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Transient heat pipe failure accident analysis of a megawatt heat pipe cooled reactor

机译:兆瓦热管冷却反应器的瞬态热管故障事故分析

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摘要

Heat pipe cooled reactors, which use heat pipes rather than fluid flow for passive cooling, are an excellent candidate for micro nuclear power sources. Heat pipes provide excellent built-in redundancy and safety margin. However, a heat pipe failure is quite likely over the reactor lifetime, which is one of the essential design basis accidents in the reactor design. This work analyzes heat pipe failure accidents to investigate the system transient performance and the response of a heat pipe cooled reactor. Single heat pipe failure and cascading heat pipe failure are simulated and analyzed by establishing a transient analysis code, including a point reactor kinetics model, a core thermal-mechanical model, and a heat pipe model. The simulation shows that in the single heat pipe failure accident, the peak monolith temperature at the failed channel increases and is only 8 K lower than the fuel center temperature. In the cascading heat pipe failure accident, the temperatures in each channel suddenly increase over a short period and rise to the maximum at 22,900 s as the heat losses increase and the core power decreases. The sensitivity of the external convection for residual heat removal is also analyzed. For small external convection coefficients less than 100 W/m(2)K, increasing the external convection coefficient significantly reduces the maximum temperatures. By contrast, for external convection coefficient greater than 500 W/m(2)K, increasing external convection coefficient still reduces the core peak temperature, but the effect is weak.
机译:热管冷却反应器,用于使用热管而不是用于被动冷却的流体流动,是微核电源的优异候选者。热管提供出色的内置冗余和安全保证金。然而,热管失效很可能是反应堆寿命,这是反应堆设计中的基本设计基础事故之一。这项工作分析了热管破坏事故,以研究系统瞬态性能和热管冷却反应器的响应。通过建立瞬态分析代码,包括点电抗器动力学模型,核心热机械模型和热管模型来模拟和分析单热管故障和级联热管故障。该模拟表明,在单热管故障事故中,失败通道处的峰值整体温度增加,并且仅比燃料中心温度低8 k。在级联热管故障事故中,每个通道中的温度在短时间内突然增加,并且随着热量损失的增加并且核心功率降低,增加到22,900秒的最大值。还分析了剩余散热的外部对流的敏感性。对于小于100 W / m(2)k的小外部对流系数,增加外部对流系数显着降低了最大温度。相比之下,对于大于500W / m(2)克的外部对流系数,增加外部对流系数仍然降低了核心峰值温度,但效果较弱。

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  • 来源
    《Progress in Nuclear Energy》 |2021年第10期|103904.1-103904.18|共18页
  • 作者单位

    Tsinghua Univ Dept Engn Phys Beijing 100084 Peoples R China|Nucl Power Inst China Sci & Technol Reactor Syst Design Technol Lab Chengdu 610213 Peoples R China;

    Tsinghua Univ Dept Engn Phys Beijing 100084 Peoples R China;

    Nucl Power Inst China Sci & Technol Reactor Syst Design Technol Lab Chengdu 610213 Peoples R China;

    Nucl Power Inst China Sci & Technol Reactor Syst Design Technol Lab Chengdu 610213 Peoples R China;

    Nucl Power Inst China Sci & Technol Reactor Syst Design Technol Lab Chengdu 610213 Peoples R China;

    Tsinghua Univ Dept Engn Phys Beijing 100084 Peoples R China;

    Nucl Power Inst China Sci & Technol Reactor Syst Design Technol Lab Chengdu 610213 Peoples R China;

    Nucl Power Inst China Sci & Technol Reactor Syst Design Technol Lab Chengdu 610213 Peoples R China;

    Nucl Power Inst China Sci & Technol Reactor Syst Design Technol Lab Chengdu 610213 Peoples R China;

  • 收录信息 美国《科学引文索引》(SCI);美国《工程索引》(EI);
  • 原文格式 PDF
  • 正文语种 eng
  • 中图分类
  • 关键词

    Heat pipe cooled reactor; Transient analysis; Core thermal-mechanical model; Heat pipe failure accidents;

    机译:热管冷却反应器;瞬态分析;核心热机械模型;热管故障事故;

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