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Grain boundaries are important microstructures in metallic materials, and dislocation nucleation on grain boundaries plays a crucial role in material deformation. Studies on P segregation on grain boundaries typically focus on segregation behavior at substitution or interstitial positions. Less attention is given to the interaction of a certain amount of P with dislocation nucleation and grain boundary structural units during deformation. Previous deformation mechanism studies often qualitatively determine mechanisms, struggling to quantitatively explain their roles. This paper offers an atomic mechanism for grain boundary dislocation nucleation, aiding in understanding of the effect of P segregation in BCC Fe on mechanical properties. In this paper, a large-scale atomic/molecular massively parallel simulation (LAMMPS) program was employed. It investigated the effect of P segregation in BCC Fe grain boundary on the tensile deformation behavior. Binary Fe-P embedded atom potential was used for the calculation of stress-strain curves and dislocation activation energy densities. The analysis of P segregation behavior on different grain boundary structures and the effect of temperature was conducted using visual analysis software (OVITO.3.10). To illustrate the elastic-plastic component of the strain during loading, a set of kinematic parameters based on continuum medium mechanics was employed. The atomic deformation tensor of the deformation gradient was calculated using the nearest-neighbor table approximation of the continuum medium field around each atom. Additionally, the mechanisms of dislocation nucleation and twinning were discussed for the symmetrically inclined grain boundaries in BCC Fe. For the (111) grain boundary system, dislocation nucleation extends and dominates the deformation behavior. For the (112) twin grain boundary system, tensile deformation is mainly controlled by twin grain boundary, resulting in BCC-FCC phase transitions. Tensile loads were applied to different grain boundaries at temperatures of 300 K and 600 K. The results indicate that, in the Fe-P system with P concentrations ranging from 0.05at.% to 0.2at.%, compared to enhancing the strength of (111) grain boundary systems, P segregation leads to a decrease in strength for (112) systems, but the effect is minor. The P segregation hinders dislocation nucleation on the (111) and (113) grain boundaries, while it promotes dislocation nucleation at (112) grain boundaries, and P segregation has no significant effect on dislocation nucleation at (114) grain boundaries. The degree of deformation of the E structural unit at the P segregated site on the (111) grain boundary is small, and the dislocations are more inclined to nucleate in the unsubstituted structural unit near the P atom.

Core structural materials such as cladding in advanced nuclear reactors face harsh service environments such as high-temperature, and their comprehensive performance is one of the bottlenecks restricting the development of advanced reactors. Mo and its alloys are key candidate, but pure Mo suffers from a lack of high-temperature strength. Mo-based alloys are superior to pure Mo in terms of mechanical properties. Research found that the addition of a certain amount of Re, W, Ti, Zr, C and other elements will enhance the high-temperature strength of Mo-based alloys. The effects produced by the ingredients are very complex and cannot be found in great detail by testing. Now materials genetic engineering, based on the prediction of data-driven material properties and optimization, shortening the cycle of materials research and development, and thus speed up the process of the entire field. Thus a CNN regression model based on Bayesian optimization was constructed based on the data obtained from literature research and preliminary experiment. The collected Mo-based alloy dataset contained 220 samples and 38 compositions as well as test temperature characteristics of tensile with the target variable being tensile strength. Found that the Re, C, O, N, Ti, Zr, Nb, La, W, Hf, Ta, Si, and Ni have a higher weight on the target value, and C, Ti, and Zr positively affect the tensile strength. Re, O, N, La, W, Hf, Ta, Si, and N have a negative tendency to correlate the characteristics with the tensile strength. Using the R2 value as an evaluation metric, the model is able to achieve 0.832 6. Combined with correlation analysis and modeling prediction analysis, it is concluded that Ti, Zr, Re and La2O3 are beneficial to the enhancement of high-temperature tensile strength of Mo-based alloy. N and Si are not conducive to the enhancement of tensile strength of Mo-based alloy at high-temperature. According to the correlation analysis conclusion, the effect of the coupling of these compositions C-Zr, Nb-Hf-Ta, Ni-Si, N-O and Zr-Ti on the tensile properties of the materials at high-temperatures will be considered subsequently, along with the simultaneous enhancement of toughness. Subsequently, by analyzing the influence of the combination of two elements or three elements on the change of tensile strength, it can get the composition design scheme that the tensile strength of Mo-based alloy reaches a certain range under specific temperature conditions.

Abstract: Reactor pressure vessel (RPV) is an irreplaceable component in pressurized water reactors. It carries internal components, fuel assemblies and primary cooling water. Its integrity plays an important role in the safe operation of the reactor. Microscopic defects caused by neutron irradiation can cause radiation hardening and radiation embrittlement of RPV steel, increasing the risk of brittle fracture of RPV steel. Accurately characterizing the microscopic defects is of great significance for predicting the macroscopic properties of RPV steel. Solute defect clusters are the dominant radiation defects in RPV steel. Because they are rich in solute atoms Cu, Mn, Ni, etc., they can also be called Cu-rich precipitates and Mn-Ni-rich precipitates. At present, there are few studies on the nucleation and growth of precipitates dominated by vacancy clusters. Due to the limitations of microscopic characterization technology, it is difficult to directly observe solute-vacancy clusters in RPV steel. In this work, the solute-vacancy clusters generated in RPV steel after irradiation were focused. Taking the FeCuMnNi alloy system as the research object, the nucleation and growth process of solute-vacancy clusters were explored. The Metropolis Monte Carlo method was used to simulate the stable configurations of the solute-vacancy clusters in different RPV model alloys. The molecular statics method was used to study the interaction mechanism between solute atoms and vacancy clusters. The results show that in FeCuMnNi alloy system, Cu atoms preferentially segregate on the surface of vacancy clusters than Mn and Ni atoms, and the solute-vacancy clusters have a spherical shell structure. Mn can promote the growth of solute-vacancy clusters, while Ni has little effect on the growth of solute-vacancy clusters. Mn and Ni elements have a synergistic effect, which can promote the growth of solute-vacancy clusters. There is a correlation between the structure and energy of solute atoms in solute-vacancy clusters. Vacancy clusters can directly change the structures of solute atoms nearby, causing the inner solute atoms have significantly higher energy than the outer atoms. The growth process of solute-vacancy clusters can change from being dominated by vacancy clusters to outer solute atoms. The results are helpful to understand the formation and evolution behaviors of microscopic defects in RPV steel under neutron irradiation, and promote the prediction of irradiation embrittlement. In the future, to fully reveal the nucleation and growth process of solute-vacancy clusters, we will carry out research from a kinetic perspective using the kinetic Monte Carlo method or the cluster dynamics method.

After neutron irradiation, RAFM steel will produce two types of interstitial dislocation loops (1/2〈111〉 loop and 〈100〉 loop), resulting in radiation hardening and embrittlement after long service. And different types of dislocation loops may lead to different hardening phenomena. However, the formation and transformation processes of different types of dislocation loops are still obscure. Therefore, it is of great significance to clarify the formation and transformation processes of different types of dislocation loops for the evolution of materials’ microstructures and prediction of their properties. In this paper, we presented an improved parallel replica method (PRD), which improved the application of the original PRD method in dislocation loop evolution simulation. Through analyzing the mechanism characteristics of dislocation loop transition and extracting Burgers vector, the improved PRD method accurately identified the state of dislocation loops, and then calculated the simulation of accelerated transition process from 〈100〉 loop to 1/2〈111〉 loop. By combing the dislocation extraction algorithm in OVITO, the synchronous analysis of Burgers vector in the simulation process was realized. During the dislocation loop transforms, its habit plane will change and eventually affect the change of Burgers vector. So Burgers vector was used to redefine the adjudication for transition events in the original PRD method. We adjudicated whether the system undergoes a transition event, based on the existence of 1/2 〈111〉 Burgers vector in the simulated system and whether the proportion of 1/2〈111〉 Burgers vector increases. With such improvements, it would help us avoid lengthy cyclic iterations at some metastable state in our simulation and realize the simulation of accelerated transition phenomenon of 〈100〉 loop to 1/2〈111〉 loop. Finally, numerical experiments were performed to evaluate the effect of our work. The results show that the improved PRD algorithm successfully simulate the changing process of small size 〈100〉 loop to 1/2〈111〉 loop at 1 000 K, and the transition time is faster than the traditional molecular dynamics method. We also observe the coexistence phenomenon of 〈100〉 dislocation segments and 1/2〈111〉 dislocation segments in large size 〈100〉 loop simulation. So far as to the case of 〈100〉 loop with Ni solute elements, it is found that the complete transition to 1/2〈111〉 dislocation loop is happened. In addition, the accuracy of program simulation was verified, and the parallel performance were also analyzed and evaluated in this paper. We achieve about 90% parallel efficiency in the parallel calculating test, which is found much higher than the original PRD method. Moreover, our program has well-fined scalability, so it can be well expanded for the large-scale parallel calculation.

Low activation ferritic martensitic steel is considered as one of the candidate structural materials for the fourth generation reactor and fusion reactor. Low activation ferritic martensitic steel is based on Fe-Cr alloy, in which W is an important solute element. Irradiation embrittlement at low temperature is one of the important problems limiting the service of low-activation ferrite martensitic steels. The precipitated clusters during the service process cause hardening and embrittlement of the material due to the obstruction of dislocation movement by the precipitated clusters. An in-depth understanding of cluster precipitation behavior during service is helpful to understand the irradiation embrittlement of low-activation ferrite martensitic steels at low temperature. In order to simulate the precipitation kinetics of Fe-Cr alloy, by introducing composition-dependent pair potential, some improvements were made on the MIET_AKMC software developed by ourselves. The precipitation kinetics of coherent Cr-rich precipitates/clusters in Fe-Cr(8%, 10%, 16%, 20%)-W(0%, 1%, 2%) alloys during thermal ageing was simulated using the atomic kinetic Monte Carlo (AKMC) method. After thermal ageing, Cr precipitate into clusters to form Cr-rich precipitates/clusters, while W is still in a solid solution state. At the beginning of coarsening process, the radius of clusters, which was about (0.57±0.03) nm, was independent on simulation temperature and the initial Cr content in the box. The precipitation process of Cr cluster can be divided into three stages: nucleation, growth and coarsening. The Cr concentration in box and simulation temperature have an influence on the precipitation rate of Cr clusters. The presence of W can delay the precipitation kinetic process and the effect is most significant when Cr content is about 10%. This is due to that when Cr content is about 10%, the absolute value of the short-range order parameter for Fe-Cr has a maximum, quite independent of the composition and the temperature. The delay effect of W was related with the stronger binding energy between W and vacancy. For the binding energy between Cr and vacancy is 0.06 eV, while the binding energy between W and vacancy is 0.14 eV. From the simulation results, it can be deduced that W is benefit for materials performance, expect for solution strengthening, it can delay the precipitation process of Cr-rich clusters, which is harmful for materials performance.

The elemental segregation behaviors and interactions between point defects and symmetrical tilt grain boundaries (GBs) in TiVTa concentrated solid–solution alloys (CSAs) have been studied through hybrid Monte Carlo/molecular dynamics (MC/MD) simulations. A pure V model, a random TiVTa CSA with randomly distributed elements, and an equilibrated TiVTa CSA with Ti segregation were constructed to investigate the influence of chemical disorder and local elemental segregation on defect–GB interactions. For defect–GB interactions, GBs interact more strongly with interstitials than with vacancies. Compared with the pure V, the vacancy absorption length scale of GBs is greater, whereas the interstitial absorption length scale of GBs is shorter in TiVTa CSAs due to the chemical fluctuation and local lattice distortion. This means a higher recombination efficiency of point defects in TiVTa CSAs. The elemental (Ti) segregation in TiVTa CSAs can further enhance the sink strength of GBs towards interstitials, while simultaneously reducing their sink strength towards vacancies. Consequently, the preference effects of GBs towards interstitials and vacancies are amplified in the equilibrated CSA due to local ordering, thereby reducing efficient defect annihilation around GBs. These results provide fundamental insights into the irradiation defect dynamics of CSAs with body-centered cubic (bcc) structure. [ABSTRACT FROM AUTHOR]

17-4PH martensitic steel is usually used as valve stems in nuclear power plants and it suffers from thermal aging embrittlement due to long-time service in a high-temperature and high-pressure environment. Here, we characterized the evolution of microstructures at the nano-scale in 17-4PH steel by in situ small-angle neutron scattering (SANS) with a thermo-mechanically coupled loading device. The device could set different temperatures and tensile so that an in situ SANS experiment could dynamically characterize the process of nanoscale structural changes. The results showed that with increasing thermal aging time, the ε-Cu phase precipitates and grows as the temperature is 475 °C and 590 °C, and the ε-Cu phase is spherical at 475 °C but became elongated cylinders at 590 °C. Moreover, the loading stress could aid in the growth of the ε-Cu phase at 475 °C. [ABSTRACT FROM AUTHOR]

The problems of underdeveloped pores and high electrochemical impedance of oxygen reduction reaction (ORR) catalysts need to be urgently addressed. Heteroatoms F and N are doped on precursor Fe‐zeolitic imidazolate framework (ZIF)‐8 to increase the activity of Fe‐ZIF‐8‐derived porous carbon materials and a noble metal‐free high‐activity ORR catalyst F–N/FeNC with 3D interconnected porous structure and high conductivity is successfully prepared. Experiments show that the synergistic effect of CF ionic bonds, pyridinic‐N, pyrrolic‐N, graphitic‐N, FeNx sites, and CF semi‐ionic bonds in F–N/FeNC can improve the ORR activity of the F–N/FeNC. The developed F–N/FeNC is advantageous with excellent specific surface area, ultra‐high electrical conductivity, and extremely high ORR mass transfer efficiency. As such, the F–N/FeNC has a high onset potential of 0.95 V (vs RHE) and a half‐wave potential of 0.84 V (vs RHE), implying excellent ORR performance even better than commercially available Pt/C catalyst. In addition, the carefully prepared F–N/FeNC demonstrated higher long‐term durability and excellent resistance to methanol influence compared with commercially available Pt/C catalyst. The developed strategy provides new insights into heteroatom‐doped metal–organic frameworks as a precursor of ORR catalysts. [ABSTRACT FROM AUTHOR]

The reactor is an extreme environment of high temperature, high pressure, and high radiation dose. Damage accumulates in structural materials over service time. It will eventually lead to material failure, such as hardening, embrittlement, and swelling. The radiation damage accumulation is an inherent, complex, and multiscale process, which has been studied extensively with multiscale modeling and simulation methods. The rapid development of high-performance computing makes it possible to accurately operate multiscale simulation for the microstructure evolution of irradiated reactor materials. The European Union, the USA, and China have put great effort into this, and many related works have been carried out. This paper first outlines the basic application of multiscale modeling and simulation technology in understanding the effects of radiation on reactor structural materials. Then, some relevant projects carried out by the USA, the European Union, and China in recent years are summarized. Next, the paper focuses on three widely used simulation techniques at different scales: molecular dynamics, kinetic Monte Carlo, and cluster dynamics. For each method, some key developments in algorithms and computer implementations are reviewed. Finally, the comparison between them is discussed. [ABSTRACT FROM AUTHOR]

Radiation-induced defects are the significant cause of radiation damage in austenitic stainless steels. In this work, the interactions between a screw dislocation and stacking fault tetrahedrons (SFTs) are studied in Fe–10Ni–20Cr (a model alloy of austenitic stainless steel) and Ni, using molecular dynamics simulations. Four interaction processes were primarily found, including (1) partial absorption, (2) sheared, (3) bypassing, and (4) restored through double cross-slip. In Fe–10Ni–20Cr alloys, there is almost no correlation between the critical resolved shear stress (CRSS) value and SFT size, but the intersection position largely impacts the results. And the CRSS value decreases with temperature increasing in Fe–10Ni–20Cr. The occurrence of cross-slip has a significant effect on the interactions, and there exist greater CRSS value in the interactions of cross-slipping. Besides, the occurrence of the cross-slip is linked to the stacking fault energy and the distribution of solute atoms of the alloys. [ABSTRACT FROM AUTHOR]