1-15 of 15
Nuclear Fuel and Engineering
Close
Save search
Follow your search
Access your saved searches in your account
Would you like to receive an alert when new items match your search?
Close Modal
Sort by
Proceedings Papers

On-the-Fly Treatment of Discrete Representation Thermal Neutron Scattering Data in RMC Code

Lei Zheng, Kaiwen Li, Kan Wang
Proc. ASME. ICONE2020, Volume 1: Beyond Design Basis; Codes and Standards; Computational Fluid Dynamics (CFD); Decontamination and Decommissioning; Nuclear Fuel and Engineering; Nuclear Plant Engineering, V001T05A001, August 4–5, 2020
Publisher: American Society of Mechanical Engineers
Paper No: ICONE2020-16066
https://doi.org/10.1115/ICONE2020-16066
View Paper titled, On-the-Fly Treatment of Discrete Representation Thermal Neutron Scattering Data in RMC Code
Open the PDF for On-the-Fly Treatment of Discrete Representation Thermal Neutron Scattering Data in RMC Code in another window
Topics: Interpolation, Neutron diffraction
Proceedings Papers

Finite Element Model to Simulate the Spacer Grid Buckling Characteristics

Youngik Yoo, Joongjin Kim, Kyongbo Eom, Hyeongkoo Kim
Proc. ASME. ICONE2020, Volume 1: Beyond Design Basis; Codes and Standards; Computational Fluid Dynamics (CFD); Decontamination and Decommissioning; Nuclear Fuel and Engineering; Nuclear Plant Engineering, V001T05A002, August 4–5, 2020
Publisher: American Society of Mechanical Engineers
Paper No: ICONE2020-16082
https://doi.org/10.1115/ICONE2020-16082
View Paper titled, Finite Element Model to Simulate the Spacer Grid Buckling Characteristics
Open the PDF for Finite Element Model to Simulate the Spacer Grid Buckling Characteristics in another window
Topics: Buckling, Finite element model, Fuel rods, Manufacturing, Mechanical behavior, Nuclear fuels, Compression, Drag (Fluid dynamics), Impact testing, Mechanical properties
Proceedings Papers

Development and Verification of PWR Physical and Thermal Coupling Calculation Software

Zhigang Li
Proc. ASME. ICONE2020, Volume 1: Beyond Design Basis; Codes and Standards; Computational Fluid Dynamics (CFD); Decontamination and Decommissioning; Nuclear Fuel and Engineering; Nuclear Plant Engineering, V001T05A003, August 4–5, 2020
Publisher: American Society of Mechanical Engineers
Paper No: ICONE2020-16106
https://doi.org/10.1115/ICONE2020-16106
View Paper titled, Development and Verification of PWR Physical and Thermal Coupling Calculation Software
Open the PDF for Development and Verification of PWR Physical and Thermal Coupling Calculation Software in another window
Topics: Computer software, Cylinders, Heat conduction, Pressurized water reactors
Proceedings Papers

Study on the Thermal-Mechanical Performance of SiC Composites Cladding Under Multiple Conditions

Chunyu Yin
Proc. ASME. ICONE2020, Volume 1: Beyond Design Basis; Codes and Standards; Computational Fluid Dynamics (CFD); Decontamination and Decommissioning; Nuclear Fuel and Engineering; Nuclear Plant Engineering, V001T05A013, August 4–5, 2020
Publisher: American Society of Mechanical Engineers
Paper No: ICONE2020-16793
https://doi.org/10.1115/ICONE2020-16793
View Paper titled, Study on the Thermal-Mechanical Performance of SiC Composites Cladding Under Multiple Conditions
Open the PDF for Study on the Thermal-Mechanical Performance of SiC Composites Cladding Under Multiple Conditions in another window
Topics: Cladding systems (Building), Composite materials, Failure, Hoop stress, Fracture toughness, Probability, Temperature, Accidents, Brittleness, Ceramics
Proceedings Papers

The Experimental Research on the Mechanical Characteristics of the Newly Designed Fuel Rod Supporting Structure of Spacer Grid

Quan-yao Ren, Zeng-ping Pu, Mei-yin Zheng, Jie Chen, Mian Qin, Yuan-Ji Han, Yuan Peng, Ping Chen, Rui-rui Zhao, Jian-wei Li
Proc. ASME. ICONE2020, Volume 1: Beyond Design Basis; Codes and Standards; Computational Fluid Dynamics (CFD); Decontamination and Decommissioning; Nuclear Fuel and Engineering; Nuclear Plant Engineering, V001T05A004, August 4–5, 2020
Publisher: American Society of Mechanical Engineers
Paper No: ICONE2020-16121
https://doi.org/10.1115/ICONE2020-16121
View Paper titled, The Experimental Research on the Mechanical Characteristics of the Newly Designed Fuel Rod Supporting Structure of Spacer Grid
Open the PDF for The Experimental Research on the Mechanical Characteristics of the Newly Designed Fuel Rod Supporting Structure of Spacer Grid in another window
Topics: Deformation, Fuel rods, Mechanical properties, Springs, Stress, Deflection, Manufacturing, Neutron flux, Strips
Proceedings Papers

Study on the Spacer Grid Dynamic Crush Strength According to Cell Sizes

Jooyoung Ryu, Hangil Woo, Chaeyoung Nam, Youngik Yoo, Namgyu Park, Seungjae Lee
Proc. ASME. ICONE2020, Volume 1: Beyond Design Basis; Codes and Standards; Computational Fluid Dynamics (CFD); Decontamination and Decommissioning; Nuclear Fuel and Engineering; Nuclear Plant Engineering, V001T05A005, August 4–5, 2020
Publisher: American Society of Mechanical Engineers
Paper No: ICONE2020-16149
https://doi.org/10.1115/ICONE2020-16149
View Paper titled, Study on the Spacer Grid Dynamic Crush Strength According to Cell Sizes
Open the PDF for Study on the Spacer Grid Dynamic Crush Strength According to Cell Sizes in another window
Topics: Accidents, Additive manufacturing, Buckling, Cladding systems (Building), Coolants, Deformation, Failure, Flow (Dynamics), Fuel rods, Fuels
Proceedings Papers

Two-Dimensional Full Core Analysis of IFBA-Coated TRISO Fuel Particles in Very High Temperature Reactors

Mohammad Alrwashdeh, Saeed A. Alameri
Proc. ASME. ICONE2020, Volume 1: Beyond Design Basis; Codes and Standards; Computational Fluid Dynamics (CFD); Decontamination and Decommissioning; Nuclear Fuel and Engineering; Nuclear Plant Engineering, V001T05A014, August 4–5, 2020
Publisher: American Society of Mechanical Engineers
Paper No: ICONE2020-16838
https://doi.org/10.1115/ICONE2020-16838
View Paper titled, Two-Dimensional Full Core Analysis of IFBA-Coated TRISO Fuel Particles in Very High Temperature Reactors
Open the PDF for Two-Dimensional Full Core Analysis of IFBA-Coated TRISO Fuel Particles in Very High Temperature Reactors in another window
Topics: Fuels, Particulate matter, Very high temperature reactors, Coating processes, Coatings, Cycles, Graphite, High temperature, Monte Carlo methods, Nuclear fuel cycle
Proceedings Papers

Study on the Impact Limiter Design in Spent Fuel Transfer Cask in Nuclear Power Plants

Yuchen Hao, Yue Li, Jinhua Wang, Bin Wu, Haitao Wang
Proc. ASME. ICONE2020, Volume 1: Beyond Design Basis; Codes and Standards; Computational Fluid Dynamics (CFD); Decontamination and Decommissioning; Nuclear Fuel and Engineering; Nuclear Plant Engineering, V001T05A006, August 4–5, 2020
Publisher: American Society of Mechanical Engineers
Paper No: ICONE2020-16157
https://doi.org/10.1115/ICONE2020-16157
View Paper titled, Study on the Impact Limiter Design in Spent Fuel Transfer Cask in Nuclear Power Plants
Open the PDF for Study on the Impact Limiter Design in Spent Fuel Transfer Cask in Nuclear Power Plants in another window
Topics: Cushioning materials, Design, Nuclear power stations, Spent nuclear fuels
Proceedings Papers

Synergetic Oxidation in Alkaline In-Situ Leaching Uranium: A Preliminary Case Study

Wensheng Liao, Weimin Que, Liming Wang, Zhiming Du
Proc. ASME. ICONE2020, Volume 1: Beyond Design Basis; Codes and Standards; Computational Fluid Dynamics (CFD); Decontamination and Decommissioning; Nuclear Fuel and Engineering; Nuclear Plant Engineering, V001T05A007, August 4–5, 2020
Publisher: American Society of Mechanical Engineers
Paper No: ICONE2020-16200
https://doi.org/10.1115/ICONE2020-16200
View Paper titled, Synergetic Oxidation in Alkaline In-Situ Leaching Uranium: A Preliminary Case Study
Open the PDF for Synergetic Oxidation in Alkaline In-Situ Leaching Uranium: A Preliminary Case Study in another window
Topics: In situ processing (Mining), Oxidation, Uranium
Proceedings Papers

Stress Analysis of the Pressurized-Water Reactor Control Rod Under Operating Conditions

Leo A. Carrilho
Proc. ASME. ICONE2020, Volume 1: Beyond Design Basis; Codes and Standards; Computational Fluid Dynamics (CFD); Decontamination and Decommissioning; Nuclear Fuel and Engineering; Nuclear Plant Engineering, V001T05A015, August 4–5, 2020
Publisher: American Society of Mechanical Engineers
Paper No: ICONE2020-16872
https://doi.org/10.1115/ICONE2020-16872
View Paper titled, Stress Analysis of the Pressurized-Water Reactor Control Rod Under Operating Conditions
Open the PDF for Stress Analysis of the Pressurized-Water Reactor Control Rod Under Operating Conditions in another window
Topics: Pressurized water reactors, Stress, Stress analysis (Engineering), Water, Cladding systems (Building), Finite element model, Fluence (Radiation measurement), Temperature, Thermal expansion, Coolants
Proceedings Papers

Preventing Nuclear Fuel Material Adhesion on Glove Box Components Using Nanoparticle Coating

Tomoomi Segawa, Koichi Kawaguchi, Katsunori Ishii, Masahiro Suzuki, Joji Tachihara, Kiyoto Takato, Takatoshi Okita, Hiroshi Satone, Michitaka Suzuki
Proc. ASME. ICONE2020, Volume 1: Beyond Design Basis; Codes and Standards; Computational Fluid Dynamics (CFD); Decontamination and Decommissioning; Nuclear Fuel and Engineering; Nuclear Plant Engineering, V001T05A008, August 4–5, 2020
Publisher: American Society of Mechanical Engineers
Paper No: ICONE2020-16215
https://doi.org/10.1115/ICONE2020-16215
View Paper titled, Preventing Nuclear Fuel Material Adhesion on Glove Box Components Using Nanoparticle Coating
Open the PDF for Preventing Nuclear Fuel Material Adhesion on Glove Box Components Using Nanoparticle Coating in another window
Topics: Adhesion, Coating processes, Coatings, Nanoparticles, Nuclear fuels
Proceedings Papers

Modelling of the H2020 INSPYRE Fuel Creep Experiment

Alexander Fedorov, Kevin Zwijsen, Sander van Til
Proc. ASME. ICONE2020, Volume 1: Beyond Design Basis; Codes and Standards; Computational Fluid Dynamics (CFD); Decontamination and Decommissioning; Nuclear Fuel and Engineering; Nuclear Plant Engineering, V001T05A009, August 4–5, 2020
Publisher: American Society of Mechanical Engineers
Paper No: ICONE2020-16231
https://doi.org/10.1115/ICONE2020-16231
View Paper titled, Modelling of the H2020 INSPYRE Fuel Creep Experiment
Open the PDF for Modelling of the H2020 INSPYRE Fuel Creep Experiment in another window
Topics: Computer simulation, Creep, Fuels, Modeling, Nuclear fuels
Proceedings Papers

Multiphysics Analysis of Thorium-Based Fuel Performance Under Reactor Steady-State and Transient Accident

Chenjie Qiu, Rong Liu, Wenzhong Zhou
Proc. ASME. ICONE2020, Volume 1: Beyond Design Basis; Codes and Standards; Computational Fluid Dynamics (CFD); Decontamination and Decommissioning; Nuclear Fuel and Engineering; Nuclear Plant Engineering, V001T05A010, August 4–5, 2020
Publisher: American Society of Mechanical Engineers
Paper No: ICONE2020-16325
https://doi.org/10.1115/ICONE2020-16325
View Paper titled, Multiphysics Analysis of Thorium-Based Fuel Performance Under Reactor Steady-State and Transient Accident
Open the PDF for Multiphysics Analysis of Thorium-Based Fuel Performance Under Reactor Steady-State and Transient Accident in another window
Topics: Accidents, Fuels, Steady state, Transients (Dynamics)
Proceedings Papers

Neutronic Study on Conceptual Lithium Fluoride Salt Cooled Fusion Driven System for Actinide Transmutation

Feryantama Putra
Proc. ASME. ICONE2020, Volume 1: Beyond Design Basis; Codes and Standards; Computational Fluid Dynamics (CFD); Decontamination and Decommissioning; Nuclear Fuel and Engineering; Nuclear Plant Engineering, V001T05A011, August 4–5, 2020
Publisher: American Society of Mechanical Engineers
Paper No: ICONE2020-16368
https://doi.org/10.1115/ICONE2020-16368
View Paper titled, Neutronic Study on Conceptual Lithium Fluoride Salt Cooled Fusion Driven System for Actinide Transmutation
Open the PDF for Neutronic Study on Conceptual Lithium Fluoride Salt Cooled Fusion Driven System for Actinide Transmutation in another window
Topics: Lithium, Coolants, Design, Fuel rods, Neutron sources, Chain, Combustion, Cycles, Fast neutron reactors, Fuels
Proceedings Papers

Development of the Fuel Behavior Analysis Code for Mechanical Fuel Cladding Failure During Reactivity Insertion Event in PWR

Yuma Higashi, Nozomu Murakami, Tadakatsu Yodo, Teruhisa Yamamoto
Proc. ASME. ICONE2020, Volume 1: Beyond Design Basis; Codes and Standards; Computational Fluid Dynamics (CFD); Decontamination and Decommissioning; Nuclear Fuel and Engineering; Nuclear Plant Engineering, V001T05A012, August 4–5, 2020
Publisher: American Society of Mechanical Engineers
Paper No: ICONE2020-16674
https://doi.org/10.1115/ICONE2020-16674
View Paper titled, Development of the Fuel Behavior Analysis Code for Mechanical Fuel Cladding Failure During Reactivity Insertion Event in PWR
Open the PDF for Development of the Fuel Behavior Analysis Code for Mechanical Fuel Cladding Failure During Reactivity Insertion Event in PWR in another window
Topics: Cladding systems (Building), Failure, Fuels, Pressurized water reactors