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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.

Paper No: NERS-22-1145

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

Published Online: March 11, 2023

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Topics: Additive manufacturing, Fuels, Particulate matter, Design, Geometry, Coolants, Heat transfer, Computational fluid dynamics, Computer software, Discrete element methods

Journal Articles

Solution of Neutron Diffusion Problems by Discontinuous Galerkin Finite Element Method With Consideration of Discontinuity Factors

Li Wanai, Gong Helin, Zhang Chunyu

ASME J of Nuclear Rad Sci. July 2023, 9(3): 031503.

Paper No: NERS-21-1114

DOI: https://doi.org/10.1115/1.4055379

Published Online: February 22, 2023

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Topics: Diffusion (Physics), Finite element methods, Neutron flux, Neutrons, Computation, Finite element model

Journal Articles

Theoretical Investigation of the Gamma and Neutron Interaction Parameters of Some Inorganic Scintillators Using Phy-X/PSD and NGCal Software

G. B. Hiremath, M. M. Hosamani, V. P. Singh, N. H. Ayachit, N. M. Badiger

ASME J of Nuclear Rad Sci. July 2023, 9(3): 032004.

Paper No: NERS-22-1046

DOI: https://doi.org/10.1115/1.4056835

Published Online: February 22, 2023

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Topics: Absorption, Atomic number, Computer software, Neutrons, Photons, Radiation (Physics)

Journal Articles

Estimate of Neutron Fluence Using Isotope Ratio Method for ZR-2.5NB Materials Irradiated in the National Research Universal Reactor

Xiaolin Wang, Youqing Shi, Grant Bickel, DeAbreu Robert

ASME J of Nuclear Rad Sci. July 2023, 9(3): 034501.

Paper No: NERS-22-1039

DOI: https://doi.org/10.1115/1.4055931

Published Online: February 8, 2023

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Topics: Fluence (Radiation measurement), Irradiation (Radiation exposure), Isotopes, Neutrons, Zirconium

Journal Articles

Benchmarking Fast Neutrons Leakage Spectrum From Copper Block With 252 Cf Source in Center

Tomáš Czakoj, Michal Košťál, Martin Schulc, Vojtěch Rypar, Evžen Novák, Evžen Losa

ASME J of Nuclear Rad Sci. July 2023, 9(3): 032001.

Paper No: NERS-22-1038

DOI: https://doi.org/10.1115/1.4056508

Published Online: February 8, 2023

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Topics: Copper, Leakage, Neutrons, Uncertainty, Spectra (Spectroscopy), Neutron sources

Journal Articles

Burnup Characteristics of Transuranic Fuel With Burnable Poison in Dual Cooled Annular Rod of VVER-1000

Md. Tanvir Ahmed, Md. Hossain Sahadath, Md. Ashraful Islam

ASME J of Nuclear Rad Sci. July 2023, 9(3): 031501.

Paper No: NERS-22-1022

DOI: https://doi.org/10.1115/1.4055522

Published Online: February 8, 2023

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Topics: Fuels, Cycles, Neutron flux, Neutrons, Manufacturing

Journal Articles

Optimized Moderator Design and Analysis of a Pin-Type Supercritical Carbon Dioxide Reactor Based On RMC

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

ASME J of Nuclear Rad Sci.

Paper No: NERS-22-1137

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

Published Online: January 1, 2023

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Topics: Design, Supercritical carbon dioxide, Coolants, Feedback, Diluents, Fuels, Leakage, Neutrons, Cross section (Physics), Nuclear fission

Journal Articles

Physics Evaluation of Alternative Uranium-Based Oxy-Carbide Annular Fuel Concepts for Potential Use in Compact High-Temperature Gas-Cooled Reactors

Daniel T. Wojtaszek, Blair P. Bromley

ASME J of Nuclear Rad Sci. January 2023, 9(1): 011504.

Paper No: NERS-22-1009

DOI: https://doi.org/10.1115/1.4055009

Published Online: September 13, 2022

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Topics: Fuels, Uranium, Neutrons

Journal Articles

Preliminary Analysis of Reactivity Feedbacks for Inner-Side Chromium Coating

Rofida Hamad Khlifa, Nicolay N. Nikitenkov

ASME J of Nuclear Rad Sci. January 2023, 9(1): 012003.

Paper No: NERS-22-1044

DOI: https://doi.org/10.1115/1.4055012

Published Online: September 1, 2022

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Topics: Boron, Coatings, Fuels, Temperature, Cladding systems (Building), Neutrons, Accidents

Journal Articles

Possible Implementation of Ex-Core Measurement in TEPLATOR Graphite Reflector

Eva Vilímová, Tomáš Peltan, Jana Jiřičková

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041505.

Paper No: NERS-21-1013

DOI: https://doi.org/10.1115/1.4050990

Published Online: June 15, 2022

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Topics: Graphite, Neutron flux, Neutrons, Sensors, Spent nuclear fuels, Water, Fuels

Journal Articles

Calculation of Gamma and Neutron Shielding Parameters for Reinforced Polymer Composites

A. M. Osman

ASME J of Nuclear Rad Sci. January 2023, 9(1): 012002.

Paper No: NERS-21-1171

DOI: https://doi.org/10.1115/1.4054547

Published Online: June 7, 2022

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Topics: Composite materials, Density, Neutrons, Polymer composites, Relaxation (Physics), Photons, Cross section (Physics), Aluminum, Boron

Journal Articles

Efficient Neutron Irradiation Procedure for Uniform Dopant Distributions in Silicon Doping

Nader M. A. Mohamed, Mosa Osman, Ibrahim Abdelrazek

ASME J of Nuclear Rad Sci. July 2022, 8(3): 034503.

Paper No: NERS-21-1109

DOI: https://doi.org/10.1115/1.4054005

Published Online: May 26, 2022

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Topics: Irradiation (Radiation exposure), Neutron flux, Neutrons, Silicon

Journal Articles

Physics Modeling for Conceptual Designs of Proposed Experiments in the Zero Energy Deuterium-2 Critical Facility for Testing Small Modular Reactor-Type Fuels

D. G. Watts, F. P. Adams, E. Masala, L. Blomeley, B. P. Bromley

ASME J of Nuclear Rad Sci. July 2022, 8(3): 031501.

Paper No: NERS-21-1118

DOI: https://doi.org/10.1115/1.4054079

Published Online: May 26, 2022

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Topics: Fuels, Modeling, Physics, Pressurized water reactors, Water, Neutrons, High temperature

Journal Articles

Reevaluation of Spectral Parameters and Neutron Fluxes in IC-7 Irradiation Channel of TRIGA MARK I IPR-R1 Research Nuclear Reactor

Radojko Jaćimović, Maria Ângela de Barros Correia Menezes

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041504.

Paper No: NERS-21-1007

DOI: https://doi.org/10.1115/1.4051249

Published Online: March 22, 2022

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Topics: Flux (Metallurgy), Irradiation (Radiation exposure), Neutrons, Neutron activation analysis, Uncertainty

Journal Articles

Analysis of the HI-TRAC VW Transfer Cask Dose Rates for Spent Fuel Assemblies Loaded in Nuclear Power Plant Krsko Storage Campaign One

Davor Grgić, Mario Matijević, Paulina Dučkić, Radomir Ječmenica

ASME J of Nuclear Rad Sci. October 2022, 8(4): 041902.

Paper No: NERS-21-1029

DOI: https://doi.org/10.1115/1.4051447

Published Online: March 16, 2022

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Topics: Accidents, Fuels, Hardware, Neutrons, Nuclear power stations, Particulate matter, Spent nuclear fuels, Storage, Uncertainty, Manufacturing

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

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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

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

Journal Articles

serpent 2 Validation for Radiation Shielding Applications

Silja Häkkinen

ASME J of Nuclear Rad Sci. October 2022, 8(4): 042001.

Paper No: NERS-21-1009

DOI: https://doi.org/10.1115/1.4051614

Published Online: March 15, 2022

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Topics: Iron, Neutrons, Photons, Spectra (Spectroscopy)

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

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

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

A. S. Bikeev, Yu. S. Daichenkova, M. A. Kalugin, D. A. Shkarovsky, V. V. Shkityr

ASME J of Nuclear Rad Sci. April 2022, 8(2): 024507.

Paper No: NERS-20-1005

DOI: https://doi.org/10.1115/1.4052340

Published Online: October 19, 2021

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Topics: Approximation, Cross section (Physics), Electromagnetic scattering, Errors, Neutrons, Radiation scattering, Scattering (Physics), Fuels, Anisotropy

2 3 4 5

Journal citation

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Supplemental Material

Manuscript

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

Solution of Neutron Diffusion Problems by Discontinuous Galerkin Finite Element Method With Consideration of Discontinuity Factors

Theoretical Investigation of the Gamma and Neutron Interaction Parameters of Some Inorganic Scintillators Using Phy-X/PSD and NGCal Software

Estimate of Neutron Fluence Using Isotope Ratio Method for ZR-2.5NB Materials Irradiated in the National Research Universal Reactor

Benchmarking Fast Neutrons Leakage Spectrum From Copper Block With 252 Cf Source in Center

Burnup Characteristics of Transuranic Fuel With Burnable Poison in Dual Cooled Annular Rod of VVER-1000

Optimized Moderator Design and Analysis of a Pin-Type Supercritical Carbon Dioxide Reactor Based On RMC

Physics Evaluation of Alternative Uranium-Based Oxy-Carbide Annular Fuel Concepts for Potential Use in Compact High-Temperature Gas-Cooled Reactors

Preliminary Analysis of Reactivity Feedbacks for Inner-Side Chromium Coating

Possible Implementation of Ex-Core Measurement in TEPLATOR Graphite Reflector

Calculation of Gamma and Neutron Shielding Parameters for Reinforced Polymer Composites

Efficient Neutron Irradiation Procedure for Uniform Dopant Distributions in Silicon Doping

Physics Modeling for Conceptual Designs of Proposed Experiments in the Zero Energy Deuterium-2 Critical Facility for Testing Small Modular Reactor-Type Fuels

Reevaluation of Spectral Parameters and Neutron Fluxes in IC-7 Irradiation Channel of TRIGA MARK I IPR-R1 Research Nuclear Reactor

Analysis of the HI-TRAC VW Transfer Cask Dose Rates for Spent Fuel Assemblies Loaded in Nuclear Power Plant Krsko Storage Campaign One

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

serpent 2 Validation for Radiation Shielding Applications

A Short Inroad to Optimized Compactification of Composite Neutronic Shields

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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