核动力工程2024,Vol.45Issue(6) :254-262.DOI:10.13832/j.jnpe.2024.06.0254

热管失效对高温热管管束及基体影响实验研究

Experimental Study on the Impact of Heat Pipe Failure on High-Temperature Heat Pipe Bundles and Matrix

王锦源 李潘潇 王成龙 张泽秦 田文喜 秋穗正 苏光辉
核动力工程2024,Vol.45Issue(6) :254-262.DOI:10.13832/j.jnpe.2024.06.0254

热管失效对高温热管管束及基体影响实验研究

Experimental Study on the Impact of Heat Pipe Failure on High-Temperature Heat Pipe Bundles and Matrix

王锦源 1李潘潇 1王成龙 1张泽秦 1田文喜 1秋穗正 1苏光辉1
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作者信息

  • 1. 西安交通大学核反应堆热工水力研究室,西安,710049
  • 折叠

摘要

为了验证小型热管反应堆的可行性,本研究设计了一种高温热管管束的实验装置.该实验装置通过拔出热管模拟热管失效,由电加热棒模拟堆芯燃料棒,用以探究热管失效对热管管束、燃料棒及堆芯基体的影响.实验发现:热管失效带来的最直接影响是堆芯基体局部温度突升,在功率 4.2 kW下,单管失效时附近基体温度平均上升约 70℃,双管失效时附近基体温度平均上升约 120℃;单管失效对其余正常热管影响较小,正常热管蒸发段平均温升 15℃,双管失效时,与失效热管相邻的燃料元件平均温升约 66℃.本文获得的热管失效下的高温热管管束实验数据可为热管堆的建模仿真提供数据支持.

Abstract

In order to verify the feasibility of small heat pipe reactor,an experimental device of high temperature heat pipe bundle is designed in this study.The experimental device simulates the failure of heat pipe by pulling out the heat pipe,and simulates the core fuel rod by electric heating rod to explore the influence of heat pipe failure on heat pipe bundle,fuel rod and core matrix.It is found that the most direct impact of heat pipe failure is a sudden local temperature rise in the nearby matrix.Under the power of 4.2 kW,the average increase of matrix temperature in the vicinity of single heat pipe failure is about 70℃,and the average increase of matrix temperature in the vicinity of double heat pipe failure is about 120℃.The failure of a single heat pipe has a minor impact on the remaining normal heat pipes,with an average temperature rise of 15℃in the evaporation section of the normal heat pipes.The average temperature increase of fuel element in the vicinity of double heat pipe failure is about 66℃.The experimental data of high temperature heat pipe bundle under heat pipe failure obtained in this study can provide data support for the modeling and simulation of heat pipe reactor.

关键词

热管反应堆/热管管束/热管失效/瞬态分析

Key words

Heat pipe reactor/Heat pipe bundle/Heat pipe failure/Transient analysis

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出版年

2024
核动力工程
中国核动力研究设计院

核动力工程

CSTPCD北大核心
影响因子:0.3
ISSN:0258-0926
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