Heat Transfer to Supercritical Water (Liquid-Like State) Flowing in a Short Vertical Bare Tube With Upward Flow

A. Zvorykin, M. Mahdi, Roman Popov, K. Barati Far, I. Pioro
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引用次数: 7

Abstract

Current Nuclear Power Plants (NPPs) equipped with water-cooled reactors (the vast majority of all NPPs) have relatively low thermal efficiencies within the range of 30–36% compared to those of modern advanced thermal power plants (SuperCritical Pressure (SCP) coal-fired — up to 55% thermal efficiency and combined cycle — up to 62%). Therefore, next generation reactors / NPPs should have higher thermal efficiencies close to those of current thermal power plants. Around 60 years ago thermal-power industry has moved from subcritical pressures to SCPs with the major objective to increase thermal efficiency. Based on this proven in power industry experience it was proposed to design SuperCritical Water-cooled Reactors (SCWRs), which are one of the six Generation-IV nuclear-reactor concepts under development in selected countries. These days, there are discussions on developing even Small Modular Reactors (SMRs) of SCPs. In spite of a large number of experiments in long bare tubes (pipes) cooled with SCW, developing SCWR concepts requires experimental data in bundle geometries cooled with SCW, which are usually shorter and will have smaller diameters. However, such experiments are extremely complicated and expensive plus each bundle geometry will have a unique Heat-Transfer (HT) characteristics due to various bundle designs. Therefore, as a preliminary and a universal approach — experiments in bare tube of shorter heated lengths and of smaller diameters to match heated lengths and hydraulic-equivalent diameters of fuel bundles are required. Current paper provides experimental data obtained in a short (0.6 m) vertical bare tube of a small diameter (6.28 mm) cooled with upward flow of SCW. Analysis of this dataset is also included. Main emphasis of this research is on liquid-like cooling within the possible conditions of future SCWRs and SCW SMRs. Two HT regimes are encountered at these conditions: 1) Normal HT (NHT) and 2) Deteriorated HT (DHT). Conditions at which the DHT regime appeared are discussed.
超临界水(液态)在短垂直裸管内向上流动的传热研究
目前配备水冷堆的核电站(绝大多数核电站)的热效率相对较低,在30-36%的范围内,与现代先进的热电厂(超临界压力燃煤电厂-热效率高达55%,联合循环电厂-高达62%)相比。因此,下一代反应堆/核电站的热效率应该接近当前的热电厂。大约60年前,火电工业已经从亚临界压力转向SCPs,主要目标是提高热效率。基于这一在电力工业中得到证实的经验,建议设计超临界水冷堆(SCWRs),这是选定国家正在开发的六个第四代核反应堆概念之一。最近,甚至出现了开发小型模块化反应堆(smr)的讨论。尽管在长裸管(管道)中进行了大量的SCW冷却实验,但开发SCWR概念需要用SCW冷却的管束几何形状的实验数据,这些管束通常更短,直径更小。然而,这样的实验是非常复杂和昂贵的,并且由于不同的束设计,每个束的几何形状将具有独特的传热(HT)特性。因此,作为一种初步和通用的方法,需要在加热长度较短和直径较小的裸管中进行实验,以匹配加热长度和燃料束的水力等效直径。本论文提供的实验数据是在一个直径较小(6.28 mm)的短(0.6 m)垂直裸管中,用SCW向上流动冷却得到的。还包括对该数据集的分析。本研究的重点是在未来scwr和scwsmr的可能条件下进行类液体冷却。在这些条件下会遇到两种高温状态:1)正常高温(NHT)和2)恶化高温(DHT)。讨论了DHT状态出现的条件。
本文章由计算机程序翻译,如有差异,请以英文原文为准。
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