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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Key Design Trades for a Near-term Lunar Fission Surface Power System

NASA and DOE have been developing concepts for lunar and Mars Fission Surface Power (FSP) systems for decades and have considered many different technologies and system variations. In 2023, three contractor teams completed Phase 1 conceptual design studies on a 40 kWe lunar FSP system in response to a modest set of NASA requirements and goals. A government team consisting of NASA Glenn Research Center (GRC), Los Alamos National Laboratory (LANL), and Idaho National Laboratory (INL) was formed to develop an independent design approach that adhered to the same requirements and goals posed to the contractor teams. GRC provided expertise, analysis, and design decisions for the power conversion, heat rejection, power management & distribution (PMAD), and mission integration. LANL and INL provided expertise, analysis, and design decisions for the reactor and shielding. The goal of the government study was to develop representative concepts that could be used in NASA architecture studies, guide government technology investments, and inform Phase 2 requirements definition. This paper provides a summary of the key technology options, system-level trades, and mission-level concept-of-operations.

99 - GENERAL AND MISCELLANEOUS↗
DOE OSTI2015

Study of Compatibility of Stainless Steel Weld Joints with Liquid Sodium-Potassium Coolants for Fission Surface Power Reactors for Lunar and Space Applications

To make a manned mission to the surface of the moon or to Mars with any significant residence time, the power requirements will make a nuclear reactor the most feasible source of energy. To prepare for such a mission, NASA has teamed with the DOE to develop Fission Surface Power technology with the goal of developing viable options. The Fission Surface Power System (FSPS) recommended as the initial baseline design includes a liquid metal reactor and primary coolant system that transfers heat to two intermediate liquid metal heat transfer loops. Each intermediate loop transfers heat to two Stirling heat exchangers that each power two Stirling converters. Both the primary and the intermediate loops will use sodium-potassium (NaK) as the liquid metal coolant, and the primary loop will operate at temperatures exceeding 600°C. The alloy selected for the heat exchangers and piping is AISI Type 316L stainless steel. The extensive experience with NaK in breeder reactor programs and with earlier space reactors for unmanned missions lends considerable confidence in using NaK as a coolant in contact with stainless steel alloys. However, the microstructure, chemical segregation, and stress state of a weld leads to the potential for corrosion and cracking. Such failures have been experienced in NaK systems that have operated for times less than the eight year goal for the FSPS. For this reason, it was necessary to evaluate candidate weld techniques and expose welds to high-temperature, flowing NaK in a closed, closely controlled system. The goal of this project was to determine the optimum weld configuration for a NaK system that will withstand service for eight years under FSPS conditions. Since the most difficult weld to make and to evaluate is the tube to tube sheet weld in the intermediate heat exchangers, it was the focus of this research. A pumped loop of flowing NaK was fabricated for exposure of candidate weld specimens at temperatures of 600°C, the expected temperature within the intermediate heat exchangers. Since metal transfer from a high-temperature region to a cooler region is a predominant mode of corrosion in liquid metal systems, specimens were placed at zones in the loop at the above temperature to evaluate the effects of both alloy component leaching and metal deposition. Microstructural analysis was performed to evaluate weld performance on control weld specimens. The research was coordinated with Oak Ridge National Laboratory (ORNL) where most of the weld samples were prepared. In addition, ORNL participated in the loop operation to assist in keeping the testing relevant to the project and to take advantage of the extensive experience in liquid metal research at ORNL.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Analysis of Alternative Core Designs for Fission Surface Power Capability Demonstration Mission (V.1.0)

NASA’s Space Technology Mission Directorate (STMD) intends to demonstrate a 10 kWe fission surface power (FSP) system on the surface of the Moon as part of a Capability Demonstration Mission. STMD tasked a joint NASA-DOE team to ‘outline an FSP reactor concept that can be readied for 2027 launch,’ and serves ‘as a pathfinder for future power modules for Mars.’ Tables 1 and 2 respectively list FSP performance requirements and design guidance provided for the study. DOE is the leading agency for the reactor module, reactor controls and radiation shielding, and shares joint responsibility for the overall FSP system. This report summarizes an analysis of alternative reactor classes and shielding configurations carried out by the DOE national laboratories. Representative reactor and shielding configurations were based on heritage and are in alignment with the industry initiatives. This report presents no recommendations on the appropriate choice of fuel or class of reactor that NASA should pursue. Instead the focus here is to provide a summary and findings of analyses carried out by DOE national laboratory personnel.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Small CaH{sub 2} Moderated Thermal Reactor for Surface Power Generation I: Conceptual Design

National Aeronautics and Space Administration (NASA) published 'Journey to Mars' on October 8, 2015, stating their plan to send human to Mars for exploration, scientific research and colonization of Mars. Mission time on Martian surface is supposed to be ∼1 year, therefore, a surface power generation system is required for human habitation on Mars. Even though solar power and chemical battery have been used for the past missions, the former cannot generate power during month-scale dust storm on Mars, and the latter is not reliable for long-term use under a cold environment. On the other hand, nuclear power provides stable power even on Mars surface and, since kilowatt class power system is required in the planned mission, a fission power system is better than a radioisotope battery. An 8-year operation is required for the Mars surface power generation system with electrical power of 10 to 100 kWe and specific power of >5 We/kg, which means the system weight needs to be below 2000 kg in the case of 10 to 100 kWe system. In the current study, stretch goals for surface fission power system were set as follows: electric power of 20 kWe and total system weight below 1000 kg. In addition, nuclear security and critical safety are essential features for the surface fission power system considering launch accidents. A sufficient level of nuclear security is needed to prevent terrorists from obtaining nuclear materials even the reactor falls in a geographic area dominated by terrorists. From this view point, it is preferable not to use highly enriched uranium with a space reactor that can be diverted to nuclear weapons. Also, a nuclear reactor core should not reach criticality during a launch accident where the reactor core can be compressed by collision with sea or land. A novel concept of surface power reactor system that satisfies the above requirements is proposed in the current study. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Kilopower - Powering the Future of Space Exploration [Poster]

The Kilopower Project was initiated by NASA’s Space Technology Mission Directorate to demonstrate subsystem-level technology readiness of small space fission power in a relevant environment (TRL 5) for space science and human exploration power needs. The Kilopower Project centerpiece is the Kilowatt Reactor Using Stirling Technology (KRUSTY) test, which consists of the development and testing of a ground technology demonstrator of a 1 kWe class fission power system. The technologies to be developed and validated by KRUSTY are extensible to space fission power systems from 1 to 10 kWe, which can enable modular surface fission power systems for human exploration, as well as higher power future potential deep space science missions.

42 ENGINEERING↗
DOE OSTI2016

Multiphysics analysis of liquid metal annular linear induction pumps: A project overview

Liquid metal-cooled fission reactors are both moderated and cooled by a liquid metal solution. These reactors are typically very compact and they can be used in regular electric power production, for naval and space propulsion systems or in fission surface power systems for planetary exploration. The coupling between the electromagnetics and thermo-fluid mechanical phenomena observed in liquid metal thermo-magnetic systems for nuclear and space applications gives rise to complex engineering magnetohydrodynamics and numerical problems. It is known that electromagnetic pumps have a number of advantages over rotating mechanisms: absence of moving parts, low noise and vibration level, simplicity of flow rate regulation, easy maintenance and so on. However, while developing annular linear induction pumps, we are faced with a significant problem of magnetohydrodynamic instability arising in the device. The complex flow behavior in this type of devices includes a time-varying Lorentz force and pressure pulsation due to the time-varying electromagnetic fields and the induced convective currents that originates from the liquid metal flow, leading to instability problems along the device geometry. The determinations of the geometry and electrical configuration of liquid metal thermo-magnetic devices give rise to a complex inverse magnetohydrodynamic field problem were techniques for global optimization should be used, magnetohydrodynamics instabilities understood –or quantified- and multiphysics models developed and analyzed. Lastly, we present a project overview as well as a few computational models developed to study liquid metal annular linear induction pumps using first principles and the a few results of our multi-physics analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Versatile Heat Transfer Module

Development of micro-reactors that can be easily transported to remote areas for civilian or military applications is an immediate focus of U.S. nuclear industry and therefore is being pursued by multiple reactor vendors (e.g. HolosGen LLC, Westinghouse, Ultra Safe Nuclear Corporation, URENCO, etc.) and actively supported by DOE-NE, DOE-ARPA-E and by the Department Of Defense (DoD). Very-high-temperature gas-cooled (VHTR) and heat pipe cooled reactor technologies are regarded as promising concepts for microreactors applications since they enable compact power conversion systems and display beneficial inherent safety features. Development of microreactors with enhanced performance must be supported by innovations in nuclear technologies. Several technologies being developed by the microreactor industry are being revisited with material technologies developed at the Argonne National Laboratory (ANL). Innovative high-temperature compatible solutions are developed to enable advanced thermal-spectrum microreactor designs featuring improved neutronics performance at high-temperature operating conditions to enable high thermal efficiency and economic competitiveness. This report focuses on the Versatile Heat Transfer Module (VHTM) technology that is developed at ANL for heat-transfer in thermal-spectrum microreactors. This technology is built on a liquid alkali metal heat pipe concept while leveraging advanced nuclear material technologies for high-temperature applications at reduced neutronic penalty with potential improvement in inherent safety behavior. The VHTM can be applied to design high-temperature thermal-spectrum microreactor systems with reduced size, weight, uranium enrichment, or increased lifetime. The INFINIT technology deriving from the VHTM enables design of a highly compact long-life microreactor concept for fission surface power (FSP). Demonstration of multiphysics modeling capability was also completed to showcase the added inherent safety benefits obtained with this concept. The VHTM technology remain at an early development stage with low Technology Readiness Level and Manufacturing Readiness Level. The current demonstration of these technologies has focused on manufacturing capability, on performance of the hydrogen diffusion barrier under thermal cycling, while radiation tolerance demonstration is planned. The next step of the demonstration plan will be the assembly of miniature VHTM for high-temperature testing, which can be used as a prototype for future pilot scale demonstration. Such research is critical to pursue development of these technologies that have the clear potential to significantly improve the performance of currently investigated micro-reactors systems.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Thermocouple and RTD Drift Due to Neutron Irradiation and Temperature for Space Nuclear Applications: A Review

Temperature measurement is a key factor in the safe and reliable operation of nuclear reactors. Thermocouples and resistance temperature detectors (RTDs) are among the most common and reliable tools deployed for temperature measurement, and some are capable of operating in extreme environments such as very high temperature and neutron radiation fields anticipated in nuclear thermal propulsion (NTP) or fission surface power. These devices are known to exhibit drifting behavior in their temperature readings throughout their operational lifetimes. Here, to help select appropriate instrumentation, it is important to identify the factors that influence drift as well as typical behavior in these environments. This work is focused on identifying contributing factors in the performance of thermocouples and RTDs. An extensive literature review is included that discusses several experimental results in different environments. Although individual instrument performance may vary with construction methods, materials selection, and vendors, this review aims to assist in the selection process and identify typical expected behaviors of several devices.

Chesser, Ryan J. [Oak Ridge National Laboratory (O↗
DOE OSTI2024

Hydrogen transport in yttrium hydride under asymmetric heat

Metal hydrides are a promising moderator material for high temperature fission reactors. Yttrium hydride can be loaded to a high hydrogen density with relatively high hydrogen stability at temperatures up to 800°C. This makes yttrium hydride a potential moderator material for microreactors as foreseen in the fission surface power program. However, during the operation of such advanced reactors temperature gradients are expected which can change the local hydrogen density in the moderator. Hydrogen diffusion in metals is driven by a concentration gradient (Fick’s law) and thermal diffusion (Soret diffusion). Thermal diffusion is the transport of hydrogen, or other species, due to a temperature gradient. For example, hydrogen might migrate from the hot side of a sample to the cold side of a sample. Measuring Fickian diffusion is achieved through various permeation or absorption experiments, however measuring thermal diffusion is challenging and has rarely been performed. The Hydrogen Experimental Apparatus for Thermal Diffusion (HEATD) experiment is designed to induce thermal diffusion in samples and quench those samples so that the hydrogen distribution can be analyzed using hot vacuum extraction (HVE). One side of the sample was heated to a high temperature e.g., 800°C, while the other side of the sample is at a lower temperature. The sample was held under the applied temperature gradient for a given time until the anticipated hydrogen diffusion has occurred. The actual time depends depend on the sample composition and hydrogen concentration. The results from HVE showed that thermal diffusion took place in the specimen and the Soret coefficient was calculated.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Thermal properties of U-Mo alloys irradiated under high fission power density

A variety of physical and thermal property measurements made on irradiated U-Mo alloy monolithic fuel samples with a Zr diffusion barrier and clad in Al alloy 6061 are reported. These properties form the basis for additional investigations in thermal behavior of U-Mo fuel under high performance research reactor irradiation conditions. Measurements were performed on samples harvested from fuel plates irradiated under high fission power density and surface heat flux and complement previous measurements on samples subjected to moderate fission power density and surface heat flux. Specific heat capacity of irradiated monolithic U-Mo increases with increasing temperature and fission density, but carries large uncertainties associated with thickness measurements used to determine the mass fraction of the layer constituents for extraction of the U-Mo specific heat capacity, specifically the very thin and nonuniform Zr diffusion barrier. Density of irradiated monolithic U-Mo decreases with increasing fission density resulting from fission gas swelling and porosity formation. Thermal diffusivity of irradiated monolithic U-Mo increased with increasing temperature and decreased with increasing fission density. Thermal conductivity of irradiated monolithic U-Mo increased with increasing temperature, and in general, appeared to decrease and become less sensitive to temperature with increasing fission density. Thermal conductivity measurements compared to a semi-empirical model agreed within ±30%. It is suspected that the main driver in deviation between model calculations and measurements is the fission gas swelling determination, which does not account for fission power density or irradiation temperature.

uranium (U), molybdenum, nuclear fuel, thermal con↗
DOE OSTI2023

AN INITIAL LOOK AT THE HEAT PIPE RESILIENCY TO REACTOR OPERATION

Several special purpose reactor designs are aiming to utilize heat pipes due to its inherently passive features allowing safe heat removal to the power conversion systems. There is insufficient data on heat pipe survivability in reactor environments, although heat-pipe failures are predicted to have low failure rates. In this paper, we perform a coupled neutronics and thermo-mechanics analyses for a 45 kWth HALEU fueled, hydride moderated homogenous design to evaluate the transient effects during both startup and a heatpipe-failure scenario. The heatpipe-failure study includes both a single failure case and a failure-propagation (cascading) case. While the model only evaluates a homogenous core, the approach could be applied to more complex models for any heat pipe cooled small reactor.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2024

Effect of impurities on hydrogen defect stability and migration barrier in yttrium dihydride crystal

The impurity or alloying atoms in YH 2 can alter the local electronic structure and so the hydrogen defect stability, as well as the H migration barrier energy. Thus, DFT calculations were employed to determine the effect of foreign elements from alkali and alkaline earth metals to transition metals and one critical impurity element, O, on H vacancy stability and retention characteristics in YH2. Results revealed that alloying elements act as hydrogen vacancy sinks by reducing the vacancy formation energy at neighboring sites. The implantation of non-magnetic foreign elements (s1, s2, and d10 valence electrons) in hydrogen energy landscape was calculated to be minor; while the hydrogen vacancy formation energy was reduced from 1.37 eV to 1.00 eV, the migration energy barrier of hydrogen was increased from 0.87 eV to 1.15 eV for non-magnetic foreign elements. The migration energy barrier monotonically decreased with increasing d-shell occupancy, reaching as low as 0.4 eV for Cr, Mo(d4), and Fe (d4). Alloying with late transition metals (d8 and d9) moderately impacted the hydrogen vacancy formation. Finally, it was found to be O addition into the YH 2- lattice did not alter the energy landscape of hydrogen vacancies. Since alloyed YH 2 has not been studied extensively, this study provides an atomistic understanding how alloying elements and impurities trap vacancies and affects hydrogen mobility YH 2 . Meanwhile, the main findings of this study may serve as guidelines for introducing alloying elements in ZrH 2 as well.

08 HYDROGEN↗
DOE OSTI2023

Space Fission Reactors

An overview of space fission reactors, both historical and current programs. Intended for university student audiences

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2016

Assembly, Test and Launch Operations for a Nuclear-enabled NASA Mission: Considerations that are Specific to Use of a Nuclear Payload

For more than five decades, Radioisotope Power Systems (RPS) have played a critical role in the exploration of space, enabling missions of scientific discovery to destinations across the solar system by providing electrical power to explore remote, challenging and extreme environments. In particular, RPS enable deep space missions where increased heliocentric distances reduce the ability of solar power to adequately meet spacecraft and instruments power requirements. Some previous notable missions that were enabled by RPS include Nimbus III, the Apollo Surface Experiments, the Pioneers 10 and 11, the Viking Mars Landers, Galileo, Ulysses, Cassini, New Horizons and Curiosity. The current operating set of missions that are enabled by RPS are Voyagers 1 and 2, Cassini, New Horizons, and Curiosity. The Multi-Mission Radioisotope Thermal Generator (MMRTG) is the current RPS used for missions. An enhanced version of this generator outfitted with higher efficiency thermoelectrics is under development for potential use in the future. Other previously deployed power systems include the Multi-hundred Watt Radisotope Thermo-electric Generator (MHW RTG) and the General Purpose Heat Source Radisotope Thermo-electric Generator (GPHS RTG). The common thread for all of these power systems is that they are fueled with Pu-238. The use of this unique isotope involves additional planning activities and requires specific actions when the devices are delivered to Kennedy Space Center (KSC), launch site, and incorporated into the assembly, test and launch operations (ATLO) process to ensure mission success, met safety and security challenges. It has been forecasted that the use of a nuclear reactor system is on the horizon. This system could be used for either specifically powering S/C propulsion system or for surface power use once the mission arrived at its destina-tion. Since a nuclear-fission-based system has never been handled or integrated into a spacecraft at KSC, this nuclear reactor system would potentially require further challenges that must be accounted for in the ATLO process. This paper will explain ATLO considerations for the recent RTG-enabled launches that have occurred and planned in the near future (Mars 2020 NASA mission). It will also describe where further challenges that need to be met for the placement of a nuclear reactor (either for propulsion or transportation for surface power use) onto a space mission. Specifically, the following topics will be addressed: • Approach for nuclear safety planning for nuclear material use and its transportation to launch site • Planning for security posture for nuclear materials at launch site • Preparation for transportation of the nuclear power system from the fueling/testing location to the launch site • Preparation of documentation and procedures for nuclear material use at launch site • Planning for coordination between power system and space mission groups • Planning for appropriate staffing and scheduling of operations

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Data and Insights for the KRUSTY Nuclear-Powered Tests

This document, and the associated files, describe all of the data from the KRUSTY warm criticals and the KRUSTY nuclear system test. The warm criticals were a group of tests where fission power heated the core and the feedback response was studied. There were 3 experiments, the 15-cent run (which was a true “free run”), followed by 30-cent and 60-cent runs, where reactivity was added to heat the core to higher temperatures. The nuclear system test (often referred to as the full-run or the final-run), engaged the power conversion system and evaluated system performance. These experiments are described in a pair of NUCLEAR TECHNOLOGY papers, which includes plots of the data that this document-set includes. Note that a lot of information in this document is based on my interpretation and could be wrong, but I’ve included it to potentially save some people a lot of time and headaches by trying to figure out which instrumentation to believe, but beware treating anything I say as fact. I’ve also included several post-irradiation photos, to show some thermocouple locations and indicate possible changes in emissivity and/or geometry (although no changes in geometry were indicated).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Development of Advanced Instrumentation for Transient Testing

A fuel safety research program centered on in-pile transient testing experiments is being developed to support assessment and qualification of advanced nuclear fuel systems using the recently restarted Transient Reactor Test (TREAT) facility at the Idaho National Laboratory (INL). While resumption of transient testing at TREAT is crucial to enable these programs, full recovery and cutting-edge transient testing capability also requires a well-coordinated and innovative instrumentation development and qualification program to support near-term and future objectives. This paper summarizes the experimental approach of transient testing to focus on measuring the response of nuclear fuel to off-normal (or power-cooling mismatch) conditions for modern and advanced reactor environments requiring capabilities extending over wide measurement and environment conditions. It also highlights unique attributes of transient testing of importance to in-pile instruments including relatively low total neutron fluence, high gamma heating, and the need for well-defined and possibly short time response. Historical approaches to instrumentation for transient testing are also reviewed to provide context to the modern instrument strategy. The instrumentation needs of a modern transient testing are detailed while summarizing several ongoing R&D activities are supporting the development of state-of-the-art and ‘advanced’ measurement technologies that will provide baseline capability for Light Water Reactor (LWR) and Sodium-Cooled Fast Reactor (SFR) experiment objectives, while extending to other advanced reactor needs and advanced sensing technology opportunities. Examples of specific sensors planned for near-term deployment with ongoing development includes prompt-response self-powered neutron detectors (SPND), miniature fission chambers, optical-fiber-coupled infrared pyrometer, cladding surface thermocouples, electrical-impedance based boiling detector, and Linear Variable Differential Transformer (LVDT) based sensors for fuel elongation and pressure measurement.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗