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Abstract: In the present study, two-dimensional numerical simulation of single bubble dynamics during nucleate flow boiling has been performed using moving particle semi-implicit (MPS) method. A set of moving particles was used to represent the liquid phase. The bubble–liquid interface was set to be a free surface boundary which can be captured according to the motion and location of interfacial particles. The interfacial heat transfer rate was determined by the energy variety of interfacial particles. The bulk liquid velocities investigated ranged from 0.07 to 0.3m/s. The surface orientations varied from vertical to horizontal through 60°, 45° and 30°. Bulk liquid subcooling varied from 0 to 6.5°C and wall superheat from 2.0 to 20.0°C. The computational results show that the bulk liquid velocity and surface orientation influenced the bubble diameter and liftoff time. Bubble would slide along the heater surface before lifting off and the sliding velocity at liftoff increased with an increase in bulk liquid velocity. Bubble dynamic was related to bulk liquid subcooling as well as wall superheat. The numerical results have been compared with the experimental data. [ABSTRACT FROM AUTHOR]

Highlights • Heat transfer experiments for super-low finned-tube bundles in MSRs are performed. • Tests are carried out under the operating conditions of the nuclear power plant. • A more accurate heat transfer correlation under wider Re number range is proposed. Abstract Experimental investigation on the heat transfer characteristics of superheat vapor through a new designed moisture separator reheater with "super-low" finned-tube bundles for Generation III+ nuclear power plant, was carried out based on a 5*30 tube bundle test section. Impacts of the vapor mass flow rates are analyzed on the heat transfer characteristics of superheat vapor through the finned-tube bundle. A group of experiments are reviewed on single-phase fluid flowing across the triangular super-low finned-tube bundles whose fin heights are less than 2.75 mm and ratios of fin height to fin diameter are from 0.05 to 0.15. Combining the presented experimental data with the other open published data for single-phase turbulent gases flowing across triangular super-low finned-tube, a new heat transfer correlation is proposed for the MSRs super-low finned-tube bundles, which shows better accuracy and has wider Reynolds number range in application. [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]

• A new method of utilizing the residual heat in reactor using Heat Pipe - Thermoelectric Generator is proposed. • Based on the numerical simulation results, the establishment of the principle prototype is improved. • The principle prototype of HPTEG assembled in reactor was built, and its thermoelectric characteristics were studied. • The thermoelectric performance of the prototype and the influence of temperature on the generating efficiency are obtained. This paper proposes a waste heat utilization method for new type reactor by Heat Pipe - Thermoelectric Generator (HP-TEG) to utilize the waste heat of lower divertor in fusion device and stirling generator in heat pipe space reactor. The feasibility of the HPTEG waste heat utilization device is verified by its principle prototype. Firstly, the numerical simulations of the single heat pipe and principle prototype were carried out to provide a numerical optimization scheme for the design of the principle prototype. The simulation results show that the heat transfer efficiency of the copper–water heat pipe is about 93%, while the overall heat transfer efficiency of the principle prototype is about 85%. Based on the simulation results, the thermal stress concentration areas were found, mainly concentrated in the thermal insulation short of the heat pipe. During the construction of the principle prototype, stainless steel sleeves were installed outside the insulation section of the heat pipe to limit the thermal expansion caused by the thermal stress concentration. Besides the principle prototype of HPTEG assembled in new type reactor was built, and its thermoelectric characteristics were studied, the results verified the feasibility of using waste heat. The test results show that the principle prototype starts smoothly, and the heat pipe has suitable isothermal properties (△T < 15 ℃). Due to the heat pipe's high heat transfer efficiency, the heat pipe obtained from the experiment is 91%–96%. The principle prototype has good heat transfer efficiency, which is 81.94%–93.25%. The maximum total electric power output of TEG is 16.37 W, in which the corresponding current is 0.52A, and the power generation efficiency is up to 1.49%. This paper verifies the feasibility of HPTEG to utilize the waste heat of lower divertor in fusion device and stirling generator in heat pipe space reactor, which provides a reference for the utilization of reactor residual heat. [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 thermal conduction heat pipe model is proposed, and the fidelity of the heat pipe model is validated by the experiments. • The numerical simulation results of HPTEG prototype are in good agreement with the experimental results. • The heat pipe failure accident in HPR increases the possibility of failure of adjacent heat pipes. Heat pipe cooled reactor featured with compact structure, high reliability, and inherent safety could be applied for the energy system for multipurpose. Numerical simulation methods are widely used to simulate the normal operating conditions of heat pipe reactors and even for accident conditions. In this paper, the ANSYS Fluent is used to perform thermal calculations on the heat pipe thermoelectric generator (HPTEG) prototype for both normal and accident conditions which include heat pipe failure and loss of heat sink accident conditions. The results show that the numerical simulation results are in good agreement with the experimental results, which proves that the fidelity of heat pipe model. Further analyses show that the failure of heat pipe would cause an increase in the power and operating temperature of the adjacent heat pipes, which may increase the possibility of failure of adjacent heat pipes. The maximum core temperature of the HPTEG prototype under the loss of heat sink accident is 1600 K, which exceeds the copper's melting point temperature 1356 K. It is obvious that the loss of heat sink accident would cause the core temperature increase sharply, which is likely to cause damage to the core. This work provides reference to safety analysis and evaluation of the heat pipe cooled reactor. [ABSTRACT FROM AUTHOR]

• A set of experiments of LBE heat transfer in a circular tube were carried out. • CFD simulations of LBE heat transfer in a circular tube were carried out using different turbulent model and Pr-t model. • Evaluation of the applicability of the numerical models was carried out combing simulation results and experiment results. The understanding of flow and heat transfer characteristics of liquid lead–bismuth eutectic (LBE) is essential for nuclear applications. In this paper, a corresponding experimental facility was established and a set of experiments of heat transfer of lead–bismuth eutectic flow in a circular tube was carried out, and a new heat transfer correlation was obtained through the experimental data to study the flow and heat transfer characteristics and evaluate a suitable computational fluid dynamics (CFD) model for LBE flow. The correlation fits well to the experimental data and the deviation of the correlation is smaller than 15%. At the same time, a CFD program was employed to evaluated the applicability of various turbulent models (standard k- ∊ , standard k- ω , SST k- ω) and turbulent Prandtl number models for LBE flows in the circular tube. Among the turbulence models, SST k- ω model shows the best agreements, the relative deviation of the simulation result from the experimental data is −3.0%∼34.7%. Among the turbulent Prandtl number models, Cheng-Tak model shows the best agreement, the relative deviation is 0.18%∼24.7%. This work provides valuable experimental data and CFD simulation method for liquid metal flow and heat transfer analysis. [ABSTRACT FROM AUTHOR]

• Several kinds of flashing model are compared for flashing in micro-crack channel. • A superior guideline for flashing model chosen is developed for micro-crack. • Leakage prediction accuracy is improved for micro-crack compared with LBB code. • Critical two phase flow phenomenon is obtained for micro-crack channel. • A comprehensive CFD results is obtained for leaking coolant migration in SG. The pressure relief rate and loss rate of coolant, which affect largely safety of fission reactor systems, are determined by the flashing process inside cracked passages. Information about flashing phenomenon is of great significance for the leakage online monitoring system of SG (Steam Generator), which ensures the normal operation of SG and safety of the reactor system during early stage of tube rupture accidents. In this research, flashing flow inside a thin micro-crack in heat transfer tube of SG has been studied using ANSYS-FLUENT. The classic cavitation model and evaporation–condensation model, in combination with both the mixture two-phase flow model and Eulerian two-fluid model, are adopted to simulate flashing phenomenon. The real geometry and operation conditions of AP1000 nuclear system are adopted to reflect the reality leakage process in SG. Two types of micro-cracks including axial crack and circumferential crack, which both could happen in the reality operations, are considered. The CFD (Computational Fluid Dynamic) results gained from this research have been compared with test data, and good agreement is demonstrated. The results show that the evaporation–condensation model under Eulerian two-fluid model framework behaves superior to the classic cavitation models in simulating the flashing phenomenon after evaluating accuracy of leakage rate and rationality of thermal hydraulic parameters distribution. In addition, two-phase choking flow phenomenon is obtained by decreasing backpressure of cracked tubes. The simulation in this research reveals the flashing process in short and micro-crack and the results could be good reference for leakage prediction of SG under LBB (Leak Before Break) operations to improve operation performance of SG and safety of the whole nuclear power system. [ABSTRACT FROM AUTHOR]

• Experimental study on heat transfer limit of potassium heat pipe is conducted. • Effect of heating power and inclination angle on heat transfer performance is considered. • Viscosity, sonic, entrainment and condensation limits of heat pipe are analyzed. Heat pipe cooled reactor (HPR) is widely applied in various environment due to its remarkable advantages compared with other nuclear power systems, such as high reliability, low noise level and compact structure. As one of the essential components in HPR, the high temperature heat pipe has an important effect on HPR operation performance. In this paper, an experimental study was carried to investigate the thermal–hydraulic characteristics and heat transfer limit of potassium heat pipe. Under the steady-state conditions, heating power and inclination angle of heat pipe were taken as sensitivity parameters. By increasing heating power, the heat transfer capacity was enhanced. For inclination angle, an increase in inclination angle would enhance heat transfer for low filling ratio heat pipe, while it does not make obvious effect on high filling ratio heat pipe. As for heat transfer limit, by analyzing the axial wall temperature distribution and predicting the vapor flow pressure drop, the particular phenomena is classified to four heat transfer limits including viscosity, sonic, entrainment and condensation limit. When heat pipe is cooled by natural air convection, the accurate prediction of condensation limit under effect of inclination is performed by air convection heat transfer relationship. Because of the friction force influence on viscosity limit when operating at low temperature state, the temperature at inlet of condenser section is 11℃ lower than adjacent position. The same reason for sonic limit when operating at low temperature state, the temperature at outlet of evaporator section is 17 ℃ higher than adjacent position. The occurrence of local dryout which lead evaporator temperature increasing sharply to 632 ℃ is entrainment limit. The experimental results provide guidance for the engineering application of high temperature heat pipe. [ABSTRACT FROM AUTHOR]

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