Review of Thermal-Hydraulic Issues and Studies of Lead-based fast reactors

Zhang, Yan; Wang, Chenglong; Lan, Zhike; Wei, Shiying; Chen, Ronghua; Tian, Wenxi; Su, G.H.

doi:10.1016/j.rser.2019.109625

Cited by 67 publications

(18 citation statements)

References 109 publications

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“…There are mainly four institutions and universities studying the thermal‐hydraulic characteristics of LBE flowing through fuel bundle by experiment, namely, Karlsruhe Liquid Metal Laboratory (KALLA) in Germany, Energia Nucleare ed Energie Alternative (ENEA) Brasimone Research Center in Italy, Belgian Nuclear Research Centre (SCK·CEN), and University of Science and Technology of China (USTC) 9,10 . The fuel bundle is located by grid or wire spacers; thus, these two critical components are considered in the experiments.…”

Section: Overview Of Previous Investigationsmentioning

confidence: 99%

“…MYRRHA is regarded as an accelerator‐driven system (ADS), while it could be able to operate in a critical mode once the spallation target is removed and the control is inserted. To study the thermal‐hydraulic behaviour of LBE flowing through the rod bundle, two experimental facilities, named Technologies for Heavy Metal Systems (THESYS) and Thermal‐Hydraulic and ADS Design (THEADES) were constructed by KALLA, 10 respectively. Based on THEADES facility, pressure drop and heat transfer characteristics for the 19‐rod bundle with grid and wire spacers were analysed, and different flow blockage conditions were considered 11‐13 .…”

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: Overview Of Previous Investigationsmentioning

confidence: 99%

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

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“…For mercury and alkali metals, a large amount of experimental data is available for developing and validating empirical relations. However, comparatively little experimental data relating to LLMs have been obtained, and few relations exist that have been specifically developed for LLMs (Pacio et al, 2015;Zhang et al, 2020b). Although it has long been believed that the heat-transfer relations of other liquid metals can also be applied to LLM, a recent study (Cheng and Tak, 2006) shows that the turbulent Prandtl numbers of LLMs are significantly higher than those of other liquid metals.…”

Section: Highlights Introductionmentioning

confidence: 99%

Study of Flow and Heat Transfer Characteristics of Lead-Based Liquid Metals in a Turbulent Tube Flow and the Impacts of Roughness

et al. 2021

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Lead-based liquid metals (LLMs) such as lead–bismuth eutectic (LBE) and lead, are currently the most interesting candidate coolants for fast reactors because of their excellent physical properties, which can improve safety and reduce costs. However, in comparison to other liquid metals, previous research on the flow and heat-transfer characteristics of LLMs has been limited. Therefore, this work carried out flow and heat-transfer experiments in LBE flowing through a circular tube in the Natural Circulation Capability Loop (NCCL) facility. The results show a significantly higher friction factor than that of water flowing in a smooth pipe. Furthermore, the Nusselt numbers were found to be lower than those found in data in the literature for experiments carried out in a smooth tube at low Péclet numbers, while they were higher at high Péclet numbers. Therefore, theoretical analyses were performed for LLMs flowing in both smooth and rough pipes, and the impacts of roughness on the heat transfer of an LLM were examined. The theoretical relations for a smooth pipe and a rough pipe were validated using experimental data from the literature and the results of the NCCL experiments, respectively. The results of the theoretical relation for a smooth pipe fitted the literature data well. The derived theoretical relation for a rough pipe with a relative roughness of 0.004 fitted the NCCL data best. Moreover, it was established from the theoretical analysis that roughness has two competitive impacts on the heat transfer of an LLM: it reduces conductive heat transfer while enhancing convective heat transfer. Because conductive heat transfer is important for liquid metals, even with turbulent flow, a small roughness will lead to heat-transfer deterioration at low Péclet numbers, and it may even deteriorate across the whole typical Péclet-number range. This discovery has important implications for the thermal–hydraulic design of LLM reactors, because corrosion and erosion by an LLM will lead to a rough surface after long operating times.

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“…During the design and safety analyses of liquid metal fast reactors, a series of issues at the assembly level still need to be addressed (Roelofs et al, 2019a;Zhang et al, 2020), and it is a fact that most of the current numerical studies focus on the normally operated subassemblies. However, the off-normal conditions also deserve great attention.…”

Section: Introductionmentioning

confidence: 99%

“…However, the off-normal conditions also deserve great attention. One of the typical off-normal conditions is a blockage accident, which can be caused by the accumulation of corrosive products in the fuel assembly (Zhang et al, 2020). Blockage accidents in LFRs may threaten the safety of reactors, leading to fuel assembly damage or melt.…”

Section: Introductionmentioning

confidence: 99%

Numerical Study on the Thermal Hydraulic Characteristics in a Wire-Wrapped Assembly of LFRs

Fang

Guo

et al. 2020

Front. Energy Res.

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The concept of wire-wrapped assembly has been widely adopted in liquid metal-cooled reactors to enhance the capability of flow and heat transfer. The detailed 3D thermal hydraulic behaviors of liquid metal in the wire-wrapped assembly directly affect the safety of the reactor. In the present work, a numerical study on a 19-pin wire-wrapped assembly cooled by a lead-bismuth eutectic was carried out using the CFD method. The geometrical model is referred to in the test section in KIT’s 19-pin experiments, which is a scaled test assembly based on the MYRRHA reactor. The established CFD model was validated against experimental data and results showed that the SST k-ω model accompanied with the modified Prt was appropriate for the current problem. Then the numerical models were extensively used in the modeling of the blocked assembly flow with a cylindrical blockage. Also, the influences of blockage longitude, position, and porosity on the thermal hydraulic characteristics in the assembly were evaluated in detail. This work is meaningful for the deep understanding of safety features under the condition of assembly blockage operation in LFRs.

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Review of Thermal-Hydraulic Issues and Studies of Lead-based fast reactors

Cited by 67 publications

References 109 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

Study of Flow and Heat Transfer Characteristics of Lead-Based Liquid Metals in a Turbulent Tube Flow and the Impacts of Roughness

Numerical Study on the Thermal Hydraulic Characteristics in a Wire-Wrapped Assembly of LFRs

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