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Proceedings Papers

Analysis and Improvement of Global-Local Self-Shielding Calculation Scheme for AIC Control Rods

Jikui Li, Tiejun Zu, Liangzhi Cao, HongChun Wu, Qingming He
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A045, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-82196
https://doi.org/10.1115/ICONE26-82196
View Paper titled, Analysis and Improvement of Global-Local Self-Shielding Calculation Scheme for AIC Control Rods
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Topics: Rods, Resonance, Errors, Absorption, Fuels, Cross section (Physics), Neutrons, Nuclear reactors, Nuclides, Physics
Proceedings Papers

Research on Coupling Scheme of Monte Carlo Burnup Calculation in RMC

Wanlin Li, Kan Wang, Ganglin Yu, Yaodong Li
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A005, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81140
https://doi.org/10.1115/ICONE26-81140
View Paper titled, Research on Coupling Scheme of Monte Carlo Burnup Calculation in RMC
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Topics: Approximation, Computational methods, Cross section (Physics), Density, Design, Errors, Geometry, Manufacturing, Modeling, Neutron flux
Proceedings Papers

Enhancement of Stochastic Sampling Capability in RMC Code

Guanlin Shi, Yishu Qiu, Kan Wang
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A035, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81982
https://doi.org/10.1115/ICONE26-81982
View Paper titled, Enhancement of Stochastic Sampling Capability in RMC Code
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Topics: Algorithms, Nuclear fission, Safety, Sampling methods, Uncertainty, Uncertainty analysis, Uncertainty quantification
Proceedings Papers

Conceptual Core Design of HAPPY200 Reactor

Xiaosheng Li, Linsen Li, Lianghui Peng, Xiaosong Chen, Zhaocan Meng, Yaodong Chen
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A040, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-82125
https://doi.org/10.1115/ICONE26-82125
View Paper titled, Conceptual Core Design of HAPPY200 Reactor
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Topics: Design, Temperature, Cycles, Fuels, Safety, Boron, Control systems, Coolants, Heating, Manufacturing
Proceedings Papers

Effect of Final Annealing Temperature on Axial Creep Property of CZ Alloys

Lin Shi, Liutao Chen, Yang Xu, Changyuan Gao, Jun Tan, Yongjun Deng
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A025, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81540
https://doi.org/10.1115/ICONE26-81540
View Paper titled, Effect of Final Annealing Temperature on Axial Creep Property of CZ Alloys
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Topics: Alloys, Annealing, Creep, Temperature, Stress, Cladding systems (Building), China, Nuclear power, Pressurized water reactors
Proceedings Papers

Polygonal Virtual Element Spatial Discretisation Methods for the Neutron Diffusion Equation With Applications in Nuclear Reactor Physics

J. Ferguson, J. Kópházi, M. D. Eaton
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A015, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81317
https://doi.org/10.1115/ICONE26-81317
View Paper titled, Polygonal Virtual Element Spatial Discretisation Methods for the Neutron Diffusion Equation With Applications in Nuclear Reactor Physics
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Topics: Diffusion (Physics), Neutrons, Nuclear reactors, Physics, Finite element methods, Algorithms, Finite difference methods, Geometry, Mesh generation, Partial differential equations
Proceedings Papers

Strength Limit of Thimble Tube With Material Plasticity Under Bending Moment and Axial Compression Force

Hisashi Koike, Masaji Mori, Daisuke Fujiwara, Takashi Shimomura
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A020, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81468
https://doi.org/10.1115/ICONE26-81468
View Paper titled, Strength Limit of Thimble Tube With Material Plasticity Under Bending Moment and Axial Compression Force
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Topics: ASME Standards, Compression, Earthquakes, Evaluation methods, Fuels, Manufacturing, Plasticity, Plastics, Pressurized water reactors, Stress
Proceedings Papers

Finite Element Analysis for Fuel Assembly Structural Behavior

Youngik Yoo, Kyounghong Kim, Kyongbo Eom, Seongki Lee, Jongsung Yoo
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A030, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81621
https://doi.org/10.1115/ICONE26-81621
View Paper titled, Finite Element Analysis for Fuel Assembly Structural Behavior
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Topics: Finite element analysis, Fuels, Manufacturing, Accidents, Computer software, Failure mechanisms, Finite element model, Mechanical behavior, Mechanical testing, Radioactive substances
Proceedings Papers

Fast Sub-Grid Scale Finite Element Method for the First Order Neutron Transport Equation

Chao Fang, Hongchun Wu, Liangzhi Cao, Yunzhao Li
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A036, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-82008
https://doi.org/10.1115/ICONE26-82008
View Paper titled, Fast Sub-Grid Scale Finite Element Method for the First Order Neutron Transport Equation
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Topics: Finite element methods, Neutrons, Convection, Finite element analysis, Vacuum
Proceedings Papers

Advances in FEMN: The Code for Nuclear Noise Analysis Based on ICEM-CFD

Baoxin Yuan, Herong Zeng, Wankui Yang, Songbao Zhang
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A006, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81169
https://doi.org/10.1115/ICONE26-81169
View Paper titled, Advances in FEMN: The Code for Nuclear Noise Analysis Based on ICEM-CFD
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Topics: Computational fluid dynamics, Ferromanganese, Noise (Sound), Geometry, Finite element methods, Neutrons, Physics, Computer software, Computer-aided design, Finite element analysis
Proceedings Papers

A Space-Time Parallel Method to Solve Space-Dependent Neutron Kinetics Equations in Hexagonal-Z Geometry

Zhizhu Zhang, Yun Cai, Xingjie Peng, Qing Li
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A011, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81213
https://doi.org/10.1115/ICONE26-81213
View Paper titled, A Space-Time Parallel Method to Solve Space-Dependent Neutron Kinetics Equations in Hexagonal-Z Geometry
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Topics: Geometry, Neutrons, Spacetime, Algorithms, Architecture, Nuclear reactor kinetics, Safety, Stability, Thread, Transients (Dynamics)
Proceedings Papers

Investigation on Breakaway Oxidation Behavior of CZ Alloys

Chang-yuan Gao, Liu-tao Chen, Lin Shi, Yang Xu, Jun Tan, Yong-jun Deng
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A026, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81541
https://doi.org/10.1115/ICONE26-81541
View Paper titled, Investigation on Breakaway Oxidation Behavior of CZ Alloys
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Topics: Alloys, Oxidation, China, Hydrogen, Nuclear power, Steam, Accidents, Coolants, Failure mechanisms, Fuels
Proceedings Papers

Preliminary Development of a TRISO Fuel Performance Analysis Code: FFAT

Jian Li, Ding She, Lei Shi, Jing Zhao
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A046, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-82242
https://doi.org/10.1115/ICONE26-82242
View Paper titled, Preliminary Development of a TRISO Fuel Performance Analysis Code: FFAT
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Topics: Fuels, Particulate matter, Failure, Very high temperature reactors, Accidents, Equipment performance, Gas cooled reactors, High temperature, Pressure vessels, Probability
Proceedings Papers

Interior Penalty Schemes for Discontinuous Isogeometric Methods With an Application to Nuclear Reactor Physics

S. G. Wilson, J. Kópházi, A. R. Owens, M. D. Eaton
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A016, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81322
https://doi.org/10.1115/ICONE26-81322
View Paper titled, Interior Penalty Schemes for Discontinuous Isogeometric Methods With an Application to Nuclear Reactor Physics
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Topics: Nuclear reactors, Physics, Diffusion (Physics), Neutrons, Errors, Finite element analysis, Fuels, Manufacturing, Numerical stability
Proceedings Papers

A Theoretical Model of the Stress Intensity Factor Threshold of DHC for Fuel Cladding Tube

Liang Chen, Lili Liu, Xiaoming Song, Hua Pang
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A031, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81665
https://doi.org/10.1115/ICONE26-81665
View Paper titled, A Theoretical Model of the Stress Intensity Factor Threshold of DHC for Fuel Cladding Tube
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Topics: Cladding systems (Building), Fuels, Stress, Temperature, Fracture (Materials), Temperature gradient, Cracking (Materials), Diffusion (Physics), Fracture (Process), Hydrogen
Proceedings Papers

Analyzing the Impact of Solutes on PKA Spectrum for Simulation of Neutron Induced-Radiation Damage in Zr-Based Metals

Guang-bo Cai, Ya-dong Zhang, Yu-xiang Han, Jia-pei Yu
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A041, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-82132
https://doi.org/10.1115/ICONE26-82132
View Paper titled, Analyzing the Impact of Solutes on PKA Spectrum for Simulation of Neutron Induced-Radiation Damage in Zr-Based Metals
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Topics: Damage, Metals, Neutrons, Radiation (Physics), Simulation, Zirconium, Computer simulation, Alloys, Atoms, Spectra (Spectroscopy)
Proceedings Papers

The Design of Nuclear Noise Frequency Spectrum Analysis System and Helium Bubble Movement Behavior Computation for Narrow Channel Core

Herong Zeng, Baoxin Yuan
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A007, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81170
https://doi.org/10.1115/ICONE26-81170
View Paper titled, The Design of Nuclear Noise Frequency Spectrum Analysis System and Helium Bubble Movement Behavior Computation for Narrow Channel Core
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Topics: Bubbles, Computation, Design, Emission spectroscopy, Helium, Noise (Sound), Spectroscopy, Data acquisition, Data acquisition systems, Fault diagnosis
Proceedings Papers

Preliminary Neutron Simulation of Ceramic Fast Reactor

Xuesong Yan, Xunchao Zhang, Yaling Zhang, Lei Yang, Wenshan Duan
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A021, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81474
https://doi.org/10.1115/ICONE26-81474
View Paper titled, Preliminary Neutron Simulation of Ceramic Fast Reactor
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Topics: Ceramics, Fast neutron reactors, Neutrons, Simulation, Absorption, Coolants, Energy generation, Nuclear fuels, Nuclear power, Nuclear reactors
Proceedings Papers

Development of Three-Dimensional Neutron Kinetics Code Based on High Order Nodal Expansion Method in Hexagonal-Z Geometry

Guo Chao, Liu Yu, He Hangxing, Liu Luguo, Wang Xiaoyu, Xin Sufang, Li Peiyang, Wu Xiaoli, Yuan Hongsheng
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A017, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81356
https://doi.org/10.1115/ICONE26-81356
View Paper titled, Development of Three-Dimensional Neutron Kinetics Code Based on High Order Nodal Expansion Method in Hexagonal-Z Geometry
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Topics: Geometry, Neutrons, Feedback, Polynomials, Approximation, Cross section (Physics), Leakage, Runge-Kutta methods
Proceedings Papers

Criticality Safety Issues in Nuclear Design of HP-SMTCs Space Reactor

Songyang Li, Dingqu Wang, Yueyuan Jiang, Weihua Li, Wenli Guo
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A047, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-82244
https://doi.org/10.1115/ICONE26-82244
View Paper titled, Criticality Safety Issues in Nuclear Design of HP-SMTCs Space Reactor
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Topics: Nuclear design, Safety, Accidents, Risk, Seawater, Absorption, Design, Heat, Neutrons, Nuclear power