核热推进系统氢气物性及流动换热模型分析

为开展关于核热推进反应堆堆芯的稳态热工水力计算，基于现有针对压水堆的系统分析程序，添加了氢气的物性模型及流动换热和摩擦阻力关系式，并采用公开文献中的数据进行验证。结果表明采用上述模型计算得到的结果与参考值符合较好，二次开发的程序适用于氢气的流动换热计算。针对一种折流式核热推进反应堆堆芯，使用该系统程序建模并计算，得到了堆芯的流量、焓升等分布情况。研究结果表明，对于折流式核热推进反应堆，内外堆芯燃料元件之间的导热会增强堆芯释热不均，对堆芯的稳态热工水力特性有较大影响，堆芯物理方案的设计应结合热工水力方面的计算。本研究可为核热推进系统内氢气流动换热计算提供借鉴。

Analysis of Thermodynamic Property,Flow and Heat Transfer Model of Hydrogen in NTP System

Nuclear thermal propulsion (NTP) has been studied extensively in the last century. The NTP technology has the advantages of high specific impulse, large thrust levels and long lifetime. It is the most promising option for in space propulsion compared with the traditional chemical and electric propulsion technologies. Hydrogen is a typical kind of propellant of the NTP system. The thermal hydraulic calculation of hydrogen flowing through the NTP reactor core is very important for the design of the system. System analysis codes are mostly adopted for thermal hydraulic calculation of the nuclear reactors. A system analysis code was modified with thermodynamic properties of hydrogen, correlations of the heat transfer coefficient and friction factor. The hydrogen models implemented into the code were validated by the data in the references. The thermodynamic properties included specific volume, specific heat capacity, thermal conductivity and viscosity. Those models were validated by the data from the experiment carried out by Taylor. The heat transfer coefficient was calculated by the modified Miller Taylor correlation. And the friction factor was calculated by the Koo correlation. The two correlations were validated by the calculation results of the ELM code. The calculation results of either the modified system analysis code or the CFD code show good agreement with the reference values. It is validated that the models are reliable and the modified code is applicable of flow and heat transfer calculation of hydrogen. A two pass NTP reactor core was simulated by the modified code. The nuclear reactor core was divided into the outer core and the inner core in this design. The outer core used molybdenum as the substrate. The inner core used tungsten for high heat resistance. As a result, the two pass design could reduce the total mass of the reactor core while ensuring safety. The fuel elements and reflectors were modeled as parallel channels. Inlet temperature, inlet mass flow rate and outlet pressure were adopted for boundary conditions. Heat conduction between the fuel elements was especially considered during modeling. The distributions of the mass flow rate and enthalpy rise were obtained. The results of the fuel elements next to the boundary of the outer and inner cores deviate largely from the average values, although the power distribution is flattened well. It indicates that the heat conduction between the fuel elements of the outer and inner cores has influence on the steady state thermal hydraulic characteristics of the reactor core. For the innermost fuel elements of the outer core, the mass flow rate is fifteen percent below the average. For the fuel elements with low mass flow rate and high heat release, the material temperature tends to exceed the limit value. The heat conduction increases the unevenness of heat release of the reactor core. Furthermore, the flow maldistribution is harmful for the system stability and safety. In conclusion, the thermodynamic property, flow and heat transfer models of hydrogen were implemented into a system analysis code. Those models were validated by the reference data. The modified code was used to simulate a typical two pass NTP reactor core. Heat conduction between the fuel elements of the outer and inner cores has large influences. Thermal hydraulic calculation should be bonded to the nuclear physics design. This study provides a reference for the flow and heat transfer calculation of hydrogen in the NTP system.