Session: 07-02: Experiments and Analyses - I

Paper Number: 132452

132452 - Thermal Hydraulics Simulation of a Typical Pressurized Water Reactor Coolant System Using Cfd Method 

Abstract: 

The thermal hydraulic modeling of the reactor coolant system (RCS) is typically performed with lumped parameter codes such as RELAP. The efficiency gained by the coarse nodilization of this approach makes it feasible to predict the behavior of the entire RCS over extended periods of time. A limitation of this approach is a reliance on the coefficients predetermined from a set of experiments and is based on a network of 1-D volumes, which could have a high calculation error, especially in the local areas with high physical values gradients. Computational fluid dynamics (CFD) method provides a tool for predicting the RCS three-dimensional phenomena under a wide variety of conditions.

In this study, the thermal hydraulics simulation of the RCS for a typical three-loop pressurized water reactor was conducted based on numerical solution of Reynolds-averaged Navier-Stokes equations by commercial CFD software. This study aims to obtain the three-dimensional, global and localized flow features of the full reactor coolant system. The completed model of the RCS was build including reactor vessel and internals, core, steam generator, primary pump and linking pipes. The core and the steam generator tubes cannot be well represented within a practical mesh size, so that the porous mediums were used for modeling the head loss，and the power of the core heat sources and the heat sinks in the steam generator was simulated by a volumetric term source in the energy balance equation.

The loop flow and the temperature at different locations were compared with the measured values from the operating nuclear power plant in order to verify the accurate description of the developed CFD model. Then, several parameters such as temperature, velocity, and pressure have been investigated under normal steady operating condition with the full core thermal power and unblanced operating condition with one reactor coolant pump fails during cold shutdown state. The local thermal hydraulic features at the reactor vessel head dome, the thermal stratification in the reactor vessel upper plenum and the hot legs, and the swirling flow of the primary pump  are characterized under different conditions to give reference to safety operation.

A further improvement may be devoted to performing coupling between the current CFD model and the system codes in progressing toward the resolution of two-phase flow and the simulation of accidental transients. This simulation practice will ultimately lead to a reduction in the uncertainties of the RCS thermal hydraulics analysis and to identify some system-level three-dimensional phenomena.

Presenting Author: Mingqian Zhang China Nuclear Power Engineering Co., Ltd.

Presenting Author Biography: Thermal-Hydraulics and Safety Analysis, Nuclear Primary Equipment Design.

Authors:

Mingqian Zhang China Nuclear Power Engineering Co., Ltd.
Run Lin China Nuclear Power Design Co., Ltd. (Shenzhen)