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

本项目旨在研究与核聚变反应堆等离子体面组件水冷却有关的除热极限和传热增强现象以及与单侧加热通道中临界热流密度有关的现象。在1996财政年度，进行了实验，以测量水在均匀加热的圆管和单向加热的矩形通道中流动的临界热流密度。前者研究了管内内径（1mm-6mm）、受热长度（受热长径比1-100）和进水过冷（0-90K）对临界热流密度的影响，并根据燃尽位置和壁面温度波动的观察对临界热流密度的机理进行了探讨。在单侧加热矩形通道中，进行了临界热流密度的测量和通过玻璃窗的流态观测。1997财政年度，在半周加热圆管和均匀加热圆管中进行了临界热流密度和流动沸腾换热实验。从这些结果，均匀加热和半圆周加热的圆管之间的区别。在1998财政年度，利用加热长度较短的两侧加热矩形通道进行了临界热流和传热实验，以冷却散裂中子源的固体目标，其中热流预计会像在聚变反应堆的等离子体面组件中一样高。在这些实验中，获得了入口区和发达区单相和沸腾两相流的换热系数，以及质量区和过冷区临界热流密度的数据。从这些结果中，获得了关于圆管和矩形管道中强制水流的传热和排热极限的系统和有用的信息。

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.