TY - JOUR
T1 - Code development and preliminary validation for lead-cooled fast reactor thermal-hydraulic transient behavior
AU - Wang, Chenglong
AU - Wang, Chen
AU - Tian, Wenxi
AU - Su, Guanghui
AU - Qiu, Suizheng
N1 - Publisher Copyright:
© 2024 Korean Nuclear Society
PY - 2024/6
Y1 - 2024/6
N2 - 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.
AB - 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.
KW - Code development
KW - Lead-cooled fast reactor
KW - Thermal-hydraulic analysis
UR - https://www.scopus.com/pages/publications/85187325591
U2 - 10.1016/j.net.2024.01.044
DO - 10.1016/j.net.2024.01.044
M3 - 文章
AN - SCOPUS:85187325591
SN - 1738-5733
VL - 56
SP - 2332
EP - 2342
JO - Nuclear Engineering and Technology
JF - Nuclear Engineering and Technology
IS - 6
ER -