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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

Small Modular Reactors: Opportunities and Challenges as Emerging Nuclear Technologies for Power Production

Lekhnath Ghimire, Edward Waller

ASME J of Nuclear Rad Sci.

Paper No: NERS-23-1006

DOI: https://doi.org/10.1115/1.4062644

Published Online: May 27, 2023

View Article titled, Small Modular Reactors: Opportunities and Challenges as Emerging Nuclear Technologies for Power Production

Topics: Energy generation, Small modular reactors, Safety, Arctic region, Design, Accidents, Climate, Construction, Electric power generation, Fuels

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

Introducing Passive Nuclear Safety in Water-Cooled Reactors: Numerical Simulation and Validation of Natural Convection Heat Transfer and Transport in Packed Beds of Heated Microspheres

Olugbenga O. Noah, Johan F. Slabber, Josua P. Meyer

ASME J of Nuclear Rad Sci. July 2023, 9(3): 031403.

Paper No: NERS-22-1020

DOI: https://doi.org/10.1115/1.4056239

Published Online: February 8, 2023

View Article titled, Introducing Passive Nuclear Safety in Water-Cooled Reactors: Numerical Simulation and Validation of Natural Convection Heat Transfer and Transport in Packed Beds of Heated Microspheres

Topics: Fluids, Heat transfer, Particulate matter, Simulation, Temperature, Flow (Dynamics), Design, Fuels, Safety, Natural convection

Journal Articles

Novel Eddy Current Array Sensor Delivery Tool for Automated Nondestructive Examination of Spent Fuel Pool Liner

Michael Smith, Emily Abbate, Joey Phillips, Byungsik Yoon

ASME J of Nuclear Rad Sci. April 2023, 9(2): 021701.

Paper No: NERS-21-1165

DOI: https://doi.org/10.1115/1.4055065

Published Online: September 2, 2022

View Article titled, Novel Eddy Current Array Sensor Delivery Tool for Automated Nondestructive Examination of Spent Fuel Pool Liner

Topics: Design, Eddies (Fluid dynamics), Eddy currents (Electricity), Inspection, Motors, Nondestructive evaluation, Sensors, Spent nuclear fuels, Rails, Railroad passenger cars

Journal Articles

Technical Evaluation Method of Mechanical Equipment Item Substitution in the Procurement and Construction Process of Nuclear Power Plant

Jindong Zhang, Chongzhi Wu

ASME J of Nuclear Rad Sci. January 2023, 9(1): 011801.

Paper No: NERS-22-1014

DOI: https://doi.org/10.1115/1.4054878

Published Online: July 19, 2022

View Article titled, Technical Evaluation Method of Mechanical Equipment Item Substitution in the Procurement and Construction Process of Nuclear Power Plant

Topics: Construction, Design, Evaluation methods, Nuclear power stations, Failure mechanisms, Nuclear power

Journal Articles

Design Study of Small Modular Reactor Class Super Fast Reactor Core for In-Vessel Retention

Ryotaro Sasaki, Akifumi Yamaji, Kyota Uchimura

ASME J of Nuclear Rad Sci. April 2023, 9(2): 021501.

Paper No: NERS-21-1135

DOI: https://doi.org/10.1115/1.4053827

Published Online: June 15, 2022

View Article titled, Design Study of Small Modular Reactor Class Super Fast Reactor Core for In-Vessel Retention

Topics: Design, Vessels, Cycles, Fuels, Temperature, Fast neutron reactors, Small modular reactors

Journal Articles

ARKADIA—For the Innovation of Advanced Nuclear Reactor Design

Hiroyuki Ohshima, Tai Asayama, Tomohiro Furukawa, Masaaki Tanaka, Akihiro Uchibori, Takashi Takata, Akiyuki Seki, Yasuhiro Enuma

ASME J of Nuclear Rad Sci. April 2023, 9(2): 025001.

Paper No: NERS-21-1157

DOI: https://doi.org/10.1115/1.4054726

Published Online: June 15, 2022

View Article titled, ARKADIA—For the Innovation of Advanced Nuclear Reactor Design

Topics: Design, Optimization, Safety, Thermal hydraulics, Nuclear design, Nuclear reactors, Computer simulation

Journal Articles

Necessity of Coordinating Nondestructive Examination Requirement for Nuclear Power Plant

Chongzhi Wu, Haoming Sun, Zhenguo Peng, Benzheng Wang

ASME J of Nuclear Rad Sci. January 2023, 9(1): 014701.

Paper No: NERS-21-1177

DOI: https://doi.org/10.1115/1.4054169

Published Online: June 7, 2022

View Article titled, Necessity of Coordinating Nondestructive Examination Requirement for Nuclear Power Plant

Topics: Design, Nondestructive evaluation, Nuclear power stations, Safety

Journal Articles

Thermal-Hydraulic Safety Assessment of Full-Scale ESBWR Nuclear Reactor Design

S. P. Saraswat, D. Ray, G. Mishra, D. Yadav, V. S. Bhadouria, P. Munshi, C. Allison

ASME J of Nuclear Rad Sci. July 2022, 8(3): 031403.

Paper No: NERS-21-1073

DOI: https://doi.org/10.1115/1.4052014

Published Online: May 26, 2022

View Article titled, Thermal-Hydraulic Safety Assessment of Full-Scale ESBWR Nuclear Reactor Design

Topics: Accidents, Boiling water reactors, Containment, Design, Pressure, Safety, Steam, Water, Heat, Economics

Journal Articles

Uncertainty Quantification in Support of Severe Accident Analysis Code User Confidence Using MELCOR-DAKOTA

Emmanuel K. Boafo, Emmanuel Numapau Gyamfi

ASME J of Nuclear Rad Sci. July 2022, 8(3): 031703.

Paper No: NERS-21-1094

DOI: https://doi.org/10.1115/1.4053050

Published Online: May 26, 2022

View Article titled, Uncertainty Quantification in Support of Severe Accident Analysis Code User Confidence Using MELCOR-DAKOTA

Topics: Accidents, Hydrogen, Sensitivity analysis, Uncertainty, Uncertainty quantification, Uncertainty analysis, Design, Optimization

Journal Articles

Suggestion on Improving Small Branch BOSS Joints Welding Quality

Chongzhi Wu, Ruifeng Zhu, Pengteng Yu, Haoming Sun, Qingli Zhou

ASME J of Nuclear Rad Sci. July 2022, 8(3): 035501.

Paper No: NERS-21-1095

DOI: https://doi.org/10.1115/1.4053690

Published Online: May 26, 2022

View Article titled, Suggestion on Improving Small Branch BOSS Joints Welding Quality

Topics: Design, Nondestructive evaluation, Welding

Journal Articles

Dr. Romney B. Duffey on His 80th Birthday

Igor Pioro, Richard R. Schultz, Dr. Hideki Kamide, Dr. Sama Bilbao y Leon, Dr. Nils Diaz, Alexander V. Bychkov, Dr. Leon Cizelj, Francesco D'Auria, Dr. Yanping Huang, Dr. Yasuo Koizumi, Tomio Okawa, Dr. Robert Walker, Dr. Robert Youngblood, Enrico Zio

ASME J of Nuclear Rad Sci. July 2022, 8(3): 030201.

Paper No: NERS-22-1030

DOI: https://doi.org/10.1115/1.4054329

Published Online: May 3, 2022

View Article titled, Dr. Romney B. Duffey on His 80th Birthday

Topics: Accidents, Design, Electricity (Physics), Governments, Innovation, Japan Society of Mechanical Engineers, Nuclear engineering, Nuclear power, Risk, Safety

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

Utilization of Nuklearna Elektrarna Krško Full Scope Simulator for Plant Operation Optimization, Nuclear Education and Engineering in 20 Years

Matjaž Žvar, Tomaž Žagar

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041701.

Paper No: NERS-21-1022

DOI: https://doi.org/10.1115/1.4052013

Published Online: March 16, 2022

View Article titled, Utilization of Nuklearna Elektrarna Krško Full Scope Simulator for Plant Operation Optimization, Nuclear Education and Engineering in 20 Years

Topics: Accidents, Design, Drills (Tools), Education, Emergency preparedness, Nuclear power stations, Safety, Optimization, Computers, Spent nuclear fuels

Journal Articles

Comparison of the ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 Libraries for the Nuclear Design Calculations of the Nuclear Power Plant Krško With the CORD-2 System

Marjan Kromar, Bojan Kurinčič

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041501.

Paper No: NERS-21-1014

DOI: https://doi.org/10.1115/1.4050991

Published Online: March 15, 2022

View Article titled, Comparison of the ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 Libraries for the Nuclear Design Calculations of the Nuclear Power Plant Krško With the CORD-2 System

Topics: Boron, Cycles, Fuels, Manufacturing, Neutrons, Nuclear design, Nuclear power stations, Nuclear fission, Cross section (Physics), Rods

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

A Short Inroad to Optimized Compactification of Composite Neutronic Shields

Nassar H. S. Haidar

ASME J of Nuclear Rad Sci. April 2022, 8(2): 024506.

Paper No: NERS-21-1173

DOI: https://doi.org/10.1115/1.4053334

Published Online: February 7, 2022

View Article titled, A Short Inroad to Optimized Compactification of Composite Neutronic Shields

Topics: Composite materials, Design, Neutrons, Optimization, Neutron beams, Diffusion (Physics), Weight (Mass)

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 citation

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Manuscript

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

Small Modular Reactors: Opportunities and Challenges as Emerging Nuclear Technologies for Power Production

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

Introducing Passive Nuclear Safety in Water-Cooled Reactors: Numerical Simulation and Validation of Natural Convection Heat Transfer and Transport in Packed Beds of Heated Microspheres

Novel Eddy Current Array Sensor Delivery Tool for Automated Nondestructive Examination of Spent Fuel Pool Liner

Technical Evaluation Method of Mechanical Equipment Item Substitution in the Procurement and Construction Process of Nuclear Power Plant

Design Study of Small Modular Reactor Class Super Fast Reactor Core for In-Vessel Retention

ARKADIA—For the Innovation of Advanced Nuclear Reactor Design

Necessity of Coordinating Nondestructive Examination Requirement for Nuclear Power Plant

Thermal-Hydraulic Safety Assessment of Full-Scale ESBWR Nuclear Reactor Design

Uncertainty Quantification in Support of Severe Accident Analysis Code User Confidence Using MELCOR-DAKOTA

Suggestion on Improving Small Branch BOSS Joints Welding Quality

Dr. Romney B. Duffey on His 80th Birthday

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

Utilization of Nuklearna Elektrarna Krško Full Scope Simulator for Plant Operation Optimization, Nuclear Education and Engineering in 20 Years

Comparison of the ENDF/B-VII.0, ENDF/B-VII.1, ENDF/B-VIII.0, and JEFF-3.3 Libraries for the Nuclear Design Calculations of the Nuclear Power Plant Krško With the CORD-2 System

Concept of a Small Modular SCWR With Horizontal Fuel Assemblies

A Short Inroad to Optimized Compactification of Composite Neutronic Shields

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

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