Code development on steady-state thermal-hydraulic for small modular natural circulation lead-based fast reactor

Pengcheng Zhao; Zijing Liu; Tao Yu; Jinsen Xie; Zhenping Chen; Chong Shen

doi:https://doi.org/10.1016/j.net.2020.05.023

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저자정보: Pengcheng Zhao (School of Nuclear Science and Technology, University of South China, Hengyang, Hunan, 421001, China) Zijing Liu (School of Nuclear Science and Technology, University of South China, Hengyang, Hunan, 421001, China) Tao Yu (School of Nuclear Science and Technology, University of South China, Hengyang, Hunan, 421001, China) Jinsen Xie (School of Nuclear Science and Technology, University of South China, Hengyang, Hunan, 421001, China) Zhenping Chen (School of Nuclear Science and Technology, University of South China, Hengyang, Hunan, 421001, China) Chong Shen (School of Nuclear Science and Technology University of Science and Technology of China Hefei Anh)

저널정보: 한국원자력학회 Nuclear Engineering and Technology Nuclear Engineering and Technology 제52권 제12호

발행연도: 2020.12

수록면: 2,789 - 2,802 (14page)

DOI: https://doi.org/10.1016/j.net.2020.05.023

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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 Leadbismuth (LBE), and small modular natural circulation LFR is one of the promising candidates for SMRsand LFRs development. One of the challenges for the design small modular natural circulation LFR is tomaster the natural circulation thermal-hydraulic performance in the reactor primary circuit, while thenatural circulation characteristics is a coupled thermal-hydraulic problem of the core thermal power, theprimary loop layout and the operating state of secondary cooling system etc. Thus, accurate predictingthe natural circulation LFRs thermal-hydraulic features are highly required for conducting reactoroperating condition evaluate and Thermal hydraulic design optimization. In this study, a thermalhydraulic analysis code is developed for small modular natural circulation LFRs, which is based onseveral mathematical models for natural circulation originally. A small modular natural circulation LBEcooled 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, andthe 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 undernominal condition, and several parameters affecting natural circulation were evaluated. What's more,two characteristics parameters that used to analyze natural circulation LFRs natural circulation capacitywere established. The analyses show that the core thermal power, thermal center difference and flowresistance 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 thereactor mass flow rate. Characteristics parameters can be used to quickly evaluate the natural circulationcapacity of natural circulation LFR under normal operating conditions.

#Code development #LFR #Natural circulation #Steady-state thermal-hydraulic

참고문헌 (11)

참고문헌 신청

H. Chen / 2016 / Conceptual design of a small modular natural circulation lead cooled fast reactor SNCLFR-100 / Int. J. Hydrogen Energy 41(17):7158~7168

Z. Chen / 2014 / Study of core flow distribution for small modular natural circulation lead or lead-alloy cooled fast reactors / Ann. Nucl. Energy 72:76~83

C. Fazio / 2015 / Handbook on Lead-Bismuth Eutectic Alloy and Lead Properties, Materials Compatibility, Thermal-Hydraulics and Technologies-2015 edition[R]

J. K. Fink / 2000 / Thermophysical properties of uranium dioxide / J. Nucl. Mater 279(1):1~18

C. Duriez / 2000 / Thermal conductivity of hypostoichiometric low Pu content (U, Pu) O2 x mixed oxide / J. Nucl. Mater 277(2-3):143~158

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Pengcheng Zhao

소속기관 School of Nuclear Science and Technology, University of South China, Hengyang, Hunan, 421001, China