Abstract
As one of the GEN-IV nuclear reactor systems proposed by GEN-IV International Forum, the sodium-cooled fast reactor features the energy generation with fast spectrum and a closed fuel cycle for fuel breeding and actinide management. In order to validate system codes for reactor system analyses and to improve code performance, benchmark activities of experimental transients of fast reactors were performed by worldwide institutions. The Shut-down Heat Removal Tests in Experimental Breeder Reactor II were proposed in the context of the International Atomic Energy Agency. In this work, loss-of-flow tests performed in Experimental Breeder Reactor II are analyzed using system thermal-hydraulic code CATHARE. Evolutions of important reactor parameters such as core power, coolant flow rate, temperatures in core and in the intermediate heat exchanger are predicted and compared against experimental data. The sodium pool is further modeled by using a 3D model in CATHARE-3 and its verification is also conducted in the benchmark. Qualitative agreement is obtained for the reactor parameters predicted by CATHARE. The inherent passive safety characteristics of Experimental Breeder Reactor II in unprotected loss-of-flow transient is demonstrated, and the stratification effects in sodium pool are able to be identified with the 3D model. The results obtained in this work are expectable to provide some valuable insights for future sodium-cooled fast reactor transient accident analyses by coupling system thermal-hydraulic codes and CFD tools.
