ProgramMaster Logo
Conference Tools for Materials Science & Technology 2019
Register as a New User
Submit An Abstract
Propose A Symposium
Presenter/Author Tools
Organizer/Editor Tools
About this Abstract
Meeting Materials Science & Technology 2019
Symposium Materials for Nuclear Applications
Presentation Title Aging Behavior and Microstructure Evolution in Ni-Cr-Mo-W (Haynes 244) Alloy After Surface Treatment by Laser Shock Peening (LSP)
Author(s) Jie Song, Anurag Sharma, Seetha Mannava, Vijay Vasudevan
On-Site Speaker (Planned) Jie Song
Abstract Scope The Nickel base superalloy Haynes-244 (Ni-22.5Mo-6W-8Cr alloy) is used in high temperature environments because of its low thermal expansion coefficient with excellent mechanical properties and resistance to creep deformation and corrosion at elevated temperatures. The strengthening of this alloy is achieved by forming coherent intragranular Ni2(Cr,Mo,W) precipitates and intergranular μ phase during aging. The effects of LSP and post-aging on near-surface microstructure and properties were studied. Instead of Ni2(Cr,Mo,W), μ phase particles form evenly at the surface in γ matrix which leads to higher mechanical properties. The sub-surface has much denser intragranular Ni2(Cr,Mo,W) compared to traditional aging and regions far away from the surface show the traditional microstructure. The gradient microstructure after aging has been found to be mainly attributed to the LSP process, which can introduce higher plastic strain at the surface and promote μ phase formation. The effects of such gradient precipitate microstructures on properties will be discussed.


A Metallic Multilayer Composite for use in Fluoride Molten Salt Reactors
Aging Behavior and Microstructure Evolution in Ni-Cr-Mo-W (Haynes 244) Alloy After Surface Treatment by Laser Shock Peening (LSP)
Anisotropic Thermal Transport in Uranium Dioxide Induced by Dislocation
Anisotropy in Thermal Creep and Creep Life Prediction of Zr-2.5%Nb Pressure Tube Alloy
Characterization of the Impact of Fission Product Inclusion on Phase Development in U3Si2 Fuel
Combined Use of In-situ and Ex-situ TEM to Characterize Irradiation Induced Dislocation Loops in F/M Steels for Nuclear Applications
Comparison of In-situ Micro- and Ex-situ Meso-scale Tensile Testing for the Evaluation of Mechanical Properties of Stainless Steels
Compatibility of U3Si2 fuel with FeCrAl and SiC/SiC Based Cladding
Computational Studies of Environmental Degradation of Silicon Carbide
Design of Alloy Chemistry to Mitigate Fuel-Cladding Chemical Interactions in Uranium-based Metallic Fuels
Development and Performance of High Temperature Irradiation Resistant Thermocouples
Effect of Ultrasonic Nanocrystalline Surface Modification (UNSM) on the Oxidation Behavior of Alloy 800HT in a Supercritical Carbon Dioxide (SCO2) Environment
Effects of Ti and Al Additions on Irradiation Behavior of FeMnNiCr Based High-Entropy Alloys
Enhancing the Properties of a Cast FeNiMnCr10 Co-free High-entropy Alloy Through Hot Rolling
Evolution of Microstructure, Deformation Mechanisms, and Internal Damage During Creep-Fatigue Testing of Alloy 709 (Fe-20Cr-25Ni)
Exploring TRISO Layer Properties and Performance for Multiple Reactor Concepts
High-Entropy Carbide Ceramics for Extreme Environments
In-situ Characterization of Zirconium Alloy Degradation to Support Nuclear Sensing Applications
In-situ Ion Irradiation Study of Silicon Carbide-Carbon Coated Nanostructured Ferritic Alloy
Materials for Capture of Uranium for Nuclear Fuel from Fertilizer
Multi-dimensional, In-situ Mechanical Testing to Evaluate Damage and Fracture of Chromium-Coated Zirconium-based Fuel Claddings
P3-62: Characterization and Oxidation of Graphite and Silicon Carbide in TRISO Nuclear Fuel
P3-63: Corrosion Behavior of Metallic Alloys in a Molten Chloride Eutectic Salt for Nuclear Applications
P3-64: Corrosion Behavior of Nanostructured Stainless Steels and High Entropy Alloys
P3-65: Effects of Deposition Conditions on the Production of ZrO2 Coatings Produced by PE-CVD as Environmental Barrier Coatings for the Molten Salt Reactor
P3-67: Investigation of Recrystallization and Grain Growth in Uranium 10 wt% Molybdenum
P3-68: Oxidation of TRISO particles and Matrix Graphite in Mixed Gas Atmospheres
Predicting Concrete’s Response to Irradiation
PVD Coating of Surrogate Fuels for Deep Space Nuclear Thermal Propulsion
Rapid Multiscale Simulation of Cladding Performance: Application to HT-9
Spectral Thermal Conductivity Predictions in UO2 with Xe Inclusions
Steam Oxidation and Microstructural Characterization of U3Si2 alloyed with Al, Cr, Nb, Y, and Zr
Temperature Impacts on Damage Response in Mixed Carbides
The Effects of Ultrasonic Nanocrystal Surface Modification at Room Temperature and Elevated Temperatures on Residual Stress, Microstructure and Mechanical Properties of Nuclear Alloys IN600 and IN690.
Thermophysical Properties of Binary Cl & F Compositions for Next Generation Molten Salt Reactors
Transmutation-induced Precipitation in Tungsten Irradiated with a Mixed Energy Neutron Spectrum
Unveiling SiC/SiC CMC Cladding Failure Mechanisms and Hermetic Performance with In-situ 3D-Digital Image Correlation
Uranium Nitride and High Temperature Irradiation Resistant Thermocouples towards Accident Tolerant Nuclear Fuel
Using ACRT with MVB Furnace to Achieve Low-cost CZT

Questions about ProgramMaster? Contact