Abstract: Power Ramp Test（PRT） of a fuel element is generally conducted with a PRT irradiation rig in a research reactor, to study the fuel behaviour and to verify its safety under power transient. Neutronics characteristics of the PRT irradiation rig within a typical HFETR（High Flux Engineering Test Reactor） core and the heat generation rates of the rig’s components are calculated with MCNP code in this paper. The range of the test fuel rod power is obtained with a coupled Nuclear-Thermal-Hydraulic calculation method which combines MCNP and CFX code.The results show that the ³He gas layer influences the neutron field intensely by reducing the thermal neutron current into the layer and decreasing the neutron flux in and near the irradiation rig. Changing the density of ³He gas can vary the PRT and its periphery neutron field, consequently changing the test fuel rod power effectively. Power of the fuel pellet in the test rod increases monotonically along with the ³He gas pressure reducing, and its calculation results have good agreement with the curve fitting by a natural logarithm function.