Conference Agenda

In validation exercises, numerical simulations are compared directly with experimental data. If the agreement is sufficiently good, then the model is considered validated (with the reported accuracy) and it can be used with a high level of confidence for the investigation of closely related scenarios that have not been or cannot be studied experimentally. A key element in this comparison is the justification of the modeling assumptions and simplifications, demonstrating that the dominant physical phenomena are well represented, and thus simulations of similar scenarios are reliable.

Rod bundles with wire spacers are used as fuel assemblies in many liquid-metal cooled fast reactor designs. The thermal-hydraulic scenario is complex due to the intricate geometry and the low Prandtl number of the coolant. While several experimental campaigns are reported in the open literature, some important aspects must be taken into account for comparing numerical and experimental results. This work discusses the impact of geometry simplification (e.g. wire shape) and uncertainty in the location of thermo-couples, as well the influence of material properties and boundary conditions. Moreover, relative errors must be defined with respect to the correct reference value in order for them to strongly support conclusions regarding the accuracy.

Two reference cases are selected for detailed analysis. They cover the study of the velocity profile in an isothermal case, and the temperature profile in a heated case with local blockages.

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Liquid Metal Fast Reactor (LMFR) rod bundles can be designed with grid spacers or wire wraps. In both types of designs, pins may deform due to tension of the pre-stressed wires, contact pressure between clad and adjacent rods and/or wires, thermal and irradiation clad creep, irradiation-caused swelling and fuel burnup. In order to analyze the effect of such deformations on the peak temperature and temperature distribution, and in order to validate a simulation methodology, a series of experiments in water and liquid metal for grid-spaced and wire-wrapped liquid metal cooled fast reactor rod bundles is being conducted in Europe. These experiments are supported by numerical analyses which will be described.

Experiments in water bundles aim at validating the simulated flow field around a bended pin, while at the same time giving a first impression on the validation of the temperature field. Similar experiments in liquid metal rod bundles aim to validate the temperature field in the simulations.

A description of the water and liquid metal experiments will be provided and subsequently the numerical support to these experiments will be discussed completed by a future outlook.

Lead-bismuth Eutectic(LBE) is a excellent coolant for the Small-Modular Reactor(SMR). And there is a Plate-type fuel assembly is considered in our research, making full use of its advantages of tight structure and high heat transfer efficiency will provide a wide prospective for the development of SMR. The comparable research about the convective heat transfer characteristics of LBE in horizontal and vertical rectangular channels is conducted in this paper. There are 9 subchannels in a assembly and the aspect ratio of each subchannel is about 21.4. The flow rate distribution characteristics and the convective heat transfers characteristics are compared and analyzed. Within the range of 370000 ≤ Re ≤650000 and 300℃ ≤ T ≤450℃, the convective heat transfer mechanism is researched. The critical working conditions for this structure are proposed for the natural convection, mixed convection and forced convection using the Large Eddy Simulation(LES) model, and the influence of buoyancy on the turbulent heat transfer process is also analyzed.

Flow blockage of the fuel assembly in the lead-based fast reactor (LFR) may produce critical local spots, which will result in cladding failure and threaten reactor safety. In this study, the flow blockage characteristics are analyzed with the sub-channel analysis method. The effects of different blockage areas and axial positions are considered. The results indicate that when a flow blockage accident occurs, larger blockage areas and blockage positions closer to the axial center result in more severe accident consequences. However, for the blockage scenarios studied in this study, all peak temperatures remain below the material limit temperatures. This work could provide a reference for the future design and development of the LFR.

Liquid metal cooled fast reactor (LMFR) designs typically utilize helically wire-wrapped pin bundles. The high thermal conductivity of the liquid metal, combined with the thermal resistance of the wire-wrapped contact points, leads to localized hot spots under high heat flux conditions characteristic of the LMFR. A separate effects test was conducted under stagnant fluid conditions to measure the local temperature peaks at the wire contact point using an infrared thermography. Galinstan was selected as a primary test fluid to simulate the thermal hydraulics characteristics of the low Prandtl number fluids in LMFRs and the analysis of local hot spots was performed with the various wire configurations and fluids. In addition to the experimental investigations, the study of hot spots was further expanded to the forced convection conditions through computational fluid dynamics (CFD) simulation, allowing for a more comprehensive analysis of the effect of the influencing parameters. Hot spots were observed when the Biot number exceeds unity, exhibiting a direct proportional relationship with the local heat flux. Based on both experimental data and CFD-generated data, an empirical correlation was proposed to predict the severity of the hot spots, which can be applied across different material and flow conditions. This study provides valuable engineering insights and recommendations for wire design strategies aimed at mitigating the undesired hot spots by reducing the contact area and enhancing the heat transfer coefficient.

Lead-bismuth reactors, with their advantages of high power density, strong inherent safety, and good maneuverability, have received widespread attention. Based on the subchannel model, a thermal-hydraulic model for a small lead-bismuth reactor assembly was established. Using a statistical analysis framework, Latin Hypercube Sampling was adopted as the sampling method, while the Wilks method and Spearman method were used for tolerance interval estimation and sensitivity analysis, respectively, to develop an uncertainty analysis program for lead-bismuth subchannels. Referring to pressurized water reactors, the selected input parameters include coolant inlet flow rate, inlet temperature, outlet pressure, power, fuel thermal conductivity, and mixing coefficient. The chosen output parameters are the average coolant temperature, the coolant temperatures in different types of channels, and the surface temperatures of different types of fuel rods. A total of 5000 samples were drawn for uncertainty analysis. The results indicate that the tolerance interval range for the average coolant temperature is smaller than that for the subchannels. The mixing coefficient has a lesser impact on the former but a greater influence on the coolant temperatures at different positions and the fuel rod temperatures.