Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors [electronic resource] /
Jackson, John H.editor.
Paraventi, Denise.editor.
Wright, Michael.editor.
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.
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 -- Development of 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 Cr5B3 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 H2origin: 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 Stainless Steels 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 on the Crack Response of Lightly Cold Worked Dual Certified Type 304/304L Stainless Steel after Sensitizing Heat Treatment -- Part 10. PWR Stainless Steel SCC and Fatigue – Fatigue -- The Effect of Load Ratio on the Fatigue Crack Growth Rate of Type 304 Stainless Steels in Air and High Temperature Water at 482\\176F -- Electrical Potential Drop Observations of Fatigue Crack Closure -- The Effect of Environment and Material Chemistry on Single-Effects Creep Testing of Austenitic Stainless Steels -- Corrosion Fatigue Behavior of Austenitic Stainless Steel in Pure D2O Environment -- Mechanistic Understanding of Environmentally Assisted Fatigue Crack Growth of Austenitic Stainless Steels in PWR Environments -- Study on Hold-Time Effects in Environmental Fatigue Lifetime of Low-alloy Steel and Austenitic Stainless Steel in Air and under Simulated PWR Primary Water Conditions -- Part 11. Special Topics I – Materials -- Evaluation of Additively Manufactured Materials for Use as Nuclear Plant Components -- Hot Cell Tensile Testing of Neutron Irradiated Additively Manufactured Type 316L Stainless Steel -- Computational and Experimental Studies on Novel Materials for Fission Gas Capture -- Hydrogen Assisted Cracking Studies of a 12% Chromium Martensitic Stainless Steel – Influence of Hardness, Stress and Environment -- Investigation of Flow Accelerated Corrosion Models to Predict the Corrosion Behavior of Coated Carbon Steels in Secondary Piping Systems -- Effect of High-Temperature Water Environment on the Fracture Behaviour of Low-alloy RPV Steels -- Corrosion Fatigue Testing of Low Alloy Steel in Water Environments with Low Levels of Oxygen and Varied Load Dwell Times -- U-1: Feasibility Study of the Internal Zr/ZrO2 Reference Electrodes in Supercritical Water Environments -- Part 12. 
Special Topics II – Processes -- Investigation Of Pitting Corrosion In Sensitized Modified High-Nitrogen 316LN Steel After Neutron Irradiation -- Quantifying Erosion-corrosion Impacts on Light Water Reactor Piping -- Effect of Molybdate Anion Addition on Repassivation of Corroding Crevice in Austenitic Stainless Steel -- Effect of pH on Hydrogen Pick-up and Corrosion in Zircaloy-4 -- Oxidation Kinetics of Austenitic Stainless Steels as SCWR Fuel Cladding Candidate Materials in Supercritical Water -- A Recent Look at CANDU Feeder Cracking: High Resolution Transmission Electron Microscopy and Electron Energy Loss near Edge Structure (ELNES) -- Part 13. Cables and Concrete Aging and Degradation – Cables -- Simultaneous Thermal and Gamma Radiation Aging of Electrical Cable Polymers -- Principal Component Analysis (PCA) as a Statistical Tool for Identifying Key Indicators of Nuclear Power Plant Cable Insulation Degradation -- How Can Material Characterization Support Cable Aging Management? -- Aqueous Degradation in Harvested Medium Voltage Cables in Nuclear Power Plants -- Frequency Domain Reflectometry Modeling and Measurement for Nondestructive Evaluation of Nuclear Power Plant Cables -- Aging Mechanisms and Nondestructive Aging Indicator of Filled Cross-linked Polyethylene (XLPE) Exposed to Simultaneous Thermal and Gamma Radiation -- Successful Detection of Insulation Degradation in Cables by Frequency Domain Reflectometry -- Capacitive Nondestructive Evaluation of Aged Cross-Linked Polyethylene (XLPE) Cable Insulation Material -- C-2: Tracking of Nuclear Cable Insulation Polymer Structural Changes using the Gel Fraction and Uptake Factor Method -- C-4: Degradation of Silicone Rubber Analyzed by Instrumental Analyses and Dielectric Spectr.
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.
Materials science.
Nuclear energy.
Characterization and Evaluation of Materials.
Nuclear Energy.
Nuclear Energy.