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

拟议的研究是为了支持加拿大第四代能源技术国家方案，以开发超临界水冷反应堆（SCWR），与现有的核反应堆相比，预计该反应堆将具有更高的安全性，更低的电力生产成本，更紧凑的尺寸和减少的核废物量。本研究的主要目的是量化以下三个重要的影响，在近临界和超临界压力下的三棒束中的流动的传热：i）流动方向的影响，即向下流动与向上流动; ii）金属丝缠绕间隔的影响;和iii）压力，质量通量和热通量的瞬变的影响。这些影响以前没有研究过与超临界水反应堆相关的条件;它们需要被量化，以改善超临界水反应堆安全分析代码，从而使它们在正常和事故条件下对超临界水反应堆燃料的热行为的预测更加现实。所有的实验研究将在渥太华大学的超临界回路中进行，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.