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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

Comparative Maps of Safety Features for Fission and Fusion Reactors

Tu Nguyen, Eann A Patterson, Richard J. Taylor, Yung-Shin Tseng, Chris Waldon

ASME J of Nuclear Rad Sci.

Paper No: NERS-22-1045

DOI: https://doi.org/10.1115/1.4062643

Published Online: May 27, 2023

View Article titled, Comparative Maps of Safety Features for Fission and Fusion Reactors

Topics: Fusion reactors, Nuclear fission, Safety, Nuclear reactors, Hazards, Regulations

Journal Articles

Neutronic Evaluation of Light Water Reactor-Based Small Modular Reactor With Transuranic Fuel and Comparison With Different Core Configuration of UO 2 and Mixed Oxide Fuel

Sushanta Biswas, Md. Hossain Sahadath, Nazia Rahim Ananya

ASME J of Nuclear Rad Sci. January 2023, 9(1): 011501.

Paper No: NERS-21-1139

DOI: https://doi.org/10.1115/1.4054208

Published Online: June 7, 2022

View Article titled, Neutronic Evaluation of Light Water Reactor-Based Small Modular Reactor With Transuranic Fuel and Comparison With Different Core Configuration of UO 2 and Mixed Oxide Fuel

Topics: Atoms, Fuels, Nuclear fission

Journal Articles

Fuel Performance Modeling at High Burn-Up by FEMAXI-6 Code

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

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041601.

Paper No: NERS-21-1030

DOI: https://doi.org/10.1115/1.4052519

Published Online: March 22, 2022

View Article titled, Fuel Performance Modeling at High Burn-Up by FEMAXI-6 Code

Topics: Fuels, Grain size, Irradiation (Radiation exposure), Modeling, Nuclear fission, Rods, Pressure, Cladding systems (Building), Stress

Journal Articles

Cross Section and Fission Yields Induced Uncertainty in the Water–Water Energetic Reactor-440 Burnup Calculation

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

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041502.

Paper No: NERS-21-1010

DOI: https://doi.org/10.1115/1.4051613

Published Online: March 16, 2022

View Article titled, Cross Section and Fission Yields Induced Uncertainty in the Water–Water Energetic Reactor-440 Burnup Calculation

Topics: Fuels, Nuclear fission, Uncertainty, Water, Nuclides, Neutrons

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

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

The Effect of Serpent 2 Calculation Parameters on Evaluated Spent Nuclear Fuel Source Term

Riku Tuominen, Ville Valtavirta

ASME J of Nuclear Rad Sci. October 2022, 8(4): 044503.

Paper No: NERS-21-1019

DOI: https://doi.org/10.1115/1.4051445

Published Online: March 15, 2022

View Article titled, The Effect of Serpent 2 Calculation Parameters on Evaluated Spent Nuclear Fuel Source Term

Topics: Boiling water reactors, Emissions, Heat, Manufacturing, Nuclear fission, Photons, Probability, Resonance, Spent nuclear fuels, Nuclides

Journal Articles

Neutronic and Thermal-Mechanical Coupling Schemes for Heat Pipe-Cooled Reactor Designs

Yugao Ma, Minyun Liu, Biheng Xie, Wenbin Han, Xiaoming Chai, Shanfang Huang, Hongxing Yu

ASME J of Nuclear Rad Sci. April 2022, 8(2): 021303.

Paper No: NERS-20-1207

DOI: https://doi.org/10.1115/1.4051612

Published Online: October 19, 2021

View Article titled, Neutronic and Thermal-Mechanical Coupling Schemes for Heat Pipe-Cooled Reactor Designs

Topics: Design, Displacement, Fuels, Geometry, Heat, Heat pipes, Molten salt reactors, Nuclear fission, Pipes, Temperature

Journal Articles

Metallic Fast Reactor Separate Effect Studies for Fuel Safety

Fidelma G. Di Lemma, Colby B. Jensen, Joshua J. Kane, Wei-Ying Chen, Xiang Liu, Luca Capriotti, Cynthia A. Adkins, Boopathy Kombaiah, Alexander J. Winston, Lingfeng He, Daniel Wachs

ASME J of Nuclear Rad Sci. October 2021, 7(4): 041602.

Paper No: NERS-20-1097

DOI: https://doi.org/10.1115/1.4049721

Published Online: April 22, 2021

View Article titled, Metallic Fast Reactor Separate Effect Studies for Fuel Safety

Topics: Bubbles, Fuels, Instrumentation, Irradiation (Radiation exposure), Nuclear fission, Safety, Transients (Dynamics), Sodium fast reactors, Testing, Fast neutron reactors

Journal Articles

Methodology of 63 Cu(n,2n) 62 Cu Reaction Rate Measurement

Martin Schulc, Michal Košťál, Jan Šimon, Evžen Novák

ASME J of Nuclear Rad Sci. April 2021, 7(2): 024503.

Paper No: NERS-20-1039

DOI: https://doi.org/10.1115/1.4048236

Published Online: November 16, 2020

View Article titled, Methodology of 63 Cu(n,2n) 62 Cu Reaction Rate Measurement

Topics: Half-life, Irradiation (Radiation exposure), Isotopes, Neutrons, Nuclear fission, Uncertainty, Gamma ray spectrometry, Emissions

Journal Articles

Toward a Mechanistic Model of Stress Corrosion Cracking in PHWR Fuel Sheaths

Andrew Oussoren, Paul Chan, Diane Wowk, Andrew Prudil

ASME J of Nuclear Rad Sci. January 2021, 7(1): 011604.

Paper No: NERS-19-1018

DOI: https://doi.org/10.1115/1.4048120

Published Online: October 5, 2020

View Article titled, Toward a Mechanistic Model of Stress Corrosion Cracking in PHWR Fuel Sheaths

Topics: Failure, Fracture (Materials), Fuels, Nuclear fission, Diffusion (Physics)

Journal Articles

The Other Metallic Phase in Spent Nuclear Fuel: A Complete Thermodynamic Evaluation of the U–Pd–Rh–Ru System

Lian-Cheng Wang, Matthew H. Kaye

ASME J of Nuclear Rad Sci. January 2021, 7(1): 011601.

Paper No: NERS-19-1094

DOI: https://doi.org/10.1115/1.4046784

Published Online: October 1, 2020

View Article titled, The Other Metallic Phase in Spent Nuclear Fuel: A Complete Thermodynamic Evaluation of the U–Pd–Rh–Ru System

Topics: Nuclear fuels, Phase diagrams, Temperature, Uranium, Modeling, Military systems, Nuclear fission, Spent nuclear fuels

Journal Articles

A Review of Circulating Voids and Gaseous Fission Product Behavior in Molten Salt Reactors

Terry Price, Zack Taylor, Ondrej Chvala

ASME J of Nuclear Rad Sci. October 2020, 6(4): 041302.

Paper No: NERS-19-1107

DOI: https://doi.org/10.1115/1.4046258

Published Online: June 3, 2020

View Article titled, A Review of Circulating Voids and Gaseous Fission Product Behavior in Molten Salt Reactors

Topics: Bubbles, Fuels, Molten salt reactors, Porosity, Pressure, Nuclear fission, Flow (Dynamics), Pumps

Journal Articles

Statistical Burnup Distribution of Moving Pebbles in the High-Temperature Reactor HTR-PM

Fanny Vitullo, Jiri Krepel, Jarmo Kalilainen, Horst-Michael Prasser, Andreas Pautz

ASME J of Nuclear Rad Sci. April 2020, 6(2): 021108.

Paper No: NERS-19-1059

DOI: https://doi.org/10.1115/1.4044910

Published Online: February 5, 2020

View Article titled, Statistical Burnup Distribution of Moving Pebbles in the High-Temperature Reactor HTR-PM

Topics: Cross section (Physics), Fuels, Simulation, Temperature, Graphite, Neutron flux, Nuclear fission, Neutrons, High temperature, Flow (Dynamics)

Journal Articles

Evaluation of Radiation Effects on Residents Living Around the NSRR Under External Hazards

Yuiko Motome, Yoshiya Akiyama, Hiroyuki Murao

ASME J of Nuclear Rad Sci. April 2020, 6(2): 021115.

Paper No: NERS-19-1073

DOI: https://doi.org/10.1115/1.4045075

Published Online: February 3, 2020

View Article titled, Evaluation of Radiation Effects on Residents Living Around the NSRR Under External Hazards

Topics: Earthquakes, Fuel rods, Fuel storage, Fuels, Radiation (Physics), Safety, Spent fuel storage, Tsunamis, Nuclear fission, Shapes

Journal Articles

Neutronic Analyses of the FREYA Experiments in Support of the ALFRED Lead-Cooled Fast Reactor Core Design and Licensing

Massimo Sarotto, Gabriele Firpo, Anatoly Kochetkov, Antonin Krása, Emil Fridman, Jerzy Cetnar, Grazyna Domanska

ASME J of Nuclear Rad Sci. January 2020, 6(1): 011402.

Paper No: NERS-18-1132

DOI: https://doi.org/10.1115/1.4044000

Published Online: November 29, 2019

View Article titled, Neutronic Analyses of the FREYA Experiments in Support of the ALFRED Lead-Cooled Fast Reactor Core Design and Licensing

Topics: Design, Fast neutron reactors, Nuclear fission, Nuclides, Uncertainty, Uncertainty analysis, Licensing, Neutrons, Rods, Simulation

Journal Articles

Multiscale Mesoscale Modeling of Porosity Evolution in Oxide Fuels

M. J. Welland, K. D. Colins, N. Ofori-Opoku, A. A. Prudil, E. S. Thomas

ASME J of Nuclear Rad Sci. January 2020, 6(1): 011105.

Paper No: NERS-19-1019

DOI: https://doi.org/10.1115/1.4044405

Published Online: November 8, 2019

View Article titled, Multiscale Mesoscale Modeling of Porosity Evolution in Oxide Fuels

Topics: Bubbles, Fuels, Grain boundaries, Irradiation (Radiation exposure), Modeling, Nuclear fission, Percolation theory, Porosity, Crystals, Tunnels

Journal Articles

Enhanced Accident Tolerance of Thoria-Based Nuclear Fuels

B. Szpunar, J. A. Szpunar

ASME J of Nuclear Rad Sci. January 2020, 6(1): 011108.

Paper No: NERS-19-1046

DOI: https://doi.org/10.1115/1.4044751

Published Online: November 8, 2019

View Article titled, Enhanced Accident Tolerance of Thoria-Based Nuclear Fuels

Topics: Accidents, Atoms, Diffusion (Physics), Fuel rods, Fuels, Helium, Nuclear fuels, Oxygen, Temperature, Temperature gradient

Journal Articles

Verification and Validation of SuperMC3.2 Using VENUS-3 Benchmark Experiments

I. K. Baidoo, J. Zou, B. Li, J. Song, B. Wu, Q. Yang, Z. Zhao

ASME J of Nuclear Rad Sci. October 2019, 5(4): 041402.

Paper No: NERS-18-1123

DOI: https://doi.org/10.1115/1.4043100

Published Online: July 19, 2019

View Article titled, Verification and Validation of SuperMC3.2 Using VENUS-3 Benchmark Experiments

Topics: Atoms, Displacement, Dosimetry, Fluence (Radiation measurement), Fuels, Geometry, Iron, Manufacturing, Modeling, Neutron flux

Journal citation

NARROW

Manuscript

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

Comparative Maps of Safety Features for Fission and Fusion Reactors

Neutronic Evaluation of Light Water Reactor-Based Small Modular Reactor With Transuranic Fuel and Comparison With Different Core Configuration of UO 2 and Mixed Oxide Fuel

Fuel Performance Modeling at High Burn-Up by FEMAXI-6 Code

Cross Section and Fission Yields Induced Uncertainty in the Water–Water Energetic Reactor-440 Burnup Calculation

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

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

The Effect of Serpent 2 Calculation Parameters on Evaluated Spent Nuclear Fuel Source Term

Neutronic and Thermal-Mechanical Coupling Schemes for Heat Pipe-Cooled Reactor Designs

Metallic Fast Reactor Separate Effect Studies for Fuel Safety

Methodology of 63 Cu(n,2n) 62 Cu Reaction Rate Measurement

Toward a Mechanistic Model of Stress Corrosion Cracking in PHWR Fuel Sheaths

The Other Metallic Phase in Spent Nuclear Fuel: A Complete Thermodynamic Evaluation of the U–Pd–Rh–Ru System

A Review of Circulating Voids and Gaseous Fission Product Behavior in Molten Salt Reactors

Statistical Burnup Distribution of Moving Pebbles in the High-Temperature Reactor HTR-PM

Evaluation of Radiation Effects on Residents Living Around the NSRR Under External Hazards

Neutronic Analyses of the FREYA Experiments in Support of the ALFRED Lead-Cooled Fast Reactor Core Design and Licensing

Multiscale Mesoscale Modeling of Porosity Evolution in Oxide Fuels

Enhanced Accident Tolerance of Thoria-Based Nuclear Fuels

Verification and Validation of SuperMC3.2 Using VENUS-3 Benchmark Experiments

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