抄録
The objective of this paper is a study on thermal-hydraulic and neutronic aspects of a pressure-channel Supercritical Water-cooled Reactor (SCWR) with a focus on determination of fuel and cladding/sheath temperatures as well as a pressure drop across a fuel channel. With these objectives, a thermal-hydraulic code has been developed in MATLAB, which calculates a fuel centerline temperature, sheath temperature, coolant temperature and heat-transfer-coefficient profiles. The developed thermal-hydraulic code is coupled with a lattice code and a diffusion code. The neutronic codes were used in order to determine a power distribution inside the core. This paper presents a fuel centerline temperature of a 73-element fuel bundle with UO_2 as a reference fuel, while results are presented for high thermal-conductivity fuels such as UC and UO_2+SiC. The results show that the maximum fuel centerline temperature is significantly lower for high thermal-conductivity fuels. The total pressure drop varied between 108 to 133 kPa per fuel channel.
