Validation of helical steam generator design for the experimental power reactor

The steam generator is an important component in any commercial nuclear power plants to deliver required steam from the reactor heat. Different with the steam generator design of PWR, other design in a small modular reactor (SMR) with smaller thermal capacity incorporates the helical-coiled tubes having advantage of additional heat transfer surface to deliver a superheated steam. The purpose of this research is to model the helical-coiled steam generator in the experimental power reactor developed by BATAN using RELAP5/SCDAP/Mod3.4 code as thermal hydraulic code to obtain a validation of technical data in the conceptual design document. After a number of steady-state simulations, operational thermal hydraulic parameters such as outlet helium and steam temperature, tube pressure drop, and heat capacity have been calculated to be close with the design data. It has been found out, that the different heat transfer characteristics in the helical-coiled tubes need to be addressed specifically in the RELAP5 input data by considering the new estimated heat transfer coefficient or by modifying the heat transfer area. In general, the developed steam generator model is ready to be integrated with the experimental power reactor model that can be used for safety anaysis purposes.