Inverted annular two-phase flow in multiphase flow systems
Inverted annular two-phase flow in multiphase flow systems
复制标题
多相流系统中的倒环形两相流
DOI：
10.1016/j.ijheatmasstransfer.2021.122340
复制
发表时间：
2022
影响因子：
5.2
通讯作者：
T. Hibiki
中科院分区：
文献类型：
--
作者：
S. Miwa;T. Hibiki
了解倒环流 (IAF) 沸腾状态对于各种工程学科至关重要。 IAF 是管或通道中强制对流流动中的干燥后或 CHF 后薄膜沸腾状态。 IAF 的流动结构的特点是通道中心存在液体核心，通过蒸汽膜包层将液体核心与周围的热源隔开。自 20 世纪 50 年代以来，它作为薄膜沸腾方式之一得到了广泛的研究。在轻水反应堆的堆芯再淹没阶段，如果发生冷却剂损失事故（LOCA），主要的传热模式之一是膜沸腾。在此过程中，流道中会产生各种流态，包括 IAF。 IAF 立即出现在 CHF 前流动状态的下游，并在骤冷前沿开始。当它向下游流动时，它逐渐发展成倒置的段塞流和分散流。为了提高工程系统（包括现代和先进核反应堆）的安全性，正确利用 IAF 的传热系数和阻力系数对于假定的 LOCA 至关重要。本研究广泛回顾了 IAF 传热和阻力的现有相关性。此外，还审查了基于液体射流破碎概念的适用于 IAF 的流态转变标准。现代反应堆安全分析代码中使用的 IAF 相关性也将被重新审视。
Understanding the inverted annular flow (IAF) boiling regime is crucial for various engineering disciplines. The IAF is a post-dryout, or post-CHF film boiling regime in forced convective flow in tubes or channels. The flow structure of the IAF is characterized by the liquid core present in the channel center, which is separated from the surrounding heating source by a vapor film envelope. It has been studied extensively as one of the film boiling regimes since 1950s. In the core reflooding phase of the light water reactors, in the event of the loss-of-coolant-accident (LOCA), one of the dominant heat transfer modes is film boiling. During that process, various flow regimes are generated in the flow channel, including the IAF. The IAF appears immediately downstream of the pre-CHF flow regime and initiates at the quench front. It gradually progresses to an inverted slug and dispersed flow as it travels downstream. For the improved safety of engineering systems, including modern and advanced nuclear reactors, proper utilization of heat transfer and drag coefficients of IAF for the postulated LOCA is crucial. The present study extensively reviews the existing correlations proposed for IAF heat transfer and drag. In addition, the flow regime transition criteria applicable for IAF, based on the liquid jet-breakup concept, are reviewed. IAF correlations utilized in modern reactor safety analysis codes will also be revisited.