In-depth design and construction of an integrated thermal-hydraulic facility (INTALI) for a prismatic gas-cooled micro reactor

To support the development of an innovative gas-cooled micro reactor (GMR), an integrated thermal-hydraulic facility (INTALI) has been designed to investigate heat transfer and fluid behaviors under both prototypical operating conditions and representative accident scenarios, thus providing a basis for thermal-hydraulic code validation. The INTALI facility consists of a helium circulation loop and a test section containing a scaled-down simulated reactor and a passive core cooling system (PCCS). This paper aims to finalize the facility design and corresponding experimental methodology. Firstly, the overall configuration and layout of the apparatus were described. An analytical model was developed to predict the helium temperature distribution along the primary loop, which is critical to facility operational control and piping integrity. A pre-test calculation under nominal conditions indicates that all helium temperatures throughout the loop remain below the piping safety limit. Parametric calculations were performed to study the influences of both heating power and the flow rate ratio of the main flow to the protective flow. Secondly, the detailed structural designs of the test section were elaborated. Specific internal components within the reactor were designed, including tenon-mortise joints for fuel assembly alignment and core baffles to suppress bypass flow. Structural analysis confirms that incorporating four reinforced ribs improves the PCCS dome’s stiffness. The layout and the installation of thermocouples (TCs) within both the reactor and the PCCS were described. A comparative calculation demonstrates that the mixer effectively enhances mixing of the main and the protective flows, thus lowering the temperature at the reactor outlet. Thirdly, an experimental methodology was proposed to simulate an anticipated transient without scram (ATWS) accident initiated by loss of forced coolant. To incorporate the temperature-reactivity feedback, a computational module solving the neutron point kinetics model was developed and validated. Furthermore, dedicated data processing along with uncertainty analysis methods were formulated to derive key physical quantities from experimental measurements: 1) friction factor of helium through a coolant channel, 2) heat exchange rate between the reactor and the PCCS, and 3) convective heat transfer coefficient on the reactor pressure vessel (RPV) wall. Finally, the current status of facility construction and the planned testing program in the next stage are outlined.