Modeling of Forced and Natural Circulation Test Loop for a Pressurized Water Reactor using Thermal-Hydraulic Method

The present paper studies an Integral Test Facility (ITF) using the thermal-hydraulic method in order to determine its safety and reliability. The scaling and thermal-hydraulic design of the Islamic Azad University Test Loop (IAUTL) is conducted, accompanied by the energy scaling methodology for a typical pressurized water reactor (PWR). The scaling findings are authenticated by the experimental data resulting from the pressurized water reactor for pure water and forced circulation (FC) flow in ۱۰۰% of core nominal power using the RELAP۵/MOD۳.۲ code. From the comparison of the simulation results of core temperature in natural and forced flow, it can be concluded that the maximum outlet temperature of the reactor core in natural and forced flow is ۱۶۹.۴ °C and ۱۷۴ °C, respectively. This temperature increase in natural flow is due to the increase in the residence time of the fluid in the natural flow regime around the core. Also, the temperature drop in the heat exchanger in natural and forced flow is ۲۰ °C and ۱۳ °C, respectively. This increase in heat loss in natural flow is due to the lower velocity of the natural flow and, as a result, the heat exchange also increases in natural flow. The pressure drop between the inlet and outlet of the reactor core, in both free and forced flow, is approximately ۲ bar. The simulation results indicate that the IAUTL core can remove the residual heat from the core using NC flow.

کلیدواژه ها:

Modeling ، Forced and Natural Circulation ، Test Loop ، Pressurized Water Reactor ، RELAP۵/MOD۳.۲