Evaluation of a SodiumeWater Reaction Event Caused by Steam Generator Tubes Break in the Prototype Generation IV Sodium-cooled Fast Reactor

The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by theKorea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant totransfer the core heat energy to the turbine. Sodium has chemical characteristics thatallow it to violently react with materials such as a water or steam. When a sodiumewaterreaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressurewaves and corrosive reaction products are produced, which threaten the structuralintegrity of the components of the intermediate heat-transfer system (IHTS) and the safetyof the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in thedesign-basis event. This event should be analyzed from the viewpoint of the integrities ofthe IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences ofthe SWR, the behaviors of the generated high-pressure waves are analyzed at the majorpositions of a failed IHTS loop using a sodiumewater advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptancecriteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated usingthe multidimensional analysis of reactor safety-liquid metal reactor code to model thewhole plant. KCI Citation Count: 4