Precipitation Behavior of Irradiated Reduced-Activation Ferritic Steels
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概要
- 論文の詳細を見る
- Atomic Energy Society of Japanの論文
- 1996-09-25
著者
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Kimura Akihiko
Institute of Advanced Energy, Kyoto University
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Kimura A
Univ. Tokyo Tokyo
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KAYANO Hideo
The Oarai Branch, Institute for Materials Research, Tohoku University
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SHIBAYAMA Tamaki
The Oarai Branch, Institute for Materials Research, Tohoku University
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Kayano Hideo
The Oarai Branch Institute For Materials Reseach Tohoku University
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Kimura Akihiko
Institute Of Advanced Energy Kyoto University
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Shibayama T
The Oarai Branch Institute For Materials Research Tohoku University
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Shibayama Tamaki
The Oarai Branch Institute For Materials Research Tohoku University
関連論文
- Characterization of Mechanically Alloyed Powders for High-Cr Oxide Dispersion Strengthened Ferritic Steel
- High Burnup Fuel Cladding Materials R&D for Advanced Nuclear Systems : Nano-sized oxide dispersion strengthening steels
- Effects of aluminum on the corrosion behavior of 16% Cr ODS ferritic steels in a nitric acid solution
- Effects of Aluminum on the Corrosion Behavior of 16%Cr ODS Ferritic Steels in a Nitric Acid Solution
- Effects of Zr Addition on the Microstructure of 14%Cr4%Al ODS Ferritic Steels
- Development of Controlled Temperature-Cycle Irradiation Technique in JMTR
- In Reactor Measurement of Electrical Conductivity of Single Crystal α-Alumina
- Vickers Microhardness Testing with Miniaturized Disk Specimens
- Development of Molybdenum Alloy with High Toughness at Low Temperatures
- New Techniques to Apply Optical Fiber Image Guide to Nuclear Facilities
- Radiation Distribution Sensor with Optical Fibers for High Radiation Fields
- Correction Techniques of Radiation Induced Errors for Raman Distributed Temperature Sensor and Experiment at the Experimental Fast Reactor : JOYO
- Effect of Specimen Geometry on Charpy Impact Test Results for Ferritic Steel Irradiated in JMTR
- Studies on hydrogen absorption-desorption proterties of U-Th-Zr alloys for developing new reactor fuel materials
- New Ternary Hydride Formation in U-Ti-H System
- Grain Boundary Phosphorus Segregation in Thermally Aged Low Alloy Steels
- Precipitation Behavior of Irradiated Reduced-Activation Ferritic Steels
- Characterization of Mechanically Alloyed Powders for High-Cr Oxide Dispersion Strengthened Ferritic Steel
- Mechanical Behavior of Oxide Dispersion Strengthened Steels Irradiated in JOYO
- Application of Small Punch Test to Evaluate Sigma-Phase Embrittlement of Pressure Vessel Cladding Material
- Effects of Aluminum on the Corrosion Behavior of 16% Cr ODS Ferritic Steels in a Nitric Acid Solution
- Development of rig for fundamental study of radiation effects on fusion materials in JMTR
- Study of Radiation Induced Electrical Degradation of Alumina in a Dynamic Pumping Condition in a Fission Reactor
- Irradiation Behavior of developed radiaiton resistance optical-fibers and observed optical radiation from their SiO_2 cores under reactor irradiation
- Development of Irradiation Techniques for Material Study in JMTR
- Low Activation V-Ti-Cr-Si Type Alloys for Fusion Applications
- Microstructure, Mechanical Properties and Fracture Behavior of α Particle Irradiated Type 316 Stainless Steel
- Factors Controlling Irradiation Hardening of Iron-Copper Model Alloy
- Analysis of the Neutron Spectrum by a Simple Method
- Effects of Neutron Irradiation on Mechanical Properties of Molybdenum
- Derivation of Energy Generated by Nuclear Fission-Fusion Reaction
- Embrittlement in Neutron Irradiated Niobium
- Current Status of Reduced-Activation Ferritic/Martensitic Steels R&D for Fusion Energy
- Microstructures of High-Purity Ferritic Steels after Helium Implantation
- Work Hardening, Sensitization, and Potential Effects on the Susceptibility to Crack Initiation of 316L Stainless Steel in BWR Environment
- Under Glorious Tradition of the Institute for Materials Research
- Material Development for Nuclear Fusion and Energy Development Using Actinoids
- Work Hardening, Sensitization, and Potential Effects on the Susceptibility to Crack Initiation of 316L Stainless Steel in BWR Environment
- Evaluation of impact properties of weld joint of reactor pressure vessel steels with the use of miniaturized specimens
- PREFACE
- Effects of Nickel Addition on Microstructural Evolution and Mechanical Properties of Reduced Activation Martensitic Steels Irradiated in the ATR-A1
- Grain Boundary Phosphorus Segregation in Thermally Aged Low Alloy Steels
- Structure-Activity Relationship of the Warburgia Sesquiterpene Dialdehydes
- Effects of temperature, strain rate, and specimen orientation on localized plastic deformation of irradiated zircaloy-2.
- Effect of neutron irradiation on fracture behavior of zirconium.
- Relation between microstructure and recovery behavior in fast-neutron irradiated titanium.
- Effect of neutron irradiation on mechanical properties of niobium and Nb-Y alloy.
- Mechanical properties of neutron-irradiated niobium.
- Effect of neutron irradiation on deformation behavior of zirconium.
- Neutron irradiation effects on thermally activated process of tensile properties of titanium.
- Irradiation Effect on Tensile Property of F82H IEA and Its Joint in TITAN Project