Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors
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...
| Συγγραφή απο Οργανισμό/Αρχή: | |
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| Άλλοι συγγραφείς: | , , |
| Μορφή: | Ηλεκτρονική πηγή Ηλ. βιβλίο |
| Γλώσσα: | English |
| Έκδοση: |
Cham :
Springer International Publishing : Imprint: Springer,
2019.
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| Έκδοση: | 1st ed. 2019. |
| Σειρά: | The Minerals, Metals & Materials Series,
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| Θέματα: | |
| Διαθέσιμο Online: | Full Text via HEAL-Link |
Πίνακας περιεχομένων:
- 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.
