Preliminary thermal-hydraulic sub-channel analysis of 61 wire-wrapped bundle cooled by lead bismuth eutectic

Lyu, Kefeng; Chen, Liuli; Chenchong, Yue; Gao, Sheng; Zhou, Tao; Huang, Qi

doi:10.1016/j.anucene.2016.01.034

Cited by 34 publications

(10 citation statements)

References 13 publications

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“…It can be seen that the thermal‐hydraulic characteristics will be affected by the turbulent mixing coefficient, and the larger it, the more uniform the temperature distribution. 0.03 was used in this study, the same as Lyu et al 19 Besides, the results show that the axially‐varied wire model in improved SACOS‐PB could simulate the temperature distributions more accurately compared with Lyu et al 19 The variation trends of temperature in sub‐channel are consistent with the experimental data, which shows the advantages of this model. For edge and corner sub‐channels (#4 and #5), although the prediction accuracy is higher than the existing code, it cannot deny that the validation deviations are indeed large, which indicates that SACOS‐PB has some shortcomings, and needs further improvement in the future.…”

Section: Code Validationsupporting

confidence: 57%

“…To measure the axial pressure drop across the assembly, a full‐scale mock‐up of the MYRRHA fuel assembly was constructed and installed in the COMPonent Loop Testing LBE facility at SCK·CEN, 9 which consists of 127 fuel rod. In China, a holistic experimental loop, KYLIN series, was constructed to carry out the analysis of the thermal‐hydraulic and safety characteristics of LBE by USTC 18,19 . A wire‐wrapped 61‐rod bundle was installed on the KYLIN.…”

Section: Overview Of Previous Investigationsmentioning

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 Validationsupporting

confidence: 57%

Section: Overview Of Previous Investigationsmentioning

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

View full text Add to dashboard Cite

show abstract

“…In Figure 10A, the experimental data have two peaks caused by the thermocouples, which cannot be presented in this calculation. Under natural circulation condition, due to the small mass flow rate, the disturbance caused by the thermocouple is small, so there are no obvious peaks 38 . As it can be seen, under both conditions, the relative errors of all the correlations are within 2%, which is an acceptable precision.…”

Section: Comparative Analysis Of Correlationsmentioning

confidence: 72%

“…The heat section contains a 61‐rod bundle, shown in Figure 6. Main parameters of the test bundle are listed in Table 5 and Figure 7 37,38 . To study the bundle heat transfer characteristics, experiments have been conducted under both forced and natural circulation condition.…”

Section: Comparative Analysis Of Correlationsmentioning

confidence: 99%

“…In the test, Lyu found the closer to the outer tube, the stronger the cross‐flow effect between subchannels, so the central subchannel is least affected by the outer tube 38 . As no cross‐flow effect is considered in this system code, the data of central subchannel #1 are used for correlation comparative analysis.…”

Section: Comparative Analysis Of Correlationsmentioning

confidence: 99%

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Development and validation of transient thermal‐hydraulic evaluation code for a lead‐based fast reactor

Wei

Wang

et al. 2021

Int J Energy Res

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Summary In this paper, development and preliminary validation of an in‐house code LETHAC (lead‐based fast reactor thermal‐hydraulic analysis code) is introduced. LETHAC is developed for predicting the thermal‐hydraulic transient performance of a lead‐bismuth reactor. The mathematical models and module relations of LETHAC are presented in detail. The most‐used LBE (lead‐bismuth eutectic) flow and heat transfer correlations are compared against data from KYLIN‐II 61‐rod wire‐wrapped bundle experiments. Results show little difference between the chosen heat transfer correlations as the relative error of these correlations is all within 2%. Novendstern correlation is recommended for wire‐wrapped bundle frictional resistance coefficient calculation as it fits the experimental data best. Besides, the transient calculation capability of LETHAC is also preliminarily evidenced by TALL transient tests. Good agreements are shown between the calculated temperatures and the experiment ones. Relative errors between calculation results and experimental data are almost within 10% during the whole transient process, which indicates LETHAC can be used for predicting the transient behavior of lead/lead‐bismuth cooled systems.

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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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Preliminary thermal-hydraulic sub-channel analysis of 61 wire-wrapped bundle cooled by lead bismuth eutectic

Cited by 34 publications

References 13 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

Development and validation of transient thermal‐hydraulic evaluation code for a lead‐based fast reactor

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

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