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Proceedings Papers

Molten Salt Reactor Simulation by a MOOSE-Based Application Suanni

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Ao Zhang, Ming Dai, Shaopeng Xia, Jingen Chen, Xiangzhou Cai
Proc. ASME. ICONE31, Volume 2: Nuclear Fuel and Material, Reactor Physics and Transport Theory, and Fuel Cycle Technology, V002T02A065, August 4–8, 2024
Publisher: American Society of Mechanical Engineers
Paper No: ICONE31-136944
https://doi.org/10.1115/ICONE31-136944
View Paper titled, Molten Salt Reactor Simulation by a MOOSE-Based Application Suanni
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Topics: Cross section (Physics), Molten salt reactors, Simulation, Errors, Diffusion (Physics), Finite element methods, Fuels, Light water reactors, Thermal reactors
Proceedings Papers

Thermo-Neutronics Coupled Simulation of a Heat Pipe Reactor Based on OpenMC/COMSOL

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Jingyu Nie, Binqian Li, Yingwei Wu, Jing Zhang, Guoliang Zhang, Qisen Ren, Yanan He, Guanghui Su
Proc. ASME. ICONE31, Volume 4: SMRs, Advanced Reactors, and Fusion, V004T04A047, August 4–8, 2024
Publisher: American Society of Mechanical Engineers
Paper No: ICONE31-135246
https://doi.org/10.1115/ICONE31-135246
View Paper titled, Thermo-Neutronics Coupled Simulation of a Heat Pipe Reactor Based on OpenMC/COMSOL
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Topics: Heat pipes, Simulation, Neutrons, Physics, Cross section (Physics), Databases, Density, Design, Modeling, Nuclear reactors
Proceedings Papers

Verification of the RMC-SaraGR Nuclear Design Code System Based on the HTTR Benchmark

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Yuan Yuan, Guoming Liu, Peng Zhang
Proc. ASME. ICONE31, Volume 10: Risk Assessments and Management; Computer Code Verification and Validation; Nuclear Education and Public Acceptance, V010T13A013, August 4–8, 2024
Publisher: American Society of Mechanical Engineers
Paper No: ICONE31-135368
https://doi.org/10.1115/ICONE31-135368
View Paper titled, Verification of the RMC-SaraGR Nuclear Design Code System Based on the HTTR Benchmark
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Topics: Neutrons, Nuclear design, Uncertainty, Very high temperature reactors, Cross section (Physics), Fuels, Modeling, Rods
Proceedings Papers

Modeling and Analysis of VENUS-2 MOX Core Experiment Based on OpenMC Monte Carlo Code

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Zelong Zhao
Proc. ASME. ICONE31, Volume 10: Risk Assessments and Management; Computer Code Verification and Validation; Nuclear Education and Public Acceptance, V010T13A001, August 4–8, 2024
Publisher: American Society of Mechanical Engineers
Paper No: ICONE31-130363
https://doi.org/10.1115/ICONE31-130363
View Paper titled, Modeling and Analysis of VENUS-2 MOX Core Experiment Based on OpenMC Monte Carlo Code
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Topics: Cross section (Physics), Dosimetry, Flux (Metallurgy), Fuels, Modeling, Nuclear fission, Pins (Engineering), Sensors
Proceedings Papers

Development and Verification of a Few-Group Parameters Calculation Code TMSR-LINK for Molten Salt Reactor

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Kailong Wang, Yong Cui, Chunyan Zou, Jingen Chen, Xiangzhou Cai
Proc. ASME. ICONE31, Volume 10: Risk Assessments and Management; Computer Code Verification and Validation; Nuclear Education and Public Acceptance, V010T13A017, August 4–8, 2024
Publisher: American Society of Mechanical Engineers
Paper No: ICONE31-136987
https://doi.org/10.1115/ICONE31-136987
View Paper titled, Development and Verification of a Few-Group Parameters Calculation Code TMSR-LINK for Molten Salt Reactor
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Topics: Cross section (Physics), Molten salt reactors
Proceedings Papers

Study on Selection Method of Shielding Nuclides in Nuclear Reactor Based on Non-Dominated Sorting

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Rui Li, Shichang Liu
Proc. ASME. ICONE29, Volume 2: Nuclear Fuel and Material, Reactor Physics and Transport Theory, and Fuel Cycle Technology, V002T02A027, August 8–12, 2022
Publisher: American Society of Mechanical Engineers
Paper No: ICONE29-90902
https://doi.org/10.1115/ICONE29-90902
View Paper titled, Study on Selection Method of Shielding Nuclides in Nuclear Reactor Based on Non-Dominated Sorting
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Topics: Nuclear reactors, Nuclides, Radiation (Physics), Neutrons, Cross section (Physics), Monte Carlo methods, Photons, Absorption, Databases, Elastic scattering
Proceedings Papers

Development and Verification of Multigroup Cross Section Generation Module GroupXS in AXSP

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Kui Hu, Xubo Ma, Xuan Ma, Chen Zhang, Guanqun Jia, Yixue Chen
Proc. ASME. ICONE29, Volume 2: Nuclear Fuel and Material, Reactor Physics and Transport Theory, and Fuel Cycle Technology, V002T02A031, August 8–12, 2022
Publisher: American Society of Mechanical Engineers
Paper No: ICONE29-91399
https://doi.org/10.1115/ICONE29-91399
View Paper titled, Development and Verification of Multigroup Cross Section Generation Module GroupXS in AXSP
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Topics: Approximation, Cross section (Physics), Errors, Intellectual property, Neutrons, Nuclear reactors, Nuclides, Physics
Proceedings Papers

Implementation and Comparison of Super Homogenization Method Based on Monte Carlo and Deterministic Codes

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Yuan Yuan, Guoming Liu, Xiaodong Huo
Proc. ASME. ICONE29, Volume 14: Computer Code Verification and Validation; Nuclear Education and Public Acceptance, V014T14A006, August 8–12, 2022
Publisher: American Society of Mechanical Engineers
Paper No: ICONE29-92520
https://doi.org/10.1115/ICONE29-92520
View Paper titled, Implementation and Comparison of Super Homogenization Method Based on Monte Carlo and Deterministic Codes
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Topics: Chemical kinetics, Cross section (Physics), Fast neutron reactors, Geometry, Manufacturing, Modeling, Physics, Relaxation (Physics), Very high temperature reactors
Proceedings Papers

Validation of NUIT for Decay Heat Predictions of PWR Spent Fuel Assemblies

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Jian Li, Ruihan Li, Ding She, Jin'gang Liang, Lei Shi
Proc. ASME. ICONE29, Volume 2: Nuclear Fuel and Material, Reactor Physics and Transport Theory, and Fuel Cycle Technology, V002T02A008, August 8–12, 2022
Publisher: American Society of Mechanical Engineers
Paper No: ICONE29-89241
https://doi.org/10.1115/ICONE29-89241
View Paper titled, Validation of NUIT for Decay Heat Predictions of PWR Spent Fuel Assemblies
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Topics: Heat, Pressurized water reactors, Spent nuclear fuels, Fuels, Neutrons, Boiling water reactors, Collapse, Cooling, Cross section (Physics), Density
Proceedings Papers

Validation of TULIP via ICSBEP Critical Benchmark

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Wenjie Chen, Xianan Du, Youqi Zheng, Yongping Wang, Rong Wang, Hongchun Wu
Proc. ASME. ICONE29, Volume 15: Student Paper Competition, V015T16A113, August 8–12, 2022
Publisher: American Society of Mechanical Engineers
Paper No: ICONE29-93353
https://doi.org/10.1115/ICONE29-93353
View Paper titled, Validation of TULIP via ICSBEP Critical Benchmark
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Topics: Computational physics, Cross section (Physics), Design, Doppler effect, Feedback, Manufacturing, Neutrons, Nuclear design, Nuclear engineering, Nuclear fission
Proceedings Papers

Evaluation of Serpent Capabilities for Hyperfidelity Depletion of Pebble Bed Cores

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Yves Robert, Massimiliano Fratoni
Proc. ASME. ICONE28, Volume 1: Operating Plant Challenges, Successes, and Lessons Learned; Nuclear Plant Engineering; Advanced Reactors and Fusion; Small Modular and Micro-Reactors Technologies and Applications, V001T04A016, August 4–6, 2021
Publisher: American Society of Mechanical Engineers
Paper No: ICONE28-65810
https://doi.org/10.1115/ICONE28-65810
View Paper titled, Evaluation of Serpent Capabilities for Hyperfidelity Depletion of Pebble Bed Cores
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Topics: Collapse, Cross section (Physics), Geometry, Machinery, Modeling, Optimization, Power grids, Simulation
Proceedings Papers

Study on Collapsed Cross Section for Radial Reflector in LWR

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Ryosuke Shibano, Tatsuya Kawano, Satoshi Takeda, Takanori Kitada, Yoshitada Masaoka, Hiroaki Nagano, Hideaki Hyoudo, Yasuhiro Kodama
Proc. ASME. ICONE28, Volume 4: Student Paper Competition, V004T14A079, August 4–6, 2021
Publisher: American Society of Mechanical Engineers
Paper No: ICONE28-65686
https://doi.org/10.1115/ICONE28-65686
View Paper titled, Study on Collapsed Cross Section for Radial Reflector in LWR
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Topics: Cross section (Physics), Light water reactors
Proceedings Papers

Mixing Characteristic Measurement of Flow in Reactor Pressure Vessel by Laser Induced Fluorescent Method

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Mingpeng Chen, Guanhui Xie, Dongyang Li, Sichao Tan
Proc. ASME. ICONE28, Volume 2: Nuclear Fuels, Research, and Fuel Cycle; Nuclear Codes and Standards; Thermal-Hydraulics, V002T07A038, August 4–6, 2021
Publisher: American Society of Mechanical Engineers
Paper No: ICONE28-65770
https://doi.org/10.1115/ICONE28-65770
View Paper titled, Mixing Characteristic Measurement of Flow in Reactor Pressure Vessel by Laser Induced Fluorescent Method
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Topics: Boron, Flow measurement, Fluids, Lasers, Reactor vessels, Slug flows, Coolants, Flow (Dynamics), Pressure vessels, Annulus
Proceedings Papers

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

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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

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

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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

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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

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

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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

One Step Method for Multigroup Adjoint Neutron Flux Through Continuous Energy Monte Carlo Calculation

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Xiaotong Shang, Guanlin Shi, Kan Wang
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A043, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-82185
https://doi.org/10.1115/ICONE26-82185
View Paper titled, One Step Method for Multigroup Adjoint Neutron Flux Through Continuous Energy Monte Carlo Calculation
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Topics: Neutron flux, Neutrons, Nuclear fission, Transients (Dynamics), Cross section (Physics), Event history analysis, Monte Carlo methods, Nuclear reactor kinetics, Probability, Transient analysis
Proceedings Papers

Towards Development of Uncertainty Library for Nuclear Reactor Core Simulation

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Hany S. Abdel-Khalik, Dongli Huang, Ondrej Chvala, G. Ivan Maldonado
Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A052, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-82385
https://doi.org/10.1115/ICONE26-82385
View Paper titled, Towards Development of Uncertainty Library for Nuclear Reactor Core Simulation
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Topics: Nuclear reactors, Simulation, Uncertainty, Physics, Algorithms, Boiling water reactors, Cross section (Physics), Degrees of freedom, Modeling, Uncertainty quantification
Proceedings Papers

Statistical Burnup Distribution of Moving Pebbles in HTR-PM Reactor

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Fanny Vitullo, Jiri Krepel, Jarmo Kalilainen, Horst-Michael Prasser, Andreas Pautz
Proc. ASME. ICONE26, Volume 9: Student Paper Competition, V009T16A005, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-81082
https://doi.org/10.1115/ICONE26-81082
View Paper titled, Statistical Burnup Distribution of Moving Pebbles in HTR-PM Reactor
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Topics: China, Construction, Cross section (Physics), Discrete element methods, Engineering simulation, Flow (Dynamics), Fuels, High temperature, Matlab, Nuclear fuel cycle