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Meeting 2023 TMS Annual Meeting & Exhibition
Symposium Mechanical Behavior of Nuclear Reactor Materials and Components III
Presentation Title A Mesoscale Model of Creep in Monolithic UMo Fuels
Author(s) Shenyang Hu, Benjamin Beeler
On-Site Speaker (Planned) Shenyang Hu
Abstract Scope Experiments show that creep is one of failure modes in monolithic UMo fuels. Because gas bubbles act as sinks of defects and has different elastic-plastic properties from that of UMo matrix gas bubble structures (volume fraction, spatial and size distributions) not only affect defect concentrations, but also stresses, hence, the creep. In this talk, we will present a mesoscale model of creep by integrating a microstructure-dependent cluster dynamics model, a phase-field model of nonequilibrium gas bubbles, and crystal plasticity theory. The integrated model aims to capture the multi-physics coupling of radiation defect evolution, gas bubble evolution, and diffusion-controlled creep. The effect of radiation conditions, inhomogeneous kinetic properties of defects, and applied stresses on gas bubble swelling and creep rate in polycrystalline UMo and Zr layer structures will be presented. The results are important inputs in the development of creep rate model for predicting the fuel performance.
Proceedings Inclusion? Planned:
Keywords Modeling and Simulation, Nuclear Materials, Mechanical Properties

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

A Mesoscale Model of Creep in Monolithic UMo Fuels
A New Microcrack Healing Mechanism in an Annealed 14YWT Nanostructured Ferritic Alloy
Additively Manufactured Digital Image Correlation for Nuclear Materials
Anisotropic Compressive Strength of Single Crystal Zirconium Pillars and the Effects of Irradiation Hardening and Temperature Through Micro-Pillar Mechanical Testing
Castable Nanostructured Alloy Steels and the Graded Transition to Tungsten for Fusion Reactors
Cladding Coating Integrity Quantified by Ring Pull and Local Strain Analysis
Data-driven Surrogate Constitutive Modeling of Mechanical Creep Behavior under Extreme Conditions
Deformation Characteristics of Additively Manufactured 316L Stainless Steels after Neutron Irradiation
Deformation Mechanisms in Gen-IV Candidate Structural Steels Studied by In-situ Micromechanical Testing Techniques
Deformation Twinning Versus Slip in Ni-based Alloys, Containing Pt2Mo-structured, Ni2Cr-typed Precipitates
Effect of Irradiation on the Tensile Strength of Select Layers and Layer Interfaces of TRISO-coated Nuclear Fuel Particles
Error in RUS Measurements Due to Geometric Uncertainties
Estimating the Strengthening Parameters for Irradiated Alloys using Atomic Scale
Evaluating ATF Cladding Mechanical Behavior
Evaluation of Size Effects in Small Scale Mechanical Testing Combining Multi-length Scale Models and Experiments
Examining Microstructural Effects on Tensile Properties in Irradiated Inconel 718 using Miniaturized Tensile Specimens
Examining Microstructures and Mechanical Properties of Neutron and Ion Irradiated T91, HT9 and 800H Alloys
Fatigue Assessment of Metastable Austenitic AISI 347 Pipe Components for Nuclear Engineering
Fracture Toughness of Highly Irradiated RPV Steels
High Throughput Nanoindentation Creep Testing in Nuclear Reactor Steels
Hydride Reorientation Behavior in ZIRLO Using Ring Compression Tests
Impact of Electrolytic Hydrogen Charging on Fatigue Crack Propagation in Reactor Steels
Impact of Thermal Treatment and Irradiation on Mechanical Behavior of Cold Spray Cr Coatings on Zr-alloy Cladding
Investigating Environmentally-Assisted Cracking in 316 Stainless Steel U-Bend Specimens Exposed to Liquid Sodium
Irradiation and Nanomechanical Performance of Additively Manufactured, In Situ Tempered Grade 91 Steel
Mechanical and Microstructural Characterization of Neutron Irradiated HT-9 Heats at LWR and Fast Reactor Relevant Temperatures
Mechanical Behavior of Bare and Cr Coated Zirconium Claddings During Burst Testing via In-situ Strain Measurements
Mechanical Martensites in Nuclear Steels
Micromechanical Aspects of Deformation and Failure of Advanced Iron-Chromium-Aluminum Alloys
Microstructure-aware Predictions of the Creep Response of Metals Subjected to Nuclear Environments
Microstructure and Mechanical Properties of Neutron Irradiated Tantalum-alloyed Ferritic Martensitic Steels
Modeling Long-term Radiation Effects on the Concrete Biological Shield
Molecular Dynamics Studies of Helium Bubble Effects on Grain Boundary Fracture Vulnerabilities in an Fe70Ni11Cr19-1%H Austenitic Stainless Steel
Multi-scale Modeling of Defect Recombination in Collision Cascade with Molecular Dynamics and Binary Collision Monte Carlo
Musings on Advanced Cladding Qualification
Non-destructive Stress Evaluation in Nuclear Materials by Positron Annihilation Spectroscopy
O-10: Mechanical Testing and Characterization of an Integrated Welding and Thermal Processing Method on Eurofer97
O-11: Migration of Intergranular He Gas Bubbles under a Thermal Gradient in Fe by Phase-field Modeling
O-12: Welding Repair : Behavior Study of the Heat-Affected Zone Regarding the Risk of HAC
O-30: Computer Vision-assisted Oxide Thickness Determination of 304 Stainless Steel in PWR Environments
O-32: Fabrication and Characterization of Oxide Dispersion Strengthened Nickel Alloys for Advanced Molten Salt Reactor Components
O-8: Effects of Helium Implantation on Mechanical Properties Near the Tungsten-carbide Interfaces of Dispersion Strengthened Tungsten Alloy
O-9: Hyper-localized Strengthening of Inconel 617 for Very High Temperature Reactor Applications
ODS Cu Materials for Fusion Application Produced by Mechanical Alloying
Optimizing Nuclear Cladding Mechanical Property Output for Hot-cell Testing
Preliminary Studies on Creep Behavior of Commercial FeCrAl Alloy (APMT)
Probing Neutron Irradiation Simulated Damage with Ion Irradiation and In-situ Mechanical Testing
Robust Constitutive Modeling with Artificial Neural Networks
Simulating Irradiation Induced Creep with Coupled Rate Theory and Plasticity Models
Structure-property Evolution of PM-HIP Fabricated Ni-Alloys 625 and 690 Neutron Irradiated to 1 and 3dpa
The Influence of Nanoindentation Orientation on Deformation Mechanisms in Irradiated Fe – P and Fe – N
The Origin of Superior IASCC Resistance of Additively Manufactured 316L Stainless Steel after Hot Isostatic Pressing in Oxygenated BWR Water
The Role of Stress-State on the Failure Mechanism, Strain to Failure and Fatigue Resistance of Zircaloy-4
Understanding the Mechanisms Involved in Chlorine-Induced Stress Corrosion Cracking of Stainless Steel 304 under a Simulated Marine Environment

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