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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

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

Superphénix Benchmark Part I: Results of Static Neutronics

Alexander Ponomarev, Konstantin Mikityuk, Liang Zhang, Evgeny Nikitin, Emil Fridman, Francisco Álvarez-Velarde, Pablo Romojaro Otero, Antonio Jiménez-Carrascosa, Nuria García-Herranz, Ben Lindley, Una Baker, Armin Seubert, Romain Henry

ASME J of Nuclear Rad Sci. January 2022, 8(1): 011320.

Paper No: NERS-21-1046

DOI: https://doi.org/10.1115/1.4051449

Published Online: October 8, 2021

View Article titled, Superphénix Benchmark Part I: Results of Static Neutronics

Topics: Density, Fuels, Rods, Sodium, Temperature, Steel, Geometry, Cladding systems (Building), Neutrons, Chemical kinetics

Journal Articles

Modeling of Reactivity Effects and Transient Behavior of Large Sodium Fast Reactor

Alexander Ponomarev, Konstantin Mikityuk

ASME J of Nuclear Rad Sci. January 2022, 8(1): 011313.

Paper No: NERS-20-1186

DOI: https://doi.org/10.1115/1.4051514

Published Online: September 13, 2021

View Article titled, Modeling of Reactivity Effects and Transient Behavior of Large Sodium Fast Reactor

Topics: Flow (Dynamics), Fuels, Modeling, Rods, Sodium, Sodium fast reactors, Temperature, Transients (Dynamics), Electrical conductance, Heat

Journal Articles

Evaluation of the Passive Reactor Shutdown System Performance in European Sodium Fast Reactor

Evaldas Bubelis, Michael Schikorr, Konstantin Mikityuk

ASME J of Nuclear Rad Sci. January 2022, 8(1): 011309.

Paper No: NERS-21-1048

DOI: https://doi.org/10.1115/1.4050563

Published Online: August 3, 2021

View Article titled, Evaluation of the Passive Reactor Shutdown System Performance in European Sodium Fast Reactor

Topics: Rods, Sodium fast reactors, Transients (Dynamics), Cooling, Circuits, Flow (Dynamics)

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

Verifying Calculation of the Parameters of AP1000 Core Design Based on MCNP Model

Xiong Wenbin, Cao Jian, Huang Chaoyun, Bie Yewang, Wang Yanqi, Li Jufeng

ASME J of Nuclear Rad Sci. July 2018, 4(3): 031018.

Paper No: NERS-17-1132

DOI: https://doi.org/10.1115/1.4039596

Published Online: May 16, 2018

View Article titled, Verifying Calculation of the Parameters of AP1000 Core Design Based on MCNP Model

Topics: Design, Fuels, Rods, Manufacturing, Nuclear power stations, Modeling, Boron, Physics

Journal Articles

Analysis of the Core Physical Characteristics of China Experimental Fast Reactor With MCNP

Xiong Wenbin, Xie Qin, Li Huwei, Yang Sengai, Mao Huan, Cao Jian

ASME J of Nuclear Rad Sci. April 2018, 4(2): 024501.

Paper No: NERS-17-1105

DOI: https://doi.org/10.1115/1.4038999

Published Online: March 5, 2018

View Article titled, Analysis of the Core Physical Characteristics of China Experimental Fast Reactor With MCNP

Topics: China, Fast neutron reactors, Safety, Design, Rods

Journal Articles

Transient Heat Transfer in an Out-of-Pile SCWR Fuel Assembly Test at Near-Critical Pressure

Thomas Schulenberg, Hongbo Li

ASME J of Nuclear Rad Sci. January 2018, 4(1): 011005.

Paper No: NERS-17-1026

DOI: https://doi.org/10.1115/1.4038061

Published Online: December 4, 2017

View Article titled, Transient Heat Transfer in an Out-of-Pile SCWR Fuel Assembly Test at Near-Critical Pressure

Topics: Heat transfer, Pressure, Supercritical water reactors, Temperature, Transient heat transfer, Fuels, Film boiling, Rods, Heat flux, Transients (Dynamics)

Journal Articles

Benchmarking the Real-Time Core Model for VVER-1000 Simulator Application on Asymmetric Core Load

Emiliya Georgieva, Yavor Dinkov, Kostadin Ivanov

ASME J of Nuclear Rad Sci. July 2017, 3(3): 031005.

Paper No: NERS-16-1132

DOI: https://doi.org/10.1115/1.4035550

Published Online: May 25, 2017

View Article titled, Benchmarking the Real-Time Core Model for VVER-1000 Simulator Application on Asymmetric Core Load

Topics: Coolants, Flow (Dynamics), Fuels, Manufacturing, Pressure, Simulation, Steam, Stress, Temperature, Transients (Dynamics)

Journal Articles

Study on Sensitivity of Control Rod Cell Model in Reflector Region of High-Temperature Engineering Test Reactor

Yuki Honda, Nozomu Fujimoto, Hiroaki Sawahata, Shoji Takada, Kazuhiro Sawa

ASME J of Nuclear Rad Sci. January 2017, 3(1): 011005.

Paper No: NERS-15-1226

DOI: https://doi.org/10.1115/1.4033813

Published Online: December 20, 2016

View Article titled, Study on Sensitivity of Control Rod Cell Model in Reflector Region of High-Temperature Engineering Test Reactor

Topics: High temperature, Neutron flux, Rods, Fuels, Design, Temperature, Computational methods

Journal Articles

A Space–Time-Dependent Study of Control Rods Withdrawal in a Large-Size Pressurized Water Reactor

Sanjeev Kumar, K. Obaidurrahman, Om Pal Singh, Prabhat Munshi

ASME J of Nuclear Rad Sci. January 2017, 3(1): 011015.

Paper No: NERS-16-1007

DOI: https://doi.org/10.1115/1.4034478

Published Online: December 20, 2016

View Article titled, A Space–Time-Dependent Study of Control Rods Withdrawal in a Large-Size Pressurized Water Reactor

Topics: Heat flux, Nucleate boiling, Pressurized water reactors, Rods, Safety, Scram, Spacetime, Thermal hydraulics, Transients (Dynamics), Temperature profiles

Journal Articles

Neutronic Design Features of a Transportable Fluoride-Salt-Cooled High-Temperature Reactor

Kaichao Sun, Lin-Wen Hu, Charles Forsberg

ASME J of Nuclear Rad Sci. July 2016, 2(3): 031003.

Paper No: NERS-15-1184

DOI: https://doi.org/10.1115/1.4032873

Published Online: June 17, 2016

View Article titled, Neutronic Design Features of a Transportable Fluoride-Salt-Cooled High-Temperature Reactor

Topics: Coolants, Design, Fuels, Nuclear fuel cycle, Safety, Rods, Particulate matter, Control systems, High temperature

Journal Articles

Computational Fluid Dynamics Investigation of Supercritical Water Flow and Heat Transfer in a Rod Bundle With Grid Spacers

Malwina Gradecka, Roman Thiele, Henryk Anglart

ASME J of Nuclear Rad Sci. July 2016, 2(3): 031015.

Paper No: NERS-15-1087

DOI: https://doi.org/10.1115/1.4032635

Published Online: June 17, 2016

View Article titled, Computational Fluid Dynamics Investigation of Supercritical Water Flow and Heat Transfer in a Rod Bundle With Grid Spacers

Topics: Computational fluid dynamics, Flow (Dynamics), Fuel rods, Heat transfer, Rods, Temperature, Water, Geometry, Boundary-value problems, Fluids

Journal Articles

A Blind, Numerical Benchmark Study on Supercritical Water Heat Transfer Experiments in a 7-Rod Bundle

M. Rohde, J. W. R. Peeters, A. Pucciarelli, A. Kiss, Y. F. Rao, E. N. Onder, P. Muehlbauer, A. Batta, M. Hartig, V. Chatoorgoon, R. Thiele, D. Chang, S. Tavoularis, D. Novog, D. McClure, M. Gradecka, K. Takase

ASME J of Nuclear Rad Sci. April 2016, 2(2): 021012.

Paper No: NERS-15-1156

DOI: https://doi.org/10.1115/1.4031949

Published Online: February 29, 2016

View Article titled, A Blind, Numerical Benchmark Study on Supercritical Water Heat Transfer Experiments in a 7-Rod Bundle

Topics: Flow (Dynamics), Heat transfer, Reynolds-averaged Navier–Stokes equations, Rods, Simulation, Temperature, Turbulence, Wall temperature, Water, Fuel rods

Journal Articles

Monte Carlo Simulation of Inclined Reactivity Control Rod and Boron Poisoning for the Canadian-SCWR Supercell

Haykel Raouafi, Guy Marleau

ASME J of Nuclear Rad Sci. April 2016, 2(2): 021015.

Paper No: NERS-15-1032

DOI: https://doi.org/10.1115/1.4030895

Published Online: February 29, 2016

View Article titled, Monte Carlo Simulation of Inclined Reactivity Control Rod and Boron Poisoning for the Canadian-SCWR Supercell

Topics: Boron, Supercritical water reactors, Coolants, Fuels, Density, Simulation, Rods

Journal Articles

A Reduced Basis Approach for Modeling the Movement of Nuclear Reactor Control Rods

Alberto Sartori, Antonio Cammi, Lelio Luzzi, Gianluigi Rozza

ASME J of Nuclear Rad Sci. April 2016, 2(2): 021019.

Paper No: NERS-14-1062

DOI: https://doi.org/10.1115/1.4031945

Published Online: February 29, 2016

View Article titled, A Reduced Basis Approach for Modeling the Movement of Nuclear Reactor Control Rods

Topics: Eigenvalues, Errors, Modeling, Neutron flux, Nuclear reactors, Rods, Shapes, Approximation, Neutrons, Diffusion (Physics)

Journal Articles

Application of Thermal Hydraulic and Severe Accident Code SOCRAT/V3 to Bottom Water Reflood Experiment QUENCH-LOCA-1

Alexander Vasiliev, Juri Stuckert

ASME J of Nuclear Rad Sci. April 2016, 2(2): 021024.

Paper No: NERS-15-1027

DOI: https://doi.org/10.1115/1.4031815

Published Online: February 29, 2016

View Article titled, Application of Thermal Hydraulic and Severe Accident Code SOCRAT/V3 to Bottom Water Reflood Experiment QUENCH-LOCA-1

Topics: Accidents, Cladding systems (Building), Hydrogen, Modeling, Temperature, Water, Thermomechanics, Rods, Steam, Heat

Journal Articles

Experimental Heat Transfer in an Annular Channel and 3-Rod Bundle Cooled With Upward Flow of Supercritical Water

V. G. Razumovskiy, Eu. N. Pis’mennyi, Kh. Sidawi, I. L. Pioro, A. Eu. Koloskov

ASME J of Nuclear Rad Sci. January 2016, 2(1): 011010.

Paper No: NERS-15-1117

DOI: https://doi.org/10.1115/1.4031818

Published Online: December 9, 2015

View Article titled, Experimental Heat Transfer in an Annular Channel and 3-Rod Bundle Cooled With Upward Flow of Supercritical Water

Topics: Flow (Dynamics), Fluids, Heat flux, Heat transfer, Temperature, Wall temperature, Water, Rods, Heat, Pressure

Journal Articles

Numerical Study of Fluid–Structure Interaction With Heat Transfer at Supercritical Pressure in a Fuel Rod Assembly

Maximilian Hartig, Thomas Schulenberg, Abdalla Batta

ASME J of Nuclear Rad Sci. January 2016, 2(1): 011013.

Paper No: NERS-15-1059

DOI: https://doi.org/10.1115/1.4031816

Published Online: December 9, 2015

View Article titled, Numerical Study of Fluid–Structure Interaction With Heat Transfer at Supercritical Pressure in a Fuel Rod Assembly

Topics: Deflection, Fluid structure interaction, Fluids, Heat transfer, Rods, Temperature, Temperature profiles, Deformation, Fuel rods, Manufacturing

Journal citation

NARROW

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

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

Superphénix Benchmark Part I: Results of Static Neutronics

Modeling of Reactivity Effects and Transient Behavior of Large Sodium Fast Reactor

Evaluation of the Passive Reactor Shutdown System Performance in European Sodium Fast Reactor

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

Verifying Calculation of the Parameters of AP1000 Core Design Based on MCNP Model

Analysis of the Core Physical Characteristics of China Experimental Fast Reactor With MCNP

Transient Heat Transfer in an Out-of-Pile SCWR Fuel Assembly Test at Near-Critical Pressure

Benchmarking the Real-Time Core Model for VVER-1000 Simulator Application on Asymmetric Core Load

Study on Sensitivity of Control Rod Cell Model in Reflector Region of High-Temperature Engineering Test Reactor

A Space–Time-Dependent Study of Control Rods Withdrawal in a Large-Size Pressurized Water Reactor

Neutronic Design Features of a Transportable Fluoride-Salt-Cooled High-Temperature Reactor

Computational Fluid Dynamics Investigation of Supercritical Water Flow and Heat Transfer in a Rod Bundle With Grid Spacers

A Blind, Numerical Benchmark Study on Supercritical Water Heat Transfer Experiments in a 7-Rod Bundle

Monte Carlo Simulation of Inclined Reactivity Control Rod and Boron Poisoning for the Canadian-SCWR Supercell

A Reduced Basis Approach for Modeling the Movement of Nuclear Reactor Control Rods

Application of Thermal Hydraulic and Severe Accident Code SOCRAT/V3 to Bottom Water Reflood Experiment QUENCH-LOCA-1

Experimental Heat Transfer in an Annular Channel and 3-Rod Bundle Cooled With Upward Flow of Supercritical Water

Numerical Study of Fluid–Structure Interaction With Heat Transfer at Supercritical Pressure in a Fuel Rod Assembly

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