Abstract
The rod assembly of nuclear heating reactor (NHR-200II) has great power distribution differences over the lifetime of operating process. Therefore, it is essential and important to predict the critical heat flux of different power distribution of rod assembly accurately under NHR-200II operating parameters, which includes non-uniformly radial and axial heating rod assembly. In this study, based on the 5 × 5 non-uniformly heating rod assembly layout of nuclear heating reactor, the CHF of rod assembly was predicted by using subchannel analysis code CTF and the CHF empirical correlations. The results show that CTF combines EPRI empirical correlation method can better predict the CHF experimental results of typical non-uniformly heating assembly, while the Barnett relation has a large error in predicting CHF. The average critical heat flux of rod assembly decreases with the increase of radial power factor ratio. On one hand, in different radial power distribution, predicted CHF and DNBR in the subchannel gradually decrease along the axial height, and CHF occurs at the outlet of the subchannel near the highest radial heating power rod. At this time, the lower mass flow rate, higher local quality and heating power lead to the occurrence of CHF. On the other hand, in different axial power distribution, CHF occurs at the outlet and middle of the subchannel near the highest radial heating power rod in outlet and inlet axial peak distribution separately. The local quality and heating power achieve relatively high values in middle position of inlet axial peak distribution, which leads to the occurrence of CHF.
