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• 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]

• Flow characteristics of LBE under rolling conditions are investigated experimentally. • The time-averaged flow frictional resistance of LBE is increased under rolling conditions. • The transverse additional force is main reason for the increase of flow resistance. • The results may be useful for the safety analysis under moving conditions. The effects of rolling motion on the flow characteristics of lead-bismuth eutectic in a vertical circular tube are investigated experimentally. The results show that the pressure drop and flow rate of the forced LBE circulation fluctuate under rolling conditions. The effect of rolling motion on the flow characteristics decreases with the increase of the corresponding static Reynolds number. The rolling motion may increase the time-averaged frictional resistance coefficient of the LBE when flow rate is low. The critical value of the impact Reynolds number is about 125,000. As a result, the time-averaged flow rate decreases. The analysis shows that it is mainly the transverse additional force caused by the rolling motion increases the frictional resistance of the LBE. An empirical correlation is established to predict the frictional resistance coefficient under rolling conditions. This study may be useful in the design and safety analysis of LBE as a working fluid under moving conditions. [ABSTRACT FROM AUTHOR]

• 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]

• 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]

• The transient response of FuSTAR in five design basis accidents are analyzed with the transient simulation code FUSY. • The sensitivity analysis is performed for some key parameters that may affect the transient behavior. • FuSTAR has a great ability to survive in URIA and ULOF. • The maximum limit of external parameters causing ULOHS and UOC are obtained by sensitivity analysis. FuSTAR is a small, modular fluoride-salt-cooled high-temperature advanced reactor, proposed by Xi'an Jiaotong University (XJTU), aiming at fulfilling the requirements of Gen IV nuclear system and meanwhile ensuring the inherent safety and economics. In order to evaluate the safety performance of FuSTAR, five typical design basis accidents are selected for safety analysis with different initial events, which are unprotected reactivity insertion accident (URIA), unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS), unprotected overcooling (UOC), and the accident combing ULOF with ULOHS. The sensitivity analysis is also performed. FuSTAR is modeled by modified system analysis code and the simulations of the steady-state are verified by the results with CFD method. The results of transient accidents indicate that safety of FuSTAR has been guaranteed in URIA and ULOF. Meanwhile, the maximum limit of external parameters causing ULOHS and UOC are obtained by sensitivity analysis. In the ULOF and ULOHS, the peak of fuel temperature, cladding surface temperature and core outlet temperature of coolant are below their safety limits. Therefore, FuSTAR exhibits acceptable safety under design basis accidents. This study will provide supports in verifying the rationality of the design and the subsequent optimization of FuSTAR. [ABSTRACT FROM AUTHOR]

• Two coupled parallel frameworks for shell and tube heat exchangers are proposed. • Fully three-dimensional parallel computation on both sides of the heat exchanger is realized. • One-to-one method achieves 85% parallel efficiency in 14 cores computation. • One-to-many method improves computational efficiency by about 20 times over serial in 28 cores computation. Shell and tube heat exchanger is widely used as a common type of heat exchanger. When the size and the number of heat exchanger tubes are enormous, it is difficult to use the direct modeling method for numerical simulation of heat exchangers. The simplified modeling method based on porous media utilizing user-defined function (UDF) has been widely applied. However, when the three-dimensional (3D) modeling is required on the tube side, the parallelization problem still restricts the computational efficiency of numerical simulation. In this paper, a one-to-one parallel method was first developed to solve the problem of coupled parameter transfer between the primary and secondary sides in the parallel framework, which was tested to achieve a parallel efficiency of 85% under 14 cores. However, the method acquires complicated data transfer between nodes, large memory occupation, and the modeling of the tube side is different from the actual structure under certain circumstances. To solve the problems, a one-to-many parallel method is proposed, and the computational efficiency is tested to be enhanced by about 20 times under 28 cores compared with single core, with a parallel efficiency of 70%. The two methods proposed in this paper are of great significance to the parallel transformation of the future 3D full-size code for shell and tube heat exchangers development and the calculation efficiency improvement. [ABSTRACT FROM AUTHOR]

• A Hi-fi space thermionic nuclear reactor system model is established, and its dynamic performance is obtained. • A model predictive controller is designed for the space reactor system using discrete state-space model and Kalman filter. • Several transient scenarios were designed to evaluate the control performance of the model predictive controller applied to space reactors from various perspectives. • Stable and rapid control of the space reactor has been achieved. Space nuclear reactors possess the inherent benefits of independence from solar energy and the capability to withstand intricate external circumstances. Distinguishing themselves from terrestrial reactors, space nuclear reactors function within the realm of outer space, operating autonomously throughout their entire operational cycles. Consequently, the establishment of a dependable automatic control system assumes utmost significance. In this paper, a high-fidelity nonlinear space thermionic nuclear reactor model is established, encompassing neutron physics, thermal–hydraulic, and thermoelectric conversion system models. An automatic control system with a model predictive controller (MPC) is designed based on the system. A simplified state-space model is established based on system identification as the internal model for the MPC controller, demonstrating good agreement with the dynamic response of the nonlinear system model. The deployment of the MPC controller for system control simulations exhibits excellent control performance under various operating conditions. In the case of a nuclear power step response, the MPC controller achieves an overshoot of 0.07 % and a settling time of 0.65 s. For an electric power step response, the MPC controller yields an overshoot of 0.28 % and a settling time of 36.6 s. These control performances are significantly better than those achieved by the PID controller. In power tracking control, the MPC controller demonstrates virtually error-free tracking performance. When external reactivity disturbances are introduced, the MPC controller outperforms the PID controller in swiftly compensating for the disturbances, maintaining the system's operating state at the desired level. [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]

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

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]

• The visual entrainment experiments in the vertical tee branch are carried out, and the test data and videos are obtained. • The effects of branch Angle, air intake mode and working medium on entrainment are revealed. • The new OLE model and entrainment rate model are proposed, which can be used for nuclear reactor safety analysis. The tee branch structure is broadly used in the nuclear power systems, and liquid entrainment in the tee branch has been studied in depth. However, most of the existing research focuses on the vertical tee branch, while the study about the entrainment phenomenon in the inclined tee branch hasn't formed mature contents. This research aims to analyze the liquid entrainment in the inclined up tee branch. The liquid level measurement method of the Automatic Depressurization and Entrainment TEst Loop (ADETEL) was improved and supplementary experiments were conducted with single-end gas inlet of vertical tee branch. In addition, the test loop was modified to conduct visualization experiments on liquid entrainment experiments of the tee branch with a 45 degree angle. The experimental result reveals that the liquid entrainment phenomenon in the inclined tee branch is significantly different from that in the vertical tee branch. Based on the experimental phenomena and data, a new OLE (Onset of Liquid Entrainment) model is proposed which is well fitted with the existing data. Then, an entrainment rate model is proposed by nondimensionalizing the key parameters affecting the steady-state entrainment, with the error within ± 15% of the experimental data. [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 obtain the behavior of axial flow field evolution under the effect of intrinsic structure on fluid excitation, we focus on the fluid excitation of fuel rod with spacer grids in axial flow by three-dimensional numerical methods. • The effects of the clamping structures and mixing vanes are investigated separately, and their effects on the excitation force are analyzed by the simulation results. • The mechanism and causes of fluid excitation force generation are described. • The results of the excitation force and the frequency spectrum is analyzed. Grids, supporting fuel rods, consist of dimples and mixing vanes. These complex structures cause turbulent flow field with complex vortex. The vibration of fuel rod by the high-velocity coolant may lead to grid to rod fretting (GTRF). In order to obtain the behavior of the fretting wear, it is crucial to investigate the induction mechanism of fuel rod by the flow field evolvement. In this paper, the effects of two structures with different geometric were discussed by numerical simulation. The generation and evolution of the fluid forces are analyzed in detail. The results show that the fluid forces mainly originate from the pressure asymmetry. Moreover, the clamping structures play a dominant role in both excitation and fluctuation of the fluid forces. The study obtained the effect of intrinsic structure on fluid excitation, and it can provide a basis for the fluid excitation boundary of the fretting wear of fuel rods. [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]

This paper proposes an optimized design for the Heat Pipe ThermoElectric Generator (HPTEG) principle prototype based on the design of the NUSTER system. The numerical analysis platform has been established and verified. It is used to analyze the steady-state and transient (start-up and shutdown conditions) thermoelectric properties of the refined HPTEG prototype. The steady-state calculation shows that when the input power of the HPTEG prototype is 5500 W, the maximum temperature is about 987 K, and the thermoelectric conversion efficiency is about 10.64%. The startup calculation show that the heat pipe startup is completed at about 14400 s. The shutdown calculation shows that at 6% shutdown power, the system tends to be stable after about 20000 s, and the maximum temperature is about 716 K. This study provides data support for the subsequent design modification of the future HPTEG principle prototype and provides ideas for the 3D transient calculations of the solid-state heat pipe reactors. [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 In-Vessel Retention analysis of the lower head under External Reactor Vessel Cooling is carried out under different stratified molten pool configurations with the annular fuel core as the object. • The effects of power density, light metal layer internal heat source, light metal layer liquidus temperature and light metal layer emissivity on the maximum wall critical heat flux ratio are analyzed. • The change of the state parameters of the light metal layer has little effect on the heat transfer of heavy metal layer and oxide layer. • The formation of heavy metal layer will threaten the integrity of bottom of lower head. • The core power density of annular fuel can be increased to about 110%. In the process of reactor severe accidents, maintaining the integrity of reactor pressure vessel to achieve IVR (In-Vessel Retention) of molten corium plays an important role in accidents mitigation and suppression. The ERVC (External Reactor Vessel Cooling) is an important technology to realize IVR. As a new economical annular fuel, the increase of zirconium content and power density in the core may threaten the integrity of lower head under severe accidents. In this paper, the IVRASA (In-Vessel Retention Analysis code in Severe Accident) is used to analyze the heat transfer characteristics of lower head molten pool corresponding to annular fuel core. The components of two-layer and three-layer configurations of molten pool are determined based on the geometric structures of annular fuel and verified heavy metal layer model. In terms of the integrity of lower head, the power density of the annular fuel core can be increased to about 110%. [ABSTRACT FROM AUTHOR]

• ASPTF was constructed to observe the flow instability in ASP for generation III 1000 MWe PWR. • The characteristics of flow instability in ASP were obtained based on experimental results. • The mechanism of flow instability in ASP was established by means of theoretical analysis. • The optimization and improvement measures of HE were summarized. • The suggestions for the design and operation of ASP were put forward. To cope with station blackout (SBO) accident in the pressurized water reactor (PWR) nuclear power plant (NPP), the advanced secondary passive residual heat removal system (ASP) was developed. A newly advanced secondary passive heat removal system test facility (ASPTF) for generation III NPP was designed and constructed to obtain the thermal hydraulic characteristics of ASP. The flow instability in ASP was observed in ASPTF, which was more likely occurred under low heating power and system pressure conditions. The effects of different influencing factors on flow instability in ASP were obtained. Then, the theoretical analysis of flow instability in ASP was conducted. Based on the experimental and theoretical analysis results, the optimization and improvement measures of ASP were summarized. [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]

• Turbulent natural convection of air between the open-ended parallel plates with one-side heating was investigated experimentally. • The influences on heat transfer and flow rate were analyzed. • A heat transfer correlation and a mass flow rate correlation considering the influence of inclination angle were proposed. • The boundary between natural convection and mixed convection was proposed in theory. In the present study, experiments on the turbulent natural convection of air for PAS were carried out. The heat transfer and flow rate induced by buoyancy were investigated. When the heated wall is inclined upward, the heat transfer coefficient increases with the decrease of inclination angle. And the pressure loss along the channel decreases because of the decrease of the velocity gradient. Based on the experimental data, the heat transfer correlation and mass flow rate correlation considering the influence of inclination angle were proposed. And they are applicable to predict the heat transfer and flow rate for PAS. In addition, a criterion for classifying the mixed and natural convection was deduced. The research results can be applied in the analysis of the thermal-hydraulics of the PAS to improve the security of nuclear power plants. [ABSTRACT FROM AUTHOR]

• The direct flow NTP reactors have poor heat exchange capacity. • The direct flow NTP reactors have strict requirements for power flattening. • The baffled reactor can enhance thermal safety. • Redundant and accurate control system design is very important. Nuclear thermal propulsion (NTP) technology has characteristics of high power and high specific impulse. In particular, CERMET fuel reactors have received a lot of research due to simple form and structure. A general CERMET fuel reactor model is established by using three-dimensional Monte Carlo code and commercial CFD program. After nuclear thermal coupling optimization, three partition schemes of baffled reactor were proposed to reduce the temperature difference between the wall and the coolant and lower the relative spike factor in the inner region. A feasible design of redundant control system was then proposed, with a cold clean excess reactivity greater than $4.59 and a minimum dry shutdown reactivity lower than -$7.94, where there was a high control accuracy of 6.7 pcm/mm. The reactor finally realized safety of the core temperature lower than 3000 K under the conditions of total exit temperature of 2650 K and thrust of 25000 lb. [ABSTRACT FROM AUTHOR]

• The flow field of a 2 × 2 subchannel is simulated using ANSYS FLUENT. • Fluid force on a fuel rod is analyzed on the time and frequency domain. • Mechanism of the generation of fluid force on a rod is discussed. Fluid excitation force is the driving source of grid-to-rod fretting (GTRF) wear, which is an important mechanism of nuclear fuel failure and breakage. In the present work, the fluid excitation force on a nuclear fuel rod was calculated and analyzed by the computational fluid dynamics (CFD) method, including time domain and frequency domain analysis. The cases of different turbulence models, inlet velocity, mixing vanes, and clamping structures were simulated and compared to study the mechanism of the generation of fluid force. The result shows that the SST k-ω model with a resolved near wall region mesh can capture the fluctuation of fluid excitation force, that the necessary condition for the generation of fluid force is enough inlet velocity and mixing vanes or clamping structures equipped, and that the necessary condition for the fluctuation of fluid force is enough inlet velocity and clamping structures. [ABSTRACT FROM AUTHOR]

• A stability criterion for spurious pressure oscillation is proposed through theoretical analysis and simulation research. • Incompressibility compensation term has great contribution on the spurious pressure oscillation. • The influence factors of pressure oscillation are carried out by hydrostatic analyses and dynamic pressure analyses. The Moving Particle Semi-implicit (MPS) method has been widely used in nuclear engineering due to its unique advantages in capturing moving interfaces. However, the unphysical numerical oscillation in MPS method is still a hotspot and difficulty in current study, and also a bottleneck hindering its high-accuracy numerical simulation of nuclear engineering. To obtain more accurate and stable numerical simulation, the spurious pressure oscillation behavior is studied theoretically and numerically. Through theoretical analysis, it is found that the incompressible compensation term is the main cause of spurious pressure oscillation. Besides, three mainstream forms of PPE are analyzed in detail. The simulations are carried out to study the effects of corrective matrix, time step, external force, numerical resolution and velocity on pressure oscillation. Finally, the stability criterion of spurious oscillation phenomenon is newly proposed to determine the time step, and reliable simulation results are obtained as well. [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 uncertainty of pressure load caused by ex-vessel steam explosion is assessed by probability analysis. • The probability distributions of pressure peak and pressure impulse are obtained. • The initial parameters that have great influence on steam explosion are clarified. In nuclear reactor severe accidents, the interaction of molten core melt and coolant may lead to steam explosion, which seriously threatens the safety of the reactor. The four initial parameters that affect the steam explosion were randomly sampled and calculated multiple times by steam explosion analysis code SEINA, and the probability distributions of steam explosion pressure peak and pressure impulse under uncertain initial conditions are obtained. The results show that the peak pressure of ex-vessel steam explosion is roughly normally distributed, and the pressure peak of 90% of the cases under submerged and non-submerged conditions within the range of 10–60 MPa and 20-80Mpa. respectively. By calculating the four key correlation coefficients between input parameters and output parameters, it is concluded that the influence of melt temperature and breach size on output parameters is greater than that of system pressure and coolant temperature. [ABSTRACT FROM AUTHOR]

• The problems of fluid–solid interaction and solid collision are Numerically investigated based on MPS-DEM method. • The movement characteristics of the debris particles during the non-heat transfer process of debris bed formation are studied. • The cake-like debris bed is observed in the 3D simulation of the QUEOS-Q21test. The core may melt down and a debris bed with specific shape may be formed on the lower head of Reactor Pressure Vessel (RPV). The solid-to-solid collision has not been concerned thoroughly in the previous studies. In order to solve the problems of fluid–solid interaction and solid collision, the MPS-DEM code is developed based on Moving Particle Semi-implicit (MPS) method and Discrete Element Method (DEM). In this paper, the MPS models, DEM models are introduced, and the coupling method is validated by particle flow and collapse tests. Besides, the two-dimensional (2D) and three-dimensional (3D) simulations of the QUEOS-Q21 test are performed. The results of numerical simulation are in accordance with that of experiments, which indicates that the MPS-DEM method has the potential to be applied to study the movement characteristics of the debris particles during the debris bed formation without heat transfer, and has application prospect in the severe accident analysis. [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 transient performance analysis code of high-temperature heat pipe with consideration of NCG effect is developed. • The heat transfer performance of a lithium heat pipe with NCG is simulated during startup from the frozen state. • Key parameters sensitivity analysis of the lithium heat pipe with NCG are carried out. The high-temperature lithium heat pipe utilized in the reactor will be irradiated by neutrons, causing non-condensable gas (NCG) to generate within and deteriorate thermal performance. This study develops a high-temperature heat pipe transient analysis code that takes into account the NCG effect. Heat pipe frozen start-up and key parameters sensitivity study are numerically evaluated. The NCG influence can be well described by the code. When compared to heat pipes without NCG, the start-up time of heat pipes with 25% NCG volume fraction increases by 54.2%. However, the existence of NCG can make the start-up process go more smoothly. The temperature drop in the NCG area changes from 40 K to 395 K as the NCG volume fraction increases from 5% to 25%. Increased input power can lower the volume fraction of NCG and hence diminish the negative thermal effects. This work provides foundation for transient analysis of high temperature heat pipe containing NCG. [ABSTRACT FROM AUTHOR]

• The evaporation of thin liquid sodium film inside high-temperature sodium heat pipe is first investigated by the molecular dynamics method. The Ar is added into the system to study the NCG's (Non-Condensable Gas) evaporation. • Equilibrium simulation and the non-equilibrium simulation are achieved. The MAC (Mass Accommodation Coefficient), the net mass flux, the heat transfer coefficient at the liquid–gas interface are acquired. The NCG's effects on evaporation are discussed. • This work can provide a reference for the optimal design and numerical simulation of high-temperature the sodium heat pipe. Understanding the evaporation of the thin liquid sodium film is important for deeply studying the heat transfer inside sodium heat pipe. In the present work, molecular dynamics is employed to investigate the evaporation of the thin liquid sodium film. The simulation system is a cuboid and consists of an upper gold wall and a bottom one. The sodium fluid exists between the walls. For studying the Non-Condensable Gas's (NCG) effect on the evaporation, the Ar is added into the evaporation system. Based on the startup and normal operation of the sodium heat pipe, eight different cases are studied. The equilibrium simulation is achieved by setting the same temperature of the solid walls. The Mass Accommodation Coefficients (MACs) of eight cases are acquired. By setting the walls at different temperatures, the non-equilibrium is built. The evolution of the evaporating liquid film is observed. The net evaporation flux (J net), the heat transfer coefficient (h) at the liquid–gas interface are obtained. It is found that the NCG's effects on the J net are achieved through its influence on the MAC. When the variation of the MAC is above 27%, NCG's effects on the J net are more obvious. In 600–700 K, 840–1020 K, the NCG suppresses the evaporation heat transfer and the maximum decrement is 150–200 kW∙m−2∙K−1. Whereas in 690–810 K, the NCG has an enhancement effect and the growing magnitude is 15–40 kW∙m−2∙K−1. This study could provide the mechanism supplement and extension for the macroscopic investigation of the sodium heat pipe. [ABSTRACT FROM AUTHOR]

• By combining the numerical simulation and multi-objective evolutionary algorithm, a co-simulation method is developed to the multi-objective optimization design of plate-type fuel. • The obtainment of Pareto optimal solutions and Pareto front proves the ability of co-simulation in actual engineering design. • Through the qualitative and quantitative analyses of the Pareto fronts, the co-simulation with NSGA-II or CMOEA-MS has better property about the design and thermal–hydraulic performance of the fuel plate. • The co-simulation method could modify the multiphysics model or add objective functions and constraints flexibly to expand the optimization objectives. Plate-type fuel element has been implemented and studied widely because of its good heat transfer capability. For the actual design of fuel assembly, the design objectives usually to be in competition, and the constraints also need to be considered. Therefore, this is a challenging constrained multi-objective optimization problem. A co-simulation method has been developed in this study to optimize the design of fuel plate for the thermal–hydraulic performance. Moreover, due to the versatility of this co-simulation method, extra objectives and constraints could be added as needed. First, a two-dimensional thermal-fluid–solid multiphysics model of plate-type fuel assembly was established and solved by the finite element method. After that, the values of relevant physical variables were verified with the three-dimensional model. Finally, the numerical simulation of multiphysics model was incorporated with the multi-objective evolutionary algorithm (NSGA-II, adaptive NSGA-III and CMOEA-MS) which could handle constraints to optimize the thickness parameters of fuel plate. Pareto optimal solutions were obtained separately after the co-simulation with three algorithms, and the qualitative and quantitative analyses of the Pareto fronts showed that CMOEA-MS and NSGA-II had better performance in this study. This co-simulation could provide comprehensive and reliable design information according to the design objectives and requirements. [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]

• 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]

• An MCCI experimental apparatus with simulant melt metal and oxide melt was established. • The high melt temperature was measured by both thermocouples and optical pyrometer. • The forming process of anchored crust was cleared recorded and analyzed. • The thicknesses of stratified layers and ablation depth were accurately measured after the splitting of crucible. Under hypothetical Reactor Pressure Vessel (RPV) failure accidents in Light Water Reactor (LWR), Molten Corium-Concrete Interaction (MCCI) will cause erosion of cavity concrete, possibly resulting in containment failure due to basemat penetration and overpressure. The CINA (Corium-Concrete Interaction Apparatus) experiment was conducted to investigate the MCCI of metallic and oxidic corium with siliceous concrete. Simulant melt material metallic iron and oxidic alumina generated by exothermic thermite chemical reaction was used to investigate the two-dimensional erosion of a cylindrical crucible in CINA experiment. The crucible was fabricated from siliceous concrete with an inner diameter of 300 mm and a height of 500 mm containing reinforcement (rebars). Decay heat in the melt was simulated by subsequent sustained addition of thermite. During the experiment, the melt temperature was monitored by two thermocouples and a two-color optical endurance pyrometer. Besides, the phenomenon that appeared on the melt surface was recorded by the video camera placed above the crucible and the formation process of the crust anchoring was observed. After the experiment, the crucible was split in half by a wire saw to accurately measure the ablation depth. Axial-radial ablation depths were 25 mm and 13 mm, respectively. Current findings contribute to the further understanding of MCCI mechanisms and the optimization of MCCI mitigation strategies. [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]

• 3D thermo-mechanical modeling of JRR-3M plate-type fuel assembly under irradiation condition with BEEs-Plates. • 1D simulation of coolant channels, considering the flow distribution. • Verification and utilization of 3D solid domain-1D fluid domain coupling method. • Analysis of the effect of burnup and uneven power distribution on temperature, stress profile of the assembly, and the flow distribution. The plate-type fuel has an excellent thermal–mechanical behavior as a type of nuclear fuel. To simulate and analyze the behavior of plate-type fuel assembly under irradiation conditions, a fuel performance analysis code BEEs-Plates was developed based on MOOSE (Multiphysics Object-oriented Simulation Environment). The properties, mechanical behavior, and irradiation effects of dispersion fuel meat were simulated by using the equivalent property models, and a one-dimensional Petrov-Galerkin stability model was used to solve the governing equations of coolant. The coupling between the three-dimensional solid domain and the one-dimensional fluid domain was also realized. The JRR-3 research reactor fuel assembly model was then used to perform the coupling simulation, and the assembly's temperature, stress, strain distribution, and coolant temperature, the size of the channels, flow distribution were analyzed. The results under the boundary conditions and input parameters applied herein show that the burnup effect has limited effects on the thermohydraulic performance of the assembly, with the variation of maximum temperature below 1 K and the change of channel width below 3.9%. However, its mechanical behavior is greatly affected by burnup, with the maximum stress of 396.2 MPa at the meat-cladding interface. The uneven power distribution causes the assembly to bend to one side, but its effect on the coolant flow distribution is negligible. [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]

• 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]

• The models of the melt evolution and the main factors affecting the jet-breakup length were introduced. • The experimental and simulation research on the morphology of debris bed were extensively summarized. • The models that can be applied to the code and the future development direction were prospected. When a severe accident occurs in a nuclear reactor, fuel–coolant interaction (FCI) may occur and cause steam explosion. Energetic FCI or steam explosion will threaten the integrity of the containment and even cause radioactive leakage. Therefore, the research and evaluation of the morphology and development of the melt in the FCI process is very important. In recent decades, many scholars have conducted theoretical and experimental research on this phenomenon, and formed a large number of mathematical and physical models for the degradation, evolution and relocation of the melt. This study retrospects and summarizes the experimental and theoretical research on the FCI phenomenon, and describes the behavior of the melt in the entire evolution process from the melt jet to debris bed. In addition, on the basis of literature review, the models of each stage are classified according to the formation mechanisms and applicable conditions. Finally, in the present study, the future development direction is prospected on the ground of the previous research results. [ABSTRACT FROM AUTHOR]

• The system analysis code RELAP5 is modified for calculation of NTP systems. • The physical property models of hydrogen are implemented into the code. • The correlations of heat transfer coefficient and friction factor are implemented into the code. • The heat conduction between the inner and outer cores should be considered when simulate a two-pass NTP reactor core. Taking advantages of long lifetime and high power density, nuclear thermal propulsion (NTP) has become an advanced propulsion option in some special power fields. Due to the high heat release and dramatic density change, the flow instability and subsequent unbalance of flow and heat transfer would happen. The thermo-hydraulic calculation of working medium is an important part of the NTP system designing work, and the system analysis methods should be developed. In this study, the system analysis code RELAP5 was modified with the thermo-physical properties, flow and heat transfer correlations of the propellant hydrogen. The modified system code has been validated against the published experiment data. The calculation results show good agreement with the experimental results and the code ELM's calculation results. The modified code was applied to thermo-hydraulic calculation of hydrogen in the analysis of a typical NTP system. The results indicate that the heat conduction between the inner and outer cores should be considered in the calculation of a two-pass NTP reactor core. This study has provided a reference for developing an NTP system analysis code. [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]

• A system analysis code applicable to offshore floating nuclear plant was developed based on existing code by adding motion condition model. • The station blackout accident was calculated under stationary and motion conditions with PRHRS in primary and secondary circuit operating respectively to compare the performance of the two PRHRS. • The heat removal capability of PRHRS in secondary circuit is better than that in primary circuit with the same arrangement. • Secondary PRHR is more susceptible to motion conditions than primary PRHRS. In order to conduct performance analysis of passive residual heat removal system (PRHRS), a system analysis code applicable to offshore floating nuclear plant (OFNP) was developed based on existing code by adding motion condition model and the OFNP system with PRHRS in primary and secondary circuit was modeled in this study. The station blackout (SBO) accident was calculated under stationary and motion conditions with PRHRS in primary and secondary circuit operating respectively to compare the performance of the two PRHRS. The results show that the heat removal capability of PRHRS in secondary circuit is better than that in primary circuit with the same arrangement. The influence of heaving condition on the primary PRHRS is greater than the secondary PRHRS, while the secondary PRHRS is more susceptible to trimming and pitching conditions and heeling and rolling conditions have little influence on both of two PRHRS. [ABSTRACT FROM AUTHOR]

• The scaling analysis and design parameters of seventeen integral test facilities were discussed. • The scaling analysis methods and scaling choices were discussed and the suggestions were proposed. • The distribution of test matrix was presented based on statistic results. • The statistical analysis of system codes developed based on the integral test facilities was completed. For the design verification and licensing of newly pressurized water reactors (PWRs), many integral test facilities were conducted. From the above integral test facilities, the obtained experimental results were adopted in the assessment and development of safety analysis codes. In this study, scaling analysis and design parameters of integral test facilities were summarized from the perspective of scaling method, basic scaling choice, equipment design, test matrix and application. The scaling method is the guideline for establishing scaling ratios of integral test facilities and different scaling methods were proposed. With the basic scaling choices of system pressure, geometrical parameters, working medium and equipment material, the design parameters of equipment were discussed. The test matrix and application of different integral test facilities were compared. This work is expected to sum up experience and provide guidance for the design and application of newly integral test facilities. [ABSTRACT FROM AUTHOR]

In this study, a modified two-dimensional geometry model was employed comprised of metal cladding, a metal matrix, and dispersed fuel particles. And an analysis method was established to simulate the thermal–mechanical behavior of dispersed plate-type fuel under the Reactivity Insertion Accident (RIA). Key fuel properties including thermal performance, mechanical behaviors and irradiation effects were considered. All these models were coded in the ABAQUS subroutines and solved. The influences of heat transfer deterioration, fission gas and burnup were investigated. Significant parameters to describe the performance of the fuel element, such as temperature, von Mises stress and strain profile, were evaluated. The results indicate that the deterioration of the heat transfer in the flow channel and the deepening burnup increase stress and strain of the dispersed plate-type fuel. The fission gas causes the crack of the fuel particles. [ABSTRACT FROM AUTHOR]

• A 2-D multiphase fluid dynamics code ACENA has been preliminarily developed. • This code is intended to simulate multiphase flow during LFR postulated accidents. • The models adopted in this code has been validated against several experiments. • The prediction results of this code agree well with experimental results. To simulate the heavy liquid metal-gas (HLM-gas) two phase flow phenomena that may occur during postulated accidents of Lead-based fast reactors (LFRs), a 2D multiphase fluid dynamics code, ACENA, is preliminarily developed and validated in this study. A brief description of theoretical models, including governing equations, auxiliary models and material properties are listed in this study. The capability of simulating flow and heat transfer characteristics between gas and liquid phases has been examined by typical air-water bubble column experiment, LBE-N 2 two phase flow experiment and Lead-Ar heat transfer experiment, and the comparison results show that this code could attain acceptable accuracy in simulating flow and heat transfer characteristics of HLM-gas two phase flow with appropriate interfacial drag coefficient and heat transfer coefficient models. [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]

• 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]

• We studied the thermo-hydraulic performances of H 2 in 2 × 2 twisted rod bundles. • The SST k-ω model predicts the heat transfer performance in twisted rod bundles well. • The twisted rods can make the lateral mixing distinct and enhance heat transfer. • The J F factor of the twisted rods is 28.0% and 6.1% higher than non-twisted rods. Nuclear thermal propulsion (NTP) is one of the most promising approaches for deep space exploration and manned interstellar voyage. As the propellant and coolant of NTP reactors, hydrogen provides higher specific impulse. In this paper, the thermohydraulic performance of high temperature hydrogen flowing through the twisted rod bundles were investigated numerically by contrasting with the cruciform rod bundles and the circular rod bundles. The heat transfer models were validated by literature data and the relative errors are within 20%. Twisted rods could enhance lateral mixing and heat transfer since the twisted lobes and valleys could enhance radial heat transfer and produce a strong rotational flow near the walls disturbing boundary layer, which reduces hot spot factors and assures the thermal safety. The largest relative difference of the J F factor between the twisted and other non-twisted channels are 28.0% and 6.1% respectively in this model. [ABSTRACT FROM AUTHOR]