TY - JOUR
T1 - Multi-scale system–subchannel coupling for PWR thermal-hydraulic analysis
T2 - Development and Validation
AU - Zhu, Xinyang
AU - Lu, Guoqing
AU - Li, Sijun
AU - Zhong, Yubao
AU - Chen, Ronghua
AU - Tian, Wenxi
N1 - Publisher Copyright:
© 2026 Elsevier Ltd. All rights are reserved, including those for text and data mining, AI training, and similar technologies.
PY - 2026/5
Y1 - 2026/5
N2 - A multiscale coupled code, VITARS-SC, was developed by externally coupling the system code VITARS (Visual Interactive Transient Analysis code for nuclear Reactor System) with the core subchannel code SACOS (Subchannel Analysis Code Of Safety) through a spatial overlapping-domain coupling strategy. The coupled framework incorporates system-level component models for key equipment such as pipes, pumps, and valves, while also enabling high-fidelity subchannel resolution within the reactor core. Specifically, VITARS provides the core inlet and outlet boundary conditions (inlet temperature/enthalpy, inlet mass flow and outlet pressure) to the subchannel code; in turn, SACOS performs detailed core calculations and feeds back a corrected core pressure drop to the system model. The accuracy and reliability of VITARS-SC were validated against the transient data of the LOFT L9-1 test, which corresponds to a loss-of-flow accident (LOFA) scenario. The predicted trends of key loop thermal–hydraulic parameters show excellent consistency with the experimental data, and the coupled code yields a more accurate prediction of the transient event sequence than the standalone VITARS calculation. Furthermore, a representative main steam line break (MSLB) scenario for a typical pressurized water reactor (PWR) was simulated as an application case. The calculated core inlet and outlet temperatures, primary-system pressure, primary-system mass flow rate, and break steam flow rate are physically reasonable and well-reflect the characteristic thermal-hydraulic behaviors of MSLB accidents. In addition, the high-fidelity predictions of core enthalpy and void fraction distributions are consistent with theoretical physical expectations. These results demonstrate that VITARS-SC has preliminary capability for full-loop multiscale thermal–hydraulic transient analyses of PWRs, offering a valuable numerical tool for PWR design optimization and safety assessment.
AB - A multiscale coupled code, VITARS-SC, was developed by externally coupling the system code VITARS (Visual Interactive Transient Analysis code for nuclear Reactor System) with the core subchannel code SACOS (Subchannel Analysis Code Of Safety) through a spatial overlapping-domain coupling strategy. The coupled framework incorporates system-level component models for key equipment such as pipes, pumps, and valves, while also enabling high-fidelity subchannel resolution within the reactor core. Specifically, VITARS provides the core inlet and outlet boundary conditions (inlet temperature/enthalpy, inlet mass flow and outlet pressure) to the subchannel code; in turn, SACOS performs detailed core calculations and feeds back a corrected core pressure drop to the system model. The accuracy and reliability of VITARS-SC were validated against the transient data of the LOFT L9-1 test, which corresponds to a loss-of-flow accident (LOFA) scenario. The predicted trends of key loop thermal–hydraulic parameters show excellent consistency with the experimental data, and the coupled code yields a more accurate prediction of the transient event sequence than the standalone VITARS calculation. Furthermore, a representative main steam line break (MSLB) scenario for a typical pressurized water reactor (PWR) was simulated as an application case. The calculated core inlet and outlet temperatures, primary-system pressure, primary-system mass flow rate, and break steam flow rate are physically reasonable and well-reflect the characteristic thermal-hydraulic behaviors of MSLB accidents. In addition, the high-fidelity predictions of core enthalpy and void fraction distributions are consistent with theoretical physical expectations. These results demonstrate that VITARS-SC has preliminary capability for full-loop multiscale thermal–hydraulic transient analyses of PWRs, offering a valuable numerical tool for PWR design optimization and safety assessment.
KW - Couple
KW - Subchannel
KW - System
KW - Thermal-hydraulic
KW - Validation
UR - https://www.scopus.com/pages/publications/105034257455
U2 - 10.1016/j.pnucene.2026.106334
DO - 10.1016/j.pnucene.2026.106334
M3 - 文章
AN - SCOPUS:105034257455
SN - 0149-1970
VL - 195
JO - Progress in Nuclear Energy
JF - Progress in Nuclear Energy
M1 - 106334
ER -