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Meeting 2018 TMS Annual Meeting & Exhibition
Symposium Accident Tolerant Fuels for Light Water Reactor
Presentation Title Modeling Radiation Defect Cluster Accumulation in Neutron Irradiated FeCrAl
Author(s) Dwaipayan Dasgupta, Brian Wirth
On-Site Speaker (Planned) Dwaipayan Dasgupta
Abstract Scope This presentation will focus on modeling neutron induced radiation effects in FeCrAl alloys proposed as accident tolerant fuel clad. Key questions about FeCrAl clad performance relate to the extent of radiation hardening and irradiation creep expected at temperatures around 340C and for neutron dose levels up to about 15 dpa. We present the results of a cluster dynamics reaction-diffusion based rate theory model to predict the size and density evolution of a/2<111> and a<100> prismatic dislocation loops, as well as the fluxes of defects that partition to network dislocations driving irradiation creep. The modeling predictions are compared to recent experimental results, and indicate a good agreement of the a<100> loop size and a/2<111> loop density. Future model extensions to improve the agreement with the experimental observations will be discussed, as well as the anticipated performance of accident tolerant fuel cladding.
Proceedings Inclusion? Planned: Supplemental Proceedings volume

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

AREVA NP’s Evolutionary Solution for Enhanced Accident Tolerant Fuel
Atomic to Mesoscale Research and Development for U3Si2 Accident Tolerant Fuel
Calculating Swelling in U3Si2 Nuclear Fuel Using a Multi-scale Computational Approach
Corrosion Products of FeCrAl Alloys in Simulated LWR Environments during In-situ Proton Corrosion-irradiation Experiment
Crystallographic and Chemical Instabilities of MAX Phases during Proton Irradiation
Density Functional Theory Study of Behavior of Selected Accident Tolerant Nuclear Fuels
Development of Alumina-forming Duplex Stainless Steels as Potential ATF Cladding Materials: Preliminary Assessments of High Temperature Steam Corrosion Behavior and Tensile Property
Development of Cold Spray Coatings for Accident Tolerant Fuel (ATF) Cladding
Effect of Dynamic Strain Aging on Mechanical Properties of Zircaloy-4
Effects of Ce Addition on the Microstructure and Mechanical Properties of Accident-tolerance Fe-Cr-Al Fuel Cladding Materials
Enhanced Accident Tolerant Zirconium-silicide Coated LWR Fuel Cladding
Ex-situ and In-situ Determination of α' Phase Formation/Dissolution in High-Cr Ferritic Alloys Using Small-angle Neutron Scattering
Experimental Characterization of Micro-scale Failure Mechanisms and Governing Properties in SiC/SiC Composites
F-12: High Temperature Oxidation Behavior of Zirconium Silicides and their Coating by Laser Cladding on the Zircaloy-4 Tube
F-13: Optimization of Process Parameters for Thin-wall Tube Fabrication of FeCrAl Alloys
Gaseous Fission Product Swelling Behavior in U3Si2 Fuel
Impact Toughness of Model and Commercial FeCrAl Alloys
Improvements to TRISO Based FCM Fuel Performance Modeling
In Situ Ion Irradiation of Multilayer (TiN, TiAlN) Ceramic Coating for Accident Tolerant Zr-alloy Fuel Claddings
Investigation of Additives, Sol-gel Process Variables, and HIP Parameters on the Density UN Microspheres
Laser Based Characterization of Microstructure and Thermal Properties in Nuclear Fuel Materials
Linking Advanced Multi-scale Modeling with Engineering Scale Fuel Performance Assessments of Accident Tolerant Fuels
Microstructure Characterization of U3Si2 Irradiated by High-energy Ions at LWR Temperatures
Microstructure Studies of Interdiffusion Behavior of U3Si2 and SiC
Mitigation of Oxidation of Zircaloy Cladding in High Temperature Steam via Cr and CrAl Coatings
Modeling Radiation Defect Cluster Accumulation in Neutron Irradiated FeCrAl
Molecular Dynamics Investigation of Interfaces in U3Si2
Multilayer Metal-ceramic Coatings for Accident Tolerant Fuel
New Zr-based MAX Phases as Accident Tolerant Fuel Cladding
ODS FeCrAl Fabrication Methodology for Optimizing Ductility and Sink Strength
Oxidation Behavior of FeCrAl Alloys at T= 300-600C for 100-1000 Hours
PCI Analysis of Coated Zircaloy Cladding under LWR Steady State and Startup Operations
Postirradiation of Accident Tolerant Fuel Concepts: Techniques, Highlights and Future Plans
Quality Optimization of Seamless Thin-wall Tube Production of ATF Wrought FeCrAl Alloys
Quantitative Characterization of Y and Ti Inclusions in a 14Cr-YWTi Nanostructured Ferritic Alloy and their Effect on High Temperature Fracture
Radiation Effects on SiC/SiC Composites for Advanced Accident Tolerant Fuel Cladding Tubes
Rate Theory Simulation of Fission Gas Behavior in U3Si2 under LWR Conditions
Relationship Between Reactive Element Particle Dispersions and Irradiation-induced Defects in Neutron Irradiated Commercial APMT Alloy
Simulation of Iron-chrome-aluminum Alloy Cladding under LOCA Conditions Using the BISON Fuel Performance Code
Simulation of SiC-SiC Composite Micro-pillar Compression as an Investigation of Fiber/Matrix Interface Properties
Spark Plasma Sintering and Microstructural Analysis of Pure and Mo Doped U3Si2 Pellets
Status of Accident Tolerant Fuel Cladding Development for LWRs
Status Update on Westinghouse EnCoreTM ATF
Steam Oxidation and Heavy Ion Irradiation Behaviors of Ti2AlC Ceramics
The Department of Energy Advanced Nuclear Fuels Campaign
The Microstructure and Fission Product Behavior in Irradiated AGR TRISO Fuel Particles
Thermal Aging Embrittlement in a Friction Stir Processed Al-bearing, High-Cr Stainless Steel
Thermal Conductivity of SiC Fiber-reinforced Composites for Accident Tolerant Fuel by the Finite Element Method
Thermal Conductivity of Uranium
Transient Swelling of SiC/SiC Composites and its Implications to Fuels and Core Designs
UB2 as Advanced Nuclear Fuel: Modelling In-reactor Evolution of Thermo-physical and Chemical Properties
Uranium Silicide Behavior in Reactor Relevant Atmospheres
ZrSiO4 as an Efficient Barrier Coating for Nuclear Applications

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