HYBRID «FUSION-FISSION» REACTOR FACILITY: NEUTRONIC RESEARCH

Abstract

Global thermonuclear research is aimed at the commercial energy production after 2050. Scientists of Korean Institute of Fusion Energy (KFE) achieve 20 seconds plasma confinement in toroidal magnetic trap (Korea Superconducting Tokamak Advanced Research) at temperature of 100 million degrees which is currently an absolute record for the implemented set of parameters. The research result of scientists allows asserting that the international program for the ITER tokamak contruction will be completed and the energy production from plasma will exceed the energy costs for obtaining and retaining plasma. Power generation stations using tokamaks will have an exceptionally large size and power, and will be build in the distant future. Our research is focused on prospect of practical using of nuclear fusion power at a shorter time. The purpose of our research is to creation a subcritical «fusion-fission» facility the concept of which is proposed and developed by the G.I. Budker Institute of Nuclear Physics SB RAS (Novosibirsk), Tomsk Polytechnic University and the Russian Federal Nuclear Center - VNIITF n.a. Academician E.I. Zababakhin (Snezhinsk). The facility under study is a hybrid reactor. The reactor core (blanket) is consisted of the assembly of fuel blocks of a unified design of a high-temperature gas-cooled thorium reactor and an extended magnetic gas-dynamic trap penetrating the axial part of the reactor core (G.I. Budker Institute of Nuclear Physics SB RAS, Novosibirsk). Such a hybrid nuclear fusion reactor has a convenient level of facility power for regional energy (~ 60 - 100 MWth), acceptable geometric dimensions and insignificant volume of spent nuclear fuel and radioactive waste generation in comparison to most common reactors such as LWR. In this paper we provide optimization neutron studies the purpose of which is to neutralize the offsets of the radial energy release field formed in the volume of the fuel part of the blanket due to prolonged operation time and pulsed work mode of the D-T-neutron plasma source. The main goal of the study was to reduce blanket's power peaking factor, that depends on time-dependent neutron flux distribution and periodic pulse mode operation parameters of the D-T fusion neutron source. Neutron transport simulations were done with Monte-Carlo code SERPENT 2.1.31. Evaluated point-wise nuclear data including S(α,β) thermal scattering data for graphite of ENDF-B/VII.0 library were used for simulation.