Code development and preliminary validation for lead-cooled fast reactor thermal-hydraulic transient behavior

Chenglong Wang; Chen Wang; Wenxi Tian; Guanghui Su; Suizheng Qiu

Nuclear Engineering and Technology (Jun 2024)

Code development and preliminary validation for lead-cooled fast reactor thermal-hydraulic transient behavior

Chenglong Wang,
Chen Wang,
Wenxi Tian,
Guanghui Su,
Suizheng Qiu

Affiliations

Chenglong Wang: Corresponding author.; Department of Nuclear Science and Technology, State Key Lab. of Multiphase Flow in Power Engineering, Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi'an Jiaotong University, Xi'an, Shaanxi, 710049, China
Chen Wang: Department of Nuclear Science and Technology, State Key Lab. of Multiphase Flow in Power Engineering, Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi'an Jiaotong University, Xi'an, Shaanxi, 710049, China
Wenxi Tian: Department of Nuclear Science and Technology, State Key Lab. of Multiphase Flow in Power Engineering, Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi'an Jiaotong University, Xi'an, Shaanxi, 710049, China
Guanghui Su: Department of Nuclear Science and Technology, State Key Lab. of Multiphase Flow in Power Engineering, Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi'an Jiaotong University, Xi'an, Shaanxi, 710049, China
Suizheng Qiu: Department of Nuclear Science and Technology, State Key Lab. of Multiphase Flow in Power Engineering, Shaanxi Key Lab. of Advanced Nuclear Energy and Technology, Xi'an Jiaotong University, Xi'an, Shaanxi, 710049, China

Journal volume & issue: Vol. 56, no. 6
pp. 2332 – 2342

Abstract

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Lead-cooled fast reactors (LFRs) have a wide range of application scenarios, which require the thermal-hydraulic characteristics of LFRs to be reliable. In the present paper, the Lead-cooled fast reactor Thermal-Hydraulic Analysis Code LETHAC was developed, including the models of pipe, heat exchanger, and pool. To verify the correctness of LETHAC, two experimental facilities and three experimental cases were selected, including GFT and PLOFA tests for NACIE-UP and Test-1 for CIRCE. The calculated results show the same and consistent trend with the experimental data, but there are some discrepancies. It can be found that LETHAC is suitable and reliable in predicting the transient behavior of lead-cooled system.

Published in Nuclear Engineering and Technology

ISSN: 1738-5733 (Print)
Publisher: Elsevier
Country of publisher: Korea, Republic of
LCC subjects: Technology: Electrical engineering. Electronics. Nuclear engineering: Nuclear engineering. Atomic power
Website: http://www.journals.elsevier.com/nuclear-engineering-and-technology

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