Generation of nuclear data for lead-cooled fast reactors using the Monte Carlo method

Abstract

The LEADER project goal is to improve and develop a scaled demonstrator of the LFR technology, ALFRED. The work in this thesis is focused in the ALFRED project framework and its mission is to obtain few-group cross section data for LFRs. Cross sections The neutron transport problem is crucial in nuclear engineering and nuclear reactor physics. Neutron transport theory and the diffusion theory applied to neutron reactions are briefly described, including their principles and hypothesis. The two different computational approaches to solve the neutron transport problem are summarized. The software used to obtain the data is based on a modification of the Monte Carlo method. Thus, some basic probability theory concepts are introduced. This section follows with the discussion of the Monte Carlo method and its principles, and how it can be applied to solve the neutron transport problem. Afterwards, the Serpent code is explained, as well as its features and characteristics. The process of creating a 2-dimension model of ALFRED fuel assembly and the elaboration of Serpent input files are detailed. Cross section data for five neutron energy groups and at different material temperatures is obtained by running several simulations using Serpent. The last section includes a brief description of LFR technology and some specific ALFRED features. Some advantages and disadvantages of LFRs are included, along with some proposals to solve the disadvantages. The last part of this section illustrates the proposed ALFRED core scheme, using the data publicly available to date.