Hydrogen Isotope Retention Research Articles

Deuterium addition to liquid Li–Sn alloys: implications for plasma-facing applications

Liquid metals are being explored actively as candidates for plasma-facing components (PFCs) in fusion reactors. Recently, Li–Sn alloys have appeared as promising alternatives that could overcome some of the challenges faced by the well-studied liquid Li system, namely, a vapor pressure that limits the operating temperature and a high hydrogen isotope retention. However, only scarce data (experimental or theoretical) are available concerning the performance of Li–Sn alloys, specifically only for the compositions of Li30Sn70 and Li20Sn80, related to their bonding and retention of deuterium (D). Here, we present a comprehensive, first-principles molecular-dynamics study of static and dynamic properties of liquid Li30Sn70 at various D concentrations. We observe the formation of D2 gas bubbles for β in Li30Sn70Dβ greater than 22.5 along with Li segregation towards D2 bubbles. To understand the effect of Sn addition on D retention in Li–Sn alloys, we perform a thermodynamic evaluation of maximum D retention in Li-rich Li–Sn alloys. Overall, this work will provide useful data and guidance in the development of Li–Sn PFCs in fusion reactors.

Finite element analysis of hydrogen retention in ITER plasma facing components using FESTIM

Abstract The behaviour of hydrogen isotopes in ITER monoblocks was studied using the code FESTIM (Finite Element Simulation of Tritium In Materials) which is introduced in this publication. FESTIM has been validated by reproducing experimental data and the Method of Manufactured Solutions was used for analytical verification. Following relevant plasma scenarios, both transient heat transfer and hydrogen isotopes (HIs) diffusion have been simulated in order to assess HIs retention in monoblocks. Relevant materials properties have been used. Each plasma cycle is composed of a current ramp up, a current plateau, a current ramp down and a resting phase before the following shot. 100 cycles are simulated. The total HIs inventory in the tokamak during resting phases reaches 1.8 × 10 − 3 mg whereas during the implantation phases it keeps increasing as a power law of time. Particle flux on the cooling channel of the monoblock is also computed. The breakthrough time is estimated to be t = 1 × 10 5 s which corresponds to 24 cycles. Relevance of 2D modelling has been demonstrated by comparing the total HIs inventory obtained by 2D and 1D simulations. Using 1D simulations, a relative error is observed compared to 2D simulations which can reach -25% during the resting phase. The error during implantation phases keeps increasing.

Secondary Effect of Nitrogen on Hydrogen Isotopes Retention in Tungsten Dust from Fusion Reactor Simulated by Wire Electrical Explosion

Secondary Effect of Nitrogen on Hydrogen Isotopes Retention in Tungsten Dust from Fusion Reactor Simulated by Wire Electrical Explosion

Time-resolved laser-induced desorption spectroscopy (LIDS) for quantified in-situ hydrogen isotope retention measurement and removal from plasma facing materials.

A laboratory scale laser induced thermal desorption spectroscopy system is developed and tested on tungsten-deuterium and titanium-deuterium codeposits, and its feasibility as a hydrogenic inventory measurement diagnostic is demonstrated over a range of retention values from 5 × 1019 m-2 to 7 × 1023 m-2 for absorbed laser power densities as low as 8 MW m-2. Codeposit layer samples are grown by magnetron sputtering and immersed in a weak argon rf plasma. A 1 kW fiber laser (λ = 1100 nm) heats the samples up to a peak surface temperature ranging from 900 to 1500 K using pulse widths of 0.5 and 1 s. Spectral line emission from Balmer series Dα and Hα from thermally desorbed deuterium and hydrogen, as well as line emission from argon, are monitored as a function of time using an optical spectrometer with maximum temporal resolution of 1 ms. To correct for wall recycling and pumping speed, and to accurately measure the time evolution of the laser-induced thermal desorption, the raw Dα signal is deconvolved with the system response function, which is obtained by injecting a short burst of D2 to approximate an impulse. Calibration is done with a standard D2 leak, and laser induced desorption spectroscopy deuterium retention values are found to be in good agreement with companion measurements made using conventional temperature programmed desorption on samples from the same codeposit batch.

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.

Hydrogen isotope permeation and retention behavior in the CoCrFeMnNi high-entropy alloy

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

Investigation of hydrogen isotope retention mechanisms in beryllium: High resolution TPD measurements

Abstract The retention of ion-implanted deuterium in beryllium poly- and single crystals at room temperature is studied using high precision temperature programmed desorption spectroscopy (TPD). Slow temperature ramps of 0.01 K/s in combination with well-defined experimental conditions are used to resolve the low temperature desorption regime for the first time revealing three sharp desorption peaks. The comparison to results of a coupled reaction diffusion system (CRDS) model shows, that the corresponding release mechanisms cannot be described by thermally activated rate processes. SEM images of a polycrystalline beryllium sample after implantation of deuterium with 2 keV per D atom show the formation of blisters of roughly 1 µm in diameter. Additionally, cracks on top of the blisters are found as well as spots, on which blisters are peeled off. Both processes are discussed to play a role in the low temperature release regime of the retained deuterium. Investigation of TPD spectra performed on single crystalline beryllium shows a jagged pattern in the low temperature release regime, which can be connected to blisters bursting up, releasing big amounts of deuterium in short time scales.

Effect of carbon impurity reduction on hydrogen isotope retention in QUEST high temperature wall

The W (tungsten) samples were placed at top, equator and bottom walls of QUEST (Q-shu University Experiment with Steady-State Spherical Tokamak) device and exposed 1238 shots of hydrogen plasma during 2016A/W (Autumn/Winter) campaign with normal wall temperature of 473 K (maximum temperature of 523 K). Thereafter, the surface morphology was evaluated by color measurement, TEM (Transmission Electron Microscope) and XPS (X-ray photoelectron spectroscopy). Thick deposition layers were formed on the samples placed at the equator and bottom walls. On the other hand, thin mixed material layer was deposited on the top wall, where large H (hydrogen) retention was observed, which would be caused by dynamic plasma wall interaction (erosion and deposition) with higher H flux. Low H retention was confirmed for bottom wall, where higher wall temperature without He discharge would contribute. The additional 1 keV D2+ was implanted into these samples and deuterium retention enhancement was estimated. It was clearly found that the irradiation damages would induce more deuterium trapping than the formation of C–D bond.

Deuterium retention in advanced steels for fusion reactor structural application

Reduced activation ferritic-martensitic (RAFM) steels, castable nanostructured alloys (CNA), and oxide-dispersion-strengthened (ODS) steels have been developed for the fusion reactor structural application with improved mechanical properties and radiation resistance. However, the hydrogen isotope retention in these steels is not well understood. In this study, 10 keV D2 implantation was performed on six steels, together with pure iron and T91 as references, at room temperature. Subsequent thermal desorption spectroscopy (TDS) measurements were carried out to investigate the deuterium retention. Possible deuterium trapping sites in the studied materials were discussed in coordination with available microstructural information. The results indicated that ODS-steels have the largest deuterium retention, which is almost one order of magnitude higher than that of pure iron, ∼6 times of RAFM, and 2–4 times of CNA steels. The difference in nanoparticles and grain/lath boundaries are primarily responsible for the different deuterium retention, while other trapping sites (e.g., vacancy clusters, dislocations, etc.) are present. Deuterium retention increases with increasing sink strength of the studied materials, although they are not linear correlated.

Molecular dynamics study of trapping and detrapping process of hydrogen in tungsten vacancy

Tungsten (W) alloys and W-based alloys are the primary candidate materials for plasma-facing components in future fusion reactors (e.g. ITER and CFETR). One of the critical issues still to be clarified in the design of the fusion reactor materials is the retention of hydrogen (H) isotopes in W, when the plasma-facing materials are supposed to sustain high-flux plasma and high-energy neutron. The dynamical behaviours of H in W with radiation defects (e.g. vacancy) are of serious concerns for understanding the mechanism of H capture, retention and permeation in W. In this work, a new model to extract the effective capture radius (ECR) and dissociation coefficient simultaneously is presented through coupling the trapping process and detrapping process of H in W vacancy. In the new model, the quantity ratio of vacancy to H atom in vacancy-H complex (VH<sub><i>x</i>+1</sub>) in the molecular dynamics (MD) simulations is described as a function of time, while the exact occurrence time of corresponding event is not required. This new model, combined with extensive MD calculations, enables the simultaneous determining of the ECR and dissociation coefficient of H in W vacancy. It is found that the parameters are dependent not only on the event type but also on temperature. The dissociation energy of H from vacancy-H complex decreases gradually with the increase of the trapped number of H atoms in the vacancy-H complex. It is also found that the common assumption (i.e. the ECR is equal to one lattice constant and the pre-exponential factor is equal to 10<sup>13</sup> s<sup>–1</sup>) in the long-term simulation methods (e.g. kinetic Monte Carlo and rate theory) is not always valid, since these calculated dynamical parameters are dispersive. The new model to obtain more reliable results with lower cost of computing resources can be easily extended into the other similar kinetic processes (e.g. H/He trapping and detrapping processes in other materials systems). These calculated dynamical parameters should be potentially helpful in supplying the initial input parameters for the long-term simulation methods.

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.

Hydrogen isotope trapping and retention in graphite and boron carbide under consecutive irradiation by deuterium and hydrogen plasma

The results of study on hydrogen and deuterium trapping and retention in MPG-8 grade graphite and boron carbide coating under consecutive irradiation by deuterium and hydrogen plasma in varying hydrogen irradiation conditions are presented in this paper. It is shown that deuterium content decreases both in graphite and boron carbide under irradiation by hydrogen plasma. It is also shown that the main mechanism of deuterium removal is sputtering for graphite, and isotope exchange for boron carbide.

Hydrogen diffusion behavior in tungsten under anisotropic strain

In future fusion devices, the incident hydrogen plasma with high mobility can diffuse deep into tungsten bulk, which is directly relevant with hydrogen isotopes permeation and retention in tungsten. In this work, density functional theory (DFT) and object kinetic Monte Carlo (OKMC) simulations are adopted to investigate the hydrogen diffusion behavior in tungsten under anisotropic uniaxial strain from −2.5% to 2.5%. As presented by our DFT calculations, there are two types of hydrogen diffusion paths when applying strain, including one path perpendicular to the strain direction and another path largely along the strain direction. The migration energy barriers of these two paths have opposite variation tendencies in tensile or compressive condition. Our OKMC calculations based on DFT input show that, in tensile condition, the hydrogen diffusion is restrained despite the lower energy barrier of the corresponding diffusion path. In compressive condition, the hydrogen diffusion along the strain direction is enhanced, while that perpendicular to the strain direction is suppressed. The hydrogen diffusivity under anisotropic strain at the temperature range from 400 K to 1800 K is determined. It is demonstrated that tensile strain can suppress the diffusivity, while compressive strain can either suppress or facilitate the diffusivity depending on the temperature and the strain value. The anisotropic strain exhibits distinct effect on hydrogen diffusivity at lower temperature but its effect is minimal as the temperature increases.

Correlation of surface chemical states with hydrogen isotope retention in divertor tiles of JET with ITER-Like Wall

To understand the fuel retention mechanism correlation of surface chemical states and hydrogen isotope retention behavior determined by XPS (X-ray photoelectron spectroscopy) and TDS (Thermal desorption spectroscopy), respectively, for JET ITER-Like Wall samples from operational period 2011–2012 were investigated. It was found that the deposition layer was formed on the upper part of the inner vertical divertor area. At the inner plasma strike point region, the original surface materials, W or Mo, were found, indicating to an erosion-dominated region, but deposition of impurities was also found. Higher heat load would induce the formation of metal carbide. At the outer horizontal divertor tile, mixed material layer was formed with iron as an impurity. TDS showed the H and D desorption behavior and the major D desorption temperature for the upper part of the inner vertical tile was located at 370 °C and 530 °C. At the strike point region, the D desorption temperature was clearly shifted toward higher release temperatures, indicating the stabilization of D trapping by higher heat load

On the Trapping and Retention of Hydrogen Isotopes in Graphite under Sequential Hydrogen Plasma Irradiation

The results of studying the trapping and retention of hydrogen and deuterium in MPG-8 graphite under sequential irradiation with ions of deuterium and hydrogen plasma are presented. The effect of the energy and density of the hydrogen-ion flux, the impurity of oxygen in the plasma, and the temperature of the irradiated samples on a decrease in the deuterium concentration and hydrogen trapping in graphite is studied. Based on the analysis of experimental data, it is suggested that the observed effect of deuterium removal can be explained by the chemical sputtering of graphite.

Retention and release of hydrogen isotopes in carbon materials priorly charged in gas phase

Thermal desorption of deuterium from isotropic graphites and a CFC material priorly charged with deuterium gas has been investigated to obtain quantitative information on hydrogen recycling and tritium inventory in fusion experimental devices and future reactors. The spectra of thermal desorption spectrometry (TDS) appear to have five peaks, which are defined as peaks 1–5 in order of the increase of temperature. Diffusion coefficients of deuterium for the absorption process are evaluated for the samples charged with deuterium gas for various absorption times. Absorption pressure dependence of TDS spectra shows a unique change in the temperatures of peak 2. As for the absorption temperature dependence of TDS spectra, peak 4 has a drastic change in the shape of spectra and also in the peak temperatures. The mechanism of desorption for peak 5 will be associated to a detrapping process with a dispersion of energies in the range of 4.0–4.9 eV.

Investigation of hydrogen bubbles behavior in tungsten by high-flux hydrogen implantation

Hydrogen isotopes retention and bubbles formation are critical issues for tungsten as plasma-facing material in future fusion reactors. In this work, the formation and growing up behavior of hydrogen bubbles in tungsten were investigated experimentally. The planar TEM samples were implanted by 6.0keV hydrogens to a fluence of 3.38×1018H⋅cm−2 at room temperature, and well-defined hydrogen bubbles were observed by TEM. It was demonstrated that hydrogen bubbles formed when exposed to a fluence of 1.5×1018H⋅cm−2, and the hydrogen bubbles grew up with the implantation fluence. In addition, the bubbles' size appeared larger with higher beam flux until saturated at a certain flux, even though the total fluence was kept the same. Finally, in order to understand the thermal annealing effect on the bubbles behavior, hydrogen-implanted samples were annealed at 400, 600, 800, and 1000 °C for 3 h. It was obvious that hydrogen bubbles' morphology changed at temperatures higher than 800 °C.

Thermal Desorption Study of the Trapping and Retention of Hydrogen Isotopes upon Irradiation of Oxidized Metal Surfaces with Hydrogen Plasma

Patterns of hydrogen isotope trapping and retention are investigated by applying hydrogen ion radiation to nickel samples free of oxidation, tungsten samples with oxide layer on their surfaces, and tungsten samples with aluminum coating with oxide layer. It is found that hydrogen isotope desorption upon hydrogen ion plasma irradiation occurs in samples of tungsten and tungsten with aluminum coating, but not in nickel samples. It is concluded that hydrogen isotope transport through the tungsten–aluminum interface is initiated when one surface of a samples is irradiated with hydrogen plasma ions.

Studies on synthesis of plasma fusion relevant tungsten dust particles and measurement of their hydrogen absorption properties

In this communication, we identify some thermal-plasma assisted synthesis techniques for the alternative production of fusion relevant tungsten dust particles in bulk amounts, to allow for the statistically meaningful study of their hydrogen absorption properties. Fast, single step synthesis of fine particles in the α-tungsten phase was demonstrated, using a high heat flux device for sizes in the micrometer region and an experimental nanoparticle reactor for dust with nanometer sizes. Production of both equilibrium polyhedral and nonequilibrium spherulitic crystal shapes was demonstrated, dust morphologies similar to which were identified before in existing tokamaks and divertor simulator systems. Their hydrogen retention characteristics were measured by the Nuclear Resonance Reaction Analysis technique, which was found to be orders of magnitude higher compared to the same metal in the bulk form. The mesoporous crystals generated during high power, high pressure synthesis conditions were assumed to be responsible for the unusually high gas retention of the nanoparticle sample. It was concluded that non-equilibrium tungsten dust morphologies would be dominant in the high power fusion machines in the future, which may aggravate the hydrogen isotope retention issues further.

Structural defect accumulation in tungsten and tungsten-5wt.% tantalum under incremental proton damage

We have performed proton irradiation of W and W-5wt.%Ta materials at 350 °C with a step-wise damage level increase up to 0.7 dpa and using two beam energies, namely 40 keV and 3 MeV, in order to probe the accumulation of radiation-induced lattice damage in these materials. Interstitial-type a/2 <111> dislocation loops are formed under irradiation, and their size increases in W-5Ta up to a loop width of 21 ± 4 nm at 0.3 dpa, where loop saturation takes place. In contrast, the loop length in W increases progressively up to 183 ± 50 nm at 0.7 dpa, whereas the loop width remains relatively constant at 29 ± 7 nm at >0.3 dpa, giving rise to dislocation strings. The dislocation loops and tangles are observed in both materials examined after a 3 MeV proton irradiation at 350 °C. Ta doping delays the evolution of radiation-induced dislocation structures in W, and can consequently impact the hydrogen isotope retention under plasma exposure.