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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 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

040226_AnnularFuel_FuelPerformance_LRSv2

This slide deck summarizes a comparative fuel performance assessment of four annular fuel concepts for pressurized water reactor applications against a reference 17×17 fuel design. Overall, annular fuel geometry improves several key thermal-mechanical performance metrics, with the large-annulus configuration providing the most favorable balance of benefits. Relative to the reference design, annular fuel reduces peak fuel temperature by lowering thermal resistance through a thinner fuel pellet and smaller heat-conduction path. At the same time, the reduced fuel volume raises local burnup and correspondingly increases fission gas release. Despite this, rod internal pressure at end of life decreases because the annular geometry provides greater internal void volume. Annular designs also tend to delay fuel-cladding gap closure, although this advantage diminishes in thin-gap configurations where the smaller initial gap accelerates closure. Similarly, hoop stress and hoop strain are generally reduced for annular fuel, but both increase as the initial gap becomes smaller, indicating a potential cladding integrity concern for thin-gap designs. Corrosion performance shows no meaningful variation among the concepts considered. Taken together, the results indicate that annular fuel can offer significant performance advantages in a PWR environment, with large-annulus designs emerging as the strongest candidate while thin-gap geometries introduce more challenging mechanical margins.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

A Combinatorial Approach to Testing and Characterization of Irradiated Fuels and Materials

•Combinatorial approaches to testing and characterization of materials is a very powerful tool to help explore some fundamentals of materials science.? •Three topics will be explored to help illustrate this:? •A fitness-for-service assessment of a CANDU component (this is work from my past)? •The use of small-scale mechanical testing of Cr-coated ATF cladding in support of qualification (this is work developed upon joining INL)? •A multi-scale characterization approach of TRISO fuel to investigate fission gas migration and microstructural features (this is work I now get to participate in on joining MFC)? •?

36 - MATERIALS SCIENCE↗
DOE OSTI2019

Fission Product Measurement and Off-Gas Scrubbing for Molten Salt Reactors

The Advanced Reactor Technology (ART) program within the US Department of Energy is developing understanding and technology to help facilitate molten salt reactor commercialization. The off-gas system for MSRs plays an important role in fission product confinement. The strategy used in MSRs is different from that used in pressurized water reactors, where the off-gas mainly mitigates severe accident conditions. Although the details vary depending on the reactor type, the off-gas system of MSRs must reduce radiological emissions to the environment to levels comparable with those mandated for non-MSRs. Systems with molten fuel salt will accumulate material in the headspace: volatile fission products, entrained salt particles, graphite fines, and material that has low solubility in the salt. Operators of thermal fluoride salt reactors with encapsulated fuel will also be concerned about tritium generation and migration. The fission and activation products include noble gases, their daughters, tritium, and aerosolized salt constituents. Their concentrations will generally increase with reactor burnup, depending on the various strategies used for mitigation. Additionally, tritium capture and removal has been reviewed by several groups.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1993

Advanced nuclear thermal propulsion concepts

In 1989, a Presidential directive created the Space Exploration Initiative (SEI) which had a goal of placing mankind on Mars in the early 21st century. The SEI was effectively terminated in 1992 with the election of a new administration. Although the initiative did not exist long enough to allow substantial technology development, it did provide a venue, for the first time in 20 years, to comprehensively evaluate advanced propulsion concepts which could enable fast, manned transits to Mars. As part of the SEI based investigations, scientists from NASA, DoE National Laboratories, universities, and industry met regularly and proceeded to examine a variety of innovative ideas. Most of the effort was directed toward developing a solid-core, nuclear thermal rocket and examining a high-power nuclear electric propulsion system. In addition, however, an Innovative Concepts committee was formed and charged with evaluating concepts that offered a much higher performance but were less technologically mature. The committee considered several concepts and eventually recommended that further work be performed in the areas of gas core fission rockets, inertial confinement fusion systems, antimatter based rockets, and gas core fission electric systems. Following the committee's recommendations, some computational modeling work has been performed at Los Alamos in certain of these areas and critical issues have been identified.

Howe, Steven D.↗
NASA NTRS1973

Fission track astrology of three Apollo 14 gas-rich breccias

The three Apollo 14 breccias 14301, 14313, and 14318 all show fission xenon due to the decay of Pu-244. To investigate possible in situ production of the fission gas, an analysis was made of the U-distribution in these three breccias. The major amount of the U lies in glass clasts and in matrix material and no more than 25% occurs in distinct high-U minerals. The U-distribution of each breccia is discussed in detail. Whitlockite grains in breccias 14301 and 14318 found with the U-mapping were etched and analyzed for fission tracks. The excess track densities are much smaller than indicated by the Xe-excess. Because of a preirradiation history documented by very high track densities in feldspar grains, however, it is impossible to attribute the excess tracks to the decay of Pu-244. A modified track method has been developed for measuring average U-concentrations in samples containing a heterogeneous distribution of U in the form of small high-U minerals. The method is briefly discussed, and results for the rocks 14301, 14313, 14318, 68815, 15595, and the soil 64421 are given.

Graf, H.↗
NASA NTRS1976

Thermodynamic properties of a high pressure subcritical UF6/He gas volume (irradiated by an external source)

A computer simulation study concerning a compressed fissioning UF6 gas is presented. The compression is to be achieved by a ballistic piston compressor. Data on UF6 obtained with this compressor were incorporated in the simulation study. As a neutron source to create the fission events in the compressed gas, a fast burst reactor was considered. The conclusion is that it takes a neutron flux in excess of 10 to the 15th power n/sec sq cm to produce measurable increases in pressure and temperature, while a flux in excess of 10 to 19th power n/sq cm sec would probably damage the compressor.

Sterritt, D. E.↗
DOE OSTI2023

Characterization of a platform for the gas transport, collection, and identification of fission products in the high-intensity laser environment

Recent progress in the production of laser accelerated high flux proton beams opens new possibilities for the study of fission in unique environments. To this end, we are currently pursuing the development of a platform for the swift gas transport, collection, and identification of fission products. Fission is induced in targets fixed inside a sealed chamber through which a carrier gas flows, transporting fission products to a carbon filter for collection and online spectroscopy. This has been recently demonstrated using target normal sheath accelerated protons at the PHELIX laser facility. There were large discrepancies between measured rates and those expected based on established fission yields and measured beam parameters. These discrepancies prompted a large number of tests at the Idaho Accelerator Center (IAC), where fission in uranium is induced by bremsstrahlung photons generated by 21 MeV electrons from a linear electron accelerator. Here, the results are compared to a series of models that account for the slowing of energetic fission fragments in the carrier gas and the fluid dynamics of the gas flow that transports fission products to the collection filter. Characterization of the apparatus reveals a few mechanisms that together account for a portion of the previously observed discrepancies at the PHELIX laser facility. However, additional research is necessary before high-accuracy experiments can be performed with the apparatus. Work performed under the auspices of the U.S. Department of Energy by LLNL under contract DE-AC52-07NA27344.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS2006

Multi-Megawatt Gas Turbine Power Systems for Lunar Colonies

A concept for development of second generation 10 MWe prototype lunar power plant utilizing a gas cooled fission reactor supplying heated helium working fluid to two parallel 5 MWe closed cycle gas turbines is presented. Such a power system is expected to supply the energy needs for an initial lunar colony with a crew of up to 50 persons engaged in mining and manufacturing activities. System performance and mass details were generated by an author developed code (BRMAPS). The proposed pilot power plant can be a model for future plants of the same capacity that could be tied to an evolutionary lunar power grid.

Juhasz, Albert J.↗
NASA NTRS2005

A molten Salt Am242M Production Reactor for Space Applications

The use of Am242m holds great promise for increasing the efficiency nuclear thermal rocket engines. Because Am242m has the highest fission cross section of any known isotope (1000's of barns), its extremely high reactivity may be used to directly heat a propellant gas with fission fragments. Since this isotope does not occur naturally, it must be bred in special production reactors designed for that purpose. The primary advantage to using molten salt reactors for breeding Am242m is that the reactors can be reprocessed continually yielding a constant rate of production of the isotope. Once built and initially fueled, the reactor will continually breed the additional fuel it needs to remain critical. The only feedstock required is a salt of U238. No enriched fuel is required during normal operation and all fissile material, except the Am242m, is maintained in a closed loop. For a reactor operating at 200 MW several kilograms of Am242m may be bred each year.

Emrich, William↗
NASA NTRS1976

Research on plasma core reactors

Experiments and theoretical studies are being conducted for NASA on critical assemblies with one-meter diameter by one-meter long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17 cm thick by 89 cm diameter beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials were measured. Finally, an 85,000 cu cm aluminum canister in the central region was fueled with UF6 gas and fission density distributions determined. These results are to be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation.

Jarvis, G. A.↗
DOE OSTI2020

Examination of Legacy Metallic Fuel Pins (U-10Zr) Tested in FFTF

The MFF series of metallic fuel (U-10wt%Zr) tests performed in the Fast Fuel Test Facility (FFTF) were the beginning tests to qualify the fuel as a driver fuel for FFTF. They all performed very well, to relatively high burnup and with no pin breaches. Tests MFF-3 and MFF-5 were chosen to be destructively examined because they were run at high peak cladding temperatures, 643 and 649°C respectively, and to modest to high burnup (138 and 101 MWd/kgM respectively). They were the only sodium fast reactor (SFR) metallic tests, with long (91.4 cm) fuel columns and clad in HT9, to be operated at these high temperatures. Extensive operating condition reconstruction was performed to accompany the examination, providing detailed operating conditions axially along the pins. In addition, detailed ORIGEN calculations provided fuel burnup along the pin length. These calculations and detailed cross-section metallography allowed a PhD dissertation to be performed where the fuel/cladding chemical interaction (FCCI) was modeled using Fickian and Soret Effect driven diffusion of rare earth fission products to the cladding surface and interact with the cladding. The examination of the MFF-3 and MFF-5 pins included neutron radiography, axial profilometry including pin bow and length measurements, precision gamma scanning, plenum gas puncturing to measure fission gas release, detailed chemical/isotopic analyses of fuel samples to confirm burnup calculations, and metallography of pin cross-sections, including micro-hardness measurements of fuel and cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Quantity of 135I Released from the AGR 5/6/7 Experiment

A series of four Advanced Reactor Technologies (ART) experiments have been conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). From 2006 through 2020, these experiments supported the development and qualification of the new U.S. tristructural isotropic (TRISO) particle fuel for Very High Temperature Reactors (VHTR). Each Advanced Gas Reactor (AGR) experiment consisted of multiple fueled capsules, each plumbed for independent temperature control using a mix of helium and neon gases. The gas leaving a capsule was routed to individual Fission Product Monitor (FPM) detectors. For intact fuel particles, the TRISO particle coatings provide a substantial barrier to fission product release. However, particles with failed coatings, whether because of a minute percentage of initially defective particles, those which fail during irradiation, or those designed to fail (DTF) particles, can release fission products to the flowing gas stream. Because reactive fission product elements like iodine and cesium quickly deposit on cooler capsule components and piping structures as the effluent gas leaves the reactor core, only the noble fission gas isotopes of Kr and Xe tend to reach FPM detectors. The FPM system utilizes High Purity Germanium (HPGe) detectors coupled with a thallium activated sodium iodide NaI(Tl) scintillator. The HPGe detector provides individual isotopic information, while the NaI(Tl) scintillator is used as a gross count rate meter. During irradiation, the 135mXe concentration reaching the FPM detectors is from both direct fission and by decay of the accumulated 135I. About 2.5 hours after irradiation (ten 15.3 minute 135mXe half lives) the directly produced 135mXe has decayed and only the longer lived 135I remains as a source. Decay systematics dictate that 135mXe will be in secular equilibrium with its 135I parent, such that its production rate very nearly equals the decay rate of the parent, and its concentration in the flowing gas stream will appear to decay with the parent half life. This equilibrium condition enables the determination of the amount of 135I released from the fuel particles by measurement of the 135mXe at the FPM following reactor shutdown. In this paper, the 135I released will be reported.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2009

A Summary of Closed Brayton Cycle Development Activities at NASA

NASA has been involved in the development of Closed Brayton Cycle (CBC) power conversion technology since the 1960's. CBC systems can be coupled to reactor, isotope, or solar heat sources and offer the potential for high efficiency, long life, and scalability to high power. In the 1960's and 1970's, NASA and industry developed the 10 kW Brayton Rotating Unit (BRU) and the 2 kW mini-BRU demonstrating technical feasibility and performance, In the 1980's, a 25 kW CBC Solar Dynamic (SD) power system option was developed for Space Station Freedom and the technology was demonstrated in the 1990's as part of the 2 kW SO Ground Test Demonstration (GTD). Since the early 2000's, NASA has been pursuing CBC technology for space reactor applications. Before it was cancelled, the Jupiter Icy Moons Orbiter (HMO) mission was considering a 100 kWclass CBC system coupled to a gas-cooled fission reactor. Currently, CBC technology is being explored for Fission Surface Power (FSP) systems to provide base power on the moon and Mars. These recent activities have resulted in several CBC-related technology development projects including a 50 kW Alternator Test Unit, a 20 kW Dual Brayton Test Loop, a 2 kW Direct Drive Gas Brayton Test Loop, and a 12 kW FSP Power Conversion Unit design.

Mason, Lee S.↗
DOE OSTI2025

Understanding grain refinement and intergranular gas bubble evolution in U-10Mo fuel using phase-field modeling

Monolithic U-10Mo fuel undergoes significant microstructural changes in the form of grain refinement and gas bubble formation during burnup, which degrades its mechanical properties. In this talk, I present a phase-field model for microstructure evolution in U-10Mo developed using the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework. Simulations demonstrate that grain refinement initiates at pre-existing grain boundaries (GBs) to eliminate the lattice distortion energy caused by the accumulation of self-interstitial loops. By employing an equation of state for xenon gas, we simulate the evolution of gas bubbles in the polycrystal microstructure. Large, interconnected bubbles are found to form along the triple junctions. The effects of defect production rate, diffusivities and GB mobility on the microstructure evolution are systematically studied. Homogenization is employed on the microstructures to obtain effective elastic constants and diffusivity as a function of fission density. The simulations provide critical insights on microstructure and property degradation in U-10Mo fuel.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1963

OUTLOOK FOR GAS-CORE NUCLEAR ROCKETS

Fission energy transferal, hydrogen propellant flow, and adequate fissionable material are main factors in gas-core nuclear rocket design concepts

GASEOUS FISSION REACTOR↗