基于FLUENT的核热耦合程序反应性反馈参数敏感性

Sensitivity analysis of feedback factors by neutronics and thermal-hydraulics coupling code based on FLUENT software

将计算流体力学模型与中子动力学模型耦合来进行反应堆瞬态安全分析的方法,由于可以开展复杂几何结构的三维流动传热分析,因此受到很大的关注。基于FLUENT用户自定义功能(UDF)开发了一套可用于池式铅堆瞬态安全分析的核热耦合程序,程序耦合了临界/次临界点堆中子动力学模型和燃料棒模型。由于反应堆处于不同寿期时,随着燃料燃耗、可燃毒物积累等因素导致反应性反馈系数有较大变化,因此使用开发的核热耦合程序对中国科学技术大学提出的小型自然循环铅冷快堆进行不同关键反馈系数下无保护的瞬态超功率事故安全分析。调整点堆模块考虑到的四个反应性反馈系数,可以发现燃料多普勒系数对堆安全的影响最大,同时定量的分析结果表明超功率事故引入时间长短对事故演化有重要影响。

With the great improvement of computer performance, analyzing the complex flow and heat transfer phenomena under transient condition by coupling CFD and neutronics has attracted lots of attention nowadays. We developed a neutronics- thermal hydraulics coupled code for transient analysis of pool type lead cooled fast reactors based on FLUENT UDF. The coupled code FLUENT/PK was validated by critical and sub-critical reactor under unprotected transient condition. The validation results show the correctness and feasibility of CFD method in safety analysis of reactors. The coupled code was used to analyze the lead cooled fast reactor SNCLFR-100 （which had been proposed by USTC） under different transient conditions. The calculated results show that there exists great difference if the introducing time of reactivity is different. The sensitivities of the feedback factors were analysed for the reason that these factors will be quite different at different operation time of the reactor core, and the result shows that Doppler constant has the most important effect on reactor safety compared with coolant temperature coefficient, fuel rod axial expansion coefficient and cladding expansion coefficient.