Flow blockage analysis for fuel assembly in a lead-based fast reactor

Wang, Chenglong; Wu, Di; Gui, Minyang; Cai, Rong-Gen; Zhu, Dahuan; Zhang, Dalin; Tian, Wenxi; Qiu, Suizheng; Su, G.H.

doi:10.1016/j.net.2021.05.007

Cited by 12 publications

(3 citation statements)

References 19 publications

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“…4) SACOS-PB (Wang et al, 2013a), which is concluded that the hexagonal fuel assembly is the best choice for the LBE fast reactor by analyzing the temperature distribution of the fuel assembly and diverse component types. Wu et al (2021) improved SACOS-PB, and Wang et al (2021) utilized it for flow blockage incident analysis. 5) ANTEO+ (Lodi et al, 2016a), which can be used to calculate liquid metal cooled fast reactors, including sodium, lead, LBE, and is suitable for a variety of geometric structures such as bare-rod, grid-type spacers, and wire-type spacers.…”

Section: Introductionmentioning

confidence: 99%

Sub-channel analysis for LBE cooled fuel assembly with grid type spacers of China initiative accelerator driven system

Chen,

Qiu,

Zhang

et al. 2023

Front. Energy Res.

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The present study focuses on the thermal hydraulic characteristics of fuel assemblies in the China Initiative Accelerator Driven subcritical System (CiADS), which utilizes grid-type spacers to support lead bismuth eutectic (LBE) cooled fast reactors. While the structural design and mechanical analysis of CiADS triangular fuel assemblies with grid-type spacers have been preliminarily investigated, there is a pressing need to study the thermal hydraulic behavior of liquid LBE within these fuel assemblies. This analysis is crucial for assessing the safety and economy of the CiADS subcritical reactor. The sub-channel analysis code, based on the lumped parameter method, plays a vital role in rapidly evaluating the safety features of LBE-cooled fast reactor fuel assemblies. To comprehensively evaluate the thermal hydraulic performance of fuel assemblies with grid-type spacers in the CiADS LBE cooling system, the flow pressure drop model, turbulent mixing model, and convective heat transfer model for the fuel assembly structure with grid-type spacers were incorporated into the existing CiADS sub-channel analysis code. This enhanced code has been successfully employed in the thermal hydraulic analysis of fuel assemblies utilizing wire-type spacers in the CiADS system. To verify the validity and accuracy of the modified CiADS sub-channel analysis code in calculating triangular fuel assemblies with grid-type spacers, the code was first utilized to replicate a 19-rods flow heat transfer experiment with grid-type spacers cooling liquid LBE. A comparison between the simulation and experimental results confirms that the CiADS sub-channel analysis code can predict the coolant temperature and fuel rod surface temperature in fuel assemblies with grid-type spacers. Subsequently, the structure and thermal hydraulic design of the latest CiADS fuel assembly with grid-type spacers were reviewed. Finally, taking into account the aforementioned enhancements to the CiADS sub-channel analysis code, a comprehensive thermo-hydraulic calculation of the CiADS fuel assembly with grid-type spacers was conducted. The results demonstrate that the coolant outlet temperature, maximum cladding temperature, and maximum fuel pellet temperature all fall within the design parameters of CiADS. Furthermore, the impact of the number of rods in different rod bundles (91, 61, 37, and 7 rods) on thermal hydraulic performance was analyzed. The findings indicate that, while ensuring calculation accuracy and efficiency, the results obtained from the 37-rod bundle can effectively reflect the thermal hydraulic behavior of the reactor core.

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Section: Introductionmentioning

confidence: 99%

Sub-channel analysis for LBE cooled fuel assembly with grid type spacers of China initiative accelerator driven system

Chen,

Qiu,

Zhang

et al. 2023

Front. Energy Res.

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“…CFD is an appropriate tool to help understand the thermal‐hydraulic characteristics in fuel assemblies after being validated with the experimental data. Also, analysis codes for the advanced fast reactor are developed 20,21 . In general, the thermal‐hydraulic characteristics and periodical distribution in fuel assemblies were explored.…”

Section: Introductionmentioning

confidence: 99%

“…Also, analysis codes for the advanced fast reactor are developed. 20,21 In general, the thermal-hydraulic characteristics and periodical distribution in fuel assemblies were explored. First, it was found that the axial velocity would periodically fluctuate because the wires sweep regularly, and the period was diverse in different sub-channel.…”

mentioning

confidence: 99%

Thermal‐hydraulic analysis of wire‐wrapped rod bundle in lead‐based fast reactor with non‐uniform heat flux

Dong

Ahmad

Khan

et al. 2022

Intl J of Energy Research

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As one of the most promising advanced reactors, the lead-based fast reactor has drawn great attention due to economic and safety advantages. Investigating thermal hydraulics is essential for the design of a lead-based reactor.In this paper, a CFD simulation of a 19-wire-wrapped-rod bundle with leadbismuth eutectic (LBE) as coolant is carried out using the Reynoldsaveraged Navier-Stokes equation method, and four different types of axial non-uniform heat flux are applied. Excellent validation of results for thermal and hydraulic aspects is obtained first by comparing simulation results using turbulent model k-ω SST with experimental data and high-fidelity large eddy simulation data. The mechanism of transverse flow variation in subchannels and faces resulting from changing locations of wires is studied.

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Safety analysis code development and validation for lead-cooled fast reactor with helical coiled once-through steam generator

Zhang,

Wang,

Tian

et al. 2024

Annals of Nuclear Energy

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Flow blockage analysis for fuel assembly in a lead-based fast reactor

Cited by 12 publications

References 19 publications

Sub-channel analysis for LBE cooled fuel assembly with grid type spacers of China initiative accelerator driven system

Sub-channel analysis for LBE cooled fuel assembly with grid type spacers of China initiative accelerator driven system

Thermal‐hydraulic analysis of wire‐wrapped rod bundle in lead‐based fast reactor with non‐uniform heat flux

Safety analysis code development and validation for lead-cooled fast reactor with helical coiled once-through steam generator

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