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A feasibility study on low-enriched uranium fuel for nuclear thermal rockets – I: Reactivity potential
Abstract
This paper is concerned with the possibility of implementing Low-Enriched Uranium (LEU) fuels in NERVA (Nuclear Engine for Rocket Vehicle Application) type Nuclear Thermal Rockets (NTR). It presents a detailed study of the requirements for implementing LEU in a NERVA type NTR revolving around the softening of the neutron spectrum. This work has been done to demonstrate the possibility to enhance the proliferation resistance of these nuclear space systems in the hope of fostering the groundwork for their eventual commercial use. The necessity of thermalizing the neutron spectrum is discussed along with a detailed study of various methods by which this can be achieved. The effect of minimizing non fission neutron loss through the selection of appropriate structural materials is then presented along with a cursory study of various options. Finally a reference conceptual reactor configuration is proposed and compared with previous NTR designs where its relative shortcomings and obvious advantages are presented and discussed. It is concluded that an LEU-NTR is not only a definite possibility, but also offers distinct advantages over existing Highly-Enriched Uranium (HEU) NTR designs, particularly the increased proliferation resistance of the system and the reduction of the fuel cost and development schedule. Finally, some suggestions are made for future work that would enable the progression of the current conceptual design to a more finalized reactor design.
... Furthermore, low enriched uranium (U-235 = 19.75%) will be utilized to achieve high thrust levels, greater power, shorter traveling time, and larger carrying payloads, eliminating any security-related issues that can arise from proliferation risks [4]. Therefore, scientists are confident that NTRs will fulfill the need for a propulsion system with a high Isp and thrustThe purpose of the present investigation is to develop an NTR design that can shorten the trip duration and increase the payload compared to any chemical rocket. ...
... However, ceramic fuels are preferable due to their availability and cost. Moreover, their properties match the thermal rocket properties of operating at high temperatures [4]. ...
... It is also stable throughout a large temperature range that can reach up to 3473 K. Graphite does not melt, however, it changes from solid to vapor state directly at around 3923 K [7]. The primary drawbacks are the opportunity of oxidation in the presence of air, displacement in the crystal dimension under the effect of radiation from the reactor and being prone to non-negligible hydrogen corrosion [4]. ...
... Meanwhile, academic research underway at the Center for Space Nuclear Research and their collaborators was focused on the viability of High-Assay Low-Enriched Uranium (HALEU) as the fuel for an NTR reactor. Research results published in 2015 It is concluded that a HALEU fueled NTR is not only a definite possibility, but also offers distinct advantages over existing HEU NTR designs, particularly in regards to the increased proliferation resistance of the system and the reduction of the fuel cost and development schedule [28] . Although the research conclusively affirmed the feasibility of a HALEU fueled NTR with a similar core mass and volume as the HEU fueled NTR, the core did not operate at the desired performance with a specific impulse of only 775 s. ...
This paper describes the current research and development effort s currently underway within the United States on Nuclear Thermal Propulsion (NTP), with a particular focus on the Demonstration Rocket for Agile Cislunar Operations (DRACO) project, a joint effort of the United States Defense Advanced Projects Agency and the National Aeronautics and Space Administration. However, to put the DRACO project into context, the prior United States’ prior effort s on NTR are described and the foundation those efforts pro- vided to enable DRACO. The impact of NTP propulsion on both human and scientific exploration of the Solar System will also be discussed. And finally, the topic of advanced NTP propulsion will be addressed, including liquid fuel NTP engines.
... A few decades of the cold war between the United States and the Soviet Union had witnessed the amount of mature and effective numerical and experimental work (Belair et al., 2013;Khatry et al., 2019;Graham, 2020) on HEU design. However, recent efforts focus on designing a feasible engine that relies on LEU fuel due to its lower cost and nuclear proliferation risk (Venneri and KIM, 2015a;Venneri and KIM, 2015b;Gates et al., 2018). In LEU design, moderator assembly is employed to cooperate with fuel assembly to support the whole structure and take some heat away. ...
The design of a nuclear thermal propulsion (NTP) reactor based on low-enriched uranium (LEU) requires additional moderator elements in the core to physically meet the critical requirements. This design softens the core energy spectrum and can provide more thermal neutrons for the fission reaction, but the heat transfer characteristics between the fuel and moderator assembly are more complex. Aiming at the typical LEU unit design, the heat transfer mathematical model is established using the principle of heat flow diversion and superposition. The model adopts the heat transfer relationship based on STAR-CCM+ simulation rather than the empirical expression used in the past literature to improve the applicability of the model. The heat transfer coefficients in the proposed model are evaluated under different Reynolds numbers and thermal power. The deviations between the proposed model and CFD simulation are analyzed. The results show that the calculation of the heat transfer coefficient between the proposed model and the CFD simulation maintains a good consistency, most of which are within 10%. It may provide a reliable and conservative temperature estimation model for future LEU core design.
... Fortunately, NTP has sprung to life and attracted much interest in recent years. The experience and achievements from NERVA program made its core become the foundation of many research and the prototype of some new designs by Venneri and Kim (2015), Houts and Mitchell (2016), Gates et al. (2018), and so on. In addition, the use of clustered NERVA engines was also addressed in the DRA 5.0 to provide an "engine-out" capability that could reduce the mission risk and increase crew safety. ...
Nuclear Thermal Propulsion (NTP) provides the leading potential due to its high thrust, improved specific impulse and long accumulated lifetime in the manned deep space exploration. Particle Bed Reactor (PBR) is the most efficient one among all proposed NTP concepts, which employs a radial flow pattern to reduce the pressure drop and fuel particles of high surface-to-volume ratio to increase the heat removal capability. The technical challenges, however, have arisen with the enhanced performance of PBR, such as the thermal hydraulic design. In this paper, a procedure based on the Non-dominated Sorting Genetic Algorithm with elitisms approach (NSGA-2) and uncertainty quantification is developed to prepare an optimal fuel element for PBR with regard to the thermal performance. The procedure has been verified through three cases, in which the thermal performance of all fuel elements has improved a lot after the optimization process.
... Maximum the exhaust core temperature, maximum will be the specific impulse value as shown in given equation. [4] √( )( ) Where, γ is adiabatic constant i.e. / R is gas constant. T is exhaust temperature of the propellant. ...
Many space agencies like NASA, SPACE-X have promised to send humans into the red planet in future. So, considering their project of mars colonization, nuclear rocket propulsion would be the better option. Replacing chemical rockets by nuclear rockets may reduce the mission duration and also can reduce the mass of the propellant used. In chemical rockets, propellant releases energy through combustion but in case of nuclear rockets, propellant i.e. hydrogen is heated up in controlled fission reaction in nuclear reactor inside the rocket engine. Specific impulse of the nuclear rocket is greater than chemical rocket. This helps in providing gigantic thrust as a result mission duration is decreased. The challenging parameter of increasing specific impulse is solved by maximizing specific impulse which is done by increasing the exhaust core temperature. The fuel is selected in such a way so that the exhaust temperature would be obtained. The (U, Zr) C –graphite fuel is selected because it has high uranium density and melting point is equivalent to exhaust core temperature which is sufficient enough to enhance the reactivity of the fissile material and thus to increase the rocket performance. A mathematical analysis shows that the percentage of mass of propellant used in mars mission will be lesser than the chemical rockets because the specific impulse is expected to be more in nuclear propulsion. The specific impulse obtained from the CFD Analysis of rocket nozzle is 979 sec with exit velocity of 9604m/s.
... With these non-negligible barriers to development, in addition to the recent push for removing HEU from all non-military applications, HEU fueled NTP systems had reached a stale-mate of sorts in terms of viability. With the recent realization that LEU fuel is in fact a viable option [7] [8], the majority of these issues have been resolved. LEU fuel requires orders of magnitude lower costs in terms of security and safety. ...
This paper presents an overview of the latest developments in the designing of low-enriched uranium nuclear thermal rockets (LEU-NTR). The concept is first introduced and explained in the context of human exploration of Mars and the development time frames associated with current and future research. The need for LEU fuel is established and the process by which LEU fuel is introduced is described. The importance of the moderator to fuel ratio is explained and the size limitations associated with the cooling requirements of the core are detailed. Once the general performance and neutronic requirements have been established, a series of design issues are identified including current trends for their successful resolution. These include the minimization of active reactivity control in reactor operation and the resolution of the full-submersion criticality accident. The implementation of spectral shift absorbers, rapid depletion neutron poisons, specialized axial and radial reflectors, and enhanced core hydrogen worth are briefly explored and compared. Following this overview of limitations and design requirements for LEU-NTRs, the possibility for different thrust levels is explored. Here, a comparison of two thrust classes is provided along with a development of requirements that govern the minimum core size for each thrust class.
... Each fuel type was studied with and without external moderating elements. To determine the effect of moderation on fuel reactivity, fuels were arranged in 1:1, 1:2, 1:3, and 2:1 moderator ratios with reference ZrH 1.8 containing moderator tie-tube elements optimized in previous studies [1,46]. Figure 3 shows the infinite lattice configurations for the 1:0, 1:1, 1:2, 1:3, and 2:1 fuel to moderator (F:M) ratios. ...
Nuclear thermal propulsion (NTP) is a non-chemical propulsion technology capable of high specific impulse (850-900 s) and high inherent thrust (100 – 2,200 kN) extensively tested in the United States and former Soviet Union. Most recent NTP development efforts have focused on recapturing fuel production capabilities and optimizing small thrust engine designs based on historic U.S. NTP development efforts. However, recent fuel production efforts have shown that fuel cannot be identically recaptured and that future development could benefit from modern manufacturing technologies, as demonstrated by the production of high density cermet fuel compacts by spark plasma sintering. Further, neutronic analyses have shown that low enriched uranium (LEU) fueled nuclear thermal rockets (NTRs)can be designed with legacy fuel systems including manufacture limits learned from past development programs. LEU engine designs are expected to significantly reduce the high maintenance cost and perceived political hurdles of developing NTP systems traditionally associated with high enriched uranium (HEU) fuel systems. All in all, these findings warrant a review of relevant materials to support future NTP design efforts. The purpose of this review is to survey a wide range of high temperature structural materials applicable to NTP fuel systems. Materials will be characterized based upon their limiting thermal-mechanical properties, chemical compatibility, neutronic performance, and known manufacturability. The focus of this paper will be largely on legacy fuel system designs; however, high temperature structural fuel matrix materials explored through recent terrestrial nuclear fuel development programs will be assessed.
Nuclear thermal propulsion (NTP) technology has characteristics of high power and high specific impulse. In particular, CERMET fuel reactors have received a lot of research due to simple form and structure. A general CERMET fuel reactor model is established by using three-dimensional Monte Carlo code and commercial CFD program. After nuclear thermal coupling optimization, three partition schemes of baffled reactor were proposed to reduce the temperature difference between the wall and the coolant and lower the relative spike factor in the inner region. A feasible design of redundant control system was then proposed, with a cold clean excess reactivity greater than 7.94, where there was a high control accuracy of 6.7 pcm/mm. The reactor finally realized safety of the core temperature lower than 3000 K under the conditions of total exit temperature of 2650 K and thrust of 25000 lb.
Nuclear Thermal Propulsion (NTP) systems are actively being developed for future crewed missions to Mars. NTP systems excel in missions where both high thrust and high specific impulse are required, but modern NTP systems currently do not have a Technology Readiness Level (TRL) high enough for use in crewed space exploration. TRLs are used to demonstrate the level of rigor with which a component/system has been tested/demonstrated for its intended use. While space systems technology in general must be qualified as a unit, nuclear technology must be first demonstrated to meet qualification level requirements both at the fuel (component) level and the reactor (subsystem) level. In this paper, historic NTP development programs are surveyed to identify a testing and development strategy that can be effectively implemented to allow for NTP reactor development. Based on this strategy, required facilities to enable such activities are identified. Current domestic experimental capabilities to support NTP qualification are limited to separate effects testing of individual components. Separate effects testing is found extensively in historic NTP development efforts but is not sufficient for full fuel and reactor qualification. Combined effects testing allows for an accurate assessment of fuel performance but is not achievable for NTP conditions in existing facilities. Assessment of historic development programs suggests that an intermediate, subscale test facility is necessary to advance NTP TRLs. A solution to meet this need is proposed, namely the Subscale Maturation of Advanced Reactor Technologies (SMART) facility. SMART will mitigate risk to NTP development by enabling performance and reactor physics demonstrations of NTP subsystems. A SMART facility could be built by modifying existing nuclear test facilities, which may potentially enable schedule and cost savings. To pursue reactor qualification beyond the subscale, a new ground test facility will be necessary. This ground test facility should be developed concurrently with SMART to allow for the facility to be operational in time for expedited NTP engine demonstration.
In order to minimise the mass of a 1MWe LEU space fission power system design, a rapid method of carrying out neutronic and thermal hydraulic analysis for the reactor is required. A two-stage deterministic analysis routine has been constructed using a core-plane method of characteristics calculation followed by a full-core SP3 calculation, within the ANSWERS© code WIMS11. This is compared to a faster route that skips the core-plane calculation and also the Monte Carlo code Serpent. Results suggest sufficiently-good agreement for the WIMS-based methods to be useful in a full system mass-minimising optimisation routine. The neutronics scheme is coupled with a simple sub-channel thermal hydraulics code with a finite-element heat conduction model for the fuel block.
Nuclear thermal propulsion is a key technology for long-range spaceflight, as demonstrated by the rising interest from space agencies, especially NASA. A preliminary design involves the exploration of many reactor configurations, until a configuration that meets all the design requirements is found. Trade-offs among different design fields, such as neutronics, thermal-hydraulics, safety and rocket performances are unavoidable. Design engineers have to run a large amount of simulations because of the multitude of possible reactor configurations and the different analyses that must be carried out. The present work investigates an optimization approach for the design of a LEU reactor with CERMET fuel, to find a configuration that fulfils neutronics, safety and NASA requirements for a nuclear thermal rocket. A neutronics analysis constitutes the basis of the work, but accidental scenarios and thermal-hydraulics analysis are performed as well, to assess reactor safety and rocket performances. The optimization procedure simulates different configurations and gradually reduces the design space to be explored, until an optimal region is found. In this way, unnecessary simulations are avoided. A Python script is developed to handle the whole analysis, from pre-processing to post-processing, including the integration between the neutronic code Serpent and MATLAB®. Once the optimal region is found, the most promising configurations are identified by comparing different performance metrics retrieved from the neutronics analysis. A safety analysis that simulates the reactor behaviour in four accidental scenarios is carried out on this smaller group of reactor configurations. Finally, reactors that fulfil safety requirements undergo a thermal-hydraulic analysis, which verifies that thermal limits are not exceeded and evaluates rocket performances. The approach successfully reduces the number of configurations to explore, and limits additional analyses to those configurations that are truly competitive. Furthermore, configurations in the optimal region achieve better performance than the initial configuration. It is also found that the most demanding constraints relate to safety: many promising configurations were discarded because they cannot meet safety criteria. This highlights the necessity of including safety analyses in the initial phase of reactor designs. It is concluded that a LEU, CERMET-fuelled reactor design can be considered challenging but feasible, and the approach presented in the work may be applied to find an optimal configuration for a preliminary reactor design.
Whether to use highly enriched uranium (HEU) or low-enriched uranium (LEU) in space reactors is a highly debated topic. Most analyses focus on performance as the principal determinant of use, where HEU has inherent advantages. This paper identifies seven dimensions along which rigorous comparisons must be made to evaluate whether HEU or LEU is an appropriate enrichment level for space nuclear systems. These dimensions are performance, safety, security and nonproliferation, timeliness of a system to come to fruition, fuel availability, cost, and ability to include commercial partners. Our analysis shows that HEU and LEU systems provide different advantages depending on the dimension of interest, and whether the United States continues to use HEU or switches to LEU is ultimately a policy decision, not a technical one.
The micro heat pipe cooled reactor system with high efficiency energy conversion units are potential candidates for the energy support in decentralized markets, such as oceanic activities and remote regions. However, its development has been limited by several unfavorable factors. In this work, the potential opportunities and challenges of micro nuclear power system with high efficiency energy conversion units are analyzed systematically. Suggestions for future work are also proposed.
Studies of Nuclear Thermal Propulsion (NTP) over the past several decades, and updated most recently with the examination of Low Enriched Uranium (LEU), have shown nuclear propulsion is an enabling technology to reach beyond this planet and establish permanent human outposts at Mars or rapidly travel to any other solar system body. The propulsion needed to propel human spacecraft needs high thrust to operate withinthe deep gravity well of a planet and provide high propulsive efficiency for rapid travel and reduced total spacecraft mass.
NTP can provide the thrust to move a spacecraft between orbits, can operate as a dual-mode system that provides power and propulsion capability, provides a strong architectural benefit to human and robotic exploration missions, and provides a path toward reusable in-space transportation systems. NTP provides smaller vehicle systems due to its specific impulse (ISP) being twice that of the best cryogenic liquid rocket propulsion and can thus provide reduced trip times for round-trip missions from Earth to Mars.
Aerojet Rocketdyne (AR) is working with NASA, other government agencies, and other industry partners to improve the design and reduce the cost of NTP engine systems. Current NTP designs focus on thrust sizes between 15,000-lbf (~67-kN) and 25,000-lbf (~111-kN). AR in 2019 has examined various reactor cores and enhancements to optimize LEU NTP designs. The enhancements improve the mission architecture robustness and provide more design margin for Mars vehicles across many mission opportunities, trip times, and mission types.
The LEU NTP design can offer mission architecture stages or elements that can be used for both Lunar and deep space exploration missions. The NTP designs enable packaging of various NTP stage designs (e.g., crew Mars vehicle stage elements, cargo stage derivatives, a deep space stage with payload, Lunar stage elements) on the NASA SLS Block 2 using the 8.4-meter fairing. The primary LEU core designs studied, for the above-mentioned missions, have relied on liquid hydrogen for the propellant and coolant and use Zirconium Hydride within a structural element as the neutron moderator. Several designs have been examined that use Beryllium Oxide with the fuel elements in the core to eliminate the structural elements.
The LEU NTP engine systems studied have typically been used only for the primary delta-V burns (e.g., earth escape, planetary capture, planetary escape, earth return capture). LEU NTP engine systems have also been examined using a the LEU reactor fuel element and moderator element approach to perform orbital maneuvering system (OMS) burns during the mission simply by permitting the reactor to keep operating at very low power levels during the entire mission.
This paper will discuss the various engine system and mission design trades performed in 2019 for Mars and lunar missions when using a single NTP or a cluster of NTP engines.
This paper explores the possibility of passively controlling the reactivity of a nuclear thermal propulsion (NTP) reactor. The objective of this study is to limit the use of the radial control drums to start-up and shutdown procedures and ensure that the exact same operation is performed for each full-power burn. To achieve the goal, this work considers several design measures, which include a low-density burnable absorber in the tie-tube components of the core, the use of variable hydrogen density in the moderator element coolant passages, and the judicious selection of a modified mission profile to maximize the decay of 135Xe after operation. In addition, the improved stability from the enhanced fuel temperature feedback due to the implementation of low-enriched-uranium fuel is also exploited for the realization of passive reactivity control. In this work, a passive reactivity control system is implemented in the Superb Use of Low Enriched Uranium (SULEU) NTP core and analyzed in terms of its ability to fulfill a NASA Mars Mission Design Reference Architecture 5.0-style mission. It is concluded that the use of the control drums can be limited to start-up and shutdown operations only, eliminating operator input in order to maintain a constant power level in the core.
Nuclear Thermal Rocket (NTR) propulsion is a viable and meritorious option for human exploration into deep-space because of its high thrust, improved specific impulse, well established technology, bimodal capability, and enhanced mission safety and reliability. The NTR technology has already been investigated and tested by the United States of America and Russia and the former Soviet Union. The representative Nuclear Engine for Rocket Vehicle Applications (NERVA) type reactors traditionally used Highly Enriched Uranium (HEU) fuels, shaped in hexagonal fuel element geometries because of the importance of making a high power reactor with a minimum size. Although the HEU-NTR designs are the best choice in terms of rocket performance and technical maturity, they inevitably provoke nuclear proliferation obstacles not only for all research and development activities by civilians and non-nuclear weapon states but also for potential commercialization. To overcome the security issues due to HEU, the non-proliferative, small-size NTR engine with low thrust levels of 41 kN–53 kN (9.2 klbf ∼ 11.9 klbf), Korea Advanced NUclear Thermal Engine Rocket utilizing a Low-Enriched Uranium fuel (KANUTER-LEU), is being designed for future generations. Its design goals are to make use of an LEU fuel for its fairly compact core, but to minimize the rocket performance sacrifice relative to the traditional HEU-NTRs. To achieve these goals, a new space propulsion reactor is conceptually designed with the key concepts of a high uranium density fuel with resistance against high heating and H2 corrosion, a thermal neutron spectrum core, and a compact and integrated fuel element core design with protective cooling capability. In addition, a preliminary design study of neutronics and thermal-hydraulics was performed to explore the design space of the new LEU-NTR reactor concept. The result indicates that the innovative reactor concept has great potential, both to implement the use of an LEU fuel and to create comparable rocket performance, compared to the existing HEU-NTR designs.
In order to elucidate the phase diagram of the zirconiumhydrogen system ; at high hydrogen contents, a pressurecomposition-temperature study in the ; temperature range 500 to 850 deg C was carried out. From the isotherms obtained, ; the position of the boundary between the two-phase ( beta + delta ) region and ; the single phase delta region was more precisely defined ( beta designates the ; hydrogen stabilized high temperature zirconium phase, and delta , the cubic ; hydride phase). The isotherms also showed the existence of a two-phase hydride ; region (cubic and tetragonal hydride co-existing) at elevated temperatures as has ; been observed at room temperature. (auth);
The SiC layer integrity in the TRISO-coated gas-reactor fuel particle is critical to the performance, allowed burn-up, and hence intrinsic efficiency of high temperature gas cooled reactors. While there has been significant developmental work on manufacturing the fuel particles, detailed understanding of the effects of the complex in-service stress state combined with realistic materials property data under irradiation on fuel particle survival is not adequately understood. This particularly frustrates the modeling efforts that seek to improve fuel performance through basic understanding. In this work a compilation of non-irradiated and irradiated properties of SiC are provided and reviewed and analyzed in terms of application to TRISO fuels. In addition to a compilation and review of literature data, new data generated to fill holes in the existing database is included, specifically in the high-temperature irradiation regime. Another critical piece of information, the strength of the SiC/Pyrolytic carbon interface, was measured and is included, along with a formalism for its analysis. Finally, recommended empirical treatments of the data are suggested.