MODELING AND SIMULATION OF FUEL DISPERSAL DURING THE LOSS-OF-COOLANT ACCIDENT
This document is the compilation of the milestone portion to a larger end of project NEUP report. The executive summary of the modeling portion is provided below: In the event of cladding rupture during a postulated LOCA in a pressurized water reactor, fuel particles, along with fission gases, can be expelled into the reactor core […]
Creep and Creep Fracture Modeling with Surrogate Creep Models and the Extended Finite Element Method
Alloy components in advanced nuclear reactors will be subjected to environmental conditions that could include high temperatures, irradiation, and exposure to corrosive salts. These conditions could lead to the formation of crack-like defects, which could grow over time in a mechanism known as creep crack growth (CCG). Predicting growth rates of these defects is important […]
Surrogate Model Integration with MOOSE XFEM for Creep Crack Growth
Ferritic-martensitic steels are key structural materials for advanced reactors but experience time-dependent deformation and damage under prolonged high temperature and irradiation, leading to creep-driven crack initiation and growth. High-fidelity models—crystal plasticity with irradiation mechanisms, phase-field for microstructural evolution, and continuum-damage viscoplasticity—capture the underlying physics but are too computationally intensive for broad design-space exploration and uncertainty […]
3D Probabilistic Fracture Mechanics / Computational Fluid Dynamics Simulation of a Reactor Pressure Vessel Under Transient Conditions
Reactor pressure vessels (RPVs) are safety-critical light-water-reactor components that, under irradiation, experience long-term material degradation in the form of embrittlement. This can increase their susceptibility to fracture under thermal-shock conditions, which could occur during off-normal transients such as loss-of-coolant accidents (LOCAs). During a LOCA, the most severe conditions for the RPV occur when emergency core […]
Assessment of Irradiation Embrittlement Effect on Fatigue Life of a Pressurized Light Water Reactor Pressure Vessel Using the Fracture Toughness Master Curve Approach
The reactor pressure vessel (RPV) is a critical structural component in pressurized-water reactors, and it is designed to withstand extreme conditions, such as high pressures, elevated temperatures, and prolonged radiation exposure. Ensuring RPV integrity is essential for the safe and reliable long-term operation of nuclear power plants, especially as aging mechanisms such as fatigue and […]
Initial Fracture Propagation Modeling of Graphite Components with Grizzly
Graphite has historically been extensively used in power reactor cores and will be used in multiple types of advanced reactors currently under development. These graphite structural components can experience significant stresses due to nonuniform volumetric strains induced by irradiation and thermal expansion, which can lead to fracture. Robust tools for predicting fracture initiation and propagation […]
Coupled alkali-silica-reaction/seismic simulation of nuclear containment vessels
Alkali-silica reaction (ASR) is a degradation mechanism that affects a variety of reinforced concrete (RC) structures and has been observed in multiple operating nuclear plants. To ensure safe long-term operation of nuclear reactors, it is essential to quantify the adverse impacts of ASR on the performance of critical nuclear power plant RC structures such as […]
Bayesian Calibration of Nuclear Graphite Property Models Accounting for Model Inadequacy and Impacts on Component Performance
Nuclear-grade structural graphite is extensively utilized in the core designs of various advanced nuclear reactors. In the reactor environment, graphite is subjected to prolonged exposure to extreme conditions, including high temperatures, radiation, and potentially molten salt and oxygen. Such exposure can induce several degradation mechanisms in graphite, including nonuniform volumetric strains caused by irradiation and […]
Development of Graphite Thermal and Mechanical Modeling Capabilities in Grizzly
Nuclear-grade graphites are used extensively in the core designs of multiple types of advanced nuclear reactors. In the reactor environment, graphite is exposed over long durations to extreme conditions, including high temperatures, radiation and potentially molten salt and oxygen. Exposure to these conditions can cause several degradation mechanisms in graphite, including nonuniform volumetric strains induced […]
Predicting thermophysical properties of molten salts in the MgCl2-NaCl-KCl-LiCl system with a shell-model potential
Ternary eutectic salts composed of MgCl2, NaCl, and KCl, referred to as MNK salts, have recently emerged as promising materials for high-temperature heat transfer fluids. In this study, we have performed classical molecular dynamics (MD) simulations to predict the densities, specific heat capacities, viscosities, and ionic self-diffusivities for MNK salts over a wide temperature range. […]