Design and thermal–hydraulic analysis of a directly heated test facility simulating plate type fuel with narrow rectangular channel for nuclear research reactors

Abstract

Abstract

Nuclear research reactors often use plate type fuel assemblies where coolant is passed through a set of narrow rectangular channels to transfer the heat from the reactor core. The channel gap between fuel plates strongly influences hydraulic resistance, fuel temperature, and the available margin to boiling, making it one of the key design parameters governing reactor safety during normal operation, anticipated operational occurrences, and accident conditions. This paper presents the design and development of the Heat-transfer Experiment Loop for Plate geometry (HELP), an out-of-pile test facility built to reproduce the thermal–hydraulic behaviour of plate-type fuel assemblies for research reactors. The facility employs a directly heated Inconel-625 test section with channel gaps of 1.7 mm, 1.8 mm, 2.0 mm, 2.2 mm and 2.4 mm, covering a large range of narrow gaps commonly adopted in research reactors. A steady-state design framework was established predicting hydraulic and thermal performance of a Directly Heated Test Facility simulating narrow rectangular channels. Pressure drop, wall temperature, heat-transfer coefficient, axial temperature distribution, and margin to onset of nucleate boiling (ONB) were evaluated using established friction-factor and heat-transfer correlations, including Shah-London, Blasius, Petukhov, Gnielinski, Dittus-Boelter, Sudo, and the SULTAN-JHR closures proposed by Ghione et al. The comparison highlights the variation among available correlations and defines the expected operating envelope of the facility. The resulting framework provides the theoretical basis for subsequent experimental measurements for validation of CFD simulations for narrow rectangular cooling channels representative of plate-type fuel for research reactors.