Neutronic analysis of a high power density hybrid reactor using innovative coolants

Abstract

In this study, neutronic investigation of a deuterium-tritium (DT) driven hybrid reactor using ceramic uranium fuels, namely UC, UO<inf>2</inf> or UN under a high neutron wall load (NWL) of 10 MW/m2 at the first wall is conducted over a period of 24 months for fissile fuel breeding for light water reactors (LWRs). New substances, namely, Flinabe or Li<inf>20</inf>Sn<inf>80</inf> are used as coolants in the fuel zone to facilitate heat transfer out of the blanket. Natural lithium is also utilized for comparison to these two innovative coolants. Neutron transport calculations are performed on a simple experimental hybrid blanket with cylindrical geometry with the help of the SCALE 4·3 System by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and an S<inf>8</inf>-P<inf>3</inf> approximation. The investigated blanket using Flinabe or Li<inf>20</inf>Sn<inf>80</inf> shows better fissile fuel breeding and fuel enrichment characteristics compared to that with natural lithium which shows that these two innovative coolants can be used in hybrid reactors for higher fissile fuel breeding performance. Furthermore, using a high NWL of 10 MW/m2 at the first wall of the investigated blanket can decrease the time for fuel rods to reach the level for charging in LWRs. © 2018 Elsevier B.V., All rights reserved.