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Tungsten (W) and W-based alloys are potential candidates for next-generation fusion reactors, which would withstand both irradiation damages and heavy heat load. In this work, we employed the molecular dynamic method to simulate the behaviors of different radiation defects under the effect of the temperature gradient field, which is induced by heat load. The rotation of the ⟨ 111 ⟩ dumbbell and habit plane of 1/2 ⟨ 111 ⟩ interstitial loops is analyzed in detail. The results show that these two behaviors are not significantly affected by the temperature gradient. Contrary to the thermal equilibrium state, temperature gradient results in the directional diffusion of ⟨ 111 ⟩ dumbbell and 1/2 ⟨ 111 ⟩ interstitial loops in tungsten from the cold to the hot region. The energy barrier is also reduced in the temperature gradient field, which accelerates the defect diffusion. These results indicate that the accumulation of radiation defects in the high-temperature region is expected in temperature gradient fields, which would lead to more severe radiation damages and degradation of materials. [ABSTRACT FROM AUTHOR]

In this work, we have carried out a series of displacement cascade simulations in tungsten (W), including seven different primary-knock-on atom (PKA) directions, and PKA energy covers 1 to 300 keV at 363 K; 10, 50, 100, and 200 keV cascades are considered at 600 K, and 10 and 50 keV cascades are simulated at 300 K. Then, a systematic database of W cascade simulations has been established, containing more than 2000 cascades. The number of Frenkel pairs, the type of defect, and its spatial distribution were all determined. The effects of PKA direction, PKA energy and temperature have been discussed in terms of defect generation, defect clustering, and dislocation loops. It is found that, the PKA direction may have an effect on the configuration of defects, such as the formation probability of 〈100〉 interstitial loops, C15 clusters and the sub-cascades, and also has an effect on the number of Frankel pairs when PKA energy larger than 50 keV. Furthermore, from 363 K to 600 K, the number of Frankel pairs is reduced by ∼10%, and high temperature can promote the clustering of interstitial. We further found that temperature may affect the distribution of dislocation loops, and the number of 〈100〉 interstitial dislocation loops is more at 363 K than at 600 K. These results can provide a database of key input parameters for the subsequent larger-scale simulations, and give an important reference for the prediction of defect behaviour, mechanical and thermal properties of W-based materials under fusion neutron irradiation. [ABSTRACT FROM AUTHOR]

• The displacement cascade simulations in pure W and W-Re alloy containing GB and/or σ-phase have been systematically investigated. • Interstitials in W-Re alloy are preferentially absorbed by GB region. • Interstitials are preferentially absorbed by σ-phase, and Re atoms tend to aggregate near the σ-phase region. • The absorption of σ phases and the GB region leads to an increase in the aggregation of interstitials near the σ phase/GB region and a reduction of vacancies in the bulk region Transmutation rhenium (Re) produced in tungsten (W) neutron irradiation not only segregates and precipitates at grain boundaries (GBs) but also forms the σ-phase (WRe) in the bulk. In this study, using a molecular dynamics (MD) method, we conducted collision cascade simulations in W and W–Re alloys with 5 or 10 at% Re atoms containing a GB/σ-phase and systematically examined the effect of GB and the evolution of σ-phase. We find the interstitials in W-Re alloy are preferentially absorbed by the GB region such that the number of vacancies in the bulk region is higher in W–Re alloys than in W; the defect distribution is sensitive to the distance of primary knock-on atoms from the GB plane. Furthermore, similar to the tendency in the GB region, interstitials are preferentially absorbed by the σ-phase, and Re atoms tend to aggregate near the σ-phase region. The absorption of σ-phases and the GB region leads to a considerable defect aggregation, which provides suitable conditions for the regrowth of σ-phase structure during long-term evolution. The results provide a reference for understanding the GB structure and precipitations in the displacement cascades of W and W-based alloy materials. [ABSTRACT FROM AUTHOR]

Tungsten (W) and W-based alloys are the leading candidates for plasma-facing materials (PFMs) in future fusion reactors. However, the high energy neutrons generated in fusion reactions not only result in cascade damages but also cause W transmutation. Both the irradiation defects and transmutation products, mainly rhenium (Re), have serious effects on the service behaviors of W PFMs. In this work, we have systematically investigated the interaction between Re and the self-interstitial atoms, self-interstitial clusters and 1/2<111> interstitial dislocation loops in bulk W using molecular dynamics and statics simulations. It is found that there is a strong attractive interaction between an interstitial W atom and a substitutional Re atom, forming a Re–W dumbbell that migrates 3-dimentionally due to the low migration and rotation energies. The small SIA clusters strongly bind with both the substitutional Re atoms and an interstitial Re atom (Re–W mixed dumbbell), thus decreasing the mobility of these clusters. The strong attractive interaction between a Re atom and a 1/2<111> interstitial dislocation loop occurs when the Re atom is located at the core of the loop, and also, their interaction distance along <111> direction is large. The mobility of the 1/2<111> interstitial dislocation loop decreases progressively with increasing Re concentration. • An optimized W–Re empirical potential has been developed. • The interactions between Re and interstitial defects and dislocation loops in W have been systematically investigated. • There is a strong attractive interaction between Re atoms and interstitial-type defects. • The presence of Re atoms reduces the mobility of the self-interstitial clusters and 1/2<111> interstitial dislocation loops. [ABSTRACT FROM AUTHOR]

• New interatomic potentials of Ta-Ta and W-Ta based on F-S formalism have been developed. • The formation energies of point defects and the non-degenerate core structure of a 1/2 〈1 1 1〉 screw dislocation in Ta bulk are well reproduced. • The developed potentials have high accuracy and good transferbility. • These potentials are applicable for describing the evolution of point defects in Ta and W-Ta alloys. Tungsten (W) and W-based alloys are regarded as the most promising candidates for plasma facing materials (PFMs) in future fusion reactors. In this work, new interatomic potentials for Ta element and W-Ta alloy have been developed based on the Finnis-Sinclair formalism, in combination with our previously developed potential for W. The potential parameters for Ta were determined by fitting to a set of experimental and first-principles data, including lattice constant, cohesive energy, elastic constants, point defects formation energies and Rose's equation of state for the bcc lattice. The W-Ta cross parameters were fitted to the first-principles data of the formation energies and binding energies of Ta atom with different point defects in bulk W. The present potentials not only reproduce some important physical properties of various point defects, but also predict the non-degenerate/compact core structure of the 1/2 〈1 1 1〉 screw dislocation in bulk Ta, which is the same as DFT calculations. The developed potentials were expected to be suitable for atomistic simulations of point defects evolution in Ta and W-Ta binary alloys. [ABSTRACT FROM AUTHOR]

Tungsten (W)-based high-entropy alloys (HEAs) have shown promising properties as nuclear fusion materials. Characterizing the primary damage is a critical step in describing and revealing the irradiation-induced damage and radiation resistance mechanism. However, there are a limited number of studies on collision cascades and primary radiation damage in W-based HEAs, owing to the large amount of calculations involved and a lack of appropriate interatomic interaction potentials. In this work, we developed a semi-empirical interatomic potential for W–Ta–Cr–V. By using the developed potential, molecular dynamics simulations of collision cascades in W-based HEAs were performed to assess the primary damage due to irradiation. Based on experimental samples, we reported defect production in W 38 Ta 36 Cr 15 V 11 and compared it to pure W for primary knock-on atom with energies ranging from 1 keV to 100 keV. Our findings showed that the number of FPs at the thermal spike and the number of surviving FPs at the end of the cascade in W 38 Ta 36 Cr 15 V 11 are more than those in pure W, mainly due to the lower threshold displacement energy, melting temperature, and formation energy of point defects. Collision cascades in the W 38 Ta 36 Cr 15 V 11 are less likely to result in the formation of dislocation loops compared to pure W. After collision cascade, in W 38 Ta 36 Cr 15 V 11 , the concentrations of Cr and V atoms in defects is significantly higher than their corresponding concentrations in the system, showing an aggregation tendency. The current collision cascade results provide insights into the primary damage of W-based HEA system under irradiation and should provide reliable guidance for describing the primary damage source terms needed in the kinetic models used to simulate radiation-induced microstructural evolution. [ABSTRACT FROM AUTHOR]