analysis code

water reactor

blanket module

nuclear fuel

fast breeder

reactivity insertion

heat removal

thermal hydraulic

reactor core

Progress in Nuclear Energy

The evaluation of thermal-hydraulics of core under decay heat removal conditions is essential to the safety evaluation of a sodium-cooled fast reactor. The inter-wrapper flow (IWF) has an influence on the thermal-hydraulics of core. However, the study of the flow and heat transfer of IWF was limited. In this paper, a 2D layered IWF model was developed, and some tests were simulated to validate the IWF model, including heat removal tests SHRT-17 and SHRT-45R conducted in the experimental fast reactor EBR-II and a natural circulation test performed during the PHENIX end-of-life experiment. In order to simulate all the components of the primary system, the IWF model is coupled with the Transient Thermal-Hydraulic Analysis Code for Sodium-cooled fast reactors (THACS). In the simulation without IWF model, the predicted peak temperature of the instrumented subassembly XX10 in EBR-II is about 150 K lower than test data, and the predicted average outlet temperature of reactor core in PHENIX is about 20 K higher than test data. While the predictions of THACS with IWF model agree well with the test data. The results show that the IWF can be accurately simulated by the IWF model, and the IWF model improves the accuracy of the simulations of the reactor core. Further, some sensitivity analyses were conducted to provide better reference for the study of sodium-cooled fast reactors.

The development and validation of the inter-wrapper flow model in sodium-cooled fast reactors