THERMAL-HYDRAULIC ANALYSIS OF A HIGH-TEMPERATURE HELIUM-GAS-COOLED REACTOR UNDER VARIOUS STEADY-STATE OPERATING CONDITIONS

This paper focuses on thermal-hydraulic analysis, which plays a critical role in system efficiency and the selection of the optimal design of nuclear reactors. The analysis is done based on a one- dimensional computer code called MIGHT that performs a subchannel thermal-hydraulic analysis of a typical gas-cooled fast breeder reactor (GCFBR) cooled by helium (He). In steady-state operation, two typical channels, the hot and average channels, with the same flow rate and pressure drop were tested. Temperature distribution profiles and the heat flux were computed and compared for different types of power distribution. The effects of coolant mass flow rate and power level on the thermal-hydraulic performance of the tested GCFBR were investigated for cosine power profile. The results demonstrate that the lowest flow rate for the tested reactor to continue operation in the safe mode at the nominal operating power ( 2530 MWt) is 80% of the nominal flow rate (10 x 10 6 kg/ h). The maximum cladding temperature stays within the suggested design limit of GCFRs (700-750 degrees C) when the power is increased by 10% and 15%. The results revealed that temperature is more sensitive to changes in power level than mass flow rate. Data of GCFBR typical reactor were used as input data and for code validation. Good agreement between tested reactor data and MIGHT code calculation concerning the reactor proves the reliability of the methodology of analysis from a thermalhydraulic perspective. The minor discrepancies could be explained by differences in the relevant physical parameters used in each method of calculation.