Investigation Of Subcooled Boiling Heat Transfer Characteristics And Void Fraction Calculating Method Of The Pulsed Reactor

In operation of the Xi’an Pulsed Reactor(XAPR), plenty of bubbles appeared at the top grid plate normally when the core is in high power operation mode; the bubbles and its moving have important effects on heat transfer of the natural convection. Based on the theory of the natural convection, the heat transfer characteristics and its probability are studied in the paper. Using the-oretical analysis, numerical simulation & experimental research methods, a comprehensive study about the bubble production mechanism, the void fraction calculation and the safety effect to the fuel elements are developed.Firstly, the heat transfer characteristics of subcooled boiling are studied in the paper. The conduction model fitting to the thick-rod fuel element of XAPR is built. Based on the mechanism of heat transfer, the conduction effect factors of the gas gap are researched, the calculating method of the gas gap conduction is constructed and the fuel element conduction model is also validated. Through the analysis of the natural convection between the clad and the coolant, the single-phase and two-phase flow heat transfer characteristics are investigated. A suit of heat transfer correla-tions in low pressure, low heat flux and less mass flow for XAPR are summarized. With the aid of the IPR-R1 reactor, the correctness of these correlations is testified. Based on the fuel element conduction model and the natural convection heat transfer of the clad and coolant, the pulsed re-actor’s fuel element heat transfer code is also accomplished.Cause of the complexity of the natural convection, the coolant mass flow and its distribution in the reactor core are difficult to confirm. With the single channel model, the hydraulic equations are constructed in the paper, the function of the reactor power and the coolant mass flow are also carried out. With the CFD software, the distribution and the flow characteristics of the coolant are also simulated. This job is of benefit to recognize the coolant micro flowage in the core.Aiming at the bubble phenomenon of the XAPR in high power operation mode, the produc-tion mechanism of the bubbles is investigated from the subcooled boiling heat transfer. In widely investigation, the void fraction calculating model of the natural convection heat transfer to the pulsed reactor core is presented. This model is based on the requirement of the XAPR, and it’s fit to the same two-phase flow analysis. Compared with the published experiment data, the correct-ness of the model obtained in the paper is verified directly.To summarize the research results of the paper, the steady state/transient calculation code is written with the subchannel method which is widely used in reactor thermal-hydraulic analysis. The thermal-hydraulic and safe parameters of XAPR are calculated with the subchannel code. With the transient function of the code, the dynamic heat transfer in pulsed operation is researched and the distribution of the time-space multidimensional safe parameters is also obtained. This job is a development to the primary single channel thermal hydraulic analysis. It has significance for safety analysis of the pulsed reactor, and it is also a beneficial reference in safety research in the same type reactors.To recognize the heat transfer characteristics in the coolant channel, and to know the varieties of the pressure field, the temperature field, the density field and the void fraction distribution, the CFD analysis model of the pulsed reactor are constructed in the paper. With ANSYS12 calculation platform, the important safety parameters of the XAPR coolant channel in steady state operation are simulated, the distribution of some three-dimensional safety parameters is obtained.In order to have a comprehensive realization of the core heat transfer characteristics, and to validate the correctness of the theoretical calculating method with experiment, the measurement experiments of the thermal hydraulic and safety parameters are carried out. The core’s temperature field at typical power and position is obtained. Based on the investigation of the measuring meth-ods of the void fraction, the local void fraction measuring method using the fiber probe is re-searched with emphasis, the possibility and the difficulty in measuring are discussed, and the ap-plication prospect of the fiber probe method is presented finally.The present dissertation constructs an integrated method in heat transfer analysis of the pulsed reactor core, and it also develops a suit of applicable calculation codes. Aming at the steady state and pulsed operation modes, a comprehensive and particular analysis is finished; furthermore some beneficial conclusions are presented. The works in the dissertation are of great significance for the reactor thermal-hydraulic analysis and safety evaluation. Furthermore these researches are of important academic value for the natural convection heat transfer and reactor numerical simula-tion.