Deuterium Retention Behavior Research Articles

Deuterium permeation and retention behavior in copper-based brazing filler materials

Copper-based brazing filler materials are commonly employed in vacuum brazing techniques to achieve joining of plasma-facing components for ITER. In this work, gas-driven permeation (GDP) and thermal desorption spectroscopy (TDS) experiments have been performed to study the hydrogen isotopes permeation and retention behavior in two types of brazing filler materials (CuMnNi and CuGeNi metals). Experimental results show that CuMnNi filler exhibits the similar deuterium (D) permeability with CuMnNi filler but lower diffusion coefficients than CuGeNi filler. Three distinct peaks are observed between 380 K and 980 K for CuMnNi filler. However, no evident desorption peak is observed for CuGeNi filler. CuGeNi filler may be considered as a potential brazing filler material for reducing tritium inventory.

Deuterium retention in heavy-ion and helium-ion sequentially irradiated tungsten

Neutron irradiation as well as the presence of helium (He) significantly affects fuel inventory in plasma-facing materials. To investigate the synergistic effects of neutron and He irradiation on deuterium (D) retention behaviors in tungsten, heavy- and He-ion sequentially irradiation experiments were performed with various He fluence and/or heavy-ion damage levels, and then the samples were exposed to low-energy D plasmas at 450 K. It is shown that even a low He concentration of 0.5 atomic parts per million (appm) increases D concentration in the heavy-ion damaged region, which increases further with increasing He concentration under the parameters selected in this work (up to a maximum He concentration of 2.1 appm). The total D inventory in tungsten bulk also increases with He fluence due to the increase in D concentration both in the heavy-ion damaged region and the region irradiated by He-ion only. Furthermore, heavy-ion and He-ion successive irradiation slightly increases D retention in tungsten compared to the individual He ions irradiation. Similar to single heavy-ion damaged tungsten, the saturation of D retention is observed as heavy-ion irradiation damage above 0.2 dpa at a fixed He fluence.

Irradiation hardening and deuterium retention behaviour of tungsten under synergistic irradiations of 3.5 MeV Fe13+ ions and deuterium plasma

This work investigates the irradiation hardening and deuterium (D) retention behaviour of tungsten (W) under synergistic irradiations of heavy ions and D plasma. 3.5 MeV iron (Fe13+) ion irradiation was performed on recrystallized W samples (RW) to produce displacement damage of 1 dpa. Then, low-energy (38 eV) D plasma exposure was conducted at 500 K. It is found that Fe ion irradiation creates substantial vacancy-type defects and dislocation loops/networks in RW. These irradiation-induced defects not only function as nucleation sites for dislocations that increase the activation probability of new dislocations and suppress the pop-in events, but also act as barriers for dislocations that result in irradiation hardening. Nano-indentation results show that the average hardness of RW-D, RW-Fe and RW-Fe-D increases from 5.26 GPa for RW to 5.28, 6.23 and 6.56 GPa, respectively. The strong interaction between dislocations and high density of damage-induced defects is suggested to be the chief source for the obvious irradiation hardening observed in the synergistic irradiation case. Besides, compared with RW-D, the total D retention in RW-Fe-D is increased by a factor of 2.65. NRA and TDS results suggest that Fe pre-irradiation not only increases D retention within the damage layer (within the first 1.3 μm), but also enhances that beyond the damage layer (> 1.3 μm) before surface blisters are formed. This work further improves the fundamental understanding on the microstructure evolution, D retention and nano-mechanical behaviour of W under synergistic irradiation effect of heavy ions and D plasma.

Effect of anisotropic grain boundaries on the surface blistering of tungsten induced by deuterium plasma exposure

The effect of anisotropic grain boundaries on the surface microstructure changes and deuterium retention behavior in tungsten under deuterium plasma irradiation were studied. The samples named as NDW and RDW were cut from RD × TD plane and TD × ND plane of a rolled W plate (ND: normal direction, RD: rolling direction, TD: transverse direction), respectively. The results suggest that both surface morphology and deuterium retention in the RDW samples are suppressed compared to that of NDW samples for 1 h and 25 h deuterium irradiation. For 1 h exposure, the total deuterium retention in the NDW samples is about 1.6 times as much as that of RDW. However, as the exposure time increased to 25 h, the total deuterium retention in the NDW sample is two times the height of RDW sample. The experimental results show that the special grain structure of RDW can alleviate the surface morphology changes and decrease the deuterium retention in tungsten, which is interpreted by the lateral stress model of blistering.

The erosion and retention properties of α-WC films by low-energy deuterium ion irradiation

Since WC can be easily formed on tungsten (W) surface through the interaction between the plasma and W surface in the C-containing impurity atmosphere in tokamaks, the investigation on erosion and deuterium (D) retention behaviors of WC is mandator for stable plasma operation. In this work, α-WC films were successfully prepared by combining reactive magnetron sputtering and thermal annealing, then irradiated by D ions. The WC shows W sputtering yield smaller than pure W, and it is almost independent with temperature. Thus, the erosion of WC could be mainly described by physical sputtering. The crystalline of WC has been improved after irradiation, and the surface shows a dominant metallic W phase due to preferential sputtering of C. D concentration in WC is 5 times higher than that in W through elastic recoil detection analysis, mainly because the Frenkel defect pairs induced by D irradiation and interstitial sites can trap D in large quantities, and their D profiles are similar. But different from W in double desorption peaks, D2 desorption on WC is a single peak whose temperature is close to the low temperature peak of W. This work provides a further understanding in erosion and deuterium retention behaviors of WC.

Deuterium retention and desorption behavior of W-Ta-Cr-V high entropy alloy

Deuterium retention and desorption behavior of W-Ta-Cr-V films under deuterium plasma irradiation were studied. The W-Ta-Cr-V films were prepared on the molybdenum (Mo) substrate by magnetron sputtering. Subsequently, all films annealed at 1000 °C were exposed to deuterium plasma at a flux of 8.13 × 1019 D/m2·s for 9 h to achieve an irradiation fluence of 2.63 ×1024 D/m2. Electron backscattered diffraction (EBSD) was used to characterize the texture and average grain size of films. Doppler broadening spectrometry of positron annihilation (DBS-PA) was carried out with an energy-variable slow positron beam to detect the concentration of vacancy-type defects in the films. The results suggest that there are fewer vacancy-type defects inside W-Ta-Cr-V grains. Deuterium retention and the desorption behavior of W-Ta-Cr-V were studied by compared to pure tungsten films. In the 300–800 K temperature range, the peaks of thermal desorption spectroscopy (TDS) of W-Ta-Cr-V films are substantially consistent with that of the W films. Notably, in the range of 800 K-1000 K, W-Ta-Cr-V film has no thermal desorption peak. We attribute this to the large lattice distortion and low vacancy formation energy, which inhibit the growth of voids inside the grains of W-Ta-Cr-V films. The total deuterium retention of W-Ta-Cr-V films is compared with that of W films. It is measured by TDS that the deuterium retention of W-Ta-Cr-V films is nearly 5 times that of W films. Based on the analysis of the TDS spectra, the higher deuterium retention in W-Ta-Cr-V films is due to the more grain boundary defects.

Deuterium Retention and Release Behavior from Beryllium Co-Deposited Layers at Distinct Ar/D Ratio

Beryllium-deuterium co-deposited layers were obtained using DC magnetron sputtering technique by varying the Ar/D2 gas mixture composition (10/1; 5/1; 2/1 and 1:1) at a constant deposition rate of 0.06 nm/s, 343 K substrate temperature and 2 Pa gas pressure. The surface morphology of the layers was analyzed using Scanning Electron Microscopy and the layer crystalline structure was analyzed by X-ray diffraction. Rutherford backscattering spectrometry was employed to determine the chemical composition of the layers. D trapping states and inventory quantification were performed using thermal desorption spectroscopy. The morphology of the layers is not influenced by the Ar/D2 gas mixture composition but by the substrate type and roughness. The increase of the D2 content during the deposition leads to the deposition of Be-D amorphous layers and also reduces the layer thickness by decreasing the sputtering yield due to the poisoning of the Be target. The D retention in the layers is dominated by the D trapping in low activation binding states and the increase of D2 flow during deposition leads to a significant build-up of deuterium in these states. Increase of deuterium flow during deposition consequently leads to an increase of D retention in the beryllium layers up to 300%. The resulted Be-D layers release the majority of their D (above 99.99%) at temperatures lower than 700 K.

Combination of in-situ ion beam analysis and thermal desorption spectroscopy for studying deuterium implanted in tungsten

We demonstrate a combinatorial approach integrating ion implantation followed by thermal annealing and simultaneous in situ ion beam analysis with thermal desorption spectroscopy in a single set-up. Atomic and molecular deuterium ions of 3 keV were implanted into bulk tungsten with doses exceeding 1 × 1022 ions m−2. Depth profiling of both, protium and deuterium was performed by elastic recoil detection analysis, while simultaneously the outgassing rates of molecular deuterium by thermal desorption spectroscopy were monitored during temperature ramps from room temperature to ≈1400 K. The combination of the two techniques in situ is shown capable to identify the distinct retention behavior of deuterium at different temperatures and in different reservoirs, e.g. located close to the surface and diffused deep into the material. Ex-situ scanning electron microscopy confirmed blister formation, and recovery of the initial surface morphology after high temperature annealing, in analogy to comprehensive ex-situ studies.

Deuterium retention and surface morphology modification in oxidized tungsten exposed to D plasma

Tungsten (W) is the most promising plasma facing material in future fusion reactor. However, oxidation occurs readily on W and might influence the hydrogen isotopes retention behavior in W. To explore the mechanism of oxidation on the hydrogen isotopes retention behavior in W, oxidized W samples have been exposed to D plasma (32.3 eV, 2 × 1020 Dm−2s−1) in a linear plasma device. Surface morphology and deuterium (D) retention behavior in W have been characterized after exposure to D plasma with fluence of 1023 D/m2. After D plasma exposure, a porous layer has been found on the surface of oxidized W. The main chemical content of the porous layer is pure W. The existence of pre-oxide layer contributes the increase of retention amount of D in the form of D2O and may contribute to the decrease of the retention amount of D in the form of HD/D2.

The influence of deuterium ions on the deuterium permeation and retention behavior in V-4Cr-4Ti alloy under plasma loading

Deuterium (D) permeation and retention behavior in V-4Cr-4Ti alloy under ECR plasma or only gas loading has been studied. The gas pressure for plasma discharge was 0.6 Pa, which was the same as that for D2 gas loading. Different bias voltages of 0 V, -50 V and -100 V were applied to the samples to investigate the effect of ion incident energy on the D permeation behavior. D retention behavior was studied by thermal desorption spectroscopy (TDS). X-ray photoelectron spectroscopy (XPS) with depth profile was utilized to characterize the distribution of surface impurities and the chemical states of vanadium element. When the ion incident energy increases under plasma loading, D permeation flux decreases. D permeation fluxes under plasma loading are generally higher than those under only gas loading except those under D plasma loading with a bias of -100 V below 650 K. In addition, the total D retention in the plasma loaded sample is lower than that in the gas loaded one. XPS analyses show that the plasma loaded sample surface has a shallower V2O3 layer. Additionally, the percentage of VC in the plasma loaded sample is less than that in the gas loaded one. It suggests that D ions especially those with high ion incident energy could reduce the content of the oxides and carbides on the sample surface, which may finally result in the abnormal D permeation and retention behavior under or after plasma loading.

Numerical analysis of deuterium migration behaviors in tungsten damaged by fast neutron by means of gas absorption method

Deuterium retention behavior in tungsten damaged by fast neutrons at high temperatures (0.43 dpa at 918 K and 0.74 dpa at 1079 K) and 6.4 MeV Fe2+ (0.3 dpa at R.T.) were investigated to evaluate the tritium retention property of fusion reactor divertors. A deuterium gas absorption method was carried out to avoid additional damage that may be induced by plasma exposure, then, deuterium retention and desorption behaviors were investigated quantitatively by means of thermal desorption spectroscopy and the following simulation code. The deuterium desorption spectra for tungsten samples were analyzed by the numerical code which includes the elementary steps of hydrogen isotope migration processes including diffusion, trapping, detrapping, and surface recombination. The evaluated deuterium detrapping energy from the irradiation defects in neutron irradiated tungsten sample was larger than that in 6.4 MeV Fe2+ irradiated tungsten. It was suggested that the dominant deuterium trapping site in the neutron irradiated tungsten would be voids which was formed by the accumulation of vacancies during neutron irradiation under high temperature and long duration.

Surface blistering and deuterium retention behaviors in pure and ZrC-doped tungsten exposed to deuterium plasma

Surface blistering and deuterium (D) retention behaviors in pure tungsten (W) and 0.5 wt% ZrC-doped W alloy (WZC) exposed to D plasma have been investigated as a function of incident D fluence up to 2.6 × 1025 D m−2. Surface observations show that large-sized blisters (∼8–10 μm) and a few small blisters (<2 μm) are formed on the W samples. For WZC, the surface of the samples is covered by a high density of small blisters (<1 μm), and no blister larger than 2 μm is found. Cross section views show that the large-sized blisters originate from subsurface grain boundaries, and the small blisters originate from intra-granular cavities at depths much closer to the surface. The intra-granular blisters are preferentially formed on the grains with a normal direction close to [111] for both types of sample. The inter-granular blisters formed in pure W are significantly suppressed in WZC, and the fluence threshold for blister formation in WZC is lower than that in W. The D depth distributions indicate that the implanted D is mainly retained in the near-surface region of WZC and W. Besides, a higher and broader D peak is observed at a depth of between 0.1–0.25 μm in WZC, but the intensity and width of the D peak is lower and thinner in W. The desorption spectra of WZC shift to the lower temperature side as compared to W, and the total retained D amount in WZC and W is comparable in the fluence range of 2.8 × 1024–2.6 × 1025 D m−2.

Deuterium permeation and retention in 316L Stainless Steel Manufactured by Laser Powder Bed Fusion

Deployment of additively manufactured materials into nuclear energy systems requires investigation of the full range of the unique environmental effects on these materials. Hydrogen isotopes are common gaseous species in nuclear reactors and exhibit ample high mobility in most materials. While hydrogen isotopes mobility in wrought stainless steel is well understood, this property is not thoroughly studied for its additively manufactured variants. In this study, we investigated the deuterium permeation and retention in 316L stainless steel manufactured by laser powder bed fusion. The results showed that the deuterium permeability in the as-built additively manufactured 316L stainless steel (AM SS316L) is greater than that of the reference wrought 316L stainless steel by a factor of 2.8. However, the stress-relieved and solution-annealed AM SS316L samples exhibit lower deuterium permeability in comparison with the as-built condition. Following the solution annealing, the deuterium permeability of AM SS316L is comparable with that of the reference wrought SS316L. Deuterium retention in the as-built AM SS316L is 45% higher than that of the wrought 316L stainless steel and slightly higher than that of the two thermally annealed AM SS316L materials. Transmission electron microscopy and positron annihilation lifetime spectroscopy were used to obtain microstructural information used to determine deuterium permeation and retention behavior in the studied materials.

Deuterium and helium desorption behavior and microstructure evolution in beryllium during annealing

Beryllium was chosen as a plasma-facing material (PFM) for the first wall in an international thermonuclear experimental reactor (ITER). In this study, the deuterium and helium retention behavior in Be and the associated microstructure evolution were investigated using thermal desorption spectroscopy (TDS) and transmission electron microscopy (TEM). Since the disappearance of bubbles was observed at desorption peak temperatures of about 773 K for deuterium and 973 K for helium, these desorption peaks could be attributed to the release from each bubble. The existence of deuterium or helium within each bubble was also clearly confirmed using electron energy-loss spectroscopy (EELS) combined with scanning transmission electron microscopy (STEM). There was a clear difference between the disappearance process of the deuterium and helium bubbles. While the deuterium bubble gradually shrunk and disappeared, the helium bubble instantly vanishes. The results showed that the stability of helium bubbles was higher than that of deuterium bubbles.

Deuterium retention in tungsten and tungsten alloys exposed to pure and impurities seeding deuterium plasma

Abstract Polycrystalline tungsten (W) and four kinds of tungsten alloys were exposed to pure and impurities seeding deuterium (D) plasma to investigate the deuterium retention behaviour. The results show that D retention in W increases with the increasing fluence and incident energy, and decreases with the increase of temperature. The doping of alloying element and compounds, such as Re, La2O3, and ZrC, suppress the D retention at lower fluence but has limited effect at fluence above 2.6 × 1025 D/m2, and D inventory in these W alloys is higher than in W at temperature of 800 K. Helium (He) and nitrogen (N) impurities seeding significantly reduce D retention in tungsten. And higher N content in plasma causes more obvious reduction of D inventory. But the argon (Ar) impurity seeding leads to the increase of D retention.

Influences of displacement damage due to heavy ion irradiation on deuterium retention in CLF-1 steel

Abstract In this study, the structure damage and its influence on deuterium retention in the reduced-activation ferritic martensitic (RAFM) steel were investigated using heavy ions to simulate the displacement damage caused by neutron irradiation. Iron (Fe) and nickel (Ni) ions were applied as self-ions and other ions to irradiate steel samples. After Fe or Ni heavy ion irradiation, some amorphous cluster structures were evident, especially at the peak displacement damage region located at a depth of around 1.0 µm in the CLF-1 steel. When the displacement damage was higher than 2.0 dpa, the grain boundaries or boundary structures became obscured by the formation of other defects, such as dislocation clusters for samples irradiated by either Fe or Ni ions. The damaged samples irradiated by Fe or Ni heavy ions were exposed to deuterium plasma, and the deuterium release and retention behaviors were measured. The results showed that after being irradiated by Ni ions, the deuterium release peak increased with no changes of the release temperature. However, after Fe ions irradiation, the deuterium release peaks shifted to lower temperatures corresponding to an increase in the number of the low-energy capture sites, such as dislocations. The increase in the total deuterium retention in CLF-1 steel after heavy ion irradiation was less than 56%, which was much lower than those of tungsten materials.

Comparison of deuterium retention in tungsten exposed to deuterium plasma and gas

Deuterium(D) retention behavior in tungsten(W) exposed to deuterium plasma and gas was studied by means of thermal desorption spectroscopy (TDS): deuterium plasma exposure in which W was exposed to D plamsa with 35 eV/D at 393 K to the fluence of 3.8E24 D/m2; D2 gas charging in which W was exposed to D2 gas of 500 kPa at 773 K for 4 hours. TDS shows that the total D retention in plasma exposure W is 1.00E22 D/m2, one order of magnitude higher than that of gas charging W; however, the D2 desorption peak of gas charging W is 952 K, much higher than 691 K of plasma exposure W. The detrapping energies of deuterium were determined experimentally from the measured peak temperatures at different heating rates and were found to be 2.17 eV for gas charging W and 1.04 eV for plasma exposure W, respectively.

Structure, morphology and deuterium retention and release properties of pure and mixed Be and W layers

Deuterium retention and release behavior were investigated in this study, for beryllium, tungsten and mixed beryllium/tungsten thin layers which are of concern for next-generation thermonuclear fusion devices like International Thermonuclear Experimental Reactor. The layers prepared with thermionic vacuum arc technology by using two different ion acceleration voltages 0 V and −700 V respectively were subsequently implanted by exposure to a steady-state deuterium plasma with ion energies of 240 eV/D for D3 and 360 eV/D for D2 at a specific fluency of 2.85 × 1020 m−2 s. Morphology is strongly influenced by layer composition and by the ion acceleration voltage applied in-situ during layer deposition. Crystalline structure analysis shows the presence of a polycrystalline W metallic phase and also highlights a dependence between tungsten crystallite size and Be/W atomic ratio. The amount of deuterium in mixed layers is lower than in pure beryllium layers indicating that W is mitigating D binding states. Results extrapolated for a ‘cold chamber and divertor scenario’ indicate a limited desorption efficiency (∼20%), if the wall baking scenario, for tritium removal procedure in ITER, is to be applied to the investigated layers.

Deuterium Gas-Driven Permeation and Retention Through Tungsten-Coated CLAM Steel

Hydrogen isotope behavior in tungsten coated on reduced activation ferritic/martensitic (RAFM) steels such as China low activation martensitic (CLAM) steel has attracted more attention in the fusion engineering research community. This paper is mainly devoted to the investigation of the effect of tungsten coating on deuterium permeation and retention behavior in RAFM steels. The permeability and diffusion coefficients of CLAM, W-CLAM, and W were determined by gas-driven permeation (GDP) tests followed by thermal desorption spectroscopy to measure deuterium retention. It was found that the observed deuterium permeability and diffusivity of the composite W-CLAM specimen was reduced to about ~60% of the pure CLAM steel, whereas deuterium retention increased, evidently owing to the W coating on the surface that caused the slower release of D into the environment and increased of the effective surface area. In addition, a key finding was that the lath martensite–coarsened and more precipitate phase was found, which may be due to the migration of lath interface during the GDP test.

D retention behavior in the CLF-1 steel after D ion irradiation and D plasma exposure

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.