3 phases in W, and the measured permeation values are comparable to Frauenfelder and Zahkarov’s data. The diffusivity of D, however, is microstructure dependent. The higher the density of defects was, the lower the effective diffusivity was measured. These results imply that dislocations/vacancies or small amounts of ZrC/La 2 O 3 phases in W will not influence the hydrogen isotopes steady state permeation flux through the first wall. [ABSTRACT FROM AUTHOR]

To evaluate the capability of Fe Cr Al ferritic steels as tritium permeation barrier in fusion systems, the oxidation behavior together with the permeation behavior of a Fe Cr Al steel was investigated. Gas driven permeation experiments were performed. The permeability of the oxidized Fe Cr Al steel was obtained and a reduced activation ferritic/martensitic steel CLF-1 was used as a comparison. In order to characterize the oxide layer, SEM, XPS, TEM, HRTEM were used. Al 2 O 3 was detected in the oxide film by XPS, and HRTEM showed that Al 2 O 3 in the α phase was found. The formation of α-Al 2 O 3 layer at a relatively low temperature may result from the formation of Cr 2 O 3 nuclei. [ABSTRACT FROM AUTHOR]

To understand the deuterium (D) bulk retention behavior in plasma facing materials under reactor-relevant conditions, tungsten (W) foils were irradiated with 122 MeV 20 Ne ions and then exposed to D 2 gas at 773 K. A quasi-homogeneous distribution of atomic displacement damage to 0.3 dpa within a depth of 50 μm was produced in W using an energy degrader in the irradiation chamber. Results of positron annihilation lifetime spectroscopy showed a long positron lifetime component of ∼400 ps appeared after irradiation, indicating the formation of vacancy clusters with up to 12 vacancies in W. Thermal desorption spectra showed a broad D desorption temperature range (730–1173 K) with a high release peak at ∼1010 K for the irradiated W specimens. The amounts of D retained in the irradiated W were significantly larger than the annealed ones, which could be attributed to the trapping effects of the vacancy clusters formed by cascade collisions and the clustering of mono-vacancies. [ABSTRACT FROM AUTHOR]

The influence of the transmutation element rhenium (Re) content with 3 and 25 wt.% on deuterium (D) permeation in tungsten (W) was investigated. It was found that the addition of Re increased D diffusion coefficient and decreased D permeability, and the D diffusion coefficient of W-3%Re was approximate to that of W-25%Re while the D permeability declined with the increase of Re content. The D retention behavior in the W-Re alloys have been studied after D plasmas exposure employing the Material and Plasma Evaluation System (MAPES) in Experimental Advanced Superconducting Tokamak (EAST). As for the higher D solution energy in W-Re alloys, the total D inventory in W was significantly decreased by Re addition. [ABSTRACT FROM AUTHOR]

Alumina is regarded as one of the most promising candidate tritium permeation barrier (TPB). Through thermal oxidization of Al-contained alloys Fe–Cr–Al, alumina layer with high tritium permeation reduction ability can be obtained. After bonding this kind of materials with structural materials such as reduced activation ferritic/martensitic (RAFM) steels, it can serve as TPB. In this work, efforts have been done to enhance the hydrogen isotope permeation reduction ability of the alumina layer on the Fe–Cr–Al ferritic steel by optimizing the oxidation process. The oxidation temperature of the Fe–Cr–Al ferritic steel is set to 973 K, which is lower than the final heat treatment temperature of the RAFM steel. Three different atmospheres have been employed for the oxidation process of the Fe–Cr–Al ferritic steel. Gas driven permeation (GDP) experiments have been performed to examine the deuterium permeability of the oxidized Fe–Cr–Al ferritic steel. The deuterium permeability of the Fe–Cr–Al ferritic steel oxidized in argon with 1700 ppm oxygen is 104 times lower than that of the RAFM steel at 823 K. The microstructure and chemical composition of the oxide layer of the Fe–Cr–Al ferritic steel oxidized in three different atmospheres has been clarified using Scanning Electron Microscopy (SEM) and X-ray Photoelectron Spectroscopy (XPS) methods. It is suggested that oxidizing the Fe–Cr–Al ferritic steel at reduced oxygen atmosphere could inhibit the growth of the iron oxide and chromium oxide in the oxide layer, thereby leading to a dense and compact alumina layer that has excellent hydrogen permeation reduction performance. [ABSTRACT FROM AUTHOR]