Thermal hydraulic study of NuScale iPWR passive containment cooling behaviour during loss-of-coolant accident
Operation of the NuScale emergency core cooling system (ECCS) was studied, with an emphasis on the crucial role of the containment vessel (CNV). Design parameters such as vessel volume were varied to explore the resulting thermal hydraulic (TH) behaviour at the system level in the case of a loss-of-...
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KeAi Communications Co., Ltd.
2025-06-01
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| Series: | International Journal of Advanced Nuclear Reactor Design and Technology |
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| Online Access: | http://www.sciencedirect.com/science/article/pii/S2468605025000493 |
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| author | Zhexi Guo Sicong Xiao |
| author_facet | Zhexi Guo Sicong Xiao |
| author_sort | Zhexi Guo |
| collection | DOAJ |
| description | Operation of the NuScale emergency core cooling system (ECCS) was studied, with an emphasis on the crucial role of the containment vessel (CNV). Design parameters such as vessel volume were varied to explore the resulting thermal hydraulic (TH) behaviour at the system level in the case of a loss-of-coolant accident (LOCA) initiated by a chemical and volume control system (CVCS) injection line break within the CNV. Results showed that an important phenomenon of concern was core fluid vaporisation, which could be suppressed with smaller vessels to keep system pressure and water levels in the core higher. Break elevation was found to be less significant, unless it were to occur below the reactor recirculation valves (RRV), which would result in backflow into the core and disrupt natural circulation. Increasing nominal power from 160 MW to 250 MW did not result in very different TH phenomena and could be managed by similar CNV designs and properties. Combined effects of a break at high elevation at the higher power of 250 MW appeared to reduce overall in-core vaporisation. Long-term cooling of the system was finally shown to be ensured up to 3 days with monotonously decreasing or constant core void fractions and pressures. |
| format | Article |
| id | doaj-art-9c737c8aee704466a47ae7c0ef7e0d4b |
| institution | DOAJ |
| issn | 2468-6050 |
| language | English |
| publishDate | 2025-06-01 |
| publisher | KeAi Communications Co., Ltd. |
| record_format | Article |
| series | International Journal of Advanced Nuclear Reactor Design and Technology |
| spelling | doaj-art-9c737c8aee704466a47ae7c0ef7e0d4b2025-08-20T02:44:53ZengKeAi Communications Co., Ltd.International Journal of Advanced Nuclear Reactor Design and Technology2468-60502025-06-017210010910.1016/j.jandt.2025.05.003Thermal hydraulic study of NuScale iPWR passive containment cooling behaviour during loss-of-coolant accidentZhexi Guo0Sicong Xiao1Singapore Nuclear Research and Safety Institute, National University of Singapore, 16 Prince George's Park, 118415, SingaporeCorresponding author.; Singapore Nuclear Research and Safety Institute, National University of Singapore, 16 Prince George's Park, 118415, SingaporeOperation of the NuScale emergency core cooling system (ECCS) was studied, with an emphasis on the crucial role of the containment vessel (CNV). Design parameters such as vessel volume were varied to explore the resulting thermal hydraulic (TH) behaviour at the system level in the case of a loss-of-coolant accident (LOCA) initiated by a chemical and volume control system (CVCS) injection line break within the CNV. Results showed that an important phenomenon of concern was core fluid vaporisation, which could be suppressed with smaller vessels to keep system pressure and water levels in the core higher. Break elevation was found to be less significant, unless it were to occur below the reactor recirculation valves (RRV), which would result in backflow into the core and disrupt natural circulation. Increasing nominal power from 160 MW to 250 MW did not result in very different TH phenomena and could be managed by similar CNV designs and properties. Combined effects of a break at high elevation at the higher power of 250 MW appeared to reduce overall in-core vaporisation. Long-term cooling of the system was finally shown to be ensured up to 3 days with monotonously decreasing or constant core void fractions and pressures.http://www.sciencedirect.com/science/article/pii/S2468605025000493NuScaleECCSASTECLOCAContainment behaviour |
| spellingShingle | Zhexi Guo Sicong Xiao Thermal hydraulic study of NuScale iPWR passive containment cooling behaviour during loss-of-coolant accident International Journal of Advanced Nuclear Reactor Design and Technology NuScale ECCS ASTEC LOCA Containment behaviour |
| title | Thermal hydraulic study of NuScale iPWR passive containment cooling behaviour during loss-of-coolant accident |
| title_full | Thermal hydraulic study of NuScale iPWR passive containment cooling behaviour during loss-of-coolant accident |
| title_fullStr | Thermal hydraulic study of NuScale iPWR passive containment cooling behaviour during loss-of-coolant accident |
| title_full_unstemmed | Thermal hydraulic study of NuScale iPWR passive containment cooling behaviour during loss-of-coolant accident |
| title_short | Thermal hydraulic study of NuScale iPWR passive containment cooling behaviour during loss-of-coolant accident |
| title_sort | thermal hydraulic study of nuscale ipwr passive containment cooling behaviour during loss of coolant accident |
| topic | NuScale ECCS ASTEC LOCA Containment behaviour |
| url | http://www.sciencedirect.com/science/article/pii/S2468605025000493 |
| work_keys_str_mv | AT zhexiguo thermalhydraulicstudyofnuscaleipwrpassivecontainmentcoolingbehaviourduringlossofcoolantaccident AT sicongxiao thermalhydraulicstudyofnuscaleipwrpassivecontainmentcoolingbehaviourduringlossofcoolantaccident |