System thermal-hydraulic analyses supporting the development of lead-cooled fast reactors

The urgent need to decarbonize the global energy system, in response to the climate crisis, has revitalized interest in nuclear power as a reliable, low-carbon, and dispatchable energy source. Nuclear energy, already a major contributor to carbon-free electricity generation, has the potential to play a central role in achieving deep decarbonization targets. To this end, the Generation IV International Forum (GIF) has promoted the development of advanced reactor concepts that aim to overcome the limitations of current technologies. Among these, Lead-cooled Fast Reactors (LFRs) have emerged as one of the most promising options. LFRs operate in a fast neutron spectrum by relying on liquid lead or lead-bismuth eutectic as the primary coolant and the deployment of advanced fuel types such as Mixed (MOX) fuel, typically composed of plutonium oxide blended with depleted or natural uranium oxide, allows for the recycling of plutonium recovered from spent nuclear fuel, thereby reducing the accumulation of fissile material in storage and supporting non-proliferation and sustainability goals. Additionally, operating in a fast neutron spectrum allows LFRs to enhance the fuel utilization enabling the use of fertile isotopes such as Uranium-238 and Thorium-232 for breeding fissile material (respectively Plutonium-239 and Uranium-233). This capability extends the energy extracted per unit of natural uranium and reducing the need for fresh uranium mining. Moreover, the fast spectrum is particularly effective in transmuting long-lived minor actinides (such as neptunium, americium, and curium) into shorter-lived or stable isotopes through fission. This transmutation process significantly reduces the long-term radiotoxicity and heat load of nuclear waste, addressing one of the key challenges facing nuclear energy in terms of public acceptance and repository requirements. The choice of Heavy Liquid Metals (HLM) as reactor coolant offers a series of heat transfer and chemical advantages over traditional coolants. Their extremely high boiling point (around 1740 °C for lead and 1670 °C for LBE) permits reactor operation at atmospheric pressure, eliminating the need for pressurized vessels and reducing the risk of pressure-induced failures. The good thermophysical properties of HLMs namely the high density and thermal conductivity enhance passive heat removal capabilities and enable compact core designs with efficient heat transfer. The employment of HLMs enables operation at higher temperatures, improving the thermodynamic efficiency and contributing to the performances of LFRs. Additionally, HLMs are chemically inert with respect to air and water, eliminating the potential for violent reactions such as hydrogen generation, which is a major concern in Sodium-cooled Fast Reactors (SFRs) or Light Water Reactors (LWRs) designs. Despite their many advantages, the use HLMs as primary coolants introduces a series of technical challenges that must be carefully addressed to ensure safe and reliable reactor operation. One of the most critical issues is corrosion: at high temperatures, lead and LBE are chemically aggressive toward structural materials, particularly steels, leading to progressive degradation of components exposed to the coolant. The formation of protective oxide layers on the surface of steels can mitigate corrosion, but their stability is highly sensitive to the oxygen concentration in the coolant. Maintaining a precise and stable oxygen level is therefore essential, requiring the implementation of robust oxygen control and monitoring systems. A related concern is erosion and mass transfer, especially in regions of high flow velocity or thermal gradients, where protective oxide layers may be mechanically removed, further accelerating material degradation. Another major challenge involves coolant activation and polonium production, particularly for LBE, which contains Bismuth-209 that can be transmuted into Polonium-210 under neutron irradiation. Polonium-210 is a highly radiotoxic alpha emitter with a relatively short half-life (~138 days), posing significant radiological hazards during maintenance, operation, and decommissioning. This issue demands stringent containment strategies and remote handling capabilities. Pure lead avoids the polonium problem but remains heavier and has higher melting and freezing points, which introduces its own operational constraints. Moreover, the opacity of HLMs complicates flow visualization, instrumentation, and refuelling operations. Moreover, HLM Thermal-Hydraulics (TH) involves complex phenomena such as transition from forced to natural circulation, pool thermal stratification and freezing. Coolant freezing is also a considerable concern: lead freezes at around 327 °C and LBE at about 125 °C, necessitating continuous heating of the coolant and system components even during shutdown or maintenance phases to prevent solidification and blockages. This imposes complexity to decay heat removal systems, which must be designed to operate reliably across a wide range of temperatures. Finally, the TH behaviour and simulation of HLMs presents unique modelling difficulties. Their high density and low Prandtl number result in flow and heat transfer characteristics that differ significantly from those of water, requiring significant efforts in Research and Development (R&D) activities. In the absence of full-scale operational reactors, the design and safety assessment of LFRs must rely heavily on high-fidelity numerical simulations supported by data from scaled experimental facilities. The qualification of numerical models and System Thermal-Hydraulic (STH) codes, and the validation of safety systems under representative conditions, remain critical steps in the development of this reactor class. This doctoral research contributes to the advancement of LFR technology through an extensive campaign of system-level TH analyses aimed at supporting the design, qualification, and safety demonstration of several LFR concepts under development in the world. The primary simulation tool used throughout the work is RELAP5, a STH code validated primarily for LWRs. The thesis is articulated around three key projects: ALFRED, the Westinghouse-LFR (WEC-LFR), and MYRRHA. In the case of ALFRED, a series of in-depth studies—including a Phenomena Identification and Ranking Table (PIRT) analysis—were performed to pinpoint the main technical challenges and to guide the prioritization of R&D activities. To assess reactor behaviour under both normal and accidental conditions, coupled Neutron Kinetic (NK) and TH simulations were conducted, focusing on unprotected transients. Particular attention was devoted to ALFRED’s innovative Decay Heat Removal (DHR) system, which relies on degradation of condensation mechanisms affected by the presence of non-condensable gases. This approach aims to conjugate two competing objectives: efficient decay heat removal and maintaining the primary coolant above its freezing point. To support and evaluate this concept, the SIRIO facility was involved in this work performing both experimental and numerical TH analyses. Additional extensive investigations were carried out to study the condensation process—both in pure conditions and with non-condensable—and to validate the TH models and correlations implemented in RELAP5. The work on the WEC-LFR involved two major facilities. The PHRF which is dedicated to study and test the Passive Heat Removal System (PHRS) of the WEC-LFR. The PHRS is based on water-to-air transition heat transfer regime causing the decreasing of heat removal. The PHRF was modelled and deeply analysed. The other experimental facility involved within this framework is the VLF, representing the Reactor Cooling System (RCS), was analysed performing a pre-test analysis to gain enough understanding of this facility both in nominal and accidental conditions. Furthermore, the inclusion of key components of the WEC-LFR’s RCS within the VLF experimental facility prompted detailed stand-alone analyses of two critical elements: the Primary Heat Exchanger (PHE) and the Fuel Pin Bundle Simulator (FPBS). The PHE, an innovative microchannel heat exchanger, represents a novel solution for which limited experience and literature exist within the nuclear industrial. As such, a comparative study was carried out using both a STH code (RELAP5) and a Computational Fluid Dynamics (CFD) tool (Ansys CFX) to better understand its behaviour. The FPBS, on the other hand, is a representative model of the WEC-LFR Fuel Assembly. To evaluate its performance, a subchannel analysis was performed using the DASSH code, followed by a comparative assessment between DASSH and RELAP5 to verify modelling consistency and identify code-specific limitations. Finally, for the MYRRHA reactor, a PIRT analysis has been performed in parallel with ALFRED, in the framework of R&D activities supporting the development of LFRs in Europe. Next, the research focused on the secondary cooling system, investigated through the HEXACOM experimental facility. This work primarily aimed at applying and refining a qualification procedure for STH numerical models of experimental facilities, with the goal of enhancing modelling accuracy and supporting future licensing processes. A comprehensive RELAP5 model of the system was developed, calibrated, and validated using both experimental data and nominal design specifications. The simulation results offered important insights into the behaviour of the water loop and played a key role in the qualification of the thermal-hydraulic model. Future work should focus on extending the validation domain of STH codes through additional high-quality experimental data, particularly under transient and accident conditions. The coupling of STH codes with high-fidelity tools such as CFD or subchannel analysis codes should be further explored to improve the resolution of local phenomena within critical components. Moreover, the integration of multi-physics approaches—including structural, NK and TH feedback—will be essential to enhance the accuracy of safety assessments. On the experimental side, additional test campaigns on advanced facilities such as SIRIO, PHRF, VLF, and HEXACOM are necessary to refine modelling strategies and reduce uncertainties in design-relevant conditions. These future directions are expected to contribute significantly to the qualification and licensing of LFRs, ultimately supporting their safe and sustainable deployment in future nuclear energy systems.