Research On Thermal-hydraulic Constitutive Models And Subchannel Analysis Code Improvement

The plate-shaped fuel element has good heat transfer characteristics,high average power density of the core,and low temperature of the fuel core,which is conducive to improving the power-to-volume ratio of the core and ensuring the safety of the core.Therefore,plate fuel is widely used in compact reactors such as research reactors,integrated reactors,and high flux reactors.At present,most thermal-hydraulic analysis programs,such as RELAP,RETRAN,THEATRE,etc.,are mostly developed for large pressurized water reactors using rod-shaped fuel,while the development of simulation programs for plate-shaped fuel element reactors is not perfect.Based on this,according to the structure and operating characteristics of the plate fuel reactor,this paper uses C++ language to compile the thermal-hydraulic constitutive relationship program of the reactor suitable for the rectangular channel of the plate fuel element through the establishment of a reasonable mathematical and physical model,and the developed The program was verified.The main tasks completed by the thesis are:(1)Established a complete set of thermal-hydraulic constitutive relationship models:including flow resistance coefficient calculation model and heat transfer coefficient calculation model.The resistance coefficient model is divided into single-phase and two-phase,and the heat transfer coefficient model is divided into single-phase.Water,subcooled boiling,saturated boiling,membrane boiling and single-phase steam;CHF model,cavitation share calculation model;two-phase flow pattern transition model,two-phase interface resistance,heat transfer coefficient calculation model,etc.(2)IAEA 10 MW Material Test Reactor(MTR)benchmark questions were used to verify the developed thermal-hydraulic constitutive relationship program.The transient conditions of two typical accident conditions defined by the benchmark problem,Reactivity Introduction(RIA)and Loss of Flow Accident(LOFA)are calculated and analyzed.In a reactive introduction accident,the process of a fast-reaction introduction and a slow-reaction introduction accident are basically the same.The difference is that the peak power and maximum fuel temperature of the reactor core when the reactivity is introduced into the accident is much higher than that of the slow-reactivity introduction accident;In a fast reactivity introduction accident,at the same reactivity introduction rate,the time for the low-enriched uranium core to reach peak power is earlier than that of the high-enriched uranium core.This is because the prompt neutron generation time of the low-enriched uranium core is It is shorter than high-enriched uranium;in addition,in the rapid reaction introduction accident,both lowenriched uranium and high-enriched uranium cores have appeared supercooled boiling phenomenon.In a stall accident,the process of fast stall and slow stall accidents is basically the same.The difference is that in a fast stall accident,the cooling capacity decreases faster,and the core temperature rises more under the action of decay heat.fast.The above results are in good agreement with most of the calculation results in the literature,so as to verify the correctness of the program.