Neutron Cross Sections generated and used during the project: "Feasibility of the use of frozen walls in molten salt fast reactors (MSFR-FW)"
The data enclosed in this repository is associated with the manuscripts submitted to the European Physical Journal - Nuclear mid February 2019 and Nuclear Engineering and Design end of March 2019. The articles are adaptations of conference papers presented at the ICONE26 Conference in London, July 2018, the CFD4NRS-7 Workshop in Shanghai, September 2018 and the NUTHOS-12 topical meeting in Qingdao, October 2018. The data are neutron cross-sections, which were generated during the project "Feasibility of the use of frozen walls in molten salt fast reactors (MSFR-FW)". The neutron cross sections take the form of ascii files that have been formatted for the DYN3D-MG code. The subdirectory coreregion contains the reactor core neutron cross sections for the fuel salt eutectic LiF-PuF3-UF4 over the temperature range 623.15 K - 1198.15 K. The cross-sections were used in the coupled neutronic and computational fluid dynamic simulation of a cylindrical reactor with a low resolution. The subdirectory coreandreflector contains the reactorcore, wall and reflector neutron cross sections for the fuel salt eutectic LiF-PuF3-UF4 over the temperature range 623.15 K - 1198.15 K. The cross-sections were used in the coupled neutronic and computational fluid dynamic simulation of a reactor with the EVOL-optimized configuration. Below 814.22 K the fuel salt eutectic is assumed solid. All cross-sections above 841.55 K temperature are liquid. Between these temperatures the media is assumed to be a temperature dependent mixture of solid and liquid phases. Details of the assumed physical properties of the fuel-salt eutectic are given in the above references. The raw data was prepared using the SCARF (scarf.rl.ac.uk). The data was generated using SERPENT (http://montecarlo.vtt.fi/). SERPENT (version 2.1.29) modelled the neutronic behaviour of the molten salt fast reactor using the Monte Carlo method. The data was reformatted for the nodal neutron diffusion solver DYN3D-MG (https://www.hzdr.de/db/Cms?pOid=11771&pNid=542) to read. This is the format in which neutron cross-sections are given.