Experimental study for critical heat flux in 2x2 rod bundles at high pressure conditions

• Published in: Nuclear engineering and design 2020-08, Vol.365 (C), p.110730, Article 110730
• Main Authors: Lyons, Kathleen, Lee, Donghwi, Anderson, Mark
• Format: Article
• Language: English
• Subjects: Bundling; CHF correlation; Fluctuations; Heat; Heat flux; Heat transfer; High pressure; Nuclear reactors; Pressure; Pressurized water reactor; Reactor design; Rod bundle; Water-cooled nuclear reactor
• ISSN: 0029-5493; 1872-759X
• DOI: 10.1016/j.nucengdes.2020.110730

•CHFs were obtained in a 2x2 rod bundle at high pressure conditions for PWR.•CHFs were evaluated in various subcooling, mass flux, and pressure conditions.•The effects of subcooling, mass flux, and pressure on CHF were analyzed.•CHF prediction shows good agreement CHF data within 20% of experimental values. 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.