Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, 2 Teile

This two-volume set represents a collection of papers presented at the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors. The purpose of this conference series is to foster an exchange of ideas about problems and their remedies in water-cooled nuclear power plants of today and the future. Contributions cover problems facing nickel-based alloys, stainless steels, pressure vessel and piping steels, zirconium alloys, and other alloys in water environments of relevance. Components covered include pressure boundary components, reactor vessels and internals, steam generators, fuel cladding, irradiated components, fuel storage containers, and balance of plant components and systems.

Offers solutions to problems in water-cooled nuclear power plants of today and the future Helps in the understanding of materials degradation, which continues to grow in importance for both economic and safety reasons Presents new insights into materials, methods, and techniques from an international,multidiscliplinary community

Auteur
The Minerals, Metals & Materials Society (TMS) is a member-driven international professional society dedicated to fostering the exchange of learning and ideas across the entire range of materials science and engineering, from minerals processing and primary metals production, to basic research and the advanced applications of materials. Included among its nearly 13,000 professional and student members are metallurgical and materials engineers, scientists, researchers, educators, and administrators from more than 70 countries on six continents.

Contenu

Part 1. PWR Nickel SCC SCC.- Scoring Process for Evaluating Laboratory PWSCC Crack Growth Rate Data of Thick-wall Alloy 690 Wrought Material and Alloy 52, 152, and Variant Weld Material.- Applicability of Alloy 690/52/152 Crack Growth Testing Conditions to Plant Components.- SCC of Alloy 152/52 Welds Defects, Repairs and Dilution Zones in PWR Water.- NRC Perspectives on Primary Water Stress Corrosion Cracking of High-chromium, Nickel-based Alloys.- Stress Corrosion Cracking of 52/152 Weldments near Dissimilar Metal Weld Interfaces.- Composite Material Stress Corrosion Crack Arrest Testing in Hydrogen Deaerated Water.- Investigation of Hydrogen Behavior in Relation to the PWSCC Mechanism in Alloy TT690.- Part 2. PWR Nickel SCC Initiation.- Crack Initiation of Alloy 600 in PWR Water.- SCC Initiation Behavior of Alloy 182 in PWR Primary Water.- Multiple Cracks Interactions in Stress Corrosion Cracking: In-situ Observation by Digital Image Correlation and Phase Field Modelling.- Stress Corrosion Cracking Initiation of Alloy 82 in Hydrogenated Steam.- Application of Ultra-high Pressure Cavitation Peening on Reactor Vessel Head Penetration, BMN and Primary Nozzles.- The Effect of Surface Condition on Primary Water Stress Corrosion Cracking Initiation of Alloy 600.- Microstructural Effects on SCC Initiation in Simulated PWR Primary Water for Cold-worked Alloy 600.- Part 3. PWR Nickel SCC - Aging Effects.- A Kinetic Study of Order-disorder Transition in Ni-Cr Based Alloys.- The Role of Stoichiometry on Ordering Phase Transformations in Ni-Cr Alloys for Nuclear Applications.- The Effect of Hardening via Long Range Order on the SCC and LTCP Susceptibility of a Nickel-30Chromium Binary Alloy.- PWSCC Initiation of Alloy 600: Effect of Long-term Thermal Aging and Triaxial Stress.- Stress Corrosion Cracking Behavior of Alloy 718 Subjected to Various Thermal Mechanical Treatments in Primary Water.- Time- and Fluence-to-fracture Studies of Alloy 718 in Reactor.- Developmentof Short-range Order and Intergranular Ccarbide Precipitation in Alloy 690 TT upon Thermal Ageing.- Part. 4. PWR Nickel SCC - Alloy 600 Mechanistic.- Diffusion Processes as a Possible Mechanism for Cr Depletion at SCC Crack Tip.- Role of Grain Boundary Cr 5 B 3 Precipitates on Intergranular Attack in Alloy 600.- Advanced Characterization of Oxidation Processes and Grain Boundary Migration in Ni Alloys Exposed to 480 °C Hydrogenated Steam.- Exploring Nanoscale Precursor Reactions in Alloy 600 in H2/N2-H2O Vapor Using In Situ Analytical Transmission Electron Microscopy.- Electrochemical and Microstructural Characterization of Alloy 600 in Low Pressure H 2 origin: initial; background-clip: initial;">-Steam.- Effect of Dissolved Hydrogen on the Crack Growth Rate and Oxide Film Formation at the Crack Tip of Alloy 600 Exposed to Simulated PWR Primary Water.- A Mechanistic Study of the Effect of Temperature on Crack Propagation in Alloy 600 under PWR Primary Water Conditions.- Part 5. PWR Nickel SCC - Alloy 690 Mechanistic.- Grain Boundary Damage Evolution and SCC Initiation of Cold-worked Alloy 690 in Simulated PWR Primary Water.- Effect of Cold Work and Grain Boundary Carbides on PWSCC Susceptibility of Alloy 690.- Relationship among Dislocation Density, Local Strain Distribution, and PWSCC Susceptibility of Alloy 690.- Morphology Evolution of Grain Boundary Carbides Precipitated near Triple Junctions in Highly Twinned Alloy 690.- A Mechanistic Study on the Stress Corrosion Crack Propagation for Heavily Cold Worked TT Alloy 690 in Simulated PWR Primary Water.- Microstructural Study on the Stress Corrosion Cracking of Alloy 690 in Simulated Pressurized Water Reactor Primary Environment.- Part 6. Effect of Strain Rate and High Temperature Water on Deformation Structure of VVER Neutron Irradiated Core Internals Steel.- Radiation-Induced Precipitates in a Self-Ion Irradiated Cold-Worked 316 Austenitic Stainless Steel Used for PWR Baffle-Bolts.-In Situ and Ex Situ Observations of the Influence of Twin Boundaries on Heavy Ion Irradiation Damage Effects in 316L Austenitic Stainless Steels.- In Situ Microtensile Testing for Ion Beam Irradiated Materials.- Development of High Irradiation Resistance and Corrosion Resistance Oxide Dispersion Strengthed Austenitic Stainless Steels.- Probing Damage Gradients in Ion-irradiated Tungsten Using Spherical Nanoindentation.- Part 7. Irradiation Damage Swelling.- Formation of He Bubbles by Repair-welding in Neutron-irradiated Stainless Steels Containing Surface Cold Worked Layer.- Predictions and Measurements of Helium and Hydrogen in PWR Structural Components Following Neutron Irradiation and Subsequent Charged Particle Bombardment.- Emulating Neutron-induced Void Swelling in Stainless Steels Using Ion Irradiation.- Carbon Contamination, Its Consequences and Its Mitigation in Ion-simulation of Neutron-induced Swelling of Structural Steels.- Void Swelling Screening Criteria for StainlessSteels in PWR Systems.- Theoretical Study of Swelling of Structural Materials in Light Water Reactors at High Fluencies.- Part 8. Irradiation Damage - Nickel Based and Low Alloy.- High Resolution Transmission Electron Microscopy of Irradiation Damage in Inconel X-750.- In-situ SEM Push-to-pull Micro-tensile Testing of in Service Inconel X-750 Annulus Spacers.- Microstructural Characterization of Proton-irradiated 316 Stainless Steels by Transmission Electron Microscopy and Atom Probe Tomography.- Part 9. PWR Stainless Steel SCC and Fatigue SCC.- Microstructural Effects on Stress Corrosion Initiation in Austenitic Stainless Steel in PWR Environments.- Oxidation and SCC Initiation Studies of Type 304L SS in PWR Primary Water.- SCC Initiation in the Machined Austenitic Stainless Steel 316L in Simulated PWR Primary Water.- High-resolution Characterisation of Austenitic Stainless Steel in PWR Environments: Effect of Strain and Surface Finish on Crack Initiation and Propagation.- SCC of Austenitic Stainless Steels under Off-normal Water Chemistry and Surface Conditions Part I: Surface Conditions and Baseline Tests in Nominal PWR Primary Environment.- SCC of Austenitic Stainless Steels under Off-normal Water Chemistry and Surface Conditions Part II: Off Normal Chemistry Long Term Oxygen Conditions and Oxygen Transients.- The Effect of Microchemistry o…