Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Code development on steady-state thermal-hydraulic for small modular natural circulation lead-based fast reactor

  • Zhao, Pengcheng (School of Nuclear Science and Technology, University of South China) ;
  • Liu, Zijing (School of Nuclear Science and Technology, University of South China) ;
  • Yu, Tao (School of Nuclear Science and Technology, University of South China) ;
  • Xie, Jinsen (School of Nuclear Science and Technology, University of South China) ;
  • Chen, Zhenping (School of Nuclear Science and Technology, University of South China) ;
  • Shen, Chong (School of Nuclear Science and Technology, University of Science and Technology of China)
  • Received : 2019.05.28
  • Accepted : 2020.05.23
  • Published : 2020.12.25
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Abstract

Small Modular Reactors (SMRs) are attracting wide attention due to their outstanding performance, extensive studies have been carried out for lead-based fast reactors (LFRs) that cooled with Lead or Lead-bismuth (LBE), and small modular natural circulation LFR is one of the promising candidates for SMRs and LFRs development. One of the challenges for the design small modular natural circulation LFR is to master the natural circulation thermal-hydraulic performance in the reactor primary circuit, while the natural circulation characteristics is a coupled thermal-hydraulic problem of the core thermal power, the primary loop layout and the operating state of secondary cooling system etc. Thus, accurate predicting the natural circulation LFRs thermal-hydraulic features are highly required for conducting reactor operating condition evaluate and Thermal hydraulic design optimization. In this study, a thermal-hydraulic analysis code is developed for small modular natural circulation LFRs, which is based on several mathematical models for natural circulation originally. A small modular natural circulation LBE cooled fast reactor named URANUS developed by Korea is chosen to assess the code's capability. Comparisons are performed to demonstrate the accuracy of the code by the calculation results of MARS, and the key thermal-hydraulic parameters agree fairly well with the MARS ones. As a typical application case, steady-state analyses were conducted to have an assessment of thermal-hydraulic behavior under nominal condition, and several parameters affecting natural circulation were evaluated. What's more, two characteristics parameters that used to analyze natural circulation LFRs natural circulation capacity were established. The analyses show that the core thermal power, thermal center difference and flow resistance is the main factors affecting the reactor natural circulation. Improving the core thermal power, increasing the thermal center difference and decreasing the flow resistance can significantly increase the reactor mass flow rate. Characteristics parameters can be used to quickly evaluate the natural circulation capacity of natural circulation LFR under normal operating conditions.

Keywords

Acknowledgement

This work is supported by the "Characteristics Analysis and Influence Research of Core Power Redistribution and Pool Complex Thermal Stratification for Pump-driven Lead-cooled Fast Reactor under Asymmetric Operation Conditions" of Natural Science Foundation of Hunan Province (Grant No. 2019JJ40239), "Pre-research on key technologies of small long-life natural circulation lead-bismuth fast reactor" of Open Foundation of Science and Technology on Reactor System Design Technology Laboratory (Grant No. HT-KFKT-10-2018001) and Opening Project of Cooperative Innovation Center for Nuclear Fuel Cycle Technology and Equipment, University of South China (Grant No. 2019KFZ08).

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