Abstract A code-to-code comparison study was performed to investigate the thermal hydraulic characteristics of the JRR-3 reactor. The COOLOD-N2 and TMAP codes were used to analyze the JRR-3 reactor under downward forced convective and natural convective flows. The geometry and operational conditions of JRR-3 used in the present study were provided by the NEA (Nuclear Energy Agency) data bank. The total power peaking factors such as the radial and axial power peaking and the engineering hot channel factors were implemented to simulate hot and average channels. Thermal hydraulic characteristics, i.e., coolant and cladding wall surface temperatures, and the flow velocity and thermal margins, i.e., minimum ONB (Onset of Nucleate Boiling) temperature margin and minimum DNB (Departure from Nucleate Boiling) ratio were estimated at a nominal power of 20 MW for forced convection cooling and at 0.2 MW for natural convection cooling. As a result, a comparison study between the COOLOD-N2 and TMAP codes showed a comparability of the thermal hydraulic analysis results except for the ONB temperature margin and the coolant velocity in the hot channel during natural convection cooling. The differences resulted from the present analyses were discussed.