Development and application of a thermal-hydraulics analysis code for small modular natural circulation lead or lead-alloy cooled fast reactors

作者：Chen, Zhao; Zhou, Guangming; Zhao, Pengcheng; Chen, Hongli^*

来源：Annals of Nuclear Energy, 2014, 73: 68-74.

DOI：10.1016/j.anucene.2014.06.031

• Small Modular Reactor
• Lead cooled
• Fast reactor
• Natural circulation
• Thermal-hydraulics

Small modular natural circulation lead or lead-alloy cooled fast reactor (LFR) is one of the potential candidates for passive cooling SMR development. In a pump-forced reactor, the coolant mass flow rate is determined by the pump power, while in a small modular natural circulation LFR, the coolant mass flow rate is determined from the balance between the driving force (due to density difference between hot and cold sides) and the resistance force (due to pressure drop in the system). For this reason, the effects of the core coolant channel geometry on the thermal-hydraulics performance in a small modular natural circulation LFR is more sensitive than that in a pump-forced reactor. This problem is not considered in existing thermal-hydraulics analysis codes, which are developed for pump-forced reactors originally. In this paper, a thermal-hydraulics analysis code for small modular natural circulation LFRs was developed, which is based on the mathematical models for natural circulation reactors originally. A validation case from Argonne National Laboratory (ANL) was chosen and calculated by using this code and the results agree fairly well with the ANL's ones. As a typical application case, the thermal-hydraulics design and core coolant channel geometry effect analysis of a 10 MW natural circulation LFR was carried out.

• 出版日期2014-11
• 单位中国科学技术大学