Conference Agenda

Sodium-cooled fast-neutron reactors (SFR) are currently considered to be the most mature type of reactor able to optimize uranium ore usage and reduce nuclear waste produced from Generation II and III reactors. CEA led studies up to 2019 on the features of a 600MWe reactor within the frame of the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) project. The chosen pool-type design offers the advantage of containing the primary sodium within a single vessel, ensuring safer operations by transferring heat to the secondary sodium circuit via Intermediate Heat Exchangers (IHX). This design eliminates the risk of water/primary sodium interaction. A tertiary loop then generates steam for power conversion. Given the safety implications of the design, careful study of the vessel's geometry is essential, particularly the IHX, which plays a critical role in heat exchange. To investigate the flow dynamics within the vessel, a scaled-down model of the ASTRID reactor was constructed. Using a similarity approach water was used as a simulant fluid due to the complexity and cost of sodium-based experiments. This model allows for adjustments in IHX geometry to conduct parametric studies on flow behavior. Particle Image Velocimetry (PIV) was employed to measure velocity near the IHX inlet across different configurations. The results align with previous studies, indicating that, whatever the configuration, flow is concentrated in the lower section of the IHX, offering valuable insights for future design improvements.

This paper presents a safety analysis performed using the AMESIM (Advanced Modeling Environment for Simulation of Engineering Systems) code for the PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor), proposing an appropriate methodology for global export market. The safety analysis for the PGSFR has been carried out with the MARS-LMR code. This research aims to develop a transient safety analysis platform for SMR (Small Modular Reactor)-powered ships. The AMESIM code offers advanced numerical methods capable of solving complex multi-physics problems, making it suitable for modeling not only thermal-fluid systems but also mechanical and electrical systems, thus fitting the modeling of ship propulsion systems. However, there are challenges in modeling nuclear systems with AMESIM. Therefore, this study defined coolant properties and modeled reactor systems of the PGSFR in AMESIM to evaluate the applicability of nuclear systems in the AMESIM SW environment. In the AMESIM code, the PGSFR consists of the Core, PHTS (Primary Heat Transport System), IHTS (Intermediate Heat Transport System), and SG (Steam Generator). In the Core, Reactivity Feedback and Point Kinetics are calculated to determine the Neutron flux. It was found that the results from AMESIM code have a good agreement with design values of the PGSFR. Furthermore, preliminary safety analysis for representative DBEs (Design Bases Events) in a PGSFR has been implemented with AMESIM code.

To enhance the safety of sodium-cooled fast reactors, the decay heat removal system under natural circulation with a dipped-type direct heat exchanger (D-DHX) installed in a hot pool of a reactor vessel (RV) has been investigated. During the D-DHX operation, the thermal-hydraulics of RV is complicated because the cold sodium from the D-DHX flows into the core and the radial heat transfer among assemblies occurs. To evaluate the RV thermal-hydraulics and core cooling performance given from these phenomena in the design study, we have been constructing the RV model using a computational fluid dynamics code (RV-CFD) with the subchannel CFD (SC) model for assemblies as a practical model which can achieve a lower computational cost while maintaining prediction accuracy (RV-CFD). However, the applicability investigation of RV-CFD was limited to several numerical analyses of steady-state. In this study, to evaluate accurately the transient response of sodium temperature using the RV-CFD, we develop the non-equilibrium thermal (NET) model in the SC model which can consider both the heat capacity and thermal resistance in simulated fuel pins. The transient analysis simulating the power reduction due to reactor scram from the steady-state operation in a sodium experimental apparatus named PLANDTL-1 is conducted. The result shows the thermal-hydraulic behavior in the RV during the transient is predicted, and the core temperature in the transient is reproduced. Thus, the RV-CFD using the NET model in the SC model can evaluate the transient temperature response.

Ensuring the safety of liquid metal-cooled reactors necessitates accurate modeling of the transition from steady-state operation to long-term passive cooling under various initiating events. A significant challenge exists in a protected loss of flow event, where thermal stratification developing in the reactor pools can impact or delay the transition to long-term cooling through natural circulation. This can induce unexpected thermal gradients which can lead to oscillating temperature fields resulting in off-normal thermal-hydraulic behavior throughout the system. This paper describes ongoing activities at Argonne National Laboratory to validate system-level software using the Thermal Hydraulic Experimental Test Article (THETA) of the Mechanisms Engineering Test Loop (METL) to enhance the system-level tools used to assess safety margins. The experimental campaign using THETA, designed to operate at scaled-down prototypical pool-type sodium-cooled fast reactors (SFRs) conditions, has been evaluated to expand the validation basis for modeling thermal stratification during and after the transition to natural circulation. A system-level computational model using SAS4A/SASSYS-1 has been developed to represent the full THETA facility, including the modeling for the facility electric heater, primary and secondary pumps, an intermediate heat exchanger, an air-cooled heat exchanger, hot and cold pools, and connected piping across both the primary and secondary systems. The THETA SAS4A/SASSYS-1 model uses a stratified volume model for both the hot and cold pools with heat transfer interactions across major components. The preliminary assessment results are discussed with key findings and potential directions for improvements of the model.

With the growing global interest in small modular reactors (SMRs), one of the key goals of the new European ESFR-SIMPLE project is to develop a compact sodium-cooled fast reactor (SFR). This system aims to address crucial SMR features, such as the transportability of main components and grid flexibility, while also leveraging the extensive experience in sodium coolant technology. Additionally, reducing core power could enhance safety by improving inherent reactivity characteristics and enabling more efficient removal of residual power, potentially paving the way for constructing a prototype SMR-SFR in Europe.

This study presents the initial results of primary system modeling for the SMR-SFR with a thermal power of 360 MW. The ATHLET system code was used to simulate the sodium coolant flow in the primary system, exploring various modeling options. Notably, the application of models for pseudo-3D flow in the large hot plenum of the primary vessel was of particular interest. The paper discusses the simulation results for selected transients and the findings from comparing the conventional zero-dimensional plenum approach with alternative pseudo-3D flow modeling options for the hot pool.