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In order to deal with the environmental pollution problem caused by excessive emission of greenhouse gases, hydrogen energy, a green low-carbon secondary energy, acts as an important role in the process of energy transformation and the realization of global carbon neutrality goals. With the development of nuclear reactors of Generation IV, there is a new way using the high-temperature residual heat of nuclear reactors to couple with advanced chemical reaction in producing hydrogen. This research based on a small Fluoride-Salt-cooled high-Temperature Advanced Reactor (FuSTAR), gives a hydrogen production method derived by biomass. Exergy analysis of this method by Aspen HYSYS shows that the material exergy consumption (including physical exergy consumption and chemical consumption) is the main part affecting exergy losses. The energy utilization of a closed-loop heating system is better than that of an open system that directly heats gasification agents. Specifically, When the temperature reaches 700 °C, the mass flow rate of CO 2 and Biomass is 1:15 and the reaction finally reaches equilibrium, the total exergy consumption of the system is 1.47 × 109 kJ/h and the mole fraction of H 2 is 49.9 % in theory. In addition, the change of temperature mainly affects physical exergy consumption while the change of mass flow rate mainly affects chemical exergy consumption. Furthermore, when the system is coupled with Brayton cycle power generation system, in which a small amount of SCO 2 is drawn out as a high-temperature gasification agent for biomass hydrogen production, the exergy consumption is barely increased. • The concept of a biomass and nuclear hydrogen production system is presented. • Supercritical carbon dioxide acts as a gasification agent and heat-transfer medium. • Detailed exergy analysis on thermal-hydraulic parameters of FuSTAR. • Exergy consumption barely increases when coupled with the Brayton power cycle. • A reference for the optimal design of comprehensive utilization of nuclear energy. [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]

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]

• 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]

• Hydrogen dissociation in NTP reactors was analyzed by the numerical method. • Thermophysical properties of H-H 2 were established and validated from 200 to 3000 K. • Degree of dissociation was small and x H,out was just 0.456% in the NRX-A6 reactor. • Power density was one of the most significant factors for hydrogen dissociation. Nuclear thermal propulsion utilizes the nuclear reactor rather than the combustion chamber to yield thermal energy. Propellant hydrogen could dissociate in the high-temperature reactor, which has an important effect on thermal hydraulic performance of the reactor. In this study, a one-dimensional steady-state analysis code has been developed for studying the behavior of hydrogen flowing through the high-temperature coolant channel. Thermal dissociation and real gas thermophysical property models of hydrogen were proposed and considered in the calculation models. It was found that the model validation deviations of thermophysical properties were within ± 5% in the range of 200 ∼ 3000 K and 0.01 ∼ 10.0 MPa. Developed models were reliable and accurate with validation. Thermal-hydraulic behaviors of hydrogen in NRX-A6 reactor channel were analyzed. When dissociation occurred, the variation of properties was larger than those without dissociation, which enhanced heat transfer. The degree of dissociation was small and x H, out was just 0.456% under the design condition. The power density was the most significant influence factor, especially under the high power density. x H, out was 3.2 times than that of design condition as the power density grew 20%. This study can provide an approach to study hydrogen dissociation phenomena in nuclear thermal propulsion reactors. [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]

A reactivity-initiated accident is a typical nuclear reactor accident during which the core fission rate and power increase unexpectedly, making the temperature of the fuel element soar, which causes a decreasing strength, large deformation, and even the failure of the element. To simulate the performance of the plate-type fuel during RIA, a neutronics–thermal hydraulics–mechanics coupling method was implanted in a fuel performance code BEEs-Plates. The coupling scheme was firstly verified and validated against the MTR benchmark. Then the RIA results with the consideration of the deformation effect were compared with the benchmark case, showing that the mechanics as well as the burnup effect have little impact on the neutronic–thermal hydraulics results when using the constant thermal properties. Lastly, the thermal-mechanical analysis was conducted for the fuel plate at different burnups (0% FIMA, 10% FIMA, 20% FIMA) under the RIA. The results show that the peak stress of the cladding increases with the burnup increases while it has an opposite trend for the fuel meat. Some special features such as over-shrinkage of the fuel meat have also been noticeable and analyzed. [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]

Among the numerous space nuclear power sources, high temperature gas-cooled reactor with closed Brayton cycle has the advantages of both high power and high energy conversion efficiency. An Open-grid MEgawatt Gas-cooled spAce nuclear reactor (OMEGA) is investigated. Furthermore, a system Transient Analysis code of scheduled Shutdown and emergency Shutdown (TASS) is developed, including core, turbine-generator-compressor, regenerator, condenser, heat pipe radiator and other modules. TASS was verified by comparing prediction value with designed value, and transient performances of system shutdown were simulated. Over the course of shutdown, emissivity of fuel cladding is closely related to cladding peak temperature. Firstly, there is no overlay coating on fuel cladding whose emissivity equals to 0.18, the peak temperature of fuel cladding during scheduled shutdown and emergency shutdown reaches melting point at 7.8 h and 3.6 h respectively. Secondly, there is overlay coating on fuel cladding and emissivity is increased to 0.80. Radiation heat transfer can be greatly enhanced, and peak temperature of fuel pellet and cladding do not exceed the melting point of relevant material, which reflects the importance of fuel cladding surface processing. This paper provides a valuable analysis support for the design of megawatt-class space gas-cooled reactor system. • Shutdown safety analysis of megawatt-class space gas-cooled reactor system is conducted. • Radiation heat transfer among fuel rod claddings is calculated by radiation heat transfer network model. • Emissivity of claddings is the most remarkable contributors at the shutdown process. [ABSTRACT FROM AUTHOR]

Plate fuel elements are commonly used in compact nuclear reactors due to the advantages of high fuel consumption. In this paper, the two-phase flow pattern in the coolant channel of plate fuel elements, i.e. the rectangular narrow channel was studied. Visualization flow pattern experiments of upward steam-water two-phase flow in narrow rectangular channels, 745 mm long and 67 mm wide with gap widths of 1.3 mm and 1.9 mm, were investigated. The experiments were performed in the conditions of pressure ranging from 1 to 2 MPa, mass flow rate ranging from 200 to 1500kg/(m2⋅s), inlet subcooling ranging from 20–100 °C and tilt angle ranging from 0 to 45°. Three flow patterns, consisting of bubble flow, churn flow and annular flow, were obtained in a wide range of thermal parameters. Flow patterns maps were plotted and the influences of pressure, mass flow rate, inlet subcooling, channel gap and tilt angle on the flow pattern transformation of narrow rectangular channel were studied. This study is of great significance for investigating the subsequent heat transfer process of two-phase flow in compact nuclear reactors. [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]

• Control rods have more worth in NTP reactor control. • The NTP reactor controlled by control rods needs appropriate three-dimensional power flattening. • Thermal stress safety deserves more attention. • Increasing the total inlet pressure is more direct and safer for improving performance. As a highly developmental potential propulsion technology, nuclear thermal propulsion (NTP) technology has characteristics of high power and high specific impulse. A large number of relevant studies for NTP reactor have been carried out, but most studies only focused on the overall reactor performance parameters, ignoring the core reactivity safety and local assembly behavior safety. In this study, a cross-scale neutronics, thermal hydraulic and mechanics coupling model of the core-assembly is established by using three-dimensional Monte Carlo code and commercial CFD program to analyze three aspects: reactor control response security characteristics, heat release homogeneity of reactor and thermal hydraulic parameters. By means of value comparison of control system, three-dimensional power flattening and sensitivity analysis of single variable thermal–hydraulic parameter, the thermal hydraulic temperature safety and mechanical thermal stress safety of assemblies are significantly improved under control safety and inherent safety, which proves the importance of the neutronics, thermal hydraulic and mechanics characteristic balance.This paper can provide a theoretical basis for study on the assembly multi-physics behavior and safety characteristics of NTP reactors. [ABSTRACT FROM AUTHOR]

P assive containment A ir-cooling S ystem (PAS) is an important passive system of the modular small nuclear reactor, ACP100. After a severe accident, the air in the annular gap between the steel shell and the concrete shell of the containment is heated, forming buoyancy-driven mixed convection which removes decay heat. Thus, the heat transfer of mixed convection in PAS is essential for the safety of nuclear reactor. However, there are very less related studies on this phenomenon. In the present study, experiments on the mixed convection of atmospheric air in a parallel-plate channel with one-side heating for PAS were carried out. The regions of Re , Gr and Bo range from 1.7 × 104 to 8.8 × 104, 1.0 × 108 to 1.0 × 1010 and 4 × 10−5 to 90, respectively. Furthermore, the effect of the inclination angle on heat transfer was analyzed for the dome which is an integral part of PAS. A heat transfer correlation was newly proposed, which can more accurately predict the heat transfer of PAS with an error within ±15%. In addition, the predicted values of heat transfer agree well with other experimental data of mixed convection in the rectangular channel. The present research puts forward a new model for the analysis of PAS heat transfer performance, and has the reference significance for the optimal design of PAS. [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 idea of multiphase flow is introduced into the MPS-DEM numerical model for fluid–solid coupling problem for the first time. • Reproduce the DKT phenomenon to validate the improved MPS code. • Simulate the slide of dry granular material and the underwater landslide problem and compare the numerical results with the experimental results and simulate the molten pool accumulation successful. Compared with the grid method, the moving particle semi-implicit method (MPS) has prominent advantages in capturing the phase interface and avoiding grid distortion at the interface. However, MPS method is not particularly effective in dealing with the fluid–solid coupling problems including the interaction between discrete solid elements. Therefore, the PArticle method software for Nuclear reactor thermal hyDraulic Analysis (PANDA) is developed based on the MPS method and coupled Discrete Element Method (DEM) by introducing the idea of multiphase flow into the MPS-DEM coupling method. In this paper, the model is validated by reappearing the phenomena of DKT(drafting, kissing, and tumbling), and the problems of the slide of dry granular material and the underwater landslide are simulated as well. The numerical results are in good agreement with the experimental results, and the relative error of the leading edge distance is less than 6.12%. Finally, numerical simulation is carried out by taking molten pool accumulation in sereve accident of nuclear reactor as an example, which further confirms the good application prospects of PANDA in engineering problems. [ABSTRACT FROM AUTHOR]

Heat pipe cooled reactor (HPR) features high power density, compact structure, long operation life and high reliability, which is an excellent candidate for small nuclear power source for deep space or underwater missions. In this paper, a conceptual Nuclear Silent Thermal-Electrical Reactor (NUSTER) is preliminarily designed, using 109 sodium heat pipes for passive cooling. The thermoelectric generators (TEGs) are employed in the reactor to convert fission heat to electric power. Based on the design, a HEat pipe cooled Advanced Reactor Transient analysis code (HEART) is developed, including point reactor kinetics model, multi-channel model, core heat transfer model, heat pipe model, thermoelectric generator model and coolant model. The models proposed in HEART are preliminarily verified against the design value and experiment data, and the maximum deviation is less than 12.3%. Steady-state performance of the NUSTER are successfully predicted by HEART, and results indicate that the solid-state core has good temperature flattening ability. The surface temperature of the heat pipe is less than 1300 K, and the average temperature drop of the TEG module in central channel is 724.1 K, which can produce an electrical power of 301.94 W. The models developed in HEART can also be applied to other heat pipe cooled reactors, providing valuable experience in design and thermal-hydraulic evaluation of HPRs. • A nuclear silent thermal-electrical reactor (NUSTER) is conceptually designed. • A specific code HEART for the thermal-electrical analysis of heat pipe cooled nuclear reactor is developed. • Code V&V and the steady-state analysis of reactor core-heat pipe-thermoelectric generator system are carried out. [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]

• CFD model for OLE based on Eulerian two-fluid methodology was developed. • CFD model for OLE was validated with experimental results. • The effects of different water inlet mass flow rates and single inlet on OLE were analyzed. • Flow characteristics during liquid entrainment were also studied. The consequences of liquid entrainment through the fourth stage Automatic Depressurization System (ADS-4) of Advanced Passive nuclear power reactor (AP1000) are decreased reactor core coolant inventory and increased resistance to quick depressurization of the reactor primary coolant system. Therefore, the accurate knowledge of the Onset of Liquid Entrainment (OLE) is critical to nuclear reactor safety (NRS) assessment. In this study, three-dimensional computational fluid dynamics (CFD) simulations of OLE were carried out using a two-fluid Eulerian model of ANSYS Fluent CFD code along with the renormalization group (RNG) κ - ε turbulence model for each phase. In OLE phenomena, the interfacial drag has a significant effect but available choices of interfacial drag coefficient in Fluent were not suitable due to the presence of droplet clusters in entrainment. Therefore, a modified interfacial drag coefficient accounting for droplet clusters was used. The transient simulations were performed with this modified CFD model and validated against ADS-4 Depressurization and Entrainment TEst Loop (ADETEL) data. The comparison between CFD calculations and experimental data shows a good agreement. Furthermore, the effects of the liquid mass flow rates and single inlet on OLE were also investigated with the validated CFD model. To improve the understanding of OLE phenomena, the liquid volume fraction and gas velocity field distributions were also studied. [ABSTRACT FROM AUTHOR]

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

• The heat transfer behavior of high temperature heat pipe is experimentally tested. • The effects of heating power and inclination angle are studied. • Overheating at low filling rate and temperature fluctuation at high filling rate are discovered. • Several theoretical models of heat transfer limits are validated with experimental data. • Dryout phenomenon in heat pipe is divided into local dryout and integral dryout. Heat pipe cooled reactor (HPR) is considered as an excellent candidate for space power systems. As a core cooling method, the operational characteristics of high-temperature heat pipes need to be researched. In this paper, the heat transfer performance of potassium heat pipes (Φ30mm × 800 mm) is studied experimentally. The thermal behavior of heat pipe is tested under various conditions, including heating power (0.2–4 kW), air/water cooling method, inclination angle (0–90°), and filling ratio (20% and 100% of void in wick). And the characteristics of heat transfer limits are analyzed, consisting of dryout, entrainment, and sonic limit. The results show heating power is beneficial to heat transfer within the capillary limit. However, overheating occurs when heating power exceeds the capillary limit. The inclination angle has a strong effect on the thermal efficiency of heat pipe for the competition between gravity and entrainment, and the angle of 30° corresponds to the best thermal performance. Sonic limit and entrainment limit are estimated with a relative error of 38.7% and 17.9%. Dryout is divided into local dryout (<200 °C/m) and integral dryout (>200 °C/m). It is concluded that the high filling ratio (100%) and the small inclination angle (30°) are helpful to realize the optimal thermal performance. This work makes it possible to validate the operation of a high-temperature heat pipe and provides a reference for the design and application of HPR. [ABSTRACT FROM AUTHOR]

• A numerical code is carried out based on the three-stage model of frozen startup of the heat pipe. • The heat transfer performance of a NaK heat pipe is simulated during startup from the frozen state. • Sensitivity analysis and heat transfer limitations of the NaK heat pipe are carried out. Heat pipe cooled reactor (HPR) has a good adaptability to portable power system, which is popular in recent years. Due to the particularity of alkali-metal working fluid, the deep understanding of startup characteristics of high temperature heat pipe from frozen state is essential for the development of HPRs. In this paper, a three-stage frozen startup model is developed to describe the thermal behavior of heat pipe during the startup process, and the continuum flow in the vapor space is modeled as a one-dimensional compressible flow. A numerical code is carried out, in which the governing equations are discretized by Finite Element Method (FEM), and then the code is used to simulate the startup performance of a NaK heat pipe in HPRs. Numerical results indicate that the heat pipe startup behavior can be well described by the three-stage model. The NaK heat pipe is successfully started with a final consistency temperature of 834 K, although the entire second stage is restricted by the sonic limitation. The startup lasts 1550 s in total, and enters the second and third stage at 230 s and 650 s, respectively. After 1500 s, the maximum Mach number of vapor flow is lower than 0.1, which verifies the rationality of the one-dimensional compressible flow model of vapor. This work could provide a reference for the design and application of HPRs. [ABSTRACT FROM AUTHOR]