Documentation: OpenNTP¶
Neutron Transport Package with Graphical User Interface
Neutron Transport Package OpenNTP (Open Neutron Transport Package from the Radiations and Nuclear Systems Group), is an open-source code written in FORTRAN90 for a pedagogical purpose to solve the steady-state multigroup neutron transport equation using either:
- Collision Probablity Method (CP) in One-Dimensional for Slab, Cylindrical or Spherical geoemtry.
- Discrete Ordinate Method (\(S_{N}\)) in One or Two-Dimensional for Cartesian Geometry.
- Method of Characteristics (MOC) in One-Dimensional for Slab Geometry.
The code, including the graphical user interface is developed and maintained by Mohamed LAHDOUR (PhD student) and Prof. Tarek EL BARDOUNI from University Abdelmalek Essaadi Tetouan Morocco .
OpenNTP’s main features are:
- free & open source software with a pedagogical purpose.
- solve the steady-state multigroup neutron transport equation in one, or two spatial dimensions.
- solve the steady-state multigroup neutron transport equation in a multiplicative medium with isotropic and anisotropic scatternig source.
- simple framework to add and test new algorithms.
- provided with a graphical user interface written in Python programing language which has been developed to simplify the use of OpenNTP.
- computed results: \(k_{eff}\) and fluxes.
Recommended publication for citing
Lahdour, M., Bardouni, T. E., Chakir, E., Benaalilou, K., Mohammed, M., Bougueniz, H., and Yaakoubi, H. E., “Ntp-ersn: A new package for solving the multigroup neutron transport equation in a slab geometry.,” Applied Radiation and Isotopes, 73:84 – 145 (2019a).
Lahdour, M., Bardouni, T. E., Mohammed, M., Ouahdani, S. E. “The discrete ordinate method for angular flux calculations in slab geometry.,” Heliyon, e02211 5 (2019).