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Digital Library/Symposia/E10/Effects of Radiation on Materials: 22nd International Symposium
Sponsored by ASTM committee E10 on Nuclear Technology and Applications
October 16 2004 - October 16 2004
Papers Published from this Symposium
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 360390°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
Use of Broken Charpy V-notch Specimens from a Surveillance Program for Fracture Toughness Determination
Plutonium-238 Alpha-Decay Damage Study of A Glass-Bonded Sodalite Ceramic Waste Form
Radiation-Induced Stress Relaxation of Welded Type 304 Stainless Steel Evaluated by Neutron Diffraction
Flow Localization Processes in Austenitic Alloys
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