摘要
FFTF LOFWOS Test#13是美国FFTF钠冷回路式快堆进行的无保护失流试验,目的是为了证明反应堆的固有安全特性。本文采用中国原子能科学研究院自主开发的FR-Sdaso程序参加了IAEA策划发起的基于该试验的基准例题项目。利用FR-Sdaso程序将一回路主泵转速、二回路流量和空气热交换器出口钠温作为边界条件,建立了FFTF基准例题模拟模型。计算结果与FFTF LOFWOS Test#13试验结果对比分析表明,FR-Sdaso程序能较好地预测无保护失流事故后反应堆功率以及一、二回路温度和流量的瞬态变化,自然循环阶段反应堆衰变功率计算值与试验值的最大相对偏差为−7.1%,一回路3个环路自然循环流量与初始稳态值的最大相对偏差为0.65%。对于第2排和第6排PIOTA组件,由于模拟中未考虑瞬态过程中堆芯功率分布变化和组件之间的传热,出口温度的计算值较试验测量值最大偏高25.5℃,计算结果更保守。FR-Sdaso程序对FFTF LOFWOS Test#13基准例题的分析初步验证了程序堆芯和一、二回路热工水力模型的正确性。
FFTF LOFWOS Test#13 is an unprotected loss of flow test conducted on the FFTF sodiumcooled loop fast reactor in the United States,aimed at demonstrating the inherent safety characteristics of the reactor.To verify advanced neutron science,thermal-hydraulic,and safety analysis programs for sodium-cooled fast reactors and improve simulation and analysis capabilities in the field of sodiumcooled fast reactors,IAEA initiated the FFTF LOFWOS Test#13 benchmark in 2017.To further verify the system program FR-Sdaso independently developed by China Institute of Atomic Energy,the code development team participated in the benchmark project.A FFTF benchmark simulation model was established using the FR-Sdaso code with the primary pump speed,secondary circuit flow rate,and sodium temperature at the outlet of the air heat exchanger as boundary conditions.The calculation of core fission power adopted a point reactor model,while the decay power adopted a lumped parameter model,considering Doppler,sodium density,axial expansion,radial expansion,control rod drive mechanism expansion,and GEM feedback reactivity.The core thermal calculation adopted a single channel model,dividing the core into 9 thermal channels.The inlet and outlet areas of the primary reactor core adopted a lumped parameter model,while the heat exchanger and pipelines adopt a onedimensional control volume model.The key parameters that could be directly compared with experimental measurement results were analyzed and compared,such as flow rate of each loop in the first circuit,core power,total reactivity of the core,sodium temperature at the outlet of PIOTA subassemblies in the second and sixth rows of the core,temperature in the cold and hot legs of the first and the secondary circuits.The comparison analysis between the calculation results and the FFTF LOFWOS Test#13 test data shows that the FR-Sdaso program can effectively predict the transient changes in reactor power,as well as the temperature and flow rate of the first and second circuits after an unprotected
作者
杨军
叶尚尚
王利霞
YANG Jun;YE Shangshang;WANG Lixia(Department of Nuclear Engineering Design,China Institute of Atomic Energy,Beijing 102413,China)
出处
《原子能科学技术》
EI
CAS
CSCD
北大核心
2024年第9期1884-1892,共9页
Atomic Energy Science and Technology
