Safety is the biggest concern of nuclear power plant and research reactors.As one of the three primary thermal design criteria for rectors, the critical heat flux (CHF) guarantee the reactor’s safety and economics. The bubble dynamics in narrow rectangular channel is didderent from that in conventional channels. Therefore the CHF characteristics obtained in the conventional channel may not be suitable in the narrow rectangular channel. To analyze CHF characteristics in the narrow rectangular channel, a visual experiment study on CHF was carried out in one side heated narrow rectangular channel under upflow condition condition. The experiments were performed at pressures ranging from 1 to 4 MPa, with inlet subcooling ranging from 65 to 120 K and mass flux ranging from 350–200 kg/(m2s). The deionized water was used as the working medium. The relevant thermal-hydraulic parameters and visualization results were collected synchronously. Then compare the CHF experimental values with the predicted values of the W-3 correlation,Mishima correlation, and Chang H correlation. The result shows the predicted values of W-3 correlation and Mishima correlation are always larger than the experimental values. Among these corrrelations, we found that the errors of Chang H correlation is within 30% in 2mm gap size.