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A thermal neutronics coupling analysis method for lead based reactor core

机译:铅基反应堆堆芯的热中子耦合分析方法

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摘要

In this research paper, a sub-channel thermal hydraulic analysis code is coupled with the point reactor neutron kinetics model with six group delayed neutron. The coupling code is mainly used to perform the transient calculation of ADS/lead based alloy cooled fast reactor. The thermal hydraulic model is used for calculating temperature distribution profile and the feedback temperature information, providing input parameters for point kinetic model. This sub-channel analysis model can provide a new approach to solve the problem of one-dimension thermal hydraulic model and simulate the temperature distribution accurately. Furthermore the accuracy and reliability of calculated results are tested by another coupled code named FLUENT/PK and good agreements are achieved. To improve computational speed, one equivalent assembly is used to replace the whole core and the study shows that using of equivalent assembly which has the same average outlet temperature with the core obtained more reasonable results. The effects of fuel rods pitch diameter P/D ratio on simulation results are discussed. The code is capable to the quick calculations and safety analysis for reactivity accidents. (C) 2017 Elsevier Ltd. All rights reserved.
机译:在本文中,子通道热工水力分析代码与具有六组延迟中子的点反应堆中子动力学模型耦合。耦合代码主要用于进行ADS /铅基合金冷却快堆的瞬态计算。热工水力模型用于计算温度分布曲线和反馈温度信息,为点动力学模型提供输入参数。该子通道分析模型可以为解决一维热工水力模型问题和准确模拟温度分布提供一种新方法。此外,通过另一个名为FLUENT / PK的耦合代码测试了计算结果的准确性和可靠性,并取得了良好的一致性。为了提高计算速度,使用了一个等效组件代替了整个磁芯,研究表明,使用与磁芯具有相同平均出口温度的等效组件可获得更合理的结果。讨论了燃料棒节径P / D比对模拟结果的影响。该代码能够对反应事故进行快速计算和安全分析。 (C)2017 Elsevier Ltd.保留所有权利。

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