Computational fluid dynamics and subchannel analysis of lead–bismuth eutectic-cooled fuel assembly under various blockage conditions

Liu, X.J.; Yang, Dongmei; Yang, Yanhua; Chai, Xiaoyan; Xiong, Juntao; Zhang, Tengfei; Xu, Cheng

doi:10.1016/j.applthermaleng.2019.114419

Cited by 25 publications

(7 citation statements)

References 8 publications

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“…Combined with statistic parameters for Equation ) (Table 10), it could be seen that the temperature differences calculated by SACOS‐PB are consistent with experimental data, and the calculation deviations are within 25%. In contrast with Liu et al, 42 SACOS‐PB has better prediction ability. Figure 9 shows the flowrate for three type sub‐channels along with the axial height.…”

Section: Code Validationcontrasting

confidence: 65%

“…Figures 7 and 8 presents a comparison of the predicted temperature differences and experimental data. Besides, the calculation results by Liu et al 42 are shown as well, wherein a sub‐channel analysis code was developed for liquid metal cooled fuel assembly. Three different axial planes at different heights (ML 1, ML 2, and ML 3) were investigated in the experiment.…”

Section: Code Validationmentioning

confidence: 99%

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Improvement and validation of a sub‐channel analysis code for a lead‐cooled reactor with wire spacers

Wang

Gui

et al. 2020

Int J Energy Res

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Summary The fuel rods in the lead‐based fast reactor (LFR) are usually supported by wire spacer, and the presence of the wire results in the complexity of the thermal‐hydraulic analysis in rod bundles. In this study, several previous experimental research about lead‐bismuth eutectic flowing through the rod bundle are investigated. Based on the collected experimental data, a series of flow and heat transfer correlations in the wire‐fixed rod bundle are compared and evaluated. The recommended correlations are then implemented into the in‐house sub‐channel analysis code SACOS‐PB. Furthermore, to account for the changes in the channel cross‐section parameters caused by the wire spacer, two optional wire geometric treatments were added to the code: axially‐averaged approach and axially‐varied approach with extra cross‐flow enhancement model. 19‐rod bundle and 61‐rod bundle tests are used to validate the optimized code. The deviations for the temperature differences of cladding and coolant in internal sub‐channels are within 25%, and the axially‐varied approach shows a higher prediction accuracy, which proves that the improvement of the SACOS‐PB are reliable. This work could provide a reference for the subsequent design and development of the LFR core.

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Section: Code Validationcontrasting

confidence: 65%

Section: Code Validationmentioning

confidence: 99%

Improvement and validation of a sub‐channel analysis code for a lead‐cooled reactor with wire spacers

Wang

Gui

et al. 2020

Int J Energy Res

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“…Moreover, the flow characteristic 32 and the hot spot issues are investigated under flow blockage accidents, which is regarded as one of the most significant issues in lead fast reactors. The effect of blockage thickness, area, and location under uniform cases are studied to evaluate the design criteria 33,34 . In summary, although there is numerous research on thermal‐hydraulics in a lead fast reactor, the study of the hot spot issues under non‐uniform heat flux is limited, especially when the blockage appears at the peak heat flux.…”

Section: Introductionmentioning

confidence: 99%

“…The effect of blockage thickness, area, and location under uniform cases are studied to evaluate the design criteria. 33,34 In summary, although there is numerous research on thermalhydraulics in a lead fast reactor, the study of the hot spot issues under non-uniform heat flux is limited, especially when the blockage appears at the peak heat flux. It is significant to investigate the maximum cladding temperature under non-uniform heat flux to practically assess the safety design of a lead fast reactor.…”

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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“…The following section devoted the literature review makes a survey of those studies which deal with local blockage burned out (Roychowdhury et al, 2000). Liu et al analyzed five blockage accidents for a liquid metal reactor using STAR-CCM + software, taking into consideration the effect of keeping the clad temperature below melting point (Liu et al, 2020). Rahm et al (2019) proposed a new design for liquid metal reactors resistant to instantaneous local blockage accidents based on a BETINa calculation code.…”

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confidence: 99%

Transient Parameter Analysis of Non-scrammed Local Melting Accidents—A VVER 1000 Case Study

et al. 2021

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The first stage of a core degradation—based on the defense-in-depth concept of nuclear power plant (NPP) safety—is prone to fuel melting due to local blockage. The flow blockage accidents with no SCRAM happening can lead to a local fuel-clad failure, consequently affecting the safety of NPP. The present study provides an analysis of Anticipated Transient Without SCRAM (ATWS), which might lead to a condition of burning out. The accidents related to the ATWS scenarios, detailed in the case of VVER-1000/V446 reactor FSAR (Final Safety Analysis Report), include pump failure, local blockage, relative power increase, and a combination of these transients. In this research, first, drawing upon MCNPX 2.7 and COBRA-EN codes, a coupling framework is developed and then validated using an authentic reference point. The obtained results reveal that the reactor SCRAM does not occur while accidents are being investigated as there is a 10% difference in the mass flow rate reduction, a 470 kPa in the channel pressure drop, and a 204°K in the clad temperature, which constitute limitations under most pessimistic scenarios. However, under these conditions, a 70% void fraction over 12 min is observed in certain channels. Hence, burnout and local fuel melting could occur under normal operational and ATWS circumstances. According to uncertainty analyses, the occurrence of the void fraction above zero is locally definite. The transient analysis outputs could be deployed as monitoring system inputs and exploited for identifying weak points in the system.

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Computational fluid dynamics and subchannel analysis of lead–bismuth eutectic-cooled fuel assembly under various blockage conditions

Cited by 25 publications

References 8 publications

Improvement and validation of a sub‐channel analysis code for a lead‐cooled reactor with wire spacers

Improvement and validation of a sub‐channel analysis code for a lead‐cooled reactor with wire spacers

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

Transient Parameter Analysis of Non-scrammed Local Melting Accidents—A VVER 1000 Case Study

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