| TÃtulo : |
Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors |
| Tipo de documento: |
documento electrónico |
| Autores: |
Jackson, John H., ; Paraventi, Denise, ; Wright, Michael, |
| Mención de edición: |
1 ed. |
| Editorial: |
[s.l.] : Springer |
| Fecha de publicación: |
2019 |
| Número de páginas: |
LXXIV, 2532 p. 1542 ilustraciones |
| ISBN/ISSN/DL: |
978-3-030-04639-2 |
| Nota general: |
Libro disponible en la plataforma SpringerLink. Descarga y lectura en formatos PDF, HTML y ePub. Descarga completa o por capítulos. |
| Palabras clave: |
Material IngenierÃa Nuclear Caracterización y Técnica AnalÃtica EnergÃa nuclear |
| Ãndice Dewey: |
620.112 |
| Resumen: |
Este conjunto de dos volúmenes representa una colección de artÃculos presentados en la 18.ª Conferencia Internacional sobre Degradación Ambiental de Materiales en Sistemas de EnergÃa Nuclear – Reactores de Agua. El objetivo de esta serie de conferencias es fomentar el intercambio de ideas sobre los problemas y sus soluciones en las centrales nucleares refrigeradas por agua de hoy y del futuro. Las contribuciones cubren los problemas que enfrentan las aleaciones a base de nÃquel, aceros inoxidables, aceros para recipientes y tuberÃas a presión, aleaciones de circonio y otras aleaciones en entornos acuáticos de relevancia. Los componentes cubiertos incluyen componentes de los lÃmites de presión, vasijas e partes internas del reactor, generadores de vapor, revestimientos de combustible, componentes irradiados, contenedores de almacenamiento de combustible y equilibrio de los componentes y sistemas de la planta. |
| Nota de contenido: |
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. |
| En lÃnea: |
https://link-springer-com.biblioproxy.umanizales.edu.co/referencework/10.1007/97 [...] |
| Link: |
https://biblioteca.umanizales.edu.co/ils/opac_css/index.php?lvl=notice_display&i |
Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors [documento electrónico] / Jackson, John H., ; Paraventi, Denise, ; Wright, Michael, . - 1 ed. . - [s.l.] : Springer, 2019 . - LXXIV, 2532 p. 1542 ilustraciones. ISBN : 978-3-030-04639-2 Libro disponible en la plataforma SpringerLink. Descarga y lectura en formatos PDF, HTML y ePub. Descarga completa o por capítulos.
| Palabras clave: |
Material IngenierÃa Nuclear Caracterización y Técnica AnalÃtica EnergÃa nuclear |
| Ãndice Dewey: |
620.112 |
| Resumen: |
Este conjunto de dos volúmenes representa una colección de artÃculos presentados en la 18.ª Conferencia Internacional sobre Degradación Ambiental de Materiales en Sistemas de EnergÃa Nuclear – Reactores de Agua. El objetivo de esta serie de conferencias es fomentar el intercambio de ideas sobre los problemas y sus soluciones en las centrales nucleares refrigeradas por agua de hoy y del futuro. Las contribuciones cubren los problemas que enfrentan las aleaciones a base de nÃquel, aceros inoxidables, aceros para recipientes y tuberÃas a presión, aleaciones de circonio y otras aleaciones en entornos acuáticos de relevancia. Los componentes cubiertos incluyen componentes de los lÃmites de presión, vasijas e partes internas del reactor, generadores de vapor, revestimientos de combustible, componentes irradiados, contenedores de almacenamiento de combustible y equilibrio de los componentes y sistemas de la planta. |
| Nota de contenido: |
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. |
| En lÃnea: |
https://link-springer-com.biblioproxy.umanizales.edu.co/referencework/10.1007/97 [...] |
| Link: |
https://biblioteca.umanizales.edu.co/ils/opac_css/index.php?lvl=notice_display&i |
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Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors / Jackson, John H. ; Paraventi, Denise ; Wright, Michael
