Abstract
The reliability of steam generator is extremely important for the sodium-cooled fast reactor nuclear power plant safety and stable operation. The convective heat transfer mechanism of the once-through steam generator (OTSG) of China Experimental Fast Reactor (CEFR) was researched. The water/steam side was divided into four areas according to the heat flux and steam quality, named subcooled, nucleate-boiling, film-boiling, and superheater. In order to accurate determine the DNB, the CHF table was used in this paper. Based on the homogeneous flow model and fixed boundary method, a thermal-hydraulic simulation system, which named OTAC, was established in this paper. To evaluate its performance, the predictions of this method were compared with PSM-W code. The maximum difference between the temperatures predicted by this model and PSM-W was ∼5K. The calculated results are consistent with the actual experiment data, which indicates the correctness of the mathematical model and simulation method. Static and dynamic characteristic researches of CEFR OTSG have done in the simulation system. And the system can be used to simulate the OTSG dynamic in real-time.
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