Materials ageing in light-water reactors : handbook of destructive assays / / François Cattant

• Autore: Cattant François
• Pubblicazione: Cham, Switzerland : , : Springer, , [2022]; ©2022
• Edizione: 2nd edition.
• Descrizione fisica: 1 online resource (2448 pages)
• Disciplina: 621.4833
• Soggetto topico: Light water reactors - Materials - Deterioration; Nuclear engineering - Materials
• 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).
• Sommario/riassunto: "Materials aging in light water reactors” is a guide to destructive testing of materials in nuclear power plants intended for engineers, researchers and experts concerned with managing the aging of materials used in nuclear power plants. underlying processes, and to acquire new knowledge one must observe. The publication of the book "Materials Aging in Light Water Reactors - Handbook of Destructive Assays" is dedicated to the observation on real samples of nuclear power plants.
• Titolo autorizzato: Materials Ageing in Light-Water Reactors
• ISBN: 9783030856007; 9783030855994
• Formato: Materiale a stampa
• Livello bibliografico: Monografia
• Lingua di pubblicazione: Inglese
• Record Nr.: 9910585779203321
• Lo trovi qui: Univ. Federico II