Search

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.

Against the backdrop of China's ambitious three-step nuclear energy development strategy, encompassing pressurized water reactors (PWR), fast reactors (FR), and fusion reactors (FNR), the purpose of this study is to meet the pivotal need for a specialized sub-channel analysis code tailored to the unique thermal-hydraulic characteristics of liquid metal fast reactors (LMFRs). Building upon the foundational SACOS sub-channel code, the approach involves the seamless integration of LMFR-specific models. These include the wire-wrapped model, turbulent crossflow model, and liquid metal convective heat exchange model. The utilization of advanced computational techniques, such as the SIMPLE algorithm and staggered grid methodology, ensures the completion of accurate sub-channel calculations, establishing SACOS-LMR as a robust code for thermal-hydraulic safety analysis in LMFRs. Validation of the SACOS-LMR code was conducted through a sodium transient experiment involving 37-pin bundles at the Karlsruhe Institute of Technology. The results not only demonstrate a commendable alignment between computed parameters (e.g. temperature distribution, pressure drop) and experimental values but also confirm the code's precision in transient analysis for LMFRs. Applying the validated SACOS-LMR code, an in-depth thermal-hydraulic safety analysis of the European Lead-cooled Fast Reactor (ALFRED) core was conducted. The calculated results are not only reasonable but also exhibit consistency with comparable codes, affirming SACOS-LMR's applicability for LMFR core design and thermal-hydraulic analysis. In conclusion, this research represents a significant step forward in the development of LMFR technology. SACOS-LMR, with its validated capability in both steady-state and transient analysis, stands as a sophisticated and reliable sub-channel analysis tool. It not only supports LMFR core design but also contributes to the broader global pursuit of sustainable and clean energy solutions in the nuclear energy landscape.

The nuclear reactor system is complex and the operating environment is harsh, resulting in the complex phenomena of multi-physics coupling. The multi-physics coupling codes developed in the early stage shows limitations on codes' scalability and generality. Therefore, it is of great significance to build a multi-physics coupling framework and conduct research on key technologies in coupling problems, which may accelerate the development process of autonomous multi-physics coupling platform in China. In this paper, the multi-dimensional and multi-physics coupling finite element analysis platform for nuclear reactor developed by XJTU-NuTHeL was introduced. The main work consisted of the development of thermal-hydraulic model, the research of fuel performance analysis technology and the establishment of multi-physics coupling framework. In terms of thermal-hydraulic calculation, XJTU-NuTHeL conducted a series of studies on pressurized water reactors and advanced reactors grounded in the advanced multi-physics coupling framework, and developed the nuclear reactor system safety analysis code, NUSAC. In addition, a subchannel analysis model tailored for liquid metal fast reactors was established, and the fully coupled subchannel transient analysis code, FLARE, was developed. NUSAC and FLARE were then verified against relevant codes and experimental data. In the realm of fuel performance analysis, considering the wide application of finite element method in solid mechanics and its versatile modeling capabilities, XJTU-NuTHeL developed a fuel performance analysis code, BEEs, based on finite element method. The code could not only conduct multi-physics coupling analysis for traditional rod fuels under steady and transient conditions, but also extends its applicability to accident tolerant fuels and other fuels with diverse geometric shapes. This paper focused on the study and analysis of coated particle dispersed fuel and plate type fuel. The multi-scale simulation results of coated particle dispersed fuels, as well as the thermomechanical and corrosion behavior of plate type fuels were shown. In the context of multi-physics coupling analysis, the efficiency and accuracy of different mesh grid mapping schemes were studied and a multi-physics coupling framework was established. An example of the framework was then presented, showcasing the integration of the fuel performance code BEEs, the Monte Carlo neutron physics code OpenMC, and reactor system safety analysis code NUSAC. The keys parameters of mechanics, thermal-hydraulic and neutronics were obtained and analyzed through the coupling different codes. The multi-dimensional and multi-physics coupling finite element analysis platform built in this paper can provide a strong support for the high-fidelity numerical simulation of nuclear reactor multi-scale and multi-physics coupling.

In recent years, heat pipe has been widely used as an efficient heat transfer equipment in fuel chemistry, electronic communication and so on. The main structure of a heat pipe mainly includes the vacuum tube containing alkali metal working fluid and the composite mesh wick on the inner wall of the tube. In the axial direction, heat pipes are generally divided into three parts:evaporation section, adiabatic section and condensation section. Heat is input into the heat pipe through the evaporation section and output through the condensation section. The adiabatic section only plays a role in connecting and mass transfer. The length of each area can be flexibly arranged. In the radial direction, heat pipes are generally divided into outer wall surface, wick and vapor area. Alkali metal high-temperature heat pipes show broad application prospects in the nuclear reactor cooling and other aspects because of their strong heat transfer capacity and inherent safety at high temperature. Because the working fluid of these heat pipes is solid at room temperature, and whose saturated vapor pressure relatively low, their frozen start-up has a complex phase change process of the working fluid, so it brings some difficulties to the study of the frozen start-up of the high-temperature heat pipes. For different application scenarios, it is necessary to study the transient and steady-state operating characteristics of high-temperature heat pipes, to provide support for the application of high-temperature heat pipes in a variety of applicable scenarios. To establish a method for predicting the transient start-up and steady-state operation characteristics of alkali metal high-temperature heat pipes, this study utilized the finite volume method (FVM) to establish the pipe wall heat conduction model, the wick flow heat transfer model, and the vapor zone model. A frozen start-up transient analysis program for alkali metal high-temperature heat pipes was developed and verified using the C programming language, with a maximum relative deviation of 9.8%. Transient start-up of a single horizontal sodium heat pipe was simulated, and sensitivity analysis was conducted. The steam zone of the heat pipe enters the continuous flow state completely 700 seconds after start-up, and after a total of 3 000 seconds, the heat pipe reaches steady-state operation. Under steady-state operation, the heat pipe exhibits good isothermal properties, with a stable axial temperature difference of 22.5 K on the outer wall and an internal pressure drop of approximately 47 Pa in the wick. Furthermore, the ambient temperature primarily impacts the time required for the heat pipe to reach steady-state, as well as the vapor pressure and velocity distribution under steady-state conditions. Meanwhile, the length of the adiabatic section of the heat pipe influences the time required for it to reach steady-state and has a significant effect on the pressure and velocity distribution of the wick.

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.