基于RELAP5 MOD3.2的钠冷快堆热工水力系统分析程序开发及验证

对大型核反应堆热工水力分析程序RELAP5 MOD3.2进行了改造,使之适用于钠冷快堆系统安全分析。在不影响原程序功能的基础上添加了气液两相钠物性和液态金属对流换热模型,并改造了相应的初始化模块和计算模块。改造后的程序可正确模拟钠的流体力学特性和热物性,搭建钠冷快堆热工水力流体网络进行分析计算。对EBR-Ⅱ试验堆基准题进行了稳态模拟和失流事故分析,其中稳态计算主要参数与实验值相对偏差小于1%,瞬态计算相对偏差小于10%,各参数变化趋势与实验值相符良好,初步验证了改造程序的可靠性。

Development and Verification of Thermal-hydraulic System Analysis Code Based on RELAP5 MOD3.2 for SFRs

The general-purpose reactor thermal-hydraulic analysis code RELAP5 MOD3.2 was modified for system analysis of sodium-cooled fast reactor (SFR).The thermodynamic and transport properties of sodium liquid and vapor were implemented into the RELAP5 MOD3.2 code,as well as the specific heat transfer correlations for liquid metal.The methods of code modifications are universal for other working fluids and will not affect the code original performance.The benchmark on loss of flow accident transient of Experimental Breeder Reactor Ⅱ (EBR-Ⅱ) for Argonne National Laboratory (ANL) was analyzed to verify the modified code.The results show that the reactor can shut down automatically relying on inherent negative feedbacks and major thermal-hydraulic parameters are lower than the safe limits.Moreover,the calculation results are consistent with the experimental data and other SFR system analysis code results.The work in the paper can demonstrate the capability and reliability of the modified RELAP5 for the analysis of SFRs further.