A three-dimensional hot pool analysis code with SIMPLE algorithm was developed based on staggered grid, which can predict thermal-hydraulic characteristic in hot pool of pool-type liquid metal fast breeder reactor (LMFBR) under Cartesian and cylindrical coordinate systems. After being incorporated into system analysis code for pool-type fast reactor in China (SAC-CFR), the coupled code was used to analyze the thermal-hydraulic behavior in the upper plenum of fast breeder reactor “MONJU” during reactor scram transient. A basic agreement was obtained, which means the present model is effective. And then the coupled code with newly developed porous medium model was used to analyze the flow field in China Experimental Fast Reactor hot pool under steady-state operation condition. The distribution characteristic of flow field in hot pool showed the effectiveness of porous medium model, which formed preparations for further development of passive residual heat removal system in-vessel.
The sodium in pool-type liquid metal fast reactor provides vital function of removing reactor generated heat. Accurate prediction of coolant thermal-hydraulic characteristic in hot pool cannot only evaluate the performance of key component in hot pool properly, but also improve the system analysis ability through more accurate intermediate heat exchanger (IHX) inlet temperature. In advanced FBR, the directly in-vessel decay heat remove system (DIDHRS) is being considered to be used to improve inherent safety, which makes it impossible to analyze the complicated phenomenon caused by DIDHRS with traditional 1D or 2D system analysis code. Therefore, the coupling of system code and 3D thermal-hydraulic analysis code is always an important challenge to take into account local 3D effects on global system behavior. Considering that the hot pool is arranged with more complex key components and filled with higher-temperature coolant than cold pool, 3D thermal-hydraulic characteristic in hot pool is a research focus in this study.
Many researches on thermal-hydraulic characteristic in hot pool have been carried out currently in some countries and organizations. In the aspect of experiment, many kinds of in and out of core experiment have been performed for their specified research purpose in America, France, Japan, Russia, etc. (Hofmann and Essig, 1993, Kasinathan, 1993 and Ieda, 1993). In the aspect of theory, besides the commercial computational fluid dynamics software such as CFX, FLUENT, many three-dimensional thermal-hydraulic analysis codes have been specially developed for evaluating the thermal-hydraulic performance in fast reactor, such as COMMIX (Chien et al., 1993) series developed by ANL, AQUA (Muramastsu et al., 1987) developed in PNC, TRIO_U (Tenchine et al., 2012)in France, FASTOR-3D (Degui et al., 1998) and DHRSC (Yishao et al., 1991) in China. During transient analysis, all these analysis codes can only predict the thermal-hydraulic performance in particular component such as hot pool at a certain boundary condition. As to the coupling of system code and 3D analysis code, many researches have been performed by organizations and researchers. For instance in Europe, the Trio_U has been coupled to CATHARE by NURISP (Emonot et al., 2012) and THINS (Xu et al., 2010) teams, and the related application, verification and validation are in progress. Also, the coupling used for PWR is also a research focus in recent years. In Japan, two-dimensional upper plenum model was incorporated into a one-dimensional system analysis code named SSC-L (Mochizuki, 2007 and Mochizuki, 2010).
From the aspect of fast reactor sustainable development in China, it is the only choice to develop system analysis code and 3D analysis code specialized for CEFR and next generation demonstration fast reactor.
In the present study, a three-dimensional hot pool analysis code with SIMPLE algorithm was developed based on staggered grid under Cartesian and cylindrical coordinate systems. After being incorporated into System Analysis Code for China Fast Reactor (SAC-CFR), the coupled code was used to analyze the thermal-hydraulic characteristic in upper plenum of MONJU during the scram transient starting from 45% thermal power operation. And then the coupled code with newly incorporated porous medium model was used to analyze the flow field in China Experimental Fast Reactor (CEFR) hot pool under steady-state operation condition.
A three-dimensional hot pool analysis code with SIMPLE algorithm was developed based on staggered grid, which can predict thermal-hydraulic characteristic in hot pool of pool-type LMFBR under Cartesian and cylindrical coordinate systems. And the coupling between newly developed 3-D hot pool model and system analysis code was finished.
After being incorporated into System Analysis Code for China Fast Reactor (SAC-CFR), the coupled code was used to analyze the thermal-hydraulic behavior in the upper plenum of fast breeder reactor “MONJU” during reactor scram test. A basic agreement between the computational results and experimental data was demonstrated, which means the present model is effective. Also, from analytical results, some characteristics of thermal stratification in the upper plenum were also discovered.
After incorporating porous medium model into hot pool analysis model, the newly developed code was used to analyze flow field in China Experimental Fast Reactor (CEFR) hot pool under 75% power steady-state operation condition. The distribution characteristic of flow field in hot pool showed the effectiveness of porous medium model, SAC-CFR with three-dimensional hot pool analysis model is now ready to simulate passive residual heat removal under natural convection for CEFR after incorporating interassembly model.
