One of the important limitations in nuclear reactor design is to avoid departure of nucleate boiling (DNB) and dry-out (DO), which may cause fuel assembly cladding failure. Critical heat flux (CHF) under different condition is a complicated physical phenomenon which could not be predicted by analytical method. Experiment is the common way to acquire the CHF correlations for complicated geometry like bundle channels. To simulate the thermal hydraulic condition in an actual reactor, different scaling and geometry rod bundles are applied. The facilities of NPIC (Nuclear Power Institute of China) which carry out the CHF experiments of full length rod bundle and primary experimental results are introduced in the article.
Thermal Hydraulic Facility for Full Length Bundle CHF Experiment
- Views Icon Views
- Share Icon Share
- Search Site
Qin, S, Lang, X, Zhuo, W, Li, P, & Liu, Y. "Thermal Hydraulic Facility for Full Length Bundle CHF Experiment." Proceedings of the 2014 22nd International Conference on Nuclear Engineering. Volume 2B: Thermal Hydraulics. Prague, Czech Republic. July 7–11, 2014. V02BT09A030. ASME. https://doi.org/10.1115/ICONE22-30899
Download citation file: