Integrated Fast Reactor Analysis and Calculation of Neutronics-thermal Coupling

Integrated fast reactor exhibits distinct advantages in enhancing the utilization rate of uranium resources and minimizing high-level radioactive waste. Through multi-physical coupling and refined computations, conducting high-precision numerical simulations of the reactor, and reducing various simplifications and approximations in traditional calculation methods, the actual operation of the reactor can be better reflected, providing verification for the reactor design scheme. This research undertook core nuclear-thermal coupling studies based on the Monte Carlo program RMC and the subchannel analysis program VERSA. With the integrated fast reactor as the research object, the active zone was equally divided into 20 nodes in the axial direction, with one layer each in the upper and lower axial conversion zones, and a total of 22 layers in the axial direction. In the nuclear-thermal coupling calculation, the initial temperature and density distributions were first set, and neutron transport calculations were carried out to obtain power distribution information, which was input into the subchannel program to conduct flow and heat transfer calculations and obtain new coolant temperature, density, and fuel rod temperature parameters. These thermal parameters were fed back to the corresponding lattice cells in the neutron physics geometric module. When conducting the next round of physical calculations, the material parameter settings were updated to calculate the new power distribution. The Doppler effect is relatively prominent in the thermal feedback. The central temperature of the fuel element affects the power distribution calculation by influencing the absorption cross-section. In this study, by creating multiple temperature libraries and adjusting the material cross-section data of the Monte Carlo calculation based on the temperature values from the thermal feedback, the nearest temperature library is assigned to each region based on the proximity principle whenever the thermal temperature is fed back to the physics, ensuring the efficiency and accuracy of the coupled calculation. After three rounds of iterative calculations, the change in the fuel center temperature is within 1%, achieving convergence. Compared with the calculation using the 900 K cross-section library, the calculated overall fuel center temperature decreases when using the interpolation cross-section library. In this research, to enhance the efficiency and accuracy of information transfer during the nuclear-thermal coupling iterative calculation, through the full core physical-thermal-hydraulic nuclear-thermal coupling interface program, the automatic identification, extraction, mapping, and conversion of power distribution and temperature field information in the core physical and thermal-hydraulic calculation results are achieved. This paper conducts a comprehensive and accurate three-dimensional refined nuclear-thermal coupling calculation study of the full core, obtaining comprehensive and accurate core power and temperature distributions. This provides the possibility for further optimizing the core flow distribution and improving economic efficiency, as well as providing a basis for accident prevention and safety evaluation.