Pre-test analysis and thermal-hydraulic characterization of the versatile loop facility

Lead-cooled fast reactors (LFRs) are a concept of advanced nuclear reactor using liquid lead or lead–bismuth eutectic (LBE) as coolant. These reactors operate with a fast neutron spectrum, enabling efficient use of nuclear fuel and the capability to burn long-lived radioactive isotopes, thereby reducing nuclear waste. LFRs offer significant safety benefits due to the properties of lead, such as high boiling point, excellent heat transfer, absence of exothermic reactions with water and air, enabling passive safety features and efficient heat removal in case of accidents. Among the various LFR concepts, a significant effort has been carried out by Westinghouse Electric Company (WEC) to design a pool-type LFR with a core power of 950 MWth. As R&D efforts continue to advance LFR technology, the focus is on optimizing reactor design, developing robust experimental facilities, and validating numerical tools to ensure the reliable and safe operation of these next-generation systems. To address knowledge gaps in liquid lead technology, dedicated experimental campaigns are necessary. In particular, for the WEC-LFR reactor, the key components of the Reactor Coolant System (RCS) of the WEC-LFR will be tested and characterized in the Versatile Loop Facility (VLF). The facility features a lead loop operating up to 650 °C, equipped with a 500-kW electric Fuel Pin Bundle Simulator (FPBS) representing the Westinghouse LFR bundle design. It also includes a hybrid microchannel-type diffusion-bonded heat exchanger, which represents the primary heat exchanger used in the Westinghouse LFR. Heat removal is managed by a supercritical water-cooling loop, designed to operate at a pressure of 330 bar and temperatures up to 620 °C. This paper details the design of the VLF, then its numerical model is presented, along with the thermal–hydraulic characterization of the facility using the system code RELAP5/Mod3.3, properly modified to include the thermo-physical properties of liquid lead. Additionally, a detailed characterization of the FPBS has been performed using the subchannel code DASSH developed by Argonne National Laboratory. Finally, starting from a steady state condition representative of the WEC-LFR, two accidental transients have been simulated: the loss of heat sink and the loss of the FPBS power. Both analyses provide important indications for the safe operation of the experimental plant.