抄録
Parallel channel system is very typical in nuclear power plant, such as in the core and the steam generator. In this paper the author developed one model and a set of codes for the thermal hydraulic transient analysis of a core parallel channel system in QinShan II nuclear power plant. The basic model was built by Clausse and Lahey (1990) and developed by Lee and Pan (1999). In the original model the channel is divided into three parts: entrance section, heater section and riser section. The integral method is used to calculate the pressure of each section. It can reflect the whole characteristics of such system except the local parameters. And the nonuniform heating is very hard to be considered in this model. And the code based on this model is very hard to be coupled with reactor physics code in the future. Hence, the control volume concept is introduced to overcome these weaknesses. The channel can be divided into sufficient control volumes to describe the details. A development model is built based on this concept. Many classical flow and heat transfer correlations and models are included. The Gear method is chosen as the solving implement of the differential equations. Then a code is compiled. And then the transient characteristics are obtained under single and two-phase conditions. The transients of blocking, reactivity insertion and loss of flow are investigated based on imaginary data. At last a coupled method is conceived.
