Neutron activation and radiation damage assessment for W-Ni-Fe tungsten heavy alloys with variable Ni content
Not Available
36 MATERIALS SCIENCE↗
Engineering topics
Kurtz, Richard J.
Publications and source records attributed to Kurtz, Richard J..
Effect of Helium Flux on Near-surface Helium Accumulation in Plasma-Exposed Tungsten
We report results of object kinetic Monte Carlo (OKMC) simulations aimed at understanding the effect of helium flux on the near-surface helium accumulation in plasma-facing tungsten, which is initially defect-free and has a W(100) surface orientation. These OKMC simulations are performed at 933 K for fluxes ranging from 10 22 to 4 × 10 25 He/m 2 s, with 100 eV helium atoms impinging on a W(100) surface up to a maximum fluence of 4×10 19 He/m 2 . In the near-surface region, helium clusters interact elastically with the free surface. The interaction is attractive and results in the drift of mobile helium clusters towards the surface as well as increased trap mutation rates. The associated kinetics and energetics of the above-mentioned processes obtained from molecular dynamics simulations are also considered. The OKMC simulations indicate that as the flux decreases, the retention of implanted helium decreases, and its depth distribution shifts to deeper below the surface in initially pristine tungsten. Furthermore, the fraction of retained helium diffusing into the bulk increases as well, so much so that for 10 22 He/m 2 s, almost all of the retained helium diffused into the bulk with minimal/negligible near-surface helium accumulation. At a given flux, with increasing fluence, the fraction of retained helium initially decreases and then starts to increase after reaching a minimum. The occurrence of the retention minimum shifts to higher fluences as the flux decreases. Although the near-surface helium accumulation spreads deeper into the material with decreasing flux and increasing fluence, the spread appears to saturate at depths between 80 and 100 nm. Finally, we present a detailed analysis of the influence of helium flux on the size and depth distribution of total helium and helium bubbles.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Dose rate effects on damage accumulation and void growth in self-ion irradiated tungsten
Not Available
Jiang, Weilin (ORCID:0000000183028313)↗
Perspectives on multiscale modelling and experiments to accelerate materials development for fusion
Prediction of material performance in fusion reactor environments relies on computational modelling, and will continue to do so until the first generation of fusion power plants come on line and allow long-term behaviour to be observed. In the meantime, the modelling is supported by experiments that attempt to replicate some aspects of the eventual operational conditions. In 2019, a group of leading experts met under the umbrella of the IEA to discuss the current position and ongoing challenges in modelling of fusion materials and how advanced experimental characterisation is aiding model improvement. This review draws from the discussions held during that workshop. Topics covering modelling of irradiation-induced defect production and fundamental properties, gas behaviour, clustering and segregation, defect evolution and interactions are discussed, as well as new and novel multiscale simulation approaches, and the latest efforts to link modelling to experiments through advanced observation and characterisation techniques.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
Effect of Confinement of SIA Cluster Diffusion by Impurities on Radiation Defect Accumulation due to 14MeV neutrons in Tungsten
Impurities acting as traps, reduce the self-interstitial atom (SIA) cluster mobility, thereby increasing recombination and in turn enhancing radiation-resistance of polycrystalline materials. However, at elevated temperatures, depending on SIA-trap (impurity) binding energy, SIA clusters may trap and detrap (detach) numerous times. In tungsten, SIA clusters of all sizes glide one-dimensionally (1D) along their Burgers vector direction. Hence, detrapped SIAs would retrace their path, confining their 1D-glide between traps. However, small SIA clusters can change the direction of their 1D-glide by overcoming a rotation barrier. In addition, impurities or solutes are known to reduce this barrier. A lower rotation barrier can result in detrapped SIA clusters that diffuse in a random 1D direction, effectively leading to 3D (net-3D) diffusion. Here we investigate the effect of two cases of SIA diffusion, namely confined-1D diffusion and net-3D diffusion, on the damage accumulation in polycrystalline tungsten under 14 MeV neutron irradiation using the object kinetic Monte Carlo method. Simulations were performed using cascade debris obtained from molecular dynamics simulations with primary knock-on atom (PKA) energies corresponding to 14 MeV neutrons. In the simulations with net-3D diffusion, SIA clusters of all sizes diffuse in a random 1D direction after the clusters detrap. While for the confined 1D diffusion case, SIA clusters larger than size 5 retain their original direction of 1D-glide. In both types of simulations, trapped vacancies are assumed to be permanently immobilized. Unexpectedly, the damage accumulation was lower with confined 1D diffusion than that with net-3D. We present a systematic comparison of the influence of the two cases of SIA diffusion on the radiation damage accumulation as a function of dose rate, detrapping barrier, and impurity concentration at 1025 K.
Accumulation and recovery of lattice damage, tungs↗
Chapter 3 - Overview of Reactor Systems and Operational Environments for Structural Materials in Fusion Reactors
This chapter presents an overview of the conditions that structural materials being developed for fusion power reactor applications would encounter. We begin by briefly describing the relevant plasma physics conditions that dictate the fusion environment, which determines the loads driving microstructure evolution of fusion structural alloys. The fundamentals of radiation damage in the fusion environment are discussed, but the reader is referred to a much more detailed treatment of this topic given in Chapter 5. Our intent is to highlight the most important degradation mechanisms operating in the fusion environment. We compare and contrast the fusion environment with that found in existing and proposed advanced fission energy systems, highlighting in particular the role that gaseous and solid transmutation products have on property evolution. The majority of the chapter is devoted to a presentation of the principal structural material choices for major subsystems of a fusion power reactor such as the first-wall/blanket, divertor, and vacuum vessel. Our aim is to highlight the essential features of the material choices for each subsystem, and not to delve into the details, since later chapters address each class of structural material in greater depth. The emphasis is on the implication of the environment to the basic physical and mechanical properties requirements of these materials and challenges they face specific to the fusion environment.
Kurtz, Richard J.↗
High-Temperature Creep Properties of NIFS-Heat-2 High-Purity Low-Activation Vanadium Alloy
A low-activation vanadium alloy with the composition of V-4 mass% Cr-4 mass% Ti is an alternative to reduced-activation ferritic/martensitic (RAFM) steels for blanket structural material in fusion reactors. The NIFS-HEAT-2 (NH2) is the world’s highest-purity V-4Cr-4Ti alloy among over-30 kg-scale heats. The processes to purify this heat have successfully enhanced workability, weldability and low-activation characteristics, however degradation of high-temperature strength was a concern. The present study evaluated high-temperature creep properties of NH2, and confirmed no degradation under projected blanket service stresses.
Nagasaka, T↗
ADVANCED-STEM-BASED DEEP LEARNING FOR SEMANTIC SEGMENTATION OF DEFECTS IN STEELS
In this work, we developed a novel deep convolutional neural network (DCNN) model, called DefectNet, for robust and automated semantic segmentation of three crystallographic defects including line dislocations, precipitates and voids commonly observed in structural metals and alloys. In previous work, we established an experimental protocol for a diffraction contrast imaging scanning transmission electron microscopy (DCI STEM) technique and tailored it specifically for imaging defects in popular iron-based structural alloy. Thus, the DefectNet was trained over a small but high-quality DCI STEM defect image sets of a HT-9 martensitic steel before and after neutron irradiation for 111.8 dpa at 412°C. For the defect quantification that typically takes at least half an hour even for an expert, with a good model the machine learning algorithm can produce more reproducible and reliable quantification in a few seconds. This work demonstrates the feasibility of using deep learning algorithms for fast and accurate defect quantitative analysis.
Machine Learning, Radiation effects, Image analysi↗
A multiscale microstructural approach to ductile-phase toughened tungsten for plasma-facing materials
Increasing fracture toughness and decreasing the ductile-brittle transition temperature of a tungsten-alloy has been shown to be feasible by ductile-phase toughening of tungsten. In this method, a ductile phase is included in a brittle tungsten matrix to prevent fracture propagation by crack bridging or crack deflection. This paper models the deformation behavior of ductile phase toughened tungsten materials such as tungsten-copper composites using a multiscale modeling approach that involves a microstructural dual-phase (copper-tungsten) region of interest where the constituent phases are finely discretized and are described by a continuum damage mechanics model. Large deformation, damage and fracture are allowed to occur and are modeled in this region that is connected to the adjacent homogenized elastic regions to form a macroscopic structure such as a test specimen. This paper illustrates such a multiscale modeling approach to analyze un-notched and single-edge notched (SENB) tungsten-copper composite specimens subjected to three-point bending. The predicted load-displacement responses and crack propagation patterns are compared to the corresponding experimental results to validate the model.
Nguyen, Ba Nghiep↗
Atomistic model of xenon gas bubble re-solution rate due to thermal spike in uranium oxide
Atomistic simulations are performed to study the response of Xe gas bubbles in UO 2 to ionizing fission products through the thermal spike approximation. A portion of the total electronic stopping power (S e ) is taken as the thermal spike energy through a ratio variable ζ. The thermal spike energy causes extreme melting within the fission track cylindrical region. Molecular dynamics is employed to quantify the probability of a Xe gas atom to be re-solved (re-dissolved) back into the UO 2 matrix. Subsequently, a re-solution model is developed and parametrized as a function of bubble radius (R), off-centered distance (r), and thermal spike energy (ζS e ). The off-centered distance measures the shift of the thermal spike axis from the bubble center. To evaluate the re-solution model, independent fission product yield of U-235 fission due to thermal neutrons (0.0253 eV), taken from the JEFF-3.3 database, is used. The kinetic energy of the fission products is taken from the EXFOR database. Subsequently, the decay of Se over distance for each fission product is simulated. Finally, the evaluated re-solution rate (re-solution probability per second) is presented as a function of bubble radius for a range of ζ.
74 ATOMIC AND MOLECULAR PHYSICS↗
A tungsten-rhenium interatomic potential for point defect studies
A tungsten-rhenium (W-Re) classical interatomic potential is developed within the embedded atom method (EAM) interaction framework. A force-matching method is employed to fit the potential to ab initio forces, energies, and stresses. Simulated annealing is combined with the conjugate gradient technique to search for an optimum potential from over 1000 initial trial sets. The potential is designed for studying point defects in W-Re systems. It gives good predictions of the formation energies of Re defects in W and the binding energies of W self-interstitial clusters with Re. The potential is further evaluated for describing the formation energy of structures in the σ and χ intermetallic phases. The predicted convex-hulls of formation energy are in excellent agreement with ab initio data. In pure Re, the potential can reproduce the formation energies of vacancy and self-interstitial defects sufficiently accurately, and gives the correct ground state self-interstitial configuration. Furthermore, by including liquid structures in the fit, the potential yields a Re melting temperature (3130 K) that is close to the experimental value (3459 K).
74 ATOMIC AND MOLECULAR PHYSICS↗
Precipitates and voids in cubic silicon carbide implanted with 25Mg+ ions
Not Available
Jiang, Weilin↗
PRECIPITATES AND VOIDS IN CUBIC SILICON CARBIDE IMPLANTED WITH 25Mg+ IONS
This study reports on the formation of 25Mg-bearing precipitates and voids in cubic phase silicon carbide (3C-SiC) single crystals implanted to 9.6×1016 25Mg+/cm2 at 673 K and subsequently annealed at 1073 and 1573 K for 2, 6 and 12 h in an Ar environment. Unidirectionally aligned tetrahedral precipitates are observed around the Mg profile peak, many of which are likely cubic MgC2 and tetragonal Mg2Si particles or their precursor structures. There is also evidence that suggests a possible formation of core (MgC2)-shell (Mg2Si) structure in some of the large precipitates. In addition, small spherical voids are observed near the surface, along with the larger faceted voids in the region of maximum vacancy concentration. 25Mg segregation takes place with a dominant fraction of small atomic clusters. The study provides important results needed to fully assess SiC for potential fusion applications.
Jiang, Weilin↗
A dual-phase microstructural approach to damage and fracture of Ti 3 SiC 2 /SiC joints
We investigate the microcracking mechanisms responsible for Ti 3 SiC 2 /SiC joint damage observed at the macroscopic scale after neutron irradiation experiments in detail. A dual-phase microstructural approach to damage and fracture of Ti 3 SiC 2 /SiC joints is developed that uses a finely discretized two-phase domain based on a digital image of an actual microstructure involving embedded Ti 3 SiC 2 and SiC phases. The behaviors of SiC and Ti 3 SiC 2 in the domain are described by the continuum damage mechanics (CDM) model reported in Nguyen et al., J. Nucl. Mater., 2017, 495:504–515. This CDM model describes microcracking damage in brittle ceramics caused by thermomechanical loading and irradiation-induced swelling. The dual-phase microstructural model is applied to predict the microcracking mechanisms occurring in a typical Ti 3 SiC 2 /SiC joint subjected to heating to 800 °C followed by irradiation-induced swelling at this temperature and cooling to room temperature after the applied swelling has reached the maximum swelling levels observed in the experiments for SiC and Ti 3 SiC 2 . The model predicts minor damage of the joint after heating but significant microcracking in the SiC phase and along the boundaries between SiC and Ti 3 SiC 2 as well as along the bonding joint during irradiation-induced swelling and cooling to room temperature. Our predictions qualitatively agree with the limited experimental observations of joint damage at this irradiation temperature.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
New understanding of nano-scale interstitial dislocation loops in BCC iron
Complex states of nanoscale interstitial dislocation loop can be described by its habit plane and Burgers vector. Using atomistic simulations, we provide direct evidences on the change of the habit plane of a ½<111> loop from {111} to {110} and {211}, in agreement with TEM observations. A new {100} habit plane of this loop is also predicted by simulations. The non-conservation of the Burgers vector is approved theoretically for: (1) dislocation reactions between loops with different Burgers vectors and (2) the transition between <100> loops and ½<111> loops. The rotation from a ½<111> to a <100> loop has also been explored, which occurs at 570 K for time on the order of 10 seconds. The dislocation-precipitate phase duality and change of habit plane are then proposed as new features for nano-scale dislocation loops.
Gao, Ning↗
Atomistic Simulation of Interstitial Dislocation Loop Evolution under Applied Stresses in BCC Iron
Evolution of an interstitial 1/2⟨111⟩ dislocation loop under tensile, shear, and torsion stresses is studied with molecular statics method. Under a tensile stress, the dependence of ultimate tensile strength on size of loop is calculated. The formation of small shear loops around the initial prismatic loop is confirmed as an intermediate state to form the final dislocation network. Under a shear stress, the rotation of a loop is observed not only by a change of the habit plane but also through a transformation between a shear and a prismatic loop. Under torsion, a perfect BCC crystal may undergo a BCC to FCC or BCC to HCP transformation. The present work indicates that a 1/2⟨111⟩ loop can delay these transformations, resulting in the formation of micro-crack on the surface.
dislocation loop↗
Modeling thermal and irradiation-induced swelling effects on the integrity of Ti 3 SiC 2 /SiC joints
Previously, results for CVD-SiC joints created using solid state displacement reactions to form a dual-phase SiC/MAX phase irradiated at 800°C and 5 dpa indicated some extent of cracking in the joint and along the CVD-SiC/joint interface. This paper elucidates the origin of cracking by thermomechanical modeling combined with irradiation-induced swelling effects using a continuum damage approach with support of micromechanical modeling. Three irradiation temperatures (400°C, 500°C and 800°C) are considered assuming experimental irradiation doses in a range leading to saturation swelling in SiC. The analyses indicate that a SiC/MAX joint heated to 400°C fails during irradiation-induced swelling at this temperature while it experiences some damage after being heated to 500°C and irradiated at the same temperature. However, it fails during cooling from 500°C to room temperature. The joint experiences minor damage when heated to and irradiated at 800°C but does not fail after cooling. The prediction agrees with the experimental findings available for this case.
damage modeling↗