TAILIEUCHUNG - Corrosion and mechanical properties of hot-rolled 0.5%Gd-0.8%Bstainless steels in a simulated nuclear waste treatment solution

The austenite and ferrite phases of the steels are about 88:12. The average austenite and ferrite grain size of the plane normal to rolling, transverse and normal directions of the hot rolled specimens are about , , , , and [mm], respectively. The average micro-hardness of the as-cast specimen is Hv, whereas, that of the hot-rolled specimen are , and [Hv] for the plane normal to RD, TD and ND, respectively. | Corrosion and mechanical properties of hot-rolled steels in a simulated nuclear waste treatment solution Nuclear Engineering and Technology 51 (2019) 207e213 Contents lists available at ScienceDirect Nuclear Engineering and Technology journal homepage: Original Article Corrosion and mechanical properties of hot-rolled stainless steels in a simulated nuclear waste treatment solution Moo Young Jung a, Youl Baik a, Yong Choi a, . Sohn b, * a Department of Materials Science and Engineering, Dankook University, Dandaer-ro 119, Cheonan, Chungnam 31116, Republic of Korea b Department of Nuclear Engineering, UNIST, 50 UNIST-gil, Eonyang-eup, Ulju-gun, Ulsan 689-798, Republic of Korea a r t i c l e i n f o a b s t r a c t Article history: Corrosion and mechanical behavior of the hot-rolled steel to develop a spent Received 16 April 2018 nuclear fuel storage material was studied in a simulated nuclear waste treatment condition with rolling Received in revised form condition. The austenite and ferrite phases of the steels are about 88:12. The 3 August 2018 average austenite and ferrite grain size of the plane normal to rolling, transverse and normal directions of Accepted 21 August 2018 Available online 10 September 2018 the hot rolled specimens are about , , , , and [mm], respectively. The average micro-hardness of the as-cast specimen is Hv, whereas, that of the hot-rolled specimen are , and [Hv] for the plane normal to RD, TD and ND, respectively. The UTS, YS and Keywords: Gd-B-stainless steel elongation of the as-cast and the hot-rolled specimen are 699, 484 [MPa], , and 654, 432 [MPa] and Corrosion , respectively. The passivity was observed both for the as-cast and the hot rolled specimens in a Neutron absorbing materials simulated .

• Năng lượng
• Nuclear engineering and technology
• Hot rolled 0
• 5%Gd 0
• 8%Bstainless steels
• Mechanical properties of hot rolled Bstainless steels
• Simulated nuclear waste treatment solution
• Integral nuclear data validation
• Experimental spent nuclear fuel compositions
• Economic Co operation and Development
• Nuclear Energy Agency
• Nuclear Criticality Safety
• Expanded spent fuel isotopic compositions
• Nuclear reactor vessel water level prediction during severe accidents
• Deep neural networks
• Nuclear power plants
• Maintain nuclear reactor integrity
• Special issue on the International Symposium
• Future Instrumentation and Control for Nuclear Power Plants
• The Nuclear Engineering and Technolog
• Parameter identifiability of Boolean networks
• Application to fault diagnosis of nuclear plants
• Fault diagnosis of nuclear plants
• Nuclear Heating Reactor
• Sensor selection strategy
• Radioactive effluents released from Korean nuclear power plants
• The resulting radiation doses to members of the public
• Korean nuclear power plants
• Nuclear power plant
• The recycling of nuclear graphite after micro oxidation
• Nuclear graphite after micro oxidation
• Whether ground micro oxidation graphite
• The oxidized graphite particles
• The recycling of nuclear graphite
• Statistical model for forecasting uranium prices
• Estimate the nuclear fuel cycle cost
• Nuclear Fuel Cycle Cost
• Nuclear fuel cycle cost calculation
• The statistical autoregressive integrated moving average
• Distinct properties of tungsten austenitic stainless alloy
• Potential nuclear engineering material
• Interchanging the molybdenum in standard SS316 by tungsten
• Gamma and neutron shielding Interchanging the molybdenum
• Nuclear engineering and design
• Brayton cycle
• Supercritical carbon dioxide
• Nuclear engineering
• Safety analysis
• Control strategy
• Seismic fragility evaluation
• The base isolated nuclear power plant piping system
• The failure criterion based on stress strain
• Nuclear power plant piping system
• Seismic fragility analysis
• BNL NRC benchmark model
• Feasibility study on photo production
• 99mTc with the nuclear resonance fluorescence
• The nuclear resonance fluorescence
• High intensity laser Compton scattering
• Improved PCA method
• Sensor fault detection and isolation in a nuclear power plant
• Sensor fault detection and isolation
• Influences of heating processes on properties
• Microstructure of porous CeO2 beads
• Surrogate for nuclear fuels fabricated
• Microfluidic sol gel process
• The ceramic nuclear fuel particles
• Sensitivity analysis of thermal hydraulic parameters
• The corrosion intensity in nuclear power plant steam generators
• Nuclear power plant steam generators
• The failure of steam generators
• The heat exchangers
• Architectural model driven dependability analysis
• Computer based safety system in nuclear power plant
• Safety system in nuclear power plant
• Embedded Real Time Safety Systems
• Mitigation of high energy arcing faults
• Nuclear power plant medium voltage switchgear
• High energy arcing fault
• Ultra high performance fiber reinforced concrete nuclear waste container
• High level waste
• Ultra high performance fiber reinforced concrete material
• The dry storage container
• Canadian Nuclear facilities
• Real time estimation of break sizes during LOCA
• Nuclear power plants using NARX neural network
• Nonlinear autoregressive with exogenous inputs
• Loss of coolant accidents
• Nordic research and development cooperation
• Strengthen nuclear reactor safety after the Fukushima accident
• The Fukushima accident
• Nordic Nuclear Safety Research
• Seismic response characteristics
• Isolated AP1000 nuclear shield building subjected
• Beyond design basis earthquake shaking
• Beyond Design Basis Earthquake
• Nuclear Shield Building
• Uncertainty analysis of containment dose rate
• Containment dose rate for core damage assessment
• Containment Radiation Monitor
• Core Damage Assessment
• An evaluation method for nuclear fuel debrisefiltering performance
• Nuclear fuel debrisefiltering performance
• Comprehensive Debris Filtering Efficiency
• Debris Filtering Performance
• Debris Induced Failure
• Development of earthquake instrumentation for shutdown
• Restart criteria of the nuclear power plant
• Multivariable decision making process
• The nuclear power plant
• Development of an information reference system
• Reconstruction models of nuclear power plants
• Work site reconstruction models
• Six nuclear power plant workers
• Nondestructive inspection of spent nuclear fuel storage canisters
• Shear horizontal guided waves
• Spent nuclear fuel storage canisters
• The cask protecting the canister
• Geometric constrains inside