In order to enhance the safety of sodium-cooled fast reactors, a direct reactor auxiliary cooling system (DRACS) under natural circulation conditions with a dipped-type direct heat exchanger (D-DHX) in an upper plenum of the reactor vessel has been investigated. During the DRACS operation, the complicated thermal-hydraulic phenomena that cold coolant from D-DHX flowed into the fuel subassemblies and narrow gaps between them, which is well-known as inter-wrapper flow (IWF) was observed. Therefore, a multi-dimensional thermal-hydraulic analysis model in the reactor vessel for computational fluid dynamics (CFD) code (RV-CFD model) has been developed to evaluate the core cooling performance under natural circulation conditions. For the design study, the RV-CFD model is demanded to simulate with reasonable calculation costs while maintaining accuracy. In this paper, the application of the subchannel analysis method by CFD code for fuel subassemblies (subchannel CFD model) to the RV-CFD model was attempted. In the subchannel CFD model, the porous media approach was used to consider local geometry in the fuel subassembly, and the effective heat conductivity coefficients in diffusion term of the energy equation were set to fit the actual radial thermal diffusion between subchannels. Analysis results were compared to the experimental data obtained in the sodium experimental apparatus PLANDTL-1 and the calculated sodium temperature in the core had good agreement with the experimental result. It was confirmed that the RV-CFD model with subchannel CFD model was applicable to the core thermal-hydraulic analysis during the DRACS with the D-DHX operation under natural circulation conditions.
A plant dynamics analysis code named Super-COPD is being developed in Japan Atomic Energy Agency (JAEA) to offer a methodology for the design and safety assessments of future commercialized sodium-cooled fast reactors (SFRs). In this study, the friction loss coefficients in the whole core thermal-hydraulic model, which is based on flow network modeling, was modified to improve the prediction accuracy of the sodium temperature distribution in a fuel subassembly under the natural circulation conditions. Super-COPD with the modified whole core model was applied to analyses of experiments, that were performed by using JAEA's test facility PLANDTL and were simulated natural circulation decay heat removal operations in SFRs, as a part of the code validation study. The obtained numerical results of sodium temperature distributions in the core showed good agreement with the measured data. It implies that the modified whole core model can properly reproduce dominant thermal-hydraulic phenomena in the core region under natural circulation conditions, i.e., flow redistribution among fuel subassemblies as well as in a fuel subassembly and inter-subassembly heat transfer.
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