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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 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
NASA NTRS2005

Heat Transfer and Thermal Stability Research for Advanced Hydrocarbon Fuel Technologies

In recent years there has been increased interest in the development of a new generation of high performance boost rocket engines. These efforts, which will represent a substantial advancement in boost engine technology over that developed for the Space Shuttle Main Engines in the early 1970s, are being pursued both at NASA and the United States Air Force. NASA, under its Space Launch Initiative s Next Generation Launch Technology Program, is investigating the feasibility of developing a highly reliable, long-life, liquid oxygen/kerosene (RP-1) rocket engine for launch vehicles. One of the top technical risks to any engine program employing hydrocarbon fuels is the potential for fuel thermal stability and material compatibility problems to occur under the high-pressure, high-temperature conditions required for regenerative fuel cooling of the engine combustion chamber and nozzle. Decreased heat transfer due to carbon deposits forming on wetted fuel components, corrosion of materials common in engine construction (copper based alloys), and corrosion induced pressure drop increases have all been observed in laboratory tests simulating rocket engine cooling channels. To mitigate these risks, the knowledge of how these fuels behave in high temperature environments must be obtained. Currently, due to the complexity of the physical and chemical process occurring, the only way to accomplish this is empirically. Heated tube testing is a well-established method of experimentally determining the thermal stability and heat transfer characteristics of hydrocarbon fuels. The popularity of this method stems from the low cost incurred in testing when compared to hot fire engine tests, the ability to have greater control over experimental conditions, and the accessibility of the test section, facilitating easy instrumentation. These benefits make heated tube testing the best alternative to hot fire engine testing for thermal stability and heat transfer research. This investigation used the Heated Tube Facility at the NASA Glenn Research Center to perform a thermal stability and heat transfer characterization of RP-1 in an environment simulating that of a high chamber pressure, regenerative cooled rocket engine. The first step in the research was to investigate the carbon deposition process of previous heated tube experiments by performing scanning electron microscopic analysis in conjunction with energy dispersive spectroscopy on the tube sections. This analysis gave insight into the carbon deposition process and the effect that test conditions played in the formation of deleterious coke. Furthermore, several different formations were observed and noted. One other crucial finding of this investigation was that in sulfur containing hydrocarbon fuels, the interaction of the sulfur components with copper based wall materials presented a significant corrosion problem. This problem in many cases was more life limiting than those posed by the carbon deposition process. The results of this microscopic analysis was detailed and presented at the December 2003 JANNAF Air-Breathing Propulsion Meeting as a Materials Compatibility and Thermal Stability Analysis of common Hydrocarbon Fuels (reference 1).

DeWitt, Kenneth↗
NASA NTRS1989

Corrosion testing of candidates for the alkaline fuel cell cathode

It is desirable to employ a corrosion screening test for catalyst or support candidates for the fuel cell cathode before entering upon optimization of the candidate or of the catalytic electrode. To this end, corrosion test electrodes, intended for complete immersion and maximum wetting, have been made with 30 to 40 vol. pct Teflon; with perovskites this is about 10 to 15 pct. The candidates were synthesized by methods intended for single-phase product without special emphasis on high surface area, although the substances tested were no coarser than 2 m squared/g. A typical loading was 25 mg/cm sq of the pure substance, usually on gold screen, a few mm squared of which were left bare for contacting. Contact to the gold lead wire was made by welding with a micro-torch or a spot-welder. Corrosion testing consisted of obtaining current-voltage data under flowing inert gas in the potential region for reduction of O2. The electrode was immersed in 30 pct KOH. Observations were made at 20 C and 80 C, and the results compared with data from gold standards. Results with some perovskites, pyrochlores, spinels, and interstitial compounds will be discussed.

Singer, Joseph↗
DOE OSTI2020

M4SF-20LL010302042: Process Model for Radionuclide Incorporation into Corrosion Products

This progress report (Level 4 Milestone Number M4SF-20LL010302042) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Crystalline Activity Number SF-20LL010302041. The research is focused on actinide and radionuclide sequestration in steel corrosion products. Fuel matrix degradation models suggest that the near field is likely to be reducing at the time of canister breaching, steel corrosion, and radionuclide release, but more oxidizing conditions may prevail in the far field. The incorporation of radionuclides into corrosion phases may limit the rate of radionuclide release by sequestering a portion of the radionuclide source term. For these reasons there is a need to evaluate the incorporation of Pu and other radionuclides into various Fe-oxide phases, and to understand the behavior of coprecipitated phases during mineral recrystallization processes and during re-oxidation events. Radionuclide coprecipitation with Fe minerals may impact long-term repository performance and is an ongoing research focus at Lawrence Livermore National Laboratory.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Steel Corrosion Impacts on Radionuclide Immobilization

This progress report (Level 4 Milestone Number M4SF-21LL010302052) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Crystalline Activity Number SF-21LL01030205. The research is focused on actinide and radionuclide sequestration in steel corrosion products. Fuel matrix degradation models suggest that the near field is likely to be reducing at the time of canister breaching, steel corrosion, and radionuclide release, but more oxidizing conditions may prevail in the far field. The incorporation of radionuclides into corrosion phases may limit the rate of radionuclide release by sequestering a portion of the radionuclide source term. For these reasons there is a need to evaluate the incorporation of Pu and other radionuclides into various Fe-oxide phases, and to understand the behavior of coprecipitated phases during mineral recrystallization processes and during re-oxidation events. Radionuclide coprecipitation with Fe minerals may impact long-term repository performance and is an ongoing research focus at Lawrence Livermore National Laboratory.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Review and Summary of Corrosion Behavior for Aluminum-Clad Spent Nuclear Fuel in Dry Storage

A description of the corrosion behavior of aluminum alloys used for cladding on aluminum-based, aluminum-clad nuclear fuel used in research reactors under potential dry storage conditions has been compiled. An evaluation is made of potential additional corrosion with water postulated to not be removed during drying. The relative humidity (RH) produced by typical dryness criteria for industrial spent nuclear fuel drying is expected to be low, particularly at high temperatures (17% RH at 20°C and lower at higher temperatures). Existing corrosion data suggests that negligible vapor-phase corrosion of aluminum is expected at such low humidity, even for nominally bare aluminum surfaces. That is, the predicted relative humidity from free water (e.g., 17% RH) is below the “critical” relative humidity (~40% RH at room temperature), below which no significant corrosion is observed for bare aluminum. Furthermore, reported room-temperature corrosion rates are very low even under saturated (100% RH) water vapor. Existing results showing significant vapor-phase corrosion corresponded specifically to conditions of high temper ed with high relative humidity. A relatively large reservoir of (chemically bound) water exists in the aluminum (oxy)hydroxide films on the SNF cladding surface. The estimated total could saturate the gas even at relatively high temperature if fully released; however, its release as molecular water is expected to only be plausible if the temperature during storage exceeds both the drying temperature and the threshold for thermal decomposition of the trihydroxides (~220°C). Therefore, the combination of high temperature and high relative humidity that could drive significant corrosion is considered implausible during sealed dry storage following an appropriate drying process, to include >220°C drying for canisters that may approach or exceed this temperature during storage. Vapor-phase testing of aluminum samples with an adherent (oxy)hydroxide layer, prepared in liquid water to resemble those on actual aluminum-clad spent nuclear fuel (ASNF), did not observe evidence of additional corrosion even at combined high-temperature (up to 180°C) and high-humidity (up to 100% RH) conditions. Instead, small net mass losses were observed for most specimens and attributed to dehydration of the samples, which had been air-dried only prior to testing. ASNF being moved to dry storage is expected to have an existing (oxy)hydroxide film, suggesting that the cladding will be less prone to additional corrosion than bare aluminum metal. If significant corrosion did occur during sealed storage, the overall extent and impact is expected to be small. Corrosion post-closure of a canister would not alter the total amount of hydrogen in the canister, so it would have no effect on the maximum H 2 release already assessed in existing bounding calculations. In addition, the total amount of water in a 25-µm dense bayerite film would consume a only ~8 µm additional aluminum metal on average, if fully consumed by oxidizing aluminum metal to Al 2 O 3 . These conclusions are consistent with ASNF-in-canister simulations to date, which included a reaction pathway for corrosion using kinetics from previous literature and believed to be conservative and predicted very low rates of corrosion.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Determination of Molecular Structure and Dynamics of Molten Salts by Advanced Neutron and X-ray Scattering Measurements and Computer Modeling

The design and development of fully functional Molten-Salt Reactors (MSR) require detailed knowledge of the molten salt properties in order to understand and predict the salt’s behavior. Fundamental properties of interest include molecular structure, speciation, and dynamics (such as diffusion coefficients) of salt components and dissolved corrosion and fission products. Computer modeling is necessary to predict changes in physical and chemical properties due to irradiation, burning of dissolved fuel, and corrosion. The modeling requires experimental data, and advanced neutron and x-ray scattering and spectroscopy provide the most reliable and direct determination of the structure (Pair-Distribution Functions, PDF), and dynamics of ions in the melt. This project dealt with both fluoride and chloride salts. The PDFs have been measured by a combination of neutron and x-ray diffraction. We utilized the techniques of isotope substitutions, a very powerful tool available for neutron-scattering, to extract the details of the liquid structure. Although similar measurements have been done before, modern advanced neutron and x-ray-scattering techniques allow collecting the data at much higher resolution and in a wider range of temperatures. Importantly, we were among the first to study fluoride salts by neutron scattering. The importance of impurities and their effects on salt properties have become apparent recently and so new methods of salt purification were developed. We took advantage of these developments to produce reliable data, which have been used for computer simulations of both clean salts and those with added fission and corrosion products most relevant for MSRs. Ab initio molecular dynamics simulations have been performed to understand the multi-component liquid solution, in particular solubility of impurities and thermodynamic interactions in relation to the ionic-cluster structure of the fluid. We applied machine learning to regress from the simulation and experimental data in order to develop a fast-acting model that can handle molten salt with an arbitrary (≥ 10) number of chemical elements and be able to predict chemical potential as a function of composition and temperature. This project resulted in a number of experimental and computer-simulation publications, a patent application, and numerous conference presentations (American Physical Society, American Chemical Society, and The Electrochemical Society among others). Multiple students and postdocs participated and collaborated on aspects of this project. This project seeded new collaborations between MIT and other institutions, such as the University of Massachusetts Lowell, the University of Illinois Urbana-Champaign, the University of California Berkeley, and Oak Ridge and Los Alamos National Labs. As such, this project has had a broad and lasting impact beyond its original scientific scope.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

L-Basin microbial monitoring program - evaluation of 20 years of monitoring

This report summarizes the results from over 20-years’ of activities for microbial monitoring of the bulk water, floc analysis, and test coupon surfaces for the Savannah River Site L Basin. The Microbial Monitoring Program provides important information for the characterization of overall water quality as it impacts the continued safe storage of spent nuclear fuel in L Basin. The potential impacts to fuel storage are: 1) microbial-influenced corrosion of the fuel; and 2) biofouling of the fuel storage systems.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Role of microstructure on CO corrosion of SiC layer in UO₂-TRISO fuel

The Advanced Gas Reactor Fuel Qualification and Development (AGR) program has focused on qualification of UCO kernel tristructural-isotropic (TRISO) particle fuel relative to UO₂ kernel TRISO particle fuel. However, a UO₂ kernel variant was included in the second AGR irradiation experiment (AGR-2) for comparison and to connect to historic fuel irradiation data. The development of a multiscale, post-irradiation examination (PIE) analysis approach through the AGR Program has allowed for a comprehensive understanding of individual particle failure. This approach has been applied to gain an understanding of SiC layer failure in UO₂ kernel TRISO fuel from AGR-2 after safety testing at 1600–1700 °C. Particle failure by intergranular CO corrosion, facilitated by a compromised inner pyrolytic carbon layer, has been confirmed through the combined application of modern x-ray computed tomography and electron microscopy techniques. In addition, a relationship between grain boundary character and CO corrosion has been identified. This finding provides an opportunity to develop mitigating strategies to improve the resilience of the SiC layer to CO corrosion in UO₂ TRISO fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

(Project 18-15531) Repair and Mitigation of Chloride-Induced Pitting and Chloride-Induced Stress Corrosion Cracking in Used Spent Nuclear Fuel Dry Cask Canister Materials (Final Technical Report)

Several experimental techniques were utilized to understand the corrosion and SCC performance of possible techniques for repairing cracks and pits in SS304 utilized for dry cask storage of spent nuclear fuel. The repairs examined included cold spray, FSW, soldering, and VFAW. The corrosion and SCC assessment techniques utilized for comparing repairs against the base material were 4-pt bending SCC testing in boiling MgCl 2 at stress levels between 0 MPa and 350 MPa, immersion testing in 40% RH brine solution meant to mimic a service-like environment, SSRT SCC testing in 40% RH brine solution. For FSW alone, scanning electrochemical cell microscopy and microelectrochemical syringe cell were utilized to understand the corrosion performance across the weld, in the retreating/advancing sides, and base material. Based on the results of this study, the FSW and cold spray repair techniques proved most viable for SCC repair in SS304 dry cask storage applications and continuing research is highly recommended to understand and quantify in further detail their SCC performance compared to unrepaired SS304. Additionally, solder and VFAW (patch or incline) would not be recommended based on the corrosion and SCC findings in this study.

42 ENGINEERING↗
DOE OSTI2020

Amphiphilic Block Copolymers for Flocculation and Hydrophobization of Legacy Waste Suspensions in Flotation Driven Dewatering Operations - 20162

Flotation has been effectively demonstrated as a low footprint, high efficiency separation process when utilising simple anionic surface modifying agents such as sodium dodecyl sulphate (SDS). These agents increase hydrophobicity to remove suspended cationic surface charged particulates such as Mg(OH){sub 2} from waste suspensions similar to those at British nuclear fuel management sites- where corrosion processes have degraded fuel cladding. The technology is advantageous as it could effectively be retrofitted into existing waste management facilities at nuclear waste management sites such Sellafield and Hanford as the simplicity (no moving parts) and size of this technology coupled with low cost of construction could be rapidly deployed to aid in risk and hazard reduction decommissioning operations with minimum impact to secondary waste generation. Previously, sodium dodecyl sulphate has been deployed as a collector agent to increase the hydrophobicity of Mg(OH){sub 2} suspended particles, achieving a maximum particulate extraction percentage of 93%. To achieve further particulate extraction, flotation performance has been determined to be a function of the hydrodynamic limitations extracting fine material (>20 μm). Fine particulates have been shown in the minerals industry to have an inability to overcome slipstreams created from rising bubbles in flotation cells, likely due to the low mass and thus momentum of the colloidal material, preventing interaction between the hydrophobized particles and air-water interfaces (bubbles). To combat this mass transfer resistance poly(acrylic acid)-block-poly(n-butyl acrylate) amphiphilic diblock copolymers were synthesised via reversible addition-fragmentation chain-transfer (RAFT) polymerisation. The anticipated mode of action of these polymers involves the hydrophilic poly(acrylic acid) block polymer acting as a flocculation agent to increase the mass of the aggregates, and the hydrophobic poly(n-butyl acrylate) modifies the hydrophobicity of the resulting flocs. The concentration of this polymer (termed a macro-collector) was varied and the change in particle size distribution of the particulate suspension was analysed using static light scattering. The flotation performance was then characterised by several methods including percentage particulate recovery, volume reduction factor and residual bulk particulate concentration. This was then benchmarked against the traditional surfactant collector agent SDS. (authors)

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DOE OSTI2020

The TRANSCEND Consortium - In-situ Identification of Surface Corrosion Products on Spent Nuclear Fuels - 20276

The management of spent nuclear fuel is a major ongoing concern for the UK owing to the cessation of reprocessing operations at Sellafield and the large, complex inventory arising from Magnox, AGR, PWR and prototype reactors. Retrieval and relocation operations for legacy fuels are imminent and therefore, any models that enhance our understanding of fuel evolution will help mitigate the risks associated with fuel storage and disposal. The TRANSCEND Consortium on nuclear waste management comprises four work packages, within this current paper we provide a summary overview of progress to date and illustrative results from Theme 3: Spent Nuclear Fuels. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

An Updated Analysis of Clad Degradation

This report provides an analysis of the clad barrier function associated with the direct disposal of dual purpose canisters (DPCs) under hypothetical conditions in a shale repository and in an alluvial repository, including the effect of a postulated criticality event inside a disposed DPC. Should a postulated criticality event occur in a hypothetical shale repository, cladding will primarily degrade by general corrosion. Stress corrosion cracking, hydride cracking, creep failure, pitting and crevice corrosion, rod pressurization, and clad unzipping are calculated to have little impact on cladding persistence. At the higher temperature expected during a postulated criticality event in a saturated shale repository, general corrosion of cladding would be rapid - on the order of 0.034 microns/yr. A few hundred years after onset of a postulated criticality event in a shale repository complete general corrosion of fuel assembly grid spacer walls and guide tubes will likely result in settling of fuel rods upon each other. This rod consolidation should displace the water moderator and possibly terminate a postulated criticality. The primary potential degradation pathway for cladding in a hypothetical alluvial repository is localized corrosion by fluoride, which cannot occur in a shale repository. Fluoride-enhanced corrosion of cladding would be accelerated under the slightly higher (< 100°C) temperatures associated with a postulated criticality event. The impact of criticality in both cases (shale and alluvial) would be to increase the amount of failed cladding. But it would require very specialized transport pathways.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1989

Corrosion testing of candidates for the alkaline fuel cell cathode

Current/voltage data was obtained for specially made corrosion electrodes of some oxides and of gold materials for the purpose of developing a screening test of catalysts and supports for use at the cathode of the alkaline fuel cell. The data consists of measurements of current at fixed potentials and cyclic voltammograms. These data will have to be correlated with longtime performance data in order to fully evaluate this approach to corrosion screening. Corrosion test screening of candidates for the oxygen reduction electrode of the alkaline fuel cell was applied to two substances, the pyrochlore Pb2Ru2O6.5 and the spinel NiCo2O4. The substrate gold screen and a sample of the IFC Orbiter Pt-Au performance electrode were included as blanks. The pyrochlore data indicate relative stability, although nothing yet can be said about long term stability. The spinel was plainly unstable. For this type of testing to be validated, comparisons will have to be made with long term performance tests.

Singer, Joseph↗
DOE OSTI2020

Drying Wet Stored and Corroded Magnox Fuel for Interim Dry Storage - 20187

To investigate the feasibility of drying corroded Magnesium alloy clad nuclear fuel, the corrosion products of inactive and unirradiated Magnox simulant dry corrosion product and dried corroded Magnox sludge were characterised by X-ray diffraction and thermogravimetric analysis. XRD identified brucite in the dried corrosion product and both brucite and hydromagnesite in the sludge. TGA identified mass loss on dehydration of ∼9.1% up to 250 deg. C and 27.9% from dehydroxylation up to 440 deg. C, with total mass loss of 41.9% up to 800 deg. C for the dried corrosion product. The sludge TGA showed 8.3% mass loss up to 250 deg. C from dehydration, 7.5% up to 330 deg. C from dehydroxylation and 31.4% mass loss up to 47.2% for decarbonation, with total mass loss of 49.2% up to 800 deg. C. Water removal up to 1.8 g was performed by cold vacuum drying (40 deg. C-120 deg. C) on a ≅14.8 g (wet) corroded Magnox sample. The process was monitored by observing pressure, dew point, temperature and gas flow changes supported by measuring the sample mass loss as water is removed. From these tests, it was observed that the dried corroded Magnox displayed some hygroscopic capacity which has implications for water retention following vacuum drying and the length of exposure required to achieve the desired dryness. Whilst at higher temperatures the water was removed faster, the increased temperature also risks the fuel cladding undergoing unwanted chemical reactions with the residual water during the process. However, at lower temperatures the achievable levels of dryness were reduced. Therefore, from these observations this work has identified that there is a temperature balance that may be necessary to optimise the drying process with respect to allowing the greatest level of dryness whilst restricting unwanted chemical reactions. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Rational design of high-performance low-loading oxygen reduction catalysts for alkaline fuel cells

The lack of mechanistic understanding and catalyst design principles for alkaline electrolytes, especially for the sluggish oxygen reduction reaction, has impeded the advancement of alkaline fuel cells. Here, in this study, we propose a modified volcano plot and apply this rationale to strategically design Pt nanosheets with PdH x nanosheets substrates. This catalyst exhibited high stability with a specific activity of 1.71 mA cm −2 at 0.95 V versus the reversible hydrogen electrode, surpassing the benchmark of Pt/C by 49-fold. Spectroscopic, electrochemical and electron microscopic characterizations revealed that such performance enhancement originated from tensile-strained Pt{111} facets, improving oxidative stability and suppressing carbon corrosion. In fuel cell testing, the catalyst enabled a peak power density of 1.67 W cm −2 with a loading of 10 µg PGM Cathode cm −2 . Further optimization delivered a peak power density of 21.7 W mg −1 PGM Cathode+Anode with a total specific catalyst cost US$\$$1.27 kW −1 , surpassing the US Department of Energy’s Pt group metal loading and cost targets. This study provides valuable insights into catalyst design for the alkaline oxygen reduction reaction.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Summary of Chloride-induced Stress Corrosion Cracking Experiments with a Spent Fuel Storage Canister Large Plate

This report provides a comprehensive summary of the crack growth rate testing and the experiments to evaluate the corrosion behavior of welded plate material taken from the Sandia National Laboratories’ full-size mockup of a spent nuclear fuel (SNF) dry storage canister. The information includes that previously reported, and new work on detailed surface characterization of the corrosion attack from the large plate experiment. The work and the results inform aging management programs for the SNF canisters.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1990

Corrosion testing of candidates for the alkaline fuel cell cathode

Current/voltage data have been obtained for specially made corrosion electrodes of some oxides and of gold materials for the purpose of developing a screening test of catalysts and supports for use at the cathode of the alkaline fuel cell. The data consist of measurements of current at fixed potentials and cyclic voltammograms. These data will have to be correlated with longtime performance data in order to evaluate fully this approach to corrosion screening.

Singer, Joseph↗
DOE OSTI2021

Challenges and opportunities to alloyed and composite fuel architectures to mitigate high uranium density fuel oxidation: uranium silicide

We report the challenges and opportunities to alloyed and composite fuel architectures designed and intended to mitigate oxidation of the fuel during a cladding breech of a water-cooled reactor are discussed in three review manuscripts developed in parallel, with the presented article focused on the oxidation performance of uranium silicide. Several high uranium density fuels are under consideration for deployment as accident tolerant and/or advanced technology nuclear reactor fuels, including UN, U 3 Si 2 , UC and UB 2 . Presented here are the literature for the U3Si2 degradation modes, thermodynamics, and oxidation performance of the pure compound and its reported alloyed and composite architectures. Furthermore, this review covers the materials and techniques for the incorporation of additives, dopants, or composite fuel architectures to improve the oxidation/corrosion behavior for high uranium density fuels for use in LWRs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗