应用于反应堆热工水力程序的核态沸腾传热关系式评价

Assessment of Nucleate Boiling Correlations Applied to Thermal-hydraulic Code of Reactor

本文以反应堆热工水力分析程序COSINE开发为背景,针对燃料棒和冷却剂换热及压力容器外部冷却时的核态沸腾两种特殊的工况,研究常用于计算热工水力程序的核态沸腾传热关系式的计算结果随影响参数的变化关系,比较不同范围内各关系式计算结果的差异程度和敏感性,为程序中用户选项的设置和进一步实验验证提供参考意见,研究表明高过热度工况最需进行实验验证,反应堆热工水力分析程序计算这两种工况下的核态沸腾传热更适宜选用Chen、Schrock-Grossman1、Wright和SchrockGrossman2公式。

This article is set in development of reactor thermal-hydraulic analysis code- COSINE, studies variation characteristic of results of nucleate boiling correlations usually used for thermal-hydraulic analysis codes with variable changing, compares difference degree of different correlations and sensitivity in different ranges based on fuel rod wall heat transfer and external reactor vessel coolant conditions, in order to provide advice on setting of code user options and further experiment researches. It draws a conclusion that the condition of high superheat degree is mostly needed experimental demonstration and Chen, Schrock-Grossmanl, Wright and Schrock-Grossman2 correlations are more suitable for calculating these conditions in reactor thermal-hydraulic analysis code.