Conference Agenda

Sodium-cooled fast-neutron reactors are currently considered to be the most mature type of reactor able to closing the fuel cycle. In France and throughout the world, pool-type reactors are selected to build generation IV power plants. Complex liquid sodium flows are known to occur in reactors in several conditions. In order to predict these phenomena, CEA developed the MATHYS code (Multi-scale Advanced Thermal-HYdraulics Simulation). This tools enables the coupling of the system thermal hydraulics code CATHARE, the sub-channel code TrioMC and of the 3D thermal-hydraulics code TrioCFD. Thanks to this coupling approach, the entire primary side of a reactor can be modelled, accounting for the feedbacks for the different scales (core, inter-wrapper flow, pools). In the late 1980s, the CLAUDINA test facility was operated at the CEA Cadarache research centre. The experimental campaigns aimed at the characterisation of the behaviour of an intermediate heat exchanger (IHX) under a variety of operation conditions. The CLAUDINA facility is a mock-up of a sodium –sodium IHX. Tests at different flow conditions were performed. The experimental data from these tests are very valuable for the validation of the intermediate heat exchanger modelling in MATHYS. In this paper, the CLAUDINA facility is first introduced. The CATHARE, TrioCFD and Neptune_CFD codes are then presented, and the models of the CLAUDINA facility are described. The results of these different modelling approaches for several tests are presented and discussed. Conclusions and recommendations are proposed.

The inherent safety features make lead-cooled fast reactors (LFRs) an attractive solution for the increased energy demand and the development of advanced nuclear power plants. Due to the limited operational experience with these reactors, simulation and analysis with system thermal-hydraulic (STH) codes become crucial to study the plant behaviour under safety-relevant conditions and to support the reactor design. In this paper, the LFR modelling has been carried out with different STH codes, such as RELAP5 Mod 3.3 version Beta, modified by University of Pisa to account for lead as working fluid, ASYST-LM and ATHLET codes.

The simulation activity aimed at assessing the code capabilities to reproduce selected phenomena occurring in LFRs under normal and accidental conditions, derived from some of the most representative transient scenarios, such as unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS), unprotected transient of overpower (UTOP) and unprotected loss of offsite power (ULOHS+ULOF). Particular attention has been paid to the establishment of natural circulation following the loss of primary pumps, which affects the sizing of the safety features derived from such operating conditions.

The obtained results will support the verification and validation efforts of the STH codes applied to LFRs. The investigated codes show a good agreement and the comparison proposes some open perspectives and future improvements.

Passive safety systems employing physical processes and phenomena are increasingly applied to contemporary nuclear reactor design. Assessment of the performance of these systems under various scenarios relies heavily on numerical analysis using codes from 1D to 3D depending on different levels of the design and safety demonstration purposes. Thermal-hydraulic (TH) phenomena in pool-type Lead-cooled Fast Reactors (LFRs) often exhibit multi-dimensional characteristics such as the development of thermal stratification and mixing during natural circulation. Accurate prediction of mutual interaction between these phenomena in the pool and its effects on loop dynamics requires 3D analysis.

Computational Fluid Dynamics (CFD) provides high-fidelity 3D TH analysis but is computationally expensive for analysis of prototypical conditions. System Thermal-Hydraulic (STH) codes (e.g., RELAP5) offer efficient calculation but are inadequate to resolve 3D phenomena. A compromised solution is to use system-level TH codes with 3D features, e.g., (GOTHIC, CATHARE). The recent development of GOTHIC enables the modeling of Lead-Bismuth Eutectic (LBE) flow while its suitability and validity for safety analysis need to be confirmed.

Therefore, this work aims to assess GOTHIC predictive capabilities for LBE 3D phenomena through code-to-code and code-to-experiment comparisons. Validation data is obtained from a forced to natural circulation transient produced in TALL-3D facility which is a 7m LBE loop featuring a 3D pool-type test section. Simulations are performed using CFD code ANSYS Fluent and system TH code GOTHIC. The focus will be pool thermal stratification and mixing in the 3D test section.

Current major international nuclear reactor system analysis codes predominantly utilize the two-fluid six-equation model to study the behavior of nuclear power plants under accident conditions, which presents considerable limitations. Most studies of the two-component two-fluid model have focused on water-steam systems, while liquid metal-gas systems at high temperatures have received relatively less attention. This paper studies the two-component two-fluid model and its rapid solution method to address the demands of full-scale simulations for both existing and conceptual nuclear reactor systems. The conservation equations of the two-component two-fluid model are discretized using a first-order upwind semi-implicit method, based on the staggered grid and finite volume difference. A system of linear equations is derived by substituting the equation of state and solved using the NRLU method. The mathematical suitability of the model is enhanced by introducing a virtual mass force. When one phase of the two-component two-fluid model is absent, the fraction of the virtual phase is assigned a small value to prevent the coefficient matrix from becoming singular. By simulating the natural circulation and gas injection-enhanced circulation conditions at varying power levels on the lead-bismuth loop test bench NACIE, the numerical accuracy and computational stability of the semi-implicit numerical algorithm for lead-bismuth-gas two-component two-fluid model are successfully demonstrated. It lays the foundation for further research on the two-component two-fluid model and the development of related code.

Lead-cooled Reactors (LFRs) are considered a promising concept in the framework of designing new Generation-IV reactors. The analysis of the thermal-hydraulics phenomena can be performed by means of numerical RANS simulations. This paper aims to evaluate the best practices for the mesh generation of a grid spaced fuel-bundle assembly. The results are compared to the experimental results provided in the LFR-T/H benchmark promoted by OECD NEA. The work focuses on different strategies to generate the background mesh and alternative modelling tools (e.g baffles) for capturing detailed geometry and dealing with the contact points in OpenFOAM. Initially, the bundle geometry without the grid is simulated under isothermal conditions and the results in terms of pressure drop are compared with existing correlations. In the second part, a single grid is included in the numerical domain, and it is characterized in terms of pressure drop as function of the flow velocity.

Liquid metal cooled fast reactors is one of the most promising fourth-generation nuclear systems. Y-junctions are commonly adpot in piping system. Non-Isothermal fluids frequently mixed in these components, lead to thermal pulsation on the solid-wall, and may induce thermal fatigue to piping system.To understand the mechanism of thermal pulsation and thermal fatigure, we independently set up a non-isothermal mixing test platform of the working fluid, and obtained the temperature distribution of the working fluid during the non-isothermal mixing process in the 90° Y-shaped component. This verifies the correctness of the Dynamic Smagorinsky Sublattice model in the non-isothermal mixing simulation, so as to correctly simulate the flow and heat transfer of liquid metal. On this basis, large Eddy Simulation (LES) approach for liquid metals in Y-junctions was applied. Angle and velocity pulsation behavior caused by the mixing of hot and cold fluids in Y-junctions under different incident angles (θ=30-90°), and momentum ratios (M_R=0.25–4.51) were discussed. The results show that the momentum ratio and the angle significantly influence the mixing characteristics of hot and cold fluids. At a 90-branch angle, fluid mixing is uniform, the thermal pulsation peak is larger. As the momentum ratio increases, the peak temperature pulsation gradually decreases. The findings offer valuable insights for the thermal-hydraulic design of future liquid metal cooled fast reactors.