Effect of multi-occupancy traps on the diffusion and retention of multiple hydrogen isotopes in irradiated tungsten and vanadium

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

Retention of hydrogen isotopes is a critical concern for operating fusion reactors as retained tritium both activates components and removes scarce fuel from the fuel cycle. Radiation-induced displacement damage in SiC influences the retention of hydrogen isotopes compared to pristine SiC. Deuterium retention in neutron irradiated high purity SiC has been compared to different microstructures of non-irradiated high purity SiC using thermal desorption spectroscopy after gas charging and low energy ion implantation. Experimental results show lower deuterium retention in single crystal SiC than in polycrystal SiC indicating that grain boundaries are key trapping features in unirradiated SiC. Deuterium is released at lower temperatures in neutron irradiated polycrystal SiC compared to pristine polycrystal SiC, suggesting weaker trapping by radiation-induced defects compared to grain boundary trapping sites in the pristine materials. Low energy ion implantation caused a high deuterium release temperature, highlighting the sensitivity of deuterium release behaviour to radiation defect characteristics. First principles calculations have been conducted to identify energetically favourable trapping sites in SiC at the HABcVSi and HTSiVC complexes, and migration barriers between interstitial sites. This helps interpret experimental results and derive effective diffusivity of hydrogen isotopes in SiC in the presence of vacancies.

For fusion test reactors and power plants, one significant concern is the retention of hydrogen isotopes in the wall materials. The build-up of the radioactive and scarce fuel isotope tritium is of special concern, but knowing the retention of the other isotopes, such as deuterium, is also important. Deuterium is known to affect the mechanical properties of the wall material and most experiments are carried out on deuterium retention as it is safer to use than tritium. In addition to affecting the mechanical properties of the wall material, deuterium retention has been observed to affect the defect accumulation in the material. In this study, we investigate the phenomena and mechanisms responsible for the greater defect accumulation observed in tungsten when deuterium is present during irradiation. This is achieved computationally, utilizing molecular dynamics simulations and appropriate analysis tools. We found that deuterium will affect both the primary defect production as well as the recombination rate of defects in irradiated tungsten.

To investigate the effect of blistering on hydrogen isotope (HI) retention, a series of deuterium plasma exposures were performed using recrystallized tungsten samples at 500 K with high fluences up to 1.0 × 1028 ions m−2 in the linear plasma device STEP. An increase of blister density and deuterium retention was observed with increasing plasma fluence. Based on the simulation of the thermal desorption spectra using TMAP, defects with different detrapping energies are found to be located at a depth of tens of microns, which coincides with the depth of the grain boundaries (GBs) close to the surface. The defect characterizations using transmission electron microscopy and positron annihilation Doppler broadening identified the defects as dislocation type and vacancy type, which were created by blistering. It is suggested that these defects can diffuse deep into the material, and the interaction between the diffusion of the defects and GBs causes a peculiar deuterium desorption spectrum over plasma fluences. Additionally, these blister-induced defects are the main source of deuterium retention. Regarding the effect of the blister-induced defects on deuterium retention, a blister-dominated retention mechanism is proposed to describe HI retention in conditions when blistering is severe as in this study. This investigation provides a new insight into the effect of blistering on retention and the modelling of retention in a tokamak edge plasma environment.

Plasma fluence at the divertor of a future magnetic confinement fusion device can accumulate up to ∼1028–1029 m−2 per year. Yet hydrogen isotope (HI) retention under such high-fluence plasma exposure has been rarely reported. To investigate deuterium (D) retention in tungsten (W) exposed to such high-fluence plasma, a series of high-flux D plasma exposures were preformed using recrystallized W samples at ∼500 K in Magnum-PSI. The highest fluence achieved was ∼1 × 1029 m−2. Surface morphology observations indicate an initial increase in the number of blisters at the sample surface with increasing fluence, followed by saturation at ∼1 × 1029 m−2. Multiple bursts of blisters with open cracks or edges were observed under the two highest fluences of ∼1 × 1028 m−2 and ∼1 × 1029 m−2. 3He nuclear reaction analysis (NRA) shows a maximum D concentration up to 0.012 at.fr., distributed within the first 4 μm from the sample surface under the highest fluence. D retention, as measured by NRA and thermal desorption spectroscopy, tends to saturate with increasing fluence. Simulations of D2 thermal desorption, performed using the TMAP rate equation code, show a maximum D trapping depth of ∼10 μm, consistent with the defect depth profile revealed by transmission electron microscopy. D retention saturation observed in this work is attributed to the sample surface morphology modifications and the saturation of plasma-induced defects. This investigation provides a valuable reference for understanding the evolution of total HI retention in W under high-fluence plasma exposure in future fusion devices.

Suppression of vacancy formation and hydrogen isotope retention in irradiated tungsten by addition of chromium

Fusion fuel retention (trapping) and release (desorption) from plasma-facing components are critical issues for ITER and for any future industrial demonstration reactors such as DEMO. Therefore, understanding the fundamental mechanisms behind the retention of hydrogen isotopes in first wall and divertor materials is necessary. We developed an approach that couples dedicated experimental studies with modelling at all relevant scales, from microscopic elementary steps to macroscopic observables, in order to build a reliable and predictive fusion reactor wall model. This integrated approach is applied to the ITER divertor material (tungsten), and advances in the development of the wall model are presented. An experimental dataset, including focused ion beam scanning electron microscopy, isothermal desorption, temperature programmed desorption, nuclear reaction analysis and Auger electron spectroscopy, is exploited to initialize a macroscopic rate equation wall model. This model includes all elementary steps of modelled experiments: implantation of fusion fuel, fuel diffusion in the bulk or towards the surface, fuel trapping on defects and release of trapped fuel during a thermal excursion of materials. We were able to show that a single-trap-type single-detrapping-energy model is not able to reproduce an extended parameter space study of a polycrystalline sample exhibiting a single desorption peak. It is therefore justified to use density functional theory to guide the initialization of a more complex model. This new model still contains a single type of trap, but includes the density functional theory findings that the detrapping energy varies as a function of the number of hydrogen isotopes bound to the trap. A better agreement of the model with experimental results is obtained when grain boundary defects are included, as is consistent with the polycrystalline nature of the studied sample. Refinement of this grain boundary model is discussed as well as the inclusion in the model of a thin defective oxide layer following the experimental observation of the presence of an oxygen layer on the surface even after annealing to 1300 K.

The influence of irradiation conditions on the retention of hydrogen isotopes in structural materials (austenitic steel) under heating is considered. The specimens under study were irradiated either in a reactor or by bombarding them with hydrogen-isotope ions of variable fluence and energy at accelerators. An investigation of irradiated specimens with an EM-300 transition electron microscope was accompanied by studying the kinetics of hydrogen release from samples with a high-vacuum mass spectrometer. Also, the kinetics of hydrogen-isotope release from specimens of structural materials treated with a deuterium plasma was studied. It was found that, under the effect of irradiation, the materials being studied develop radiation defects, which appear to be efficient traps for hydrogen atoms, retaining them up to rather high temperatures (650 K). It is also shown that blisters formed in the materials treated with a hydrogen plasma contain both molecular hydrogen and hydrocarbons—in particular, methane.

Hydrogen isotope permeation and retention behavior in the RAFM steel manufactured by laser powder bed fusion

Hydrogen isotopes retention in JT-60U

FOREWORD

The retention of hydrogen isotopes (HIs) (H, D and T) in the first, plasma exposed wall is one of the key concerns for the operation of future long pulse fusion devices. It affects the particle-, momentum- and energy balance in the scrape off layer as well as the retention of HIs and their permeation into the coolant. The currently accepted picture that is used for interpreting current laboratory and tokamak experiments is that of diffusion hindered by trapping at lattice defects. This paper summarises recent results that show that this current picture of how HIs are transported and retained in W needs to be extended: the modification of the surface (e.g. blistering) can lead to the formation of fast loss channels for near surface HIs. Trapping at single occupancy traps with fixed de-trapping energy fails to explain isotope exchange experiments, instead a trapping model with multi occupancy traps and fill level dependent de-trapping energies is required. The presence of interstitial impurities like N or C may affect the transport of solute HI. The presence of HIs during damage creation by e.g. neutrons stabilises defects and reduces defect annealing at elevated temperatures.

Annealing of hydrogen trap sites in displacement-damaged EUROFER

Modelling deuterium release from tungsten after high flux high temperature deuterium plasma exposure

Typical materials for components, type 316 stainless steel (316-SS), were chosen as a sample and hydrogen isotope was charged by various methods, water adsorption, electrolysis and ion irradiation to elucidate hydrogen isotope behavior on/in SS. The chemical states of SS surface were studied by XPS and the hydrogen isotope retention and its desorption behavior were analyzed by TDS. Two types of surface finish, namely non-pretreated sample and pretreated sample by polish and annealing were prepared. It was found that the oxy-hydroxide and hydroxide were formed on the surface layer. The hydrogen isotope desorption stages consisted of three stages, namely the desorption stages from oxy-hydroxide, hydroxide and bulk hydrogen. A large amount of deuterium was trapped by the oxy-hydroxide layer for the non-pretreated sample with electrolysis. The hydrogen isotope trapping by this layer would have a large influence on the hydrogen isotope retention. The surface finish would be one of the effective improvement for decreasing its retention on SS.