Experimental study of the influence of inter-wrapper flow on liquid-metal cooled fuel assemblies

Abstract

Most liquid metal fast reactor (LMFR) designs consider a regular hexagonal lattice of fuel pins in the core. Detailed thermal-hydraulic analyses of the core are based on an accurate description of the flow in fuel assemblies (FAs), which are arrangements of pins with a specific spacer design, and enclosed in a wrapper tube. Treating them as an isolated unit with adiabatic outer boundary condition is a reasonable approach for a conservative estimation of the maximum pin wall temperature and for a detailed comparison of experiments and simulations. However, there are conditions that are not covered by this approach.Each FA is separated from its neighbors by a small gap of a few millimeters, mainly due to mechanical considerations, including expansion and bending during irradiation. During normal operation, this inter-wrapper gap region is filled by static liquid-metal coolant, as it is separated from the main flow, driven e.g. by the main circulation pumps. An upward flow in this region is expected to occur in scenarios of decay heat removal by natural circulation, when cold liquid from the upper plenum penetrates at the edges and rises at the center. This inter-wrapper flow (IWF) phenomenon has been observed in scaled experiments with water in Germany and with sodium in Japan. The extent of this phenomenon depends on the characteristics of the pool and pressure losses in the core, which are specific to each LMFR design. As a rough estimation, the resulting upward velocities in the gap region can be similar to the mean values inside the FAs.In order to obtain a more accurate description of the heat transfer processes related to IWF, a unique experimental campaign is performed at the Karlsruhe Liquid Metal Laboratory (KALLA). It considers three electrically-heated neighboring hexagonal rod bundles with wire spacers, and a fourth channel in the gap region, representing the IWF. It is possible to control the three thermal powers and the four flow rates independently. The selected local instrumentation is focused on capturing the heat transfer process around the IWF region, with thermo-couples at two measuring levels and at several axial positions in the gap center, as well as at the gap outlet, where a movable Pitot probe is installed. The test section geometry is representative of the core arrangement in the MYRRHA reactor, and the coolant fluid is lead-bismuth eutectic (LBE), at reactor-like operating conditions at decay heat removal in terms of temperature and power density.The results of this experimental campaign are presented in this work. A preliminary assessment indicates that a significant fraction of the thermal power is transported by the IWF. This effect is particularly relevant for non-symmetric cases, e.g. with one blocked bundle operating at full decay heat and reduced flow rate. These empirical data on the influence of the IWF represents an important step toward the safety evaluation of the complete core.