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

Author: John H Jackson; Denise Paraventi; Michael Wright
Publisher: Cham, Switzerland : Springer, [2018].
Series: Minerals, metals & materials series.
Edition/Format:   eBook : Document : Conference publication : EnglishView all editions and formats
Summary:
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,  Read more...
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Genre/Form: Electronic books
Conference papers and proceedings
Congresses
Material Type: Conference publication, Document, Internet resource
Document Type: Internet Resource, Computer File
All Authors / Contributors: John H Jackson; Denise Paraventi; Michael Wright
ISBN: 9783319672441 3319672444
OCLC Number: 1005921894
Notes: International conference proceedings.
Includes indexes.
Description: 1 online resource.
Contents: 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.
Series Title: Minerals, metals & materials series.
Responsibility: John H Jackson, Denise Paraventi, Michael Wright, editors.

Abstract:

Components covered include pressure boundary components, reactor vessels and internals, steam generators, fuel cladding, irradiated components, fuel storage containers, and balance of plant  Read more...

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    schema:description "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."@en ;
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    rdfs:comment "eBook available for UOIT via SpringerLink. Click link to access" ;
    .

<http://worldcat.org/isbn/9783319672441>
    a schema:ProductModel ;
    schema:isbn "3319672444" ;
    schema:isbn "9783319672441" ;
    .

<http://worldcat.org/issn/2367-1181> # The minerals, metals & materials series,
    a bgn:PublicationSeries ;
    schema:hasPart <http://www.worldcat.org/oclc/1005921894> ; # Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems -- Water Reactors. Volume 1
    schema:issn "2367-1181" ;
    schema:name "The minerals, metals & materials series," ;
    .

<http://www.worldcat.org/title/-/oclc/1005921894>
    a genont:InformationResource, genont:ContentTypeGenericResource ;
    schema:about <http://www.worldcat.org/oclc/1005921894> ; # Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems -- Water Reactors. Volume 1
    schema:dateModified "2018-04-07" ;
    void:inDataset <http://purl.oclc.org/dataset/WorldCat> ;
    .

<https://library.smu.ca/login?url=http://link.springer.com/10.1007/978-3-319-67244-1>
    rdfs:comment "Volume 1" ;
    rdfs:comment "Access restricted: SMU users only" ;
    .

<https://library.smu.ca/login?url=http://link.springer.com/10.1007/978-3-319-68454-3>
    rdfs:comment "Volume 2" ;
    rdfs:comment "Access restricted: SMU users only" ;
    .


Content-negotiable representations

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