Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors: Volume 1

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  DegreesC 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
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-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 Spectroscopy.- Part 14. Cables and Concrete Aging and Degradation - Concrete.- Automated Detection of Alkali-silica Reaction in Concrete Using Linear Array Ultrasound Data.- Coupled Physics Simulation of Expansive Reactions in Concrete with the Grizzly Code.- Overview of EPRI Long Term Operations Work on Nuclear Power Plant Concrete Structures.- The Effects of Neutron Irradiation on the Mechanical Properties of Mineral Analogues  of Concrete Aggregates.- Part 15. Accident Tolerant Fuel Cladding.- Accident Tolerant FeCrAl Fuel Cladding: Current Status towards Commercialization.- Interdiffusion Behavior of FeCrAl with U3Si2.- Mechanical Behavior of FeCrAl and Other Alloys Following Exposure to LOCA Conditions Plus Quenching.- Mechanical Behavior and Structure of the Advanced Fe-Cr-Al Alloy Weldments.- Investigating Potential Accident Tolerant Fuel Cladding Materials and Coatings.- Steam Oxidation Behavior of FeCrAl Cladding.- In-situ Proton Irradiation-corrosion Study of ATF Candidate Alloys in Simulated PWR Primary Water.- Hydrothermal Corrosion of SiC Materials for Accident Tolerant Fuel Cladding with and without Mitigation Coatings.- Characterization of the Hydrothermal Corrosion Behavior of Ceramics for Accident Tolerant Fuel Cladding.- Corrosion of Multilayer Ceramic-coated ZIRLO Exposed to High Temperature Water.- Part 16. General SCC and SCC Modeling.- Calibration of the Local IGSCC Engineering Model for Alloy 600.- Prediction of IGSCC as a FEM Post Analysis.- Monte Carlo Simulation Based on SCC Test Results in Hydrogenated Steam Environment for Alloy 600.- Protection of the Steel Used for Dry Cask Storage System from Atmospheric Corrosion by TiO2 Coating.- Predictive Modeling of Baffle-former Bolt Failures in Pressurized Water Reactors.- Technical Basis and SCC Growth Rate Data to Develop SCC Disposition Curve for Alloy 82 in BWR Environments.- Part 17. BWR SCC and Water Chemistry.- SCC and Fracture Toughness of XM-19.- On the Effect of Preoxidation of Nickel Alloy X-750.- Microstructures of Oxide Films Formed in Alloy 182 BWR Core Shroud Support Leg Cracks.- Effect of Chloride Transients on Crack Growth Rates in Low Alloy Steels in BWR Environments.- Electrochemical Behavior of Platinum Treated Type 304 Stainless Steels in Simulated BWR Environments under Startup Conditions.- Investigations of the Dual Benefits of Zinc Injection on 60Co Uptake and Oxide Film Formation under Boiling Water Reactor Conditions.- SCC Mitigation in Boiling Water Reactors: Platinum Deposition and Durability on Structural Materials.- Confirmation of On-line NobleChem (TM) (OLNC) Mitigation Effectiveness in Operating Boiling Water Reactors (BWRs).- E-1: Development of the Fundamental Multiphysics Analysis Model for Crevice Corrosion Using a Finite Element Method.- E-2: In-situ Electrochemical Study on Crevice Environment of Stainless Steel in High Temperature Water.- Part 18. Zirconium and Fuel Cladding.- Corrosion Fatigue Crack Initiation in Zr-2.5Nb.- Cluster Dynamics Model for the Hydride Precipitation Kinetics in Zirconium Cladding.- Modeling of Oxidation Kinetics of Zirconium Alloys in Loss of Coolant Accident (LOCA).- Progressing Zirconium-alloy Corrosion Models Using Synchrotron XANES.- Advanced Characterization of Hydrides in Zirconium Alloys.- Influence of alloying elements and effect of stress on anisotropic hydrogen diffusion in Zr-based alloys predicted by accelerated kinetic Monte Carlo simulations.- Part 19. Stainless Steel Aging and CASS.- Influence of d-Ferrite Content on Thermal Aging Induced Mechanical Property Degradation in Cast Stainless Steels.- Microstructure and Deformation Behavior of Thermally Aged Cast Austenitic Stainless Steels.- Microstructural Evolution of Cast Austenitic Stainless Steels under Accelerated Thermal Aging.- Electrochemical Characteristics of Delta Ferrite in Thermally Aged Austenitic Stainless Steel Weld.- Effect of Long-term Thermal Aging on SCC Initiation Susceptibility in Low Carbon Austenitic Stainless Steels.- Crack Growth Rate and Fracture Toughness of CF3 Cast Stainless Steel at ~3 dpa.- Effects of Thermal Aging and Low Dose Neutron Irradiation on the Ferrite phase in a 308L Weld.- Microstructural Evolution of Welded Stainless Steels on Integrated Effect of Thermal Aging and Low Flux Irradiation.- Part 20. Welds, Weld Metals, and Weld Assessments.- The Use of Tapered Specimens to Evaluate the SCC Initiation Susceptibility in Alloy 182 in BWR and PWR Environments.- Effect of Thermal Aging on Fracture Mechanical Properties and Crack Propagation Behavior of Alloy 52 Narrow-gap Dissimilar Metal Weld.- Distribution and Characteristics of Oxide Films Formed on Stainless Steel Cladding on Low Alloy Steel in PWR Primary Water Environments.- Microstructural Characterization of Alloy 52 Narrow-gap Dissimilar Metal Weld after Aging.- A Statistical Analysis on Modeling Uncertainty through Crack Initiation Tests.- Part 21. Plant Operating Experience.- Laboratory Analysis of a Leaking Letdown Cooler from Oconee Unit 3.- Root Cause Analysis of Cracking in Alloy 182 BWR Core Shroud Support Leg Cracks.- Microbially Induced Corrosion in Fire Fighting Systems - Experience and Remedies.- Managing the Ageing Degradation of Concealed Safety Relevant Cooling Water Piping in European S/KWU LWRs.- F-1: Identification of PWR Stainless Steel Piping Safety Significant Locations Susceptible to Stress Corrosion Cracking.- Part 22. IASCC Testing - Characterization.- On the Use of Density-based Algorithms for the Analysis of Solute Clustering in Atom Probe Tomography Data.- Comparative Study on Short Time Oxidation of Un-irradiated  and Protons Pre-irradiated 316L Stainless Steel in Simulated PWR Water.- Hydrogen Trapping by Irradiation-induced Defects in 316L Stainless Steel.- Grain Boundary Oxidation of Neutron Irradiated Stainless Steels in Simulated PWR Water.- Irradiation Assisted Stress Corrosion Cracking (IASCC) of Nickel-base Alloys in Light Water Reactors Environments Part I: Microstructure Characterization.- Irradiation Assisted Stress Corrosion Cracking (IASCC) of Nickel-base Alloys in Light Water Reactor Environments Part II: Stress Corrosion Cracking Behavior.- O-2: Solute Clustering in As-irradiated and Post-irradiation Annealed 304 Stainless Steel.- Part 23. IASCC Testing - Initiation and Growth.- Irradiation-Assisted Stress Corrosion Cracking Initiation Screening Criteria for Stainless Steels in PWR Systems.- Novel Technique for Quantitative Measurement of Localized Stresses Near Dislocation Channel - Grain Boundary Interaction Sites in Irradiated Stainless Steel.- IASCC Susceptibility of 304L Stainless Steel Irradiated in a BWR and Subjected to Post Irradiation Annealing.- Irradiation Assisted Stress Corrosion Cracking Susceptibility of Alloy X750 Exposed to BWR Environments.- Evaluation of Crack Growth Rates and Microstructures Near the Crack Tip of Neutron-irradiated Austenitic Stainless Steels in Simulated BWR Environment.- Effect of Specimen Size on the Crack Growth Rate Behavior of Irradiated Type 304 Stainless Steel.- Plastic Deformation Processes Accompanying Stress Corrosion Crack Propagation in Irradiated Austenitic Steels.- Part 24. PWR Oxides and Deposits.- Effect of Grain Orientation on Irradiation Assisted Corrosion of 316L Stainless Steel in Simulated PWR Primary Water.- Finite Element Modelling to Investigate the Mechanisms of CRUD Deposition in PWR.- Properties of Oxide Films on Ni-Cr-xFe Alloys in a Simulated PWR Water Environment.- Part 25. PWR Secondary Side.- Effect of Applied Potential and Inhibitors on PbSCC of Alloy 690TT.- Corrosion of SG Tube Alloys in Typical Secondary Side Local Chemistries Derived from Operating Experience.- Investigation on the Effect of Lead (Pb) on the Degradation Behaviour of Passive Films on Alloy 800.- Influence of Alloying on a-a\' Phase Separation in Duplex Stainless Steels.- Stress Corrosion Crack Growth Rate of Alloy 800NG in an Acidic Secondary Side Crevice Environment.- Using Modern Microscopy to \"Fingerprint\" Secondary Side SCC in Ni-Fe Alloys.