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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]

• Validation of THACS and OpenMC for SFR has been enhanced by simulating FFTF LOFWST Test #13. • A new modeling method multi-parallel channel plus local subchannel model is proposed for reactor core. • Important local phenomenon like inter wrapper flow, thermal stratification is analyzed by the specific model. In order to support collaborative efforts within international partnerships on the development of sodium-cooled fast reactor (SFR), the International Atomic Energy Agency (IAEA) proposes a coordinated research project (CRP) about the unprotected loss-of-flow-without-scram test (LOSWOS) performed at Fast Flux Test Facility (FFTF) in 2018, which could validate simulation tool and numerical models for SFR. As one of the participants, the Nuclear THermal-hydraulic Laboratory at Xi'an Jiaotong University (NuTHeL) participates with the monte carol code OpenMC and system thermal hydraulic code THACS. This paperdeal with the standalone neutronic modeling by OpenMC and thermal hydraulic modelling by THACS. For neutronic part, the axial homogenous model was modeled to analyze the reactivity feedback coefficients and steady radial power distribution. For thermal hydraulic part, the system model, including 2D CFD outlet plenum model, subchannel model and inter-wrapper flow model, was modeled to analyze both system response and local phenomena during the LOSWOS. In general, both neutronic results and thermal hydraulic results agreed well with other participants or experiments, the reactivity feedback of GEM could provide major negative feedback, which could be used as a novel passive shutdown device. However, the remaining discrepancies were the underestimated peak temperature of PIOTA at the begin of transients, it may be caused by unknow heat transfer coefficients around hexagon tube, which has no correlation and should be further investigated by experiments. In addition, when considering IWF model, the maximum temperature of the outlet for the active fuel region was 10 K larger than the subassemblie' outlet in the PIOTA, this phenomenon should be carefully treated in loss of flow accidents of SFR. [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 boiling two-phase heat transfer experiments of liquid sodium are carried out. • The variation of parameters in typical boiling process is obtained. • The influence law of thermal parameters on heat transfer coefficient is revealed. • The new correlation of the heat transfer coefficient of boiling two-phase liquid sodium is developed. Sodium boiling will occur in some hypothetical accident cases (such as additional reactivity introduction and heat discharge capacity deterioration, and Local fault propagation). Thus sodium boiling experiments are necessary for the safety analysis and the development of serious accident programs in sodium-cooled fast reactors. In the present study, the boiling two-phase heat transfer experiments of liquid sodium were carried out on the boiling liquid sodium test loop. The experimental results showed that the entire boiling process can be divided into four stages. The heat transfer coefficient of the boiling two phases in the rod bundle channel increases gradually with the increase of heat flux and grows very slowly with the increase of system pressure. Based on the 54 groups of rod channels obtained, the new correlation of transfer heat coefficient of boiling two-phase liquid sodium was developed in a 7-rod bundle. The new correlation can well predict the experimental data of boiling two-phase liquid sodium obtained by other scholars, and the prediction error is within ± 50 \% . Finally, mean relative deviation (MRD), absolute mean relative deviation (MARD), and root mean square deviation (RMSRD) were introduced to evaluate the accuracy of the correlation quantitatively. [ABSTRACT FROM AUTHOR]

• The boiling two-phase pressure drop experiments of liquid sodium are conducted in a 7-rod bundle. • Based on the two-phase pressure drop data, the existing correlations for predicting the two-phase frictional multiplier of liquid sodium are evaluated. • A new friction multiplier for calculating the boiling two-phase flow in rod bundle channels is proposed. • The predicted value of the developed correlation is within 30% of the experimental data. The two-phase pressure drop characteristics of liquid sodium are very significant for the safety and serious incident program development analysis of sodium-cooled fast reactor (SFR). In this paper, the boiling two-phase pressure drop experiments of liquid sodium were conducted in a 7-rod bundle with the mass flux range of 80 ∼ 390 kg/(m2·s), heat flux up to 120 kW/m2 and the absolute pressure range of 7.5 ∼ 100 kPa. The experimental results show that the frictional multiplier of boiling two-phase liquid sodium in the rod bundle channel decreases with the increase of X LM . Some existing two-phase pressure drop correlations in the literatures were assessed and compared with the experimental data obtained in the present study. Results showed that these correlations could not predict the current experiments well because of the different geometries and measurements of high temperature environment. The new correlation for the two-phase friction multiplier of liquid sodium was developed in the 7 rod bundle. The predicted value of the new correlation was within 30 % of the experimental data in the rod bundle, which sufficiently demonstrates that the new correlation developed in this paper can well predict the boiling two-phase frictional pressure drop of liquid sodium in the rod bundle. [ABSTRACT FROM AUTHOR]

• An experimental phenomenon was used to verify numerical model and method. • The modeling criteria of simulating sodium entrainment by other liquids were determined by simulation results. • The relation of critical re-submergence velocity can be accurately concluded and validated. • An improved wide-applicable prediction method was proposed for entrainment situation judgment. Liquid fall type entrainment, a kind of gas entrainment phenomenon which may undesirably occur in sodium-cooled fast reactors (SFRs) due to complex immersed structures and high-velocity flowing coolants of SFRs, will involve bubbles into coolants and lead a series of safety risks. Aiming at providing a guideline for avoiding this phenomenon, the computational fluid dynamics (CFD) method was adopted to investigate the influences of fluid properties on entrainment initiation. Before the simulation case study, a small-scale experiment was conducted for verification which demonstrated the numerical model and VOF method are suitable for this study. The modeling criteria of simulating sodium entrainment by other liquids were determined by simulation results. It was also found that the relation between critical inlet velocity and σ / ρ , and the relation between critical re-submergence velocity and critical inlet velocity can be accurately estimated by the case validation. Based on which, an improved wide-applicable prediction method was proposed for entrainment situation judgment. [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 transient analysis code SALS is developed for space nuclear power system with stirling engine. • The operation characteristics of lithium-cooled reactor with solar heat flux model is simulated and analyzed. • The heat sink loss accident and reactivity insertion accident are simulated. As the power requirements for deep space exploration continue to increase, space nuclear power systems are imperative. In this paper, a set of models are established for the lithium-cooled space reactor combined with the Stirling engine. The space lithium-cooled reactor system is modeled, and the result of the steady-state is checked with the maximum relative error of 13.3%. Moreover, the characteristics under the unprotected reactivity insertion accident (URIA) and the unprotected loss of heat sink accident (LOHA) are obtained and analyzed. The results showed that: a) the solar heat flux cause the radiator temperature to fluctuate but has limited impact, b) under LOHA, the temperature of hot spot decreases rapidly from 1436 K to 1423 K, verifying the inherent safety of the system, c) under URIA, the hot spot temperature rises to 1546 K within 160 s. This work provides a solid basis for the design and analysis of space nuclear systems. [ABSTRACT FROM AUTHOR]

• Validation base of TRACE for safety analysis to SFR has been enhanced by simulating FFTF LOFWST Test #13. • Transient movement of sodium free level in Gas Expansion Modules (GEM) was reasonably reproduced. • An accurate method for thermal-hydraulic simulation of the inter-assembly heat transfer was recommended. The US NRC system code TRACE has been modified at PSI for application to liquid- metal-cooled reactor. An unprotected loss-of-flow-without-scram test performed at the Fast Flux Test Facility (FFTF) provides an opportunity to enhance the validation base of TRACE to transient analysis for sodium-cooled fast reactor (SFR). The FFTF primary system model was created with TRACE and initial core flow distribution and pressure drop in each segment of primary loop were reproduced using available data. In addition, a full-core model was built with the Serpent-2 Monte Carlo code to compute reactivity feedback parameters and delayed neutron information for point kinetics model in TRACE. Transient movement of sodium free level in Gas Expansion Modules (GEM) which was designed as a passive safety device of FFTF was simulated with TRACE using a level tracking model. A good agreement between measured and calculated total reactivity indicated a reasonable validity of modeling of feedback effects and of predicted sodium level in GEM. Multi-dimensional thermal-hydraulics effects in the FFTF vessel especially thermal stratification phenomenon which was directly related to natural circulation flow rate in primary loop were simulated with three three-dimensional VESSEL components in TRACE. Transient evolution of sodium temperatures at the Post-Irradiation Open Test Assembly (PIOTA) outlet was predicted in a good agreement with the measurements. The need of a more accurate thermal–hydraulic simulation of the inter-assembly gaps corresponding to the fuel region was discovered to obviously improve the estimation of inter-assembly heat transfer. This study represented an important step towards the validation of the TRACE code to SFR and some suggestions for further development work are proposed. [ABSTRACT FROM AUTHOR]

• Heat transfer mechanisms between subchannels of bar rod bundle has been investigated through the CFD approach. • New correlations have been proposed for the turbulent mixing and the rod conduction through fitting of numerical results. • New correlations are applicable for conditions with the P/D range of 1.08–1.30 and the Re number range of 6000–70,000. The flow field in the fuel assembly of Sodium cooled Fast Reactor (SFR) is highly non-uniformed which results in quite complex heat transfer phenomena between subchannels of rod bundles.From the design and safety purpose of the reactor core, mechanisms behind these heat transfer phenomena must be analysed thoroughly under all working conditions. However, due to the decrease of the fast reactor activity and the huge experimental costs with liquid sodium, few researches have been conducted to deepen the understanding of the heat transfer mechanisms after the 1980's in spite of the large uncertainty and limited parameter ranges of existing empirical correlations. In this paper, a Computational Fluid Dynamic (CFD) model of two subchannels with adjacent fuel rods is used to investigate the heat transfer mechanisms under a wide range of geometry and flow conditions. The turbulent heat transfer of the liquid metal is resolved through the SST k-ω turbulent model with modified turbulent Prandtl number, and the CFD model is verified with experimental data. Data analysis of the simulation results shows clearly dependence of the heat transfer mechanisms on the geometry parameters and the flow parameters. By fitting of the numerical results with the least square method, new correlations for the turbulent mixing and the rod conduction are developed. [ABSTRACT FROM AUTHOR]

• Thermal fatigue in the core outlet of SFRs due to mixing phenomenon of coolant is a critical safety issue. • Large eddy simulation method is employed to study the coolant mixing phenomenon of a simplified assembly head model of SFR. • The coolant velocity and temperature fields with vortices of different scales are obtained and analyzed. • The normalized parameters of temperature in axial and radial measuring lines and temperature fluctuations at specific points are obtained and analyzed. It is a critical safety issue of thermal fatigue in the core outlet of sodium cooled fast reactors (SFRs) due to the mixing phenomenon of coolant with different temperatures flowing from different assemblies. In this paper, the large eddy simulation (LES) method was employed to study the coolant mixing phenomenon of a simplified assembly head model of SFR. The coolant velocity and temperature fields with vortices of different scales were obtained and analyzed. It shows that the lateral velocity is more sensitive to the inlet velocity than to the inlet temperature difference. In addition, the normalized parameters of temperature in axial and radial measuring lines and temperature fluctuations at specific points were obtained. Results show that there is no obvious dominant frequency of the temperature fluctuations with the energy of temperature fluctuations distributed in the frequency range of 0–100 Hz. The results would be beneficial for the design and structure optimization of SFRs. [ABSTRACT FROM AUTHOR]

• The PIRT of SBO for China pool type SFR has been developed by 15 experts. • A math method to evaluate different opinion about PIRT results has been proposed. • The results could be developed V&V matrix and a better understanding for SFR. In order to satisfy analysis of safety requirements for China pool type sodium-cooled fast reactor, a phenomena identification and ranking table (PIRT) is developed for station blackout (SBO) accidents by 15 experts. The SBO accidents could be divided into two phases, i.e. phase 1: force circulation and phase 2: natural circulation, and the most important phenomena but maximum disagreements among experts are the gap conductance model between fuel and clad, the inter wrapper flow, the heat transfer between coolant and inter wrapper. Those phenomena should be carefully assessed and done more theorical and experiment analysis to develop more accurate and realistic model, to reduce uncertainty of safety analysis. The PIRT could be also used for developing verification and validation matrix for thermal hydraulic safety code, guiding researchers and designers to conduct experiments to close technological gap and giving related researcher and designer a comprehensive understanding for pool-type sodium-cooled fast reactor. [ABSTRACT FROM AUTHOR]