SEDL / Past Symposia for Committee E10 / Effects of Radiation on Materials: 22nd International Symposium

Effects of Radiation on Materials: 22nd International Symposium

Sponsored by ASTM committee E10 on Nuclear Technology and Applications

October 16 2004 - October 16 2004

• Microstructural and Mechanical Characterization of Radiation Effects in Model Reactor Pressure Vessel Steels
• Effects of Neutron Irradiation on Precipitation in Reactor Pressure Vessel Steels
• Deformation Mechanism Maps of Unirradiated and Irradiated V-4Cr-4Ti
• Small Angle Neutron Scattering Study of Irradiated Martensitic Steels: Relation Between Microstructural Evolution and Hardening
• Atomic-Scale Simulation of Defect Cluster Formation in High-Energy Displacement Cascades in Zirconium
• Effects of Proton Irradiation on Reactor Pressure Vessel Steel and Its Model Alloys
• Fracture Toughness, Thermo-Electric Power, and Atom Probe Investigations of JRQ Steel in I, IA, IAR, and IARA Conditions
• Post-Irradiation Tensile Behavior and Residual Activity of Several Ferritic/Martensitic and Austenitic Steels Irradiated in Osiris Reactor at 325°C up to 9 dpa
• Mechanical Properties of Cubic Silicon Carbide after Neutron Irradiation at Elevated Temperatures
• Radiation Resistance of Advanced Ferritic-Martensitic Steel HCM12A
• Radiation- and Thermally-Induced Phosphorus Inter-Granular Segregation in Pressure Vessel Steels
• Correlated Formation and Stability of SIA Loops and Stacking Fault Tetrahedra in High Energy Displacement Cascades in Copper
• Neutron Flux Effect on the Irradiation Hardening of Type 304 Stainless Steel
• Development of Fuel Clad Materials for High Burn-up Operation of LWR
• Modeling the Effects of Oversize Solute Additions on Radiation-Induced Segregation in Austenitic Stainless Steels
• Recent Surveillance Data and a Revised Embrittlement Correlation
• Extrapolation of Fracture Toughness Data for HT9 Irradiated at 360–390°C
• Dynamic Finite Element Modeling of Fracture in Charpy V-Notch Specimens of Weld Material 72W
• Microstructure Evolution in ZrC Irradiated with Kr ions
• Behavior of Irradiated Type 316 Stainless Steels under Low-Strain-Rate Tensile Conditions
• The Role of Grain Boundary Engineering on the High Temperature Creep of Ferritic-Martensitic Alloy T91
• Assessment of Neutron Irradiation-Induced Grain Boundary Embrittlement by Phosphorous Segregation in a Reactor Pressure Vessel Steel
• Effects of Neutron Dose, Dose Rate, and Irradiation Temperature on the Irradiation Embrittlement of a Low-Copper Reactor Pressure Vessel Steel
• Creep-Fatigue Behavior in High Strength Copper Alloys
• Tensile, Flexural, and Shear Properties of Neutron Irradiated SiC/SiC Composites with Different Fiber-Matrix Interfaces
• Low Strain-Rate Microstructural Deformation Behavior in 316 Stainless Steel Irradiated in EBR-II
• Mechanical Property Changes in Reactor Vessel Materials Thermally Aged for 209 000 H at 260°C
• Use of Broken Charpy V-notch Specimens from a Surveillance Program for Fracture Toughness Determination
• Radiation-Induced Stress Relaxation of Welded Type 304 Stainless Steel Evaluated by Neutron Diffraction
• Flow Localization Processes in Austenitic Alloys