Transient three-dimensional numerical simulation on asymmetrical flow and thermal characteristics of CEFR sodium pool under one secondary pump shaft stuck accident

Abstract

The China Experimental Fast Reactor (CEFR) primary circuit adopts a pool-type structure. The sticking of the main circulating pump shaft in the secondary circuit is an important design-based accident for CEFR, and it is necessary to investigate the three-dimensional thermal-hydraulic phenomena in the sodium pool under the accident. However, conducting relevant experiments directly is almost impossible due to the large dimensions and intricate flow paths within the reactor. In the present work, the “modular modeling and integrated coupling calculation” method is used to build the detailed model and calculate the primary circuit of CEFR in a 1:1 scale. The three-dimensional thermal-hydraulic phenomena inside the sodium pool under the accident condition of one secondary circuit pump shaft stuck are simulated. According to the calculation results, the backflow phenomenon of the fault loop pressure pipe and the intermediate heat exchanger (IHX) begins to appear around 50 s. An obvious thermal stratification phenomenon occurs in the hot pool at the beginning of the accident. As the accident progresses, the temperature difference between the hot and cold pools decreases gradually, and the thermal stratification of the hot pool gradually diminishes. The temperature difference between the inlet and outlet of the same IHX decreases from a maximum of 170 °Cto 2 °C. During the course of the accident, the peak temperatures calculated for the fuel assembly cladding and fuel pellet are 541.78 °C and 1040.17 °C, respectively, which is far below the designed temperature limit of 800 °C and 2800 °C, respectively. Moreover, comparative analysis is conducted on the simulation results of pump shaft stuck in the primary and secondary circuits. The results suggest that when the shaft of one secondary circuit pump is stuck, the average temperature of the sodium at the outlet of the core is higher and the temperature difference between the inlet of different loop pumps is greater, which places more substantial demands on the reactor component structure. It provides significant numerical reference for the safety design and assessment of CEFR.