Study on applicability of fast reactor plant dynamics analysis code to core thermal hydraulics under natural circulation decay heat removal conditions

Abstract

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.