期刊
ANNALS OF NUCLEAR ENERGY
卷 82, 期 -, 页码 74-84出版社
PERGAMON-ELSEVIER SCIENCE LTD
DOI: 10.1016/j.anucene.2014.08.022
关键词
MORET; Monte Carlo; Neutron transport; Criticality; Reactor modeling; Validation
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. (C) 2014 Elsevier Ltd. All rights reserved.
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