Scaling Analysis Of Steam Generator During Natural Circulation Phase Under Station Blackout Accident

Advanced pressurized water reactor nuclear power plant in the event of station blackout accident, after the reactor shutdown, the reactor core decay heat could transfer to the second circuit by the establishing of natural circulation between core and steam generator (SG). The heat-carrying capacity of the steam generator which as the primary loop heat trap is an important factor affecting the reactor safety. In order to verify the advanced pressurized water reactor nuclear power technology's strategy to response to station blackout accident, needing to carry out the engineering verification test. In order to ensure that important phenomenon between prototype plants and test is similar, needing to carry out the scaling analysis. In this paper, take the natural circulation phase of the main circuit for the nuclear power plant as research objects, carrying out the scaling analysis using the H2TS method. Focusing on the thermal hydraulic behaviors on the steam generator, obtaining the similar criterion between the prototype SG and test SG under the station blackout accident, giving the distortion evaluation. Provide the theoretical support for the test design and nuclear safety review.