A pressure wave propagation across the second loop of a sodium-cooled fast reactor may lead to severe damage to the pipes and the equipment due to a large leakage sodium-water reaction accident. Therefore, the pressure source and the pressure wave propagation calculation and analysis can be significant for a sodium-cooled fast reactor’s design and operation.
A mathematical model with code was built to calculate and predict both the pressure source and the pressure wave propagation after the large leak sodium-water reaction accident occurred in an SFR steam generator.
In the pressure source model, the sodium-water reaction produces hydrogen and squeezes sodium nearby. It is assumed that the sodium-water reaction carries out instantaneously, and a spherical-to-column bubble is assumed to grow up at the reaction zone. Under such assumptions, the pressure source model can obtain the hydrogen bubble’s volume and pressure in the steam generator.
In the pressure wave propagation model, flow is assumed to be one-dimensional. With the basic theory of mass conservation, momentum conservation, and the Isentropic relation of the working fluid, the numerical solution is calculated by a grid divided by the method of characteristics, considering both precision and efficiency.
However, plenty of parameters, both of the pressure source and the parallel channels, can affect the accident in different ways and scales. Considering that to practice a large leakage sodium-water reaction experiment can be somewhat difficult because of its danger, inconvenience, and high cost, it is quite essential to conduct a sensitivity analysis through the program for the purpose of facilitating the design of an SFR.
Since most sodium-cooled fast reactor designs are multimodules, a parallel channel structure with multiple pressure boundaries is modeled for sensitivity analysis. Sensitivity analysis was performed on multiple sets of important parameters, including different hydrogen bubble properties, different pressure boundaries, and different equipment types. Following the sensitivity study, the code can be used for the accident analysis in a future commercial demonstration fast reactor by indicating various parameter guidance for the design and construction.