Space Reactor Research Articles

Numerical Investigation of the Heat Transfer Characteristics of Liquid Lithium Metal in Spiral Tubes

Abstract The application of liquid lithium as a coolant for the forthcoming era of space nuclear reactors exhibits significant potential, and spiral tube heat exchanger components are commonly used in steam generators for reactors. However, the heat transfer characteristics of liquid lithium in spiral tubes are not yet fully understood. This research establishes a non-isothermal heat transfer model incorporating a modified turbulent Prandtl number to analyze the flow of liquid lithium through spiral tubes with varying geometries. Numerical analysis is carried out focusing on the influence of inlet velocity, the distribution of related parameters, and the geometry of spiral tubes. The results demonstrate that in the range of the dimensionless Dean number 8165–13,063, the Nusselt number and the pressure drop present approximately linear relations with the Dean number. For the distribution law of relevant physical quantities, the inner side of the tube displays an eye-anatomy low-flowrate area and a high-temperature area, while a low-pressure area forms on the inner pipe wall. Finally, the pitch and spiral radius are found to be reduced as much as possible to ensure high liquid lithium-based heat transfer performance with a small pressure drop. The optimized design parameters reveal that within the actual design range of non-dimensional pitch of 0.667–10.667 and curvature of 0.0556–0.1667, the non-dimensional pitch and curvature are 0.667–2.667 and 0.1667, respectively. This study offers valuable insights into the heat transfer properties of liquid lithium within heat exchangers of the spiral tube design, promoting its application in space nuclear reactor power supply.

Key features and highly effective prediction of complete startup from frozen state for high-temperature heat pipe in heat pipe reactor

High-temperature heat pipe is an efficient phase change heat transfer device, widely applied in the space nuclear reactors, the China Fusion Engineering Test Reactor, solar collectors, and the thermal protection of hypersonic vehicles. In order to understand the startup characteristics of the heat pipe and to predict its transient behavior with high accuracy, this paper first experimentally investigated the startup performance of a sodium heat pipe from the frozen state under different heating powers from 534 W to 1897 W. Then, a new compressible flow model for startup from the frozen state was developed and verified, solved the independence problems of mesh number and time-step used in the numerical treatments of most traditional heat pipe codes. Finally, based on the assessed model and experimental data, the startup procedure and the influences of heating power on the working temperature, kick-in time, and temperature difference were analyzed respectively. The comparisons indicated that the new model is extremely informative and effective, and the results provided knowledge for understanding the safe and optimum working regime of a high-temperature heat pipe using in heat pipe reactor.

Study of Soot Deposits during Continuous Methane Pyrolysis in a Corundum Tube

Methane pyrolysis is one of the promising methods for producing low-carbon hydrogen, while one of the main problems of methane pyrolysis technology is soot clogging of the reactor space. In this work, soot deposits were studied during continuous methane pyrolysis in a corundum tube with an inner diameter of 50 mm. Experiments were carried out at temperatures of 1000 °C, 1050 °C, 1100 °C, 1200 °C and 1400 °C with methane flow rates of 1 L/min and 5 L/min. Each experiment lasted 1 h. The formed soot accumulated inside the reactor (corundum tube) and the connected filter, where the gaseous product of methane pyrolysis was separated from the soot. The gaseous product was studied by gas chromatography. The soot was studied by SEM, BET and ICP-MS. With an increase in the temperature of the pyrolysis process from 1000 to 1200 °C, the hydrogen yield increased from 28.64 to 92.74% and from 1.10% to 72.09% at a methane flow rate of 1 and 5 L/min, respectively. The yield of soot increased from 1.28 g at 1000 °C to 43.9 g at 1400 °C (at a methane flow rate of 1 L/min). With an increase in the flow rate of methane from 1 to 5 l/min, the yield of soot at 1200 °C increased by almost two times to 75.65 g. It was established that in the region of the reactor where maximum heating occurs, the accumulated soot sinters and forms dense growths. At 1050 °C, the particle size of soot varies from 155 to 650 nm, at 1200 °C—from 157 to 896 nm, and at 1400 °C—from 77 to 532 nm. The specific surface of soot was 3.5 m2/g at 1000 °C and 8.0 m2/g at 1400 °C. The purity of the produced carbon black was about 99.95%. This study is useful in the selection of materials and technical solutions for a pilot plant for methane pyrolysis.

A super-real-time three-dimension computing method of digital twins in space nuclear power

Digital twins (DTs) have attracted widespread attention in academia and industry in recent years. It can accurately reflect the physical world in real-time, enabling online monitoring, control, and prediction operations. Their foundation is super-real-time computing and high data representation capabilities. However, current DTs do not achieve 3D super-real-time computing. This study proposes a novel 3D computational method for solving fluid–solid coupling problems in a super-real-time. The method is based on a mixed solution framework that combines traditional numerical methods with deep learning operators. Specifically, the method employs multi-core CPU parallel acceleration to solve the solid equations while leveraging the computing power of GPU to solve the fluid equations. The fluid–solid coupling is achieved through information exchange between the GPU and the multi-core CPU. In addition, the proposed method introduces a new deep learning operator framework based on the DeepONET. The framework is accompanied by a database structure that facilitates model training and validation and a loss function that guides the training. The space nuclear reactor, an improved TOPAZ-II system, was selected to demonstrate its feasibility. Four non-training transient conditions were simulated to test the generalization performance. The results show that the proposed method achieves an average error between the calculated results and reference values below 2.5%, with the average error of thermodynamic parameters below 1.5%. The average deviation between system parameter peak values during the transient process and the reference value was less than 5 s. The result meets the acceptable error level and satisfies the super-real-time requirements with a time acceleration ratio of approximately 1.17, which is 60 times faster than traditional numerical methods. The results demonstrate the accuracy and efficiency of the proposed method for DT.

Control strategy study and cascade control system design for space thermionic nuclear reactor

Nuclear reactors are gaining increasing attention for use in spacecraft owing to their long life and high power-density. Because the dynamic characteristics and operation strategy of a space nuclear reactor (SNR) are different from those of a traditional reactor, its control strategy must be investigated to design an appropriate control system. In this study, the dynamic characteristics of the time and frequency domains are analyzed. Based on this analysis, a control strategy is proposed. The cascade control method is applied to the design of control systems, including electric power and coolant temperature control systems. The controller parameters are tuned using the frequency-domain analysis method. To verify the control performance, step and ramp disturbances of the electric power and coolant temperature setpoints are introduced, and comprehensive comparisons are made. Additionally, the startup process is simulated; it reveals that the cascade control systems perform better compared with the traditional proportional-integral-derivative (PID) control. Moreover, the control performance of the wide-range power regulation is smooth and steady. Thus, the control strategy and cascade control system designed in this study can be applied to the thermionic nuclear reactors in space.

Dehydrogenation of cycloalkanes over Pt/SBA-15 for endothermic cooling

Dehydrogenation of cyclohexane, methylcyclohexane, and decalin was studied on Pt/SBA-15 between 773 and 873 K for application to engine cooling in hypersonic aircraft. While the reactions were unstable at 1 bar, the reactions showed no evidence for deactivation at pressures of 20 and 60 bar for at least 4 h. Selectivities for dehydrogenation of cyclohexane to benzene and methylcyclohexane to toluene were greater than 90% in all cases. The reaction of decalin proceeded sequentially, forming primarily tetralin at low reactor space times and naphthalene at higher space times. The reaction of 50% mixtures of n-dodecane and decalin at 773 K showed negligible conversion of n-dodecane but the reaction of decalin was similar to what was observed for a pure decalin feed. Heat measurements gave endothermicities that were in good agreement with standard reaction enthalpies.

Transient thermal–hydraulic analysis of the thermionic space reactor TOPAZ-II

Space nuclear reactor system is of great significance in aerospace applications such as deep-space exploitation and exploration, high-power earth-orbiting satellites. A transient reactor system analysis code, RESYS, is developed for advanced non-light water reactor safety analysis recently. In this paper, the model of the thermionic space nuclear reactor TOPAZ-II including the thermionic fuel elements, thermionic energy conversion system, reactor core, coolant loop and radiator is established with RESYS code. The point reactor equation with six-group delayed neutron is utilized to calculate the fission power. The reactivity effect of various structural components and the reactivity insertion effects of control drums are considered. The thermionic emission model of cesium thermionic converter is based on the phenomenological model established by Rasor. Then, the steady state operation and three transient accidents including reactivity insertion accident (RIA), partial loss flow accident (LOFA), load resistance shortcircuit accident (LRSCA)are simulated. In the steady-state condition, the output electric power is about 5.8 kW with thermoelectric conversion efficiency of 5.0%. The results of steady calculation agree well with the design values. And in the three transient accidents, the moderator introduces a large temperature reactivity feedback which makes the reactor fission power increasing. The results indicate that the TOPAZ-II reactor system has a sufficient safety margin during these accidents. In addition, the calculation results preliminarily demonstrate the capability of RESYS code in modeling and simulation of thermionic space reactor.

Steady-state performance analysis of loop heat pipe radiator in space reactor

The loop heat pipe has a good application prospect in the field of space reactor radiator because of its characteristics of the separation of gas/liquid pipelines and the long-distance efficient heat transfer. The comprehensive performance analysis of loop heat pipe radiator for space reactor is carried out by the heat balance calculation method based on energy coupling iteration proposed in this paper. The loop heat pipe radiator models of different configurations were developed using the mass and energy conservation theory. The models were validated by experimental data of the loop heat pipe under various heat loads. The model in this paper can correctly simulate the steady-state operation characteristics of loop heat pipe, and evaluate the design of loop heat pipe radiator for space reactor. The performance of loop heat pipe radiator was analyzed in this paper under various design conditions such as power, evaporation temperature, and heat sink temperature. The results show that the thermal operating parameters have significant effect on characteristics of loop heat pipe radiator. The design length of the condenser pipe was considered as an important factor to achieve steady state heat balance under different design power. Meanwhile, the operation characteristics of loop heat pipe radiator with different configurations were compared, which demonstrated that the performance of panel-type radiator was better than fin-type radiator.

State-space model construction and dynamic analysis for a space thermionic nuclear reactor

The space nuclear reactor (SNR) is an important energy pattern for spacecrafts due to its high power density, long life and stabilization. A linear dynamic model is necessary to analyze the dynamics and design the appropriate control system analytically. A simplified nonlinear model based on conservation equations including point reactor kinetics model, thermal-hydraulic model, coolant model and heat pipe and radiator model was constructed. The state-space linear model was derived using perturbation theory. To verify the linear model, comprehensive comparisons were made in both time domain and frequency domain. The dynamic characteristics with respect to reactor power and reactivity were obtained in frequency and time domains. It is concluded that the reactor is an unstable system with a large time delay and strong nonlinearity at different power levels. This study provides solid foundation for future control system design.

Thermal-mechanical coupling characteristics and heat pipe failure analysis of heat pipe cooled space reactor

Heat pipe cooled reactor has wide application prospects in space environment because of the advantages of high reliability and small mass. Heat Pipe-Segmented Thermoelectric Module Converters (HP–STMCs) is a typical design of space heat pipe cooled reactor. In this paper, a new thermal–mechanical coupling analysis method for the reactor was proposed. A two-dimensional heat pipe code for multiphase heat transfer was developed, with simplification of the wick as a purely conductive process and consideration of the flow of compressible vapor. A thermal stress computational model of the core was established using COMSOL Multiphysics, with the heat pipes divided into six channels. Based on various power distribution conditions, different heating powers were employed to simulate the heat pipe operation within each channel. Normal and accident operating characteristics were analyzed by the model. Under normal operation, the maximum temperature is 1641.9 K, located in the center of the core, and heat pipes in the fifth channel have the highest heat transfer power, reaching 13.674 kW. During accidents, thermal stress emerges as the main threat to the reactor's safety. A single heat pipe failure at any location poses no safety risk, but the thermal stress exceeds the safety limit if two heat pipes fail at the reactor core's edge, reaching 269.92Mpa. With three failed heat pipes, stress exceeds limits even at the well-cooled middle position and raises the local maximum temperature to 1959.8 K. This research provides valuable experience for local effects assessment and safety analysis of heat pipe cooled reactors.

Experimental investigation on the heat transfer characteristics of argon space containing sodium vapor

A large quantity of argon is typically provided at the top of the main vessel of a sodium-cooled fast reactor (SFR) to serve as covering gas. The sodium coolant in an SFR evaporates and mixes with the argon covering gas during reactor operation. This mixed gas has a critical influence on the heat transfer at the top of the main vessel, particularly the temperature of the roof slab. The temperature distribution in the roof slab significantly affects reactor safety as it serves as the pressure boundary of the main vessel. Research on the behavior of sodium vapor has typically employed numerical simulations, and any experimental verifications have generally used simplified experimental models. To study the heat transfer characteristics of sodium vapor and eliminate the influence of simplifications on the results, the Heat transfer Experimental Facility for Argon space at the top of the main vessel (HEFA) was established based on a scaled-down top of the main vessel. The HEFA was employed to analyze influence of sodium vapor on the heat transfer characteristics of the reactor space by evaluating the effects of various heating temperatures with and without liquid sodium on the temperature distributions in the argon and roof slab. Furthermore, the emissivity of the liquid sodium surface was measured. The results show that the radiative absorption effect of sodium vapor primarily occurs near the liquid sodium surface. When the heating temperature was less than 400 °C, the presence of sodium vapor increased the temperature of the main vessel. When the heating temperature was 540 °C, the radiative absorption effect of the sodium vapor was enhanced, reducing the roof slab temperature. The observed phenomena can be considered to ensure reactor safety during the design of SFRs.

Interval type-2 fuzzy logic control for a space nuclear reactor core power system

The uncertain environment and delayed communications bring great challenges for the space nuclear reactor core power control. The imprecise knowledge about nonlinear and time varying parameters of the reactor make the problem further difficult. In this work, an anti-interference method is proposed for the core power control of space reactors. A new core power controller is designed to compensate the power deviation caused by model uncertainties and external disturbances. To further approximate nonlinear mapping functions between the control drum angle position and the deviation of the reactor core power, an interval type-2 fuzzy logic system is employed to deduce the correction of control drum angle position for its general approximation ability. Moreover, in order to solve control difficulties caused by physical constraints of the control drum and high stability requirements of the core power control system, an optimized iteration algorithm is proposed to ensure convergence to the optimal solution. A 1MWe space nuclear reactor is used as a reference object to carry out the research and verification of the newly proposed power control scheme. The numerical results verify the feasibility of the novel controller and show that load tracking performance of the system can be realized effectively.

Investigation of the heat transfer enhancement and deterioration induced by vortex generators in low Prandtl number sodium-potassium alloy liquid

Liquid metals, being characterized by low molecular viscosity and high molecular thermal conductivity, appear to be a stupendous choice for heat transfer fluids in deep-space nuclear reactors. Nevertheless, because of space nuclear reactor weight and volume constraints, the heat transfer capability of the heat exchanger must be substantially improved for a deeper space exploration mission. In this study, the influence of the vortex generator applied to the liquid metal was investigated to further promote the heat transfer capability of the liquid metal. And the thermo-fluid behaviors of the sodium-potassium alloy (NaK-78) are compared among the channel with the Body-Centered Cubic (BCC) lattice array, the rectangular ring rib array, and the Kagome lattice array under identical porosity. The numerical results indicate that the heat transfer deterioration phenomena occurred in the case of rectangular ring rib channel, which is a relatively rare occurrence in conventional fluids with the local Nu number even being lower than that of the smooth channel at the same location. Therefore, the rectangular ring rib should be carefully considered and evaluated when it is applied to heat transfer enhancement in NaK alloy fluid. The Kagome structure and the BCC structure could effectively promote the heat transfer performance of the NaK alloy fluid by means of the induced composite vortex system. In comparison to the smooth channel, the average Nu of the cases with the Kagome array and the BCC array yield 17.26–37.32% and 51.98–78.85% higher, respectively.

The rule of law: A fragile tool for the development of emerging nuclear technologies

Before the Fukushima Daiichi Nuclear Power Plant accident in Japan in 2011, the nuclear industry had strongly promoted the idea that the time of nuclear renaissance had come after a long, fallow period in the wake of the nuclear accidents at Three Mile Island (1979) and Chernobyl (1986). During the post-Fukushima period, there were few new projects, but growing demand for energy and anxieties raised by climate change have brought us to a turning point. Despite the Fukushima accident, which led to some nuclear projects being delayed or cancelled, there is still a great deal of interest in the use of nuclear power for civil purposes. This is primarily because, as the International Energy Agency’s (IEA) Executive Director Fatih Birol has rightly pointed out: “Without an important contribution from nuclear power, the global energy transition will be that much harder.”1 In this regard, in 2010, 67 reactors were under construction, 120 planned and 441 in operation.2 In 2022, 11 years after the Fukushima Daiichi accident, there were 60 under construction, 104 planned and 338 proposed.3 The International Atomic Energy Agency (IAEA) estimates that nuclear electric power capacity will have increased by as much as 23% by 2030 and more than doubled by 2050.4 In addition to this renewed interest, many projects are now in development, opening up new prospects for the use of the atom for civil purposes. Nuclear fusion, small modular reactors (SMRs), the use of artificial intelligence, floating, underwater and space reactors, and nuclear batteries, to name but a few of the projects on the table, lead us to think that the “nuclear renaissance” is slowly shifting to a “nuclear spring”. In this view, where the concept of a renaissance involves new impetus for nuclear energy, with the construction of new facilities, the concept of “spring” refers to a determination to break with nuclear traditions, in terms of concepts, means and players. Consequently, this phenomenon calls for new legal rules which, in some cases, have already started to be debated.

Methanogen colonization and ‘end-of-life’ use of spent lignocellulose from a solid-state reactor as an inoculum source

Methanogen colonization on lignocellulosic feedstocks has been known, but few attempts have been made to characterize and quantify it. This study used total specific methanogenic assay (TSMA) to quantify the methanogenic activity build-up on ten lignocellulose and brought out distinct variations and patterns of conferred activity. TSMA of up to 53L CH4/d/kg residual TS was recorded. Based on the high methanation potential, the ‘end-of-life’ use of spent lignocellulose as a high-performance inoculum source was proposed for the solid-state reactors. The TSMA was devised to quantify the methanation potential of the spent lignocellulose-based solid-inoculum source. Further, “S/I analog” was proposed in solid-state reactor calculation- organic loading rate and inoculation volume. The study also used the two derivatives- TSMA and VS flux rates, to study the implication of the colonization on the reactor's waste stabilization potential. Here, a new concept of “equalization point” was presented. It referred to the stage of digestion where cumulative methanogen colonization build-up was adequate to convert cumulative VS generated from decomposing feedstock, i.e., the relative ratio of ≥1. This new direction would generate insights into the upper limits to which volumetric feed rates can be supported, enabling a higher reactor space economy.

Shutdown safety analysis of megawatt-class space gas-cooled reactor system

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.

Thermal hydraulic design and mass optimization of a 100-kWe S-CO2 cooled Mars-surface fission reactor system

Fission reactor with electrical power of 100 kW to megawatt is a promising power option for upcoming deep space exploration. To reduce the number of space launches and the costs, it is necessary to design an energy supply system coupling fission reactor with energy conversion system of high efficiency and compactness. The combination of gas-cooled fast reactor and supercritical carbon dioxide (S-CO2) Brayton cycle has the advantages of simple system layout, compact structure and high thermal efficiency.This paper aims to carry out researches on the thermal hydraulic design and mass optimization of 100kWe reactor system for Mars-surface application.An analysis code which includes thermodynamic models for different components in the system, as well as mass evaluation models for radiator, recuperator and reactor, is developed and validated to provide an analysis tool. An S-CO2 Brayton cycle directly cooled reactor system is proposed, which applies Tube-In-Duct fuel assembly to make up the core and heat pipe radiator as heat rejection system. The thermal hydraulic evaluation of the reactor core, thermodynamic performance evaluation as well as overall mass evaluation of the whole system are carried out to achieve an optimized 100kWe fission reactor system for Mars surface with minimum mass. The optimal design has an operating pressure of 16 MPa, core outlet temperature of 600 °C, radiator panel area of 120 m2 and an overall mass of 1862.98 kg considering the reactor and key components of Brayton cycle. S-CO2 Brayton cycle directly cooled reactor system has a comparable specific mass as He-Xe cooled reactor with lower operating temperature, which is beneficial for material selection and provides another option for space reactor development.

Study of the thermal expansion effects of a space nuclear reactor with an integrated honeycomb core design using OpenMC and ANSYS

The negative reactivity feedback in the nuclear fuel plays an important role in maintaining the stable operation of nuclear reactors. For heat pipe cooled space reactors using highly enriched U, the fuel temperature reactivity feedback mainly due to Doppler effect and the thermal expansion of solid fuel. In this paper, based on an integrated heat pipe space reactor design, a one-way coupling of the Monte Carlo code OpenMC and finite element method ANSYS is used to evaluate the reactor expansion and the impacts to the reactor reactivity. The average expansion was 0.126 cm for the fuel block from cold state to full power operation state. The fuel thermal expansion introduces a negative reactivity feedback of −59 pcm, and the reflector expansion introduces another −675 pcm, resulting a temperature coefficient of −0.638 pcm/K.

Thermal-hydraulic analysis of He–Xe gas mixture in 2×2 rod bundle wrapped with helical wires

Gas-cooled space reactor, which adopts He–Xe gas mixture as working fluid, is a better choice for megawatt power generation. In this paper, thermal-hydraulic characteristics of He–Xe gas mixture in 2 × 2 rod bundle wrapped with helical wires is numerically investigated. The velocity, pressure and temperature distribution of the coolant are obtained and analyzed. The results show that the existence of helical wires forms the vortexes and changes the velocity and temperature distribution. Hot spots are found at the contact corners between helical wires and fuel rods. The highest temperature of the hot spots reach 1600K, while the mainstream temperature is less than 400K. The helical wire structure increases the friction pressure drop by 20%–50%. The effect extent varies with the pitch and the number of helical wires. The helical wire structure leads to the reduction of Nusselt number. Comparing thermal-hydraulic performance ratios (THPR) of different structures, the THPR values are all less than 1. It means that gas-cooled space reactor adopting helical wires could not strengthen the core heat removal performance. This work provides the thermal-hydraulic design basis for He–Xe gas cooled space nuclear reactor.

Numerical Analysis of Liquid Potassium Flow and Heat Transfer in a Core for Space Nuclear Reactor Based on Entransy Dissipation Theory

AbstractThe research on the flow and heat transfer of alkali metals is an important technical means to strengthen the heat transfer of the core. In this paper, the 15 cooling channels with different structures are established. Under Re = 2.85 × 104–2.17 × 105, several operating parameters, such as core inlet temperature (T = 350 K, 390 K, 450 K, 550 K, 850 K), fuel rod pitch ratio (P = 1.2, 1.4, 1.6, 1.8, 2.0), and channel shape (triangle and square) on the heat transfer performance are numerically simulated. At the same time, this work introduces an evaluation criterion, entransy dissipation, to measure the irreversibility of the heat transfer process. The results show that the triangular channel has a stronger heat transfer capacity than the square. And the increase of the rod diameter ratio is a positive means to enhance the heat transfer. Among the studied channels, Model “I” (triangle channel, P = 2.0) has the best heat transfer performance. The above results have some guiding value to realize core heat transfer enhancement by adjusting fuel rod arrangement.