Conference Agenda

Given the rising global energy demand, Generation IV lead-cooled fast fission reactors (LFRs) are emerging as a promising technology, offering inherent safety, reliability, and sustainability. In this framework, newcleo is planning to demonstrate the feasibility of building a first-of-a-kind 30 MWe LFR (LFR-AS-30) by the early 2030s.

One of the decay heat removal systems (DHRSs) envisaged for newcleo’s LFR, designed to remove the residual heat generated by the core after a shutdown, is based on the dip cooler architecture: several tens of bayonet tubes, working in parallel, are directly submerged into the reactor’s primary coolant pool. Water, the secondary fluid, flows through each bayonet tube undergoing phase change.

To assess potential instabilities that may occur within the dip cooler DHR, the Dip Cooler Instability (DCI) Test Facility was designed. The facility will be operated to investigate the behavior of two bayonet tubes. The primary focus of the activity has been the computational modeling of the DCI Test Section using thermal-hydraulic system codes. The selected system codes are ATHLET 3.4.1 (2023.2) and RELAP5/MOD3.3. The models comprise two coupled bayonet tubes operating in parallel, with a uniform and constant temperature applied to the outer surface of both risers. To support the facility design phase and test matrix definition, a code-to-code benchmark was performed prior to experimental validation.

The results of the different codes are compared to highlight the level of agreement. The current models will be extended to the entire facility and will be validated against the upcoming experimental campaigns.

Due to its unique safety and economic advantages, the lead bismuth eutectic (LBE) cooled fast reactor has been extensively studied. A multi-application experimental circuit (MATH) for LBE was constructed, and the steady-state and heat transfer characteristics of the circuit were investigated with different heat flux. The fluid temperature distribution in each section of the test section was measured to obtain the convective heat transfer coefficient. The experimental results indicate that the LBE exhibits stable flow characteristics in the heating power range of 13-20 kW. the Pe number remains basically constant across different heating powers, indicating that the flow characteristics are independent of the heating power. Experimental and theoretical analyses demonstrate that for upward flow, the heat transfer coefficient decreases with increasing heat flux, indicating that the buoyancy effect enhances the impairment of heat transfer. Based on the experimental data, a new LBE convective heat transfer correlation is proposed, and its relative error with the experimental data is less than 5%.

Lead-bismuth alloy has the characteristics of high thermal conductivity, low melting point and high boiling pointcite{xing2025comparative}, which enables the lead-bismuth liquid metal-cooled fast reactor to operate at atmospheric pressure and achieve a high core average temperature, and thus it has a unique advantage over the traditional pressurized water reactors in terms of safety and economy, making it a fourth-generation nuclear energy system and has a wide range of application prospects. According to the different uses of lead-bismuth reactors, it is of theoretical value and practical engineering significance to carry out related technical research.

This paper takes the small integrated lead-bismuth reactor as the research object, and establishes the simulation model including the lead-bismuth reactor vessel and the main cooling circuit, electromagnetic pump, helical coil tube type once-through steam generator model and voltage regulator based on the system analysis program NUMAP, and establishes control modelsd for the reactor power, the steam generator, the gas pressurizer, and the electromagnetic pump, based on the operational characteristics of the lead-bismuth reactor. Through simulations under normal operating conditions and accidental conditions such as control drum stoppage, it is demonstrated that the established simulation model accurately reflects the steady state characteristics of the system. The verification of transient lift power is also completed, and the control model effectively regulates the system. This lays a foundation for in-depth research on the operational and control characteristics of the lead-bismuth reactor power unit.

Argonne National Laboratory participates in the International Atomic Energy Agency (IAEA) coordinated research project (CRP) on “Benchmark of Transition from Forced to Natural Circulation Experiment with Heavy Liquid Metal Loop (NACIE)”. The benchmark project includes three experiments from the NACIE lead-bismuth eutectic (LBE) experimental loop located at the ENEA Brasimone Research Center, in Italy. All benchmark tests are on transition from forced to natural circulation initiated by shut-off of the gas lift-off pump. The main difference between the tests, ADP10, ADP06, and ADP07, is which heater pins are activated (heated) during the tests, meant to approximate partial flow blockage in a fuel assembly.

Argonne’s work with the SAS4A/SASSYS-1 code on the CRP includes system-level and sub-channel simulations. Via comparisons against experimental measurements from the NACIE tests, these benchmark simulations are being performed to expand the validation basis of the SAS4A/SASSYS-1 code. The paper presents progress on NACIE test simulations for simulation of the NACIE benchmark tests with SAS4A/SASSYS-1 code.

All the results obtained so far for NACIE tests and presented in this paper in general show good agreement with the available experimental data. However, in some cases, model modifications were needed to obtain that good level of agreement - those model modifications are also presented in the paper, along with the identification of the remaining differences and approaches for how to resolve them in future work.

Utilizing a fuel assembly cooled by a lead-bismuth alloy within a narrow rectangular channel offers significant benefits for thermal exchange. This design enhances the thermal transfer area within the core, enabling the efficient removal of excess heat. In this study, we focus on understanding the role of surface roughness in narrow rectangular channels for lead-bismuth alloy. By using detailed numerical simulations, we explore how variables like the type, height, and spacing of the roughness affect the flow and heat transfer characteristics of the alloy. Our findings indicate that the channels with a rough interior have a much higher Nusselt number and friction resistance compared to channels with a smooth interior. The disturbance of velocity distribution around the roughness elements significantly affects surface resistance, turbulent mixing, and heat transfer. When the fluid flows over these roughness elements, the fluid behind them is disturbed and forms vortices, which disrupt the flow and heat transfer boundary layers, thus enhancing heat transfer and also increasing flow resistance. These results offer valuable insights for the design of high flux reactor core.