Sodium Heat Pipe Research Articles

The evaporation of nanoscale sodium liquid film on the non-ideal nanostructure surface: A molecular dynamics study

The nanoscale liquid sodium film inside the microporous wick structure is of great importance to understanding the evaporation mechanism of the sodium heat pipe. The novel optimized wick structure is made of several layers of special screen. The surface of each screen exhibits a nanostructure type. Some non-ideal nanostructures may result from experimental faults or limits. And they will have an effect on the evaporation of film. In the present study, molecular dynamics is adopted to investigate this effect. The simulation system consists of the liquid sodium film and the solid surface. The flat surface is set as the reference. Based on the three non-ideal shapes of deposition, the sinusoidal nanostructures, conical nanostructures, and spherical nanostructures are built. The results indicate that the evaporation is suppressed by the above nanostructure surfaces. The weakening effect is through three forms: the potential gradient of the liquid film is intensified and the evaporation difficulty is increased; the heat transfer in the solid-liquid contact region is impeded; the collision heat transfer inside the liquid film is affected due to the delay of the aggregation variation between liquid atoms.

Molecular dynamics study of the wettability effect on the evaporation of thin liquid sodium film

The nanoscale liquid sodium film exists in the porous wick of the sodium heat pipe. Understanding its evaporation is essential for further exploring the heat transfer inside the heat pipe. Due to the solid–liquid interaction, surface wettability has significant effects on evaporation. In this work, molecular dynamics is adopted to study these effects. A cuboid evaporation system is constructed. It consists of the bottom solid surface and thin liquid sodium film. By changing the potential well depth between solid and liquid atoms, four surfaces with different wettability are built. The 400 K equilibrium evaporation is achieved. The self-diffusion coefficients of liquid atoms on four surfaces are acquired. The non-equilibrium is introduced by increasing the heat source wall’s temperature to 600 K. The main phase of evaporation is in 0–8 ns. The variation of gas atoms number is observed. The evaporation rate, average heat flux, and heat transfer coefficient are acquired. The results indicate that evaporation could be improved with surface wettability. For the 0°, 58° hydrophilic surface, the heat transfer efficiency could reach 96.9 %, 96.3 % respectively. The surface wettability effects are achieved by two paths: it enhances the heat transfer in the solid–liquid contact region and intensifies the atoms’ collision heat transfer inside the liquid film.

Preliminary conceptual design and analysis of a 100 kW e level Nuclear Silent Thermal‐Electrical Reactor ( NUSTER ‐100)

To meet the requirements of the unmanned underwater vehicles (UUVs) for the power source, a conceptual design of a 100-kWe level Nuclear Silent Thermal-Electrical Reactor (NUSTER-100) is proposed. The NUSTER-100 is a 1-MWt reactor that has been designed to couple with a 100-kWe thermoelectric generator system. The high temperature sodium heat pipes are adopted to cool the reactor core. The cascaded wide temperature range thermoelectric generation (TEG) system is used to achieve thermoelectric conversion, which ensures high power generation efficiency and low operating noise. The neutronic and thermal-mechanical analysis results are provided for the design. The core power distribution and energy spectrum are obtained. The neutronic analysis results are imported into ANSYS as input for thermal-mechanical analysis. In this paper, the fuel-HPs channel and the 1/8 symmetrical core are selected as the analysis objects, and the temperature distribution and statics parameters under the normal operation condition are obtained. The analysis results show that under normal operation conditions, the core parameters meet the design requirements. These design parameters are all lower than the safety limit. This work provides a certain reference for the design and application of the heat pipe reactor power system.

Experimental study of explosion characteristics caused by High-Temperature sodium leakage from a heat pipe into air

Heat pipe cooled reactor is an advanced reactor design that uses an array of in-core heat pipes for passive heat transfer.In heat pipe leakage accident, combustion or even explosions produced by high-temperature sodium reacting with the air can cause core damage. In this paper, the explosion characteristics of the sodium leakage are studied by analyzing the physical phenomena and studying the pressure pulse and the temperature rise. Based on the explosion phenomena, this paper proposes a phenomenological model for sodium heat pipe leakages. A small explosion, accompanied by yellow light, occurred at the instant of the sodium heat pipe leakage. The overpressure observed, increased with an increase in the initial temperature and also with the break size. The results of this paper include experimental phenomena and quantitative analysis of explosion, which provide a basis for the development of subsequent mechanistic model and safety evaluation of the heat pipe cooled reactor.

Performance of large-diameter high-temperature sodium heat pipes at high heat flux

高温热管广泛应用于航空航天、核工业、太阳能发电等领域, 其启动特性和均温性是高温热管复杂工况下的传热性能的重要指标. 根据现有研究, 随着热管直径的增大, 高温热管的总传热量应随之增加, 但热管的启动性能和均温性能与其总传热量是两个相对立的性能表现. 在高热流密度环境下, 常规直径的高温热管受限于结构尺寸已无法快速完成热量疏导, 而大直径的高温热管由于其很高的总传热量, 在高热流密度条件下具有广泛的发展潜力和应用前景. 本文制备了大直径高温热管, 研究了其在高热流密度条件下的性能; 对比了不同倾角和加热热流密度下高温热管的启动性能, 分析了大直径高温热管轴向、径向均温性以及等效热阻. 结果表明: 在高热流密度和自然对流条件下, 高温热管能够在-45°～ 90°范围内正常启动, 具有一定的抗重力特性. 根据努森数将蒸汽状态分为自由分子流、过渡分子流和连续分子流, 以此为基础可将大直径高温热管启动过程划分为6个阶段, 分别为: 蒸发段启动阶段、绝热段过渡阶段、绝热段启动阶段、冷凝段过渡阶段、冷凝段启动阶段和稳态阶段, 启动期间温度曲线形成4个“阶梯”. 这6个阶段和4个温度“阶梯”在实验中均被观测到, 与实验结果吻合，由此归纳出了大直径高温热管典型的6阶段升温曲线. 此外, 该高温热管在高热流密度下表现出优异的启动性、均温性和低热阻特性, 在420 kW/m2热流密度时, 高温热管在750 s内完成启动, 轴向平均温差小于80℃, 最小等效热阻可以达到0.02 K/W; 最后, 大直径的高温热管会引起沿热管管壁和毛细芯径向方向的热阻增加, 增大液态工质回流的阻力, 进而影响高温热管性能. 实验发现, 高温热管径向温差最大可达50℃, 但较高的热流密度能够改善大直径高温热管径向均温性.

Experimental evaluation on heat transfer limits of sodium heat pipe with screen mesh for nuclear reactor system

• Wall and vapor temperature of heat pipes are measured. • Capillary limits are characterized experimentally and are predicted by the model of Chi under horizontal condition. • Sonic limits are characterized and occur during the frozen startup process. • A model of entrainment limit is proposed and verified by the experiments. • Transition temperature of vapor is measured experimentally and is affected by the inclination angle. Heat pipe cooled reactor (HPR) is an ideal technology and a potential future niche where reliability and simplicity are key requirements. The heat transfer capacity of heat pipes determines the application range of HPR. Heat transfer limits (HTLs) of heat pipes are investigated, and the models of HTLs are evaluated in this study. Various HTLs of heat pipes with different filling ratios are tested. For the capillary limit, there is a sudden dryout in the evaporator section, in which the model of Chi is applicable for the horizontal conditions with a deviation of 19.0% but not for the inclined conditions. There is a possibility of sonic limit for a temperature rise in the adiabatic and condenser sections. The sonic limit predicted by Levy model is always higher than the experimental results. The entrainment limit is accompanied by a sudden rise in temperature of the evaporator section and temperature fluctuation in the condenser section, which can be estimated by using the wave-induced model with a relative error of 45.3% and is affected by the wick and inclination angle. The Knudsen number is used to determine the vapor state and the recommended critical Knudsen number of 0.01 is less than experimental data, which neglects the effect of inclination angle. The experimental results provide references for the heat pipe design and the accuracy of the present models are estimated.

Design optimization of gap distance for the capillary limitation of a heat pipe with annular-type wick structure

In this study, an experimental investigation was conducted on the rising height and contact angle of fluid in an annular wick-type heat pipe. The annular wick-type heat pipe was characterized by a small gap between the wick structure and tube wall, which compensated for the pressure drop along the porous media and created additional capillary force. To describe and model the advantage of this gap, the rising of a wetting liquid in the gap between a vertical solid plate and a mesh (with a small angle between them) was experimentally measured and analyzed. An additional experiment was performed to investigate the effect of curvature on the capillary rise using tubes and meshes of varying radii. Resultantly, we confirmed that the linear combination of the contact angles of the solid plate and mesh could be applied to calculate the rising height from the Laplace–Young equation. Furthermore, the effect of curvature on the rising height of the liquid was negligible. These results were extended to the investigation of finding the optimal gap distance for the annular wick-type heat pipe by referring to previous studies. We observed that a gap distance of 1.27 mm provided the largest permeability (K) over the effective pore radius (reff) value for a heat pipe with ethanol, which in turn resulted in the highest capillary limitation. For a sodium heat pipe, a gap distance of 0.84 mm resulted in the highest capillary limitation.

HEART, a specific code for thermal-electrical analysis of heat pipe cooled nuclear reactor

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.

An Improved Model of the Heat Pipe Based on the Network Method Applied on a Heat Pipe Cooled Reactor

As the unique piece of heat transport equipment in a heat pipe cooled reactor, an accurate simulation of the heat pipe is helpful for understanding the real behavior of the reactor core and the reactor system. Based on the network method, an improved model for the heat pipe which considers the heat conductance in the wall, the vapor flow in the vapor space, and the liquid flow in the wick is proposed. Meanwhile, the gravity term is also added to the flow equation. Compared with the experimental results of a copper-water heat pipe, the validity of this model is verified. Then, a high-temperature sodium heat pipe of 1.0 m length is selected as the study object. Based on the analysis, it can be found that the total temperature difference of the heat pipe is 31.7 K, and the temperature drop caused by the vapor flow is only 2.6 K. As for the flow pressure drop in the heat pipe, the pressure drop is mainly concentrated in the wick region, which is 8,422.47 Pa, and the pressure drop in the vapor space is only 896.68 Pa. In cases of non-uniform heating and cooling, high heat leakage, and inclined operation, results indicate that the greater the non-uniformity of heating or cooling, the greater will be the temperature drop of the heat pipe. With the increase of heat leakage, the operating temperature of the heat pipe decreases significantly, and the total temperature drop increases. The heat pipe can operate at all positive inclination angles, but when the inclination angle exceeds−30°, the heat pipe will reach the capillary limit, and it may be damaged. All of these results can provide support for the design and simulation of a heat pipe cooled reactor.

Molecular dynamics study of liquid sodium film evaporation and condensation by Lennard-Jones potential

Deeply understanding the phase change of thin liquid sodium film inside wick pore is very important for further studying high-temperature sodium heat pipe's heat transfer. For the first time, the evaporation and condensation of thin liquid sodium film are investigated by the Lennard-Jones potential of molecular dynamics. Based on the startup and normal operation of the sodium heat pipe, three different cases are simulated. First, the equilibrium is achieved and the Mass Accommodation Coefficients of the three cases are 0.3886, 0.2119, 0.2615 respectively. Secondly, the non-equilibrium is built. The change of liquid film thickness, the number of gas atoms, the net evaporation flux (Jnet), the heat transfer coefficient (h) at the liquid-gas interface are acquired. Results indicate that the magnitude of the Jnet and the h increase with the basic equilibrium temperature. In 520–600 K (the startup of the heat pipe), the h has approached 5–6 W m−2 K−1 while liquid film thickness is in 11–13 nm. The fact shows that during the initial startup of the sodium heat pipe, the thermal resistance at the liquid-gas interface can't be negligible. This work is the complement and extension for macroscopic investigation of heat transfer inside sodium heat pipe. It can provide a reference for further numerical simulation and optimal design of the sodium heat pipe in the future.

Molecular dynamics simulation of the evaporation of liquid sodium film in the presence of non-condensable gas

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 (Jnet), the heat transfer coefficient (h) at the liquid–gas interface are obtained. It is found that the NCG’s effects on the Jnet are achieved through its influence on the MAC. When the variation of the MAC is above 27%, NCG’s effects on the Jnet 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.

The development of high temperature heat-pipe transient model for system analysis of heat pipe cooled microreactor

The miniaturization of nuclear reactor power is the direction of nuclear energy development in the future. A nuclear power system with a small solid core and cooled by a high-temperature alkali metal heat pipe has excellent safety performance. The start-up of the heat pipe cooled reactor depends on the heat pipe start-up characteristics from the frozen state, so transient analysis models of high-temperature heat pipe from the frozen state are established in this paper for the system characteristics analysis of the heat pipe cooled microreactor. The models include three stages of high-temperature heat pipe start-up from the frozen state: the molecular diffusion model of the free molecular stage, the flat-front model of the transitional flow process, and the 2-D thermal resistance network model of continuous flow. The energy conservation model is also used to consider the phase transition between solid, liquid, and gas. And The heat transfer limit model of heat pipe has been also coupled in the model. The models have been verified by the start-up experimental dates of the sodium heat pipe and the transient experimental dates of the water heat pipe. Then the models are applied to the system analysis code of the heat pipe reactor to analyze the transient experiment characteristics of a typical heat pipe reactor system. The results show that the predicted trend by HPMR system code is reasonable and the deviation with experimental data is less than 2%, which further demonstrates the applicability and effectiveness of the heat pipe transient models.

EBlock37 Microreactor Electrical Demonstration Unit

The eBlock37 is a subscale electrically heated and heat-pipe-cooled prototype of a fast spectrum microreactor that is under development at Los Alamos National Laboratory. The prototype consists of an electrically heated core and gas-cooled heat exchanger. These subassemblies, both built from 316 L stainless steel, are thermally linked by an array of 37 sodium heat pipes that transfer a nominal 100 kW from the core at 700°C. An overarching objective of this effort is to overcome challenges associated with core block and heat exchanger manufacture and integration of high-temperature heat pipes into the assembly. Components that would be safety critical in an actual reactor, such as the heat pipe wicks, are being built under a Nuclear Quality Assurance 1 quality program. The completed assembly is intended for non-nuclear electrically heated testing, which will be conducted at a demonstration facility at Idaho National Laboratory. This paper provides a top-level summary of the efforts to date.

Effect of inclination angle on the startup of a frozen sodium heat pipe

Screen-wick sodium heat pipes are highly effective passive devices for heat transport, which are mainly used in a horizontal orientation. For practical applications, the sodium heat pipe may operate at an inclination. Previous studies have mainly focused on large inclination angles where gravity dominates. By contrast, fewer experiments were conducted at a near-horizontal inclination, where neither the capillary nor gravity dominate. The coupling effects may significantly change heat pipes’ startup and operation characteristics. This work experimentally investigated the temperature variations during a sodium heat pipe frozen startup at near-horizontal inclination angles, focusing on the effects of inclination angle (−15°, −10°, 0°, 5°, 10°, 15°, 30°, 45°) with a heating power of 0–1.4 kW. Various startup power regimes are identified for the various conditions. Critical startup power transits at 15° due to the competing effects of the capillary and gravitational forces. During the transient startup, the outer wall temperatures oscillated at positive inclination angles when the evaporator bottom temperature is over 580 ℃, with the oscillations suddenly stopping at 780 ℃. The inclination angle significantly affects the oscillation periods and amplitudes, which first increased and then decreased with increasing inclination angle, with peaks at 15°.

Thermal-electrical coupling characteristic analysis of the heat pipe cooled reactor with static thermoelectric conversion

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.

Assessment of Screen-Covered Grooved Sodium Heat Pipes for Microreactor Applications

New nuclear reactor designs that incorporate heat pipes are being investigated for possible near-term deployment in terrestrial applications. This study explores the use of screen-covered axially grooved sodium heat pipes and their applicability for providing heat removal for microreactors. A sodium working fluid is appropriate for microreactors operating in the 5 to 20 MW(thermal) range at approximately 650°C. HTPIPE, a legacy software code, was validated for the case of screen-covered grooves and used to perform steady-state analyses to determine the performance limits of a proposed heat pipe design. The performance limits of a sodium heat pipe with a screen-covered square grooved wick structure is compared to that of an equivalent heat pipe with an annular wick. In a horizontal orientation at an operating temperature of 650°C,the performance limits for the heat pipe with an annular wick configuration are 15% higher than for the screen-covered grooved wick. At operating temperatures below 777°C, the annular wick outperforms the screen-covered grooved wick, and at temperatures above 777°C, the screen-covered grooved wick outperforms the annular wick. However, the marginal performance gain at higher temperatures may not justify the use of heat pipes with a screen-covered grooved wick structure due to increased manufacturing costs.

Reactor core design of UPR-s: A nuclear reactor for silence thermoelectric system NUSTER

The heat pipe reactor is featured with the passive heat discharge, anti-single point failure and silent operation. Therefore, it could be applied for the power system of underwater vehicles. In this paper, the concept of Unattended Uranium-fueled Portable Reactor (UPR) for the NUclear Silence ThermoElectric Reactor (NUSTER) power system is introduced and the reactor core with a thermal power of 1 MW is neutronics designed. In the core, highly enriched UO2 fuel, sodium heat pipe, molybdenum alloy matrix, beryllium oxide and beryllium reflector are adopted and the reactivity is controlled by four sliding reflectors and four safety rods. In design process, the NECP-MCX code is used to carry out the neutronics calculation. At the same time, the heating unit, core loading, control system and power distribution of the core are described. Furthermore, criticality safety, burnup characteristics, power distribution of the core are specifically analyzed. The results show that the proposed core design meets the requirements of criticality safety and operating life.

Rovibrationally resolved radiative lifetime and collisional cross section measurements of the 6[formula omitted] state of molecular sodium

We report on radiative lifetime and collisional cross section measurements of the Na2 61Σg+(v=6, J=31) and 61Σg+(v=8, J=31) molecular levels by time-resolved laser spectroscopy. The excitation to the 61Σg+(3s+5s) electronic state was achieved using two synchronized pulsed lasers, directed to the sodium heatpipe oven and molecular fluorescence decay time as a function of buffer gas was recorded via time-correlated single photon counting technique. Radiative lifetimes were measured using a Stern-Volmer plots. From the lifetime measurements, collisional cross sections between the sodium molecules in the 61Σg+ state and the ground state argon atoms were extracted.

Analysis of unprotected loss of heat sink accident in compact high temperature reactor cooled under natural circulation of LBE coolant

India is developing 100kWth Compact High Temperature Reactor (CHTR) to demonstrate the technology enabling high-temperature process heat applications such as hydrogen production. The reactor core is envisaged with 233U-Th based TRISO coated fuel particles and beryllium oxide moderator in prismatic fuel assemblies for high-temperature operation. The core heat is designed to be removed by lead–bismuth eutectic (LBE) coolant under natural circulation in vertical channels of CHTR at nominal power. The heat generated in the primary circuit is to be transferred to the secondary side using high-temperature sodium heat pipes. The design and safety of the prototype core necessitate comprehensive, integrated neutronics-thermal hydraulics analyses of anticipated transients. In this context, the Unprotected Loss of Heat Sink Accident (ULOHSA) at full power during the beginning of the core life cycle is investigated with indigenous 3D multi-physics code ARCH-TH. The key parameters defining neutronics and thermal-hydraulics of the core have been assessed in the steady-state as well as during the transient. The findings, along with the model of the simulation and the code validation are discussed in this paper. The analyses indicate that the core is being shut down passively due to reactivity feedbacks and the loss of heat sink leads to reverse flow of the coolant in the low powered channels during the transient. However, the coolant and fuel peak temperatures are observed to stay well below the leak-tightness limit of TRISO particles and boiling point of LBE, even considering the failure of shutdown systems. The modifications in the design of CHTR are also investigated to pacify the condition of flow reversal during such transients.

Heat Transport and Power Conversion of the Kilopower Reactor Test

The Kilopower reactors have been designed to provide a steady-state thermal power range between 4 and 40 kW and to convert the heat generated to an electrical output of 1 to 10 kW(electric), providing an overall system efficiency of 25%. This range of thermal and electrical power has been derived from two basic designs: the small 1-kW(electric) design and the larger 10- kW(electric) electric design intended to support science and human exploration missions for surface and in-space power. The Kilowatt Reactor Using Stirling TechnologY (KRUSTY) experiment was built using the 1-kW(electric) Kilopower design to make the test affordable by using existing infrastructure and to complete it in a 3-year timeframe. The data from the smaller, lower-mass system could be extended to the larger 10-kW(electric) system, knowing that the materials and neutronic design are similar. Each of these designs use the same fuel, heat transport systems, and power conversion systems at the appropriate scale to produce the desired electrical output power for mission use. The heat transport system uses multiple heat pipes that operate passively and do not require any electrical pumps or other parasitic loads to cool the reactor core. This type of reactor cooling provides several layers of redundancy and makes it ideal for coupling a self-regulating reactor to a variable-output power conversion system. The power converters accept the reactor heat that has been delivered by the heat pipes and create the needed electrical power through their thermodynamic Stirling cycle and linear alternator. This paper provides details about the sodium heat pipes used in the experiment, the Stirling power converters that create the electricity, and the overall power system that make up the 1-kW(electric) Kilopower reactor.