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• The CFD-DEM method is a new method developed for the formation of debris bed from irregularly shaped debris. • The simulation for DEFOR-E7 is performed and analyzed with the debris size obtained by the SEINA code. • The deposition morphology of a debris bed with a height of 0.1 m, a repose angle of 35° and a porosity of 72 % in DEFOR-E7 test are obtained. In a severe accident in nuclear reactor, the core meltdown occurs when nuclear fuel rods cannot be sufficiently cooled, thus forming debris beds with specific shapes on the lower head of the Reactor Pressure Vessel (RPV). In this paper, the fluid–solid interaction during debris bed formation is studied based on Computational Fluid Dynamics method coupled with Discrete Element Method (CFD-DEM). The CFD models are used to simulate the flow and pressure field of fluid, the DEM models are used to calculate the collision between solids and the collision between solids and wall, as well as the fluid–solid interaction force can be calculated by the two-way coupling models. The DEM models and CFD-DEM models were validated by different experiments. Based on the above models, the DEFOR-E7 test was simulated, and a conical debris bed formed by irregular debris with a static angle of 35° and a porosity of 72% was obtained. Therefore, the method and findings in this paper are significative on theory and application for the study of nuclear severe accidents. [ABSTRACT FROM AUTHOR]

• A conceptual design of FuSTAR is developed, and validated by neutronics and thermal-hydraulics analysis. • Detailed core design and neutronics analysis with near-critical configuration and burnup calculation are carried out. • Detailed thermal-hydraulics analysis with turbulent flow and heat transfer is carried out. The fluoride-salt-cooled high-temperature advanced reactor (FuSTAR) is an advanced small modular reactor, which aims to the electric supplement and clean heat production to western China. In order to have good economy and safety, FuSTAR needs to have the characteristics of small size, light weight, deep burnup, and inherent safety. A conceptual design of FuSTAR is proposed in this study with neutronics and thermal-hydraulics assessments. In this study, its concept is established with two-region fuel to flatten core power, and the neutron spectrum of the core is in the thermal spectrum to improve the utilization of thermal neutrons and reduce fuel inventory. Neutronics and thermal-hydraulics analyses are also carried out, and established the core layout and material ratio under the near-critical condition. Burnup depth of the fuel, power distribution, neutron flux distribution, flow and heat transfer in rod-bundles channel, and 1/6 core porous media analysis are also investigated. The parameters of the designed core are evaluated quantitatively from various aspects. The results show that the core design in this study has sufficient shutdown depth, the equilibrium cycle is 3.24 years, and the discharge burnup depth is 106 GWd/MtU. The helical cruciform fuel (HCF) adopted by this reactor has a better flow and heat transfer capacity than cylindrical-type fuel at the inner channel. The analysis of 1/6 core porous media shows that the peak fuel temperature under normal operating condition is 1009.3 K, which is far below the limit of 1573 K. Relevant design and analysis results can provide the reference for the subsequent small modular FHR design and construction. [ABSTRACT FROM AUTHOR]

• Sliding mesh method was used to simulate moving conditions. • Comparing the numerical results with the experimental results. • Two calculation methods of simulating moving conditions were compared. • It is concluded that the effect of rolling motion is the most sever. • Secondary flows and flow instabilities under moving conditions were analyzed. The investigation of flow and heat transfer of lead-bismuth eutectic (LBE) under moving conditions is vital to the design optimization of LBE loop system in offshore engineering. In this paper, we have developed a method for simulating moving conditions using sliding mesh. Moreover, we have conducted experiments to validate the numerical model, and the comparison results demonstrate a good agreement between them. The calculation results demonstrate that the period of the flow rate fluctuation is consistent with the motion period, and the amplitude of the fluctuation will increase with the increase of the maximum range of motion and the decrease of the motion period. In addition, the influence of rolling on system parameters is more severe than that of heaving. These results obtained in this study have important engineering application value for the design of LBE loop system under moving conditions. [ABSTRACT FROM AUTHOR]

Numerical simulation analyses of LIFUS5/MOD2 experiment facility for hyperbaric steam-water jet inside liquid lead bismuth eutectic (LBE) were conducted using computational fluid dynamics (CFD). Alchagirov surface tension correlation and Lee evaporation model were applied in Eulerian-Eulerian two-phase flow model. Benchmark study was performed by comparing the CFD results against experimental data. Four thermal-hydraulics phenomena of sensitivity analysis were established under three extended conditions. These include 1) argon zone compression risk, 2) pressure rising and initial pressure peak, 3) argon entrained by LBE fluctuation, and 4) LBE level rising and steam cavity diffusion path. Overall, CFD models have a great advantage simulating pressure change and pressure wave transfer with 0.37% maximum error. LBE level rising and the radial and even diffusion of steam cavity should happen simultaneously under steam generator tube rupture accident (SGTR). This study provides the engineering basis for Lead-cooled Fast Reactor (LFR) design to improve its safety margin. [ABSTRACT FROM AUTHOR]

• Thermal-hydraulics of LBE is studied and an experiment is carried out. • A new empirical correlation of LBE is developed, and validated by XJTU 7-rod bundle, KIT 19-rod bundle, and ENEA 37-rod bundle. • Detailed CFD is carried out to analyze the characteristics of flow and heat transfer of LBE. • Turbulent heat transfer model of LBE is discussed and verified. Lead-bismuth eutectic (LBE) is a liquid with low Prandtl number and high thermal conductivity, generally used as the coolant in lead-cooled fast reactors of Generation-IV. The traditional Reynolds analogy is not appropriate for LBE, hence the theoretical analysis and experimental measurements are studied to interpret the flow and heat transfer of LBE in this article. In this way, a single-rod experiment on flow and heat transfer for LBE is performed and a new empirical correlation is fitted. To analyze the behavior of flow and heat transfer for low Prandtl number fluids, and also to validate the turbulence model and new correlation in the hexagonal rod bundles. The CFD simulations are performed for the XJTU 7-rod bundle, the KIT 19-rod bundle, and the ENEA 37-rod bundle based on ANSYS Fluent. The convective heat transfer coefficient of LBE, the temperature of heating wall at different sub-channels, and the temperature non-uniformity are investigated in this paper, to reveal the flow and heat transfer characteristics of LBE. The results show that the deviations of the section averaged Nusselt number are in 25%, and the deviations of the local temperature of LBE and heating wall are in 10℃. Grid spacers will result in a 30% rise in the section averaged Nusselt number, which will improve the turbulent intensity and heat transfer. Temperature non-uniformity is reduced as the Peclet number is increased. The validations demonstrate that the new empirical correlation provides accurate convective heat transfer coefficient predictions for LBE flow and heat transfer in hexagonal rod bundles with a relative derivation in 20%, which can be used as a reference for additional LBE simulations and experiments. [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]

• TaSNAM 2.0 is developed to solve 3D neutronics problem based on FVM. • The neutron parameters are obtained from the pre-established neutron parameters library. • A three-dimensional steady-state simulation of the whole core in AP1000 was completed. The Nuclear Thermal-hydraulic Laboratory (NutheL) at Xi'an Jiaotong University developed a Temporal and Spatial Neutronics Analysis Module (TaSNAM) based on the User Defined Function (UDF) and User Define Scalar (UDS) in Fluent. The neutron diffusion equations were solved based on the Finite volume method (FVM). In this paper, the TaSNAM is extended the original function to solve 3D problems and obtain neutron parameters based on the neutron parameters library, and improved to TaSNAM 2.0. The 3D-IAEA benchmark problem was selected and simulated to validate. The calculated results show good agreement with the reference values and the difference from the reference value of k eff is 40pcm. A three-dimensional steady-state simulation of the whole core in AP1000 was completed employing the TaSNAM 2.0. The calculated results of important parameters agree well with the design values and the detailed three dimensional distributions of neutron flux and power were achieved. [ABSTRACT FROM AUTHOR]

• A three dimensional CFD model for liquid entrainment based on coupled Eulerian-Eulerian VOF formulation is proposed. • The CFD model is validated with experimental results. • The effect of gas mass flow rates and vertical to horizontal branch diameter ratio on liquid entrainment are determined. • Flow characteristics during liquid entrainment phenomena through vertical T-junction are studied. In this study, a three-dimensional (3D) transient computational fluid dynamics (CFD) model is presented to investigated liquid entrainment in gas–liquid flow through vertical T-junction of fourth stage automatic depressurization system (ADS-4) in AP1000. A specialized CFD model for the liquid entrainment phenomenon study, which is based on coupled Eulerian-Eulerian VOF (volume of fluid) formulation with suitable interfacial drag and standard κ - ε turbulence model for each phase was proposed. The entrainment rate was calculated and results were validated with ADETEL experimental data, which was built at XJTU-NuTheL. The good agreement between CFD calculations and experimental data demonstrated that the proposed CFD model could reasonably predict entrainment within the studied range. The effect of vertical to horizontal branch diameter ratio and gas mass flow rates on liquid entrainment were also studied. In addition, the liquid volume fraction distributions and velocity field were investigated to develop an understanding of the entrainment process. The relatively high demand for computational resources due to very small timestep size and small grid size to accommodate high flow velocities was found to be a challenge. [ABSTRACT FROM AUTHOR]