Comparison Between Experimental Data and Numerical Modeling for the Natural Circulation Phenomenon
The study of natural circulation phenomenon has been object of crescent interest in scientific community in recent years. The new generation of compact nuclear reactors uses natural circulation of the fluid as a residual heat removal cooling system in case of accident or shutdown. The objective of this paper is to present a comparison between experimental data and numerical simulation results for the natural circulation phenomenon in single and two-phase flow regime. Experimental data were obtained from a circuit built with glass tubes, through which thermal hydraulic phenomenon visualization was possible. The circuit is composed of an electric heater as the heat source, a heat exchanger as the heat sink and an expansion tank to accommodate fluid density changes. Instrumentation data acquisition is performed through thermocouples and pressure meters, and controlled by a computer interface developed using LABVIEW. Numerical modeling and simulations were done with the thermal hydraulic code RELAP5, which is widely used for this purpose. The cyclic reverse flow observed in the circuit was well represented by the numerical model. The comparison demonstrated that the numerical simulations are very close to the experimental data.
Simulation of Operational Conditions of Fassip-02 Natural Circulation Cooling System Experimental Loop
The natural circulation is considered in the design of emergency passive core cooling system in a nuclear power plant. In that context, in order to investigate the characteristics of the natural circulation, FASSIP-02 experimental loop is designed. This paper simulates the various operational conditions with different condition of the heater power, the pipe insulation and the expansion tank's valve using RELAP5 code. The objective is to obtain the best operational conditions of FASSIP-02 once it is built. The simulation results show that the until 50,000 s the steady state condition could not be achieved yet when the heater power greater than 10 kW. The pipe insulation reduced the heat loss to the environment and in turn it causes faster increase of the water temperature inside the pipe. While, if the expansion tank's valve is closed during the operation, the pressure inside the loop would increase, faster when the heater power is higher. It is concluded that in all cases to avoid the saturation condition, the heater power should be maintained lower than 10 kW, especially when the loop pipe is insulated.
RELAP5/3.2 assessment against low pressure onset of flow instability in parallel heated channels
Abstract Best Estimate computer codes have been, so far, developed for safety analysis of nuclear power plants and were extensively validated against a large set of separate effects and integral test facilities experimental data relevant to such kind of reactors. Their application to research reactors is not fully straightforward. Modelling problems generally emerge when applying existing models to low pressure and more particularly to subcooled flow boiling situations. The objective of the present work is to investigate the RELAP5/3.2 system code capabilities in predicting phenomena that could be encountered under abnormal research reactor’s operating conditions. For this purpose, the separate effect related to the static onset of flow instability is investigated. The cases considered herein are the flow excursion tests performed at the Oak Ridge National Laboratory thermal hydraulic test loop (THTL) as well as some representative Whittle and Forgan (W & F) experiments. The simulation results are presented and the capabilities of RELAP5/Mod 3.2 in predicting this critical phenomenon are discussed.