Deuterium Retention Behavior Research Articles

Influence of dynamic annealing of irradiation defects on the deuterium retention behaviors in tungsten irradiated with neutron

Tungsten (W) samples were damaged by neutron and 6.4 MeV Fe-ion irradiation above 1000 K simulating the divertor operation temperature. Deuterium (D) retention properties were examined by decorating the damaged W with D and subsequent thermal desorption spectroscopy (TDS) measurements. Vacancy clusters were the major D trapping site in the W irradiated with Fe-ion at 873 K, although D retention by vacancy clusters decreased in the W irradiated with Fe-ion at 1173 K due to dynamic annealing. The D de-trapping activation energy from vacancy clusters was found to be 1.85 eV. D retention in neutron damage W was larger than that damaged by Fe-ion due to the uniform distribution of irradiation defects. The D desorption behaviors from neutron damaged W was simulated well by assuming the D de-trapping activation energy to be 1.52 eV.

Effect of C-He simultaneous implantation on deuterium retention in damaged W by Fe implantation

Deuterium (D) retention behaviors for the 3 keV Helium (He+) implanted damaged-Tungsten (W) and 10 keV Carbon (C+) - 3 keV He+ simultaneous implanted damaged-W were evaluated by thermal desorption spectroscopy (TDS) to understand the synergetic effect of defect formation and C/He existence on D retention behavior for W with various damage level. For the He+ implantation, the retention of D trapped by dislocation loops was controlled by 3 keV He+ fluence. The D retention in the deeper region was reduced by He+ implantation with higher He+ fluence due to the formation of He bubbles and dense defects at the surface region which would reduce the effective D diffusion coefficient. In addition, in the case of the simultaneous C+ - He+ implantation, the reduction of D retention trapped in the deeper region was also found by the higher C+ - He+ fluence. It can be said that D retention behavior was controlled by the formation of He induced defects and accumulation of He near the surface even if the damages were introduced in the deeper region.

Surface or bulk He existence effect on deuterium retention in Fe ion damaged W

To evaluate Helium (He) effect on hydrogen isotope retention in tungsten (W), He+ was introduced into W bulk by 201 – 1000 keV He+, or W surface by 3 keV He+ for 6 MeV Fe ion damaged W at room temperature. The deuterium (D) retention behavior was evaluated by thermal desorption spectroscopy (TDS). In addition, the amount of tritium (T) at surface and bulk were separately evaluated by beta-ray induced X-ray spectroscopy (BIXS). The experimental results indicated that the formation of He-void complexes reduced the D trapping in vacancies and voids which have higher trapping energy by the bulk He retention. The BIXS measurement also supported the He enhanced the D reduction in the W bulk region. On the other hand, the He ion irradiation near the surface region enhanced D trapping by dislocation loops or surface, indicating the existence of He near surface interfered the D diffusion toward the bulk. It was concluded that the He existence in bulk or surface will significantly change the D trapping and diffusion behavior in damaged W.

Effects of high-energy C ions irradiation on the deuterium retention behavior in V-5Cr-5Ti

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.

Deuterium retention behavior in simultaneously He+–D2+ implanted tungsten

Poly-crystalline tungsten (W) samples were simultaneously irradiated with Helium (He) and Deuterium (D) ions using the triple-ion implantation device. He effect on D retention and transportation was studied using different combination of ion energies and He/D flux ratios in the simultaneous implantation. The experimental results show that D trapping at dislocation loops is significantly reduced in the case of 3 keV He+–3 keV D2+at He/D flux ratios over 0.6. D trapping by stronger trapping sites such as vacancies and vacancy clusters showed less dependence on the flux ratio. On the contrary, the D retention increases at each He/D flux ratio in the case of 3 keV He+–1 keV D2+compared to only D2+ implantation even the He/D flux ratio reaches a value of 1.0. TEM observations confirmed that dense dislocation loops are formed rather than He bubbles, which is responsible for the enhanced D retention in W.

A Study on Deuterium Retention Behavior of Plasma-Facing Materials for EAST

To investigate the deuterium (D) retention behavior of plasma-facing materials (PFMs) in the Experimental Advanced Superconducting Tokamak (EAST), SiC-coated graphite, bulk W, and W coating were exposed to D plasmas employing the material and plasma evaluation system in EAST. A comparative study was performed in an electron cyclotron resonance (ECR) linear plasma facility as well. For the EAST experiments, the local electron temperature and density were $T_{e}\approx$ 10–15 eV and $n_{e}\approx$ (1–3) $\times$ 1018 m−3, respectively. In the ECR facility, the plasma parameters were measured to be $T_{e} \approx 2.4$ eV and $n_{e} \approx 1.5 \times 10^{17}$ m−3. The surface morphology was examined by scanning electron microscopy, and D retention properties of various PFMs were examined by thermal desorption spectroscopy. No significant modification of the surface morphology was observed after tokamak exposure experiments. Bulk W and W coating exhibit similar D retention behavior, while D inventory in SiC-coated graphite has been found to be much larger than that in W materials.

Deuterium retention behavior of pure and Y2O3-doped tungsten investigated by nuclear reaction analysis and thermal desorption spectroscopy

Pure and Y2O3-doped tungsten samples were simultaneously exposed to deuterium (D) plasma. The following exposure parameters were investigated: ion energy of 38 eV/D, sample temperatures of 370, 450 and 570 K, and incident fluences of 6 × 1024 and 7.5 × 1025 D/m2. The deuterium retention behavior of these tungsten materials was investigated by nuclear reaction analysis and thermal desorption spectroscopy. After exposure at 450 K Y2O3-doped tungsten shows a D depth profile comparable with that of pure tungsten. But at each investigated exposure condition Y2O3-doped tungsten shows a higher total D inventory than pure tungsten. This is attributed to the presence of intrinsic defects with high trapping energies. Investigation of the D retention behavior of four different Y2O3-doped tungsten materials, which contain the same amount of Y2O3, but have various microstructures, suggests that sub-micron pores are the underlying intrinsic defects responsible for the appearance of a high temperature D release peak.

Deuterium retention and release behaviours of tungsten and deuterium co-deposited layers

Tungsten (W) layer deposited in argon and deuterium atmosphere by magnetron sputtering was used as a model system to study the deuterium (D) retention and release behavior in co-deposited W layer. After deposition several selected samples were exposed in deuterium plasma at 370 K with a flux of 4.0 × 1021 D/(m2 s) up to a fluence of 1.1 × 1025 D/m2. Structures of co-deposited W layers are investigated by field-emission scanning electron microscopy (FESEM) and X-ray diffraction (XRD), and the corresponding D retention and release behaviors are studied as functions of deposition and exposure parameters using thermal desorption spectroscopy (TDS). Two main D release peaks were detected from TDS spectra located near 600 and 800 K in these W and D co-deposited layers, and total deuterium retention increased linearly as a function of W layer's thickness. After deuterium plasma exposure, the total D retention amount in W layer increases significantly and D release peak shifts to lower temperature. Clearly, despite the high density of defects expected in co-deposited W layers, the initial deuterium retention before exposure to the deuterium plasma is low even for the samples with a W&D layer. But due to the high densities of defects, during the deuterium plasma exposure the deuterium retention increases faster for co-deposited layer than for the bulk W sample.

Impact of Annealing on Deuterium Retention Behavior in Damaged W

The annealing effects on deuterium (D) retention for 0.1–1.0 dpa iron (Fe) ion damaged W were studied as a function of annealing duration. The D2 spectra for Fe damaged W with lower defect concentration showed that D trapped by vacancy clusters was clearly decreased as increasing annealing duration due to the recovery of vacancy clusters. On the other hand, at higher defect concentration, the desorption peak of D trapped by voids was shifted toward higher temperature side, which would be caused by aggregation of vacancies and vacancy clusters. It can be said that the recovery and aggregation behavior of defects are controlled by defect concentration. By disappearing of desorption of D trapped by vacancy clusters after annealing for longer duration, the desorption of D trapped by vacancies was increased, which could be explained by following two possibilities. One is that the retention of hydrogen isotope trapped by monovacancy was increased. The other is that number of vacancies during annihilation process of vacancy cluster were formed by annealing.

Deuterium retention in RAFM steels after high fluence plasma exposure

Deuterium retention and detrapping behavior in the ferritic-martensitic steels EUROFER’97 and P92 after exposure to plasma at high fluences ≥ 1026D/m2 was studied using thermal desorption spectroscopy (TDS), supported by nuclear reaction analysis. Low-temperature irradiation at 450K and fluences ≥ 1026D/m2 with low impact energy D+ / D++He+ ions of 40eV at PSI-2 resulted in a deuterium inventory of 7–18 ×1019D/m2 predominantly at depths ≥8.6µm. Helium admixture led to a reduction of total D retention in both steels, irrespective of surface erosion and composition. The deuterium spectra of both steels displayed one D2 desorption peak at ∼ 540–570K and HD maxima at 540–590, 700–730 and 900–930K. It is suggested that deuterium is mostly retained in the bulk of steel material on interfaces of carbide precipitates and on grain boundaries.

Effect of sequential Fe2+ − C+ implantation on deuterium retention in W

Deuterium (D) retention behavior for the sequential 6MeV iron (Fe) and 10keV carbon (C) implanted tungsten (W) were evaluated by thermal desorption spectroscopy (TDS) and β-ray-induced X-ray spectroscopy (BIXS) to understand the synergetic effect of defect formation and C existence on D retention behavior for W under various damage distribution profiles. The experimental results indicated that retention of D trapped by dislocation loops was controlled by 10keVC+ implantation. The D retention was reduced in the sequential Fe2+ − C+ implanted W with higher C+ fluence in comparison to that with lower C+ fluence due to the formation of C-W layer which suppressed D diffusion toward the bulk and dense defects at the surface which reduce effective D diffusion coefficient. On the other hand, the amount of D trapped by the defects in the deeper region than C+ implantation region (50nm) was increased due to the formation of dense defects by 6MeV Fe2+ implantation within the depth of 1.5μm.

Tungsten nitride coatings obtained by HiPIMS as plasma facing materials for fusion applications

In this work, tungsten nitride coatings with nitrogen content in the range of 19–50 at% were prepared by reactive multi-pulse high power impulse magnetron sputtering as a function of the argon and nitrogen mixture and further exposed to a deuterium plasma jet. The elemental composition, morphological properties and physical structure of the samples were investigated by Rutherford backscattering spectrometry, atomic force microscopy and X-ray diffraction. Deuterium implantation was performed using a deuterium plasma jet and its retention in nitrogen containing tungsten films was investigated using thermal desorption spectrometry. Deuterium retention and release behaviour strongly depend on the nitrogen content in the coatings and the films microstructure. All nitride coatings have a polycrystalline structure and retain a lower deuterium level than the pure tungsten sample. Nitrogen content in the films acts as a diffusion barrier for deuterium and leads to a higher desorption temperature, therefore to a higher binding energy.

Deuterium retention and desorption behavior in an advanced reduced-activation alloy

We present the first experimental results of the deuterium retention and desorption behavior in an advanced reduced-activation alloy (ARAA) under development in Korea. For the in-situ measurement of desorbed gases from samples immediately after irradiation, a thermal desorption spectroscopy (TDS) system clustered with an inductively coupled plasma ion source has been built. Samples were and were not irradiated with helium ions at energies of 1.4, 3.5, and 5.0 keV and then continuously irradiated with 1.7-keV deuterium ions. TDS measurements were performed in situ immediately after deuterium irradiation and after exposure to air for one week. The amount of desorbed deuterium is the largest for the sample without helium irradiation from the TDS results measured in situ immediately after irradiation. Further, the amount of desorbed deuterium is significantly lowered when the helium energy is increased to 3.5 keV with no significant changes thereafter, indicating that the layer formed by implanted helium at near or deeper than the stopping range for 1.7-keV deuterium ions effectively acts as a barrier against deuterium diffusion into the depth. Because of the strong diffusivity of deuterium into the ambient atmosphere, the amounts of desorbed deuterium are greatly reduced for the samples without helium irradiation and with 1.4-keV helium irradiation after exposure to air for one week. In addition, our deuterium results for the ARAA are also compared with the results for F82H by other authors.

Deuterium permeation and retention behaviors in erbium oxide-iron multilayer coatings

Hydrogen isotope migration behaviors in ceramics-metal multilayer coatings have been elucidated for a further improvement of fueling system and radiological safety. Erbium oxide (Er2O3) coatings were fabricated by filtered vacuum arc deposition (VAD) on reduced activation ferritic/martensitic steel substrates. An iron (Fe) layer was fabricated by radio-frequency magnetron sputtering or covered with an Fe foil on the Er2O3 coating. An Er2O3-Fe-Er2O3 three-layer coating was also fabricated by the VAD on the Er2O3-Fe coating. The grain boundary diffusion model was constructed based on the results of grain structure analysis and deuterium depth profile for the Er2O3 single-layer coating from our previous works. The Er2O3-Fe two-layer coatings with different Fe layer thickness showed no significant difference in deuterium permeability. The Er2O3-Fe-Er2O3 three-layer coating showed a PRF of up to 104 due to a contribution of two diffusion barriers of the inner and outer Er2O3 layers. In addition, the three-layer coating had three times higher deuterium concentration than the Er2O3-Fe coating, although the concentration will be negligibly small from the perspective of tritium inventory in the fuel system.

Plasma exposure behavior of re-deposited tungsten on structural materials of fusion reactors

To evaluate the effects of re-deposited tungsten (W) on the surface modification and hydrogen isotope retention behavior of fusion structural materials, the plasma exposure behavior of re-deposited W samples prepared by magnetron sputtering on the F82H steel, the V-5Cr-5Ti alloy as well as bare substrate samples was investigated. All the samples were exposed to 367 shots of deuterium plasmas in the 2015 spring EAST campaign. After the plasma exposure, large area of W layer was exfoliated, while big blisters were found at the interface between the remaining W layer and the substrate materials. The deuterium retention behavior of the samples with re-deposited W layer was characterized by thermal desorption spectroscopy and compared with the bare substrate samples.

Influence of He ions irradiation on the deuterium permeation and retention behavior in the CLF-1 steel

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 3.5MeV 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 623K, 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.

Deuterium retention in molten salt electrodeposition tungsten coatings

Molten salt electrodeposition is a promising technology to manufacture the first wall of a fusion reactor. Deuterium (D) retention behavior in molten salt electrodeposition tungsten (W) coatings has been investigated by D-plasma exposure in the EAST tokamak and D-ion implantation in an ion beam facility. Tokamak exposure experiments demonstrate that coatings prepared with lower current density exhibit less D retention and milder surface damage. Deuterium-ion implantation experiments indicate the D retention in the molten salt electrodeposition W is less than that in vacuum plasma spraying W and polycrystalline W.

Annealing effects on deuterium retention behavior in damaged tungsten

Effects of annealing after/under iron (Fe) ion irradiation on deuterium (D) retention behavior in tungsten (W) were studied. The D2 TDS spectra as a function of heating temperature for 0.1dpa damaged W showed that the D retention was clearly decreased as the annealing temperature was increased. In particular, the desorption of D trapped by voids was largely reduced by annealing at 1173K. The TEM observation indicated that the size of dislocation loops was clearly grown, and its density was decreased by the annealing above 573K. After annealing at 1173K, almost all the dislocation loops were recovered. The results of positron annihilation spectroscopy suggested that the density of vacancy-type defects such as voids, was decreased as the annealing temperature was increased, while its size was increased, indicating that the D retention was reduced by the recovery of the voids. Furthermore, it was found that the desorption temperature of D trapped by the voids for damaged W above 0.3dpa was shifted toward higher temperature side. These results lead to a conclusion that the D retention behavior is controlled by defect density. The D retention in the samples annealed during irradiation was less than that annealed after irradiation. This result shows that defects would be quickly annihilated before stabilization by annealing during irradiation.

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.

Effect of Heating Temperature on Deuterium Retention Behavior for Helium/Carbon Implanted Tungsten

The effect of heating temperature on deuterium (D) retention behavior for helium (He+) / carbon (C+) implanted tungsten (W) was studied. It was found that D retention behavior for He+ implanted W was not limited by the size of the He bubbles. The microstructure observation showed that the large helium bubbles were formed near the surface for He+ implanted W at 1173 K, suggesting that the D retention was reduced by the growth of the helium bubbles. In addition, to evaluate the effect of implantation ion species at high temperature, D retention behavior for He+ implanted W at 1173 K was compared with that for C+ implanted W at 673 K. It is concluded that the D retention depends on ion species, which makes different kinds of damages like He bubbles for He+ implantation and vacancy-ion complex (voids) for C+ implantation.