A Preliminary Analysis Of The Source Terms Of The Pebble Bed Fluoride Salt-cooled High-temperature Reactor In Normal Operation

Source term is the actual release or the potential release information of the radioactive substances from a given source to the environment. The release information contains the type, quantity and form of the nuclides and other characteristics varying with time. The source term analysis of a reactor is an important part of the reactor safety analysis, which can provide a basis for reactor design, operation, decommissioning, waste management, environmental impact analysis and emergency planning. Fluoride salt-cooled high-temperature reactor or FHR is a new type of reactor, which is still in the preliminary design and research stage. FHR is different from other types of reactors in many aspects such as core design, material composition and operation conditions. For example, the molten FLi Be salt coolant produces tritium when irradiated by neutrons. In the high-temperature environment, metal structure is highly permeable to tritium. This phenominon should be seriously considered in the FHR design. In this thesis, we established a preliminary method to calculate the source terms of the pebble-bed FHR during normal operation, We used the 10 MWth solid fuel molten salt reactor or TMSR-SF1 designed by the Shanghai Institute of Applied Physics, Chinese Academy of Sciences as the baseline design and obtained the information of the radioactive source terms of the reactor during normal operation.The transport-burnup coupling method was adopted to calculate the inventory of the fission products and the activated products of the in-pile structure material. The pressure vessel damage model was used to evaluate the performance of the TRISO coated particle fuel under irradiation, and the impact of burnup, temperature and attribute of SiC layer on the performance of TRISO was discussed. The diffusion model was applied to calculate the migration of nuclides from fuel to the primary coolant. The transport equations of nuclides in the TMSR-SF1 were established and the distributions of radioactive nuclides in the primary coolant, cover gas and confinement were calulated. The radioactivity of 60 Co accumulated on the inner wall of the heat exchanger was analyzed using the temperature gradient mass transfer theory.In this thesis, the source terms within the boundary of the confinement were calculated assuming the TMSR-SF1 operates for 200 days at its full power of 10 MWth. The results of the calculation show no damage to the fuel since the reactor opearates in low burnup and low fuel temperature condition. Manufacturing defects, heavy metal contaimination and isotope migrations are the main reasons for the fission products to enter the primary coolant. The radioactive nuclides in the primary coolant come from the fuel, the activation of the coolant itself, and the activation of the corrosion products of the structural materials. The majority of the radioactive nuclides remain in the coolant salt and the rest of them, notably, noble gas, iodine, tritium and some noble metals, will enter the coolant cover gas. The cover gas of the primary coolant is the major radioactive source term of the potential environmental release. Tritium release, 14 C production, 60 Co accumulation in the heat exchanger, and the high-level radiation of the primary coolant during full power operation are issues that need careful considerations in the source term analysis.