Volume 4 Issue 1 | ASME J. of Nuclear Rad Sci.

Topics: Flow (Dynamics) , Heat transfer , Supercritical water reactors , Thermal hydraulics , Water , Computational fluid dynamics , Annulus , Fuels , Carbon dioxide , Fluids

Research Challenges of Heat Transfer to Supercritical Fluids

X. Cheng, X. J. Liu

ASME J of Nuclear Rad Sci. January 2018, 4(1): 011003. doi: https://doi.org/10.1115/1.4037117

Abstract

View articletitled, Research Challenges of Heat Transfer to Supercritical Fluids

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Topics: Heat transfer , Supercritical fluids , Temperature , Wall temperature , Water , Numerical analysis , Heat flux , Fluids , Turbulence , Flow (Dynamics)

Assessment of Convective Heat Transfer Correlations Against an Expanded Database for Different Fluids at Supercritical Pressures

Hussam A. M. Zahlan, Laurence K. H. Leung, Yan-Ping Huang, Guang-Xu Liu

ASME J of Nuclear Rad Sci. January 2018, 4(1): 011004. doi: https://doi.org/10.1115/1.4037720

Abstract

View articletitled, Assessment of Convective Heat Transfer Correlations Against an Expanded Database for Different Fluids at Supercritical Pressures

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Topics: Carbon dioxide , Databases , Flow (Dynamics) , Fluids , Fuel rods , Water , Heat transfer , Uncertainty , Annulus

Transient Heat Transfer in an Out-of-Pile SCWR Fuel Assembly Test at Near-Critical Pressure

Thomas Schulenberg, Hongbo Li

ASME J of Nuclear Rad Sci. January 2018, 4(1): 011005. doi: https://doi.org/10.1115/1.4038061

Abstract

View articletitled, Transient Heat Transfer in an Out-of-Pile SCWR Fuel Assembly Test at Near-Critical Pressure

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Topics: Heat transfer , Pressure , Supercritical water reactors , Temperature , Transient heat transfer , Fuels , Film boiling , Rods , Heat flux , Transients (Dynamics)

Critical Heat Flux of Flowing Water in Tube for Pressure Up to Near Critical Point—Experiment and Prediction

Yuzhou Chen, Minfu Zhao, Keming Bi, Bin Yang, Dongxu Zhang, Kaiwen Du

ASME J of Nuclear Rad Sci. January 2018, 4(1): 011006. doi: https://doi.org/10.1115/1.4038215

Abstract

View articletitled, Critical Heat Flux of Flowing Water in Tube for Pressure Up to Near Critical Point—Experiment and Prediction

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Topics: Critical heat flux , Pressure , Water , Subcooling

Large Eddy Simulation Study on Forced Convection Heat Transfer to Water at Supercritical Pressure in a Trapezoid Annulus

Muhsin Mohd Amin, Yu Duan, Shuisheng He

ASME J of Nuclear Rad Sci. January 2018, 4(1): 011007. doi: https://doi.org/10.1115/1.4038161

Abstract

View articletitled, Large Eddy Simulation Study on Forced Convection Heat Transfer to Water at Supercritical Pressure in a Trapezoid Annulus

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Topics: Flow (Dynamics) , Forced convection , Pressure , Temperature , Water , Heat flux , Large eddy simulation , Heating , Annulus , Fluids

Computational Fluid Dynamic Simulations of Heat Transfer From a 2 × 2 Wire-Wrapped Fuel Rod Bundle to Supercritical Pressure Water

Krishna Podila, Yanfei Rao

ASME J of Nuclear Rad Sci. January 2018, 4(1): 011008. doi: https://doi.org/10.1115/1.4037747

Abstract

View articletitled, Computational Fluid Dynamic Simulations of Heat Transfer From a 2 × 2 Wire-Wrapped Fuel Rod Bundle to Supercritical Pressure Water

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Topics: Computational fluid dynamics , Flow (Dynamics) , Fuel rods , Heat transfer , Pressure , Simulation , Turbulence , Water , Wire , Temperature

Burnable Poison Design for Supercritical Water Cooled Reactor With Annular Fuel

Zhao Chuanqi, Wang Kunpeng, Cao Liangzhi, Zheng Youqi

ASME J of Nuclear Rad Sci. January 2018, 4(1): 011009. doi: https://doi.org/10.1115/1.4037119

Abstract

View articletitled, Burnable Poison Design for Supercritical Water Cooled Reactor With Annular Fuel

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Topics: Design , Fuels , Manufacturing , Water , Density

Fuel Assembly Concept of the Canadian Supercritical Water-Cooled Reactor

Metin Yetisir, Holly Hamilton, Rui Xu, Michel Gaudet, David Rhodes, Mitch King, Kittmer Andrew, Ben Benson

ASME J of Nuclear Rad Sci. January 2018, 4(1): 011010. doi: https://doi.org/10.1115/1.4037818

Abstract

View articletitled, Fuel Assembly Concept of the Canadian Supercritical Water-Cooled Reactor

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Topics: Fuels , Manufacturing , Water , Pressure

Investigation of Fuel Burnup Impacts on Nuclear Reactor Safety Parameters in the Canadian Pressure Tube Supercritical Water-Cool Reactor

Ahmad Moghrabi, David Raymond Novog

ASME J of Nuclear Rad Sci. January 2018, 4(1): 011011. doi: https://doi.org/10.1115/1.4037895

Abstract

View articletitled, Investigation of Fuel Burnup Impacts on Nuclear Reactor Safety Parameters in the Canadian Pressure Tube Supercritical Water-Cool Reactor

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Topics: Coolants , Fuels , Nuclear fission , Temperature , Supercritical water reactors

Improved Core Design of a High Breeding Fast Reactor Cooled by Supercritical Pressure Light Water

Sukarman, Akifumi Yamaji, Takayuki Someya

ASME J of Nuclear Rad Sci. January 2018, 4(1): 011012. doi: https://doi.org/10.1115/1.4037719

Abstract

View articletitled, Improved Core Design of a High Breeding Fast Reactor Cooled by Supercritical Pressure Light Water

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Topics: Coolants , Design , Fuels , Manufacturing , Pressure , Water , Fast neutron reactors , Power density

Study on Optimization Design for CSR1000 Core

Lianjie Wang, Ping Yang, Di Lu, Wenbo Zhao

ASME J of Nuclear Rad Sci. January 2018, 4(1): 011013. doi: https://doi.org/10.1115/1.4037669

Abstract

View articletitled, Study on Optimization Design for CSR1000 Core

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Topics: Cycles , Design , Fuels , Optimization , Steady state , Flow (Dynamics) , Coolants , Manufacturing

Numerical Simulation Research on the Performance of SCWR Fuel Rod

Changbing Tang, Shuo Xing, Hua Pang, Ping Chen, Yi Zhou

ASME J of Nuclear Rad Sci. January 2018, 4(1): 011014. doi: https://doi.org/10.1115/1.4038060

Abstract

View articletitled, Numerical Simulation Research on the Performance of SCWR Fuel Rod

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Topics: Cladding systems (Building) , Computer simulation , Fuel rods , Fuels , Irradiation (Radiation exposure) , Stainless steel , Supercritical water reactors , Temperature , Thermal expansion , Thermal conductivity

Microstructure Study of NiCrAlY and FeCrAlY Exposed to Superheated Steam at 800 °C

Alberto Sáez-Maderuelo, Michael McTaggart, Xiao Huang, César Maffiotte

ASME J of Nuclear Rad Sci. January 2018, 4(1): 011015. doi: https://doi.org/10.1115/1.4037668

Abstract

View articletitled, Microstructure Study of NiCrAlY and FeCrAlY Exposed to Superheated Steam at 800 °C

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Topics: Alloys , Coatings , Corrosion , Grain boundaries , Oxidation , Steam , Superheating , Temperature , Weight (Mass) , Supercritical water reactors

Miniature Autoclave and Double Bellows Loading Device for Material Testing in Future Reactor Concept Conditions—Case Supercritical Water

Sami Penttilä, Pekka Moilanen, Wade Karlsen, Aki Toivonen

ASME J of Nuclear Rad Sci. January 2018, 4(1): 011016. doi: https://doi.org/10.1115/1.4037897

Abstract

View articletitled, Miniature Autoclave and Double Bellows Loading Device for Material Testing in Future Reactor Concept Conditions—Case Supercritical Water

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Topics: Bellows (Equipment) , Pressure , Stress , Water , Temperature , Materials testing , Light water reactors , Testing

Corrosion Behaviour of Bare and NiCrAlY Coated Alloy 214 in Supercritical Water at 700 °C

Suzan Bsat, Xiao Huang, Sami Penttila

ASME J of Nuclear Rad Sci. January 2018, 4(1): 011017. doi: https://doi.org/10.1115/1.4037324

Abstract

View articletitled, Corrosion Behaviour of Bare and NiCrAlY Coated Alloy 214 in Supercritical Water at 700 °C

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Topics: Alloys , Corrosion , Water , Temperature , Oxidation , Weight (Mass) , Supercritical water reactors

Numerical Simulation of Supercritical Carbon Dioxide Critical Flow in the Nozzle Tube

Zhou Yuan, Wang Yangle, Chen Jingtan, Xia Zhaoyang, Wang Junfeng

ASME J of Nuclear Rad Sci. January 2018, 4(1): 011018. doi: https://doi.org/10.1115/1.4037896

Abstract

View articletitled, Numerical Simulation of Supercritical Carbon Dioxide Critical Flow in the Nozzle Tube

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Topics: Flow (Dynamics) , Nozzles , Pressure , Simulation results , Temperature , Supercritical carbon dioxide , Errors , Computer simulation , Carbon dioxide

Characterization of Alloy 214 After Exposure to Superheated Steam at 800 °C

Nick Tepylo, Rainier Garcia Sanchez, Xiao Huang

ASME J of Nuclear Rad Sci. January 2018, 4(1): 011019. doi: https://doi.org/10.1115/1.4037718

Abstract

View articletitled, Characterization of Alloy 214 After Exposure to Superheated Steam at 800 °C

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Topics: Alloys , Steam , Superheating , Weight (Mass) , Oxidation , Temperature

Editorial

Editorial Greeting to Journal Readers 2018

Reviewer's Recognition

ASME/NED Chair's Message

Professor Pavel Leonidovich Kirillov on His 90th Birthday

Guest Editorial

Special Issue: Selected Papers from the 8th International Symposium on Supercritical Water-Cooled Reactors

Research Papers

Summary on the Results of Two Computational Fluid Dynamic Benchmarks of Tube and Different Channel Geometries

Thermal-Hydraulics Program in Support of Canadian SCWR Concept Development

Research Challenges of Heat Transfer to Supercritical Fluids

Assessment of Convective Heat Transfer Correlations Against an Expanded Database for Different Fluids at Supercritical Pressures

Transient Heat Transfer in an Out-of-Pile SCWR Fuel Assembly Test at Near-Critical Pressure

Critical Heat Flux of Flowing Water in Tube for Pressure Up to Near Critical Point—Experiment and Prediction

Large Eddy Simulation Study on Forced Convection Heat Transfer to Water at Supercritical Pressure in a Trapezoid Annulus

Computational Fluid Dynamic Simulations of Heat Transfer From a 2 × 2 Wire-Wrapped Fuel Rod Bundle to Supercritical Pressure Water

Burnable Poison Design for Supercritical Water Cooled Reactor With Annular Fuel

Fuel Assembly Concept of the Canadian Supercritical Water-Cooled Reactor

Investigation of Fuel Burnup Impacts on Nuclear Reactor Safety Parameters in the Canadian Pressure Tube Supercritical Water-Cool Reactor

Improved Core Design of a High Breeding Fast Reactor Cooled by Supercritical Pressure Light Water

Study on Optimization Design for CSR1000 Core

Numerical Simulation Research on the Performance of SCWR Fuel Rod

Microstructure Study of NiCrAlY and FeCrAlY Exposed to Superheated Steam at 800 °C

Miniature Autoclave and Double Bellows Loading Device for Material Testing in Future Reactor Concept Conditions—Case Supercritical Water

Corrosion Behaviour of Bare and NiCrAlY Coated Alloy 214 in Supercritical Water at 700 °C

Numerical Simulation of Supercritical Carbon Dioxide Critical Flow in the Nozzle Tube

Characterization of Alloy 214 After Exposure to Superheated Steam at 800 °C

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