Study on optimization of radial space between tritium breeding and neutron shielding for CFETR blanket

The China Fusion Engineering Test Reactor (CFETR) is a tokamak device designed to bridge the gap between ITER and DEMO, with the primary goal of generating useable fusion energy through D–T reactions. To balance the self-sustaining cycle requirements of CFETR tritium fuel and the irradiation safety of superconducting magnets, this paper conducts optimization research on the neutron shielding performance of the blanket and its rear structure. A 1D cylindrical neutron model was established based on the radial dimensions and material composition of the main components of CFETR. MCNP was used to calculate the radial distribution of fast neutron flux density and nuclear heating deposition under different neutron wall load (NWL) conditions, and the 3D correction factor of ITER was introduced to correct the calculation results. In the updated design, adjustments were made to the blanket thickness, breeding material ratio, cooling structure, vacuum chamber and coil box materials, and neutron shielding block. The fusion power was set to 1.5 GW, and the NWL range was set to 0.5–2.0 MW/m2. The results indicate that the optimized material and structural scheme can effectively reduce the fast neutron flux at the insulation layer of the TF coil and improve the shielding effect of the superconducting magnet. Under equivalent operating conditions of 10 full-power years and NWL of 0.5, 1.0, 1.5, and 2.0 MW/m2, the recommended shielding zone thicknesses are ∼34, 40, 45, and 48 cm, respectively. This study can provide a reference for the integrated design of CFETR blanket breeding shielding and subsequent 3D neutron optimization.