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.