SEDL / Topics / Nuclear Science and Technology / Recent 25 JAI
Corrosion of M5 in PWRs: Quantification of Li, B, H and Nb in the Oxide Layers Formed Under Different Conditions
Understanding Crack Formation at the Metal/Oxide Interface During Corrosion of Zircaloy-4 Using a Simple Mechanical Model
Studies Regarding Corrosion Mechanisms in Zirconium Alloys
Microstructural Studies of Heat Treated Zr-2.5Nb Alloy for Pressure Tube Applications
Detailed Analysis of the Microstructure of the Metal/Oxide Interface Region in Zircaloy-2 after Autoclave Corrosion Testing
Ultra Low Tin Quaternary Alloys PWR Performance—Impact of Tin Content on Corrosion Resistance, Irradiation Growth, and Mechanical Properties
Hydrogen Absorption Mechanism of Zirconium Alloys Based on Characterization of Oxide Layer
Neutron Radiography: A Powerful Tool for Fast, Quantitative and Non-Destructive Determination of Hydrogen Concentration and Distribution in Zirconium Alloys
Experimental Investigation of Irradiation Creep and Growth of Recrystallized Zircaloy-4 Guide Tubes Pre-Irradiated in PWR
ZIRLO
®
Irradiation Creep Stress Dependence in Compression and Tension
In Situ Scanning Electron Microscope Observation and Finite Element Method Analysis of Delayed Hydride Cracking Propagation in Zircaloy-2 Fuel Cladding Tubes
Improved Zr-2.5Nb Pressure Tubes for Reduced Diametral Strain in Advanced CANDU Reactors
Polycrystalline Modeling of the Effect of Texture and Dislocation Microstructure on Anisotropic Thermal Creep of Pressurized Zr-2.5Nb Tubes
Characterization of Oxygen Distribution in LOCA Situations
Damage Build-Up in Zirconium Alloys During Mechanical Processing: Application to Cold Pilgering
Effects of Secondary Phase Particle Dissolution on the In-Reactor Performance of BWR Cladding
Study on the Role of Second Phase Particles in Hydrogen Uptake Behavior of Zirconium Alloys
In Situ Studies of Variant Selection During the
α
-
β
-
α
Phase Transformation in Zr-2.5Nb
Hydride Platelet Reorientation in Zircaloy Studied with Synchrotron Radiation Diffraction
Multiscale Analysis of Viscoplastic Behavior of Recrystallized Zircaloy-4 at
400
°
C
Statistical Analysis of Hydride Reorientation Properties in Irradiated Zircaloy-2
Study of the Initial Stage and Anisotropic Growth of Oxide Layers Formed on Zircaloy-4
Hydrogen Solubility and Microstructural Changes in Zircaloy-4 Due to Neutron Irradiation
Radiation Damage of E635 Alloy Under High Dose Irradiation in the VVER-1000 and BOR-60 Reactors
Shadow Corrosion-Induced Bow of Zircaloy-2 Channels
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