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Proceedings Papers

Multiphysics Modeling of Thorium-Based (Th, U)O 2 and (Th, Pu)O 2 Fuel Performance in a Light Water Reactor

Rong Liu, Jie-Jin Cai, Wen-Zhong Zhou, Ye Wang
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A012, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81237
https://doi.org/10.1115/ICONE26-81237
View Paper titled, Multiphysics Modeling of Thorium-Based (Th, U)O 2 and (Th, Pu)O 2 Fuel Performance in a Light Water Reactor
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Topics: Fuels, Light water reactors, Modeling, Temperature, Nuclear fission, Pressure, Mechanical properties, Nuclear reactors
Proceedings Papers

Development of a Spatial Domain Decomposition Scheme for Monte Carlo Neutron Transport

Manuel García, Diego Ferraro, Victor Hugo Sanchez-Espinoza, Luigi Mercatali, Jaakko Leppänen, Ville Valtavirta
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A042, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-82144
https://doi.org/10.1115/ICONE26-82144
View Paper titled, Development of a Spatial Domain Decomposition Scheme for Monte Carlo Neutron Transport
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Topics: Neutrons, Algorithms, Particulate matter, Simulation, Computer software, Geometry, Parallel processing
Proceedings Papers

A Neutron Transport Calculation Method for Deep Penetration and its Preliminary Verification

Wankui Yang, Baoxin Yuan, Songbao Zhang, Haibing Guo, Yaoguang Liu, Li Deng
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A032, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81709
https://doi.org/10.1115/ICONE26-81709
View Paper titled, A Neutron Transport Calculation Method for Deep Penetration and its Preliminary Verification
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Topics: Computational methods, Neutrons, Geometry, Modeling, Monte Carlo methods, China, Cross section (Physics), Finishes, Slabs, Tin
Proceedings Papers

Effect of Al 2 O 3 and SiC Content in Pack Cementation Powders on the Microstructure of SiC Coatings on HTR Graphite Spheres

Hongsheng Zhao, Ping Zhou, Ziqiang Li, Xiaoxue Liu, Kaihong Zhang, Taowei Wang, Bin Wu, Bing Liu
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A008, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81174
https://doi.org/10.1115/ICONE26-81174
View Paper titled, Effect of Al 2 O 3 and SiC Content in Pack Cementation Powders on the Microstructure of SiC Coatings on HTR Graphite Spheres
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Topics: Coatings, Graphite, Sintering, Bonding, Oxidation, Porosity, Raman spectroscopy, Raw materials, Scanning electron microscopy, Vacuum
Proceedings Papers

Development and Application of a 2D/1D Fusion Code With Leakage Reconstruction Method

Liang Liang, Zhouyu Liu, Hongchun Wu, Sheng Wang, Qian Zhang, Qiang Zhao
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A022, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81507
https://doi.org/10.1115/ICONE26-81507
View Paper titled, Development and Application of a 2D/1D Fusion Code With Leakage Reconstruction Method
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Topics: Leakage, Approximation, Diamonds, Manufacturing, Physics, Tin
Proceedings Papers

Experiences on Radioactive Materials Safe Transport in CIRP

Zhang Jiangang, Li Guoqiang, Wang Renze, Sun Hongchao, Zhuang Dajie, Sun Shutang, Meng Dongyuan
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A037, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-82009
https://doi.org/10.1115/ICONE26-82009
View Paper titled, Experiences on Radioactive Materials Safe Transport in CIRP
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Topics: Radioactive substances, China, Test facilities, Accidents, Containers, Design, Probabilistic safety assessment, Process monitoring, Radiation (Physics), Safety
Proceedings Papers

Analysis on the Dynamic Buckling Behavior of the Spacer Grid Structure

Yan Guo, Chenglong Gu, Wei Tian, Weicai Li
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A048, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-82298
https://doi.org/10.1115/ICONE26-82298
View Paper titled, Analysis on the Dynamic Buckling Behavior of the Spacer Grid Structure
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Topics: Buckling, Design, Fuels, Manufacturing, Mechanical properties, Pressurized water reactors, Stiffness, Stress, Structural elements (Construction), Vibration
Proceedings Papers

Fuel Cycle Economy of Accident Tolerant Fuel Assemblies

Xu Duoting, Liu Tong, Huang Heng
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A018, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81384
https://doi.org/10.1115/ICONE26-81384
View Paper titled, Fuel Cycle Economy of Accident Tolerant Fuel Assemblies
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Topics: Accidents, Economics , Fuels, Nuclear fuel cycle, Manufacturing, Pressurized water reactors
Proceedings Papers

A Monte Carlo Method for Simulating Stochastic Neutron Fields in Criticality Transients

Qi Xu, Zhe Wang, Gang Xiao
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A028, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81595
https://doi.org/10.1115/ICONE26-81595
View Paper titled, A Monte Carlo Method for Simulating Stochastic Neutron Fields in Criticality Transients
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Topics: Monte Carlo methods, Neutrons, Transients (Dynamics), Probability, Simulation, Statistical distributions, Accidents, Neutron sources, Nuclear fission
Proceedings Papers

Thermal and Mechanics Analysis Code of the PWR Nuclear Fuel Performance Based on FEM

Yongbo Hui, Bo Zhang, Wenhua Zhang, Di Yun, Peichao Zhai
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A013, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81295
https://doi.org/10.1115/ICONE26-81295
View Paper titled, Thermal and Mechanics Analysis Code of the PWR Nuclear Fuel Performance Based on FEM
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Topics: Finite element methods, Nuclear fuels, Pressurized water reactors, Deformation, Fuel rods, Coolants, Fuels, Safety, Stress, Temperature
Proceedings Papers

Influence of Xe-135 Dynamic Behavior on Core Operation Safety for a Molten Salt Reactor

Wu Jian-hui, Li Xiao-xiao, Hu Ji-feng, Chen Jin-gen, Yu Cheng-gang, Zou Chun-yan, Cai Xiang-zhou
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A050, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-82352
https://doi.org/10.1115/ICONE26-82352
View Paper titled, Influence of Xe-135 Dynamic Behavior on Core Operation Safety for a Molten Salt Reactor
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Topics: Molten salt reactors, Safety, Neutrons, Fuels, Helium, Nuclear reactors, Absorption, Chain, Computational methods, Density
Proceedings Papers

Modelling and Simulation Activities in Support of the UK Nuclear R&D Programme on Digital Reactor Design

B. Lindley, D. Allen, J. Lillington, A. Smethurst, P. Smith, D. Bowman, L. Dwyer, K. Lai, A. Levers, K. Vikhorev, D. Litskevich, B. Merk, E. Patelli, E. Patterson, M. Bankhead, A. Peakman, S. de Haas, C. Jackson, J. Draup, E. Galenne, G. Parks, E. Shwageraus, M. D. Eaton
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A004, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81090
https://doi.org/10.1115/ICONE26-81090
View Paper titled, Modelling and Simulation Activities in Support of the UK Nuclear R&D Programme on Digital Reactor Design
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Topics: Design, Modeling, Simulation, Pressurized water reactors, Physics, Bricks, Graphite, Nuclear design
Proceedings Papers

Experimental and Modeling Research on Leading Betavoltaic Technology

Thomas E. Adams, Shripad T. Revankar, Darrell S. Cheu, Peter Cabauy, Jesse Grant, Bret Elkind
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A056, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-82475
https://doi.org/10.1115/ICONE26-82475
View Paper titled, Experimental and Modeling Research on Leading Betavoltaic Technology
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Topics: Modeling, Electrolytes, Leakage, Temperature, Corrosion, Defense industry, Electronics, Explosions, Fire, Hazards
Proceedings Papers

Greatly Enhanced Thermal Conductivity of Fully Inert Matrix Dispersion Pellet (IMDP) Produced by Spark Plasma Sintering (SPS) Technique

Zhaodandan Ma, Tong Liu, Rui Li, Maozhou Sun, Zhiwei Lu
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A057, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-82536
https://doi.org/10.1115/ICONE26-82536
View Paper titled, Greatly Enhanced Thermal Conductivity of Fully Inert Matrix Dispersion Pellet (IMDP) Produced by Spark Plasma Sintering (SPS) Technique
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Topics: Plasmas (Ionized gases), Sintering, Thermal conductivity, Fuels, Temperature, Finite element methods, Finite element model, Geometry, Heat capacity, High temperature
Proceedings Papers

Parallel Computation of the Point Neutron Kinetic Equations Using Parallel Revisionist Integral Deferred Correction

Yun Cai, Xingjie Peng, Qing Li, Zhizhu Zhang, Zhumin Jiang, Guo Rui
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A010, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81205
https://doi.org/10.1115/ICONE26-81205
View Paper titled, Parallel Computation of the Point Neutron Kinetic Equations Using Parallel Revisionist Integral Deferred Correction
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Topics: Computation, Errors, Neutrons, Safety
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