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The start-up of a high temperature heat pipe is a complex multi-phase transition and flow heat transfer process, which may result in heat transfer oscillations, causing fluctuations in the heat pipe temperature and thus affecting the operational safety of the reactor. In this study, for the high temperature sodium heat pipe with a length of 1 000 mm and a diameter of 20 mm, the evolution law of the heat transfer oscillation was summarized through start-up experiments under vertical and horizontal conditions. The finite element thermal-mechanical simulation and thermal fatigue program analysis of a localized heat pipe reactor core design with generalized characteristics were carried out, and the stress response and the consequences of fatigue damage induced by heat pipe heat transfer oscillations were investigated. The experimental results show that the change of sodium evaporation rate causes the change of heat pipe pressure under vertical condition, which affects the evaporation and reflux balance of working medium, and causes the oscillation of heat transfer in different characteristics during start-up process. Enhancing the input power will shorten the frequency and decrease the amplitude of the oscillation. When the heat flow density increases from 9 973.71 W/m2 to 12 452.28 W/m2, the temperature oscillation amplitude decreases by nearly 30% and the oscillation period is shortened by about 40%. In the heat transfer oscillation state, the heat pipe will still maintain the heat transfer oscillation when the input power is suddenly reduced, but the waveform changes from a sawtooth shape to a trapezoid shape. The oscillation is obviously suppressed under horizontal condition. Simulation results show that, when the high temperature heat pipe heat transfer oscillation occurs, the thermal fatigue failure risk area of the localized reactor core is the inner wall of the evaporation section of the heat pipe. Fatigue damage is the largest in same-phase oscillation. The average value of the stress increases when the peak temperature difference increases due to the phase difference of oscillation, but because the decrease in the amplitude of stress of different-phase condition, the fatigue damage is reduced compared to the same-phase condition. Considering the effect of the conservatism deviation and creep on the fatigue damage, the thermal fatigue analysis is based on the safety factor K. Fatigue safety analysis results show that when the safety factor K is lower than 1.4 there is almost no risk of fatigue failure. K reaches 1.4 or more fatigue damage is gradually significant. When K=1.6 the fatigue life of the pipe wall material reduces to 3.29 years. This study is of great significance in analyzing the mechanism of heat transfer oscillation of high temperature sodium heat pipe and improving the reliability assessment of heat pipe.

BackgroundHeat pipe cooled reactor (HPR) has many characteristics, such as reliability, inherent safety, small volume, modularity, and solid core. The nuclear fuel of solid core is seriously affected by high temperature, strong irradiation, and solid constraint when operating, which affect the heat transfer performance and mechanical properties of the core seriously. The stress and gap heat transfer caused by the contact between monolith and other components change nonlinearly with the increase of burnup, and they influence each other. Therefore, the coupled irradiation-thermal-mechanical behavior of the monolith is a complex multi-physics phenomena.PurposeThis study aims to develop a coupled irradiation-thermal-mechanical model to explore the characteristics of gap variation, heat transfer and mechanics during the lifetime of solid core.MethodsFirst of all, based on the geometric parameter and material of a typical solid core of HPR with fuel rod composed of UO2 pellets and 316 stainless steel cladding, a coupled irradiation-thermal-mechanical model was developed and applied to the finite element multi-physics field analysis software COMSOL. The calculation parameter settings mainly referred to the design parameters of the MegaPower reactor. Then, a thermal conductivity model changing with the increase of burnup for UO2, the gap heat transfer model and mechanical contact were introduced in the gaps in the solid core, and both irradiation-induced deformation effect including densification and fission product swelling, and creep effect of UO2 pellets and 316 stainless steel monolith were taken into account. Finally, the model was applied to calculating the typical HPR and the characteristics of gap variation, heat transfer and mechanics were analyzed.ResultsAnalysis results show that pellet temperature and creep of monolith and cladding increase after complete contact between monolith and cladding. A smaller average number of heat pipes around the fuel rod result in higher temperature and stress distribution in the nearby area, and the cladding in this area has a risk of creep failure during its lifetime caused by internal pressure of the fuel rod and contact pressure between the monolith and cladding.ConclusionsThe gap contact can affect the heat transfer and mechanical properties of the solid core of HPR, and even result in an increase in the risk of cladding failure.

The power system of heat pipe cooled nuclear reactor can be divided into dynamic conversion system and static conversion system. Thermoelectric generator (TEG) is a typical static conversion device, which can directly convert heat energy into electric energy. The combination of heat pipe cooled nuclear reactor and TEGs can bring many advantages, such as compact structure, high inherent safety, no auxiliary mechanical equipment and low noise. However, the maximum core temperature could exceed 1 000 K, and ordinary thermoelectric generators can not operate in such high temperature. To solve this problem, the use of segmented thermoelectric generator (STEG) in the power system of heat pipe cooled nuclear reactor is a method worthy of research. STEG can play an important role in the heat pipe cooled nuclear reactor. But there is not enough literature to study the application of STEG in heat pipe cooled nuclear reactor. STEG usually splices two or three thermoelectric materials with the highest power generation efficiency in different temperature ranges. In this way, STEG can obtain a high thermoelectric conversion efficiency in a wide temperature range (usually above 600 K). Therefore, a kind of STEG for heat pipe cooled nuclear reactor was designed. In this paper, a conceptual design of a heat pipe cooled nuclear reactor with STEGs was proposed. The single STEG and single channel model were explored by COMSOL software. A three-dimensional finite element model of STEG’s performance in heat pipe cooled nuclear reactor was established. STEG was designed to compose of bismuth telluride and skutterudite, and its geometric structure and performances were optimized. Numerical simulation was carried out under steady-state conditions to determine the optimal STEG’s geometry. Then, the optimized STEG was connected with the heat pipe to form a single channel model for simulation to explore its performances. When the temperature difference reaches 550 K, the maximum STEGs’ output power reaches 0.374 W, and the maximum conversion efficiency is 15.75%. In single channel model, when the heating power is 300 W, the maximum STEGs’ output power reaches 47.29 W, and the maximum conversion efficiency is 15.63%. The simulation also show that the maximum stress of single STEG and single channel model are 270 MPa and 547.5 MPa, respectively. And the maximum stress appears on the copper conductor near the hot side. This paper could provide a preliminary basis for the numerical simulation of the combination of STEG and heat pipe cooled nuclear reactor. This paper can also be used as a reference for the thermoelectric conversion of heat pipe cooled nuclear reactors.

Safe and reliable energy supply is the key foundation for the development of unmanned underwater vehicle (UUV). Unlike land and space missions, underwater missions are more demanding on energy, which requires higher power level, longer operating time, faster response time, and better concealment. Heat pipe cooled reactor has the advantages of large power capacity, simple structure, easy control of reactivity, and rapid thermal response. The solid-state reactor core does not require coolant and relies entirely on heat pipes to dissipate heat, thereby providing good inherent safety, controllability and concealment. Therefore, heat pipe cooled reactor is considered as one of the most promising options for UUV energy supply. In this study, a conceptual design of Nuclear Silent Thermal-Electrical Reactor (NUSTER-100) was proposed to meet the energy requirement of China’s heavy ocean UUV research and development. 109 sodium heat pipes were arranged in the reactor core for passive cooling, and the thermoelectric generators (TEGs) were employed in the reactor to convert fission heat to electric power. A set of heat pipe cooled reactor system-wild mathematical and physical models were established, including core neutron physics model, core channel heat transfer model, heat pipe model, thermoelectric conversion model and cold junction heat transfer mode. Based on the efficient and robust numerical algorithm and modular modeling ideas, heat pipe advanced reactor transient analysis code HEART was developed with independent intellectual property rights, and the key modules of HEART were validated and verified by heat pipe experiment and thermoelectric power experiment. The steady-state, cold-start transient and reactivity insertion transient conditions of NUSTER-100 were calculated and analyzed by HEART, and the full power operation characteristics of NUSTER-100 were obtained. Steady-state performance of the NUSTER-100 indicates that the solid-state core has good temperature flattening ability. The surface temperature of the heat pipe is less than 1 300 K, and the average temperature drop of the TEG module in central channel is 724 K, which can produce an electrical power of 1 207.8 W. Based on the cold-start transient thermal-hydraulic analysis, a three-stage heat pipe cooled reactor start-up scheme with high safety was proposed, and the stability and safety of heat pipe reactor under reactivity insertion transient were evaluated. The results show that the three-stage heat pipe cooled reactor start-up scheme can improve the reactor start-up efficiency. This study can provide support for the development of UUV and heat pipe cooled reactor technology in China.

As a kind of passive equipment with high efficiency, reliability and long-distance heat transfer, heat pipe is widely used in the field of nuclear energy. The reactor with heat pipe as the main heat transfer component came into being, which is heat pipe reactor. The main difficulty in the research is the heat transfer of heat pipe. There are some heat transfer limits in the work of heat pipe, which hinder the heat transfer of heat pipe. At this stage, there are few studies on the heat transfer limit, and the occurrence law of many limits has not been mastered. The theoretical and experimental researches on the heat transfer limit mechanism of liquid metal high-temperature heat pipe are of great significance to improve the heat transfer performance of heat pipe. In this paper, the existing heat transfer limit of high-temperature heat pipe was briefly described, the experimental research was carried out on the heat transfer limit of capillary driven high-temperature heat pipe with working medium of sodium, and the experimental platform for heat transfer limit analysis and test of high-temperature heat pipe was designed and built. The heat transfer limit of capillary driven heat pipe with working medium of sodium and liquid filling of 158 g and 208 g was experimentally and theoretically studied. The horizontal inclination working condition was designed from 0° to 90° and the heating power was from 1 kW to 7 kW. In terms of experiment, an experimental platform for heat transfer limit test and analysis of high-temperature heat pipe was designed and built, and the changes of heat transfer power of liquid metal high-temperature heat pipe under different horizontal inclinations and different heating powers were studied. In theory, the correctness of the theoretical models of continuous flow limit and entrainment limit was verified, and the occurrence laws of the two limits were summarized. It is found that the occurrence of continuous flow limit of heat pipe affects the start-up stage of heat pipe. Because the transition temperature is higher when the horizontal included angle is large, the continuous flow limit is more likely to occur in the heat pipe at large angle. The prediction error of the empirical model is less than 6.58% at small angle and more than 28% at large angle. When the entrainment limit occurs, the temperature in the evaporation section of the heat pipe rises sharply and the temperature in the condensation section fluctuates. The greater the inclination angle of the heat pipe, the easier the entrainment limit occurs. The empirical model has errors at different angles, and the error at large angles exceeds 100%. This paper summarizes the occurrence laws of continuous flow limit and entrainment limit, which provides a reference for the design of heat pipe in advanced nuclear reactor system.

Atomization and sprays are widely used in industry and agriculture. An appropriate atomization simulation method is essential in analyzing the liquid film-breaking process and atomization performance, especially in large-scale atomization field calculations. This study innovatively proposes a portioned method that combines existing fundamental atomization calculation models to balance computational accuracy and speed, finally achieving a full-scale numerical study of large atomization fields. This study employs the volume of fluid (VOF) model to measure the two-phase flow in the inner flow field and applies the discrete particle model (DPM) to analyze droplet behavior in the far-atomization field. In the near-atomization field, the VOF-to-DPM method connects the nozzle with the jet space, providing an effective numerical simulation of the liquid film formation and droplet breakup processes. Additionally, experiments on atomization using a pressure-swirl nozzle at different flow rates were conducted. Experimental data, such as atomization cone angle, flow distribution, and droplet particle size distribution, were obtained, and numerical calculations were performed using the large atomization field partitioned calculation model. The simulation results are utilized to explain the mechanisms of liquid film disintegration, while the experimental results are employed to validate the accuracy of the numerical model. The comparison revealed that the calculated results of the partitioned simulation approach align well with the experimental data. The maximum error in flow characteristics is 9.53%, in atomization cone angle is 6.16%, and in flow distribution is 3.67%, and there is a good agreement in particle size distribution with a maximum error of 17.58% in Sauter mean diameter, validating the accuracy of the portioned calculation method for large atomization fields. [ABSTRACT FROM AUTHOR]