Capabilities overview of the MORET 5 Monte Carlo code
| Autor: | Olivier Jacquet, Alexis Jinaphanh, L. Heulers, B. Cochet |
|---|---|
| Přispěvatelé: | PSN-EXP/SNC, Institut de Radioprotection et de Sûreté Nucléaire (IRSN) |
| Jazyk: | angličtina |
| Rok vydání: | 2015 |
| Předmět: |
Radiation transport
[PHYS]Physics [physics] Neutron transport Theoretical computer science Computer science 020209 energy Monte Carlo method 02 engineering and technology 01 natural sciences Calculation methods Computational science 010104 statistics & probability Neutron transport theory Nuclear Energy and Engineering Criticality Monte carlo code 0202 electrical engineering electronic engineering information engineering Code (cryptography) 0101 mathematics Convection–diffusion equation |
| Zdroj: | Annals of Nuclear Energy Annals of Nuclear Energy, Elsevier Masson, 2015, 82, pp.74-84. ⟨10.1016/j.anucene.2014.08.022⟩ |
| ISSN: | 0306-4549 |
| DOI: | 10.1016/j.anucene.2014.08.022⟩ |
| Popis: | International audience; The MORET code is a simulation tool that solves the transport equation for neutrons using the Monte Carlo method. It allows users to model complex three-dimensional geometrical configurations, describe the materials, define their own tallies in order to analyse the results. The MORET code has been initially designed to perform calculations for criticality safety assessments. New features has been introduced in the MORET 5 code to expand its use for reactor applications. This paper presents an overview of the MORET 5 code capabilities, going through the description of materials, the geometry modeling, the transport simulation and the definition of the outputs. © 2014 Elsevier Ltd.All rights reserved. |
| Databáze: | OpenAIRE |
| Externí odkaz: |
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