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• Subchannel code for the thermal hydraulic analysis of liquid metal cooled reactor. • Simulation of transition from forced to natural circulation by subchannel code. • Effect of non-uniform heating on the thermal field. • Validation of YAOGUANG code under steady and transient conditions for LBE coolant. Based on the experimental campaign performed on the NACIE-UP (NAtural Circulation Experiment-Upgraded) facility, the benchmark activity was proposed by ENEA to qualify and validate the numerical tools for the Heavy Liquid Metal (HLM) system. Xi'an Jiaotong University participated in the benchmark exercise of subchannel validation and the self-developed subchannel code YAOGUANG was applied to the thermal hydraulic analysis. The performed two tests with different power distributions on the Fuel Pin bundle Simulator (FPS) were simulated, characterized by transition from forced to natural circulation. The applicability of this code to the stable state and mass flow rate transient was evaluated by comparing the numerical outcomes with the experimental results. Then the effect of the non-uniform heating on the thermal field was analyzed. The results indicate that this subchannel code has good accuracy in simulating both steady state and transient conditions with uniform power distribution. For the non-uniform case, the subchannel calculations can capture the general trend of temperature distributions, with a larger simulation deviation observed. This discrepancy is mainly due to the limitations of existing models adopted because they didn't consider the impacts of non-uniform conditions on the flow and heat transfer in rod bundles. Finally, the influence of non-uniform heating is mainly manifested as the increase of axial and radial temperature gradients and it helps to flatten the radial thermal field to some degree. [ABSTRACT FROM AUTHOR]

• In order to study the fluid-induced vibration of the fuel rod of a lead-cooled fast reactor core, we developed a geometric model of the wire-wrapped fuel assembly. • A detailed flow field analysis of the multi-pitch wire winding assembly case was carried out using large eddy simulation method. • A unidirectional fluid–structure coupling method was used to simulate the vibration of the fuel rod due to the fluctuation of fluid excitation force. • A combination of time domain and frequency domain analysis was used to analyze the fluid excitation force fluctuations and fuel rod vibration. • The amplitude analysis of the fuel rod at different positions in the assembly was analyzed, and the statistical histograms were drawn. Fluid excitation force is the root cause of flow-induced vibration in the fast reactor core, which is an important cause of nuclear fuel failure and breakage phenomena. In this study, the fluid–structure interaction (FSI) of a filament-wound rod bundles consisting of seven rods in a lead–bismuth fast reactor was simulated. And the fluctuation of fluid excitation force, as well as the vibration displacement and vibration frequency of fuel rods were calculated. It was found that when the fluid flow through the winding wire, the turbulence intensity increases significantly and pressure difference appear on the winding wire windward and leeward surface, which leads to fluctuation in the fluid force, which is the root cause of the emergence of flow-induced vibration. In this structural model, the vibration eventually reach a stable amplitude as well as position, and no pin-to-pin contact occurs. [ABSTRACT FROM AUTHOR]

• The three-dimensional CFD model of ELSY primary system lead pool is established. • The calculation process is simplified by multiple reasonable UDF settings, and the steady state operation condition and reactor shutdown transient process was simulated. • The thermal hydraulic phenomena under rated and natural circulation conditions are analyzed. • This work can provide an appropriate and effective reference for the design and optimization of ELSY in the future. • A complete set of thermal stratification calculation method of LFR is formed, which can provide safety assessment advice for the same type of fast reactor. Lead-cooled fast reactor (LFR) adopts the liquid metal lead as the coolant and is beneficial to the fuel sustainability and high safety. LFR is suffering from the serious thermal stratification due to the pool type design and high operation temperature, leading to the threatening thermal stress on the structure. In this paper, the three-dimensional CFD model of ELSY lead pool is established and the thermal stratification phenomena under the forced and natural circulations are investigated. The results show that under steady-state conditions, ELSY has temperature stratification along the wall in the upper lead pool where the steam generator is located, and a high temperature concentration zone and a speed stagnant zone in the lower plenum. Under natural circulation conditions, when the water decay heat removal system (W-DHR), isolation condenser (IC) and reactor vessel air cooling system (RVACS) are put into operations, there is also obvious temperature stratification in the steam generator area. This paper puts forward some corresponding optimization suggestions on the problems in the simulation. The above results can provide an appropriate and effective reference for the design of ELSY in the future. [ABSTRACT FROM AUTHOR]

• A numerical simulation method for the coupling flow and heat transfer process of liquid metal HCOTSG is proposed. • The correctness of the method is validated by comparing with relevant experimental research, and the maximum error is less than 25%. • The flow and heat transfer characteristics of liquid metal HCOTSG under different working conditions are analyzed. Much attention has been focused on the Liquid Metal Fast Reactor (LMFR) as one of the most promising concepts for GEN-IV reactors. Helical coil once-through tube steam generator (HCOTSG) is a proposed form of steam generator. Due to its unique advantages, HCOTSG is widely used in various reactor power systems, including LMFR. In this paper, a numerical approach is proposed to simulate heat transfer from the primary side (liquid metal) and secondary side (water/steam). The liquid lead-bismuth eutectic (LBE) is flowing in the shell side and pressurized water is flowing in the tube side. FLUENT CFD commercial code is adopted for the simulations. The robustness of the method is validated by comparing it with relevant experimental research, and the maximum error is less than 25%. Based on this method, the distribution of thermal-hydraulic parameters inside HCOTSG is obtained and the flow and heat transfer characteristics are analyzed. The influence of boundary conditions and geometric parameters on the flow and heat transfer in HCOTSG is analyzed. The comprehensive performance of different models is compared with the comprehensive performance evaluation index. The optimal geometric arrangement scheme for the working conditions is recommended. Compared with the reference model, the comprehensive performance can be improved by up to 14%. This study provides a numerical simulation method reference for the study of flow and heat transfer characteristics and structural design optimization of liquid metal HCOTSG. [ABSTRACT FROM AUTHOR]

• A mathematical model for high-temperature heat pipes is developed. • The relative error are 1.04% at steady state and 23.4% during startup. • The vapor velocity decrease with power. Heat pipe cooled reactor (HPR) is an innovative solid-core reactor which utilizes heat pipes to extract heat from the core to the energy conversion system. In this paper, a mathematical model for high-temperature heat pipes from a frozen state to a steady state is developed based on experimental data. The model is simplified and solved numerically. The two-dimensional transient heat conduction model of the wall and wick and the one-dimensional quasi-steady compressible vapor flow model are coupled to simulate the operation of a heat pipe when vaporization and condensation occur at the liquid–vapor interface. And the equivalent heat capacity model and the rarefied vapor model are introduced to simulate the process of heat pipe startup from a frozen state. Numerical results indicate that the model of heat pipe can predict the operation of high-temperature heat pipe with an error of less than 1.04% at steady state and 23.4% during the startup. Based on the simulation results, the velocity of vapor decreases with the heat transfer power in the steady state and the velocity of potassium vapor is 58.4% of sodium vapor velocity under the same operation condition. During the startup, when the rarefied-continuum interface located in the end of the evaporator section, the maximum vapor velocity occurs which is 1.3 times of steady velocity. This work makes it possible to predict the operation of a high-temperature heat pipe and provides a reference for the design and application of HPR. [ABSTRACT FROM AUTHOR]

• Radiative heat transfer and evaporative model of liquid metals was established. • Process of cooling and evaporation, and system performance of LDR were analyzed. • The performance of LDR adopted different liquid metals was simulated and compared. Liquid droplet radiator (LDR) is the frameless space radiator designed for heat dissipation of large-power space nuclear reactors. This paper analyzes the radiative heat transfer and evaporative characteristics of liquid metals adopted in LDR including lithium, aluminum and tin. Firstly, the cooling process of different working fluids is simulated and the temperature distribution at the droplet collector is obtained. The radiant energy of per unit mass fluid with different optical thicknesses and temperatures is calculated. Secondly, the evaporative characteristics of the working fluid are studied. Finally, the system operation performances are evaluated including the system life and radiant energy with different optical thicknesses and temperatures. Results show that liquid lithium is appropriate for heat dissipation system working under 600 K, while liquid aluminum and tin are more appropriate for working under 1200 K and 1500 K, respectively. This work provides the theoretical support for liquid droplet radiator thermal design. [ABSTRACT FROM AUTHOR]