Numerical Simulation Research on Natural Circulation Flow of the Reactor Coolant System

Zhang Mingqian; Lin Run; Li Zhenguang

doi:10.13832/j.jnpe.2024.050038

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Article Navigation > Nuclear Power Engineering > 2025 > Accepted Manuscript

Zhang Mingqian, Lin Run, Li Zhenguang. Numerical Simulation Research on Natural Circulation Flow of the Reactor Coolant System[J]. Nuclear Power Engineering. doi: 10.13832/j.jnpe.2024.050038

Citation:

Zhang Mingqian, Lin Run, Li Zhenguang. Numerical Simulation Research on Natural Circulation Flow of the Reactor Coolant System[J]. Nuclear Power Engineering. doi: 10.13832/j.jnpe.2024.050038

Zhang Mingqian, Lin Run, Li Zhenguang. Numerical Simulation Research on Natural Circulation Flow of the Reactor Coolant System[J]. Nuclear Power Engineering. doi: 10.13832/j.jnpe.2024.050038

Citation:

Zhang Mingqian, Lin Run, Li Zhenguang. Numerical Simulation Research on Natural Circulation Flow of the Reactor Coolant System[J]. Nuclear Power Engineering. doi: 10.13832/j.jnpe.2024.050038

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Numerical Simulation Research on Natural Circulation Flow of the Reactor Coolant System

doi: 10.13832/j.jnpe.2024.050038

1.
State Key Laboratory of Nuclear Power Safety Technology and Equipment, Shenzhen, Guangdong, 518172, China
2.
China Nuclear Power Design Co., Ltd.(Shenzhen), Shenzhen, Guangdong, 518172, China

Received Date: 2024-05-15
Rev Recd Date: 2024-07-17

Available Online: 2025-04-22

Abstract

Abstract

Computational Fluid Dynamics(CFD) program is employed to enable the high-fidelity modeling of the reactor coolant system (RCS) for a typical three-loop pressurized water reactor, and the completed model of the RCS is build including reactor vessel and internals, core, steam generator, primary pump and linking pipe. The three-dimensional, global and localized flow features have been investigated under natural circulation flow condition with lower core thermal power, and the temperature at different locations are compared with the measured values from the operating nuclear power plant in order to verify the accurate description of the developed CFD model. The results show that the natural circulation flow rate is about 4.5% of the full power flow rate while the temperature of the core outlet is stable, and the residual core heat could be effectively removed. The phenomenon of the thermal stratification in the reactor pressure vessel head dome shows that the measured temperature value of the detector position in nuclear power plant could not provide the highest value. The coolant from different loops could be more fully mixed due to the local convection flow. There is a swirling flow at the outlet of primary pump, and the tangential velocity near the pipe wall is large while the local convection occurs at the central area. This analysis practice provides an effective evaluation for the system-level three-dimensional thermal hydraulics phenomena of the reactor coolant system.
- Computational fluid dynamics(CFD),
- Reactor coolant system,
- Natural circulation flow

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References(18)

References

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