Abstract Critical heat flux (CHF) phenomena has been a topic of study for over fifty years yet remains a serious concern across industries and especially in water-cooled nuclear reactor design. CHF data in the high-pressure range is limited, and the effects of bundle geometry and non-uniform heat flux profile are challenging to quantify. In this study, CHF experiments were performed in upward flowing water in a 2 × 2 rod bundle with a cosine profile heat flux. Data were collected between 16.5 and 18 MPa at two mass flux conditions with three inlet subcooling conditions. Under these conditions, average CHF was shown to decrease with increasing pressure. However, pressure had less of an effect than decreasing the inlet subcooling or the mass flux, both of which reduce the CHF value considerably. Interestingly, correlations that have been developed for lower pressure continued to predict CHF occurrences with moderate accuracy outside their ranges of validity. Nearly all predicted values were within 20% of experimental values.

中文翻译：

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高压条件下 2x2 棒束临界热通量的实验研究

摘要 50 多年来，临界热通量 (CHF) 现象一直是研究的主题，但仍然是各行各业的严重问题，尤其是在水冷核反应堆设计中。高压范围内的 CHF 数据有限，而且束几何形状和非均匀热通量分布的影响难以量化。在这项研究中，CHF 实验是在具有余弦分布热通量的 2 × 2 棒束中向上流动的水中进行的。在 16.5 到 18 MPa 的两个质量通量条件和三个入口过冷条件下收集数据。在这些条件下，平均 CHF 会随着压力的增加而降低。然而，与降低入口过冷度或质量通量相比，压力的影响要小，这两者都会显着降低 CHF 值。有趣的是，为较低压力开发的相关性继续以中等准确度预测 CHF 发生率超出其有效性范围。几乎所有预测值都在实验值的 20% 以内。