Search

Summary: In this study, two‐phase density wave instability in parallel‐twin rectangular channels was investigated with axially nonuniform heat profiles in the reactor core combined with static and rolling conditions. A parallel‐channel thermal‐hydraulic model was built using the method of two‐phase homogeneous flow developed in previous work, while the drift‐flux approach for void fraction and profile‐fit model for subcooled boiling were implemented. Although the rolling condition was chosen as the typical motion in this work, the additional force caused by the motion with six degrees of freedom was derived. The theoretical analysis was performed based on the method of small power perturbation for parallel‐twin rectangular channels. The flow oscillation caused by rolling was studied for different system parameters, including inlet resistance, exit resistance, pressure, and axial heating profile. The influence on flow instability of rolling parameters such as period, amplitude, and distance between channels was analyzed. The results showed that it would destabilize the system if a larger additional force was generated by rolling parameters. The influence of different axial heat profiles on flow instability was also studied under inlet‐peaked, cosine‐shape, and outlet‐peaked heat fluxes. The coupling effect of rolling motion and axial nonuniform heating was finally studied. The stability boundaries under different conditions were compared to the inherent boundary under the static condition with uniform heating. The results indicated that the influence of nonuniform heating was more evident and should be paid more attention to. [ABSTRACT FROM AUTHOR]

In this paper, a board type sodium cooled breed and burn reactor core is further designed and applied to perform minor actinide (MA) transmutation. MA is homogeneously loaded in all the fuel sub-assemblies with a weight fraction of 2.0 wt.%, 4.0 wt.%, 6.0 wt.%, 8.0 wt.%, 10.0 wt.% and 12.0 wt.%, respectively. The transmutation efficiency, transmutation amount, power density distribution, neutron fluence distribution and neutronic safety parameters, such as reactivity, Doppler feedback, void worth and delayed neutron fraction, are compared under different MA weight fraction. Neutronics and depletion calculations are performed based on the self-developed MCNP–ORIGEN coupled code with the ENDF/B-VII data library. In the breed and burn reactor core, a number of breeding sub-assemblies are arranged in the inner core in a board type way (scatter load) to breed, and a number of absorbing sub-assemblies are arranged in the inner side of the outer core to absorb neutrons and reduce power density in this area. All the fuel sub-assemblies (ignition and breeding sub-assemblies) are shuffled from outside in. The core reached asymptotically steady state after about 22 years, and the average and maximum discharged burn-up were about 17.0% and 35.3%, respectively. The transmutation amount increased linearly with the MA weight fraction, while the transmutation rate parabolically varied with the MA weight fraction. Power density in ignition sub-assembly positions increased with the MA weight fraction, while decreased in breeding sub-assembly positions. Neutron fluence decreased with the increase of MA weight fraction. Generally speaking, the core reactivity and void worth increased with the MA weight fraction, while the Doppler feedback and the effective delayed neutron fraction decreased with the increase of the MA weight fraction. [ABSTRACT FROM AUTHOR]

In this paper, a preliminary board type radial fuel shuffling sodium cooled breed and burn reactor core is designed. In the current design, a number of breeding subassemblies are arranged in the center core to ensure enough breeding. A self-developed MCNP-ORIGEN coupled system with the ENDF/B-VI data library is applied to perform neutronics and burn-up calculations. For a 2.0 m radius and 2.5 m height core, the results demonstrate the feasibility of the board type radial fuel shuffling strategy. Breeding mainly occurs in the breeding subassemblies during the first 6 fuel cycles as they are moved to the burning/breeding region. The core will become asymptotically stable after about 24 years. The discharged burn-up of most subassemblies is about 15.0–30.0%. The influences of the core size on the major core parameters, such as initial k eff , steady k eff , maximum power density, peak burn-up and burn-up ratio between breeding and ignition subassemblies are calculated and investigated. The results indicate that the initial k eff increases with fuel height and core radius and finally reaches stability; the steady k eff increases with fuel height and core radius, then reaches peak value and finally decreases; the maximum power density, the peak burn-up and the burn-up ratio between breeding and ignition subassemblies decrease with the increase of fuel height and core radius; if core radius is less than 1.875 m, they increase sharply with the decrease of core radius. [ABSTRACT FROM AUTHOR]

The SACOS code is thermal-hydraulic subchannel analysis code designed to meet steady-state conditions and operational transients in different types of reactor cores. It uses a homogeneous flow or drift-flux model to simulate single-phase and two-phase flow characteristics. In order to achieve a more refined two-phase flow modeling for LWR cores, a new eight-equation two-fluid based analysis module of SACOS is developed in this study. The phasic equations for mass continuity, momentum, and energy are spatially discretized by the finite volume method (FVM) with a staggered grid structure. To reduce numerical diffusion and improve numerical accuracy, the high-resolution scheme with nonlinear flux limiter is implemented. The code assessment is performed with linear advection flow problem and the preliminary code validation is presented by comparing with the experimental results of Weiss 14 × 14 blockage test, CNEN 4 × 4 mixing test, Studsvik 3 × 3 mixing test, and ISPRA 4 × 4 mixing test covering single/two-phase and unheated/heated conditions. The application of SACOS implies that the code can be used to model and perform assembly-level core safety analysis with visualization capabilities for hexagonal and rectangular geometry. Development of a two-fluid based thermal-hydraulic subchannel analysis code and its application. [Display omitted] • A new eight-equation two-fluid based analysis module of SACOS is developed. • The high-resolution scheme with nonlinear flux limiter is implemented into code. • The validity of the higher-order scheme is verified with linear advection flow problem. • The two-fluid model is validated based on reported bundle experiments. [ABSTRACT FROM AUTHOR]

• Two quick-analysis models are investigated and compared in theory and application. • The simulations verify and compare the two models via the typical sodium heat pipe frozen startup experiment. • Inherent defects and targeted model improvements of the two models are proposed. The alkali-metal high-temperature Heat Pipe (HP) is the common component connecting the reactor core and energy conversion device in Heat Pipe cooled Reactor (HPR). Due to the particularity of the working medium, the frozen startup of HP makes a great influence on the startup process of HPR system. In this paper, two quick-analysis models, the Improved Thermal Resistance Network (ITRN) method and the Pseudo Wick Thermal Conductivity (PWTC) method have been investigated and compared. The simulations of a sodium HP in a typical experiment are carried out by two methods in COMSOL. For simulation speed, the convergence time of a single HP using the quick-analysis models is about tens to hundreds of seconds depending on the computing equipment. The convergence of PWTC is 26.7% faster than that of ITRN. For simulation accuracy, except for the inevitable error near the flat-front, the relative error of ITRN is less than 6.94% and the relative error of PWTC is less than 8.24%. The maximum effective error of ITRN method is 34.84 K, while that of PTWC method is 58.21 K at the initial stage of startup. The T Transition = 651 K in PWTC is lower than the T exp. = 710 K in the experiment and the T Tr = 701 K in ITRN, which affects the accuracy at the initial stage of startup. The error will decrease gradually in the thawing front advancing stage. For simulation analysis, some inherent defects and targeted model improvements of the two models are proposed. Through the detailed comparative study in this work, researchers can easily apply the appropriate model according to the specific requirements and avoid the defects during the simulation process. [ABSTRACT FROM AUTHOR]

• Concept of "LESS is MORE" is proposed in the design of lead-bismuth cooled reactor. • Core design of lead-bismuth cooled small modular reactor is conducted. • Neutronic and thermal-hydraulic performances of reactor core are analyzed and improved. • Preliminary core design of LBE cooled reactor could satisfy the design criteria and is reasonable. Small modular reactors (SMR) are in great demands in the near future and the lead-based fast reactor is one of the most important directions in the development of SMR. In this paper, core design and analysis of a LEad-bismuth Small MOdular Reactor (LESMOR) was developed and examined to realize the notion of "LESS is MORE". LESMOR is a 300 MWt pool-type reactor where all facilities are mounted in a reactor vessel. Advanced nitride fuel is adopted, and the spent-plutonium is used as the driven fuel and thorium is used as the fertile fuel. Sub-channel analysis was performed in the assembly design using an in-house sub-channel code SUBAS, and the 11 × 11 scheme with a pitch-to-diameter (P/D) ratio of 1.4 was adopted. The full core neutronic performance was studied and evaluated in this paper. Results show that the reactor can be maintained critical for 20 years and the maximum radial power factor is 1.28. The steady-state thermal-hydraulic performance of the core was analyzed, including the characteristics of the mass flow distribution and the hottest assembly. The peaking fuel cladding temperature is 519.2 °C, and the maximum fuel center temperature is 723.4 °C, both within the temperature limit. This paper provides valuable information in the neutronic and thermal-hydraulic design and analysis of LBE nuclear reactor. [ABSTRACT FROM AUTHOR]

• An innovative three-dimensional thermal–hydraulic code, CorTAF, was developed based on OpenFOAM. • The CorTAF is more computationally efficient with the spatial resolution of subchannel level. • The CorTAF is validated against experiment benchmarks and subchannel analysis codes. • The steady state and transient simulations of a full scale PWR core were carried out using CorTAF. In this paper, a thermal hydraulic analysis method based on open source CFD platform OpenFOAM was proposed. The models of coolant flow and heat transfer, fuel rod heat conduction and coupled heat transfer were established according to the bundle structure characteristics of PWR core, and then a nuclear reactor three-dimensional thermal–hydraulic characteristics analysis code CorTAF was developed based on finite volume method with the spatial resolution of subchannel level. The international benchmarks, including GE3 × 3, Weiss and PNL2 × 6 fuel assembly flow and heat transfer experiments, were selected to carry out the validation. The calculation results were basically consistent with the experimental data, illustrating that CorTAF is suitable for predicting the flow and heat transfer characteristics of coolant in the rod bundle fuel assembly. The thermal–hydraulic characteristics of fuel assembly and PWR core under full power operation and blockage accident conditions were simulated using CorTAF. The steady-state and transient spatial distribution of physical quantities in subchannel-scale including coolant and fuel rod temperature were obtained, and the influence of blockage condition on flow and heat transfer characteristics was analyzed. This work has reference significance for the development of core thermal hydraulic analysis tools for PWR. [ABSTRACT FROM AUTHOR]

Summary: To meet the requirements of the unmanned underwater vehicles (UUVs) for the power source, a conceptual design of a 100‐kWe level Nuclear Silent Thermal‐Electrical Reactor (NUSTER‐100) is proposed. The NUSTER‐100 is a 1‐MWt reactor that has been designed to couple with a 100‐kWe thermoelectric generator system. The high temperature sodium heat pipes are adopted to cool the reactor core. The cascaded wide temperature range thermoelectric generation (TEG) system is used to achieve thermoelectric conversion, which ensures high power generation efficiency and low operating noise. The neutronic and thermal‐mechanical analysis results are provided for the design. The core power distribution and energy spectrum are obtained. The neutronic analysis results are imported into ANSYS as input for thermal‐mechanical analysis. In this paper, the fuel‐HPs channel and the 1/8 symmetrical core are selected as the analysis objects, and the temperature distribution and statics parameters under the normal operation condition are obtained. The analysis results show that under normal operation conditions, the core parameters meet the design requirements. These design parameters are all lower than the safety limit. This work provides a certain reference for the design and application of the heat pipe reactor power system. [ABSTRACT FROM AUTHOR]

• This study utilizes the increasing flow regime experimental data to extend the prediction range and improve the accuracy. • Integrated flow regime prediction models considering vertical and horizontal flow direction were developed and evaluated. • The MLP, RF, and SVM algorithms were utilized to develop the flow regime prediction model respectively. • The flow regime prediction model shows better performance than existing models in three directions respectively. The nuclear reactor core is the pivotal component in a nuclear power plant to generate and transfer heat, so accurate prediction of reactor core flow and heat transfer characteristics is a crucial problem for the reactor system design. Flow regime is closely related to the thermal–hydraulic characteristics of two-phase fluid. Still, flow regime models used in thermal–hydraulic calculation codes rely heavily on earlier experimental data, featuring a narrow application range and unsatisfactory accuracy. To extend the prediction range and improve the accuracy of flow regime prediction by making full use of the increasing experimental data, integrated flow regime prediction models considering vertical flow direction for light water reactor (LWR) and horizontal flow direction for Canadian Natural Deuterium Uranium (CANDU) were developed and evaluated based on three standard machine learning algorithms in this study. Firstly, the experimental data of horizontal and vertical flow collected from the literature was modified and preprocessed to be training data. Then, the multi-layer perceptron (MLP) algorithm, random forest (RF) algorithm, and support vector machine (SVM) algorithm were utilized to develop the flow regime prediction model respectively. The performance of the three models was evaluated and compared and the results showed that the flow regime prediction model based on RF was the optimal model with higher prediction and more efficiency than the other two models. Finally, the flow regime prediction model was compared with existing models in three directions respectively, which indicated that the range and accuracy of the model based on RF were superior to the existing models. This study provides an implantable and scalable approach for flow regime prediction, and the application range and accuracy of flow regime prediction can be continuously improved with updating training data. [ABSTRACT FROM AUTHOR]

Fuel pins assembled in Pressurized Water Reactor (PWR) and Liquid Metal-cooled Fast Reactor (LMFR) core are usually cooled by single-phase coolant such as water or liquid metal. Researches on the coolant flow and heat transfer phenomena through rod bundles is quite essential for the nuclear reactor core design. This paper briefly summaries the current developments of Computational Fluid Dynamics (CFD) studies on the thermal hydraulic analysis of single-phase flow in the rod bundles of nuclear reactor core. On account of the tremendous amount of publications related to each aspect above, this brief review would focus on the most representative researches in the past decades. The aim of this paper is to provide an outline and present a comprehensive overview of the CFD research and guidelines on single-phase flow through bare rod bundle rod with mixing vane spacer and wrapped wire spacer in reactor core. In addition, the analysis and validation methods for various CFD calculating methods are concluded. The applications of CFD approach with different type of rod bundles are summarized. Finally, some remaining challenges for future work are proposed and discussed. [ABSTRACT FROM AUTHOR]

The lead-bismuth cooled fast reactor core adopts box-type design, and the fuel rod bundle is fixed by wrapping wires. The complex geometric structure of the fuel assemblies significantly affects the flow and heat transfer status of the liquid metal coolant. Building upon the open-source CFD platform, OpenFOAM, this work established conservation equations for liquid metal coolant, fuel rod thermal characteristic models, inter-wrapper coolant analysis models, and coupled heat transfer models, and developed the subchannel level thermal-hydraulic analysis code CorTAF-LBE for whole-core analysis of LBR core using the finite volume method. The wire-wrapped rod bundles internal flow and heat transfer experiments of lead-bismuth, carried out based on the THEADES and KYLIN-II circulation loops, were selected to perform verification and validation. The results match well with the experimental data and CFD fine-modeling prediction. The steady-state numerical simulation of three wire-wrapped rod bundle assemblies with inter-wrapper flow was carried out under normal operation condition. The temperature and pressure distributions of the coolant in each subchannel and inter-wrapper flow region, the temperature of cladding and other key thermal-hydraulic parameters were obtained. The role of the inter-wrapper coolant in heat removal from the fuel assemblies was revealed. The work in this paper has referential value for the optimization of lead-bismuth cooled fast reactor design and safety assessment during operation. [ABSTRACT FROM AUTHOR]

Summary: Numerical simulation of the Space Reactor Power System is of great importance to reactors development and application in aerospace. In this paper, thermal‐hydraulic and thermoelectric characteristics analysis based on 1/6th full thermionic reactor core of the TOPAZ‐II were performed using 3‐D computational fluid dynamics software STAR‐CCM+. A Benchmark study was conducted by comparing the numerical results with experimental data and design values with a maximum error of 5.4%‐11.8%. For the thermal‐hydraulic analysis: (a) the highest temperature point occurred at the center hole of the center thermionic fuel element (TFE) with the temperature of 2303 K; (b) the maximum temperature difference located in the fuel region and interelectrode gap with the value about 200 K and 1000 K, respectively; (c) axial temperature distribution of coolant is similar in each flow channel, and the maximum temperature difference appears at the reactor core outlet with the value of 10 K. For the thermoelectric analysis, the potential on center or the edge TFE is much higher or lower than the others, the calculated electrical power of the full thermionic reactor core is 4.95 kW, the distribution trend of current and potential is in a good agreement with the axial emitter temperature. [ABSTRACT FROM AUTHOR]

Summary: Closed type assembly leads to complicated core thermohydraulics which are significant for design and safety analysis in sodium‐cooled fast reactor (SFR). Most of the major phenomena can be dealt with by general parallel channel network and two‐dimensional model for inter‐wrapper flow (IWF) in system code transient thermal‐hydraulic code for analysis of SFR (THACS). A main deficiency left is the overestimation of inter‐assembly heat transfer effect due to lump temperature message that one‐dimensional assembly model can only provide. A methodology was thus developed to compute wall‐subchannel temperature for simulating energy transportation through duct wall. Intra‐assembly thermohydraulics were classified as two mechanisms and non‐dimensional parameter was introduced for characterizing each of them. Ratio between linear heat flux density through hexagonal wrapper surface and pin surface qw/p was employed to account for radial heat transfer. Flow redistribution effect arouse in mixed and natural convection was concretized by a importance symbol of buoyancy force, Richardson number Ri. Two criteria were also proposed for dividing flow state into three parts and specific model for each section was defined. Two validation cases against 3‐D computational fluid dynamics (CFD) simulation of China experimental fast reactor whole core and test data from plant dynamics test loop with direct heat exchanger experiment were performed with THACS. Noticeable improvement after incorporation of this model demonstrated its value in system analysis. Although the procedure for parameter computations is geometry related, this method can be reproduced to any type of assembly directly. [ABSTRACT FROM AUTHOR]

Summary: The development of deep space exploration requires space nuclear reactor with higher power and better endurance. In order to meet the increasing high power requirements for space applications, an Open‐grid MEgawatt Gas‐cooled spAce nuclear reactor (OMEGA) is proposed in the paper, which is featured with low specific mass and high energy conversion coefficients. A set of models, including neutron kinetics model, reactivity feedback model and heat transfer model are established. Besides, the space nuclear system analysis code (System Analysis code of MEgawatt gas‐cooled space Reactor [SAMER]) is developed by finite volume method and staggered grid technique for the preliminary safety analysis of the OMEGA conceptual design. Furthermore, the calculation results provided by the SAMER code agree well with the computational fluid dynamics simulation, validating the reliability of the code. Moreover, the transient thermal‐hydraulic responses of the OMEGA under the unprotected reactivity insertion accident (URIA) and the unprotected loss of partial coolant flow accident (LOFA) are obtained and analyzed. The maximum fuel temperature after the URIA has a margin of 262 K from the fuel temperature limit value. The total power after the LOFA is about 68.7% of the rated power, meanwhile, the fuel radial temperature distribution tends to be more uniform with the fuel average temperature unchanged. It can be found that the OMEGA design shows satisfactory safety characteristic and reliability under the URIA and the LOFA scenarios. This paper is the stage summary of the conceptual design work of the OMEGA, and it may provide useful reference for megawatt space nuclear system design and thermal‐hydraulic analysis. [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]

The employment of heat pipe reactors is gaining significant traction due to their exceptional advantages in terms of mobility, safety, and applicability across a wide range of scenarios, including strategic defense locations, remote communities, and situations of power shortage. Importantly, most of the heat pipe reactors are designed as solid-state reactors, which avoids the challenges associated with two-phase flow problems that commonly arise from coolant transformation and makes it possible to analyze the reactor through the integration of multiple physical field coupling. In this paper, we conduct a 3D neutronic-thermal-mechanical coupled performance analysis of a heat pipe reactor operating under steady-state conditions. This analysis employs a combination of the Monte Carlo code and finite element analysis software, facilitating in-depth insights into the behavior of reactor. Coupling calculation results show that under normal operation, the maximum temperature and thermal expansion of core is 1338.94 K and 2.53 mm. In the steady-state coupling process, the influence of temperature parameters on neutronic calculation is mainly reflected in k eff and its power level, the maximum temperature drops from 1347.26 K to 1338.94 K, but it has little influence on power distribution. By coupling calculations, the maximum stress of the reactor is reduced by 5.78 MPa. The research results will support the subsequent development of multi-physics coupled transient analysis code for heat pipe reactors. [ABSTRACT FROM AUTHOR]

The thermal hydraulic and neutronics coupling analysis is an important part of the high-fidelity simulation for nuclear reactor core. In this paper, a thermal hydraulic and neutronics coupling method was proposed for the plate type fuel reactor core based on the Fluent and Monte Carlo code. The coupling interface module was developed using the User Defined Function (UDF) in Fluent. The three-dimensional thermal hydraulic model and reactor core physics model were established using Fluent and Monte Carlo code for a typical plate type fuel assembly, respectively. Then, the thermal hydraulic and neutronics coupling analysis was performed using the developed coupling code. The simulation results with coupling and noncoupling analysis methods were compared to demonstrate the feasibility of coupling code, and it shows that the accuracy of the proposed coupling method is higher than that of the traditional method. Finally, the fuel assembly blockage accident was studied based on the coupling code. Under the inlet 30% blocked conditions, the maximum coolant temperature would increase around 20°C, while the maximum fuel temperature rises about 30°C. The developed coupling method provides an effective way for the plate type fuel reactor core high-fidelity analysis. [ABSTRACT FROM AUTHOR]

Summary: A thermoelectric analysis of space nuclear power reactor is of great importance to the space reactors development. In this paper, the reactor core model of TOPAZ‐II designed by Soviet Union, including neutronics model, thermal‐hydraulic model, and electrical circuit model, is established based on reasonable assumptions. A system analysis code is developed to analyze the thermoelectric characteristics of the TOPAZ‐II under the condition of steady‐state operation, start‐up procedure, power change mode, and reactor shutdown. The code has been benchmarked with experimental data, and the maximum relative error is 16.6%. Numerical results show that for the steady state, the simulated electrical power is 5.2 kW within the design value range. For transient state, the reactor thermoelectric characteristics are mainly affected by the electrode temperature: (a) For the start‐up procedure, when the emitter temperature increases above 1700 K, electrical system begins to work and reach full power in 5 minutes. (b) For power change mode, the emitter temperature decreases by 5%, while the electrical power decreases by 67%; (c) For reactor shutdown, electrical power reduces to 0 kW as the emitter temperature decreases to 1400 K in 100 seconds. This study provides valuable theoretical supports for the design and analysis of the thermionic space nuclear power reactor. [ABSTRACT FROM AUTHOR]

• A new boron diffusion model is proposed for the nuclear reactor core analysis. • The effect of turbulent diffusion is taken into consideration. • The model is implemented in the subchannel analysis code. • The CFD method is used to verify the new model. The diffusion characteristic of borated water in nuclear reactor core is important for the reactivity control during some possible reactivity transient scenarios. The non-uniform distributions of boron concentration increase the risk of reactor accidents and lead to some reactor safety problems. Hence it is necessary to develop an effective boron diffusion model into the reactor thermal-hydraulic analysis code. In this paper, a quasi-three dimensional boron tracking model, in which the turbulent diffusion effect was considered carefully, was proposed and successfully implemented in the subchannel analysis code. The reliable CFD method with k-omega turbulent model validated against International Standard Problem (ISP No. 43) experimental data was used to verify the feasibility of proposed boron tracking model. A typical Pressurized Water Reactor (PWR) 3 × 3 fuel bundle model was established and the borated water was injected into one specified channel and pure water into other channels. The distributions and variations of boron concentration along the flow direction were achieved and analyzed. Results indicate that the proposed boron tracking model could predict the boron diffusion process in the fuel rod bundles more accurately compared with the original model. This work provides an essential supplement for the nuclear reactor subchannel analysis method. [ABSTRACT FROM AUTHOR]

In this paper, a full scale three-dimensional thermal-hydraulic analysis method for plate-type fuel reactor core is proposed through integrating traditional sub-channel analysis concepts into CFD platform. The coupled heat transfer model between fuel plates and narrow rectangular coolant channels is established. The three-dimensional thermal-hydraulic analysis code CorTAF-PT is developed based on open-source CFD platform OpenFOAM. The CorTAF-PT is validated against with the Qiu-TAKAHASHI experiment and IAEA 10 MW MTR. The code simulation results agree well with the experiment and MTR operation data with relative error within 5%. The full-scale MTR simulation under steady state and LOFA is carried out through channel-level resolution. In partial blockage accident, maximum temperature of coolant, fuel pellet and cladding increase by 9.75 K, 60.65 K and 60.52 K respectively. Furthermore, the double temperature peaks of fuel plates and local coolant reflux with −0.121 m s−1 are revealed under partial blockage accident. This work provides reference value for correlational studies on thermal-hydraulic characteristics of plate-type fuel reactor core. • CorTAF-PT, a full scale three-dimensional thermal-hydraulic analysis code for plate-type fuel reactor core is developed. • CorTAF-PT can efficiently save the computing resource while remaining refined simulation accuracy. • The CorTAF-PT is validated against with the Qiu-TAKAHASHI experiment and IAEA 10 MW MTR. • Transient analysis under partial blockage accident of MTR is carried out while the local coolant reflux is revealed. [ABSTRACT FROM AUTHOR]

Lead-cooled Fast Reactors (LFRs) has become a main candidate choice among the fourth-generation reactors, and Steam Generator Tube Rupture accident (SGTR) is one kind of its important Design Extension Condition (DEC). Due to the limitations of existing theoretical research and simulation tools, many SGTR thermal-hydraulic experiments have been carried out recently. This paper reviews thermal-hydraulics stages into four stages: 1) water-LBE interaction together with steam explosion, 2) high-pressure water jet and structures impact, 3) steam bubbles transport in primary system, 4) bubble entrained and stayed in the reactor core. The existing SGTR experiments are classified as separate SGTR effect test experimental facilities and separate LBE-water interaction experimental facilities. The methods of high-pressure injection, LBE status change and parameters measurement of separate SGTR effect experimental facilities are introduced. Four experimental prospects for further research are suggested as pressure transmission and structure impact, steam lifting and visualization, bubble transport in full-scale or reduced-scale model of reactor, behaviors of LBE-water mixture. • An LFR experimental review of SGTR thermal-hydraulic experiments. • SGTR thermal hydraulic phenomena are divided. • SGTR thermal hydraulic experiments are classified into two categories. • Experimental methods introduction and further research suggestion. [ABSTRACT FROM AUTHOR]

In order to explore the core corrosion characteristics of horizontal lead-bismuth reactor in oxygen-controlled environment, a liquid lead-bismuth eutectic (LBE) corrosion model including the growth and removal process of oxide layer was established in this paper. The coupling method of double-layered oxidation model and computational fluid dynamics (CFD) was used to simulate the long-term oxidation corrosion of horizontal LBE reactor core numerically. The results show that the average total oxide layer thickness of fuel rod surface is 1.59 μm, and the thickest oxide layer is 3.48 μm on the fuel rod surface at the core outlet after 20,000 h. The fuel rods maintain a double-layer oxide layer structure. The growth rate and removal rate of the oxide layer on fuel rod surface increase with the increase of inlet temperature and the decrease of inlet flow rate. The growth rate of the oxide layer on fuel rod surface increases with the increase of inlet oxygen concentration, while the removal rate decreases. At higher temperature above 420℃, lower flow velocity below 0.25 m/s and lower oxygen concentration below 2 × 10−8wt%, the magnetite layer will first be completely removed from the surface of the fuel rods at the core outlet. The liquid LBE oxidation corrosion model and coupled CFD method presented in this paper can be used for numerical simulation of oxidation corrosion in different units of LBE cooling system. [ABSTRACT FROM AUTHOR]

Thermal-hydraulic performance in the reactor core was an essential factor in the nuclear power plant design. In this study, we analyzed the three-dimensional (3-D) thermal-hydraulic characteristic of reactor core based on porous media method. Firstly, a 3-D rector pressure vessel (RPV) model was built, including the inlet leg nozzle, downcomer, lower plenum, reactor core, upper plenum and outlet leg nozzle. Porous media model was used to simplify the reactor core and upper plenum. The commercial CFD code ANSYS CFX was employed to solve the governing equations and provide the 3-D local velocity, temperature and pressure field. After appropriate parameters and turbulent model being carefully selected, the simulation was validated against the 1:5 scaled steady-state hydraulic test. The predicted hydraulic parameters (normalized flowrate distribution and pressure drop) were in good agreement with the test results. And the predicted thermal parameters agreed well with the designed values. The validation indicated that this method was practicable in analyzing the 3-D thermal-hydraulic phenomena in the RPV. Finally, the thermal-hydraulic features in reactor core were analyzed under the condition of the Steam Generator Tube Rupture (SGTR) accident. The simulation results showed that the coolant temperature increased gradually from the center to the periphery in the reactor core in the accident. But the temperature decreased to safety level rapidly after the reactor shutdown and safety injection operation. The reactor core could keep in a safe state if appropriate safety operations were performed after accidents. [ABSTRACT FROM AUTHOR]

Core degradation evaluation of probability, progression and consequences of a core degradation accident is critical for evaluation of risk as well as its mitigation. However, research and modeling of severe accidents to date are limited, and their accuracy in predicting severe accident consequences is still insufficient. It is therefore important to explore the mechanisms of core degradation and to develop mitigation measures for severe accidents. PHEBUS FPT1 is a typical and classic core degradation experiment. MELCOR is a world famous severe accident analysis code developed by Sandia National Lab that has seen wide application, a broad user base, and a number of supporting experiments. The PHEBUS experiment was simulated using MELCOR in this paper. Experimental data on, thermal power and steam mass flow rates are used to determine average pressure, energy distribution, molten mass, temperature of the fuel, and hydrogen generation. Data from the PHEBUS experiment and Cho’s calculations are used to compare the average pressure, several fuel temperatures and the hydrogen generation rate. The results confirm the validity of MELCOR’s simulation of the PHEBUS experiment. The temperature distribution of the core is provided. These results are used to determine the mode and behavior of core degradation with the intent of building a foundation for further research. [ABSTRACT FROM AUTHOR]

Severe accident analysis is important for the safety evaluation of a nuclear power plant (NPP). In this paper, analysis has been performed for a Chinese three-loop Pressurized Water Reactor (PWR) severe accident induced by loss of coolant accident (LOCA) along with Station- Block-Out (SBO) using the MIDAC code. The simulation results show the influence of different break sizes in the cold leg on the severe accident progression. Important parameters, such as primary coolant system (RCS) pressure, core melting time, pressure vessel rupture time. are analyzed with three different break sizes in the cold leg. With the break size of 0.002 m 2 , the reactor core starts to melt at 1481 s and the lower head fails at 10,317 s after the accident occurs. With a larger break size of 0.05 m 2 , the reactor core starts to melt at 1066 s and the lower head fails at 2473 s. If the break size is further enlarged to 0.2 m 2 , the reactor core starts to melt at 422 s and the lower head fails at 1757 s. Another case in which a hot leg break is with the size of 0.002 m 2 is performed. The reactor core starts to melt at 3641 s and the lower head fails at 14,744 s. It is less severe than the cold leg break accident. The severe accident prevention and mitigation measures are summarized based on the computation results. The results are helpful to develop the management strategies and guideline for the severe accident of the similar types of PWRs. [ABSTRACT FROM AUTHOR]

The state of the nuclear reactor core is significant, and the corrosion products will be deposited in the core, threatening its safety. In this paper, the analytical model of fouling deposition and migration process is established, and the core corrosion deposition-thermal-hydraulic coupling analysis method is proposed to study the fouling deposition and its effect on the flow and heat transfer state under the long-term operation of the typical 5 × 5 rod bundle channel. The results show that the established model is reasonable. The distribution of fouling in the bundle is related to the position. Under the operating conditions calculated in this paper, the amount of fouling in the outer layer bundle is higher than that in the inner layer. And the fouling rate is positively correlated with the wall temperature. The fouling rate will be greater in the place with higher wall temperature in the bundle. The content of this study could provide reference for the study of reactor core fouling and its heat transfer characteristics. [ABSTRACT FROM AUTHOR]

Experience gained in decades of nuclear safety research and previous nuclear accidents direct to the investigation of passive safety system design and accident-tolerant fuel (ATF) system which is now becoming a hot research point in the nuclear energy field. The ATF system is aimed at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the current UO 2 –Zr system when the reactor is in normal operation. By virtue of advanced materials with improved properties, the ATF system will obviously slow down the progression of accidents, allowing wider margin of time for the mitigation measures to work. Specifically, the simulation and analysis of a large break loss of coolant accident (LBLOCA) without ECCS and extended station blackout (SBO) severe accident are performed for a pressurized water reactor (PWR) loaded with ATF candidates, to reflect the accident-tolerance of ATF. [ABSTRACT FROM AUTHOR]

• A multi-physics coupling analysis platform is established for the static heat pipe reactor NUSTER-100. • The thermal-electrical coupling characteristics of the static heat pipe reactor are investigated under different operation conditions. • The unloading condition will cause the core temperature to rise significantly. Heat pipe cooled reactor with static conversion is gaining more and more attention because of its unique advantages such as mobility, high reliability, simple structure, long lifetime, and high-power density. A 100 kW heat pipe reactor with the three-segment type thermoelectric generators namely NUSTER-100 is designed by Xian Jiaotong University. Based on the finite element software COMSOL Multiphysics, a multi-physics coupling analysis platform is established for the static heat pipe reactor NUSTER-100. The heat transfer models of the heat pipe and thermoelectric models of the three-segment type TEGs are supplemented in the simulation model. The thermal-electrical coupling characteristics of the static heat pipe reactor are investigated under different operation conditions. For the steady state condition, the temperature difference of the heat pipe between evaporator and condenser stabilizes at around 110 K, the main temperature gradient about 850 K appears between the two side of TEGs. The total output power of the NUSTER system is 123 kW and the system thermoelectric conversion efficiency is 12.3%. For the three rows unloading condition, the maximum temperature increases to 1483 K which is 133 K higher than that on the steady state condition. The output electrical power of the reactor system is 114.5 kW and the system thermoelectric conversion still could reach 11.5%. For the whole reactor unloading condition, the reactor core temperature increases by 393 K compared with the steady state and the maximum temperature is 1743 K which is lower than the core melting temperature. The operating temperatures of the heat pipes are still within the normal operating temperature range of the sodium heat pipe. The heat transfer capacity of the heat pipe is still sufficient to ensure that the core temperature is kept within the safe limits. In summary, the unloading condition will cause the core temperature to rise significantly. Therefore, the system safety characteristics under accident transient conditions need to be focused coupling with the unloading condition in the further research. [ABSTRACT FROM AUTHOR]

Sub-channel analysis can improve the accuracy of reactor core thermal design. However, the important initial parameters contain various uncertainties during reactor operation. In this work, the Sub-channel Analysis Code of Supercritical reactor (SACOS) code, which is also applicable for Pressurized Water Reactor (PWR), was used to study the coolant flow characteristic and fuel rod heat transfer characteristic of 1/8 assembly which has the maximum linear power density in 300 MWe PWR core firstly. Then the Wilks' method and Response Surface Method (RSM) were utilized to determine the influence of sub-channel input parameters uncertainties on the highest temperature of reactor core fuel rod and Minimum Departure from Nucleate Boiling Ratio (MDNBR). The results show that in the most conservative conditions, the maximum temperature of the fuel rod and MDNBR were 2167.4 °C and 1.08, respectively. Considering the uncertainties of assembly inlet flow rate, inlet coolant temperature and system pressure, the 95% probability values (with 95% confidence) of fuel rod maximum and MDNBR calculated using response surface methodology were 2144.0 °C and 1.6, while they were 2137 °C and 1.74 calculated by Wilks' approach. Results show that the uncertainty analysis methods can provide larger reactor design criteria margin to improve the economy of reactor. Furthermore, the code was developed to have the capacity to perform the uncertainty study of sub-channel calculation. [ABSTRACT FROM AUTHOR]

• The thermal behavior of a horizontal high-temperature heat pipe under motion conditions are experimentally investigated. • The motion conditions include transitory inclination, vertical heave, and periodic swing. • The impacts of motion types on temperature variation and heat transfer deterioration of heat pipes are summarized. High temperature heat pipe is widely applied to multiple industrial fields, including in small modular reactors to cool the reactor core, or as heat exchangers. The thermal behavior of the high temperature heat pipe has been widely researched by theoretical and experimental approaches, including designing, manufacture, numerical analysis, and experimental test on its thermal characteristics of start-up process and operation. As the heat pipe has a potential application protest on mobile nuclear reactors, this paper carried out experimental study on the thermal behavior of the high temperature heat pipe under various motion conditions. The temperature distributions on the heat pipe surface under transitory inclination, vertical heave, and periodic swing, are measured to evaluate the impact of motion conditions on the heat transfer of heat pipe at horizontal position. The transitory inclination causes an obvious temperature variation at both evaporator and condenser. The vertical heave, of which the motion direction is perpendicular to the heat pipe, has negligible impact on the isothermality of the heat pipe. The periodic swing leads to periodic temperature fluctuation, of which the amplitude increase with the decease of the period time. Besides, the dry out phenomenon is easier to happen at the evaporator when the condenser is longer. This paper summarizes the impact of different motion types on the thermal behavior of a horizontal heat pipe and provides a reference for the design and application of the high temperature heat pipe used in nuclear reactors. [ABSTRACT FROM AUTHOR]

Due to the independence of coolant channels, the occurrence of channel blockage will affect the flow and heat transfer characteristics of plate-type fuel assembly significantly, which may lead to coolant boiling and threaten the structural integrity of reactor core. In this work, based on open source platform OpenFOAM, the wall boiling model under two fluid framework was established, and the validation was performed against the subcooled boiling experiment conducted by Bartolemei et al. After that, a typical structure with four parallel coolant channels were modeled, and two kinds of flow blockage conditions, including the thin-wall blockage and plate-bending blockage conditions, were performed to investigate the influence of blockage on the flow and heat transfer features. Simulation results show that the proportion of mass flow rate decreases significantly with the value of 7.82% and 10.06% and the outlet temperature increases correspondingly to the value of 326.69K and 325.81K in blocked coolant channels under the thin-wall blockage and plate-blending blockage conditions respectively. Subcooled boiling occurs at the end of blocked channel under the high power level, and the coolant returns to the subcooled state after mixing in the upper chamber. The thin-walled blockage causes sudden change of the channel shape, leading to the flow stagnation zone and weak zone accompanied by the local liquid temperature rise. [ABSTRACT FROM AUTHOR]

• Turbulent flow in 3 × 3 rod bundle with mixing vane was predicted by SEM method. • Meshing strategy of pure hexahedral mesh in complex geometry was introduced. • A simplified treatment of inlet velocity boundary was proposed and tested. • Predicting capability of LES and URANS models are validated with measured data. In this paper, the turbulent flow in rod bundles with spacer grid and mixing vanes in nuclear reactor core was studied using spectral element method with large eddy simulation (LES) and unsteady RANS (URANS) method. Combining with the radial periodical boundary conditions, the turbulent mixing phenomena in central subchannels were investigated, especially focusing on the flow field after the spacer grid and mixing vanes. Other than setting up long bare rod bundle to establish the fully developed turbulent flow, only a 2-cm-long bare rod bundle was extruded upstream the spacer grid and mixing vanes, while the fully developed laminar velocity profile was applied on the inlet surface for velocity boundary settings. The results indicate that both of the models could predict time-averaged velocity at 0.5Dh, 1Dh, 4Dh downstream mixing vanes properly, while the URANS model underestimate the root-mean-square (RMS) value of fluctuating velocity due to the inlet treatment effect. Furthermore, turbulent kinetic energy contour with transverse velocity vector was combined to investigate the behaviors of vortices formed by split-type mixing vanes, revealing the behavior of vortices for preliminary dynamic and mechanic analysis in design process of fuel assembly. [ABSTRACT FROM AUTHOR]

• The full-scale model for UPTF facility is established. • The transient coolant mixing process in RPV is simulated and analyzed. • The CFD best practice guidenline for the coolant mixing in RPV is concluded. The emergency coolant cooling system (ECCS) is designed to inject coolant into reactor pressure vessel (RPV) during loss of coolant accidents, ensuring the sufficient coolant inventory in the nuclear reactor core. During this process, the cold water mixes with high temperature coolant in the primary loop, making the wall of RPV and cold legs suffer severe pressurized thermal shock (PTS) and threatens the integrity of primary loop. In this study, coolant mixing phenomena in the experiment UPTF, which is a 1/1 full-scale model of the PWR with four coolant loops in Germany, was studied numerically using the CFD software Fluent. The best practice guideline for coolant mixing study using CFD method in the reactor primary loop and RPV, including the suitable turbulent model, geometry simplification and mesh conditions, was concluded. The detailed three dimensional thermal hydraulic parameter distributions were achieved and the effect of coolant injection rates on the mixing process is also obtained. Results show that the mixing effect is enhanced due to the existence of reverse flow at the core entrance sections. The high injection rate provides more timely coolant supplement and reduces the possible core exposure time. However, as the injection rate increases, more intensive disturbance on the temperature and velocity fields appear, which is more likely to cause the thermal fatigue. [ABSTRACT FROM AUTHOR]

• The Phenix natural circulation test was analyzed by system program THACS. • The predicted results of THACS are in good agreement with the test results. • The effect of Inter-wrapper flow on decay heat removal of pool-type SFR was analyzed. • The analytical capability of THACS program for natural circulation of pool-type SFR was evaluated. For the pool type sodium-cooled fast reactor (SFR), whether the primary circuit can establish a stable natural cycle after the reactor accident, and whether the primary natural cycle can take away the decay heat of the reactor core in the long-term, are the key issues in the design and safety analysis of SFR. In order to verify the natural circulation capacity of the primary circuit, a natural circulation test was carried out on pool type SFR Phenix by CEA before the reactor is decommissioned. One objective of this test is to verify the system codes used to simulate sodium natural convection phenomena. In this paper, the system code THACS is used to analyze Phenix natural circulation test, and the capacity of THACS to analyze the natural circulation of SFR is evaluated. Firstly, this paper introduces the components of Phenix reactor and the process of natural circulation test. Then, the models of program and the computational modeling of Phenix reactor are described. The calculation involves the whole process of the test: dry out of steam generators, scram, stop of pumps, two phases of natural circulation. Then, the results are compared with the test data, and the influence of inter-wrapper flow (IWF) on core outlet temperature is emphasized. The results show good applicability of THACS to analyze natural circulation in SFR, and the IWF can reduce the peak value of core outlet temperature at the beginning of natural cycle. Finally, the deviation between calculation and test results are analyzed, and the limitation of the models are evaluated. [ABSTRACT FROM AUTHOR]

• A containment design is proposed for the I2S-LWR, including passive safety systems. • Assessment of the safety systems demonstrates indefinite passive cooling. • The impact of the ADS valves and accumulator on plant response is investigated. The Integral Inherently Safe Light Water Reactor (I2S-LWR) is a novel reactor design concept which aims at delivering an electric power output level comparable to that of large LWRs (approximately 1000 MWe), while at the same time achieving an overall level of safety that is enhanced with respect to large Generation III+ LWRs. One of the main safety goals is to achieve indefinite cooling following design basis accidents (DBAs) using atmosphere air as ultimate heat sink. In order to accommodate these goals, the I2S-LWR incorporates several innovative safety features, including an integral Reactor Pressure Vessel (RPV), enhanced passive Decay Heat Removal (DHR) systems and several containment passive cooling systems. The present work is focused on a passive and reliable containment design, which plays a significant role in LOCA scenario and is the last boundary to prevent the release of radioactivity to the environment. In this paper, several innovative passive systems located in the I2S-LWR containment are proposed, including Core Make-up Tank (CMT), Accumulator (ACC), Passive Suppression System (PSS), Automatic Depressurization System (ADS), Passive Containment Cooling System (PCCS) and Passive Reactor Cavity Cooling System (PRCCS). The best-estimate thermal-hydraulic code RELAP5 has been used to model the I2S-LWR RPV and containment passive systems. The inadvertent opening of PORV (Power Operated Relief Valve) accident scenario has been simulated in order to study the containment response, including the coupling between RPV and containment. The results show that, through the activation of the ADS, pressure equilibrium between containment and reactor pressure vessel is achieved, while maintaining the reactor core covered at all times. The containment passive cooling systems ensure that the containment pressure remains at acceptable levels throughout the transient. A sensitivity study on the number of operating ADS valves, and on the availability of accumulators is also performed. [ABSTRACT FROM AUTHOR]

• Code development and analysis of heat pipe cooled PRHRS of MSR are conducted. • Reactor model shows a good agreement with MSRE data and results. • Heat pipe cooled PRHRS model is validated with experimental data of 2.8% maximum error. The high efficient and reliable Passive Residual Heat Removal System (PRHRS) is of great importance to the development of Molten Salt Reactor (MSR), whether the liquid fuel MSR or solid fuel MSR. In this research, based on the structure and operation performance of the new conceptual PRHRS, using high temperature heat pipe as heat transportation, a set of reasonable mathematical and physical models have been built, mainly including reactor core model, high temperature heat pipe model and PRHRS model. The new conceptual Passive Residual heat removal system Analysis Code (PRAC) has been developed using FORTRAN language. The verification and validation for PRAC is conducted by using existed Molten Salt Reactor Experiment (MSRE) data provided by Oak Ridge National Laboratory and the PRHRS experimental data in our previous work. The results show that the developed PRAC could predict the MSR transients reasonably and show high fidelity. The numerical results of reactor core neutronic and thermal–hydraulic model show a good agreement when compared with MSRE pump startup and shutdown transients. The numerical results by PRHRS model agree well with experimental data which the maximum error of the fluoride salt temperature and heat pipe wall temperature are within 1.2% and 2.8%. This research provides a valuable software basis for the design and analysis of PRHRS of MSR. [ABSTRACT FROM AUTHOR]

• A code is developed for studying the operating thermal-hydraulic characteristics of the reactor core, which consists of core thermal-hydraulic model, neutron kinetics model and heat pipe startup model. • In order to verify the code, a series of experiments about potassium heat pipe are operated. • Steady-state and startup transient are performed by the code. • This research provides valuable experience for designing and formulating control strategies of the heat pipe cooled reactor. Heat pipe cooled reactor has been proposed for applying in the space station and underwater exploration featured with high reliability, low noise level and compact structure. A typical design of a heat pipe cooled reactor system is HP-STMCs designed by the University of New Mexico. It is conceptually designed as a fast reactor, lithium-filled heat pipes that transfer heat to potassium heat pipes radiator and thermoelectric generators. In this paper, a code is developed for studying the operating thermal–hydraulic characteristics of the reactor core, which consists of core thermal–hydraulic model, neutron kinetics model and heat pipe startup model. In order to verify the code, a series of experiments about potassium heat pipe are operated. Steady-state and startup transient are performed by the code. The highest temperatures of fuel and matrix in reactor core show at the 2nd channel, which reach 1720 K and 1545 K from the results of steady-state. As for the startup transient, the code for heat pipe cooled reactor core successfully predicts key parameters during startup transient, such as each layer material temperature of reactor core, a single heat pipe's temperature response and heat rejection power. The code has initially predicted the transient response of the heat pipe cooled reactor under power increasing rates of 800 W/s and130W/s. It is discussed that the startup strategy of power increasing rates is based on heat pipe transition temperature (916 K). This research provides valuable experience for designing and formulating control strategies of the heat pipe cooled reactor. [ABSTRACT FROM AUTHOR]

The lead-based reactor is one of the fourth-generation nuclear energy systems with excellent development prospects. It has attracted much attention because of its unique advantages in fuel transmutation and breeding, passive safety, and economic efficiency. In this study, the main features of typical lead-based fast reactors (LFRs) worldwide are reviewed and current challenges in their development are pointed out. Moreover, the research progress in four main thermal-hydraulic aspects is overviewed and discussed, including the flow and heat transfer characteristics of the coolant, thermal-hydraulic analysis of the reactor core, lead pool and reactor system. These aspects are further divided into specific issues according to the developing needs of LFR. For each issue, up-to-date research findings, existing difficulties, and future trends are presented. This study can serve as a reference for the design and thermal-hydraulic analysis of LFRs. • Research on the main thermal-hydraulic features of LFRs is discussed. • Flow and heat transfer characteristics of lead-based coolants are summarized and compared. • Up-to-date research and developments of Core/Pool/System Level thermal-hydraulics are reviewed. [ABSTRACT FROM AUTHOR]