Fuel transport is an indispensable task for nuclear power plants. For pressurized water reactors (PWR) and boiling water reactors (BWR), many research projects have been completed for designing and testing the transport casks for fresh fuel as well as spent fuel [1–3]. To ensure the safety of nuclear fuel during the transportation, many aspects should be analyzed and examined for the casks with fuel inside, such as heat transfer and temperature calculation, radiation protection, nonproliferation issues, etc. The transport cask discussed in this paper is especially for new spherical fuel elements, which should be designed in accordance with the stipulations in the GB11806 Regulations for the Safe Transport of Radioactive Material [4].
The Transport Cask for spherical fuel elements used in molten salt reactor (MSR) should be designed in accordance with the safety standards for transport of radioactive material. It is necessary to evaluate the thermal performance of the transport cask separately in normal transport condition and in accident transient. The MSR fuel sphere elements cask is in a circular cylinder shape and composed of inner container and outer shell cask. The objective of the thermal analysis of the cask under hypothetical accident conditions is to demonstrate that the cask containment boundary structural components are maintained within their safe operating temperature ranges. The heat transfer process (conduction, convection, and radiation) is simulated by ANSYS-CFX in this paper and it is demonstrated that the components of cask are maintained in safe operating temperature ranges. The calculation results are below limit temperatures, indicating that the thermal design of the cask could meet the Standard Regulations. The result is also compared with the fire test, which shows the calculation model is conservative and rational.
Densification of matrix graphite for spherical fuel elements used in molten salt reactor via addition of green pitch coke
