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

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 exchang...

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Bibliographic Details
Published in:Nuclear engineering and technology 2024, Vol.56 (6), p.2332-2342
Main Authors: Chenglong Wang, Chen Wang, Wenxi Tian, Guanghui Su, Suizheng Qiu
Format: Article
Language:Korean
Online Access:Get full text
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Summary: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.
ISSN:1738-5733
2234-358X