Deuterium Trapping Research Articles

Analysis of lattice locations of deuterium in tungsten and its application for predicting deuterium trapping conditions

Analysis of lattice locations of deuterium in tungsten and its application for predicting deuterium trapping conditions

The effect of High-Temperature Pre-Damage on Vacancy-Type defects and deuterium retention in tungsten

This work investigated tungsten samples that were pre-damaged at room (300 K) and high temperatures (573 and 873 K) using 3.5 MeV iron ions and then exposed to the high-flux deuterium plasma. The vacancy-type defects and the behavior of deuterium retention in the investigated tungsten were characterized and compared to study the effect of high-temperature pre-damage. Results of slow positron annihilation with Doppler broadening spectroscopy indicate the introduction of numerous defects by the room-temperature damage and the formation of larger-sized vacancy clusters by the high-temperature pre-damage. Surface observation demonstrates that the pre-damage at room temperature effectively suppressed the formation of intra-granular blisters, while the pre-damage at high temperature did not exhibit such a significant suppression effect on blister formation. Thermal desorption results show that pre-damage at room temperature significantly increases deuterium desorption. Pre-damage at high temperatures also strengthens deuterium desorption, not as significantly as room-temperature pre-damage, and meanwhile produces additional desorption at around 900 K. These results indicate differences between room-temperature and high-temperature pre-damage in the generation of defects and their impact on deuterium retention. Compared to room-temperature pre-damage, high-temperature pre-damage tends to generate defects with a larger average size, weakening the suppression effect on deuterium aggregation and blistering within the grains of tungsten, reducing the quantity of deuterium trapping sites and producing higher-energy deuterium traps.

Annealing effect on deuterium retention in W-Cr-Y alloy

Deuterium (D) trapping in the W-11.4Cr-0.6Y alloy manufactured by field assisted sintering technology (FAST) and pre-annealed at temperatures of 1050, 1250, 1473 K was studied by in-situ thermal desorption spectroscopy (TDS). D incorporation in W-11.4Cr-0.6Y alloy was carried out by irradiation with deuterium ions (670 eV/D+) to a fluence of 5 × 1019 D/m2. The change in Cr concentration in the bulk of the W-Cr-Y alloy was studied using energy dispersive spectroscopy (EDS). The elemental composition on the surface of alloy samples before and after annealing was investigated using low energy ion scattering spectroscopy (LEIS). The Cr concentration increases on the surface of alloy sample during annealing at 1000 and 1050 K due to intensive Cr segregation, while it decreases at temperatures of 1250 K and above by dominant Cr sublimation. Pre-annealing of W-Cr-Y alloy at 1050 K leads to a decrease in the D trapping in the alloy which is connected with an increase in the Cr concentration on the surface. At high annealing temperatures of 1250 K and above, the appearance of new phases and decomposition-induced grain refinement leads to new types of defects with high binding energy for D.

Modelling of deuterium diffusion and thermal desorption in tungsten exposed to high-flux deuterium plasma

In nuclear fusion, hydrogen isotopes transportation in tungsten has been identified as a crucial process which could affect various key processes in the operation of a fusion device, such as tritium self-sustaining and operational safety. To enhance the understanding of deuterium diffusion and trapping in millimeter-thick tungsten under high-flux plasma conditions, we carried out a systematic simulation investigation on the deuterium plasma exposure and subsequent thermal desorption using the TMAP7 code based on experimental results. The simulation reveals that the defect concentration with a low de-trapping energy , as estimated solely from simulating thermal desorption, is lower than its actual density due to the temperature history of plasma exposure and subsequent cooling processes. Moreover, the growth rate of deuterium mobile concentration exhibits nonmonotonic behavior over time during plasma exposure, which strongly depends on the ratio of defect concentration with a high de-trapping energy to that with a low de-trapping energy. Additionally, we observed that the maximum deuterium diffusion depth does not follow a proportional relationship with the square root of exposure time, which is attributed to the growing number of plasma-induced defects and the reducing injection coefficient with increasing fluences. This work provides valuable insights into the understanding of deuterium transportation in tungsten during high-flux plasma exposure which could contribute to understanding and modelling of hydrogen isotopes recycling in future fusion devices.

DEUTERIUM RETENTION IN N-SEEDED AND UNSEEDED TUNGSTEN LAYERS OBTAINED WITH DCMS AND M-HIPIMS TECHNIQUE

The retention of hydrogen isotopes in tungsten co-deposited layers is a subject severely understudied in the literature compared to hydrogen isotope implantation in bulk tungsten specimens. However, impurity seeding in future next-generation fusion devices can reduce the sputtering threshold of tungsten and potentially increase the contribution of the tungsten layers to the global tritium retention problem. This paper reports on the development of fusion-relevant tungsten-deuterium and nitrogen-seeded tungsten deuterium layers using direct current (DCMS) and multipulse high-power impulse magnetron sputtering techniques (HiPIMS). The four layers resulting from co-deposition in argon-deuterium and argon-deuterium-nitrogen atmosphere were investigated from the deuterium retention and release point of view. Supplementary the crystalline structure of all samples and particularly the composition of the nitrogen-seeded layers were also studied. The tungsten-deuterium layers produced using m-HiPIMS technology showed a higher crystalline coherence compared to layers deposited with DCMS in part owned to the enhanced ion bombardment of the forming layers during deposition. On the other hand, the nitrogen-seeded layers showed without exception the formation of the tungsten-nitride compound. Here on the contrary the crystalline coherence as revealed by X-ray diffraction measurements measurement showed that nitride formation is favored for DCMS compared with m-HiPIMS. The elemental areal densities of the layer constituents were analyzed using Rutherford Backscattering Spectrometry. In this case, only the two nitrogen-seeded layers were analyzed revealing that a higher nitrogen content by a factor of ~2 is retained in DCMS deposited layer compared to HiPIMS.The release of deuterium was higher in the case of m-HiPIMS layers and it was shown that despite the deposition method the release behavior occurs predominantly at temperatures above 800 K suggesting a high energy trapping of deuterium in tungsten layers. Nitrogen seeding influences both the retention mechanisms and increases the deuterium retention by two orders of magnitude in seeded layers compared to unseeded ones.

Deuterium retention in W-Cr-Y alloy: Impact of the manufacturing method and helium presence

The retention of deuterium in tungsten-chromium-yttrium alloys, W-11.4Cr-0.6Y and W-10Cr-0.5Y, was investigated by means of in-situ thermal desorption spectroscopy (TDS). The first alloy was manufactured by field-assisted sintering technology (FAST), while W-10Cr-0.5Y alloy was produced by hot isostatic pressing followed by heat treatment (HIP+HT). Both alloys were irradiated with D3+ ions (670 eV/D) and a fluence of 1021 D/m2 at temperatures of 300, 600, 900 K. In the case of W-11.4Cr-0.6Y alloy, deuterium retention was investigated during sequential irradiation with helium (3 keV) and deuterium (670 eV/D) ions at room temperature with fluences of 1019–7 × 1022 He/m2 and 1021 D/m2, respectively.The material structure has a great influence on deuterium retention, as evidenced by the significantly increased deuterium trapping in W-11.4Cr-0.6Y alloy at both room and elevated irradiation temperatures due to the presence of high-energy trapping centers compared to W-10Cr-0.5Y alloy. In both alloys, the observed amount of deuterium released during TDS was higher than in pure tungsten.The general trend of helium influence on the deuterium retention in W-11.4Cr-0.6Y alloy is very similar to bare tungsten. Namely, deuterium retention increased until helium implantation reached a fluence of 1021He/m2, and then sharply decreased when the fluence exceeded this value.

Deuterium trapping behavior in tungsten surface due to low-energy ion irradiation

Deuterium (D) irradiation at 421–597 K was performed on polycrystalline tungsten with 100 eV D ions up to a fluence of 4.64 × 1023 m−2. A supersaturated layer in the ion stopping range (< 20 nm) with the concentration of 1.2 at.% under the present surface resolution of ∼ 12 nm, a shallow minimum D concentration area of > 100 nm and a subsequent significant increase in the sub-surface were revealed by combining D (12C, D) 12C elastic recoil detection and D (3He, p) 4He nuclear reaction analysis. The first-principle calculations indicated that the trapping of vacancies for hydrogen (H) atoms can greatly reduce the recombination probability of Frenkel pairs and then act to stabilize the Frenkel pair. The stabilizing effect can be enhanced with increasing number of H atoms. Furthermore, the vacancies from Frenkel pairs can trap more surrounding H atoms, resulting in further H enrichment which could account for the formation of supersaturated layers. The typical location of the blister cavities coincides rather well with the subsequent significant increase in the sub-surface area. D-induced damage in the supersaturated layer and the sub-surface layers was examined by combining scanning electron microscope, transmission electron microscope and positron annihilation measurements.

The effect of pre-damage distribution on deuterium-induced blistering and retention in Tungsten

Heavy ions have been widely used to simulate the effect of neutron irradiation-induced crystal defects on the plasma-wall interaction in the future fusion device. In this work, energetic argon ions were utilized to damage tungsten with the different pre-damage distributions. The pre-damage distribution is controlled by tuning ion energy and fluence. After exposure to deuterium plasma, the pre-damaged tungsten targets were analyzed to explore the impact of pre-damage distribution on deuterium-induced surface blistering and retention. Surface observation results reveal that increasing pre-damage depth and peak dpa significantly reduce the surface blister's area density, especially in diameters below ten micrometers. Meanwhile, thermal desorption spectra indicate that the given change in the pre-damage distribution barely change the type of irradiation-induced defects. The role of pre-damage increases in deuterium trapping and aggregation is discussed based on the experimental results.

EION-IMPLANTED DEUTERIUM RELEASE BEHAVIOR FROM Li-BASED ADVANCED CERAMICS

The behavior of ion-implanted deuterium in advanced lithium orthosilicate (Li4SiO4) pellets with addition of lithium metatitanate (Li2TiO3), and in reference Li4SiO4 pellets, has been studied. Thermal desorption spectroscopy (TDS) was used to examine the deuterium interaction with radiation defects in materials. Deuterium gas release from studied ceramics has revealed a clear dependence of deuterium trapping behavior on phase composition. The trap strength of radiation defects formed in lithium titanate was found to be higher than in lithium orthosilicate.

Effects of self-irradiation on deuterium retention and reflectivity of molybdenum, fusion plasma-facing material: Combined experimental and modeling study

Molybdenum is used as plasma-facing material in tokamaks and as material for plasma optical diagnostics mirrors. Harsh conditions of neutron irradiation, exposure to hydrogen isotopes and helium ions, and high operating temperatures result in degradation of the molybdenum surface and ultimately limit their lifetime in a fusion power plant. In the current paper, intake and subsequent thermal release of deuterium from self-irradiated by high energy (1 MeV) ions molybdenum as a function of irradiation dose are investigated. Several characteristic temperature regions where deuterium release takes place are identified and attributed to trapping of deuterium in intrinsic and radiation-induced microstructure defects. This attribution is further validated by molecular dynamics modeling, which confirms that increase and saturation of vacancy concentration found in simulations follows increase and saturation of experimentally determined deuterium content. Deuterium inventory and vacancy content saturate at a damage level of around 0.2 dpa (displacement per atom), similar to recent modeling and experimental studies of iron and tungsten. Reflectivity measurements of irradiated molybdenum show that it is only slightly affected by damage up to 1 dpa.

EFFECT OF IRRADIATION AND DEUTERIUM BEHAVIOR IN LI-BASED TRITIUM BREEDING CERAMICS

The behavior of implanted deuterium in advanced lithium orthosilicate (Li4SiO4) pellets with addition of lithium metatitanate (Li2TiO3), and in reference Li4SiO4 pellets, has been investigated. Thermal desorption (TD) spectroscopy was used to study the deuterium interaction with radiation defects in materials. Computational evaluation of deuterium desorption within the framework of the diffusion-trapping model allowed to associate characteristics of experimental TD spectra with specific trapping sites in the material. It was found that deuterium desorption is limited mainly by intragranular diffusion of deuterium and its trapping by radiation defects associated with Li-vacancy traps. Deuterium gas release from both ceramics demonstrated similar trend indicating weak dependence of deuterium trapping behavior on phase composition. The change of the morphology and elemental composition of the pellets surface has been analyzed. SEM examination indicated that ion irradiation and subsequent thermal desorption annealing of two-phase ceramics leads to an increase in surface destruction processes.

Effect of noble-gas bubbles on deuterium trapping behavior in argon pre-implanted stainless steel

Ion implantation technique and nuclear reaction analysis were used to study the trapping behavior of deuterium by radiation defects in 316 austenitic stainless steel. Defect structure with a different ratio of displacement per atom (dpa) to argon gaseous impurity along with a depth has been formed by pre-irradiation of SS316 with 1.4 MeV argon ions. The deuterium migration between layers containing various defect types served as an indicator of traps strength. Computational evaluation of deuterium trapping process within the framework of the diffusion-trapping model has shown that gas-bubbles associated traps turn out to be a strongest deuterium traps with a binding energy of 0.74 eV. Other irradiation-induced defect types have a weaker effect on deuterium retention in austenitic steel. Experimental observations and calculation results are discussed with respect to the possibility of molecular hydrogen formation in radiation-induced nanovoids under operating conditions of water-cooled reactor internals.

Effect of titanium doping on deuterium behaviour in tungsten under low-energy deuterium plasma irradiation

Surface modification, deuterium trapping, microstructure, hardness of tungsten – titanium alloys (W-5Ti) and reference tungsten (W) were investigated. All the specimens were exposed to deuterium plasma (100 eV) for 2 h and 4 h with a density of 1.8 × 1021 D (m−2 s−1) at 470 K. Surface blistering and D retention in tungsten is strongly suppressed by Ti doping. Two typical regions are observed on W-5Ti alloys, namely, W–Ti solid solute area and Ti-enriched region. Most of blisters appeared on Ti-enriched area while no blisters were observed on the W–Ti solid solute region. In addition, the hardness of W is significantly improved by Ti-doping.

Reaction–diffusion simulations of hydrogen isotope trapping and release from cavities in tungsten, I: Single cavity

We present simulations of deuterium thermal desorption spectra (TDS) from a single nm-sized spherical cavity in tungsten. We study both D2 gas-filled cavities and cavities with trapping only at surface sites. The simulations are based on the diffusion theory and the kinetic model of deuterium interaction with metal surfaces. We show that the previously used approaches based on local thermodynamic equilibrium between D2 gas and the subsurface solute deuterium are inadequate for simulating TDS spectra. We demonstrate the influence of cavity radius, D2 gas pressure, the parameters of the surface and the bulk. The deuterium release rate from a gas-filled cavity at a fixed temperature (below the TDS peak) stays constant until the D2 gas is depleted, in contrast to exponential decay observed for conventional deuterium trapping sites. Using the simulations and a simplified analytic model we demonstrate that for cavities the dependence of the TDS peak position on the heating rate does not follow exactly the Kissinger equation.

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.

An inertial electrostatic confinement fusion system based on graphite

Inertial electrostatic confinement (IEC) devices use concentric electrodes to accelerate ions to sufficient energies to produce nuclear fusion. In a previous publication, we have indicated that, when operating at low power, fusion events largely occur when high energy ions impact neutral molecules that are adsorbed on the cathode surface. The selection of the cathode material therefore plays an important role in determining the absolute fusion output of an IEC machine. A study is presented in which a pair of matching IEC cathodes were constructed from 316 stainless steel and graphite and the fusion characteristics of the grids examined as a function of system pressure and discharge power. Graphite is shown to be an excellent cathode material, producing fusion rates 2.2–4 times that of stainless steel. Due to the excellent deuterium trapping properties of graphite, it is likely this enhancement factor will continue to grow as operating power is further increased.

Identification by deuterium diffusion of a nitrogen-related deep donor preventing the p-type doping of ZnO

Deuterium diffusion is investigated in nitrogen-doped homoepitaxial ZnO layers. The samples were grown under slightly Zn-rich growth conditions by plasma-assisted molecular beam epitaxy on m-plane ZnO substrates and have a nitrogen content [N] varied up to 5 × 1018 at cm−3 as measured by secondary ion mass spectrometry (SIMS). All were exposed to a radio frequency deuterium plasma during 1 h at room temperature. Deuterium diffusion is observed in all epilayers, while its penetration depth decreases as the nitrogen concentration increases. This is strong evidence of a diffusion mechanism limited by the trapping of deuterium on a nitrogen-related trap. The SIMS profiles are analyzed using a two-trap model including a shallow trap, associated with a fast diffusion, and a deep trap, related to nitrogen. The capture radius of the nitrogen-related trap is determined to be 20 times smaller than the value expected for nitrogen–deuterium pairs formed by coulombic attraction between D+ and nitrogen-related acceptors. The (N2)O deep donor is proposed as the deep trapping site for deuterium and accounts well for the small capture radius and the observed photoluminescence quenching and recovery after deuteration of the ZnO:N epilayers. It is also found that this defect is by far the N-related defect with the highest concentration in the studied samples.

Deuterium trapping in the subsurface layer of tungsten pre-irradiated with helium ions

The effect of He-induced defects in tungsten on the efficiency of trapping of deuterium ions in the subsurface layer was studied using thermal desorption spectroscopy (TDS). The W sample was pre-irradiated with 3 keV helium ions at room temperature and various fluences in the range of 1019 – 5 × 1021 He/m2. Then, it was exposed to a probe fluence of 1019 D/m2 of 2 keV D3+ (670 eV/D) ions, and in-situ TDS was performed. The de-trapping energy for D atoms increased with the increase of the He pre-irradiation fluence. On the other hand, a strong decrease in the D retention was observed if the He fluence increased above 1021 He/m2. At the highest He fluence of 5 × 1021 He/m2 deuterium trapping was possible only after partial release of He atoms. By comparison of experimental TDS spectra with modeling, the de-trapping energies of D atoms from various defects were estimated.

Studies on the near-surface trapping of deuterium in implantation experiments

Surface-shifted deuterium profiles are re-examined in deuterium-ion irradiation experiments by using a combined experimental and modelling approach. Recrystallized tungsten foil samples were irradiated with energetic deuterium ions and the defect and deuterium depth profiles were studied using positron annihilation spectroscopy and secondary ion mass spectroscopy. We report direct experimental evidence of trapping of deuterium at the vacancies created by the deuterium ions themselves during the implantation by using positron annihilation studies. The deuterium profile is simulated using a Monte-Carlo diffusion model by taking into account the defect-aided diffusion of deuterium due to the local strain field created by the vacancies. The simulations also elucidate the role of the anisotropy in the diffusion and trapping of deuterium in ion-implantation experiments in metals.

Post-mortem analyses of gap facing surfaces of tungsten tiles of T-10 ring limiter

• The entire surface of gaps was covered by Li evaporated and splashed from Li limiter. • Heavy carbide precipitates were found deep in narrow gaps. • Deuterium and helium are accumulated in the tiles due to ion implantation. • Mainly regular but not cleaning discharges were responsible for effects in gaps. • High frequency auto-oscillating discharges were suggested to play a role in processes in gaps. Surfaces facing the gap between W tiles of the ring limiter of tokamak T-10 were analyzed after T-10 decommissioning using LIBS, SEM/EDA, XRD, TDS, and NRA techniques. Gaps with the width of 5 mm and 0.1 mm were nearly completely covered to their full depths of 22 and 15 mm, respectively, by a deposited film. The film was formed mainly by deposition of lithium that came from Li limiter and transformed in air to Li 2 CO 3 and Li 2 O. Carbon was deposited from volatile hydrocarbons sputtered from the tokamak walls. Besides, carbon appeared due to chemical reaction with lithium in air. Chemical interactions of W with C, O, and Li led to formation W 2 C, WC, WO 2 , and Li 2 WO 4 . Carbides formed in W over the entire surface to the full depth of the gaps. Trapping of deuterium and helium in tiles was demonstrated. Possible influence of auto-oscillating discharges on ionization and ion trapping of C,D, and He in gaps is discussed.