摘要
基于斯特林技术的千瓦级热管反应堆实验(Kilowatt Reactor Using Stirling Technology,KRUSTY)开展了目前国内外唯一完成且公开发表的热管冷却反应堆带核实验,实验开展的工况包括冷态启动、负载变化、热管失效、反应性引入及热阱丧失事故等,这些工况下的带核实验数据对于热管堆瞬态分析程序的验证至关重要。本文自主研发了适用于热管反应堆的瞬态分析程序TAPIRS-D,采用KRUSTY带核实验数据对该程序展开了验证,对实验开展的冷态启动、负载变化、热管失效、反应性引入及热阱丧失事故进行了模拟计算和对比。验证结果表明:TAPIRS-D程序计算结果与实验数据符合较好,各工况下模型计算燃料温度的最大相对误差小于2%,整体功率的计算平均误差小于10%,预测的堆芯功率与温度瞬态响应与实验数据趋势符合一致,验证了TAPIRS-D程序的准确性与可靠性。本文研究可为其他新开发的热管堆热工安全分析程序的验证工作提供参考。
[Background]The kilowatt reactor using stirling technology(KRUSTY)is a heat-pipe-cooled reactor experimental system that uses a Stirling engine to convert thermal energy to electricity,it is the only one published experimental data for heat-pipe-cooled reactor systems.The KRUSTY experimental data under different working scenarios include the cold startup and load change processes,heat pipe failure,reactivity insertion,and heat sink loss.[Purpose]This study aims to validate the self-developed system transient analysis code named TAPIRS-D for the heat-pipe-cooled reactor concept using KRUSTY experimental data.[Methods]Firstly,an in-house system code for a heat-pipe-cooled reactor named TAPIRS-D was introduced,with the main theoretical module briefly explained,including the reactor power calculation module,heat transfer module for fuel assembly,and heat pipes.Then,the TAPIRS-D was applied for the first time to the simulation of the key processes of the KRUSTY prototypic reactor test under normal operation and accident conditions.Finally,comparison between the simulation data and experimental data was conducted for the validation of this analysis code.[Results]Comparison results demonstrate that the maximum relative prediction error for the fuel temperature is less than 2%,and the reactor power average prediction error is less than 10%.[Conclusions]The prediction trend of the numerical simulation by TAPIRS-D fits well with the experimental data on key parameters such as core power and the temperature of fuel and heat pipes,which indicates that TAPIRS-D is well developed and is capable of conducting safety analysis for heat pipe cooled reactor concepts.The validation of this system analysis code provides a good reference for other newly developed system codes for heat pipe reactors.
作者
吴攀
欧阳泽宇
朱煜
单建强
WU Pan;OUYANG Zeyu;ZHU Yu;SHAN Jianqiang(School of Nuclear Science and Technology,Xi'an Jiaotong University,Xi'an 710049,China)
出处
《核技术》
CAS
CSCD
北大核心
2023年第11期110-121,共12页
Nuclear Techniques
