The nuclear heating reactor (NHR), which is independently developed and designed by Tsinghua University, adopts full power natural circulation and has the characteristic of low mass flow rate. Due to the different operating conditions between NHR and PWR, it is necessary to analyze the thermal hydraulic performance of the fuel assembly in NHR core. The thermal hydraulic performance of the fuel assembly is mainly analyzed by the subchannel analysis code which can accurately calculate the two-phase parameters. Therefore, this research employs subchannel analysis code COBRA-TF to simulate radial uniform and non-uniform heating assembly under the operating conditions of low mass flow rate from 300kg/m2s to 700kg/m2s. The simulations are validated on the thermal hydraulics experiments of GE and Studsvik. The result shows that the calculated mass flow rate and void fraction are in good agreement with the experimental results. Under radial nonuniform heating condition, high outlet mass velocity appears in low heating power region, while the high outlet void fraction appears in high heating power region. Compared with the uniform heating condition, the distribution of the mass flow rate and the void fraction in subchannels under non-uniform heating conditions varies greatly. In addition, the transverse flow and the counterflow also can be observed in non-uniform heating conditions under low mass flow rate conditions. The appearance of such transverse flow mainly results from the radial pressure difference, rather than the turbulent mixing and void drift. Therefore, it is essential to predicate the distribution of the mass flow rate and the void fraction under non-uniform heating and low mass flow rate conditions.