In this paper the thermal-hydraulic characteristics of the primary loop of the Experimental Breeder Reactor (EBR-II), including the temperature and the flow characteristics of the core, the intermediate heat exchanger (IHX) and the experiment subassembly XX09 and XX10, were analyzed with the transient thermal-hydraulic code THACS. The THACS code contains the core, the pumps, IHX, the sodium pool and some other modules, and each module could operate separately. All of the primary–loop components are simulated one-dimensional, and in the core calculation the incompressible model for the single phase. The multiple-channel model is applied to simulate the core subassemblies, including the average, hot, XX09, XX10, the reflector and the blanket channels. The neutron physics is calculated with the point reactor kinetics, and the reactivity feedbacks caused by the Doppler effect, coolant density, axial expansion of fuel rods and radial expansion of core are considered. Two tests, namely the SHRT-17 and SHRT-45R tests, are simulated to validate our tools and models. The THACS simulation results show that the EBR-II type sodium cooled fast reactor could shut down automatically relying on inherent negative feedbacks in the two tests.
Validation of THACS for Sodium Cooled Fast Reactor Based on Benchmark Analysis of EBR-II
- Views Icon Views
- Share Icon Share
- Search Site
Nina, Y, Zaiyong, M, Benxue, H, Qiu, S, & Su, G. "Validation of THACS for Sodium Cooled Fast Reactor Based on Benchmark Analysis of EBR-II." Proceedings of the 2014 22nd International Conference on Nuclear Engineering. Volume 2A: Thermal Hydraulics. Prague, Czech Republic. July 7–11, 2014. V02AT09A064. ASME. https://doi.org/10.1115/ICONE22-30485
Download citation file: