Experiments on heat transfer in rod-bundle flows as supercritical pressures

这项拟议的研究是为了支持加拿大国家第四代能源技术计划，该计划旨在开发一种超临界水冷反应堆（SCWR），与现有的核反应堆相比，这种反应堆有望提高安全性，降低电力生产成本，体积更小，核废料量更少。本研究的主要目的是量化近临界和超临界压力下三杆束流动中传热的以下三个重要影响：1)流动方向的影响，即向下流动与向上流动；Ii)绕丝垫片的作用；压力、质量通量和热通量的瞬变效应。这些影响以前没有研究过与SCWRs相关的条件；为了改进SCWR安全分析规范，从而使SCWR燃料在正常和事故条件下的热行为预测更加现实，需要对其进行量化。所有的实验研究都将在渥太华大学的超临界循环中进行，以二氧化碳作为替代流体。

The proposed research is in support of the Canadian National Program on Generation IV Energy Technologies for the development of a Super Critical Water-cooled Reactor (SCWR), which, compared to existing nuclear reactors, is expected to have increased safety, lower-cost electricity production, more compact size and reduced volume of nuclear wastes. The main objective of this research is to quantify the following three important effects on the heat transfer in flows in a three-rod bundle under near-critical and supercritical pressures: i) effect of flow direction, namely downflow vs. upflow; ii) effect of wire wrap spacers; and iii) effect of transients in pressure, mass flux and heat flux. These effects have not been investigated previously for conditions relevant to SCWRs; they need to be quantified in order to improve the SCWRs safety analysis codes and thus make their prediction of thermal behaviour of the SCWR fuel during normal and accidental conditions more realistic. All experimental investigations will be performed in the University of Ottawa's supercritical loop with carbon dioxide as a surrogate fluid.