Junling Chen

Clustering of H and He, and their effects on vacancy evolution in tungsten in a fusion environment

The behaviours of hydrogen and helium in tungsten are vitally important in fusion research because they can result in the degradation of the material. In the present work, we carry out density-functional theory calculations to investigate the clustering of hydrogen and helium atoms at interstitial sites, vacancy and small vacancy clusters (Vacm, m = 2, 3), and the influence of hydrogen and helium on vacancy evolution in tungsten. We find that hydrogen atoms are extremely difficult to aggregate at interstitial sites to form a stable cluster in tungsten. However, helium atoms are energetically favourable to cluster together in a close-packed arrangement between (1 1 0) planes forming helium monolayer structure, where the helium atoms are not perfectly in one plane. Both hydrogen and helium prefer to aggregate stably in vacancy and small vacancy cluster forming VacmXn (X = H, He). The concentrations of VacmHn (m = 1) clusters relative to temperature are evaluated through the law of mass action. The present calculations also show that the emission of a 〈1 1 1〉 dumbbell self-interstitial atom (SIA) from Hen to form VacHen and from VacHen to form Vac2Hen may take place for n > 5 and n > 9, respectively. According to the present results, we predict that a helium monolayer structure could nucleate for He atom platelet lying on (1 1 0) plane in tungsten, and the helium platelet formation on (1 1 0) plane in molybdenum observed by the experiment may be due to the initial monolayer arrangement of He atoms at interstitial sites. Meanwhile, our results contribute to the understanding for nucleation and the development of the voids and blisters in tungsten that are observed in the experiments.

An energetic and kinetic perspective of the grain-boundary role in healing radiation damage in tungsten

Radiation-induced damage in tungsten (W) and W alloys has been considered as one of the most important issues in fusion research, because radiation-produced defects not only degrade the mechanical property but also change the behaviours of H and He in W significantly, such as the retention of H. Nano-structured W has been developed to reduce accumulation of defects within grains and further mitigate radiation-induced damage. However, the fundamental role of a grain boundary (GB) in healing radiation damage in W is not yet well understood. Using molecular dynamics and statics, we evaluate energetically and kinetically the role of a GB in defect evolution (vacancy and interstitial segregation and their annihilation) near the GB in W, by calculating the vacancy (interstitial) formation energy, segregation energy, diffusion barrier, vacancy?interstitial annihilation barrier near the GB and the corresponding influence range of the GB. We find that, as reported and expected, interstitials are preferentially trapped into GBs over vacancies during irradiation, with vacancies dominant near the GB and interstitials highly localized at the GB. On the one hand, the GB serves as a sink both for vacancies and interstitials near itself by reducing their formation energy and diffusion barrier. The formation energy of the vacancy decreases only by ?0.86?eV, but 7.5?eV is reduced for the formation energy of the interstitial in the GB core, indicating that the sink is unexpectedly stronger for interstitials than vacancies. The average barrier of vacancy diffusion is 0.98?eV much less than 1.8?eV in the bulk; the interstitial migrates into the GB via a barrier-free process. On the other hand, the GB acts as a catalyst for the vacancy?interstitial recombination at the GB by lowering the annihilation barrier. The annihilation with the average barrier as low as 0.31?eV works even at a low temperature of 121?K; besides, the annihilation of a close vacancy?interstitial pair is spontaneous. Meanwhile, the annihilation mechanism near the GB is modified due to the exceptionally large reduction in the interstitial formation energy. The influence range of the GB is small (1?1.5?nm), leading to a small volume fraction of the GB region working as the sink and the catalyst. This indicates that GBs in fine W grains may play a limited role in improving radiation performance.

Mechanical properties and microstructural change of W-Y2O3 alloy under helium irradiation.

A wet-chemical method combined with spark plasma sintering was used to prepare a W–Y2O3 alloy. High-temperature tensile tests and nano-indentation microhardness tests were used to characterize the mechanical properties of the alloy. After He-ion irradiation, fuzz and He bubbles were observed on the irradiated surface. The irradiation embrittlement was reflected by the crack indentations formed during the microhardness tests. A phase transformation from α-W to γ-W was investigated by X-ray diffraction (XRD) and transmission electron microscopy (TEM). Polycrystallization and amorphization were also observed in the irradiation damage layer. The W materials tended to exhibit lattice distortion, amorphization, polycrystallization and phase transformation under He-ion irradiation. The transformation mechanism predicted by the atomic lattice model was consistent with the available experimental observations. These findings clarify the mechanism of the structural transition of W under ion irradiation and provide a clue for identifying materials with greater irradiation resistance.

PSI issues toward steady state plasmas in the HT-7 tokamak

Abstract The main efforts of HT-7 superconducting tokamak are directed to quasi-steady state discharges and relevant physics. New doped graphite with an SiC gradient coating as a limiter material and ferritic steel used to reduce the ripples has been developed. Significant progress in obtaining high-performance discharges under quasi-steady state in HT-7 has been realized. The long pulse discharges with T e ∼1 keV and central density ∼1×10 19 m −3 have been obtained with a duration up to 20 s. Recycling has been studied in LHCD plasmas. The characteristic times of the recycling coefficients seem to be correlated with the physical absorption and adsorption. The uncontrolled density increase is accompanied by the impurity influx originated mainly from the parts of the inner vessel, located far from the plasma edge, which are caused mainly by slow heating by the radiated power and/or fast ion loss.

Effects of alloying and transmutation impurities on stability and mobility of helium in tungsten under a fusion environment

The behaviour of helium in metals is particularly significant in fusion research due to the He-induced degradation of materials. A small amount of impurities introduced either by intentional alloying or by transmutation reactions, will interact with He and lead the microstructure and mechanical properties of materials to change. In this paper, we present the results of first-principles calculations on the interactions of He with impurities and He diffusion around them in tungsten (W), including the interstitials Be, C, N, O, and substitutional solutes Re, Ta, Tc, Nb, V, Os, Ti, Si, Zr, Y and Sc. We find that the trapping radii of interstitial atoms on He are much larger than those of substitutional solutes. The binding energies between the substitutional impurities and He increase linearly with the relative charge densities at the He occupation site, indicating that He atoms easily aggregate at the low charge density site. The sequence of diffusion energy barriers of He around the possible alloying elements is Ti > V > Os > Ta > Re. The present results suggest that Ta might be chosen as a relatively suitable alloying element compared with other possible ones.

Plasma facing components for the Experimental Advanced Superconducting Tokamak and CFETR

Three generations of plasma facing components (PFCs) have been used in Experimental Advanced Superconducting Tokamak (EAST): stainless steel for the initial campaign, fully actively cooled doped graphite PFC and a combination of a graphite divertor with a molybdenum first wall. The doped graphite with a ∼100xa0μm-thick SiC coating has shown good performance for the past 5 years with a heat load removal capacity of 2xa0MWxa0m−2. An ITER-like W mono-block divertor and W flat-type tiles will be used in the second phase of EAST with long pulse operation beginning in 2013. A hot W wall over 350xa0°C will be used in the third phase of EAST operation from 2017 when nearly 40xa0MW of heating and current drive power will be utilized. Combining these experiences of PFCs together with divertor operation optimization, a solid technical basis will be established for the Chinese Fusion Engineering Testing Reactor.

Oxidation behavior of fine-grained SiC–B4C/C composites up to 1400 °C

Abstract Fine-grained B4C–SiC/C composites were fabricated using a ball-milling dispersion process. The oxidation behaviors of both fine-grained B4C–SiC/C composites and coarse-grained B4C–SiC/C composites at temperatures of up to 1400xa0°C were analyzed by the differential thermal analysis technique, and the surface morphology of the composites after isothermal oxidation at 800, 1200 and 1400xa0°C was examined by scanning electron microscopy (SEM). The results indicated that fine-grained B4C–SiC/C composites had excellent oxidation resistance with self-healing properties at 1400xa0°C. A general model and mechanism for self-protection against oxidation of carbon materials were proposed.

Extension of operational limits on EAST

The first plasma has been achieved successfully in the Experimental Advanced Superconducting Tokamak (EAST). Boronization by the glow discharge (GDC) method was studied in experiments. The plasma performance was obviously improved by GDC boronization. Extension of the operational region and improvement in the plasma performance were obtained. Sawtooth discharges were observed by means of soft x-ray signals, electron cyclotron emission signals and line averaged electron density after boronization. Lower qa and more stable operation were also achieved following GDC boronization. The plasma current ramp-up rate was also improved as a result of decreased impurity content and low averaged loop voltage due to boronization. PLEASE NOTE: THERE HAS BEEN A RETRACTION PUBLISHED FOR THIS ARTICLE.

The preparation of fine-grain doped graphite and its properties

Fine-grain doped graphite was prepared by the ball-milling dispersion method for the first time. Such composite has not only high thermal conductivity and excellent bending strength (116 MPa), but also better oxidation resistance at elevated temperature and outgassing properties than those of composite doped with normal size carbides. Correlations between microstructure and properties of such composites are also discussed in detail.

First-principles calculations of transition metal solute interactions with hydrogen in tungsten

We have performed systematic first-principles calculations to predict the interaction between transition metal (TM) solutes and hydrogen in the interstitial site as well as the vacancy in tungsten. We showed that the site preference of the hydrogen atom is significantly influenced by the solute atoms, which can be traced to the charge density perturbation in the vicinity of the solute atom. The solute-H interactions are mostly attractive except for Re, which can be well understood in terms of the competition between the chemical and elastic interactions. The chemical interaction dominates the solute-H interaction for the TM solutes with a large atomic volume and small electronegativity compared to tungsten, while the elastic interaction is primarily responsible for the solute-H interaction for the TM solutes with a small atomic volume and large electronegativity relative to tungsten. The presence of a hydrogen atom near the solute atom has a negative effect on the binding of other hydrogen atoms. The large positive binding energies among the solute, vacancy and hydrogen suggest that they would easily form a defect cluster in tungsten, where the solute-vacancy and vacancy-H interaction contribute greatly while the solute-H interaction contributes a little. Our result provides a sound theoretical explanation for recent experimental phenomena of hydrogen retention in the tungsten alloy and further recommends a suitable W–Re–Ta ternary alloy for possible plasma-facing materials (PFMs) including the consideration of the hydrogen retention.