Abstract
The two-phase flow phenomenon can be encountered in whether nuclear reactor normal running condition or emerged accidents, and imposes significant influences on reactor safety. One of the essential characteristics of a two-phase flow is the interfacial drag, which can influence the two-phase interface’s deformation, the distribution of the two phases, the flow pressure drop, etc. In this paper, the interfacial drag models in RELAP5 are assessed using GE mixture level swell experiments and FRIGG tests. GE Small Level Swell experiment and GE Large Level Swell experiment were respectively based on small and large experimental sections to study critical flow, void fraction distribution, two-phase level, and other characteristics under reactor blowdown conditions. FRIGG tests, using steam and water as the two-phase working fluid, studied the steady-state distribution characteristics of axial void fraction in the rod bundle channel under different conditions. They are separate effects tests and used to assess the interfacial drag models for many system analysis codes. For the GE Small Level Swell experiment, the results indicate that the RELAP5’s prediction of the void fraction at the middle and top part of the test section are in good agreement. However, at the bottom of the test section, the prediction is much higher than the experimental data and beyond the experimental data’s error bar. For FRIGG tests, the most simulation results of RELAP5 are similar to the simulation results of TRACE. For high flow rate, high-level power, and large inlet subcooling operation conditions, the prediction of the low void fraction region (below 0.3) and the high void fraction region (beyond 0.6) will have some differences comparing with the experimental data. All the differences between simulation results and experimental results can be attributed to the adopted interfacial drag models’ applicability in RELAP5.
