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 modelling, the transport simulation and the definition of the outputs.
Depletion codes such as VESTA are used to calculate the evolution of a material subjected to radiation (be it neutrons or another type of particle) for a wide variety of applications in the fields of nuclear safety, radiation protection and environmental health safety. For these applications, experimental validation is paramount. In this paper we will describe the experimental validation of the latest version of VESTA using a set of 76 samples consisting of radiochemical assay data and decay heat measurements. We will describe the general calculation procedure that has been applied to determine the uncertainty on each individual nuclide measurement as well as the general tendencies and detailed results.
The MORET 5 Monte Carlo code includes two calculation routes: a multi-group route based on cross-sections calculated from various preliminary cell codes and a continuous energy calculation route. The criticality experimental validation database is made up of 2255 benchmarks, mainly taken from the Handbook of the International Criticality Safety Benchmark Evaluated Experiments Project or performed in French facilities. Most of them have been investigated with the two different routes. The calculation results are generally in good agreement with the benchmarks, depending on the nuclear data used.
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