Qingyu Wang

Harbin Engineering University · College of Nuclear Science and Technology

Refractory high-entropy alloys (RHEAs) possess remarkable high-temperature mechanical properties and have significant potential for use in nuclear fission reactors. However, the phase stabilities of RHEAs are relatively poor in the intermediate temperature range. The coarsening of the nanoprecipitates weakens the plasticity of RHEAs, which imposes...

Pyrochlore oxides (A2B2O7) are potential nuclear waste substrate materials due to their superior radiation resistance properties. We performed molecular dynamics simulations to study the structural properties and displacement cascades in ytterbium titanate pyrochlore (Yb2Ti2O7) and high-entropy alloys (HEPy), e.g., YbYTiZrO7, YbGdTiZrO7, and Yb0.5Y...

Pyrochlore oxides (A2B2O7) are potential nuclear waste substrate material due to their superior radiation resistance properties. We performed molecular dynamics simulations to study the structural properties and displacement cascades in ytterbium titanate pyrochlore (Yb2Ti2O7). We computed threshold displacement energy ( Ed ) and lattice constant (...

The integrity of structural components is a major concern in reactor safety. Cladding materials can fail when hydrogen uptake is in excess of the terminal solid solubility (TSS). Previous studies on hydrogenation or secondary hydriding (SH) focused mainly on loss of coolant accident (LOCA) and fuel ballooning. Most of these studies are carried out...

Dislocation-oxide interaction in Y2O3 embedded Fe: A molecular dynamics simulation study

Controlled defects population and their movement constitute the foundation for describing micro-structural evolution in any material systems for nuclear applications. Molecular dynamics (MD) simulations were performed to study temperature (1200–1400 K) dependent displacement cascades with a U primary knock on atom (pka) carrying 2 keV kinetic energ...

This study employed classical molecular dynamics (MD) simulations to investigate radiation damage and Cerium (Ce4+) incorporation as surrogate of plutonium (Pu) at Zirconium (Zr4+) site in delta (δ) phase Y4Zr3O12 material. The results indicate that Oxygen (O) disorder is dominant due to its higher number of survived defects with 2 keV, 5 keV and 1...

To optimize fission fuel and protect cladding integrity, this work investigates shadow corrosion in a one-fourth circular electrode geometry. The anodic corrosion of Zircaloy-2 (Zry-2) was investigated in a circular geometry electrode configuration under reactor operating conditions. The impact of gamma and neutron radiations on water conductivity...

Molecular dynamics (MD) simulations were performed to examine the continuous stiffness measurements (CSM) and nanoindentation behavior at 300 K temperature for body-centered cubic molybdenum (Mo). The nanomechanical behavior was examined by using a spherical indenter and the values of elastic modulus and hardness were extracted. In the nanoindentat...

The delta hydrides formation in zirconium cladding under irradiation and thermal aging is studied by the rate theory modelling. The precipitates may lead to cracking of the clad and are investigated both experimentally and theoretically. The rate theory studies published so far are based on the well-established classical nucleation theory, but use...

We performed molecular dynamics simulation on nanoindentation ofγ phase Uranium Molybdenum alloys using spherical indenter. A ternary potential developed for UMoXe was utilized. We calculated the updated values for hardness and reduced elastic modulus at different concentrations of Mo. The whole process of deformation and dislocation analysis was v...

One of the most common threats to the integrity of reactor fuel cladding is the geometric imperfections such as the missing pellet surface (MPS) that produces a remarkable surge in the local fuel-clad gap. The cooling water could occupy this gap leading to secondary hydriding (SH) and hydrogen embrittlement. Most studies on this subject have identi...

The underline challenge of low thermal conductivity which is responsible for high centerline temperature in Uranium-dioxide (UO2) nuclear plant fuel type associated with the current generation of commercial reactors remains a huge concern to the nuclear power industry. Although, researchers in the nuclear industry have proposed uranium mononitride...

One of the most common threats to the integrity of reactor fuel cladding is the geometric imperfections such as the missing pellet surface (MPS) that produces a remarkable surge in the local fuel-clad gap. The cooling water could occupy this gap leading to secondary hydriding (SH) and hydrogen embrittlement. Most studies on this subject have identi...

Phase field modelling technique is critical to contextualizing material microstructures and to represent the composition of microstructural evolution. This work utilizes the periodic boundary condition to numerically solve the Cahn-Hilliard equation. To enhance computation and improve flexibility, Python programming language is introduced to develo...

Yttrium (Y) doped aluminum nitride (AlN) thin films are prepared using reactive magnetron sputtering method in a nitrogen atmosphere at room temperature. Thermal annealing at a temperature of 900 • C is performed after the deposition for the homogeneity and removal of porosity in the films. The prepared samples are irradiated at a proton fluence of...

Thulium (Tm) doped aluminum nitride (AlN) thin films are deposited by radio frequency (RF) magnetron sputtering in a pure nitrogen atmosphere. As-deposited thin films are irradiated at room temperature with 335 keV protons and a fluence of 1×10¹⁴ ions/cm². The effects of irradiation on the structural and optical properties of the deposited thin fil...

The concept of a molten salt reactor began at the Oak Ridge National Laboratory (ORNL) in the United States. Due to its high energy densities, high operating temperatures, and distinguished safety characteristics compared to those of conventional water reactors, molten salt reactors are receiving increasing attention. The dual fluid reactor (DFR),...

The development of next generation nuclear energy systems such as Gen-IV and fusion reactors need materials that resist radiation damage effects. Therefore, advanced nuclear reactors require the development of new materials that will satisfy their design specifications. A considerable amount of research has been performed on some crystalline oxides...

Molecular Dynamics simulation was employed to study precipitate composition dependence on strengthening. Edge dislocation interaction with pure, 80at.%, and 60at.% Cr precipitates of different sizes in a matrix of Fe-15at.%Cr was investigated. The precipitates were found to be relatively hard. This is evident from the absence of shearing mechanism...

The aim of this research is to produce materials that will be useful in complex applications such as nuclear reactor core, fuel rods especially in Gen-IV reactors which operate at higher temperatures than the present reactors, as well as in waste immobilization, high temperature, stress and fatigue applications which are tolerant to a number of rad...

Various researches have indicated the relative advantages of ferritic/martensitic (FM) steels over austenitic and other steels currently in use in light water reactors, and have regarded them as the most promising structural materials in both present and future reactors. Fe-Cr are the model alloys of these steels. Continuous exposure to severe irra...

Neodymium (Nd) and Tungsten (W) doped Aluminum Nitride (AlN) thin films deposited on Si (110) substrates were fabricated by reactive magnetron sputtering technique. The thin films were irradiated at room temperature with protons with a dose of 1 × 1014 ions/cm2 carrying energy of 335 keV. Before and after irradiation, the films were characterized b...

Radiation response of a material is a consequence of defects’ evolution in any radiation damage event. The radiation-induced defects can significantly alter the mechanical properties of a material. Radiation damage initiates from incident neutron by bombardment on solid material causing production and evolution of Frenkel defects. Since voids are f...

Oxide dispersed strengthened (ODS) steel is an important candidate for Gen-IV reactors. Oxide embedded in Fe can help to trap irradiation defects and enhances the strength of steel. It was observed in this study that the size of oxide has a profound impact on the depinning mechanism. For smaller sizes, the oxide acts as a void; thus, letting the di...

In this study, we have investigated the effect of the grain boundary (GB) on the diffusion of a Phosphorus (P) atom in alpha-Fe using molecular dynamics simulations. A Fe-P mixed <110> dumbbell is created in the six symmetric tilt grain boundary (STGB) models. The dumbbells are allowed to migrate at different temperatures from 400 to 1,000 K, with...

Various researches have indicated the relative advantages of ferritic/martensitic (FM) steels over austenitic and other steels currently in use in light water reactors, and have regarded them as the most promising structural materials in both present and future reactors. Fe-Cr are the model alloys of these steels. Continuous exposure to severe irra...

Oxide-dispersed-strengthened (ODS) steel has excellent mechanical, thermodynamic and radiation resistant properties, which makes it an important candidate material for hightemperature reactors applications. Radiation stability of oxide is very important to study at the atomic level. In present case molecular dynamics (MD) simulation is used to stud...

Radiation damage (RD) studies on structural materials provide a way to understand materials' behaviour in a NPP. Oxide Dispersed Strengthened Steel (ODSS) alloys have been considered as one of the promising candidate materials in fast nuclear fission and fusion reactors. There are multiple mechanisms responsible for the microstructural changes inst...

Oxide Dispersed Strengthened (ODS) alloys are considered as a potential candidate for future generation reactors and can even be graded as better radiation resistant materials than the ones already in use. Molecular dynamics (MD) simulation study for defect sink properties of ODS alloys have been investigated at 700K using LAMMPS (Atomic/Molecular...

Radiation Damage (RD) studies on structural materials provide a way to understand a material’s behaviour in a NPP. Oxide Dispersed Strengthened Steel (ODSS) alloys have been considered as one of the promising candidate materials in fast nuclear fission and fusion reactors. There are multiple mechanisms responsible for the microstructural changes in...

Based on the Mott cross sections of relativistic electron collisions with atoms, we calculate displacement creation by electron beams of arbitrary energies (up to 100 MeV) in thin films of arbitrary atomic numbers (up to Z = 90). In a comparison with Mont Carlo full damage cascade simulations, we find that total number of displacements in a film ca...

Understanding defect kinetics in a stress field is important for multiscale modeling of materials degradation of nuclear materials. By means of molecular dynamics and molecular statics simulations, we calculate formation and migration energies of self-interstitial atoms (SIA) and SIA clusters (up to size of 5 interstitials) in alpha Fe and identify...

Continuous silicon carbide (SiC) fiber-reinforced SiC (SiCf/SiC) composites have been considered to be used as structural materials in advanced nuclear reactors for its excellent properties. Their mechanical properties have been greatly improved during the last decade. But the radiation damage at the SiC and pyrolytic carbon interface would degrade...

SiCf/SiC composite materials have been considered as candidate structural materials for several types of advanced nuclear reactors. Both experimental and computer simulations studies have revealed the degradation of thermal conductivity for this material after irradiation. The objective of this study is to investigate the effect of SiC/graphite int...

The objective of this study is to investigate the effect of alloying element indium on the microstructure, mechanical properties, corrosion behavior and in vitro cytotoxicity of Ti-In binary alloys, with the addition of 1, 5, 10 and 15at.% indium. The phase constitution was studied by optical microscopic observation and X-ray diffraction measuremen...

The aim of this study is to investigate the microstructure, martensitic transformation behavior, shape memory effect and superelastic property of Ti49.6Ni45.1Cu5Cr0.3 alloy, with Cu and Cr substituting for Ni. After annealing, the alloy showed single step A–M/M–A transformations within the whole annealing temperature range of 623K to 1273K even in...

The aim of this study was to investigate the electrochemical behavior of Ti50Ni47.2Co2.8 alloy in deaerated artificial saliva solutions with binary NiTi alloy as reference and to characterize the composition and structure of the passive film after polarization tests. The corrosion behavior of NiTiCo alloy was systematically studied by open circuit...

The aim of this study was to investigate the electrochemical behavior of Ti(49.6)Ni(45.1)Cu(5)Cr(0.3) (TiNiCuCr) alloy in artificial saliva solutions with a wide rage of pH values and to characterize the surface passive film after polarization tests. This article represents the ideal, static environment and associated electrochemical response and c...