Thermal-hydraulic analysis of a representative Westinghouse lead-cooled fast reactor fuel bundle using CFD

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design Pub Date : 2025-05-05 DOI:10.1016/j.nucengdes.2025.114117

D. Wilson , H. Iacovides , E. Tatli , C. Stansbury , P. Ferroni

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引用次数: 0

Abstract

Lead-cooled fast reactors (LFRs) are Generation IV reactor technologies that use molten lead as the primary coolant. Whilst lead offers advantages for economics, safety, and sustainability, its low Prandlt number and challenging experimental characteristics pose difficulties for thermal–hydraulic modelling and validation. To support LFR development, this study aims to advance modelling capabilities and understanding of the relevant physical phenomena through a series of Computational Fluid Dynamics (CFD) simulations of a Fuel Pin Bundle Simulator (FPBS) that shares design features with the Westinghouse LFR fuel assembly.

Three geometrical configurations of the FPBS have been modelled, using the Reynolds-averaged Navier-Stokes (RANS) approach: a bare pin bundle (without spacer grids), a T-junction upstream of the main test section, and the full-length 360° main FPBS test section including spacer grids and instrumentation wires. The sensitivity of the results to modelling choices, including turbulence models and approaches for the turbulent Prandtl number, is explored.

The original contributions of this study are in the assessment of different RANS models of the Reynolds stresses, the assessment of different values and functions of the turbulent Prandtl number for the modelling of the turbulent heat fluxes, the exploration of the entry conditions on the flow and thermal development along the fuel bundle and the determinations of the effects of the intrusive instrumentation on the measured quantities.

The bare bundle simulations showed only minor sensitivity to the turbulence model and produced friction factors in excellent agreement with existing correlations. Predictions in the upstream T-junction indicated the generation of significant swirl that enters the main test section, but the spacer grid acts as an effective flow straightener. Nusselt number predictions in the main FPBS test section showed good agreement with established correlations for liquid metal rod bundles. Instrumentation wires had only a minor effect on the temperature field and increased the pressure drop by 2.7 %. A sensitivity analysis of the turbulent Prandtl number (

\Pr_{t}

) showed that Kay’s correlation produced Nusselt numbers that were closest to the empirical correlation of Ushakov et al. (1977), with a mean deviation of 1.1 %. In contrast, a constant

\Pr_{t} = 0.9

resulted in an overprediction of 19 %.

查看原文本刊更多论文

基于CFD的典型西屋铅冷快堆燃料束热水力分析

铅冷快堆（LFRs）是使用熔融铅作为主冷却剂的第四代反应堆技术。虽然铅具有经济性、安全性和可持续性的优势，但其低普朗特数和具有挑战性的实验特性给热液建模和验证带来了困难。为了支持LFR的发展，本研究旨在通过一系列燃料销束模拟器（FPBS）的计算流体动力学（CFD）模拟来提高建模能力和对相关物理现象的理解，FPBS与西屋公司的LFR燃料组件具有相同的设计特征。采用reynolds -平均Navier-Stokes （RANS）方法，对FPBS的三种几何构型进行了建模：裸引脚束（不含间隔网格）、主测试段上游的t型结、以及包含间隔网格和仪表线的全长360°FPBS主测试段。结果对模型选择的敏感性，包括湍流模型和湍流普朗特数的方法，进行了探讨。本研究的原始贡献在于评估雷诺应力的不同RANS模型，评估湍流普朗特数的不同值和函数以模拟湍流热通量，探索沿燃料束流动和热发展的进入条件以及确定侵入式仪器对测量量的影响。裸束模拟对湍流模型的敏感性很小，产生的摩擦因子与现有的相关性非常吻合。在上游t型交界处的预测表明，产生了明显的涡流进入主测试段，但间隔网格起到了有效的流动矫直器的作用。FPBS主测试段的Nusselt数预测与液态金属棒束的建立相关性很好地吻合。仪表线对温度场的影响很小，并使压降增加了2.7%。对紊流普朗特数（Prandtl number, Prt）的敏感性分析表明，Kay的相关得到的Nusselt数最接近Ushakov et al.（1977）的经验相关，平均偏差为1.1%。相反，恒定的Prt=0.9导致19%的过度预测。

本文章由计算机程序翻译，如有差异，请以英文原文为准。

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来源期刊

Nuclear Engineering and Design 工程技术-核科学技术

CiteScore

3.40

自引率

11.80%

发文量

377

审稿时长

5 months

期刊介绍： Nuclear Engineering and Design covers the wide range of disciplines involved in the engineering, design, safety and construction of nuclear fission reactors. The Editors welcome papers both on applied and innovative aspects and developments in nuclear science and technology. Fundamentals of Reactor Design include: • Thermal-Hydraulics and Core Physics • Safety Analysis, Risk Assessment (PSA) • Structural and Mechanical Engineering • Materials Science • Fuel Behavior and Design • Structural Plant Design • Engineering of Reactor Components • Experiments Aspects beyond fundamentals of Reactor Design covered: • Accident Mitigation Measures • Reactor Control Systems • Licensing Issues • Safeguard Engineering • Economy of Plants • Reprocessing / Waste Disposal • Applications of Nuclear Energy • Maintenance • Decommissioning Papers on new reactor ideas and developments (Generation IV reactors) such as inherently safe modular HTRs, High Performance LWRs/HWRs and LMFBs/GFR will be considered; Actinide Burners, Accelerator Driven Systems, Energy Amplifiers and other special designs of power and research reactors and their applications are also encouraged.