DOAJ Open Access 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

Abstrak

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.

Topik & Kata Kunci

Nuclear engineering. Atomic power

Penulis (5)

Chenglong Wang

Chen Wang

Wenxi Tian

Guanghui Su

Suizheng Qiu

Format Sitasi

APA MLA BibTeX

Wang, C., Wang, C., Tian, W., Su, G., Qiu, S. (2024). Code development and preliminary validation for lead-cooled fast reactor thermal-hydraulic transient behavior. https://doi.org/10.1016/j.net.2024.01.044

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Informasi Jurnal

Tahun Terbit: 2024
Sumber Database: DOAJ
DOI: 10.1016/j.net.2024.01.044
Akses: Open Access ✓