The supercritical pressure light water cooled fast reactor (Super Fast Reactor), which is a once-through water cooled fast reactor supplying supercritical pressure steam at high temperature to the turbine system, has an innovative and economical potential. During startup process and shutdown procedure of the Super Fast Reactor, the pressure of water passes through the critical pressure condition. In the region of the pressure slightly below the critical pressure, critical heat flux condition leading to the rapid rise of fuel rod surface temperature tends to occur at relatively low heat flux. In the present study, the critical heat flux at the near-critical pressure is measured for vertically upward flow in circular-tube and sub-bundle channels, in order to understand the critical heat flux condition in the near-critical pressure region. HCFC22 is used as the test fluid. HCFC22 is easy for handling due to its low critical pressure and temperature, and therefore the experimental conditions can be set easily to make systematic data. Based on the experimental data, characteristics of the critical heat flux are clarified, including maximum wall temperature and critical enthalpy changes with pressure, effects of heat flux and mass flux, the difference between pressure-increasing and pressure-decreasing processes, the comparison between tube and bundle channels, and the comparison of measured data with predicting correlation.