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Journal Articles

Optimized Moderator Design and Analysis of a Pin-Type Supercritical Carbon Dioxide Reactor Based on Reactor Monte Carlo Code

Xingyu Zhao, Minyun Liu, Rongyi Cui, Shanfang Huang, Kan Wang, Chuan Lu

ASME J of Nuclear Rad Sci. October 2023, 9(4): 041303.

Paper No: NERS-22-1137

DOI: https://doi.org/10.1115/1.4056772

Published Online: May 31, 2023

View Article titled, Optimized Moderator Design and Analysis of a Pin-Type Supercritical Carbon Dioxide Reactor Based on Reactor Monte Carlo Code

Topics: Coolants, Design, Diluents, Fuels, Leakage, Neutrons, Supercritical carbon dioxide, Nuclear fission, Nuclides, Feedback

Journal Articles

Fabrication Aspects and Performance Characterization of α-Al 2 O 3 /ALPO 4 Based Sandwich Configuration Flow Channel Inserts and Coatings for High Temperature Liquid Metal Applications

Abhishek Saraswat, Rajendraprasad Bhattacharyay, Paritosh Chaudhuri, Sateesh Gedupudi

ASME J of Nuclear Rad Sci.

Paper No: NERS-23-1007

DOI: https://doi.org/10.1115/1.4062646

Published Online: May 27, 2023

View Article titled, Fabrication Aspects and Performance Characterization of α-Al 2 O 3 /ALPO 4 Based Sandwich Configuration Flow Channel Inserts and Coatings for High Temperature Liquid Metal Applications

Topics: Coatings, Flow (Dynamics), High temperature, Liquid metals, Manufacturing, Performance characterization, Insulation, Coolants, Corrosion, Fusion reactors

Journal Articles

System Thermal-hydraulics Model for Fluoride Salt-Cooled Reactor Based On Small Advanced High Temperature Reactor (smAHTR) Design Concept

Shu Jun Wang, Xianmin Huang, Blair Bromley

ASME J of Nuclear Rad Sci.

Paper No: NERS-23-1011

DOI: https://doi.org/10.1115/1.4062500

Published Online: May 5, 2023

View Article titled, System Thermal-hydraulics Model for Fluoride Salt-Cooled Reactor Based On Small Advanced High Temperature Reactor (smAHTR) Design Concept

Topics: Design, High temperature, Thermal hydraulics, Fuels, Coolants, Temperature, Cycles, Engineering simulation, Flow (Dynamics), Simulation

Journal Articles

Neutronic and Thermal-Fluidic Analyses for an Additive Manufactured Reactor With SiC Matrix and TRISO Particle Fuel

Wenbin Han, Jian Deng, Qi Lu, Chong Chen, Youyou Xu, Tao Zhang, Shanfang Huang

ASME J of Nuclear Rad Sci. October 2023, 9(4): 041501.

Paper No: NERS-22-1145

DOI: https://doi.org/10.1115/1.4062119

Published Online: April 17, 2023

View Article titled, Neutronic and Thermal-Fluidic Analyses for an Additive Manufactured Reactor With SiC Matrix and TRISO Particle Fuel

Topics: Additive manufacturing, Computational fluid dynamics, Design, Fuels, Particulate matter, Temperature, Coolants, Neutrons, Heat transfer, Computer software

Journal Articles

Eddy Current Flow Meter Measurements in Liquid Sodium at High Temperatures

Nico Krauter, Alexandru Onea, Gunter Gerbeth, Sven Eckert

ASME J of Nuclear Rad Sci.

Paper No: NERS-22-1185

DOI: https://doi.org/10.1115/1.4062239

Published Online: March 31, 2023

View Article titled, Eddy Current Flow Meter Measurements in Liquid Sodium at High Temperatures

Topics: Eddies (Fluid dynamics), Eddy currents (Electricity), Flowmeters, High temperature, Sodium, Sensors, Flow (Dynamics), Liquid metals, Coolants, Engineering prototypes

Journal Articles

Development of a Passive Reactor Shutdown Device to Prevent Core Disruptive Accidents in Fast Reactors: A Study on Device Specifications

Koji Morita, Wei Liu, Tatsumi Arima, Yuji Arita, Isamu Sato, Haruaki Matsuura, Yoshihiro Sekio, Hiroshi Sagara, Masatoshi Kawashima

ASME J of Nuclear Rad Sci. October 2023, 9(4): 041302.

Paper No: NERS-22-1136

DOI: https://doi.org/10.1115/1.4056854

Published Online: March 14, 2023

View Article titled, Development of a Passive Reactor Shutdown Device to Prevent Core Disruptive Accidents in Fast Reactors: A Study on Device Specifications

Topics: Coolants, Flow (Dynamics), Fuels, Sodium, Temperature, Accidents, Transients (Dynamics), Pins (Engineering)

Journal Articles

Impact of Thermal-Hydraulic Feedback and Differential Thermal Expansion on European Sodium-Cooled Fast Reactor Core Power Distribution

Ben Lindley, Francisco Álvarez Velarde, Una Baker, Janos Bodi, Paul Cosgrove, Alan Charles, Carlo Fiorina, Emil Fridman, Jiri Krepel, Jean Lavarenne, Konstantin Mikityuk, Evgeny Nikitin, Alexander Ponomarev, Stefan Radman, Eugene Shwageraus, Brendan Tollit

ASME J of Nuclear Rad Sci. July 2023, 9(3): 031301.

Paper No: NERS-22-1119

DOI: https://doi.org/10.1115/1.4056930

Published Online: March 13, 2023

View Article titled, Impact of Thermal-Hydraulic Feedback and Differential Thermal Expansion on European Sodium-Cooled Fast Reactor Core Power Distribution

Topics: Coolants, Density, Feedback, Fuels, Sodium fast reactors, Temperature, Thermal expansion, Manufacturing, Electrical conductance, Heat

Journal Articles

Development of a Passive Reactor Shutdown Device to Prevent Core Disruptive Accidents in Fast Reactors: A Preliminary Study

Koji Morita, Wei Liu, Tatsumi Arima, Yuji Arita, Isamu Sato, Haruaki Matsuura, Yoshihiro Sekio, Hiroshi Sagara, Masatoshi Kawashima

ASME J of Nuclear Rad Sci. April 2023, 9(2): 021301.

Paper No: NERS-21-1161

DOI: https://doi.org/10.1115/1.4056834

Published Online: March 2, 2023

View Article titled, Development of a Passive Reactor Shutdown Device to Prevent Core Disruptive Accidents in Fast Reactors: A Preliminary Study

Topics: Accidents, Coolants, Fuels, Temperature, Fast neutron reactors, Flow (Dynamics), Pins (Engineering), Transients (Dynamics)

Journal Articles

Validation of Feedback Reactivity Evaluation Models for Plant Dynamics Analysis Code During Unprotected Loss of Heat Sink Event in Sodium-Cooled Fast Reactors

Kazuo Yoshimura, Norihiro Doda, Kennichi Igawa, Masaaki Tanaka, Hidemasa Yamano

ASME J of Nuclear Rad Sci. April 2023, 9(2): 021601.

Paper No: NERS-21-1160

DOI: https://doi.org/10.1115/1.4055328

Published Online: December 14, 2022

View Article titled, Validation of Feedback Reactivity Evaluation Models for Plant Dynamics Analysis Code During Unprotected Loss of Heat Sink Event in Sodium-Cooled Fast Reactors

Topics: Feedback, Sodium fast reactors, Temperature, Fuels, Sensitivity analysis, Density, Heat, Heat sinks, Dynamics (Mechanics), Coolants

Journal Articles

Flash-Boiling Characterization During Ingress of Coolant Event for Dust Issue in ITER

Benjamin Blaisot, Emmanuel Porcheron, Olivier Praud, Véronique Roig

ASME J of Nuclear Rad Sci. July 2022, 8(3): 032301.

Paper No: NERS-21-1149

DOI: https://doi.org/10.1115/1.4053828

Published Online: June 1, 2022

View Article titled, Flash-Boiling Characterization During Ingress of Coolant Event for Dust Issue in ITER

Topics: Boiling, Coolants, Drops, Dust, Ejectors, Flow (Dynamics), Pressure, Sprays, Vacuum, Water

Journal Articles

Thermal Fatigue of Pressurized Water Reactor Coolant System Loop Drain Lines Due to Outflow Activities

Yue Zou, Brian Derreberry

ASME J of Nuclear Rad Sci. July 2022, 8(3): 031801.

Paper No: NERS-21-1040

DOI: https://doi.org/10.1115/1.4053013

Published Online: May 26, 2022

View Article titled, Thermal Fatigue of Pressurized Water Reactor Coolant System Loop Drain Lines Due to Outflow Activities

Topics: Chemistry, Coolants, Fatigue, Flow (Dynamics), Outflow, Pipes, Pressurized water reactors, Stress, Temperature, Thermal stratification

Journal Articles

Development and Processing of Thermal-Hydraulic Model for GFR 2400 Fast Reactor Design and NESTLE Coupled Transient Code

Filip Osuský, Branislav Vrban, Štefan Čerba, Jakub Lüley, Vladimír Nečas

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041401.

Paper No: NERS-21-1012

DOI: https://doi.org/10.1115/1.4051515

Published Online: March 22, 2022

View Article titled, Development and Processing of Thermal-Hydraulic Model for GFR 2400 Fast Reactor Design and NESTLE Coupled Transient Code

Topics: Coolants, Design, Fast neutron reactors, Fuels, Heat, Temperature, Transients (Dynamics), Temperature distribution, Simulation, Cladding systems (Building)

Journal Articles

Testing the Modified Subchannel tempa-sc Code in Comparison With Experiments and Other Computer Codes

A. M. Baisov, A. N. Churkin, V. I. Deev, V. S. Kharitonov

ASME J of Nuclear Rad Sci. July 2022, 8(3): 031101.

Paper No: NERS-21-1101

DOI: https://doi.org/10.1115/1.4053049

Published Online: March 22, 2022

View Article titled, Testing the Modified Subchannel tempa-sc Code in Comparison With Experiments and Other Computer Codes

Topics: Coolants, Fluids, Fuel rods, Heat transfer, Momentum, Temperature, Water, Wall temperature, Flow (Dynamics), Testing

Journal Articles

Impact of Approximations in Operating History Data on Spent Fuel Properties With Serpent 2

Silja Häkkinen

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041901.

Paper No: NERS-21-1015

DOI: https://doi.org/10.1115/1.4051444

Published Online: March 15, 2022

View Article titled, Impact of Approximations in Operating History Data on Spent Fuel Properties With Serpent 2

Topics: Approximation, Coolants, Cycles, Irradiation (Radiation exposure), Manufacturing, Nuclides, Spent nuclear fuels, Heat, Boron, Photons

Journal Articles

Concept of a Small Modular SCWR With Horizontal Fuel Assemblies

Thomas Schulenberg, Ivan Otic

ASME J of Nuclear Rad Sci. July 2022, 8(3): 031104.

Paper No: NERS-21-1074

DOI: https://doi.org/10.1115/1.4052191

Published Online: March 15, 2022

View Article titled, Concept of a Small Modular SCWR With Horizontal Fuel Assemblies

Topics: Coolants, Design, Fuels, Heat, Supercritical water reactors, Water, Temperature, Light water reactors, Cladding systems (Building), Natural convection

Journal Articles

On Some Differences of the Thermoelectrical Effects in BN-600 and Superphenix

Serguei Podobed

ASME J of Nuclear Rad Sci. April 2022, 8(2): 024505.

Paper No: NERS-20-1167

DOI: https://doi.org/10.1115/1.4052591

Published Online: January 11, 2022

View Article titled, On Some Differences of the Thermoelectrical Effects in BN-600 and Superphenix

Topics: Coolants, Currents, Design, Excitation, Liquid metals, Magnetic fields, Pressure, Sodium, Heat exchangers, Pumps

Journal Articles

Optimization of the Nodalization of Nuclear System Thermal-Hydraulic Code Applied on Primary Loop Benchmark

Hong Xu, Aurelian Florin Badea, Xu Cheng

ASME J of Nuclear Rad Sci. April 2022, 8(2): 021404.

Paper No: NERS-21-1036

DOI: https://doi.org/10.1115/1.4050770

Published Online: October 19, 2021

View Article titled, Optimization of the Nodalization of Nuclear System Thermal-Hydraulic Code Applied on Primary Loop Benchmark

Topics: Accidents, Coolants, Steady state, Transients (Dynamics), Errors, Flow (Dynamics), Thermal hydraulics, Leakage, Simulation

Journal Articles

Evaluation of Sodium Boiling Models Using KNS-37 Loss of Flow Experiments

Sara Perez-Martin, Marine Anderhuber, Laurent Laborde, Nathalie Girault, Calogera Lombardo, Luca Ammirabile, Konstantin Mikityuk, Stephane Mimouni, Christophe Péniguel, Werner Pfrang

ASME J of Nuclear Rad Sci. January 2022, 8(1): 011310.

Paper No: NERS-21-1049

DOI: https://doi.org/10.1115/1.4050769

Published Online: August 3, 2021

View Article titled, Evaluation of Sodium Boiling Models Using KNS-37 Loss of Flow Experiments

Topics: Boiling, Coolants, Flow (Dynamics), Sodium, Sodium fast reactors, Temperature, Temperature profiles, Transients (Dynamics)

Journal Articles

Small Modular Reactors: Learning From the Past

Esam M. A. Hussein

ASME J of Nuclear Rad Sci. July 2021, 7(3): 034508.

Paper No: NERS-20-1152

DOI: https://doi.org/10.1115/1.4050478

Published Online: May 26, 2021

View Article titled, Small Modular Reactors: Learning From the Past

Topics: Coolants, Fast neutron reactors, Fuels, Gas cooled reactors, High temperature, Liquid metals, Molten salt reactors, Pressurized water reactors, Small modular reactors, Licensing

Journal Articles

CANDU-Like Features in Emerging Small Modular Reactor Designs

Esam M. A. Hussein

ASME J of Nuclear Rad Sci. July 2021, 7(3): 034507.

Paper No: NERS-20-1151

DOI: https://doi.org/10.1115/1.4050477

Published Online: May 26, 2021

View Article titled, CANDU-Like Features in Emerging Small Modular Reactor Designs

Topics: Fuels, Small modular reactors, Water, Coolants, Drainage, Heat sinks, Design, Construction, Lithium, Heat

Journal citation

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Optimized Moderator Design and Analysis of a Pin-Type Supercritical Carbon Dioxide Reactor Based on Reactor Monte Carlo Code

Fabrication Aspects and Performance Characterization of α-Al 2 O 3 /ALPO 4 Based Sandwich Configuration Flow Channel Inserts and Coatings for High Temperature Liquid Metal Applications

System Thermal-hydraulics Model for Fluoride Salt-Cooled Reactor Based On Small Advanced High Temperature Reactor (smAHTR) Design Concept

Neutronic and Thermal-Fluidic Analyses for an Additive Manufactured Reactor With SiC Matrix and TRISO Particle Fuel

Eddy Current Flow Meter Measurements in Liquid Sodium at High Temperatures

Development of a Passive Reactor Shutdown Device to Prevent Core Disruptive Accidents in Fast Reactors: A Study on Device Specifications

Impact of Thermal-Hydraulic Feedback and Differential Thermal Expansion on European Sodium-Cooled Fast Reactor Core Power Distribution

Development of a Passive Reactor Shutdown Device to Prevent Core Disruptive Accidents in Fast Reactors: A Preliminary Study

Validation of Feedback Reactivity Evaluation Models for Plant Dynamics Analysis Code During Unprotected Loss of Heat Sink Event in Sodium-Cooled Fast Reactors

Flash-Boiling Characterization During Ingress of Coolant Event for Dust Issue in ITER

Thermal Fatigue of Pressurized Water Reactor Coolant System Loop Drain Lines Due to Outflow Activities

Development and Processing of Thermal-Hydraulic Model for GFR 2400 Fast Reactor Design and NESTLE Coupled Transient Code

Testing the Modified Subchannel tempa-sc Code in Comparison With Experiments and Other Computer Codes

Impact of Approximations in Operating History Data on Spent Fuel Properties With Serpent 2

Concept of a Small Modular SCWR With Horizontal Fuel Assemblies

On Some Differences of the Thermoelectrical Effects in BN-600 and Superphenix

Optimization of the Nodalization of Nuclear System Thermal-Hydraulic Code Applied on Primary Loop Benchmark

Evaluation of Sodium Boiling Models Using KNS-37 Loss of Flow Experiments

Small Modular Reactors: Learning From the Past

CANDU-Like Features in Emerging Small Modular Reactor Designs

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