Nuclear Thermal Propulsion Research Articles

Post-irradiation examination of UN-Mo-W fuels for space nuclear propulsion

The National Aeronautics and Space Administration's return to space nuclear propulsion stems from the need for a more efficient method of space travel. Nuclear thermal propulsion systems have been shown to be two times more efficient than chemical propulsion. NASA's Sirius program was created to fabricate and test fuels for space nuclear propulsion, specifically to determine their performance under prototypical startup conditions. The Sirius project featured 4 test capsules, Sirius-1 featured uranium nitride fuel dispersed in a matrix of tungsten and rhenium, while Sirius-2A, -2B, and -3 featured uranium nitride-molybdenum-tungsten fuel (UN-Mo-W). This study discusses the Sirius-2A and -2B irradiation experiments at the Idaho National Laboratory, specifically their performance under irradiation at the Transient Reactor Test Facility. It was found that the fuel samples overall did not exhibit significant cracking, though the Sirius-2A fuel did have one large crack on the surface of the fuel. There was minimal hydrogen absorption in the samples, though it is unknown if the absorption occurred during irradiation or during fabrication. Mechanical testing indicated that the UN fuel demonstrated ceramic behavior as expected, and the Mo/W matrix demonstrated linear elastic behavior to failure.

Lumped Parameter Analysis of Autogenous Propellant Pressurization System for Nuclear Thermal Rocket

This work proposes a new propellant management configuration for an ammonia-fueled nuclear thermal propulsion system. The suggested configuration maximizes the advantage deriving from the autogenous pressurization of ammonia by exploiting the thermal power lost by the nuclear reactor toward the vacuum space due to leaking radiation. In this layout, a tank containing ammonia in saturated conditions is placed near the nuclear reactor and receives an input thermal power proportional to the dose of gamma rays and neutrons absorbed by the tank walls and the ammonia itself. Such a thermal power accelerates the vaporization process of the saturated ammonia, thus increasing the pressure in the tank. A pressure regulator valve exploits this overpressure to pressurize the ammonia propellant contained in a run tank to the level required by the mission by connecting the two ammonia volumes. The pressure achieved inside the run tank pushes the propellant with an adequate mass flow rate inside the nuclear reactor. The developed lumped parameter analysis shows how this propellant management system can provide a constant mass flow to the nuclear reactor without using a turbopump assembly. Moreover, it is shown how the proposed concept allows for a reduction in the radiation shield mass.

The Bimodal Epithermal Astronuclear Reactor (BEAR): A Long-Lived, Universal Space Nuclear Powerplant

A neutronic model of a Nuclear Thermal Propulsion reactor based on the NERVA Peewee design using High Assay Low Enriched Uranium (HALEU) fuel in a Zirconium Carbide matrix has been thoroughly characterised and continuously modified over the last few years at the CSNR. The initial design (Emu, for the flightless bird) created in the summer of 2020 and presented at the NETS 2021 conference, has undergone significant adjustments in this time. Two major investigations have been undertaken and evolved over time; one into the requisites for the provision of electrical power to the propelled spacecraft via the circulation of a noble gas mixture in a closed Brayton cycle, and another into the use of planetary regolith as a reflector, reducing neutron leakage at what would otherwise be the reactor’s end of life in vacuo, rekindling it for an extended period of time as well as providing a very easy ultimate disposal solution, replacing the control drums with rods. It was determined that there are no insurmountable obstacles to performance in either mode.

Hot hydrogen testing of Mo30W matrix surrogate cermets

High temperature hydrogen exposure is one of the most challenging material issues for nuclear thermal propulsion (NTP) fuel development. Under legacy NTP programs, ceramic-metallic (cermet) fuel forms with a refractory metal matrix and dispersed uranium dioxide (UO2) fuel particles were developed and showed promising performance following hot hydrogen testing. However, since the conclusion of those programs, established fabrication techniques, material feedstocks, and the ability to use highly enriched have been reduced or lost all together. In this study, a cermet consisting of a solid solution alloy of molybdenum with 30 wt percent tungsten (Mo30W) was fabricated using spark plasma sintering. Fabrication process parameters were selected to optimize the cermet microstructure using lessons learned from historic NTP programs. Yttria stabilized zirconia particles (50 to 70 % volumetric loading) were used as a fuel particle surrogate. To evaluate whether as-fabricated microstructures exhibited similar resilience as legacy cermet fuels to a hot hydrogen environment, samples were exposed to hot flowing hydrogen from 1920 to 2500 °C (∼2290 to 2770 K). The cermets performed well with minimal mass loss, minor to no cracking, and good retention of internal surrogate particles. Based on these findings, recommendations for future studies with Mo30W-UO2 cermets as an NTP fuel form are provided.

Steady-state thermal–hydraulic analysis of an NTP reactor core based on the porous medium approach

Nuclear thermal propulsion (NTP) is a promising advanced technology which has attracted wide attention in recent years. The reactor core is an essential component of an NTP system and the corresponding thermal–hydraulic analysis is necessary. In this study, the porous medium approach was applied to the simulation of a two-pass NTP reactor core which consists of the porous prismatic cermet fuel elements. The thermodynamic property models of hydrogen and the fuel element materials were implemented, as well as the empirical correlations of the heat transfer coefficient and the friction factor. The three-dimensional simulation of a single fuel element was carried out and the results were compared against another code. The code-to-code comparison verified the applicability of the porous medium approach. The three-dimensional model of the two-pass NTP reactor core was established and the steady-state simulation was carried out. The distribution patterns of the parameters are determined by the thermal–hydraulic characteristics of the reactor core, including the nonuniform heat release, contact heat conduction and folded-flow scheme. The full-core heat-flow adaptability analysis is realized, which provides a reference for the thermal–hydraulic safety analysis of the NTP reactor.

Effect of ultra-high temperature treatment on the rolled pure molybdenum for nuclear thermal propulsion

Molybdenum (Mo)-based uranium dioxide (UO2) fuel is considered one of the most promising fuel forms for deep-space exploration. The rolling process has been proposed for the preparation of Mo matrix due to its excellent performance. However, there has been limited research on the ultra-high temperature stability of Mo matrix used in nuclear thermal propulsion (NTP). Therefore, in this study, annealing experiments were carried out on rolled Mo plates within a relatively wide temperature range of 1200 to 2300 °C for 1 h to investigate the effects of temperature on the phase composition, microstructure, and mechanical property. The results showed that, despite the body-centered cubic structure of the Mo plates remaining unchanged without undergoing phase transition after high-temperatures, the intensity of the crystal diffraction peaks experienced a significant increase. The preferred growth orientation shifted from the (110) direction at lower temperatures to the (200) direction at 2300 °C. The grain shape of the rolled Mo plates changed from elongated to equiaxed after annealing, and the grain size increased from the initial 1 μm to tens of microns. No discernible surface pores or fracture bubbles were observed when annealed at 1200 to 2100 °C. However, subsequent annealing at 2300 °C revealed a proliferation of polyhedral bubbles, ranging in size from tens of nanometers to several microns, distributed both in grain boundaries and within grains. Nevertheless, there is no apparent change in the density from 1200 to 2300 °C. Bright-field TEM micrographs show that after annealing at 2300 °C, the dislocation density is significantly reduced, and the dislocation shape evolves from screw dislocation to long and straight line. In addition, the Vickers hardness value of the rolled Mo plates undergoes a significant decrease, from 247.1 to 199.6 Hv, after annealing at 1200 °C. Subsequently, in the temperature range of 1200 to 2300 °C, a gradual decline in the hardness value is observed. This work will deepen our understanding of the high-temperature resistance of rolled Mo plates and provide data to support their applications in ultra-high-temperature conditions.

Vaporization in Liquid-Core Nuclear Thermal Rockets

This work provides an overview of vaporization phenomena in liquid-core nuclear thermal rockets to aid in unifying modeling assumptions by characterizing the impact of vaporization. Historical variations in vapor modeling have led to reported specific impulses ranging from 1000 to 2000s following model refinements in the 1960s and the omission of vaporization consideration post 1990. The only preexisting vapor transport study to not rely on the Reynold’s analogy was performed for the radiator design. This paper performs the first such analysis for the bubbler design, given its renewed interest. Consideration is given to potential centrifugal species separation, proposed fuel mixtures, interplay between chamber pressure and vaporization, impacts on propellant thermoproperties, evaporative cooling, power requirements for vapor separation, and the complexity of hydrocarbon interactions under the proposed chamber conditions. The bubbler design is shown to be well approximated as fully saturated, whereas the radiator and droplet designs may be configured to reduce vaporization to 20% saturation. Characterization of vaporization within liquid cores is essential to early design decisions, such as overall archetype, defining the operating point, and liquid media composition. For a threshold temperature corresponding to around 10% of the propellant being vapor by mass, further increases in temperature begin reducing the specific impulse. Below this threshold temperature, the impact of vaporization on thermal modeling with the chamber is likely negligible. Above this threshold temperature, higher-fidelity thermal modeling will be required, and the work required to separate the vapor begins exceeding 100 kW/(kgH/s). Nozzle kinetics are applied to liquid cores for the first time, showing that specific impulses increase with increasing chamber pressure contrary to historical findings. Historically, modeled uranium-zirconium-carbon mixtures are predicted to reach a specific impulse up to 1260 s. A uranium-tungsten-carbon fuel mixture yielded the best theoretical specific impulse of slightly under 1400 s; however, no nucleonic or thermal analysis has been conducted with this fuel composition.

Demonstrating autonomous controls on hardware test beds is a necessity for successful missions to Mars and beyond

NASA and the Department of Defense are planning for a mission to Mars in the 2030s–2040s using nuclear thermal propulsion (NTP). NTP uses a nuclear reactor to heat flowing hydrogen and create thrust. A serious concern for crewed and uncrewed missions to Mars is the loss of reactor control. The reactor startup and initial rocket impulse are initiated in cislunar or near-earth orbital regions; therefore, radio communications between ground control and the NTP engine should occur in real time. However, radio communications can take more than 20 min, depending on planet positions, to reach Mars orbiters from ground control. To address this delay, local autonomous controls are implemented onboard the NTP engine to ensure acceptable operation. However, autonomous controls have not been demonstrated or implemented in research or power reactor contexts because of safety and reliability concerns. To enable autonomous controls development, demonstration, and validation, Oak Ridge National Laboratory has created a nonnuclear hardware-in-the-loop test bed. Sensors throughout the test bed relay system status and hardware response to the user control algorithm, including measurements of temperature, flow, pressure of a loop, control drum position, and drum speed. This paper discusses the development of this facility and user accessibility.

Numerical Study of Porous Cylindrical Containment for the Fuel Element of a Centrifugal Nuclear Thermal Rocket

For deep-space propulsion and interplanetary exploration, the centrifugal nuclear thermal rocket (CNTR) has the ability to achieve a very high specific impulse (Isp) metric beyond that of conventional chemical rockets or solid-core nuclear thermal propulsion systems. The high Isp allows the rocket to use less propellant or achieve a higher velocity for shorter transit times. However, the cylindrical containment structure of a CNTR fuel element encounters extreme conditions, as it houses molten uranium at temperatures exceeding 1408 K, leading to challenges such as dissolution, chemical reactions, and thermal stresses that conventional materials struggle to withstand. This study aims to address this issue by analyzing appropriate materials for constructing the cylindrical containment component. The operating conditions of the annular porous medium that confines the liquid uranium in the centrifugal fuel element are simulated by conducting a comprehensive one-dimensional numerical analysis using a range of candidate porous materials, including Mo, W, zirconium carbide, and silicon carbide. The porous structure facilitates the flow of the hydrogen propellant into the internal molten uranium section, where it gains significant thermal energy while simultaneously cooling the cylinder. The containment cylinder has an internal temperature of 1478.1 K, exceeding the melting point of uranium, while the external gas temperature of the hydrogen propellant is much lower. This temperature difference induces significant thermal stresses in the cylinder. The porous containment cylinder made from molybdenum was able to maintain elastic deformation throughout the thickness of the cylinder, showcasing its ability to handle these extreme thermal stress conditions. Tungsten, on the other hand, experienced plastic deformation at the cylinder’s edges and elastic deformation through the middle radial locations. In contrast, the stresses experienced by the ceramic materials far exceeded their failure stress values, leading to brittle failure. These findings will help with the refinement of the CNTR design, edging it closer to practical implementation.

Integrated Steady-State System Package for Nuclear Thermal Propulsion Analysis Using Multi-Dimensional Thermal Hydraulics and Dimensionless Turbopump Treatment

Integrated Steady-State System Package for Nuclear Thermal Propulsion Analysis Using Multi-Dimensional Thermal Hydraulics and Dimensionless Turbopump Treatment

Exploring the Feasibility of Kuiper Belt Missions Supported by Centrifugal Nuclear Thermal Propulsion

Missions to the Kuiper belt have previously been carried out only as flybys and with very small payloads. Investigating launch windows for Kuiper belt missions supported by centrifugal nuclear thermal propulsion (CNTP) contributes to defining its operational use case. Results indicate that CNTP enables rendezvous missions to the Kuiper belt, both with direct transfer trajectories or planetary gravity assist trajectories, although there are many challenges to making these mission architectures feasible. The direct trajectories have transfer times of roughly 14 to 16 years while combining CNTP with gravity assists from Jupiter could lower transfer time to as low as 10 to 12 years to Kuiper belt objects such as Pluto and Quaoar. These missions are then shown to inform the architecture of the CNTP injection stage vehicle, which can be supported by heavy and super-heavy commercial launch vehicles with a single launch. Last, drawbacks of the mission and vehicle architectures are given that impose limits on the use case for CNTP on these missions.

Developing a Three-Dimensional Bubble Dynamics Model for Centrifugal Nuclear Thermal Propulsion

Nuclear thermal propulsion (NTP) is a highly efficient type of propulsion system that could yield a specific impulse up to 800 s. High-performance NTP allows for possibilities of various mission types from flyby to interplanetary flight, manned or unmanned. Centrifugal nuclear thermal propulsion (CNTP) is a type of NTP that uses liquid fissile material instead of a solid core in a standard reactor configuration. CNTP requires an interaction between liquid fissile fuel and gaseous propellant to perform heat transfer and generate thrust in a spinning enclosure. Because of the liquid state of the fissile fuel, it is essential to monitor bubble behavior and trajectory. For example, the bubble dwell time within the liquid and hydrogen flow rates can influence the liquid volume fraction and volume, which in turn can affect reactivity. A model is developed and presented that, as it matures, will predict the bubbles’ behavior based on the various types of liquid and gas and their corresponding properties as well as assist in cross-validation with the concurring experiment. The model primarily utilizes the smooth particle hydrodynamics (SPH) approach to simulate the liquid and gas interaction within the design space. The physics include buoyancy and drag on the bubbles. The pressure field within the liquid is modeled using hydrostatics, in which centrifugal motion introduces a radial gradient. Boundary conditions are devised to confine the liquid and gas within the enclosure. In this paper, we present the progress to date in two centrifugal fuel element–relevant subscale devices, the so-called “Antfarm” and “Blender II.” Antfarm is a static stage in a rectangular enclosure where buoyancy is purely gravity driven. Blender II refers to a rectangular enclosure that spins at up to 7000 rpm. Both Antfarm and Blender II are experiments that provide important data for validation against our SPH model. Two liquid and gas combinations are modeled in the present study using the Antfarm setup: water-air, and Galinstan-Nitrogen. The bubbles’ behavior was comparable to the experiment, and the velocity was at the same order of magnitude. The liquid simulation performed for the Blender II model shows that the pressure gradient within the liquid in the radial direction matches that predicted analytically from the centrifugal acceleration. The Blender II liquid model awaits experimental data for validation and verification.

Effect of finned configuration of circular tube based on fluid-structure coupling on hydrogen flow characteristics

Nuclear thermal propulsion has the characteristics of high specific impulse, large thrust, green and efficient, and is the primary choice of manned deep space exploration propulsion system. Because the circulating flow of propellant hydrogen in the internal pipeline of the engine is always affected by the harsh environment of high temperature and high pressure, it is very important to study the flow characteristics of hydrogen in the pipeline, the heat exchange characteristics of hydrogen and pipeline and the deformation characteristics of pipeline under heat stress. In this paper, the flow phenomena of hot hydrogen in round tubes in a laminar flow state were analyzed by numerical simulation with COMSOL Mulitiphisics6.0, and the fluid-structure coupling between hot hydrogen fluid and pipe was investigated by a multi-physics field. It is concluded that the smoother the inner wall of the pipeline is, the smaller the hydrogen flow velocity is, the smaller the surface pressure of the pipe wall is, and the smaller the fluid extrusion and impact deformation of the pipe in the typical pipe with the inner fin is. It inspires studying the application of hot hydrogen flow and pipeline configuration in nuclear thermal rocket engines, including improving heat transfer energy and uniformity.

Design, preparation and diffusion research of the W coating used in W-based cermet for nuclear thermal propulsion

The fissile fuel agglomeration and loss in the tungsten (W) based ceramic-metallic (cermet) fuel element are the major challenges for application in nuclear thermal propulsion (NTP). The previous research indicated that a W coating deposited on the surface of uranium dioxide (UO2) by the fluidized bed chemical vapor deposition (CVD) technique is an efficient way to achieve the uniform distribution of fuel particles and reduce fuel loss simultaneously. However, the W layers' resulting thicknesses differ, which is not conducive to achieving high fuel loading. Therefore, in this study, the appropriate thickness of the W coating is first investigated by COMSOL Multiphysics modeling. Through contrastive analyses, the W layer with a thickness of about 5–10 μm is considered to be the optimal choice, where the maximum loading capacity of the fissile fuel can be achieved with a volume fraction of about 60%. Based on the above calculations, the CVD technique is employed to produce the optimized thickness of the W layer by precisely adjusting the deposition time at 900 °C. This approach is taken after determining the critical influencing factor that governs the performance of the W layer. Experimental results show that the highly dense W layer presents a uniform thickness and fine-grained size. As the only structural and protective layer, the application of this W coating provides a barrier for blocking the inward diffusion of hydrogen (H2) to stop the detrimental chemical reactions between UO2 and H2, thus lowering fuel loss. The spark plasma sintering (SPS) demonstrates that zirconium dioxide (ZrO2) particles as the surrogate for UO2 exhibit a homogeneous distribution without any direct contact in the W matrix because of the presence of the W coating. Despite the fact that the interface between the W coating and the W matrix exhibits remarkable bonding, making it difficult to clearly discern the boundaries due to surface diffusion and atomic binding, the existence and integrity of the W coating after high-temperature sintering is unequivocally confirmed through interface TEM observations and the uniform dispersion of coated particles within the W-Y2O3 matrix. Furthermore, the fine-grained W coating has been demonstrated to effectively hinder the diffusion of H2 isotopes through permeation and diffusion tests. This comprehensive study not only aids in establishing the optimal geometrical thickness for the W coating but also underscores its pivotal role in W-based cermet fuels for NTP system.

A Continuous Flow Boiling Curve in the Heating Configuration Based on New Cryogenic Universal Correlations

To enable the design of future in-space cryogenic propellant vehicles such as Lunar and Martian ascent and descent stages, fuel depots, and nuclear thermal propulsion systems, high-accuracy models of various phases of the propellant transfer process are required. This paper focuses on modeling steady-state flow boiling through the transfer line that connects a propellant tank to an engine or customer receiver tank, which is required to set limits on the allowable heat flux into the line. Using the largest-ever collection of available cryogenic heated tube data, universal cryogenic flow boiling correlations were recently developed for various regimes of the boiling curve and their transition points. However, to model flow boiling in heated tubes, these individual correlations must be patched together to provide a continuous predictive curve of wall superheat as a function of preponderant parameters. This paper provides an overview of the individual flow boiling correlations along with the logic and rationale for patching the correlations together to produce a single continuous boiling curve. Resulting flow boiling curves are presented for different possible permutations of independent inlet and flow conditions for illustration.

Resilience of uranium mononitride/zirconium carbide composites and uranium-zirconium carbonitride in hot hydrogen for nuclear thermal propulsion

Nuclear thermal rockets require fuels capable of withstanding flowing hydrogen propellant up to 3200 K. Presently, there has not been a fuel type that reliably operates at these conditions. A promising candidate anticipated to endure this demanding environment is a ceramic-ceramic composite comprising of uranium mononitride and zirconium carbide. This investigation assesses the behavior and resilience of variations of this composite and its resultant homogenized form (uranium-zirconium carbonitride) under two hot hydrogen conditions (2273 K and 3000 K). The findings revealed that composites that homogenize into UZrCN exhibit superior structural integrity in hydrogen compared to heterogeneous counterparts. Consequently, this study underscores the potential of homogenized uranium-zirconium carbonitride for enhanced performance in nuclear thermal propulsion applications.

Hot Hydrogen Testing of W-Coated UN Kernels in a Mo30W Matrix

Ceramic uranium mononitride (UN) is being considered as a reactor fuel for nuclear thermal propulsion. To avoid or reduce the dissociation of UN at the high temperatures needed, embedding it in a metallic matrix (cermet) has been proposed. To assess the viability of this concept, hot hydrogen testing of tungsten-coated UN kernels embedded in a Mo-30 wt% W (Mo30W) alloy matrix has been performed at temperatures from 1800°C to 2300°C. Both the isolated kernels and kernels consolidated by spark plasma sintering in the Mo30W matrix were tested. In addition to direct observations and mass loss measurements, the samples were analyzed by X-ray diffraction (XRD) and scanning electron microscopy (SEM)/energy dispersive X-ray spectroscopy (EDS) after each run. The decomposition of UN started at 1800°C despite the coating and matrix, and increased at 2000°C. Uranium seeped through the tungsten grain boundaries of the coating at all temperatures. The consolidated sample expanded irregularly at 2000°C through the formation of voids, and SEM/EDS analysis showed uranium-containing veins in the matrix consisting of U2Mo according to the XRD data. The observed pore generation at 2000°C was explained by the formation of water vapor from residual oxides and diffused hydrogen. At 2200°C and above, both the kernels and the consolidated samples melted through the formation of uranium or low–melting point uranium-molybdenum alloys.

Characterization of fluidized bed chemical vapor deposition ZrC coatings on PyC/YSZ kernels deposited under differing conditions

Coated fuel particle architectures with ZrC coatings are candidate fuels for advanced power reactors and space nuclear propulsion (SNP) concepts. Owing to its relevance to SNP, the composition, microstructure, and mechanical properties of eight ZrC coatings prepared by fluidized bed chemical vapor deposition were evaluated. Evaluation by SEM and EBSD showed that all grains were columnar. Across the various examined samples, minor axis diameters varied between 0.3 and 1.1 μm, and major axis diameters varied between 0.4 and 2.3 μm. Major and minor diameters increased with thickness particularly at higher deposition temperatures in which the major grain axis (from an ellipse fit to the grain shape) increased by 2.5 μm over the entire coating. Coatings with higher reactive gas flows and Zr/C concentrations closer to 1 were observed to contain nanocrystalline graphite deposits. Reactive gas flow doubling led to increases in coating thickness from around 10–15 μm to around 22–27 μm.

Multiphysics and multi-scale design and analysis of nuclear thermal propulsion pellet bed reactor based on spherical cermet fuel element

Due to the advantages of high thrust and high specific impulse, nuclear thermal propulsion (NTP) is a preferred near-term rocket engine technology for future Mars missions. This article introduces the reactor design based on spherical cermet fuel element, with the aim of studying the feasibility of applying this new fuel in NTP system. First, design requirements are determined based on the unique features of NTP. In terms of neutronics, the effect of rocket launch accidents and fuel loss caused by high-temperature hydrogen are considered. Thermal hydraulics refer to the requirements of NASA's Mars Design Reference Architecture 5.0 (DRA 5.0) for impulse, thrust, and thrust to weight ratio. Second, based on the requirements above, the optimal reactor configuration parameters are obtained with the goal of maximizing the thrust to weight ratio: reactor radius 51.1 cm, height 106 cm, aspect ratio 1.04, reflector 13 cm, and reactor mass 3940 kg. Finally, the performances of the optimal reactor design are analyzed. The neutron spectrum shows a characteristic of fast neutron spectrum, and fuel loss analysis shows that the core can withstand a fuel mass loss of about 4 wt%. In addition, the core axial and radial power distribution are calculated, with an overall power peak factor of 1.26. The thermal hydraulic calculation provides a range of core operating conditions that satisfy the design requirements of DRA 5.0 and temperature, with power of 620 ∼ 950 MW and flow rate of 16 ∼ 24 kg/s. Compared with the ANL200/2000 reactor using prismatic cermet, it has the advantage of higher specific impulse. The main disadvantage is the large pressure drop, which prevents it from being applied to the scenarios of high power and high thrust. Therefore, for NTP reactor using spherical cermet elements, it should be applied to low to medium power scenarios.

Assessment of Electric-Pump-Fed Nuclear Thermal Propulsion for Near-Term Missions

This paper describes an analysis of the viability of electric-pump-fed nuclear thermal propulsion (EPFS NTP) for space missions. EPFS is an alternative to the turbopumps typically employed by high-impulse rocket engines, but it is not yet clear if they are suitable for this application. The method employed by this analysis utilizes the rocket equation to determine whether EPFS NTP spacecraft are capable of delivering payloads to Jupiter and Saturn, subject to the mass and geometry constraints imposed by near-term launch vehicles. The analysis found that 7.5 klbf EPFS NTP can deliver up to 3000 kg to Jupiter and 2000 kg to Saturn, and that spacecraft with higher thrust EPFS NTP engines deliver smaller payloads. This analysis provides confidence that EPFS NTP can enable a space mission of interest, and therefore EPFS can be considered as an alternative to turbomachinery for near-term NTP engine development.