The 1-D computer code MITH was used in this paper to perform sub-channel thermal-hydraulic analyses of a typical (Westinghouse model) pressurized water reactor. Two typical channels, hot and average, with the same flow rate and pressure drop, were tested under steady-state operating conditions. In this analysis, the channel with the highest temperature is designated as the hot channel. For the calculations, the channel model was divided into 20 parts. The thermal-hydraulic performance of the tested reactor was affected by power distribution, power level, and coolant mass-flow rate. Temperature distribution profiles of the fuel element and coolant are obtained for the average and hottest channels. A critical heat flux qncr analysis is also carried out and the heat fluxes in both channels were calculated. The W-3 correlation is employed to examine qncr in the hottest channel. Some data from the pressurized water reactor typical data sheet were used as input data, while others were used to validate the code. The code faithfully reproduced the Westinghouse model reactor results, including coolant, cladding, centerline, and surface fuel temperatures, quality and local heat flux qnloc, qncr and minimum departure from nucleate boiling ratio.

The steady-state thermal-hydraulic analysis of the core of the Boiling Water Reactor (BWR/6) at nominal operating conditions is presented in this paper. The BWR/6 is produced by General Electric USA. The analysis' goal is to keep the thermal safety margin under control and the core integrity intact under steady-state operating conditions. The effects of operating conditions such as power distribution, power level, and coolant mass flow rate on the pro- posed core's performance are investigated. For this purpose, the one-dimensional computer code MITH was used. The code's reliability was tested using the General Electric benchmark 3579 MW reactor. Two-channel models were tested (the average and the hot channel). Ther- mal-hydraulic parameters such as fuel-centerline, fuel-surface, outer clad surface and coolant temperature, critical and actual local heat flux, critical and minimum critical heat flux ratio and pressure drop are evaluated along the tested channels. Temperatures, as well as actual and critical heat flux distribution profiles, were obtained. The tested operating conditions had a significant influence on these parameters, and also on the thermal-hydraulic performance. The obtained results are in good agreement with the data from the tested core. The obtained results are well within the safety margins. The good agreement between tested reactor data and MITH code calculation concerning the reactor demonstrates the reliability of the analysis methodology from a thermal-hydraulic perspective.

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 × 10⁶ kg/h). The maximum cladding temperature stays within the suggested design limit of GCFRs (700-750°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 thermal-hydraulic perspective. The minor discrepancies could be explained by differences in the relevant physical parameters used in each method of calculation.