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Study on the Limit of Water Cooling of High Heat Flux Components

高热通量部件水冷极限研究

基本信息

批准号：
07458114
负责人：
MISHIMA Kaichiro
金额：
$ 4.03万
依托单位：
KYOTO UNIVERSITY
依托单位国家：
日本
项目类别：
Grant-in-Aid for Scientific Research (B)
财政年份：
1995
资助国家：
日本
起止时间：
1995 至 1997
项目状态：
已结题

来源：
https://kaken.nii.ac.jp/grant/KAKENHI-PROJECT-07458114/
关键词：
heat transfer enhancement limit of heat removal critical heat flux one-side heated uniformly-heated fusion reactor components solid target of intense neutron source 半周加熱

项目摘要

This project aims at investigation phenomena related to the limit of heat removal and heat transfer enhancement as well as those related to critical heat flux in one-side heated channels in connection with the water cooling of plasma facing components of a fusion reactor. In Fiscal Year 1996, experiments were performed to measure critica heat flux for water flowing in a uniformly-heated round tube and a one-side heated rectangular channel. In the former, effects of tube inner diameter (1mm-6mm), heated length (heated length-to-diameter ratio : 1-100) and inlet water subcooling (0-90K) on critical heat flux were investigated, and some discussions were given on the mechanism of critical heat flux based on the burnout location and the observation of wall-temperature fluctuations. In the one-side heated rectangular channel, measurement of critical heat flux and observation of flow pattern through the glass window were performed. In Fiscal Year 1997, experiments were conducted to measure critical heat flux and flow boiling heat transfer in a half-circumferenially heated round tube and a uniformly heated round tube. From these results, the difference between uniformly heated and half-circumferentially heated round tubes. In Fiscal Year 1998, critical heat flux and heat tranfer experiments were conducted with use of a both-sides heated rectangular channel with a short heated length in relation to the cooling of a solid target of a spallation neuton source in which very high heat flux as in a plasma facing component of a fusion reactor is anticipated. In these experiments, data were taken for heat transfer coefficients for sigle phase and boiling two-phase flows in the entrance and developed regions, as well as critical heat flux in the quality and subcooled regions. From these results, systematic and useful information was obtained for heat transfer and the limit of heat removal by forced flow of water in round tubes and rectangular channels.

本项目旨在研究与聚变反应堆面向等离子体部件的水冷连接的单侧加热通道中的排热和传热增强极限以及与临界热通量相关的现象。在1996财政年度，进行了实验，以测量水在均匀加热的圆管和一侧加热的矩形通道中流动的临界热流。在第一部分中，研究了管内径（1 mm ~ 6 mm）、加热长度（加热长径比1-100）和进水过冷度（0- 90 K）对临界热流密度的影响，并根据燃烬位置和壁温波动对临界热流密度的影响机理进行了讨论。在单侧加热的矩形通道中，进行了临界热流密度的测量和通过玻璃窗的流型观察。在1997财政年度，进行了实验，以测量临界热流密度和流动沸腾传热在半周向加热圆管和均匀加热圆管。从这些结果，均匀加热和半周向加热圆管之间的差异。在1998财政年度，临界热通量和热传递实验进行了使用两侧加热矩形通道与短加热长度与冷却的固体靶的spiritoneuton源，其中非常高的热通量，如在等离子体面对组件的聚变反应堆中的预期。在这些实验中，单相和沸腾的两相流在入口和发达地区的传热系数，以及在质量和过冷区的临界热通量的数据。从这些结果中，获得了系统的和有用的信息，传热和限制的水强制流动的传热在圆管和矩形通道。