Fusion Science and Technology / Volume 63 / Number 1T / May 2013 / Pages 322-324
The MCNPX numerical code has been used to model the neutron transport in a mirror based fusion-fission reactor. The purpose is to find a principal design of the fission mantle which fits to the neutron source and to calculate the leakage of neutrons through the mantle surface of the fission reactor.
The fission reactor part has a cylindrical shape with an outer radius 1.66 m and a 4 m length. The fuel has the isotopic composition of the spent nuclear fuel from PWR after uranium-238 removal. Inside the fission reactor core is a vacuum chamber with a radius 0.5 m containing a 4 m long hot plasma producing fusion neutrons. To sustain the hot ion plasma which is responsible for the fusion neutron production, neutral beam injection is considered.
Calculation results for the radial leakage of neutrons through the mantle surface of the fission reactor are presented. These calculations predict that the power released with neutrons from the reactor to outer space would be small and will not exceed the value of 6 kW while the reactor thermal power is 1 GWth.