3D thermal hydraulic characteristics analysis of pool-type upper plenum for lead-cooled fast reactor with multi-scale coupling program

Zou, Ze-Ren; Shen, Chong; Zhang, Xilin; Wang, Shuai; Chen, Hongli

doi:10.1016/j.nucengdes.2020.110892

Cited by 10 publications

(3 citation statements)

References 11 publications

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“…The heat transfer between the hot and cold sides is calculated by using the average temperature. The heat transfer coefficient is assumed to be constant [11][12][13].…”

Section: The Thermal-hydraulic Methodsmentioning

confidence: 99%

“…However, the various prompt neutrons per fission and the numerous delayed neutrons of precursor per fission affect the velocity of the delayed neutron decay constants of the precursor, and generation of neutron time respectively. ߬ shows the transit time [12]. The fuel salt passes through the core, pipe section #1, heat exchanger, and pipe section #2, and then it enters the core again as shown in Figure 1.…”

Section: Mathematical Derivation and Reactor Neutrons Of The Msrmentioning

confidence: 99%

See 1 more Smart Citation

The Development of a Thermal-Hydraulic and Nonlinear Dynamic System for Molten Salt Reactors

Ali

Liu

Raza³

et al. 2021

Eng. Technol. Appl. Sci. Res.

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The Molten Salt Reactor (MSR) is the most important system suggested by Generation IV for the future direction in the nuclear reactor field. For more development of the MSR reactor, the core system inside the tube is proposed by naturally circulating molten fuel salt. The nonlinear kinetic equations form a linearized function and are obtained in state-space form. Reactivity feedback and delayed neutrons are extremely important for reactor control. In this paper, a thermal-hydraulic system for the commercial computation dynamic model is proposed. Currently, there is no commercial software to simulate the natural circulation flow. The proposed method can be easily employed to detect faults and can provide a feasible overall system performance.

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“…The heat transfer between the hot and cold sides is calculated by using the average temperature. The heat transfer coefficient is assumed to be constant [11][12][13].…”

Section: The Thermal-hydraulic Methodsmentioning

confidence: 99%

Section: Mathematical Derivation and Reactor Neutrons Of The Msrmentioning

confidence: 99%

The Development of a Thermal-Hydraulic and Nonlinear Dynamic System for Molten Salt Reactors

Ali

Liu

Raza³

et al. 2021

Eng. Technol. Appl. Sci. Res.

View full text Add to dashboard Cite

show abstract

“…Zou et al [ 96 ] performed a 3D thermal-hydraulic multi-scale coupling assessment of the top plenum in a pool-shaped LBE fast reactor employing the ATHLET-OpenFoam code. The ATHLET code reveals regions where stratification occurred.…”

Section: Analysis Of Thermal Hydraulics Activities Performed On Hlm-c...mentioning

confidence: 99%

Liquid metal cooled fast reactor thermal hydraulic research development: A review

Agbevanu

Arthur

Debrah

2023

Heliyon

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Verification of Solver for Coupled Simulation of Fluid and Fuel Pin in LFR Based on Openfoam

Gong

Pan

et al. 2023

Springer Proceedings in Physics

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The Computational Fluid Dynamics (CFD)-based thermal-hydraulics and safety analyses of Lead-based Fast Reactor (LFR) have attracted great attentions in recent years. Commercial CFD tools have been widely used in the 3D simulations of pool-type reactors owing to their powerful abilities in geometric modeling and meshing. Compared with the commercial CFD tools, OpenFOAM is a free open-source CFD code, which is more flexible to perform multi-physics coupling activities. In this paper, in order to develop a solver for simulating the coupled flow and heat transfer behaviors of fluid (coolant) and fuel pin in LFR, the fuel pin Heat Conduction (HC) model was coupled to the modified icoFoam solver of OpenFOAM. Verifications were conducted by the steady-state coupled simulation of fluid and fuel pin heat transfer behaviors, comparing with the MPC-LBE code which has been verified by the benchmarks for LFR fuel pin channel. The results simulated by the coupled solver proposed in this paper agreed well with the ones provided by the MPC-LBE code. This study lays a foundation for the further development of transient safety analysis code for LFR in our future work.

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3D thermal hydraulic characteristics analysis of pool-type upper plenum for lead-cooled fast reactor with multi-scale coupling program

Cited by 10 publications

References 11 publications

The Development of a Thermal-Hydraulic and Nonlinear Dynamic System for Molten Salt Reactors

The Development of a Thermal-Hydraulic and Nonlinear Dynamic System for Molten Salt Reactors

Liquid metal cooled fast reactor thermal hydraulic research development: A review

Verification of Solver for Coupled Simulation of Fluid and Fuel Pin in LFR Based on Openfoam

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