Deuterium Depth Profiles Research Articles

Influence of near-surface blisters on deuterium transport in tungsten

The effect of near-surface blisters on deuterium transport in tungsten is studied by means of nuclear reaction analysis (NRA) and scanning electron microscopy (SEM). Gentle deuterium plasma loading of different durations and subsequent NRA depth profiling is performed in heavily pre-blistered and unblistered areas on self-damaged tungsten samples. Comparison of the deuterium depth profiles reveals a considerable reduction of the deuterium transport into the bulk due to the presence of near-surface blisters. SEM and NRA results identify the enhanced re-emission of deuterium from the sample due to open blisters as the underlying mechanism, which reduces the deuterium flux into the bulk. Based on a simple analytical hydrogen retention model, the re-emitted deuterium flux by open blisters is determined to be 80% of the implanted deuterium flux in the present conducted experiment. In addition, the deuterium flux into the bulk is reduced by 60% compared to the unblistered case. Hence the presence of blisters is not a general disadvantage in the context of retention, but can be beneficial in slowing down the build up of a certain hydrogen inventory and in reducing the permeation flux.

Direct depth distribution measurement of deuterium in bulk tungsten exposed to high-flux plasma

Understanding tritium retention and permeation in plasma-facing components is critical for fusion safety and fuel cycle control. Glow discharge optical emission spectroscopy (GD-OES) is shown to be an effective tool to reveal the depth profile of deuterium in tungsten. Results confirm the detection of deuterium. A ∼46 μm depth profile revealed that the deuterium content decreased precipitously in the first 7 μm, and detectable amounts were observed to depths in excess of 20 μm. The large probing depth of GD-OES (up to 100s of μm) enables studies not previously accessible to the more conventional techniques for investigating deuterium retention. Of particular applicability is the use of GD-OES to measure the depth profile for experiments where high deuterium concentration in the bulk material is expected: deuterium retention in neutron irradiated materials, and ultra-high deuterium fluences in burning plasma environment.

Hydrogen isotope accumulation in the helium implantation zone in tungsten

The influence of helium (He) on deuterium (D) transport and retention was studied experimentally in tungsten (W). Helium was implanted 1 µm deep into W to a maximum calculated concentration of 3.4 at.%. To minimize the influence of displacement damage created during the He implantation on D retention, so-called self-damaged W was used. W was damaged by 20 MeV W ion bombardment and defects were populated by low-temperature D plasma at room temperature before He implantation. Deuterium depth profiling was performed in situ during isochronal annealing in the temperature range from 300 K to 800 K. It is shown for the first time unambiguously that He attracts D and locally increases D trapping. Deuterium retention increased by a factor of two as compared to a non-He implanted W reference after sample annealing at 450 K. Rate equation modelling can explain the measured D depth profiles quantitatively when keeping the de-trapping parameters unchanged but only increasing the number of traps in the He zone. This bolsters the confidence in the theoretical calculations predicting that more hydrogen isotopes can be stored around a He cluster zone.

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.

Deuterium retention in MeV self-implanted tungsten: Influence of damaging dose rate

Recrystallized, polycrystalline tungsten was self-damaged by 20MeV tungsten ions up to a calculated damage dose in the damage peak of 0.23dpa. The time to acquire this dose and hence the average damaging dose rate was varied from 6×10-3 to 4×10-6dpa/s, the latter coming close to the damage dose rate expected from fusion neutrons in future devices such as ITER and DEMO. One series was conducted at 295K and one at 800K to check for possible effects of defect evolution at elevated temperature. The created damage was decorated afterwards with a deuterium plasma at low ion energy of <15eV and low flux of 6×1019D/m2 until saturation to derive a measure for the defect density that can retain hydrogen isotopes. 3He nuclear reaction analysis (NRA) was applied to derive the deuterium depth profile and the maximum concentration in the damage peak. Neither for the 295K nor for the 800K series a variation in deuterium retention with damage dose rate was found.

Mitigated blistering and deuterium retention in tungsten exposed to high-flux deuterium–neon mixed plasmas

Surface morphology and deuterium retention in tungsten exposed at surface temperature of ~550 K to mixed deuterium–neon plasmas of different neon concentrations are investigated. It is found that the addition of neon up to 20% mitigates blistering on the surface. Cross-section view of the surface shows the formation of pores near the surface in the depth less than 100 nm. Deuterium depth profile is featured by an enhanced deuterium concentration within a depth of 16 nm but a mitigated penetration in depth larger than 1 µm. Deuterium retention is reduced by up to a factor of four. It is suggested the open pores formed in the surface serves as an escaping channel, mitigates deuterium penetration towards bulk and retention in the bulk.

Optimization of the depth resolution for deuterium depth profiling up to large depths

The depth resolution of deuterium depth profiling by the nuclear reaction D(3He,p)α is studied theoretically and experimentally. General kinematic considerations are presented which show that the depth resolution for deuterium depth profiling using the nuclear reaction D(3He,p)α is best at reaction angles of 0° and 180° at all incident energies below 9MeV and for all depths and materials. In order to confirm this theoretical prediction the depth resolution was determined experimentally with a conventional detector at 135° and an annular detector at 175.9°. Deuterium containing thin films buried under different metal cover layers of aluminum, molybdenum and tungsten with thicknesses in the range of 0.5–11μm served as samples. For all materials and depths an improvement of the depth resolution with the detector at 175.9° is achieved. For tungsten as cover layer a better depth resolution up to a factor of 18 was determined. Good agreement between the experimental results and the simulations for the depth resolution is demonstrated.

Deuterium removal from radiation damage in tungsten by isotopic exchange with hydrogen atomic beam

The tungsten samples were pre-irradiated with self-ions to create radiation-induced defects and then exposed to the deuterium atomic beam. The deuterium removal was studied by isotopic exchange with atomic hydrogen beam. Modification of the deuterium depth profile in self-ion irradiated tungsten under isotopic exchange up to a depth of 6 μm was measured in- situ by nuclear reaction analysis. The total deuterium retention after isotopic exchange was measured by thermal desorption spectroscopy. It is shown that the efficiency of the deuterium removal increases with increasing of the hydrogen incident flux, incident energy and temperature of the tungsten sample.

Observation of oversaturation-induced defect formation in tungsten irradiated by low energy deuterium ion

The type of induced material damage in the tungsten irradiated by using deuterium ions was investigated for various value of the fluence at low energy. Experiments were carried out in an electron cyclotron resonance (ECR) plasma source that provided an ion flux of 2.8 × 1021D2+/m2s and a sheath energy of 100 eV/D2+ on the tungsten target. The energy of irradiated ions was much smaller than the threshold energy for generating cascade collisional damage (∼ 250 eV) in tungsten and was similar of the plasma at the first wall of KSTAR. The target temperature was kept as 700 − 800 K by using an active cooling system. Secondary ion mass spectroscopy (SIMS) was employed to observe the depth profiles of deuterium. The peak of the trapped deuterium concentration in the irradiated tungsten was located near 16 − 17 nm for 2.0 − 4.0 × 1025D2/m2, which is far deeper than the 1.6 nm for ion implantation at 100 eV/D2+ ions. Thermal desorption spectroscopy (TDS) data were analyzed to determine the binding energy (Eb = 1.45 eV) of trapped deuterium, which corresponded to an oversaturation-induced vacancy. This observation is very important for understanding the refueling property of the retained deuterium during steady-state fusion plasma operation.

Comparative Study of Hydrogen- and Deuterium-Induced Degradation of Ferroelectric (Pb,La)(Zr,Ti)O3 Capacitors Using Time-of-Flight Secondary Ion Measurement.

Ferroelectric (Pb,La)(Zr,Ti)O3 (PLZT) capacitors were fabricated with Pt, Al:ZnO (AZO), or Sn:In2O3 (ITO) top electrodes. Hydrogen- or deuterium-induced degradation was investigated for the three capacitors by annealing in a 3% H2/balance N2 or 3% D2/balance N2 ambient environment at 200 °C and 1 torr. The remnant polarization of all capacitors decreased after annealing in both H2 and D2 ambient after 45 min, and the remnant polarization of the Pt/PLZT/Pt capacitor significantly decreased after 45-min annealing compared with that of the AZO/PLZT/Pt and ITO/PLZT/Pt capacitors, even though the initial remnant polarization for the Pt/PLZT/Pt capacitor was larger. Time-of-flight secondary ion mass spectrometry showed slight differences in hydrogen content for the three different capacitors after H2 annealing. In contrast, the deuterium content of the Pt/PLZT/Pt and AZO/PLZT/Pt or ITO/PLZT/PT capacitors was significantly different after deuterium annealing. Deuterium depth profiles for the Pt/PLZT/Pt capacitor after annealing showed that deuterium conformally exists in the PLZT layer of the Pt/PLZT/Pt capacitor, and deuterium accumulation under the Pt bottom electrode was also observed. This result suggests that diffusion of deuterium in Pt was much higher than that in PLZT. AZO and ITO top electrodes could act as a hydrogen barrier layer for ferroelectric films.

The influence of the annealing temperature on deuterium retention in self-damaged tungsten

The influence of the annealing temperature on deuterium retention was studied for self-ion damaged tungsten in the range from 600–1200 K. Samples were damaged by 20 MeV W ions at room temperature to the peak damage level of 0.5 dpa. Samples were then annealed at the desired temperature for 1 h and exposed to deuterium atom beam with the flux of for 144 h to populate the remaining defects. An unannealed sample was also used as a reference. Nuclear reaction analysis technique was used for deuterium depth profile analysis and thermal desorption was performed on the same samples to measure the amount of total retained D. Scanning transmission electron microscopy was used for the calculation of dislocation densities in the samples. After annealing at 1200 K approximately 66% of those initial defects which retain deuterium were annealed.

The effect of ion flux on plasma-induced modification and deuterium retention in tungsten and tungsten–tantalum alloys

The paper presents the results of an experimental study of deuterium retention in W and W–Ta alloy that were exposed to first-wall relevant low flux (∼1020m−2s−1) deuterium plasma in the ECR plasma generator PlaQ. Subsequent analysis included surface imaging by optical microscopy, deuterium depth profiling by nuclear reaction analysis (NRA) and measurements of deuterium content by thermal desorption spectroscopy (TDS). It was found that under investigated exposure conditions the deuterium content was higher in W–Ta alloy than in W. Combined with the previously reported results showing that under high-flux (∼1024m−2s−1) retention is higher in W instead, this gives rise to a peculiar flux effect – dependence of relative retention between different materials on exposure flux. We interpret this effect as evidence that at different flux ranges different populations of trapping sites determine the retention, namely pre-existing microstructural traps at low-flux exposure and plasma-induced ones at high-flux exposure.

Comparison of Laser and Oven Annealing Effects on Hydrogen and Microstructure in Thin Film Silicon

ABSTRACTLaser and oven annealing effects on hydrogen concentration, hydrogen diffusion and material microstructure in hydrogenated amorphous silicon films deposited on crystalline silicon substrates are compared. For laser annealing, a 6 W green (532 nm) continuous wave laser with 100 µm focus diameter was applied and samples of about 1 cm2were scanned in ambient with a line distance of 50 µm and at a speed of 1 – 100 mm/s. Hydrogen content and microstructure were measured by infrared spectroscopy, and hydrogen diffusion was investigated by secondary ion mass spectroscopy (SIMS) measurements of depth profiles of deuterium and hydrogen in layered structures of deuterated and hydrogenated material. The results show that in both annealing experiments hydrogen diffuses predominantly in form of atoms although some formation of H2molecules cannot be excluded. By comparison of laser and oven treatment, an effective temperature describing the laser treated state can be defined. Furthermore, the temperature of the thin silicon film during laser treatment is estimated.

Development of an analytical diffusion model for modeling hydrogen isotope exchange

We create a model for H retention depth profiles in W and subsequently model how this profile changes after isotope exchange. This is accomplished by calculating how trapping defects in W accumulate D (or H) inventory as W is being exposed to plasma. Each trapping site is characterized by a trapping rate and a release rate, where the only free parameters are the distribution of these trapping sites in the material. The filled trap concentrations for each trap type are modeled as a diffusion process because post-mortem deuterium depth profiles indicate that traps are filled well beyond the ion implantation zone (2–5nm). Using this retention model, an isotope exchange rate is formulated. The retention model and isotope exchange rate are compared to low temperature isotope exchange experiments in tungsten with good agreement. The limitations of the current model highlight important physics and motivate future work.

Influence of helium on hydrogen isotope exchange in tungsten at sequential exposures to deuterium and helium–protium plasmas

Hydrogen isotopes exchange in tungsten was investigated after sequential exposures to low energy deuterium (D) and helium–seeded protium (He–seeded H) plasmas at sample temperatures of 403 and 533K. Deuterium depth profiles were measured by the D(3He, p)4He nuclear reaction with 3He+ energies between 0.69 and 4.5MeV allowing determination of the D concentration up to a depth of 8μm. It was found that a significant part of the deuterium initially retained in tungsten after D plasma exposure was released during sequential exposure to a protium plasma. However, exposure of the D-plasma-exposed W samples to the He–seeded H plasma reduces the amount of released deuterium as compared to pure H plasma exposure.

Retention of deuterium in damaged low-activation steel Rusfer (EK-181) after gas and plasma exposure

Reduced-activation ferritic–martensitic steels (RAFMS) are advanced structural materials for the construction of future fusion reactors with high fluxes of neutrons, such as DEMO or a Fusion Neutron Source (FNS).In the present work the influence of different damages on deuterium retention in the RAFM Rusfer (Chernov et al., 2007) was investigated. Three different types of damage were applied:(i)irradiation by 20MeV W6+ ions to a damage fluence of 0.89dpa (1.4×1018ions/m2). Tungsten ion irradiation was used as proxy for displacement damage created by neutrons;(ii)heat loads in the QSPA-T facility with 10 pulses of 0.5MJ/m2 with a duration of 0.5ms;(iii)low-temperature hydrogen plasma irradiation in the LENTA facility at 320 and 600K to a fluence of 1025H/m2.The hydrogen isotope retention properties of the damaged and undamaged material were investigated by exposure to deuterium gas several weeks after damaging. The deuterium retention was investigated in the temperature range of RT–773K at pressure 104Pa. Deuterium depth profiles were measured a month after gas exposure by nuclear reaction analysis (NRA) using the D(3He,p)α nuclear reaction. Deuterium retention in damaged and undamaged Rusfer in the temperature range of 300–600K has a maximum at 500K for all types of damage investigated. The typical value of deuterium concentration in the bulk is 10−3at.%. Peculiarities of D retention in damaged samples are discussed.

Surface modification and deuterium retention in reduced activation ferritic martensitic steels exposed to low-energy, high flux D plasma and D2 gas

Samples prepared from steels F82H and EUROFER97 were irradiated with 20 MeV W ions at 300 K to 0.54 displacements per atom at the damage peak. Damaged and undamaged samples were exposed at elevated temperatures both to deuterium plasma at ion energies of 60 and 200 eV to a fluence of ≈1026 D m−2 and to D2 gas at a pressure of 100 kPa. The surface modification after plasma exposure was examined by scanning electron microscopy and Rutherford backscattering spectroscopy. Deuterium depth profiles were determined by the D(3He, p)4He nuclear reaction. In damaged steels loaded with deuterium, deuterium decorates the damage profile and the D concentration decreases with increasing temperature. After exposure of the F82H steel to the D plasma W-enriched near-surface layers are formed. The effective concentration of W in the near-surface steel layer depends on plasma exposure conditions.

In situ nuclear reaction analysis of D retention in undamaged and self-damaged tungsten under atomic D exposure

In situ deuterium depth profile measurements by nuclear reaction analysis (NRA) were performed on undamaged and neutron-like damaged W during D atom exposure at several sample temperatures. Damaged W was produced by high energy W-ion irradiation. In situ results of D retention after exposure termination are in good agreement with ex situ measurements for damaged W. In contrast, in situ NRA during D exposure on undamaged W showed significantly higher D retention than ex situ NRA measurement. It is also shown that this increased retention is lower at the location non-irradiated by the probing 3He beam. These observations indicate that the D retention is influenced by secondary damage produced by the probing 3He beam, however the dynamic retention could also partially contribute to the observed retention increase.

Modeling hydrogen isotope behavior in fusion plasma-facing components

In this work, we focus on understanding hydrogen isotope retention in plasma-facing materials in fusion devices. Three common simulation methods are usually used to study this problem that includes Monte Carlo, molecular dynamics, and numerical/analytical methods. A system of partial differential equations describing deuterium behavior in tungsten under various conditions is solved numerically to explain recent data compared to other methods. The developed model of hydrogen retention in metals includes classic, intercrystalline and trapped-induced Gorsky effects. The bombardment and depth profile of 200eV deuterium in single crystal tungsten are simulated and compared with recent work. The total deuterium retention at various temperatures and fluences are also calculated and compared with available data. The results are in reasonable agreement with data and therefore, this model can be used to estimate deuterium inventory and recovery in future fusion devices.

Hydrogen trapping in 3He-irradiated Fe

The characteristics of irradiation-induced hydrogen (deuterium) traps in pure Fe were investigated for quantitative evaluation of tritium retention in fusion reactor components. The deuterium depth profiles of an Fe disk sample exposed to deuterium plasma were observed by means of nuclear reaction analysis (NRA) before and after irradiation with 0.8MeV or 1.3MeV 3He ions. Irradiation generated a number of traps, and deuterium retention was drastically increased subsequent to irradiation. Steady-state deuterium concentration in the trap and the solution sites were obtained by continuously charging the sample with deuterium during the NRA. Based on these values, the trapping energy, which is the enthalpy difference between the two sites, was estimated to be 0.38eV. The number ratio of the trap to atomic displacement was 0.013. Some of the traps were annihilated around 523K. The annihilation temperature, the trapping energy, and the equilibrium constant suggest that the trap is a dislocation loop introduced by the irradiation. It is deduced that the tritium inventory in the Fe components of a reactor should be drastically increased by neutron irradiation due to the formation of traps, but may be significantly reduced by high temperature operation of the components.