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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 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Investigating Radiation-Induced Actinide Species in Solution

Our fundamental understanding of actinide radiation-induced redox chemistry is crucial due to their unavoidable exposure to ionizing radiation fields, both inherent and from in-process applications. Plutonium (Pu) and americium (Am) both possess multiple oxidation states, the careful manipulation of which are essential in the study and utilization of their rich chemistry, developing new nuclear technologies, and securing the long-term sustainability of nuclear power. However, knowledge in this area is far from complete. Consequently, we have studied the radiation-induced chemistry of both Pu and Am through a variety of techniques. Temperature-controlled electron pulse radiolysis has been used to study Am for the first time, determining the feasibility of Am redox reactions under used nuclear fuel reprocessing conditions, (e.g. nitric acid, non-ambient temperature). Additionally, we developed an experimentally evaluated multi-scale computer model for the prediction of gamma radiation-induced Pu redox chemistry due to radiolysis and disproportionation reactions in concentrated nitric acid solutions.

actinide↗
NASA NTRS2010

Progress Towards the Development of a Long-Lived Venus Lander Duplex System

NASA has begun the development of a combined Stirling cycle power and cooling system (duplex) to enable the long-lived surface exploration of Venus and other harsh environments in the solar system. The duplex system will operate from the heat provided by decaying radioisotope plutonium-238 or its substitute. Since the surface of Venus has a thick, hot, and corrosive atmosphere, it is a challenging proposition to maintain sensitive lander electronics under survivable conditions. This development effort requires the integration of: a radioisotope or fission heat source; heat pipes; high-temperature, corrosion-resistant material; multistage cooling; a novel free-displacer Stirling convertor for the lander; and a minimal vibration thermoacoustic Stirling convertor for the seismometer. The first year effort includes conceptual system design and control studies, materials development, and prototype hardware testing. A summary of these findings and test results is presented in this report.

Dyson, Roger W.↗
NASA NTRS2011

Progress Towards the Development of a Long-Lived Venus Lander Duplex System

NASA has begun the development of a combined Stirling cycle power and cooling system (duplex) to enable the long-lived surface exploration of Venus and other harsh environments in the solar system. The duplex system will operate from the heat provided by decaying radioisotope plutonium-238 or its substitute. Since the surface of Venus has a thick, hot, and corrosive atmosphere, it is a challenging proposition to maintain sensitive lander electronics under survivable conditions. This development effort requires the integration of: a radioisotope or fission heat source; heat pipes; high-temperature, corrosion-resistant material; multistage cooling; a novel free-displacer Stirling convertor for the lander; and a minimal vibration thermoacoustic Stirling convertor for the seismometer. The first year effort includes conceptual system design and control studies, materials development, and prototype hardware testing. A summary of these findings and test results is presented in this report.

Dyson, Rodger, W.↗
NASA NTRS1978

A 5-GWe nuclear satellite power system conceptual design

This paper presents the results of a brief study performed for MSFC on the conceptual design of a nuclear satellite power station which delivers 5 GWe net power to earth by microwave transmission. The system contains 26 modules each consisting of a reactor, fuel processing plant, Brayton PCU, space radiator, and nuclear shield. A high-temperature, gas-cooled, pebble-bed plutonium breeder concept was selected which is resupplied with fertile U-238. Sections of this core are periodically replaced and the spent fuel is chemically processed, the radioactive wastes separated, and stored for eventual space disposal. Fresh fuel pellets, formed from the U-238 and the bred plutonium, are recycled back to the reactor. The hot (1317 C) helium gas exiting the reactor serves as the working fluid in a 30%-efficient Brayton PCU.

Goodman, M.↗
DOE OSTI2025

Feasibility of Recycling Discharged Microreactor Heavy Metal in Light Water and Sodium-Cooled Fast Reactors: A Neutronics Analysis

Nuclear microreactors (MRs) offer unique advantages, such as rapid deployment, potability, low maintenance requirements, and operational flexibility. Their compact size makes them a promising solution for decentralized power generation, particularly in remote areas, military bases, and disaster-stricken regions. However, MRs face challenges, including unutilized fissile material at the end of life, economic inefficiency, increased heavy metal (HM) waste complicating disposal, and the accumulation of plutonium (Pu) with high 239 Pu concentrations raising proliferation risks. Here, this study investigated the neutronics feasibility of a novel three-stage fuel cycle where discharged HM from MRs is recycled and burned in light water reactors and sodium-cooled fast reactors. This approach converts discharged HM into valuable fuel, enhancing the efficiency of MR deployments while improving the safeguardability of their final waste products. Neutronics analysis demonstrated that the safety characteristics of reactor designs in each stage were minimally impacted by the proposed cycle. For two representative MR designs, a fast-spectrum MR with solid pellet fuel and a thermal-spectrum MR with TRISO (TRi-structural-ISOtropic) fuel compacts, the proposed fuel cycle reduced the uranium disposal mass flow rate by ~60%, decreased the 235 U enrichment of the discharge fuel to ~1 wt%, eliminated plutonium disposal, and increased the cumulative fuel burnup to ~580 gigawatt-day per metric ton of initial heavy metal (GWd/t-iHM) or 60% fissions per initial metal atom. Despite the significant differences between the two MR designs, the performance and infrastructure requirements of the developed fuel cycles were remarkably similar, indicating its generalizability to a broader class of MRs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

The CWS Experiments – An Experimental Study of the Effects of Chlorine on Thermal Neutron Absorption

The Chlorine Worth Study (CWS) experiments consist of layers of Zero Power Physics Reactor plates moderated by high density polyethylene, polyvinyl chloride, and chlorinated polyvinyl chloride with the goal to produce a thermal critical experiment. A thick high density polyethylene reflector surrounds the experiments. Aluminum fuel trays and frames were used to conduct the intrinsic heat generation of the plutonium plates away from the core and out to the top plate and the platen, both acting as heat sinks. The goal of the CWS experiments was to provide benchmark critical experiments that were sensitive to chlorine and matched the sensitivity of criticality safety applications for aqueous chloride operations at PF-4 at Los Alamos National Laboratory. The specific aqueous chloride PF-4 applications were solutions of 30 g/L plutonium, 300 g/L plutonium, and 600 g/L plutonium. The CWS experiments were designed using prototype methods developed by the ARCHIMEDES and EUCLID projects at Los Alamos National Laboratory. The CWS experiments were performed in December of 2021 at the National Criticality Experiments Research Center at the Nevada Nuclear Security Site on the Planet critical assembly. It is intended that the CWS experiments will be evaluated and submitted to the ICSBEP in the future as criticality safety benchmarks.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Materials for Small Nuclear Reactors and Micro Reactors, Including Space Reactors

Small nuclear reactors, including small modular reactors (SMRs) or reactors for space applications, rely on different materials than those typically applied in large-scale nuclear power plants. Examples include molten salts as cooling medium or fuel carrier, metal hydrides as high-temperature moderators, and fuels allowing for higher burnup. All of these also require novel structural materials, for which material interactions have to be understood. Fissionable and fissile materials, such as uranium or plutonium, are rarely considered in materials design other than for nuclear fuels. Similarly, the aspects of radiation damage, occurring during irradiation when a reactor operates, are unique to nuclear materials research. The handling of these materials puts further limitations on the materials science conducted for nuclear materials. All of these issues move research for these materials off the “main stream” of materials science, and cause it to be more easily conducted at national laboratories. However, nuclear reactors offer unique opportunities for carbon-neutral energy generation and have great potential to address if not solve problems arising from global warming. This special topic, sponsored by the TMS Nuclear Materials Committee, focuses on materials research for small nuclear reactors, both experimental and simulation/modeling.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Parabolic oxidation kinetics of a plutonium alloy at room temperature

Auger electron spectroscopy (AES) was used to characterize the oxide scale formed on an ~0.5 wt. % Ga plutonium metal alloy. Analysis of changes to the characteristic Auger transitions for Pu and O during sputter depth profiles were used in combination with previously measured sputter yields to determine the thickness of the oxide scale and provide insight into its chemistry. From this data, the parabolic kinetics for the Pu-Ga alloy were determined as a function of relative humidity, RH = 14–92%, and oxygen partial pressure, p(O 2 ) = 31–157 Torr, at 25°C. For each relative humidity investigated, the parabolic rate constant was found to be independent of oxygen partial pressure. Furthermore, power-law fits to the relative humidity effect on the rate constant was found to have a power of n = 1.265, indicating the diffusion species which rate-limits the reaction isn’t molecular water, which subsequent dissociates at the oxide-metal interface, or hydroxyl from dissociation at the gas-surface interface.

36 MATERIALS SCIENCE↗
NASA NTRS1987

Space reactor assessment and validation study

The present difficulties experienced by the United States in launching payloads into space has suggested a number of problems which are associated with the handling of hazardous materials in spacecraft. The question has arisen as to the safety of launching highly radioactive material such as plutonium-238, related to the possibility of its dispersion into the atmosphere during a launch vehicle explosion. An alternative is the use of a small nuclear reactor which is not started until it is in space and contains little or no radioactivity at launch. A first order assessment of six small reactor concepts with power levels up to 100 MWe was performed. Both the nuclear feasibility of these concepts to operate at their rated power levels between 7 and 10 years and the capability of these concepts to remain subcritical both before and during launch and also in the case of water immersion during a potential launch failure or abort were investigated.

Gedeon, Stephen↗
DOE OSTI2025

A Semi-Empirical Density Law for Ternary, Homogeneous PuCl 3 /HCl/H 2 O Solutions

Nuclear material operations pose unique hazards that are not encountered in other chemical, energy, or manufacturing industries. One of these hazards is the potential for a nuclear criticality accident when handling fissile isotopes such as 235 U and 239 Pu. These hazards are particularly high when fissile material is dissolved in solution as the neutron behaviors of the system can change rapidly with the physical and chemical changes accessible in solution. Current estimates of solution density used for criticality safety are outdated and hinder fissile material handling. Developing new estimates for these safety calculations requires experimental characterization and the derivation of empirical density models. We have derived a density law describing PuCl 3 /HCl/H 2 O solutions from experimental data characterizing solution density. Density data was treated using a Pitzer-derived eight-parameter equation, defining density as a function of analyte concentrations, temperature, and interactions between these variables. The model is predictive across the concentration and temperature ranges from which it was derived. The potential effects of varying oxidation states of plutonium, which are easily accessible in aqueous media, on the bulk solution density of the ternary system were also investigated. The resulting Pitzer-derived density law was applied to a nuclear criticality safety model, and the impact of the experimental characterization of solution density relative to previous estimates was demonstrated to be significant and suggest that the current approach to estimating density in nuclear criticality safety calculations may lead to overly conservative controls.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Getting things done onsite: Heat Source production never ramps down

The COVID-19 pandemic has changed many norms in the world and at LANL. During the rapid transition of sending many to work from home in March, some programs were deemed mission critical, with the expectation to continue driving their missions forward. One of those essential programs is Heat Source Technologies, a group within the Actinide Material Power & Processing (AMPP) Division at TA-55. Within the Associate Laboratory Directorate of Weapons Production (ALDWP), the group responsible for the Pu-238 heat sources continued their work nearly uninterrupted as many others made the mandatory transition to work remotely. This group, known as AMPP-1, Heat Source Technologies, works on the production line producing heat sources from Plutonium (Pu)-238, which is an essential mission for National Security. This group is comprised of nearly 90 employees and operates out of the TA-55 Plutonium Facility (PF-4 and PF-5) producing heat sources. Heat source production includes the operations that purify, process, and weld Pu-238 oxide into encapsulated heat sources that are essential for space and defense missions.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗
NASA NTRS1997

Feasibility Investigation for Performing Fireball Temperature Tests

NASA Johnson Space Center White Sands Test Facility (WSTF) was requested by the Launch Abort Subpanel and the Power Systems Subpanel of the Interagency Nuclear Safety Review Panel to investigate the feasibility of using spectroscopic techniques to measure propellant fireball gas temperatures. This report outlines the modeling and experimental approaches and results of this investigation. Descriptions of the theoretical particle temperature and mass effusion models are presented along with the results of the survivability of small plutonium dioxide (less than or equal to 1000 microns diameter) particles entrained in various propellant fireball scenarios. The experimental test systems used to measure the hydroxide radical, water, and particle graybody spectral emissions and absorptions are discussed. Spectral results along with temperatures extracted by analyzing the spectral features are presented for the flames investigated in the laboratory environment. Methods of implementing spectroscopic measurements for future testing using the WSTF Large-scale Hydrogen/Oxygen Explosion Facility are discussed, and the accuracy expected for these measurements is estimated from laboratory measurements.

Tapphorn, Ralph M.↗
NASA NTRS2004

Thermal Energy Conversion Branch

The Thermal Energy Conversion Branch (5490) leads the way in designing, conducting, and implementing research for the newest thermal systems used in space applications at the NASA Glenn Research Center. Specifically some of the most advanced technologies developed in this branch can be broken down into four main areas: Dynamic Power Systems, Primary Solar Concentrators, Secondary Solar Concentrators, and Thermal Management. Work was performed in the Dynamic Power Systems area, specifically the Stirling Engine subdivision. Today, the main focus of the 5490 branch is free-piston Stirling cycle converters, Brayton cycle nuclear reactors, and heat rejection systems for long duration mission spacecraft. All space exploring devices need electricity to operate. In most space applications, heat energy from radioisotopes is converted to electrical power. The Radioisotope Thermoelectric Generator (RTG) already supplies electricity for missions such as the Cassini Spacecraft. The focus of today's Stirling research at GRC is aimed at creating an engine that can replace the RTG. The primary appeal of the Stirling engine is its high system efficiency. Because it is so efficient, the Stirling engine will significantly reduce the plutonium fuel mission requirements compared to the RTG. Stirling is also being considered for missions such as the lunar/Mars bases and rovers. This project has focused largely on Stirling Engines of all types, particularly the fluidyne liquid piston engine. The fluidyne was developed by Colin D. West. This engine uses the same concepts found in any type of Stirling engine, with the exception of missing mechanical components. All the working components are fluid. One goal was to develop and demonstrate a working Stirling Fluidyne Engine at the 2nd Annual International Energy Conversion Engineering Conference in Providence, Rhode Island.

Bielozer, Matthew C.↗
DOE OSTI2022

Final Design for Thermal/Epithermal eXperiments using High 240 Pu Content Plutonium/Aluminum Zero Power Research Reactor Plates with Polyethylene Moderator (IER 520 Final Design CED-2 Report)

The US Department of Energy Nuclear Criticality Safety Program (NCSP) convened a multinational Thermal Epithermal eXperiments (TEX) meeting in July of 2011 to discuss the data and experimental needs of criticality safety practitioners. The number one and two priority integral experiment data needs were for 239 Pu and 240 Pu, with special emphasis on cross section performance in the intermediate energy range (from 0.625 eV to 100 keV). LLNL measured five critical configurations with LANL for the plutonium test bed (IER-184) and published the experiments as International Criticality Safety Benchmark Evaluation Project evaluation PUMET-MIXED-002. Modeling of the benchmark configurations using ENDF/B-VIII.0 nuclear data showed significant overprediction of reactivity for configurations that had a large percentage of fissions in the intermediate energy regime. This report documents a variation on the TEX plutonium test bed to provide a test of 240 Pu cross sections, with sensitivity of the configuration to 240 Pu radiative capture and fission cross sections a priority for the design.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS2003

Neutron Measurements for Radiation Protection in Low Earth Orbit - History and Future

The neutron environment inside spacecraft has been of interest from a scientific and radiation protection perspective since early in the history of manned spaceflight. With 1:.1e exception of a few missions which carried plutonium-fueled radioisotope thermoelectric generators, all of the neutrons inside the spacecraft are secondary radiations resulting from interactions of high-energy charged particles with nuclei in the Earth's atmosphere, spacecraft structural materials, and the astronaut's own bodies. Although of great interest, definitive measurements of the spacecraft neutron field have been difficult due to the wide particle energy range and the limited available volume and power for traditional techniques involving Bonner spheres. A multitude of measurements, however, have been made of the neutron environment inside spacecraft. The majority of measurements were made using passive techniques including metal activation fo ils, fission foils, nuclear photoemulsions, plastic track detectors, and thermoluminescent detectors. Active measurements have utilized proton recoil spectrometers (stilbene), Bonner Spheres eRe proportional counter based), and LiI(Eu)phoswich scintillation detectors. For the International Space Station (ISS), only the plastic track! thermoluminescent detectors are used with any regularity. A monitoring program utilizing a set of active Bonner spheres was carried out in the ISS Lab module from March - December 200l. These measurements provide a very limited look at the crew neutron exposure, both in time coverage and neutron energy coverage. A review of the currently published data from past flights will be made and compared with the more recent results from the ISS. Future measurement efforts using currently available techniques and those in development will be also discussed.

Golightly, M. J.↗
DOE OSTI2024

Pu-238 Production Progress at INL From December 2022 to December 2023

Idaho National Laboratory (INL) has an ongoing effort to produce Pu-238 for NASA deep space missions. Recent work at INL has consisted of irradiation of Pu 238 production targets in the Advanced Test Reactor (ATR), generating more than an estimated 400 grams of Pu-238 heat source material between March and October 2023. Additionally, INL began qualifying Pu-238 production targets with a higher loading of Np-237 to further increase the production of Pu-238 production in later years. INL has also updated the analysis as a result of operational changes at ATR. One instance was updating the analysis to enable a single Pu-238 production target to be run in ATR’s South Flux Trap (SFT), rather than the previous seven production targets, to make use of a spare target from a discontinued design. Higher lobe powers in ATR were also analyzed due to potential changes in planned lobe powers.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Protecting People and Planet

To produce energy, civilian nuclear power plants require fuel. The majority of existing plants rely on solid ceramic fuels that contain concentrations of about 3 to 5% of the element uranium-235. This specific isotope is well suited to sustaining the type of chain reaction required for a nuclear reactor to harness the heat produced by nuclear fission. Title I of the United States Atomic Energy Act of 1954 defines “special nuclear material” as plutonium, uranium-233, or uranium enriched in the isotopes uranium-233 or uranium-235. This label identifies materials that at high concentrations can be used to produce nuclear weapons. The United States and broader international community are fully committed to ensuring the peaceful use of nuclear energy by committing to practices known as safeguards.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2020

Preliminary Evaluation of Driver Fuel Plates and Reduced Sodium Temperature to Increase Fast Test Reactor Experimental Neutron Irradiation Capabilities

All sodium-cooled fast test reactors operated so far have used essentially the same fuel geometrical ar-rangement, namely cylindrical fuel pins assembled into a triangular array in hexagonal fuel assemblies. Fuel compositions and dimensions may vary but the overall geometrical pattern remains identical. Fur-thermore, most sodium-cooled reactors have operated, and still operate, with core-average inlet and outlet sodium temperatures of, respectively, 350-400°C and 500-550°C. The present preliminary study revisits the standard fuel assembly geometry as well as operating tempera-tures to evaluate potential impact on fast test reactor neutron irradiation capabilities. The reference fuel is U-20Pu-10Zr. Pin (reference) and plate geometries are considered together with two sets of core-average sodium inlet/outlet temperatures: 350/500°C (reference) and 200/400°C. More prototypic environments could be obtained inside ad-hoc irradiation testing vehicles that are independent from the reactor primary coolant. Everything else being the same, fuel temperature is significantly lower in plates than in pins. Therefore, plates could in principle accommodate higher plutonium content, thus, reducing the need for uranium enrichment or allowing the use of lower quality plutonium (U-xPu-10Zr thermal conductivity and solidus temperature decrease as x increases, hence, necessitating additional thermal margins). For the reference pin configuration, lowering the inlet/outlet sodium temperatures from the reference 350/500°C down to 200/400°C provides additional thermal margins that can be used to increase the peak fast flux from about 4.5 × 10 15 n/cm 2 -s to 6 ×10 15 n/cm 2 -s for the same core power of 300 MW. Assuming 300 Equivalent Fuel Power Days (EFPD) of operation per calendar year, a (steel) test article could accu-mulate up to 75 dpa/year. The use of fuel plates provides even more thermal margins which may allow the peak fast flux to reach values as high as 8 × 10 15 n/cm 2 -s for both sets of inlet/outlet temperatures. In this environment, a test article could accumulate up to 95 dpa/year assuming 300 EFPD/year, hence greatly accelerating irradiation testing. Allowing a fast test reactor to operate over a wide range of inlet/outlet temperatures could add significant flexibility to its neutron irradiation capabilities. Steel present in the driver fuel assemblies is not expected to accumulate more than 90 displacements per atom (dpa) over its lifetime. Hence, steel alloys that are not suited for very high dpa but can operate over a wide range of temperatures (from low to high) would be appropriate for this application. Finally, because pin and plate assemblies fit on the same grid plate, a fast test reactor could in principle start-up with standard fuel pin assemblies and, later on, move to a plate-type Mark-II fuel to further in-crease its neutron irradiation capabilities while potentially reducing the level of uranium enrichment needed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗