Volume 5 Issue 1 | ASME J. of Nuclear Rad Sci.

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Editorial

Igor L. Pioro

ASME J of Nuclear Rad Sci. January 2019, 5(1): 010201. doi: https://doi.org/10.1115/1.4041694

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Topics: China , Nuclear engineering , Radiation (Physics) , Electricity (Physics) , Nuclear power

Reviewer's Recognition

ASME J of Nuclear Rad Sci. January 2019, 5(1): 010202. doi: https://doi.org/10.1115/1.4041823

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ASME Nuclear Engineering Division Chair’s Message

Leon Cizelj

ASME J of Nuclear Rad Sci. January 2019, 5(1): 010203. doi: https://doi.org/10.1115/1.4041695

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Topics: Nuclear engineering

The CNS President's Message

Shoujun Wang

ASME J of Nuclear Rad Sci. January 2019, 5(1): 010204. doi: https://doi.org/10.1115/1.4041888

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Topics: China , Nuclear engineering , Nuclear power , Japan Society of Mechanical Engineers , Nuclear industry

Guest Editorial

A Happy New Year…

Koji Okamoto

ASME J of Nuclear Rad Sci. January 2019, 5(1): 010301. doi: https://doi.org/10.1115/1.4041822

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Topics: Accidents , Fukushima nuclear disaster, Japan, 2011 , Japan Society of Mechanical Engineers , Nuclear engineering , Nuclear industry , Nuclear power , Nuclear power stations , Safety , Nuclear decommissioning , Collaboration

The AP1000 Plant and the China Project Commissioning Success

Bryan N. Friedman

ASME J of Nuclear Rad Sci. January 2019, 5(1): 010302. doi: https://doi.org/10.1115/1.4041973

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View articletitled, The AP1000 Plant and the China Project Commissioning Success

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Topics: China , Safety , Cooling , Containment , Heat , Nuclear power stations , Plant design , Design , Accidents

Research Papers

Development and Validation of SAS4A Code and Its Application to Analyses on Severe Flow Blockage Accidents in a Sodium-Cooled Fast Reactor

Yoshitaka Fukano

ASME J of Nuclear Rad Sci. January 2019, 5(1): 011001. doi: https://doi.org/10.1115/1.4040649

Abstract

View articletitled, Development and Validation of SAS4A Code and Its Application to Analyses on Severe Flow Blockage Accidents in a Sodium-Cooled Fast Reactor

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Topics: Cladding systems (Building) , Control systems , Coolants , Flow (Dynamics) , Fuels , Sodium fast reactors , Accidents , Safety , Melting , Signals

Thermo-Mechanical Safety Analyses of Preliminary Design Experiments for ²³⁸Pu Production

Christopher J. Hurt, James D. Freels, Prashant K. Jain, G. Ivan Maldonado

ASME J of Nuclear Rad Sci. January 2019, 5(1): 011002. doi: https://doi.org/10.1115/1.4041269

Abstract

View articletitled, Thermo-Mechanical Safety Analyses of Preliminary Design Experiments for 238Pu Production

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Topics: Heat , Irradiation (Radiation exposure) , Safety , Temperature , Manufacturing , Heat transfer , Design , Shrinkage (Materials) , Geometry , Cycles

Breeding Properties Study on High-Power Thorium Molten Salt Reactor

Yafen Liu, Rui Guo, Xiangzhou Cai, Rui Yan, Yang Zou, Bo Zhou

ASME J of Nuclear Rad Sci. January 2019, 5(1): 011003. doi: https://doi.org/10.1115/1.4041272

Abstract

View articletitled, Breeding Properties Study on High-Power Thorium Molten Salt Reactor

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Topics: Fuels , Molten salt reactors , Nuclides , Security , Temperature , Transients (Dynamics) , Fluids , Nuclear fission , Graphite

Experimental Investigations on Carryover in a Gravity Separation-Based Steam Drum

R. K. Bagul, D. S. Pilkhwal, P. K. Vijayan, J. B. Joshi

ASME J of Nuclear Rad Sci. January 2019, 5(1): 011004. doi: https://doi.org/10.1115/1.4041791

Abstract

View articletitled, Experimental Investigations on Carryover in a Gravity Separation-Based Steam Drum

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Topics: Drops , Flow (Dynamics) , Steam , Water , Particulate matter , Separation (Technology) , Gravity (Force)

A Thoria and Thorium Uranium Dioxide Nuclear Fuel Performance Model Prototype and Knowledge Gap Assessment

J. S. Bell, P. K. Chan, A. Prudil

ASME J of Nuclear Rad Sci. January 2019, 5(1): 011005. doi: https://doi.org/10.1115/1.4039778

Abstract

View articletitled, A Thoria and Thorium Uranium Dioxide Nuclear Fuel Performance Model Prototype and Knowledge Gap Assessment

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Topics: Fuels , Temperature , Uranium , Irradiation (Radiation exposure) , Deformation , Thermal conductivity , Modeling , Nuclear fission

Corrosion in Iron and Steel T91 Caused by Flowing Lead–Bismuth Eutectic at 400 °C and 10⁻⁷ Mass% Dissolved Oxygen

Carsten Schroer, Valentyn Tsisar, Adeline Durand, Olaf Wedemeyer, Aleksandr Skrypnik, Jürgen Konys

ASME J of Nuclear Rad Sci. January 2019, 5(1): 011006. doi: https://doi.org/10.1115/1.4040937

Abstract

View articletitled, Corrosion in Iron and Steel T91 Caused by Flowing Lead–Bismuth Eutectic at 400 °C and 10−7 Mass% Dissolved Oxygen

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Topics: Corrosion , Iron , Oxidation , Oxygen , Steel

Seismic Damage at the Fukushima Daiichi Nuclear Power Station With Emphasis on Unit 1

Kazuhiro Akimoto

ASME J of Nuclear Rad Sci. January 2019, 5(1): 011007. doi: https://doi.org/10.1115/1.4040651

Abstract

View articletitled, Seismic Damage at the Fukushima Daiichi Nuclear Power Station With Emphasis on Unit 1

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Topics: Damage , Earthquakes , Tsunamis

New Exploration on TMSR: Redesign of the TMSR Lattice

J. K. Zhao, S. Y. Si, Q. C. Chen, H. Bei

ASME J of Nuclear Rad Sci. January 2019, 5(1): 011008. doi: https://doi.org/10.1115/1.4041192

Abstract

View articletitled, New Exploration on TMSR: Redesign of the TMSR Lattice

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Topics: Design , Fuels , Graphite , Optimization , Molten salt reactors , Neutrons , Cladding systems (Building) , Safety

Predicting Leak Rate Through Valve Stem Packing in Nuclear Applications

Ali Salah Omar Aweimer, Abdel-Hakim Bouzid, Mehdi Kazeminia

ASME J of Nuclear Rad Sci. January 2019, 5(1): 011009. doi: https://doi.org/10.1115/1.4040493

Abstract

View articletitled, Predicting Leak Rate Through Valve Stem Packing in Nuclear Applications

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Coupled Three-Dimensional Neutronics and Thermal-Hydraulics Analysis for SCWR Core Typical Transients

Wang Lianjie, Lu Di, Zhao Wenbo

ASME J of Nuclear Rad Sci. January 2019, 5(1): 011010. doi: https://doi.org/10.1115/1.4041693

Abstract

View articletitled, Coupled Three-Dimensional Neutronics and Thermal-Hydraulics Analysis for SCWR Core Typical Transients

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Topics: Safety , Supercritical water reactors , Thermal hydraulics , Transient analysis , Transients (Dynamics) , Accidents , Fuels , Temperature , Water , Density

Cycle Calculations of a Small-Scale Heat Removal System With Supercritical CO₂ as Working Fluid

Marcel Straetz, Joerg Starflinger, Rainer Mertz, Dieter Brillert

ASME J of Nuclear Rad Sci. January 2019, 5(1): 011011. doi: https://doi.org/10.1115/1.4039884

Abstract

View articletitled, Cycle Calculations of a Small-Scale Heat Removal System With Supercritical CO2 as Working Fluid

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Topics: Cycles , Glass , Heat , Pressure , Pressurized water reactors , Temperature , Turbines , Compressors , Fluids , Supercritical carbon dioxide

Obstructed View Factor Calculations for Multiple Obstructions in Closed Cavities

Fiaz Mahmood, Huasi Hu

ASME J of Nuclear Rad Sci. January 2019, 5(1): 011012. doi: https://doi.org/10.1115/1.4041563

Abstract

View articletitled, Obstructed View Factor Calculations for Multiple Obstructions in Closed Cavities

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Topics: Cavities , Geometry , Pellet fusion , Radiation (Physics) , Shades and shadows , String , Computer software , FORTRAN

Evaluation of Interfacial and Permeation Leaks in Gaskets and Compression Packing

Ali Salah Omar Aweimer, Abdel-Hakim Bouzid

ASME J of Nuclear Rad Sci. January 2019, 5(1): 011013. doi: https://doi.org/10.1115/1.4041691

Abstract

View articletitled, Evaluation of Interfacial and Permeation Leaks in Gaskets and Compression Packing

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Topics: Gaskets , Leakage , Packing (Shipments) , Packings (Cushioning) , Stress , Finishes

Performance of Aluminide and Cr-Modified Aluminide Pack Cementation-Coated Stainless Steel 304 in Supercritical Water at 700 °C

Nick Tepylo, Xiao Huang, Shengli Jiang, Sami Penttilä

ASME J of Nuclear Rad Sci. January 2019, 5(1): 011014. doi: https://doi.org/10.1115/1.4040889

Abstract

View articletitled, Performance of Aluminide and Cr-Modified Aluminide Pack Cementation-Coated Stainless Steel 304 in Supercritical Water at 700 °C

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Topics: Coating processes , Coatings , Oxidation , Spallation (Nuclear physics) , Stainless steel , Supercritical water reactors , Water , Weight (Mass) , X-ray diffraction , Alloys

Simulation on Pellet–Cladding Mechanical Interaction of Accident Tolerant Fuel With Coated Cladding

Yangbin Deng, Yingwei Wu, Dalin Zhang, Wenxi Tian, G. H. Su, Suizheng Qiu

ASME J of Nuclear Rad Sci. January 2019, 5(1): 011015. doi: https://doi.org/10.1115/1.4041194

Abstract

View articletitled, Simulation on Pellet–Cladding Mechanical Interaction of Accident Tolerant Fuel With Coated Cladding

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Topics: Accidents , Cladding systems (Building) , Fuels , Stress , Pressure , Simulation , Coatings , Zirconium

Guided Waves as an Online Monitoring Technology for Long-Term Operation in Nuclear Power Plants: Experimental Results on a Steam Discharge Pipe

Mauro Cappelli, Francesco Cordella, Francesco Bertoncini, Marco Raugi

ASME J of Nuclear Rad Sci. January 2019, 5(1): 011016. doi: https://doi.org/10.1115/1.4041193

Abstract

View articletitled, Guided Waves as an Online Monitoring Technology for Long-Term Operation in Nuclear Power Plants: Experimental Results on a Steam Discharge Pipe

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Topics: Nuclear power stations , Pipes , Probes , Steam , Waves , Symmetry (Physics)

Coupling of a Reactor Analysis Code and a Lower Head Thermal Analysis Solver

Hiroshi Madokoro, Alexei Miassoedov, Thomas Schulenberg

ASME J of Nuclear Rad Sci. January 2019, 5(1): 011017. doi: https://doi.org/10.1115/1.4041278

Abstract

View articletitled, Coupling of a Reactor Analysis Code and a Lower Head Thermal Analysis Solver

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Topics: Cooling , Heat , Heat transfer , Temperature , Thermal analysis , Vessels , Heating , Accidents , Water , Heat flux

Calibration of HPGe Detector Efficiencies With Self-Absorption Correction of Gas Sphere Sources

Waseem Khan, Chaohui He

ASME J of Nuclear Rad Sci. January 2019, 5(1): 011018. doi: https://doi.org/10.1115/1.4041338

Abstract

View articletitled, Calibration of HPGe Detector Efficiencies With Self-Absorption Correction of Gas Sphere Sources

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Topics: Absorption , Sensors , Photons , Calibration , Nuclides , Simulation , Density

Analyses of the Costs Associated With Very High Turbine Entry Temperatures in Helium Recuperated Gas Turbine Cycles for Generation IV Nuclear Power Plants

A. Gad-Briggs, P. Pilidis, T. Nikolaidis

ASME J of Nuclear Rad Sci. January 2019, 5(1): 011019. doi: https://doi.org/10.1115/1.4041276

Abstract

View articletitled, Analyses of the Costs Associated With Very High Turbine Entry Temperatures in Helium Recuperated Gas Turbine Cycles for Generation IV Nuclear Power Plants

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Topics: Compressors , Cooling , Cycles , Helium , Nuclear power stations , Temperature , Turbines , Gas turbines , Pressure , Turbine blades

Transport Mechanism of Steam Methane Reforming on Fixed Bed Catalyst Heated by High Temperature Helium for Hydrogen Production: A Computational Fluid Dynamics Investigation

Feng Wang, Ziqiang Yang, Long Wang, Qiang Wen

ASME J of Nuclear Rad Sci. January 2019, 5(1): 011020. doi: https://doi.org/10.1115/1.4040377

Abstract

View articletitled, Transport Mechanism of Steam Methane Reforming on Fixed Bed Catalyst Heated by High Temperature Helium for Hydrogen Production: A Computational Fluid Dynamics Investigation

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Topics: Helium , Hydrogen production , Methane , Temperature , Catalysts , Steam , Thermal efficiency , High temperature

Technical Brief

A Study on Hydrogen Emission of Zirconium Hydride

Fu Xiaogang, Qin Bo, Ma Haoran, Zhang Jinquan, Long Bin

ASME J of Nuclear Rad Sci. January 2019, 5(1): 014501. doi: https://doi.org/10.1115/1.4041274

Abstract

View articletitled, A Study on Hydrogen Emission of Zirconium Hydride

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Topics: Emissions , Hydrogen , Zirconium , Cladding systems (Building)

Editorial

Editorial

Reviewer's Recognition

ASME Nuclear Engineering Division Chair’s Message

The CNS President's Message

Guest Editorial

A Happy New Year…

The AP1000 Plant and the China Project Commissioning Success

Research Papers

Development and Validation of SAS4A Code and Its Application to Analyses on Severe Flow Blockage Accidents in a Sodium-Cooled Fast Reactor

Thermo-Mechanical Safety Analyses of Preliminary Design Experiments for ²³⁸Pu Production

Breeding Properties Study on High-Power Thorium Molten Salt Reactor

Experimental Investigations on Carryover in a Gravity Separation-Based Steam Drum

A Thoria and Thorium Uranium Dioxide Nuclear Fuel Performance Model Prototype and Knowledge Gap Assessment

Corrosion in Iron and Steel T91 Caused by Flowing Lead–Bismuth Eutectic at 400 °C and 10⁻⁷ Mass% Dissolved Oxygen

Seismic Damage at the Fukushima Daiichi Nuclear Power Station With Emphasis on Unit 1

New Exploration on TMSR: Redesign of the TMSR Lattice

Predicting Leak Rate Through Valve Stem Packing in Nuclear Applications

Coupled Three-Dimensional Neutronics and Thermal-Hydraulics Analysis for SCWR Core Typical Transients

Cycle Calculations of a Small-Scale Heat Removal System With Supercritical CO₂ as Working Fluid

Obstructed View Factor Calculations for Multiple Obstructions in Closed Cavities

Evaluation of Interfacial and Permeation Leaks in Gaskets and Compression Packing

Performance of Aluminide and Cr-Modified Aluminide Pack Cementation-Coated Stainless Steel 304 in Supercritical Water at 700 °C

Simulation on Pellet–Cladding Mechanical Interaction of Accident Tolerant Fuel With Coated Cladding

Guided Waves as an Online Monitoring Technology for Long-Term Operation in Nuclear Power Plants: Experimental Results on a Steam Discharge Pipe

Coupling of a Reactor Analysis Code and a Lower Head Thermal Analysis Solver

Calibration of HPGe Detector Efficiencies With Self-Absorption Correction of Gas Sphere Sources

Analyses of the Costs Associated With Very High Turbine Entry Temperatures in Helium Recuperated Gas Turbine Cycles for Generation IV Nuclear Power Plants

Transport Mechanism of Steam Methane Reforming on Fixed Bed Catalyst Heated by High Temperature Helium for Hydrogen Production: A Computational Fluid Dynamics Investigation

Technical Brief

A Study on Hydrogen Emission of Zirconium Hydride

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