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

本项目旨在研究与聚变反应堆面向等离子体部件的水冷却有关的散热和强化传热的极限现象以及与单侧加热通道中的临界热流密度有关的现象。在1996财政年度，进行了水在均匀加热圆管和单侧加热矩形通道中流动的临界热流密度测量实验。在前者中，研究了管子内径(1 mm~6 mm)、加热长度(加热长径比：1~100)和进水过冷度(0~90K)对临界热流密度的影响，并根据烧毁位置和壁温波动的观测对临界热流密度的机理进行了讨论。在单面加热的矩形通道中，测量了临界热流密度，观察了玻璃窗内的流态。在1997财政年度，对半周加热圆管和均匀加热圆管的临界热流密度和流动沸腾换热进行了测量。根据这些结果，得出了均匀加热和半周加热的圆管之间的区别。在1998财政年度，对散裂中子源固体靶的冷却，采用加热长度较短的两侧加热矩形通道进行了临界热流密度和热传递实验，其中预计聚变反应堆面向等离子体的部件将有很高的热流密度。在这些实验中，获得了入口区和发展区单相流和沸腾两相流的换热系数，以及质量区和过冷区的临界热流密度。这些结果为圆管和矩形通道中水的强制流动换热和换热极限提供了系统和有用的信息。

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.