Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-07-02
Abstract： [Background]: As the commercial application of sodium-cooled fast reactors continues to develop, there is a need to develop an integrated analysis code for sodium-cooled fast reactors in order to enhance the capability for safety assessment. [Purpose]: Sodium boiling is a critical phenomenon prior to severe accidents in sodium-cooled fast reactors, and accurate prediction of thermodynamic parameters of the coolant at the onset of sodium boiling is essential for accident analysis. [Methods]: This paper introduces the development of the integrated analysis code ISAA-Na for sodium-cooled fast reactors, focusing on the core thermal-hydraulic model. The ISAA-Na code was used to simulate and analyze the loss-of-flow experiment CABRI-BI1. [Results]: Comparison with experimental data shows that ISAA-Na provides more accurate predictions of coolant temperature and pressure prior to boiling compared to other codes such as SAS4A, ASTEC-Na, and SIMMER. However, after the onset of boiling, overestimations of bubble growth rate and two-phase interface movement were observed due to the lack of fuel melting and cladding failure models. [Conclusions]: Overall, the application of the multi-bubble liquid slug boiling model in ISAA-Na for the analysis of the CABRI-BI1 experiment is deemed reasonable and reliable.
Subjects: Nuclear Science and Technology >> Nuclear Science and Technology submitted time 2024-07-01
Abstract： [Background]: As the commercial application of sodium-cooled fast reactors continues to develop, there is a need to develop an integrated analysis code for sodium-cooled fast reactors in order to enhance the capability for safety assessment. [Purpose]: Sodium boiling is a critical phenomenon prior to severe accidents in sodium-cooled fast reactors, and accurate prediction of thermodynamic parameters of the coolant at the onset of sodium boiling is essential for accident analysis. [Methods]: This paper introduces the development of the integrated analysis code ISAA-Na for sodium-cooled fast reactors, focusing on the core thermal-hydraulic model. The ISAA-Na code was used to simulate and analyze the loss-of-flow experiment CABRI-BI1. [Results]: Comparison with experimental data shows that ISAA-Na provides more accurate predictions of coolant temperature and pressure prior to boiling compared to other codes such as SAS4A, ASTEC-Na, and SIMMER. However, after the onset of boiling, overestimations of bubble growth rate and two-phase interface movement were observed due to the lack of fuel melting and cladding failure models. [Conclusions]: Overall, the application of the multi-bubble liquid slug boiling model in ISAA-Na for the analysis of the CABRI-BI1 experiment is deemed reasonable and reliable.
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