Neutronic analyses of port impact on blankets and superconducting coils of CFETR

Abstract

Abstract Chinese Fusion Engineering Test Reactor (CFETR) is an ITER-like test superconducting TOKAMAK fusion reactor. In CFETR, blankets face core plasma directly and are in charge of tritium breeding and neutron shielding, while the condition of superconducting coils needs to be maintained in a low temperature for providing a steady magnetic field to torus. Thus, doing neutronic analyses for blankets and superconducting coils is important for the safety and steady operation of fusion reactor. In addition, ports are used for remote handing, plasma diagnose and other necessary operations, which will cause a different neutron transport behavior in the reactor. In order to evaluate the neutronic performance of CFETR in port opening operations, port impact on blankets and superconducting coils is necessary to be analyzed. In this paper, global TBR and neutron radiation damage for blanket first walls and nuclear heat deposition for toroidal field coils are calculated in different port schemes. Different port opening scenarios are also given in order to meet the requirement of tritium self-sufficiency. According to the result, nuclear heat on the toroidal field coils near the top port exceeds the limited value. Thus, neutronic protection to decrease the leakage of neutron from top port cannot be ignored. Neutronic shielding of superconducting is necessary to be strengthen as well.