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Meeting 2017 TMS Annual Meeting & Exhibition
Symposium Energy Materials 2017: Materials for Nuclear Energy
Presentation Title Friction Stir Processing of Degraded Austenitic Stainless Steel Nuclear Fuel Dry Cask Storage System Canisters
Author(s) Ben Sutton, Kenneth Ross, Glenn Grant, Gary Cannell, Greg Frederick, Robert Couch
On-Site Speaker (Planned) Ben Sutton
Abstract Scope Chloride-induced stress corrosion cracking (CISCC) of austenitic stainless steel dry cask storage system (DCSS) canisters has been identified as an industry concern. Typical DCSS canisters are constructed from Types 304 or 316 stainless steel or their variants via conventional fusion welding processes. The presence of residual tensile stress and Cr-carbide precipitation within the weld heat affected zone (HAZ) places canisters near salt-bearing environments at an elevated risk for CISCC. The current study evaluates the suitability of friction stir processing (FSP) to repair SCC and remediate sensitized fusion weld HAZs. FSP was applied to furnace sensitized Type 304 specimens containing laboratory-generated SCC and evaluated using liquid penetrant inspection, phased array ultrasonic inspection, and optical microscopy. In addition, fusion welded Type 304L specimens were fabricated, subjected to FSP, and destructively analyzed via ASTM A262 and optical microscopy. Results demonstrate that FSP is a viable option for SCC repair and sensitization remediation.
Proceedings Inclusion? Planned: Stand-alone book in which only your symposium’s papers would appear (indicate title in comments section below)


A Preliminary Investigation on the Phase Transformation Kinetics Behavior of an U-10wt%Mo Cast and Homogenized Alloy
Advanced ODS FeCrAl Alloys for Accident-tolerant Fuel Cladding
Bonding Characteristics and Site Occupancies of Si Atoms in M6C Carbides from First Principles and Experimental Study
C-17: Effect of Heat Treatments on the Microstructure and Mechanical Properties of Zr-1NB-1SN-0,1Fe Alloy used in the Nuclear Industry
C-18: Effects of Irradiation on Thermal Conductivity of Nickel Alloys
C-19: Reduced Deuterium Retention in Simultaneously Damaged and Annealed Tungsten
C-20: Studies of the Differential Thermal Analysis and Microstructural Characterization of Gd-containing Stainless Steel
Calculation of Phase Equilibria and Properties in Multi-Component Molten Salt Systems
Comparative Study of Thermal Conductivity of SiC and BeO from Ab Initio Calculations
Comparison of Corrosion Properties of Alloy 800 and Alloy 690 by In-situ Scratching Repassivation Behavior in High-temperature Pressurized Water
Compatibility Research of Fission Product Tellurium and Alloy N in Molten Salt Reactor
Development of a Novel Structural Material (SIMP steel) for Nuclear Equipment with Balanced Resistances to High Temperature, Radiation and LBE Corrosion
Ductile Phase Toughening of 90-97W-NiFe Heavy Alloys
EBSD and TEM Assessment of Deformation Localization in 718 Alloy
Effect of Steam Pressure on the Oxidation Behaviour of Alloy 625
Effects of Fe Concentration on Ion-irradiation Induced Defect Evolution and Hardening in Ni-Fe Binary Alloys
Enhancing the High-Cycle Fatigue Property of 316 Austenitic Stainless Steels through Introduction of Mechanical Twins by Cold-Drawing
Environmental Assisted Cracking of the Additively Manufactured Austenitic Stainless Steel in High Temperature Water
First Principles Investigations of Alternative Nuclear Fuels
First Principles Study of Electronic Structure and Thermo-mechanical Properties of the Components of Accident Tolerant Nuclear Fuel: UO2 and UB2
Friction Stir Processing of Degraded Austenitic Stainless Steel Nuclear Fuel Dry Cask Storage System Canisters
Fuel Cladding Materials for Supercritical Water Cooled Reactor
IASCC Behavior of Nickel-based Alloys in Light Water Reactors (LWRs)
Impact of Neutron Irradiation on Helium Desorption Behavior in Iron
In-situ Observation on the Oxides Stability under Laser and/or Electron Beams Irradiations in 9Cr-ODS Steel
Investigation of Oxidation/Carburisation Mechanisms of 9Cr Ferritic Steel Heat Exchanger Tubes
Irradiation Defects in UO2, CeO2 and (U, Ce)O2 Leached in Oxidizing Water: An In-situ Raman Study
Is There a Role for Advanced Materials in Light Water Reactors?
Microstructure Evolution of a Reactor Pressure Vessel Steel during High-temperature Tempering
Minimizing Hydrogen Diffusion through FeCrAl Alloy Accident Tolerant Fuel Cladding
Morphology of Y-Ti Nano-oxides in ODS Alloys Irradiated with High Energy Heavy Ions
Oxidation of Alloy 690 in Simulated Pressurized Water Reactor Primary Environment
Research and Development of Pressure Vessel Steels for Advanced Pressurized Water Reactors in China
Size Effects in Ion-irradiated 800H Steel at High Temperatures Utilizing Nanoindentation and Microcompression Testing
Systematic Studies on Dispersoid Stability and Swelling Resistance in ODS Alloys under Ion Irradiation Conditions
The Mechanical Response of Advanced Claddings during Proposed Reactivity Initiated Accident Conditions
Thermal Conductivity Reduction of Tungsten Plasma Facing Material Due to Helium Plasma and Cu2+ Ion Irradiation
Understanding Transuranic Binding Mechanisms and Speciation on Stainless Steel

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