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Journal Articles

Calculation and Uncertainty Analysis of Core Parameters of Advanced Lead-Cooled Modular Nuclear Reactor Using New Nuclear Data Libraries

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Thanh Mai Vu, Le Quang Linh Tran, Thi Hong Bui, Nhu Viet Ha Pham
Journal: Journal of Nuclear Engineering and Radiation Science
Publisher: ASME
Article Type: Research-Article
ASME J of Nuclear Rad Sci. July 2024, 10(3): 031501.
Paper No: NERS-23-1039
https://doi.org/10.1115/1.4064780
Published Online: April 17, 2024
View Article titled, Calculation and Uncertainty Analysis of Core Parameters of Advanced Lead-Cooled Modular Nuclear Reactor Using New Nuclear Data Libraries
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Topics: Eigenvalues, Fast neutron reactors, Neutrons, Nuclear fission, Nuclear reactors, Uncertainty, Uncertainty analysis, Cross section (Physics), Electromagnetic scattering, Radiation scattering
Journal Articles

Improvement of Few-Group Homogenized Cross-Sections for RSG-GAS In-Core Fuel Management Code Batan-FUEL

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Surian Pinem, Donny Hartanto, Peng Hong Liem, Wahid Luthfi
Journal: Journal of Nuclear Engineering and Radiation Science
Publisher: ASME
Article Type: Research-Article
ASME J of Nuclear Rad Sci. July 2023, 9(3): 031502.
Paper No: NERS-22-1040
https://doi.org/10.1115/1.4056603
Published Online: February 8, 2023
View Article titled, Improvement of Few-Group Homogenized Cross-Sections for RSG-GAS In-Core Fuel Management Code Batan-FUEL
Open the PDF for Improvement of Few-Group Homogenized Cross-Sections for RSG-GAS In-Core Fuel Management Code Batan-FUEL in another window
Topics: Fuel management, Fuels, Cross section (Physics), Cycles
Journal Articles

Calculation of Gamma and Neutron Shielding Parameters for Reinforced Polymer Composites

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A. M. Osman
Journal: Journal of Nuclear Engineering and Radiation Science
Publisher: ASME
Article Type: Research-Article
ASME J of Nuclear Rad Sci. January 2023, 9(1): 012002.
Paper No: NERS-21-1171
https://doi.org/10.1115/1.4054547
Published Online: June 7, 2022
View Article titled, Calculation of Gamma and Neutron Shielding Parameters for Reinforced Polymer Composites
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Topics: Composite materials, Density, Neutrons, Polymer composites, Relaxation (Physics), Photons, Cross section (Physics), Aluminum, Boron
Journal Articles

On the Self-Shielding in the Unresolved Resonance Range

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Daniel Lopez Aldama, Andrej Trkov, Dermott E. Cullen
Journal: Journal of Nuclear Engineering and Radiation Science
Publisher: ASME
Article Type: Research-Article
ASME J of Nuclear Rad Sci. October 2022, 8(4): 041503.
Paper No: NERS-21-1026
https://doi.org/10.1115/1.4052377
Published Online: March 16, 2022
View Article titled, On the Self-Shielding in the Unresolved Resonance Range
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Topics: Cross section (Physics), Resonance, Spectra (Spectroscopy), Probability, Leakage
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

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Marjan Kromar, Bojan Kurinčič
Journal: Journal of Nuclear Engineering and Radiation Science
Publisher: ASME
Article Type: Research-Article
ASME J of Nuclear Rad Sci. October 2022, 8(4): 041501.
Paper No: NERS-21-1014
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
Open the PDF for 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 in another window
Topics: Boron, Cycles, Fuels, Manufacturing, Neutrons, Nuclear design, Nuclear power stations, Nuclear fission, Cross section (Physics), Rods
Journal Articles

Using the Few-Group Approximation for Calculating Some Neutron-Physical Characteristics of VVER-1000 Core by Means of the Monte Carlo Universal Code

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A. S. Bikeev, Yu. S. Daichenkova, M. A. Kalugin, D. A. Shkarovsky, V. V. Shkityr
Journal: Journal of Nuclear Engineering and Radiation Science
Publisher: ASME
Article Type: Technical Briefs
ASME J of Nuclear Rad Sci. April 2022, 8(2): 024507.
Paper No: NERS-20-1005
https://doi.org/10.1115/1.4052340
Published Online: October 19, 2021
View Article titled, Using the Few-Group Approximation for Calculating Some Neutron-Physical Characteristics of VVER-1000 Core by Means of the Monte Carlo Universal Code
Open the PDF for Using the Few-Group Approximation for Calculating Some Neutron-Physical Characteristics of VVER-1000 Core by Means of the Monte Carlo Universal Code in another window
Topics: Approximation, Cross section (Physics), Electromagnetic scattering, Errors, Neutrons, Radiation scattering, Scattering (Physics), Fuels, Anisotropy
Journal Articles

Pressurized Water Reactor Mixed Oxide/UO 2 Transient Benchmark Calculations Using Monte Carlo Serpent 2 Code and Open Nodal Core Simulator ADPRES

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Muhammad Imron, Donny Hartanto
Journal: Journal of Nuclear Engineering and Radiation Science
Publisher: ASME
Article Type: Research-Article
ASME J of Nuclear Rad Sci. July 2021, 7(3): 031603.
Paper No: NERS-20-1064
https://doi.org/10.1115/1.4048764
Published Online: March 16, 2021
View Article titled, Pressurized Water Reactor Mixed Oxide/UO 2 Transient Benchmark Calculations Using Monte Carlo Serpent 2 Code and Open Nodal Core Simulator ADPRES
Open the PDF for Pressurized Water Reactor Mixed Oxide/UO 2 Transient Benchmark Calculations Using Monte Carlo Serpent 2 Code and Open Nodal Core Simulator ADPRES in another window
Topics: Boron, Coordinate measuring machines, Cross section (Physics), Fuels, Manufacturing, Pressurized water reactors, Transients (Dynamics), Temperature, Diffusion (Physics), Geometry
Journal Articles

The Impact of Fueling Operations on Full Core Uncertainty Analysis in CANDU Reactors

Open Access
M. D. Tucker, D. R. Novog
Journal: Journal of Nuclear Engineering and Radiation Science
Publisher: ASME
Article Type: Research-Article
ASME J of Nuclear Rad Sci. July 2020, 6(3): 031401.
Paper No: NERS-19-1065
https://doi.org/10.1115/1.4045485
Published Online: March 9, 2020
View Article titled, The Impact of Fueling Operations on Full Core Uncertainty Analysis in CANDU Reactors
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Topics: Algorithms, Cross section (Physics), Feedback, Fuels, Simulation, Uncertainty, Uncertainty analysis, Uranium, Physics, Diffusion (Physics)
Journal Articles

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

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Fanny Vitullo, Jiri Krepel, Jarmo Kalilainen, Horst-Michael Prasser, Andreas Pautz
Journal: Journal of Nuclear Engineering and Radiation Science
Publisher: ASME
Article Type: Research-Article
ASME J of Nuclear Rad Sci. April 2020, 6(2): 021108.
Paper No: NERS-19-1059
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
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Topics: Cross section (Physics), Fuels, Simulation, Temperature, Graphite, Neutron flux, Nuclear fission, Neutrons, High temperature, Flow (Dynamics)
Journal Articles

A Study of Superhomogenization Applied to PHWR Lattices

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Thomas A. Ferguson, Eleodor M. Nichita
Journal: Journal of Nuclear Engineering and Radiation Science
Publisher: ASME
Article Type: Research-Article
ASME J of Nuclear Rad Sci. January 2020, 6(1): 011109.
Paper No: NERS-19-1049
https://doi.org/10.1115/1.4044748
Published Online: November 8, 2019
View Article titled, A Study of Superhomogenization Applied to PHWR Lattices
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Topics: Fuels, Heavy water reactors, Light water reactors, Water, Diffusion (Physics), Cross section (Physics)
Journal Articles

Investigation of 127 I(n,2n) 126 I and 23 Na(n,2n) 22 Na Reactions Using 252 Cf Neutron Source

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Martin Schulc, Michal Košťál, Evžen Novák, Jan Šimon, Nicola Burianová
Journal: Journal of Nuclear Engineering and Radiation Science
Publisher: ASME
Article Type: Research-Article
ASME J of Nuclear Rad Sci. July 2019, 5(3): 030918.
Paper No: NERS-18-1012
https://doi.org/10.1115/1.4039774
Published Online: April 16, 2019
View Article titled, Investigation of 127 I(n,2n) 126 I and 23 Na(n,2n) 22 Na Reactions Using 252 Cf Neutron Source
Open the PDF for Investigation of 127 I(n,2n) 126 I and 23 Na(n,2n) 22 Na Reactions Using 252 Cf Neutron Source in another window
Topics: Chemical kinetics, Cross section (Physics), Emissions, Irradiation (Radiation exposure), Neutron sources, Neutrons, Uncertainty, Nuclear fission, Sodium, Germanium
Journal Articles

Measurement of Selected Differential Cross Sections in 235 U Spectrum

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Nicola Burianová, Michal Košt´ál, Martin Schulc, Jan Šimon, Martin Mareček, Jan Uhlíř
Journal: Journal of Nuclear Engineering and Radiation Science
Publisher: ASME
Article Type: Research-Article
ASME J of Nuclear Rad Sci. July 2019, 5(3): 030906.
Paper No: NERS-18-1051
https://doi.org/10.1115/1.4042850
Published Online: April 16, 2019
View Article titled, Measurement of Selected Differential Cross Sections in 235 U Spectrum
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Topics: Chemical kinetics, Computational methods, Cross section (Physics), Gamma ray spectrometry, Manganese dioxide, Neutrons, Nuclear fission, Sensors, Uncertainty, Irradiation (Radiation exposure)
Journal Articles

U233 Data Evaluation for Criticality Study

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Mohammad Alrwashdeh, Wang Kan
Journal: Journal of Nuclear Engineering and Radiation Science
Publisher: ASME
Article Type: Technical Briefs
ASME J of Nuclear Rad Sci. July 2016, 2(3): 034501.
Paper No: NERS-15-1118
https://doi.org/10.1115/1.4032814
Published Online: June 17, 2016
View Article titled, U233 Data Evaluation for Criticality Study
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Topics: Bayes methods, Computers, Cross section (Physics), Databases, Fittings, Thermal energy, Nuclear fission, Computer software, Neutrons
Journal Articles

Cross-Section Influence on Monte Carlo-Based Burn-Up Codes Applied to a GFR-Like Configuration

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Davide Chersola, Guglielmo Lomonaco, Guido Mazzini
Journal: Journal of Nuclear Engineering and Radiation Science
Publisher: ASME
Article Type: Research Papers
ASME J of Nuclear Rad Sci. July 2015, 1(3): 031004.
Paper No: NERS-14-1027
https://doi.org/10.1115/1.4029521
Published Online: May 20, 2015
View Article titled, Cross-Section Influence on Monte Carlo-Based Burn-Up Codes Applied to a GFR-Like Configuration
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Topics: Cross section (Physics), Density, Flux (Metallurgy), Fuels, Nuclides, Spectra (Spectroscopy), Nuclear fission, Fast neutron reactors