Neutronic analysis of mixed thorium-uranium fuel bundle for CANDU reactors

International Conference on the Sustainable Energy and Environmental Development

生态环境科学

Oettingen, M.^1 ; Skolik, K.^1 ; Malicki, M.^1

AGH University of Science and Technology, Mickiewicza Av. 30, Krakow

30-059, Poland^1

关键词: Continuous energy;    Grid infrastructures;    Heterogeneous geometry;    Neutron multiplication;    Neutron transport;    Neutronic analysis;    Science and Technology;    Trajectory analysis;   

Others  :  https://iopscience.iop.org/article/10.1088/1755-1315/214/1/012024/pdf
DOI  :  10.1088/1755-1315/214/1/012024

学科分类：环境科学（综合）

来源: IOP

【 摘 要 】

The paper shows neutronic analysis of the CANDU (CANada Deuterium Uranium) nuclear reactor fuel channel with mixed thorium-uranium fuel bundles. The numerical model of the fuel channel was designed using The Monte Carlo Continuous Energy Burn-up Code - MCB developed at the AGH University of Science and Technology, Faculty of Energy and Fuels, Department of Nuclear Energy. The super-computer Prometheus available in the frame of the Pl-Grid Infrastructure at the Academic Computer Centre Cyfronet AGH was used for multi-scale calculations. The fuel bundles are composed of two clusters of fuel rods. The neutronic analysis considers detailed numerical simulation of neutron transport in fully heterogeneous geometry of the fuel channel. Moreover, burnup simulations were performed using Transmutation Trajectory Analysis method implemented in the MCB code. In the analysis we mainly consider time evolutions of neutron multiplication factor, fissile 233U, 235U and 239Pu and fertile 238U and 232Th. The simulations were performed for eight scenarios with various fuel composition.