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In the fusion environment, a complex stress field is generated in materials, which affects the evolution of radiation defects. In this study, the behaviors of radiation-induced defects under the effect of stress gradient field in tungsten are carefully simulated at the atomic scale with the molecular dynamics (MD) method. It was found that the stress gradient field affects the migration properties of interstitial defects, resulting in the energy barriers changing with the stress and stress gradient. In the axial stress gradient field, the movement of the 1/2 <111> interstitial dislocation loop is significantly accelerated, and it tends to move toward the region where the stress is concentrated. Within the time scale of the classical MD simulation, the stress gradient has little effect on the migration of vacancies. These results suggested that the stress gradient would cause interstitial defects to accumulate to the region where the stress is concentrated, thereby significantly changing the properties of the tungsten 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]

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]

l Ta does not significantly influence the primary damage state, in terms of number of surviving defects or clustered fraction. l Ta reduces the mobility of the interstitial clusters and 1/2<111> interstitial loops. l The presence of Ta has a stabilizing effect on the 〈100〉 loops at high-temperature. Tungsten (W) and W-based alloys are competitive candidates for plasma-facing materials in future fusion reactors. W-tantalum (Ta) alloys get the most attention amongst these materials. The present study uses molecular dynamics simulations to study the displacement cascades of pure W and W-Ta alloy systems with recoil energy up to 100 keV at 300 K and the further evolution of cascade defects at 1,000 K. The effects of Ta concentration on the generation and evolution of the defects, produced with primary knock-on atom energies, are quantitatively analysed. This study's results show that the presence of randomly distributed Ta atoms does not significantly affect the average number of surviving Frenkel pairs or the fractions of clustered vacancies and interstitials. In addition, there is no completed vacancy dislocation loop in all systems after cascading. The produced dislocation loops are mainly 1/2 〈111〉 loops companied with a small amount of 〈100〉 loops and mixed loops. However, the existence of Ta atoms slows down the motion of defects and defect mobility decreases as the Ta concentration increases in bulk W. This will affect the size and density of defects. In addition, the presence of Ta also inhibits the transition of the 〈100〉 dislocation loops to 1/2 〈111〉 dislocation loops under the high-temperature condition. This will affect the types of dislocation loops and the material performance. [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]