Lu Guangyao, Ren Junsheng, Huang Wenyou, Xiang Wenyuan,

Zhang Chengang and Lv Yonghong

China Guangdong Nuclear Power Holding Co. Ltd.

China

1. Introduction

Two-phase flow exits in many nuclear power equipments, such as containment sump strainer, steam generator, steam turbine, control rod drive mechanism and so on. Experimental investigations are carried out to study the two-phase flow patterns and their transitions in these nuclear power equipments. And the results show that the two-phase flow patterns and their transitions are quite different from those in normal circular tubes.

For tube-bundle channel heat transfer enhancement technique, it has great advantages in high heat transfer efficiency and compact configuration without complex machining or additional surface processing, which has been successfully used in steam generator in nuclear power station and other industrial equipments. The characteristics of boiling flow and heat transfer have an important impact on these industrial equipments. It was found that heat transfer characteristics of fluid flowing in tube-bundle channels were different from those in circular tubes on account of the special geometric frame and different flow patterns in tube-bundle channels (Petigrew & Taylor, 1994). Boiling flow and heat transfer is a complex issue of two — phase flow, and many studies have been conducted in this research area.

Experiments of boiling flow in tube-bundle channels were carried out in order to simulate boiling flow and heat transfer in fuel module of nuclear reactor (Bergles, 1981). The experimental results showed that there were many different aspects of flow patterns and their transitions in tube-bundle channels. Grant & Chisolm (1979) and Ma (1992) made studies of air-water two-phase flow in tube-bundle heat exchangers, in which the vertical flow and horizontal flow experiments were conducted respectively. Grant & Chisolm (1979) found that there were mist flow, bubbly flow, intermittent flow, stratified-mist flow and stratified flow in tub-bundle channels. But Ma (1992) detected stratified flow and wave flow in horizontal channels, and bubbly flow and intermittent flow in vertical channels, which was different from the results gained by Grant & Chisolm (1979). Chan & Shoukri (1987) conducted visualization experiments of refrigerant-113 flowing in tube-bundle channels, of which the tubes were arranged 1×1, 3×1, 3×3, and 9×3 respectively. Sadatomi & Kawahara (2004) carried out experiments in tube-bundle channels, of which the tubes were arranged 2×3. On the basis of experiments, Sadatomi protracted flow pattern maps.

On the basis of the results obtained before, studies are carried out to investigate the characteristics of refrigerant 113 flowing in a tube-bundle channel, of which the tubes are

arranged 2×2. Furthermore, flow visualization experiments are carried out with high-speed camera. Through the comparison between the experimental results and other works, it is found flow regimes and their transitions in the tube-bundle channel are different from other normal circular tubes.

In the event of LOCA (Loss-of-Coolant Accident) or HELB (High Energy Line Break) within the containment of a light-water reactor (LWR), the primary safety concern regarding the long-term recirculation is that accident-generated debris and resident debris may be transported to the recirculation sump screens, which would result in adverse blockage effects and loss of the pumps net positive suction head (NPSH) of the emergency core cooling system (ECCS) and the containment spray system (CSS). The debris may starve the sump and the head loss of water flow through the containment sump strainers may be so large that it would exceed the available NPSH margin of pumps of ECCS and CSS systems. The pressure loss due to the accident-generated debris accumulated on the sump screens would be computed through the debris types and contents which are determined to be destroyed and transported. So the computational researches were carried out to investigate the characteristics of containment sump strainers by Lu et al. (2011a, 2011b, 2011c).

For control rod drive mechanism (CRDM) in nuclear power station, it works in high pressure and high temperature condition and single-phase water is adopted as the working liquid. But two-phase flow would come into being in the cold-state test and hot-state test. The high speed camera is also used in the CRDM visualization tests in cold-state to investigate the characteristics of two-phase flow.

2. Experiments