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Validation of deterministic code DRAGON5 for the fuel depletion analysis of a PWR pin-cell benchmark

Radiation physics and chemistry (Oxford, England : 1993), 2021-09, Vol.186, p.109545, Article 109545

2021

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Neutronic and burn‐up calculations of the (ThO2‐UO2) pin cell benchmark using DRAGON5 and MCNP6.2 codes with ENDF/B‐VIII.0 nuclear data library

International journal of energy research, 2021-06, Vol.45 (8), p.11538-11551

2021

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Eigenvalue sensitivity and nuclear data uncertainty analysis for the Moroccan TRIGA Mark II research reactor using SCALE6.2 and MCNP6.2

Nuclear engineering and design, 2021-07, Vol.378, p.111160, Article 111160

2021

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Validation of deterministic code OpenNTP for the analysis of the C5G7-1D benchmark with different configurations

Applied radiation and isotopes, 2022-09, Vol.187, p.110317-110317, Article 110317

2022

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First principles calculation for generating new thermal neutron scattering data for Zirconium hydride

Applied radiation and isotopes, 2022-09, Vol.187, p.110313-110313, Article 110313

2022

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Validation and Neutronic Analysis of the CENM TRIGA MARK II SCALE Research Reactor Model: Computations Versus Experiment

Nuclear technology, 2024-01, Vol.210 (1), p.72-83

2024

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Modeling and simulation of a TRIGA MARK-II research reactor using WIMSD-5B and CITATION codes

Applied radiation and isotopes, 2019-06, Vol.148, p.64-75

2019

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Sensitivity and uncertainty quantification of neutronic integral data in the TRIGA Mark II research reactor

Nuclear Engineering and Technology, 2022, 54(2), , pp.523-531

2022

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Analysis of Depletion and Inventory of the Fuel for the MNSR Research Reactor Using the Deterministic Code DRAGON5

Nuclear technology, 2020-04, Vol.206 (4), p.620-636

2020

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NTP-ERSN: A new package for solving the multigroup neutron transport equation in a slab geometry

Applied radiation and isotopes, 2019-03, Vol.145, p.73-84

2019

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Neutronic modeling and calculation of the Nuclear Heating Reactor NHR-5 by the deterministic codes DRAGON5 & DONJON5

Progress in nuclear energy (New series), 2021-12, Vol.142, p.104000, Article 104000

2021

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Benchmark Comparisons of OpenMC and DRAGON5 Codes for Calculating the Doppler Coefficient of Reactivity

Moscow University physics bulletin, 2022-12, Vol.77 (6), p.922-929

2022

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Fuel reloads optimization for TRIGA research reactor using Genetic Algorithm coupled with neutronic and thermal-hydraulic codes

Progress in nuclear energy (New series), 2021-03, Vol.133, p.103637, Article 103637

2021

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The discrete ordinate method for angular flux calculations in slab geometry

Heliyon, 2019-08, Vol.5 (8), p.e02211-e02211, Article e02211

2019

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NTP-ERSN verification with C5G7 1D extension benchmark and GUI development

Nuclear Engineering and Technology, 2021, 53(4), , pp.1079-1087

2021

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Analysis of the Recent Nuclear Data through a Fast Benchmarks Highly Enriched Uranium Study Using the Mcnp6.1 Code

Moscow University physics bulletin, 2020, Vol.75 (1), p.63-69

2020

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OpenNTP: Implementation of the S N method in cartesian 2D geometry and the CP method in cylindrical and spherical 1D geometry

Computer physics communications, 2021-04, Vol.261, p.107812, Article 107812

2021

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ERSN-OpenMC-Py: A python-based open-source software for OpenMC Monte Carlo code

Computer physics communications, 2024-06, Vol.299, p.109121, Article 109121

2024

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Development and implementation of a numerical simulation system for analyzing the thermal-hydraulic (T-H) characteristics of the MNSR research reactor fuel assembly

Nuclear engineering and design, 2024-01, Vol.416, p.112759, Article 112759

2024

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OpenNTP: Implementation of the SN method in cartesian 2D geometry and the CP method in cylindrical and spherical 1D geometry

Computer physics communications, 2021-04, Vol.261, Article 107812

2021

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Neutronic and burn‐up calculations of the ( ThO2‐UO2 ) pin cell benchmark using DRAGON5 and MCNP6 .2 codes with ENDF / B‐VIII .0 nuclear data library

International journal of energy research, 2021-06, Vol.45 (8), p.11538-11551

2021

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Calculation of the activity inventory of the CENM TRIGA MARK II reactor for decommissioning planninig

Radiation physics and chemistry (Oxford, England : 1993), 2023-06, Vol.207, p.110837, Article 110837

2023

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Sensitivity and uncertainty quantification of neutronic integral data in the TRIGA Mark II research reactor

Nuclear engineering and technology, 2022, Vol.54 (2), p.523-531

2022

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NTP-ERSN verification with C5G7 1D extension benchmark and GUI development

Nuclear engineering and technology, 2021, Vol.53 (4), p.1079-1087

2021

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Validation study of the reactor physics lattice transport code DRAGON5 & the Monte Carlo code OpenMC by critical experiments of light water reactors

Journal of King Saud University. Science, 2019-10, Vol.31 (4), p.1271-1275

2019

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Validation of deterministic code DRAGON5 for the fuel depletion analysis of a PWR pin-cell benchmark

Neutronic and burn‐up calculations of the (ThO2‐UO2) pin cell benchmark using DRAGON5 and MCNP6.2 codes with ENDF/B‐VIII.0 nuclear data library

Eigenvalue sensitivity and nuclear data uncertainty analysis for the Moroccan TRIGA Mark II research reactor using SCALE6.2 and MCNP6.2

Validation of deterministic code OpenNTP for the analysis of the C5G7-1D benchmark with different configurations

First principles calculation for generating new thermal neutron scattering data for Zirconium hydride

Validation and Neutronic Analysis of the CENM TRIGA MARK II SCALE Research Reactor Model: Computations Versus Experiment

Modeling and simulation of a TRIGA MARK-II research reactor using WIMSD-5B and CITATION codes

Sensitivity and uncertainty quantification of neutronic integral data in the TRIGA Mark II research reactor

Analysis of Depletion and Inventory of the Fuel for the MNSR Research Reactor Using the Deterministic Code DRAGON5

NTP-ERSN: A new package for solving the multigroup neutron transport equation in a slab geometry

Neutronic modeling and calculation of the Nuclear Heating Reactor NHR-5 by the deterministic codes DRAGON5 & DONJON5

Benchmark Comparisons of OpenMC and DRAGON5 Codes for Calculating the Doppler Coefficient of Reactivity

Fuel reloads optimization for TRIGA research reactor using Genetic Algorithm coupled with neutronic and thermal-hydraulic codes

The discrete ordinate method for angular flux calculations in slab geometry

NTP-ERSN verification with C5G7 1D extension benchmark and GUI development

Analysis of the Recent Nuclear Data through a Fast Benchmarks Highly Enriched Uranium Study Using the Mcnp6.1 Code

OpenNTP: Implementation of the S N method in cartesian 2D geometry and the CP method in cylindrical and spherical 1D geometry

ERSN-OpenMC-Py: A python-based open-source software for OpenMC Monte Carlo code

Development and implementation of a numerical simulation system for analyzing the thermal-hydraulic (T-H) characteristics of the MNSR research reactor fuel assembly

OpenNTP: Implementation of the SN method in cartesian 2D geometry and the CP method in cylindrical and spherical 1D geometry

Neutronic and burn‐up calculations of the ( ThO2‐UO2 ) pin cell benchmark using DRAGON5 and MCNP6 .2 codes with ENDF / B‐VIII .0 nuclear data library

Calculation of the activity inventory of the CENM TRIGA MARK II reactor for decommissioning planninig

Sensitivity and uncertainty quantification of neutronic integral data in the TRIGA Mark II research reactor

NTP-ERSN verification with C5G7 1D extension benchmark and GUI development

Validation study of the reactor physics lattice transport code DRAGON5 & the Monte Carlo code OpenMC by critical experiments of light water reactors

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