Neutron photon transport-thermal hydraulic coupling analysis of high-flux lead-bismuth reactor

BackgroundThe high-flux reactor is an important nuclear facility for research and application, and has broad application prospects. Lead-bismuth alloy has high melting point and boiling point, weak moderation capability, hard energy spectrum and small neutron absorption cross section, demonstrating inherent potential for engineering implementation as a high-flux reactor system. However, the increase of neutron flux in high-flux lead-bismuth reactor is primarily constrained by the need for co-optimization of fuel assembly power density with heat load capacity. And there is obvious gamma heating effect in the reactor, which puts forward higher requirements for the calculation and analysis accuracy of core physical and thermal characteristics.PurposeThis study aims to develop a neutron photon transport and thermal-hydraulics coupled analysis code for high-precision modeling and analysis of high-flux lead-bismuth cooled reactors.MethodsThe fixed-point iteration method combined with the relaxation factor coupling Monte Carlo program OpenMC and the subchannel analysis code SubChanFlow were applied to the neutron photon transport-thermal hydraulic coupling scheme. With combination of the whole-core energy deposition model and the display heat source distribution method, the physical-thermal coupling program was constructed in this scheme. Based on the interface developed by python, the grid and mapping relationship were automatically built in the coupling code, and the convergence of the coupling process was controlled by the second-order norm. Finally, the VERA benchmark was used to verify the coupling code, and the high-flux lead-bismuth reactor core performance analysis was carried out based on the neutron photon transport-thermal hydraulic coupling code.ResultsCalculation results show that there is an obvious gamma heating effects in the non-fuel region of the high-flux lead-bismuth reactor, which significantly changes the neutron flux and power distribution in each region of the core. The neutron flux is decreased significantly, with a maximum decrease of 5.9×1014 n·cm-2·s-1, a decrease of 5.9%, while the coolant power in the active zone is increased by 4.15%, the relative power factor of the central assembly and the hottest rod is decreased by 0.05, and the core life is increased from 90 d to 100 d, which deepened the core burnup and the maximum difference in reactivity reached 161 pcm. If only the neutron transport is considered, a large error will be caused in the results of temperature and neutron flux density, and the error increases with the increase of power and neutron flux density.ConclusionsCompared with the neutron transport-thermal-hydraulic coupling model, the neutron photon transport-thermal-hydraulic coupling model proposed in this study demonstrates superior accuracy in analyzing high-flux reactor core performance, improving the accuracy of parameters such as core power distribution and neutron flux. It significantly changes the core power and neutron flux distribution, and deepens the core burnup.