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Ion-Irradiation-Induced Damage in Nuclear Materials : Case Study of a-SiO₂ and MgO / / by Diana Bachiller Perea
| Ion-Irradiation-Induced Damage in Nuclear Materials : Case Study of a-SiO₂ and MgO / / by Diana Bachiller Perea |
| Autore | Bachiller Perea Diana |
| Edizione | [1st ed. 2018.] |
| Pubbl/distr/stampa | Cham : , : Springer International Publishing : , : Imprint : Springer, , 2018 |
| Descrizione fisica | 1 online resource (191 pages) |
| Disciplina | 621.4833 |
| Collana | Springer Theses, Recognizing Outstanding Ph.D. Research |
| Soggetto topico |
Materials science
Energy systems Nuclear fusion Nuclear energy Characterization and Evaluation of Materials Energy Systems Nuclear Fusion Nuclear Energy |
| ISBN | 3-030-00407-4 |
| Formato | Materiale a stampa  |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | eng |
| Nota di contenuto | Introduction -- Part I Materials and Methods -- Studied Materials: a-SiO2 and MgO -- Ion-Solid Interactions and Ion Beam Modification of Materials -- Experimental Facilities -- Experimental Characterization Techniques -- Part II Ion Beam Induced Luminescence in Amorphous Silica -- General Features of the Ion Beam Induced Luminescence in Amorphous Silica -- Ionoluminescence in Silica: Role of the Silanol Group Content and the Ion Stopping Power -- Exciton Mechanisms and Modeling of the Ionoluminescence in Silica -- Part III Ion-Irradiation Damage in MgO -- MgO under Ion Irradiation at High Temperatures -- Ion Beam Induced Luminescence in MgO -- Conclusions and Prospects for the Future. |
| Record Nr. | UNINA-9910298591403321 |
Bachiller Perea Diana
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| Cham : , : Springer International Publishing : , : Imprint : Springer, , 2018 | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. Federico II | ||
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Materials ageing in light-water reactors : handbook of destructive assays / / François Cattant
| Materials ageing in light-water reactors : handbook of destructive assays / / François Cattant |
| Autore | Cattant François |
| Edizione | [2nd edition.] |
| Pubbl/distr/stampa | Cham, Switzerland : , : Springer, , [2022] |
| Descrizione fisica | 1 online resource (2448 pages) |
| Disciplina | 621.4833 |
| Soggetto topico |
Light water reactors - Materials - Deterioration
Nuclear engineering - Materials |
| ISBN |
9783030856007
9783030855994 |
| Formato | Materiale a stampa |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | eng |
| Nota di contenuto |
Intro -- 2010 Foreword -- 2020 Foreword -- For a Better Next Decade -- Acknowledgments -- Goal of the 2020 Revision -- Contents -- About the Author -- Acronyms -- Part I Fundamentals, Degradation Mechanisms, Failures of Nickel Alloys, Heat Exchangers and Cold Worked Stainless Steels -- 1 Introduction -- 2 Fundamentals of Light Water Reactors -- 2.1 Background -- 2.2 Basics of Pressurized Water Reactors -- 2.3 Basics of Boiling Water Reactors -- 3 Failure and Ageing Mechanisms -- 3.1 Background -- 3.2 Corrosion -- 3.2.1 Aqueous Corrosion -- 3.2.2 Atmospheric Corrosion -- 3.2.3 Hot Oxidation -- 3.3 Cavitation Erosion -- 3.3.1 Mechanism Identification -- 3.3.2 Application Domain -- 3.3.3 Mechanism Description -- 3.3.4 Mechanism Impact -- 3.3.5 Influencing Conditions -- 3.3.6 Components Susceptible to Cavitation Erosion -- 3.3.7 Preventing Cavitation Erosion -- 3.4 Fatigue -- 3.4.1 Mechanism Identification -- 3.4.2 Application Domain -- 3.4.3 Mechanism Description -- 3.4.4 Understanding and Keeping Fatigue Under Control -- 3.4.5 Mechanism Impacts -- 3.4.6 Initiation Parameters List -- 3.4.7 Potentially Susceptible Components -- 3.4.8 Preventing Fatigue -- 3.5 Vibration Fatigue -- 3.5.1 Foreword -- 3.5.2 Definitions -- 3.5.3 Mechanism Identification -- 3.5.4 Application Domain -- 3.5.5 Mechanism Description -- 3.5.6 Mechanism Consequences -- 3.5.7 Influent Parameters -- 3.5.8 Potentially Concerned Components -- 3.5.9 Preventing Vibration Fatigue -- 3.6 Environmentally Assisted Fatigue -- 3.6.1 Mechanism Identification -- 3.6.2 Mechanism Description -- 3.6.3 Mechanism Consequences -- 3.6.4 Influent Parameters -- 3.6.5 Potentially Susceptible Components -- 3.6.6 Preventing Primary Water Assisted Fatigue -- 3.6.7 Corrosion Fatigue -- 3.7 Excessive Deformation and Plastic Instability -- 3.7.1 Definitions -- 3.7.2 Materials and Components of Concern.
3.8 Elastic or Plastic Instability-Buckling -- 3.8.1 Definitions -- 3.8.2 Materials and Components of Concern -- 3.9 Progressive Deformation -- 3.9.1 Definition -- 3.9.2 Materials and Components of Concern -- 3.10 Fast Fracture in the Ductile Regime -- 3.10.1 Definition -- 3.10.2 Materials and Components of Concern -- 3.11 Fast Fracture in the Brittle Regime and in the Fragile/Ductile Region -- 3.11.1 Definition -- 3.11.2 Background -- 3.11.3 Metallurgical and Mechanical Aspects of Cleavage Fracture -- 3.11.4 Main Materials and Components of Concern -- 3.11.5 Materials Mechanical Characterization -- 3.11.6 Fracture in the Brittle-Ductile Transition Domain -- 3.11.7 Intergranular Fracture -- 3.12 Austenitic Stainless Steels Irradiation Embrittlement -- 3.12.1 Mechanism Identification -- 3.12.2 Application Domain -- 3.12.3 Mechanism Description -- 3.12.4 List of Influent Parameters -- 3.12.5 Components of Potential Concern -- 3.12.6 Preventing Austenitic SSs Irradiation Embrittlement -- 3.13 Austenitic Stainless Steels Irradiation Creep -- 3.13.1 Foreword -- 3.13.2 Mechanism Identification -- 3.13.3 Application Domain -- 3.13.4 Mechanism Description -- 3.13.5 Mechanism Impacts -- 3.13.6 List of Influent Parameters -- 3.13.7 Potentially Affected Components -- 3.13.8 Preventing Austenitic Stainless Steels Irradiation Creep -- 3.14 Irradiation Assisted Stress Corrosion Cracking of Austenitic Stainless Steels -- 3.14.1 Mechanism Identification -- 3.14.2 Application Domain -- 3.14.3 Mechanism Description -- 3.14.4 Various IASCC Mechanisms -- 3.14.5 List of Influent Parameters -- 3.14.6 Potentially Affected Components -- 3.14.7 Preventing IASCC -- 3.15 RPV Steels Neutron Irradiation Embrittlement -- 3.15.1 Description of the 2 Embrittlement Classes (Hardening and not Hardening) -- 3.15.2 Mechanism Identification -- 3.15.3 Application Field. 3.15.4 Mechanism Description -- 3.15.5 Mechanism Impact -- 3.15.6 List of Influent Parameters -- 3.15.7 How to Mitigate RPV Steels Neutron Irradiation Embrittlement -- 3.16 Swelling Under Irradiation -- 3.16.1 Mechanism Identification -- 3.16.2 Application Domain -- 3.16.3 Mechanism Description -- 3.16.4 Preventing Swelling -- 3.17 Cast Stainless Steels Thermal Ageing -- 3.17.1 Mechanism Identification -- 3.17.2 Application Domain -- 3.17.3 Mechanism Description -- 3.17.4 Mechanism Consequences -- 3.17.5 List of Influent Parameters -- 3.17.6 Potentially Concerned Components -- 3.17.7 How to Prevent and Mitigate Cast SSs Thermal Ageing -- 3.18 α′ Precipitation Ageing of Martensitic Stainless Steels -- 3.18.1 Mechanism Identification -- 3.18.2 Application Domains -- 3.18.3 Mechanism Description -- 3.18.4 Mechanism Effects -- 3.18.5 Influent Parameters List -- 3.18.6 Potentially Affected Components -- 3.18.7 How to Prevent and Mitigate α′ Precipitation Ageing -- 3.19 Low Alloy Steels and Carbon Steels Thermal Ageing or Temper Embrittlement -- 3.19.1 Foreword -- 3.19.2 Mechanism Identification -- 3.19.3 Domains of Relevance -- 3.19.4 Phenomenon Description -- 3.19.5 Mechanism Consequences -- 3.19.6 List of Influent Parameters -- 3.19.7 Susceptible Components -- 3.19.8 Preventing and Mitigating Temper Embrittlement -- 3.20 Thermal Ageing of 30% Chromium Nickel Base Alloys, Ordering -- 3.20.1 Mechanism Description -- 3.20.2 Preventing SRO and LRO -- 3.21 Wear -- 3.21.1 General Description -- 3.21.2 Ashby Maps: Wear Mechanisms -- 3.21.3 Influence of Particles -- 3.21.4 Wear Consequences -- 3.21.5 Influent Parameters -- 3.21.6 Preventing Wear -- References -- 4 Materials Properties -- 4.1 Austenitic Stainless Steels -- 4.2 Ni Alloys -- 4.3 High Strength Alloys -- 4.4 Carbon and Low Alloy Steels -- 4.5 Hard-Facing Alloys -- 4.6 Copper Alloys. 4.7 Titanium Alloys -- 4.8 Materials Forbidden in the Containment Building -- 4.8.1 Materials in the Containment Building Atmosphere -- 4.8.2 Polluting Materials -- 5 Nickel Base Alloys -- 5.1 Background -- 5.2 Destructive Examinations Related to Reactor Pressure Vessel Issues-Results and Remediation -- 5.2.1 Reactor Pressure Vessel Outlet Nozzle Cracking -- 5.2.2 Reactor Pressure Vessel Outlet Nozzle Repair Cracking -- 5.2.3 Reactor Pressure Vessel Outlet Nozzle Dissimilar Weld Cracking -- 5.2.4 Reactor Pressure Vessel Outlet Nozzle Leak -- 5.2.5 Destructive Examination of a Boat Sample Removed From a Leaking Bottom Mounted Instrumentation Nozzle at a W Plant -- 5.2.6 Laboratory Analysis of a Boat Sample Removed From a Leaking Bottom Mounted Instrumentation Nozzle at a CE Plant (Hyres 2015) -- 5.2.7 Laboratory Analysis of a Bottom Mounted Instrumentation Nozzle at an Areva Plant (Derniaux 2018) -- 5.2.8 Synthesis of the Destructive Examinations Carried Out on EDF Reactor Pressure Vessel Head Penetrations -- 5.2.9 Destructive Examination of a Boat Sample Harvested From a Leaking Penetration of a B& -- W Unit -- 5.2.10 Replica of a Leaking Control Rod Drive Mechanism Penetration From an MHI Unit -- 5.2.11 Destructive Examination of a Boat Sample Harvested From a Penetration of a W Unit -- 5.2.12 Destructive Examination of a Retired Reactor Pressure Vessel Head -- 5.2.13 Destructive Examination of a Control Element Drive Mechanism Repaired with A52 -- 5.2.14 Leak of a Reactor Pressure Vessel Head Vent Nozzle at a KHIC-CE Unit ([PRI-10-02, 2010]) -- 5.3 X-750 Field Experience -- 5.3.1 Destructive Examinations of X-750 Split Pins-Results and Remediation -- 5.3.2 Destructive Examination of X-750 Clevis Bolts (Hyres 2014) -- 5.4 Destructive Examinations Related to Pressurizer Issues-Results and Remediation. 5.4.1 Destructive Examination of an Instrumentation Nozzle Leaking at First Outage -- 5.4.2 Destructive Examination of a Leaking Instrumentation Nozzle -- 5.4.3 Laboratory Analysis of a Pressurizer Safety Nozzle -- 5.5 Destructive Examinations Related to Steam Generator Issues-Results and Remediation -- 5.5.1 Destructive Examination of a Leaking SG Blowdown Nozzle of a Framatome Unit -- 5.5.2 Destructive Examinations of Leaking SG Blowdown Nozzles of KHIC-CE Units (Chung 2007 (Hwang et al. 2008) [PRI-07-10] [PRI-08-06]) -- 5.5.3 Synthesis of the Steam Generator Channel Heads Destructive Examinations -- 5.5.4 Destructive Examination of a Steam Generator Inlet Nozzle Dissimilar Metal Weld -- References -- 6 Steam Generator Tubes, Plugs, Sleeves and Heat Exchangers -- 6.1 Background -- 6.2 Materials Properties -- 6.3 Steam Generator Tubes Examinations-Results and Remediation -- 6.3.1 Tubes with Primary Water Stress Corrosion Cracking -- 6.3.2 Tubes with OD Initiated Corrosion (IGSCC, IGA, TGSCC) -- 6.3.3 Tubes with ID and OD Initiated Corrosion -- 6.3.4 Tubes with Wear -- 6.3.5 Tube with Defects in the Tubesheet -- 6.3.6 Tubes with Bulging Above the Tubesheet -- 6.3.7 Fatigue Cracking of U-bends (Boccanfuso et al. 2014b -- Duisabeau et al. 2014) -- 6.4 Steam Generator Tubes Plugs-Destructive Examination Results and Remediation -- 6.5 Steam Generator Tubes Sleeves-Destructive Examination of a Welded Sleeve -- 6.6 Steam Generators Blowdown Heat Exchangers Degradations in Operation (Praud et al. 2014) -- References -- 7 Stress Corrosion Cracking of Cold Worked Stainless Steels -- 7.1 Background -- 7.2 Destructive Examinations-Results and Remediation -- 7.2.1 Destructive Examination of 2 Thermocouple Clamping Devices (Staples) -- 7.2.2 Destructive Examination of a Cracked Alloy A-286 Vent Valve Jackscrew (Fyfitch et al. 2014). 7.2.3 Stress Corrosion Cracking of A286 Reactor Coolant Pump Turning Vane Bolts (Ickes and Ruminski 2019). |
| Record Nr. | UNINA-9910585779203321 |
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Cattant François
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| Cham, Switzerland : , : Springer, , [2022] | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. Federico II | ||
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Materials in Nuclear Applications
| Materials in Nuclear Applications |
| Pubbl/distr/stampa | [Place of publication not identified], : American Society for Testing & Materials, 1960 |
| Descrizione fisica | 1 online resource (vi, 344 pages) |
| Disciplina | 621.4833 |
| Soggetto topico |
Materials - Effect of radiation on
Nuclear reactors - Materials |
| ISBN |
9780803156722
0803156723 |
| Formato | Materiale a stampa |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | eng |
| Record Nr. | UNINA-9910164757303321 |
| [Place of publication not identified], : American Society for Testing & Materials, 1960 | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. Federico II | ||
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Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors / / edited by John H. Jackson, Denise Paraventi, Michael Wright
| Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors / / edited by John H. Jackson, Denise Paraventi, Michael Wright |
| Edizione | [1st ed. 2019.] |
| Pubbl/distr/stampa | Cham : , : Springer International Publishing : , : Imprint : Springer, , 2019 |
| Descrizione fisica | 1 online resource (2,532 pages) |
| Disciplina | 621.4833 |
| Collana | The Minerals, Metals & Materials Series |
| Soggetto topico |
Materials - Analysis
Nuclear engineering Characterization and Analytical Technique Nuclear Energy |
| ISBN |
9783030046392
3030046397 |
| Formato | Materiale a stampa |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | eng |
| Nota di contenuto | 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 -- Developmentof 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 StainlessSteels 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 Toolfor 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 Spectroscop. |
| Record Nr. | UNINA-9910337929403321 |
| Cham : , : Springer International Publishing : , : Imprint : Springer, , 2019 | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. Federico II | ||
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Properties of Reactor Structural Alloys after Neutron or Particle Irradiation
| Properties of Reactor Structural Alloys after Neutron or Particle Irradiation |
| Autore | Baroch C. J |
| Pubbl/distr/stampa | [Place of publication not identified], : American Society for Testing & Materials, 1975 |
| Descrizione fisica | 1 online resource (619 pages) : illustrations |
| Disciplina | 621.4833 |
| Soggetto topico |
Nuclear reactors - Materials
Alloys - Effect of radiation on |
| ISBN | 0-8031-4653-1 |
| Formato | Materiale a stampa |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | eng |
| Record Nr. | UNINA-9910164739603321 |
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Baroch C. J
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| [Place of publication not identified], : American Society for Testing & Materials, 1975 | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. Federico II | ||
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Research reactors for the development of materials and fuels for innovative nuclear energy systems / / IAEA
| Research reactors for the development of materials and fuels for innovative nuclear energy systems / / IAEA |
| Pubbl/distr/stampa | Vienna, Austria : , : International Atomic Energy Agency, c2017, , 2017 |
| Descrizione fisica | 1 online resource (44 pages) : illustrations |
| Disciplina | 621.4833 |
| Collana | IAEA Nuclear Energy Series |
| Soggetto topico |
Nuclear reactors - Materials
Materials testing reactors |
| Soggetto genere / forma | Electronic books. |
| ISBN | 92-0-136619-1 |
| Formato | Materiale a stampa |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | eng |
| Record Nr. | UNINA-9910467058403321 |
| Vienna, Austria : , : International Atomic Energy Agency, c2017, , 2017 | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. Federico II | ||
| ||
Research reactors for the development of materials and fuels for innovative nuclear energy systems / / IAEA
| Research reactors for the development of materials and fuels for innovative nuclear energy systems / / IAEA |
| Pubbl/distr/stampa | Vienna, Austria : , : International Atomic Energy Agency, c2017, , 2017 |
| Descrizione fisica | 1 online resource (44 pages) : illustrations |
| Disciplina | 621.4833 |
| Collana | IAEA Nuclear Energy Series |
| Soggetto topico |
Nuclear reactors - Materials
Materials testing reactors |
| ISBN | 92-0-136619-1 |
| Formato | Materiale a stampa |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | eng |
| Record Nr. | UNINA-9910796203903321 |
| Vienna, Austria : , : International Atomic Energy Agency, c2017, , 2017 | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. Federico II | ||
| ||
Research reactors for the development of materials and fuels for innovative nuclear energy systems / / IAEA
| Research reactors for the development of materials and fuels for innovative nuclear energy systems / / IAEA |
| Pubbl/distr/stampa | Vienna, Austria : , : International Atomic Energy Agency, c2017, , 2017 |
| Descrizione fisica | 1 online resource (44 pages) : illustrations |
| Disciplina | 621.4833 |
| Collana | IAEA Nuclear Energy Series |
| Soggetto topico |
Nuclear reactors - Materials
Materials testing reactors |
| ISBN | 92-0-136619-1 |
| Formato | Materiale a stampa |
| Livello bibliografico | Monografia |
| Lingua di pubblicazione | eng |
| Record Nr. | UNINA-9910814726903321 |
| Vienna, Austria : , : International Atomic Energy Agency, c2017, , 2017 | ||
| Materiale a stampa | ||
| Lo trovi qui: Univ. Federico II | ||
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