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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]

• Detailed heat transfer in blistered plates' coolant channels is modeled using CFD. • Multiple blisters can cause flow maldistribution beyond design threshold. • Single blister leads to a rather high fuel meat maximum temperature. • Multiple blisters' influence on fuel temperature is independent. Due to the special structure, the plate-type fuel element possesses a unique failure mode, i.e. blistering. Blistering is a kind of plastic deformation, induced by thermal stress and pressure of gaseous fission products at the interface of fuel meat and cladding. A blistering cladding surface will alter the flow channel shape and thus affect the flow and heat transfer characteristics in a fuel assembly. The influence is significant because the parallel channels are closed and narrow. In this study, three plates and channels were modeled to represent parallel channels in a fuel assembly. Two kinds of blisters were considered, namely round blister and pillow-like blister. Six cases with various blistering distribution were simulated to investigate its influence on pressure drop, mass flow distribution, and fuel temperature. The turbulent model and mesh condition were carefully chosen and validated. Detailed flow structures were visualized and their relations with flow and heat transfer characteristics were analyzed. Results show that multiple blisters can induce the flow distribution factor to exceed the design threshold value. A single blister, due to the large thermal resistance of gaseous fission products, can cause a relatively high maximum temperature in the fuel plate and lead to additional structure rupture. Multiple blisters' influence on fuel temperature is independent and similar with that of a single blister. Larger blisters lead to higher maximum fuel meat temperature and therefore are more dangerous. [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]

Micro Heat Pipe-cooled Reactor (HPR) power source is featured with lower noise level, higher power output, longer lifetime than conventional power sources. It could be applied for the energy system of space or underwater vehicles. The HPR power source is considered as an ideal candidate for the space and underwater reactor concept. In this paper, a 120 kWe lithium HPR power source applied for multiple use is neutronics designed. Uranium nitride fuel with 70% enrichment and lithium heat pipe are adopted in the reactor core. Tungsten and water are used as shields on both sides of the core. The reactivity is controlled by 6 control drums with B 4 C neutron absorbers. Monte Carlo code MCNP is used to obtain kinetics parameters, the power distribution, shield analysis, reactivity coefficient and core criticality safety. A code MCORE coupling MCNP and ORIGEN is used to analysis the fuel depletion characteristics of the designed reactor core. A 14-year once-through fuel cycle is adopted according to optimization analysis. Overall, the designed core parameters preliminary satisfy the safety requirements and the rector is neutronics safe. This work provides reference to the design and application of the micro nuclear power source. [ABSTRACT FROM AUTHOR]

After the Fukushima Dai-ichi nuclear power plant severe accident, the advanced facilities that could improve the nuclear safety performance are preferred by the public. This paper discusses the preliminary design and thermal-hydraulic analysis of a heat pipe cooled passive heat removal system for the transportable fluoride-salt-cooled high-temperature reactor of 20 MWth proposed by Massachusetts Institute of Technology in 2014. Under severe accident, the high temperature fuel and fluoride salt sustain at the reactor core, the heat pipes are inserted by gravity to remove the decay heat to the final heat sink of silo cooling system. To illustrate the feasibility of the designed passive heat removal system, the heat pipe startup performance is firstly numerically investigated by self-developed code HPTAC (Heat Pipe Transient Analysis Code). Four heat transfer limits are adopted as criteria for the success of heat pipe operation, 1) Viscous 2) Entrainment 3) Sonic 4) Capillary limits. The benchmark of heat pipe startup compared with experimental data is conducted as well as a sensitivity study for heat pipe inclination angle. Secondly, the transient performance of passive heat removal system based on reasonable assumptions of heat pipe model is analyzed. Results show that the heat pipe reaches the normal operation in 6 min, in good agreement with experimental data with the maximum discrepancy of 16%. Most of fuel decay heat can be removed in 5 h by heat pipe passive heat removal system and all the key temperature are kept below the allowed temperature limits. [ABSTRACT FROM AUTHOR]

Heat pipe radiator, featured with remarkable advantages in heat transfer efficiency and inherent safety with small specific mass and little weight, are widely adopted to the heat-rejection system for space nuclear power reactors. In this paper, physical and numerical models are developed to obtain the startup and transient behaviors of radiator unit with a potassium (K) heat pipe covered by fin under space environment. The heat transfer limit theory is adopted as criteria for heat pipe operation success. Numerical results indicated that according to the internal vapor flow regimes, the K heat pipe startup could be divided into three distinct stages. The K heat pipe started up from frozen state successful and rapidly until the expected operation state is reached. Among the different heat transfer limits, only the sonic limit due to choked flow restricts the K heat pipe during the second stage. Overall, for the heat-rejection system of space nuclear power reactor, the heat pipe radiator unit can effectively radiated waste heat to the space environment in 5 min and responses fast under transient conditions. [ABSTRACT FROM AUTHOR]

An alkali metal heat pipe radiator is proposed for the Russian TOPAZ-II space reactor power system to replace the original pumped loop radiator with three advantages. First, the single point failure problem of the original radiator can be avoided. Second, because only the NaK manifold needs to be armored, the specific mass would be reduced and the reliability and survivability can be improved. Third, the heat pipe has nearly isothermal characteristics and high heat transfer capabilities. In the present paper, the high temperature heat pipes using potassium as working fluid, wire screen mesh as wick layer, made of stainless steel are adopted to remove waste heat by radiation. An integral carbon-carbon fin covering and connecting heat pipes as whole heat transfer radiator is selected for its higher thermal conductivity and lower self-weight. A detailed steady state numerical calculation using the finite element method coupled with finite difference method is performed to predict the heat rejection characteristics of the heat pipe radiator. Heat transfer performances of the new concept heat pipe radiator and the pumped loop radiator are compared with key parameters. The results show that the designed heat pipe radiator satisfies the waste heat rejection requirements of the TOPAZ-II power system under the normal operating conditions and has an ideal redundancy. The isothermal and safety of heat pipe radiator is also better than that of pumped loop radiator. [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]

• 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]

• 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]

• A mathematical model for high-temperature heat pipes is developed. • The relative error are 1.04% at steady state and 23.4% during startup. • The vapor velocity decrease with power. Heat pipe cooled reactor (HPR) is an innovative solid-core reactor which utilizes heat pipes to extract heat from the core to the energy conversion system. In this paper, a mathematical model for high-temperature heat pipes from a frozen state to a steady state is developed based on experimental data. The model is simplified and solved numerically. The two-dimensional transient heat conduction model of the wall and wick and the one-dimensional quasi-steady compressible vapor flow model are coupled to simulate the operation of a heat pipe when vaporization and condensation occur at the liquid–vapor interface. And the equivalent heat capacity model and the rarefied vapor model are introduced to simulate the process of heat pipe startup from a frozen state. Numerical results indicate that the model of heat pipe can predict the operation of high-temperature heat pipe with an error of less than 1.04% at steady state and 23.4% during the startup. Based on the simulation results, the velocity of vapor decreases with the heat transfer power in the steady state and the velocity of potassium vapor is 58.4% of sodium vapor velocity under the same operation condition. During the startup, when the rarefied-continuum interface located in the end of the evaporator section, the maximum vapor velocity occurs which is 1.3 times of steady velocity. This work makes it possible to predict the operation of a high-temperature heat pipe and provides a reference for the design and application of HPR. [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]

• A new conceptual design of a megawatt power level heat pipe cooled space reactor power system. • The heatpipe-fuel element was adopted. • Neutronic and thermal–hydraulic analysis of the reactor. • The He-Xe Brayton conversion unit was preliminarily designed. A new conceptual design of a megawatt power level heat pipe cooled space reactor (HPCR) power system adopted the integrated heatpipe-fuel modules is developed. A moderated thermal-neutron spectrum reactor, is designed with the thermal power of 3.2 MW. It consists of 234 HP-Fuel elements, which is a central lithium heat pipe surrounded by highly enriched UN fuel pellet with cladding on both radial sides. The dual closed Brayton dynamic conversion using binary He-Xe gas as working fluid is adopted. System waste heat is transported by the sodium–potassium (NaK) alloy cooling circuit to the potassium heat pipe radiation panels. The neutronic and CFD thermal–hydraulic analysis results are provided for the current design. The normal operation condition and event of a heat pipe failure are calculated and assessed. The redundancy and reliability of the core design is evaluated. The main design parameters of the dual-Brayton energy conversion system and heat pipe radiation panel are also illustrated. System heat balance analysis is performed. The potassium heat pipe radiator is designed and evaluated for the thermal performance. [ABSTRACT FROM AUTHOR]

• Code development and analysis of heat pipe cooled PRHRS of MSR are conducted. • Reactor model shows a good agreement with MSRE data and results. • Heat pipe cooled PRHRS model is validated with experimental data of 2.8% maximum error. The high efficient and reliable Passive Residual Heat Removal System (PRHRS) is of great importance to the development of Molten Salt Reactor (MSR), whether the liquid fuel MSR or solid fuel MSR. In this research, based on the structure and operation performance of the new conceptual PRHRS, using high temperature heat pipe as heat transportation, a set of reasonable mathematical and physical models have been built, mainly including reactor core model, high temperature heat pipe model and PRHRS model. The new conceptual Passive Residual heat removal system Analysis Code (PRAC) has been developed using FORTRAN language. The verification and validation for PRAC is conducted by using existed Molten Salt Reactor Experiment (MSRE) data provided by Oak Ridge National Laboratory and the PRHRS experimental data in our previous work. The results show that the developed PRAC could predict the MSR transients reasonably and show high fidelity. The numerical results of reactor core neutronic and thermal–hydraulic model show a good agreement when compared with MSRE pump startup and shutdown transients. The numerical results by PRHRS model agree well with experimental data which the maximum error of the fluoride salt temperature and heat pipe wall temperature are within 1.2% and 2.8%. This research provides a valuable software basis for the design and analysis of PRHRS of MSR. [ABSTRACT FROM AUTHOR]

• A code is developed for studying the operating thermal-hydraulic characteristics of the reactor core, which consists of core thermal-hydraulic model, neutron kinetics model and heat pipe startup model. • In order to verify the code, a series of experiments about potassium heat pipe are operated. • Steady-state and startup transient are performed by the code. • This research provides valuable experience for designing and formulating control strategies of the heat pipe cooled reactor. Heat pipe cooled reactor has been proposed for applying in the space station and underwater exploration featured with high reliability, low noise level and compact structure. A typical design of a heat pipe cooled reactor system is HP-STMCs designed by the University of New Mexico. It is conceptually designed as a fast reactor, lithium-filled heat pipes that transfer heat to potassium heat pipes radiator and thermoelectric generators. In this paper, a code is developed for studying the operating thermal–hydraulic characteristics of the reactor core, which consists of core thermal–hydraulic model, neutron kinetics model and heat pipe startup model. In order to verify the code, a series of experiments about potassium heat pipe are operated. Steady-state and startup transient are performed by the code. The highest temperatures of fuel and matrix in reactor core show at the 2nd channel, which reach 1720 K and 1545 K from the results of steady-state. As for the startup transient, the code for heat pipe cooled reactor core successfully predicts key parameters during startup transient, such as each layer material temperature of reactor core, a single heat pipe's temperature response and heat rejection power. The code has initially predicted the transient response of the heat pipe cooled reactor under power increasing rates of 800 W/s and130W/s. It is discussed that the startup strategy of power increasing rates is based on heat pipe transition temperature (916 K). This research provides valuable experience for designing and formulating control strategies of the heat pipe cooled reactor. [ABSTRACT FROM AUTHOR]

• The thermoelectric space reactor system model was proposed. • The reactor model, energy conversion assembly model, heat radiator model were established. • An electric efficiency model was integrated in the ECA model. • Heat resistance network was utilized to model the heat pipe. • Demonstration transients were calculated and analyzed. The thermoelectric space reactor power system conceptual design incorporates fast reactor, liquid-lithium primary coolant loops that transfer heat to potassium-filled high temperature heat pipes and thermoelectric generators. An integrated system analysis model was developed to study the operating characteristics of the space power system, which consisted of reactor thermal-hydraulic model, neutron kinetics model, thermoelectric energy conversion assembly (ECA) model, heat pipe radiator model etc. One-dimensional thermal-hydraulic model was applied to model the coolant circuit. Quasi two-dimensional analysis models for heat transfer of fuel element, ECA and the heat pipe fin were established. Considering the thermoelectric conversion phenomena, an equivalent electric efficiency model was inserted in the ECA model. Both finite element method and thermal resistance network were applied to simulate the potassium heat pipe system. The normal operation condition and two demonstration accident scenarios (unprotected inadvertent movement of a sliding reflector and loss of heat pipes) were calculated to prove the capabilities of the new system model. [ABSTRACT FROM AUTHOR]