Irradiation effects in crystalline solid
Thermodynamics and kinetics of point defects in concentrated alloys
Crystalline solids, which most alloys, semiconductors and ceramics are, consist of atoms located on periodic lattices. Point defects such as unoccupied lattice sites, known as vacancies, and atoms not on lattice sites, known as interstitials, have critical importance on many material properties. For instance, they dictate the time scale for a crystal to attain its natural crystal structure and composition, and the behind physics centers on the thermodynamic and kinetic properties of the point defects themselves.
The classic theory for the thermodynamic and kinetic properties of point defects is established primarily using a simplified picture: in which all atoms are of the same chemical type. In conjunction with the rapidly increasing interest in developing concentrated alloys, including the well known high entropy alloys, there is an urgent need for theoretical description of point defects in concentrated alloys. Combining atomistic calculations and statistical mechanics, this project aims to establish an efficient framework for obtaining thermodynamic and kinetic properties of point defects in concentrated alloys. In the figure below, the distribution of vacancy formation energy in U-29Zr (atomic percent) calculated by molecular dynamics simulations using a MEAM interatomic potential is shown. From the distribution, the equilibrium vacancy concentration and the effective vacancy formation energy in the harmonic regime can also be derived.
The research is supported by an INL LDRD and DOE ARPA-E, in collaboration with INL, Auburn, UMich, and UWy.
Defect self-organization under irradiation
Radiation has been widely known to produce damage, i.e., disorder, in materials by creating lattice defects (e.g., vacancies and self-interstitial-atoms, SIA, loops, voids). It also alters kinetics of mass transport by introducing ballistic mixing and radiation enhanced diffusion, and changes materials chemistry due to the creation of lattice defects and impurities and their subsequent reaction and interaction. This leads to the observation of distinct self-organization, including nanoscale compositional patterning in alloys which results from the competing kinetics of mass transport, and the formation of void and gas bubble superlattices. While such self-organized nanostructures may have superior properties, their formation mechanisms are yet to be discerned. Combining rate-theory based instability analysis and mesoscale simulations, two possible anisotropic factors that are responsible for nanoscale void and gas bubble superlattices, namely anisotropic elasticity and anisotropic interstitial diffusion, are investigated. Symmetry breaking processes that lead to the formation of nanoscale concentration waves and the selection of wave vectors are theoretically predicted and observed in atomic kinetic Monte Carlo and phase field simulations.
The research is supported by DOE, Office of Science, Basic Energy Science, in collaboration with INL and BNL.
Radiation induced segregation in stainless steels
Stainless steels are widely used in nuclear energy applications for their superior mechanical properties. They suffer from irradiation assisted stress corrosion cracking (IASCC), which is the accelerated stress corrosion cracking growth rate induced by irradiation. It has been shown that certain minor additives added into 316 stainless steels can effectively reduce radiation induced segregation (RIS), which is regarded as one of the mechanisms responsible for IASCC. However, such mitigating effect can be temporary and vanishes at higher irradiation dose. Without a mechanistic understanding on how additives affect RIS, searching for additives that can mitigate IASCC becomes a tedious task.
Here, a rapid alloy design approach combining integrated computation material engineering (ICME), additive manufacturing, and out-of-pile irradiation test, and advanced characterization are used to search for additives that are beneficial for RIS and thus IASCC. First-principles based density functional theory (DFT) calculations are performed to elucidate how solute atoms interact with point defects. AKMC and phase field models are developed and parameterized by DFT results to study RIS in FeNiCr model alloys. Such a model can be used to rapidly search for additives based on solute-defect interaction. An example phase field simulation of concurrent RIS and grain growth can be found in the below.
The research is supported by an INL LDRD, in collaboration with INL, Auburn, UMich, and UWy.
Multiscale fuel and materials performance modeling
Buffer tearing in TRISO fuel particles
Tristructural isotopic (TRISO) particle fuel is regarded as “the most robust fuel on earth” because they “cannot melt in a reactor and can withstand extreme temperatures that are well beyond the threshold of current nuclear fuels” (DOE_TRISO). They are of interest of may fuel vendors such as X-energy and Kairos Power. Still, the most robust fuel can fail, and one responsible cause is the buffer tearing which can induce radial crack that propagates to outer layers. An open question is: Can we minimize buffer tearing to make TRISO fuel even more robust?
This project aims to establish the correlation between buffer microstructure, more specifically the pore morphology, and fracture, which is driven by irradiation induced dimensional change of the TRISO layers. Combining advanced characterization and measurement and multiscale modeling, this project will investigate several critical factors impacting buffer tearing, including the heterogeneity in buffer microstructure and the stochastic differences among particles, buffer-IPyC bonding strength, swelling of fuel kernel, and irradiation temperature. Multiscale modeling and experiments will be carried out to develop microstructure-property correlations, for the use in finite element modeling of in-pile TRISO fuel particle behavior using the BISON code. Statistical simulations considering the heterogeneity in buffer microstructure and the stochastic differences between particles will be performed to understand and predict the probability of buffer tearing in AGR-1 and GAR-2 particles. The research outcomes are expected to i) provide accurate data on buffer microstructure and mechanical properties, ii) advance the fundamental understanding on microstructure-property correlation of buffer, iii) provide a modeling tool for predicting the in-pile buffer tearing behavior, and iv) assess the stress state and the probability of fracture in IPyC. The simulation results will be validated using the results from TRISO fuel irradiation tests being carried out by the AGR program. Targeting both a fundamental understanding and a predictive modeling tool, the proposed research will provide critical feedbacks for future design of AGR particles with improved performance.
The research is supported by an INL LDRD and DOE ARPA-E, in collaboration with Profs. Thevamaran and Sridharan at UW, INL, ORNL, & TAMU.
Molten salt reactor (MSR) systems have attracted worldwide interests owing to their inherent safety, high efficiency of fuel utilization and low production of nuclear waste. The harsh environment in MSRs that combines high operating temperature, intensive radiation and salt attack postulates a big challenge for structural materials, for which various types of Ni-based alloys have been proposed. One critical issue that limits the performance of these alloys is the loss of Cr due to salt corrosion during operation. Albeit the significant efforts made in the past, a full understanding is yet to be achieved on the effects of salt chemistry, alloy microstructure and composition, and irradiation in corrosion of Ni-based alloys. Here, we propose to develop a phase field model that is capable of describing the dissolution of Cr and the infiltration of vacancies and the subsequent formation of pores, to study the effects of microstructure and irradiation on molten salt corrosion in polycrystalline Ni-based alloys.
The research is supported by the startup funding of Prof. Zhang.
Multiscale modeling of fracture in oxide and metallic fuel
Fracture has been frequently observed in nuclear fuels such as UO2 used in most current light water reactors and UMo used in research reactors. The causes include the thermal gradient in UO2 fuel and the irradiation induced dimensional change in UMo fuel. Fracture of fuels has important consequence on fuel performance as it affects thermal transport, fission gas release and fuel cladding mechanical interaction. The fracture properties and the effects of the underlying microstructure of fuels are required to understand the predict the fracture behavior during fuel operation. Here, a multiscale approach combining molecular dynamics simulations and phase field fracture modeling is used to investigate the fracture behavior of both oxide and metallic fuels, taking the effect of fission gas bubbles into consideration.
The research is supported by the DOE NNSA USHPRR program, in collaboration with INL & NCSU.