Gas-cooled thorium reactor with fuel block of the unified design

Scientific investigation aimed at the implementation of advance technological platform is carried out in Russia based on the ideas of nuclear fuel breeding within closed fuel cycle and on the physical principles of fast reactors. Innovation design of low-power reactor facilities also fall under the new technological platform. High-temperature gas-cooled nuclear reactors operated with thorium fuel load possessing advantageous features of transportability, factory prefabricated manufacturing, short on-site assembly and start-up period and ability to work during extended time periods without fuel reloading represent promising direction of development in this area of nuclear power generation. It is specifically this type of low-power nuclear reactors brought to commercially competitive level that must form the basis of regional power generation in Russia. The objective of the present study is to develop the concept of low-power inherently safe thorium-fueled nuclear power installation based on the unified design of the fuel block.Scientific research and numerical experiments were performed using verified computer codes of the MCU-5 series, advanced libraries of evaluated nuclear data (ENDF/B-VII.0, JEFF-3.1.1, JENDL-4.0, ROSFOND, BROND, BNAB and others) and multi-group approximations.Analysis of information materials pertaining to the use of thorium as fuel element in rector facilities of the new generation and of its future potential was performed in the present study. Results of the first phase of neutronics studies of 3D model of high-temperatures gas-cooled reactor facility on the basis of unified design of the fuel block are presented. Calculation 3D model was developed using the software code of the MCU-5 series. Several optimal configurations of the reactor core were selected according to the results of comparison of neutronics characteristics of the examined options for the purpose of development of small-size modular nuclear power installations with power up to 60 MW. Results of calculations of reactivity margin of the reactor, neutron flux distribution and power density profiles are presented for the selected options of reactor core configuration.