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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 OSTI2024

Removal and Deactivation of Bond Sodium from Fast Reactor Blanket Materials

The disposition of sodium-bonded spent nuclear fuel and blanket materials in a repository is complicated by the presence of sodium metal that is used as a thermal bond between the uranium metal fuel and blanket slugs and their cladding. The concern is that the metallic sodium could react with water, producing explosive hydrogen gas, or could exhibit a pyrophoric character. Thus, experimental studies were performed to investigate and demonstrate the removal and deactivation of bond sodium from blanket material in a dry environment. Specifically, bond sodium was removed from unirradiated Fermi-1 blanket elements and an assembly via a melt-drain-evaporate process using elevated temperature and reduced pressure. The effectiveness of sodium metal removal from the blanket materials and their associated cladding was =99.9998%, based on post-test quantitative analyses. The separated sodium metal was collected and subsequently deactivated by reacting it in a molten state with a controlled addition of ammonium chloride particles atop a molten salt medium. In this process sodium chloride is formed and assimilates into the salt pool. The subsequent deactivation of the bond sodium produced a solid ingot of sodium chloride, potassium chloride, lithium chloride, and cesium chloride that was devoid of sodium metal based on post-test analyses of the salt product. Both the sodium removal and deactivation operations were conducted within a dry inert atmosphere enclosure. The results of this study substantiate a path forward for the disposition of sodium-bonded blanket materials, including 34 metric tons heavy metal in irradiated Fermi-1 blanket material currently stored at Idaho National Laboratory.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1991

An analysis of LDEF-exposed silvered FEP teflon thermal blanket material

The characterization of selected silvered fluorinated ethylene propylene (FEP) teflon thermal blanket material which received 5 years and 9 months of exposure to the LEO environment on the Long Duration Exposure Facility is reported. X-ray photoelectron spectroscopy, infrared, and thermal analyses did not detect a significant change at the molecular level as the result of this exposure. However, various microscopic analyses revealed a roughening of the coating surface due to atomic oxygen erosion which resulted in some materials changing from specular reflectors of visible radiation to diffuse reflectors. The potential effect of silicon-containing molecular contamination on these materials is addressed.

Young, Philip R.↗
NASA NTRS1991

An investigation of the degradation of Fluorinated Ethylene Propylene (FEP) copolymer thermal blanketing materials aboard LDEF in the laboratory

Samples of fluorinated ethylene propylene copolymer thermal blanketing material, recovered from the Long Duration Exposure Facility (LDEF), were investigated to determine the nature and the extent of degradation due to exposure to the low-Earth-orbit environment. Samples recovered from the ram-facing direction of LDEF, which received vacuum-ultraviolet (VUV) radiation and atomic-oxygen impingement, and samples from the trailing edge, which received almost exclusively VUV exposure, were investigated by scanning electron microscopy and atomic force microscopy. The most significant result of this investigation was found on samples that received only VUV exposure. These samples possessed a hard, embrittled surface layer that was absent from the atomic-oxygen exposed sample and from unexposed control samples. This surface layer is believed to be responsible for the 'synergistic' effect between VUV and atomic oxygen. Overall, the investigation revealed dramatically different morphologies for the two samples. The sample receiving both atomic-oxygen and VUV exposure was deeply eroded and had a characteristic 'rolling' surface morphology, while the sample that received only VUV exposure showed mild erosion and a surface morphology characterized by sharp high-frequency peaks. The morphologies observed in the LDEF samples, including the embrittled surface layer, were successfully duplicated in the laboratory.

Stiegman, A. E.↗
NASA NTRSDate not supplied

Preliminary experimental studies into the storage capacity of cryogenic hydrogen in aerogel blanket materials

The abundance and diversity of hydrogen applications necessitates continued and accelerated research into advanced storage technologies. Traditionally, hydrogen has been stored as either a high-pressure, warm gas; or a low-pressure, cryogenic liquid. Methods such as cryo-supercritical and cryo-adsorbed have been explored, but are not yet mainstream. Cryo-adsorbed is attractive because higher storage densities at higher temperatures than liquid may be achieved. Recently NASA, in partnership with Eta Space, Southwest Research Institute, the University of Central Florida, and Air Liquide, have been exploring the use of inexpensive, commercially available silica aerogel blanket materials for cryo-adsorbed hydrogen storage. Unlike most adsorbents, aerogel blanket is not a powder, but a robust, composite material that can be formed into complex shapes to aid in more efficient storage system designs, and has already been proven to uptake large quantities of fluids such as nitrogen and oxygen. Recent experimental efforts into the uptake of low-pressure hydrogen gas at 77 K, and liquid hydrogen at normal boiling point (NBP) will be discussed. Although preliminary in nature, the test results are promising, showing up to a 49% increase in storage density at 77 K over the gas alone, and greater than a one-to-one volume equivalency with NBP LH 2 .

A M Swanger↗
NASA NTRS1990

Vacuum ultraviolet radiation and thermal cycling effects on atomic oxygen protective photovoltaic array blanket materials

The importance of synergistic environmental exposure is demonstrated through the evaluation of DuPont 93-1 in simulated LEO environment. Changes in optical properties, surface condition, and mass loss data are described. The qualitative results indicate the necessity for exposure of materials to a series of simulated LEO environments in order to properly determine synergistic effects and demonstrate the overall LEO durability of candidate materials. It is shown that synergistic effects may occur with vacuum thermal cycling combined with VUV radiation followed by atomic oxygen exposure. Testing the durability of candidate solar array blanket materials in a test sequence with necessary synergistic effects makes it possible to determine the appropriate material for providing structural support and maintaining the proper operating temperature for solar cells in the SSF Photovaltaic Power System.

Brady, J.↗
NASA NTRS1982

Thermal insulation blanket material

A study was conducted to provide a tailorable advanced blanket insulation based on a woven design having an integrally woven core structure. A highly pure quartz yarn was selected for weaving and the cells formed were filled with a microquartz felt insulation.

Pusch, R. H.↗
DOE OSTI2022

Synergies between H, He and radiation damage in dual and triple ion irradiation of candidate fusion blanket materials

Three ferritic/martensitic alloys were studied to understand the synergistic effect between single ion beam (Fe 2+ ), dual ion beam (Fe 2+ +He 2+ and Fe 2+ +H + ), and triple ion beam (Fe 2+ +He 2+ +H + ) irradiations on cavity evolution. A commercial alloy, F82H, a castable nanostructured alloy, CNA3, and a high purity model alloy, Fe8Cr2W, were irradiated at 400°C to 600°C to a damage level of 50 dpa at a damage rate of 1 × 10 –3 dpa/s with He and H injection rates of 10 and 40 appm/dpa, respectively. Post-irradiation characterization via bright field transmission electron microscopy and high-angle annular dark-field scanning transmission electron microscopy was performed on all irradiated conditions to characterize the cavity size distribution and determine the effects of H/He injection on cavity microstructure. In all three alloys, hydrogen co-injection with helium resulted in an increased cavity number density and maximum cavity size, producing an increase in swelling over that from helium injection alone. Swelling in F82H appears to peak between 450°C and 500°C. At 600°C, swelling was minimal and cavities of high density and small size were confined to grain boundaries and dislocations while at 400°C, swelling is also low with a nearly homogeneous, high density, distribution of very small cavities throughout. Swelling was least in the commercial alloy F82H due to the high sink strength. The CNA3 alloy underwent dissolution of precipitates that lowered the sink strength and resulted in higher swelling than F82H, but less than the model alloy. Electron energy loss spectroscopy (EELS) elemental mapping revealed hydrogen forming a halo-like structure about the periphery of the cavities and helium residing within the cavities themselves. Finally, this observation suggests that hydrogen reduces the surface energy of helium-filled cavities which results in both increased cavity number density and cavity size in triple beam irradiation over dual beam irradiation.

36 MATERIALS SCIENCE↗
NASA NTRS1991

Durability evaluation of photovoltaic blanket materials exposed on LDEF tray S1003

Several candidate protective coatings on Kapton and uncoated Kapton were exposed to the LEO environment on the LDEF in order to determine whether the coatings could be used to protect polymeric substrates from degradation in the LEO environment. These materials are used for flexible solar array panels in which the polymer is the structural member that supports the solar cell and current carriers. Arrays such as these are used on the Hubble Space Telescope and will be used on Space Station Freedom. The results of the experiments are presented.

Rutledge, Sharon K.↗
NASA NTRS1992

Durability evaluation of photovoltaic blanket materials exposed on LDEF tray S1003

Several candidate protective coatings on Kapton and uncoated Kapton were exposed to the low Earth orbital (LEO) environment on the Long Duration Exposure Facility (LDEF) to determine if the coatings could be used to protect polymeric substrates from degradation in the LEO environment. The coatings that were evaluated were 700 A of aluminum oxide, 650 A of silicon dioxide, and 650 A of a 4 percent polytetrafluoroethylene-96 percent silicon dioxide mixed coating. All of the coatings evaluated were ion beam sputter deposited. These materials were exposed to a very low atomic oxygen fluence (4.8 x 10 exp 19 atoms/sq. cm) as a result of the experiment tray being located 98 degrees from the ram direction. As a result of the low atomic oxygen fluence, determination of a change in mass was not possible for any of the samples including the uncoated Kapton. There was no evidence of spalling of any of the coatings after the approximately 33,600 thermal cycles recorded for LDEF. The surface of the uncoated Kapton, however, did show evidence of grazing incidence texturing. There was a 7-8 percent increase in solar absorptance for the silicon dioxide and aluminum oxide coated Kapton and only a 4 percent increase for the mixed coating. It appears that the addition of a small amount of fluoropolymer may reduce the magnitude of absorptance increase due to environmental exposure. Thermal emittance did not change significantly for any of the exposed samples. Scanning electron microscopy revealed few micrometeoroid or debris impacts, but the impact sites found indicated that the extent of damage or cracking of the coating around the defect site did not extend beyond a factor of three of the impact crater diameter. This limiting of impact damage is of great significance for the durability of thin film coatings used for protection against the LEO environment.

Sharon K Rutledge↗
NASA NTRS2009

Aerogel Blanket Insulation Materials for Cryogenic Applications

Aerogel blanket materials for use in thermal insulation systems are now commercially available and implemented by industry. Prototype aerogel blanket materials were presented at the Cryogenic Engineering Conference in 1997 and by 2004 had progressed to full commercial production by Aspen Aerogels. Today, this new technology material is providing superior energy efficiencies and enabling new design approaches for more cost effective cryogenic systems. Aerogel processing technology and methods are continuing to improve, offering a tailor-able array of product formulations for many different thermal and environmental requirements. Many different varieties and combinations of aerogel blankets have been characterized using insulation test cryostats at the Cryogenics Test Laboratory of NASA Kennedy Space Center. Detailed thermal conductivity data for a select group of materials are presented for engineering use. Heat transfer evaluations for the entire vacuum pressure range, including ambient conditions, are given. Examples of current cryogenic applications of aerogel blanket insulation are also given. KEYWORDS: Cryogenic tanks, thermal insulation, composite materials, aerogel, thermal conductivity, liquid nitrogen boil-off

Coffman, B. E.↗
DOE OSTI2023

Tritium Breeding Ratio Evaluation of Solid Breeder Concepts for the FESS-FNSF

This paper presents a parametric study of the Fusion Energy System Studies-Fusion Nuclear Science Facility’s (FNSF’s) tritium breeding performance for several solid breeder concepts, neutron multiplying materials, and blanket materials, assuming volume fractions based on the most recent FNSF design as a realistically representative fusion facility. In this study, we initially surveyed the tritium breeding ratio (TBR) of several solid breeder concepts by employing a simplified but efficient one-dimensional (1-D) infinite cylinder reduced-order model (ROM). Parametric studies were performed with the ROMs for the full range of breeder-to-multiplier ratios to identify the optimum mixture compositions for each breeder type that would lead to a maximum TBR. These optimized breeder-multiplier combinations were then homogenized with FNSF blanket component materials to estimate their impacts on the TBR. Subsequently, as a validation step for the optimal designs, TBR calculations were performed using a more realistic modified 1-D ROM with inner and outer breeding regions, as well as with a fully detailed 22.5-deg three-dimensional (3-D) sector of the FNSF to assess the impact of geometry details on the TBR. The differences between the two 1-D models were negligible, while the ROMs were able to correctly predict trends and identify the maximum and minimum TBR cases, as well as show consistent biases relative to the results produced by the full 3-D, 22.5-deg sector for specific breeder/multiplier combinations. Solid breeder concepts such as Li 2 O, Li 4 SiO 4 , and Li 8 ZrO 6 outperformed all others in this study in terms of TBR performance when combined with all the neutron multiplier materials selected. Here, an underlying goal of this study was to develop and improve rapid and reliable ROMs to aid designers during parametric optimizations of highly complex and computationally expensive fusion models.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

Estimate of Gamma Dose Rates from Arrays of Fermi-1 Blanket Elements During the MEDE Process

The Enrico Fermi Atomic Power Plant Unit 1 (Fermi-1) was a sodium-cooled fast breeder reactor located in Monroe County, Michigan. The reactor was powered by a core of enriched uranium metal alloy driver fuel, which was enveloped by an axial and radial blanket material consisting of depleted uranium metal alloyed with 3 wt.% molybdenum. There are 406 axial and 559 radial irradiated sodium-bonded Fermi-1 blanket assemblies in storage at INL, totaling 34 metric tons of heavy metal. Disposal of the Fermi-1 blanket material directly into a geological repository is prohibited due to the reactive characteristic of its bond sodium. A melt drain evaporate (MEDE) process can effectively remove bond sodium from Fermi-1 blanket material. Consequently, planning is underway to apply a MEDE process to treat the 34 metric tons of heavy metal of irradiated Fermi-1 blanket material. Given the irradiated Fermi-1 blanket material’s relatively low power history and 50+ years of decay, it can be treated in a shielded glovebox. To assess the requisite shielding, the dose rates of the Fermi-1 blanket elements in various process configurations are needed. The method to perform the dose rate calculations is to first generate an average source term and associated photon source spectra for the Fermi-1 blanket material in SCALE and then use the associated photon spectra to calculate dose rates using MCNP6.2 in seven representative geometries. Dose rates in rem/h were calculated on contact (1 cm from outer geometry surface), 30 cm away from outer geometry surface, and 1 m from outer geometry surface at axial heights spanning the length of the blanket material within the geometry. The maximum average dose rate for the single Fermi-1 radial blanket element is ~0.60 rem/h, for the Fermi-1 radial blanket assembly ~1.10 rem/h, for the single Fermi-1 axial blanket element ~0.5 rem/h, for the Fermi-1 axial blanket assembly ~1.01 rem/h, for the MEDE can ~ 0.98 rem/h, for the Fermi Storage Canister ~ 0.83 rem/h, and for the MEDE cans in the DOE Standard Canister ~0.51 rem/h. Based on these maximum average values, the bounding dose rates are assumed to be 3× the average dose rates calculated for the axial blanket material and 7.6× the average dose rates calculated for the radial blanket material, based on the distribution of Fermi-1 blanket material assembly burnup.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1992

Synergistic effects of ultraviolet radiation, thermal cycling, and atomic oxygen on altered and coated Kapton surfaces

The photovoltaic (PV) power system for Space Station Freedom (SSF) uses solar array blankets which provide structural support for the solar cells and house the electrical interconnections. In the low Earth orbital (LEO) environment where SSF will be located, surfaces will be exposed to potentially damaging environmental conditions including solar ultraviolet (UV) radiation, thermal cycling, and atomic oxygen. It is necessary to use ground based tests to determine how these environmental conditions would affect the mass loss and optical properties of candidate SSF blanket materials. Silicone containing, silicone coated, and SiO(x) coated polyimide film materials were exposed to simulated LEO environmental conditions to determine there durability and whether the environmental conditions of UV, thermal cycling and oxygen atoms act synergistically on these materials. A candidate PV blanket material called AOR Kapton, a polysiloxane polyimide cast from a solution mixture, shows an improvement in durability to oxygen atoms erosion after exposure to UV radiation or thermal cycling combined with UV radiation. This may indicate that the environmental conditions react synergistically with this material, and the damage predicted by exposure to atomic oxygen alone is more severe than that which would occur in LEO where atomic oxygen, thermal cycling and UV radiation are present together.

Dever, Joyce A.↗
NASA NTRS1992

Synergistic effects of ultraviolet radiation, thermal cycling and atomic oxygen on altered and coated Kapton surfaces

The photovoltaic (PV) power system for Space Station Freedom (SSF) uses solar array blankets which provide structural support for the solar cells and house the electrical interconnections. In the low earth orbital (LEO) environment where SSF will be located, surfaces will be exposed to potentially damaging environmental conditions including solar ultraviolet (UV) radiation, thermal cycling, and atomic oxygen. It is necessary to use ground based tests to determine how these environmental conditions would affect the mass loss and optical properties of candidate SSF blanket materials. Silicone containing, silicone coated, and SiO(x) coated polyimide film materials were exposed to simulated LEO environmental conditions to determine their durability and whether the environmental conditions of UV, thermal cycling and oxygen atoms act synergistically on these materials. A candidate PV blanket material called AOR Kapton, a polysiloxane polyimide cast from a solution mixture, shows an improvement in durability to oxygen atoms erosion after exposure to UV radiation or thermal cycling combined with UV radiation. This may indicate that the environmental conditions react synergistically with this material, and the damage predicted by exposure to atomic oxygen alone is more severe than that which would occur in LEO where atomic oxygen, thermal cycling and UV radiation are present together.

Dever, Joyce A.↗
DOE OSTI2021

Deuterium transport and retention properties of representative fusion blanket structural materials

Reduced activation ferritic-martensitic (RAFM) steels have been developed for decades for use as fusion blanket structural materials, and have advantages in both mechanical properties and irradiation resistance following careful engineering of the microstructure. However, the hydrogen isotope behavior in these proposed fusion structural materials is not well understood, but is important to assess since it impacts the fusion reactor safety and self-sufficient tritium fuel cycle. Here, we investigated deuterium transport and retention in representative advanced RAFM steels, including castable nanostructured alloys (CNAs), and oxide-dispersion-strengthened (ODS) steels. A gas-driven permeation (GDP) system was used to measure the permeability, diffusivity and solubility of the studied materials, covering the temperature range from 623 K to 873 K, and the loading pressures from 1.8 x 10 4 to 1.0 x 10 5 Pa. The results indicated that the deuterium permeability has little material dependence. In contrast, the deuterium diffusivity of the studied materials showed significant variation. The deuterium diffusivity in ODS steels is one order of magnitude lower than that in RAFM steels and CNAs, and correspondingly, have an effective solubility that is 2–10 times larger than RAFM steels and CNAs. In addition, thermal desorption spectroscopy (TDS) measurements were performed to assess the deuterium retention and desorption of these materials following a static thermal deuterium charging at 723 Kfor 1 hour under the deuterium pressure of 1.0 x 10 5 Pa. It was found that ODS steels exhibit the highest deuterium retention and have broader desorption peaks. Microstructural features contributing to deuterium retention and impacting deuterium transport are discussed to rationalize the observed deuterium behavior in the studied RAFM steels.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Relevance of Tritium Breeder Irradiation Testing in a Fusion Prototypic Neutron Source

A fusion prototypic neutron source (FPNS) will play a fundamental role in testing and developing fusion materials that can withstand the harsh fusion environment. We conducted a thorough assessment to determine whether blanket materials would benefit from FPNS testing. Because of the cross sections for the various blanket materials, lithium-containing materials can generally be tested adequately using existing fission irradiation capabilities instead of waiting for or relying solely on future FPNS experiments. However, non-lithium blanket materials, such as beryllium multipliers and flow channel inserts, should be tested under a fusion neutron spectrum.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗