SEARCH · Search NASA
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.
Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.
Potentiostatic and corrosion resistance tests for metallic ingots prepared as rotating disk electrodes - cathodic electrocatalysts for use in low temperature hydrogen-oxygen fuel cells
Production of green gasoline, jet, and diesel biofuels from waste greases was achieved using a novel hydrothermal, continuous-flow catalytic process operating under supercritical water conditions, with recycled water the only added chemical. Thermally and chemically stable catalysts were explored to optimize yields of liquid biofuels and to minimize production of gases and acidic compounds. A 50:50 mixture of brown and yellow waste greases converted into 76.6 wt% liquid biocrude (BC); the remainder converted to water and gases. Less than 0.2% of the FS formed carbon char (mainly amongst catalyst particles). Various tubular reactors (Inconel, ® Hastelloy®, titanium, stainless) showed no interior defects, erosion, or mass loss after runs. Here, the titanium catalyst was fully recovered and regenerated back to its original potency. The BC was further refined into 28%, 48%, 20%, and 4 wt%, respectively for green gasoline, jet, diesel, and bunker. Biofuels were analyzed for compound class compositions and reaction mechanisms were proposed. The hundreds of identified fuel products (C3-C35) from processing oleic acid as a pure model compound were identified. The neat green gasoline and diesel biofuels along with a 50% green jet blend (with petroleum Jet A) were tested in appropriate spark ignition, turbine, and diesel engines at University of MN Engine Labs. Biofuels achieved 107.7%, 97.2%, and 101.3% engine power performance levels relative to petroleum fuels (91-Octane, Jet A, #2 Diesel) along with lower CO and pollutant emissions. Biofuels passed ASTM fuel specifications (D4814, D7566, D975) including mandated corrosion and low sulfur limits of all three biofuels.
Poster summarizing modeling of UO2 (SIMFUEL) corrosion using the Zuzax code suite
Operation of a molten salt fueled reactor is governed, more so than any other type of reactor, by the thermal properties of the molten salt fuel because it is in a liquid form. The composition of the fuel salt will evolve during use in a reactor, as fission products are generated and corrosion of structural materials release corrosion products into the salt. The recently released NUREG/CR7299 states that fuel qualification for molten salt fuels will be based upon maintaining fuel salt properties within an acceptable range that results in achievement of fundamental safety functions (NUREG/CR-7299). Understanding the dependence of fuel properties on temperature and composition across the anticipated range of operating temperatures, burn-up and anticipated levels of contamination is required to qualify molten salt fuels. Measuring aspects of salt chemistry is appropriately a major thrust of the Molten Salt Reactor (MSR) campaign under the Advanced Reactor Technology program.
Accident Tolerant Fuels (ATFs) are designed to increase coping time following an accident scenario while preserving or improving current steady state reactor operational performance. A potential ATF concept is SiC-SiC composite claddings. Fuel performance simulations were conducted on a SiC-SiC based cladding concept utilizing a multilayered approach for improved performance. This cladding concept referred to as the Duplex concept is a duplex structure composed of a monolithic SiC layer placed on the outside of a SiC-SiC composite. A liquid metal is added to fuel-cladding gap for improved heat dissipation from the fuel. The monolithic SiC layer is used to improve the coolant corrosion characteristics and protect the SiC-SiC composite layer from exposure to the coolant. The fuel performance code BISON was used to conduct fuel performance simulations on the cladding concepts. Comparisons are made with a current prototypic fuel rod design (UO 2 fuel enclosed in Zircaloy-4 cladding). Representative steady-state cases were considered for normal power and two cycle power histories. Additionally, a PCI ramp case was simulated to analyze potential anticipated operational occurrences. Transient response during a Loss of Coolant Accident and a Reactivity Initiated Accident were also simulated. This computational study demonstrated that for normal operating conditions, the SiC concept cladding performed as well as the baseline for the standard power cases evaluated. The ramping evaluations indicate potential fracturing of the SiC-SiC composite of the composite cladding compared to the Zircaloy-4 cladding due to the temperature gradient and the subsequent differential thermal conductivity degradation and swelling across the composite thickness. In conclusion, the rod fails early at low enthalpy for RIA but survives a LOCA with minimal material loss due to high temperature steam corrosion.
Various vacuum jacketed cryogenic supply lines at the Space Shuttle launch site at Kennedy Space Center use convoluted flexible expansion joints. The atmosphere at the launch site has a very high salt content, and during a launch, fuel combustion products include hydrochloric acid. This extremely corrosive environment has caused pitting corrosion failure in the thin walled 304L stainless steel flex hoses. A search was done to find a more corrosion resistant replacement material. The study focussed on 19 metal alloys. Tests which were performed include electrochemical corrosion testing, accelerated corrosion testing in a salt fog chamber, and long term exposure at a beach corrosion testing site. Based on the results of these tests, several nickel based alloys were found to have very high resistance to this corrosive environment. Also, there was excellent agreement between the electrochemical tests and the actual beach exposure tests. This suggests that electrochemical testing may be useful for narrowing the field of potential candidate alloys before subjecting samples to long term beach exposure.
FeCrAl alloys are promising candidate materials for the accident tolerant fuel (ATF) cladding applications due to their excellent corrosion resistance to the elevated temperature steam environment. Currently, the handbook on FeCrAl material properties contains only limited data regarding the fracture toughness properties of any FeCrAl alloy. This includes alloys currently under investigation within the Advanced Fuels Campaign (AFC) at Oak Ridge National Laboratory (ORNL). In this project, a series of irradiation capsules have been irradiated in the High Flux Isotope Reactor (HFIR) at ORNL with two Generation II FeCrAl candidate alloys, i.e., C06M and C36M, to assess the fracture response of these alloys after neutron irradiation. These alloys represent the “book-end” compositions for C26M, the alloy currently being developed as the leading candidate for LWR cladding. A total of six irradiation capsules were irradiated in HFIR at target temperatures of 200°C, 330°C, and 500°C up to target damage doses of 8 displacements per atom (dpa) and 16 dpa. These damage doses represent the expected middle and end of life damage levels for typical LWR cladding while the irradiation temperature regimes will provide insight into the role of varying microstructural features on the fracture toughness properties of neutron irradiated FeCrAl alloys. To date, irradiation of all capsules has been completed in HFIR. This report summarizes the latest results of microhardness and fracture toughness PIE for the 16 dpa capsules (FCAB2, FCAB4, and FCAB6), for which the measured irradiation conditions were: 204°C/17.6dpa, 343°C/18.3dpa, and 507°C/18.6dpa. The main conclusions of this study can be summarized as follows: 1) After the 204°C/17.6dpa irradiation, both C06M and C36M exhibited significant irradiation hardening and embrittlement 2) After the 343°C/18.3dpa irradiation, both C06M and C36M exhibited small irradiation hardening without irradiation embrittlement 3) After the 507°C/18.6dpa irradiation, both C06M and C36M exhibited irradiation softening without irradiation embrittlement 4) Comparing the microhardness and Master Curve reference temperature T 0q before and after neutron irradiation, we did not observe a linear correlation between the two parameters for both C06M and C36M. This should be mainly due to a flat response of the Master Curve reference temperature T 0q to the irradiations at 166-204°C and 315-343°C ranges 5) C06M showed a lower T 0q , meaning better toughness, than C36M at the unirradiated condition and such trend was kept even after neutron irradiation except for the 166-204°C irradiation where both materials had similar T 0q . 6) In terms of hardening and embrittlement, the irradiation effect on both C06M and C36M appeared to saturate after an irradiation dose of 7 dpa.
Dry storage canisters of used nuclear fuels are fabricated using SUS 304L stainless steel. Chloride-induced stress corrosion cracking (CISCC) is one of the major failure modes of dry storage canisters. The cracked canisters can be repaired by friction stir welding (FSW), a low-heat input ‘solid-phase’ welding process. It is important to evaluate the ClSCC resistance of the friction stir welded material. Stress corrosion cracking (SCC) studies were carried out on mill-annealed base materials and friction stir welded 304L stainless U-bend specimens in 3.5% NaCl + 5 N H2SO4 solution at room temperature and boiling MgCl2 solution at 155 °C. The engineering stress on the outer fiber of the FSW U-bend specimen was ~60% higher than that of the base metal (BM). In spite of the higher stress level of the FSW, both materials (FSW and BM) showed almost similar SCC failure times in the two different test solutions. The SCC occurred in the thermo-mechanically affected zone (TMAZ) of the FSW specimens in the 3.5% NaCl + 5 N H2SO4 solution at room temperature, while the stirred zone (SZ) was relatively crack-free. The failure occurred at the stirred zone when tested in the boiling MgCl2 solution. Hydrogen reduction was the cathodic reaction in the boiling MgCl2 solution, which promoted hydrogen-assisted cracking of the heavily deformed stirred zone. The emergence of the slip step followed by passive film rupture and dissolution of the slip step could be the SCC events in the 3.5% NaCl + 5 N H2SO4 solution at room temperature. However, the slip step height was not sufficient to cause passivity breakdown in the fine-grained SZ. Therefore, the SCC occurred in the partially recrystallized softer TMAZ. Overall, the friction-stirred 304L showed higher tolerance to ClSCC than the 304L base metal.
Not Available
Measurements of thermophysical properties of molten salts are needed for modeling and simulation activities that support the development of molten salt reactor (MSR) technologies. Properties of interest including transition temperatures, phase behavior, heat capacity, density, volumetric thermal expansion, surface tension, viscosity, thermal diffusivity, thermal conductivity, and vapor pressure are being performed at Argonne. Results of these property measurements are suitable for use in evaluating reactor performance during startup and the early operating life of the reactor. Ingressions of oxygen and moisture into the fuel salt are expected to occur at different times during the operating life of the reactor due to system leaks, maintenance, and refueling activities. The presence of these environmental contaminants induces corrosion of structural materials. The introduction of corrosion-derived species, oxygen and moisture is expected to affect the physical and chemical properties of the salt and operation of the reactor. Previous work performed at Argonne evaluated the effects of fission product dopants on the thermal properties of eutectic LiF-NaF-KF (FLiNaK). Properties of FLiNaK are commonly used to represent those of fluoride-bearing fuel salts. Metallic corrosion products such as chromium and nickel ions together with dissolved oxygen are expected to affect system redox differently than the accumulation of fission products. Work summarized in this report was performed to measure the effects of corrosion products and dissolved oxygen on the phase transitions, and specific heat capacity of FLiNaK. Thermophysical property measurements were made using four salts that were prepared by doping aliquots of a eutectic mixture of FLiNaK with surrogate corrosion products. Controlled additions of CrF 3 and NiF 2 were used as surrogates for corrosion product contamination anticipated during extended reactor operations in which fuel salt is in contact with steel reactor components. Controlled additions of UF 4 from two sources containing known amounts of UO 2 at different concentrations were used to represent oxygen contamination. The phase transitions and specific heat capacities of the four salts were measured by using differential scanning calorimetry. Measurements were made at temperatures spanning the range of 500−900 °C, which is the expected operating range of MSRs. Measured property values were compared to values measured previously with eutectic FLiNaK without dopants. Differences between property values measured for the doped and non-doped salts were compared with the uncertainties of the measurements to determine the significance of the effect of corrosion products and oxygen on salt properties.
Biomass-derived oils are renewable fuel sources and commodity products and are proposed to partially or entirely replace fossil fuels in sectors generally considered difficult to decarbonize such as aviation and maritime propulsion. Bio-oils contain a range of organic compounds with varying functional groups which can lead to polarity-driven phase separation and corrosion of containment materials during processing and storage. Polar compounds, such as organic acids and other oxygenates, are abundant in bio-oils and are considered corrosive to structural alloys, particularly to those with a low-Cr content. To study the corrosion effects of small carboxylic acids present in pyrolysis bio-oils, type 410 stainless steel (SS410) specimens were exposed in bio-oils with varying formic, acetic, propionic and hexanoic acid contents at 50 °C during 48 h exposures. The specific mass change data show a linear increase in mass loss with increasing formic acid concentration. Interestingly, a mild corrosion inhibition effect on the corrosion of SS410 specimens was observed with the addition of acetic, propionic and hexanoic acids in the bio-oil.
Sixteen candidate iron base Stirling engine heater head tube alloys are evaluated in a diesel fuel fired simulator materials test rig to determine their oxidation and corrosion resistance. Sheet specimens are tested at 820 C for 3500 hr in 5 hr heating cycles. Specific weight change data and an attack parameter are used to categorize the alloys into four groups; 10 alloys show excellent for good oxidation and corrosion resistance and six alloys exhibit poor or catastrophic resistance. Metallographic, X-ray, and electron microprobe analyses aid in further characterizing the oxidation and corrosion behavior of the alloys. Alloy compositions, expecially the reactive elements aluminum, titanium, and chromium, play a major role in the excellent oxidation and corrosion behavior of the alloys. The best oxidation resistance is associated with the formation of an iron nickel aluminum outer oxide scale, an intermediate oxide scale rich in chromium and titanium, and an aluminum outer oxide scale adjacent to the metallic substrate, which exhibits a zone of internal oxidation of aluminum and to some extent titanium.
Oxidation modeling in modern nuclear fuel performance codes is currently limited by the lack of coupling with mechanics, thus preventing proper description of how high-temperature oxidation impacts mechanical properties. This is mostly due to the fact that the finite difference formalism adopted in corrosion models is incompatible with the direct coupling with mechanics in the finite element modeling employed in modern nuclear fuel performance codes. In this study, a physically based zirconium alloy corrosion model called the Coupled-Current Charge Compensation (C4) model, which was initially developed for operating temperature conditions, has been updated to include high-temperature corrosion in order to provide additional critical information (e.g., oxygen concentration profile) under loss-of-coolant accident (LOCA) conditions—information lacking in existing empirical models. The C4 model was implemented in the MOOSE finite-element framework developed at Idaho National Laboratory, enabling it to be used in the BISON nuclear fuel performance code based on the MOOSE framework. To precisely track the different interfaces at a relatively low computational cost, the eXtended Finite Element Method (X-FEM) was applied in MOOSE. The model’s results were compared to those of existing empirical models as well as metallographic analysis of high-temperature oxidized Zircaloy-4 coupons. Oxygen diffusivities in the a and ß phases resulting from this comparison closely agree with those found in the literature. The C4 model implemented with X-FEM in MOOSE now has the capability to accurately predict oxide, oxygen-stabilized a, and prior ß phase layer growth kinetics under isothermal exposure at high temperature (1000–1500°C). Furthermore, in contrast with the empirical models, the C4 model accounts for the finite thickness of the fuel cladding. It can predict the oxygen concentration profile evolution through the whole cladding, enabling evaluation of the remaining ductile thickness—a crucial variable for modeling the mechanical behavior of the fuel cladding under LOCA. Furthermore, this implementation allows direct coupling with mechanics, at a low computing cost, using finite-element-based nuclear fuel performance codes such as BISON.
This report summarizes the current actives in FY21 related to the effort by Sandia National Laboratories to identify and test coating materials for the prevention, mitigation, and repair of spent nuclear fuel dry storage canisters against potential chloride-induced stress corrosion cracking. This work follows up on the details provided in Sandia National Laboratories FY20 report on the same topic, which provided a detailed description of the specific coating properties desired for application and implementation on spent nuclear fuel canisters, as well as provided detail into several different coatings and their applicability to coat spent nuclear fuel canisters. In FY21, Sandia National Laboratories has engaged with private industry to create a Memorandum of Understanding and established a collaborative R&D program building off the analytical and laboratory capabilities at Sandia National Laboratories and the material design and synthesis capabilities of private industry. The resulting Memorandum of Understanding included four companies to date (Oxford Performance Materials, White Horse R&D, Luna Innovations, and Flora Coating) proposing six different coating technologies (polyetherketoneketone, modified polyimide/polyurea, modified phenolic resin, silane-based polyurethane hybrid with and without a Znrich primer, and a quasi-ceramic sol-gel polyurethane hybrid) to be tested, evaluated, and optimized for their potential use for this application. This report provides a detailed description of each of the coating systems proposed by the participating industry partners. It also provides a description of the planned experimental actives to be performed by Sandia National Laboratories including physical tests, electrochemical tests, and characterization methods. These analyses will be used to identify specific ways to further improve coating technologies toward their application and implementation on spent nuclear fuel canisters. In FY21, Sandia National Laboratories began baseline testing of the base metal material in according with activities of the Memorandum of Understanding. In FY22, Sandia National Laboratories will receive coated coupons from each of the participating industry partners and begin characterization, physical, and electrochemical testing following the test plan described herein.
The overarching goal of this research project is quantitative microscopic modeling of phase diagrams and solubility of actinides and lanthanide-fission products in molten salts for molten salt reactors. Application of molten salts as coolants and liquid fuels for next-generation reactors requires quantitative understanding of phase diagrams, equations of state and transport properties of binary and ternary molten salt mixtures, as well as solubility of fuel, fission products and reactor-wall corrosion products, but these remain beyond the reach of current first-principles computational techniques. We will address this fundamental capability gap using a novel hybrid framework for thermodynamic integration to directly predict absolute free energies of solid and molten salts at any specified conditions of composition, temperature and pressure. This framework will combine a hierarchy of first-principles, molecular dynamics and continuum techniques to deliver first-principles-level accuracy at a fraction of the computational cost of conventional ab initio molecular dynamics approaches. In particular, this technique is expected to provide first-principles accuracy, while off-loading most of the computational work on much less expensive classical molecular dynamics and continuum techniques.
Innovative thin film multilayer metal/ceramic barrier coatings have been developed by Argonne National Laboratory to protect fuel cladding against different phenomena that can be encountered under the harsh operating conditions of advanced nuclear reactors. One phenomenon is Fuel-Cladding Chemical Interaction (FCCI), which can represent a challenge to the performance of steel-clad metallic fuels in sodium-cooled fast reactors (SFRs), especially when operating to high burnups at elevated cladding temperatures, and long fuel residence time. Another phenomenon is the corrosion of zirconium-based cladding under prototypic boiling water reactor (BWR) conditions and associated cladding during off normal conditions, including loss of coolant accidents (LOCA). This work provides details of the development and testing of a thin film (few microns) multilayer metal/ceramic (CrY/Y 2 O 3 ) coatings on steel cladding and its performance against FCCI. The coating architecture evolved through systematic mechanical testing, radiation tolerance assessments, and high-temperature diffusion studies from a preliminary multilayer concept to an optimized high-performance design. The optimized structure leverages stoichiometric Y 2 O 3 for its exceptional radiation stability, chemical inertness, and ultra-low fission product diffusivity (e.g., Ce diffusivity ~10 -20 m 2 /s in irradiated Y 2 O 3 at 650 °C). This is paired with ductile chromium-rich interlayers (Cr 95 Y 5 ), which significantly enhance fracture toughness and mechanical resilience. Design refinements, including increased ceramic layer thickness and optimized Cr–Y interlayer composition, eliminated fracture propagation and preserved coating integrity under extreme mechanical stress (>10× typical reactor conditions) and prolonged radiation exposure (~300 dpa). Those conditions are well beyond expected conditions during normal operations in SFR, and cover some of the transient conditions as well. In addition, the coating deposition was demonstrated to be applied uniformly to the inner surface of steel tubes representing prototypic SFR cladding. Another phenomenon under consideration in this work is the corrosion of zirconium-based cladding under prototypic boiling water reactor (BWR) conditions and associated off normal conditions including loss of coolant accident (LOCA). Another multilayer thin film coating is considered for mitigation of this phenomenon, where a coating of yttrium aluminum and yttrium aluminum ceramic/metal thin film coating. Preliminary results on the coating performance during high temperature thermal cycling (up to 1000 C) have shown promising performance of the coating. On-going optimization of the coating will continue in FY26.
With the resurgence of interest in molten salt reactors, there is a need for new experiments and modeling capabilities to characterize the unique phenomena present in this fluid fuel system. A Versatile Experimental Salt Irradiation Loop (VESIL) is currently under investigation at Idaho National Laboratory to be placed in the Advanced Test Reactor (ATR). One of the key phenomena this proposed experiment plans to elucidate is fission product speciation in the fuel-salt and the subsequent effects this has on the fuel-salt properties, source term generation, and corrosion control. Specifically, noble gases (Xe & Kr) will bubble out to a plenum or off-gas system, and noble metals (Mo, Tc, Te, etc.) will precipitate and deposit in specific zones in the loop. This work extends the mass transfer and species interaction models in CTF (Coolant-Boiling in Rod Arrays—Two Fluids) and applies these models to give a preliminary estimation of fission product behavior in the proposed VESIL design. A noble metal–helium bubble mass transfer model is coupled with the thermal-hydraulic results from CTF to determine the effectiveness of this insoluble fission product (IFP) extraction method for VESIL. Amounts of IFP species extracted to the off-gas system and species distributions in VESIL after a 60-day ATR cycle are reported.
A ternary chloride (MgCl2-NaCl-KCl) molten salt pump has been developed to operate up to 720°C. The design is the result of a thermal, structural and material study of critical components. The vertical overhung pump incorporates first of a kind salt-wetted hydrodynamic bearings. Several bearing material pairings were tested which include: colmonoy grades, stellite grades, and a novel NiWC-based alloy (i.e. HybrimetTM NiWC3b) used as a High Velocity Oxygen Fuel (HVOF) coating. The coating withstands the corrosion effects of the molten salt, and is also used to protect a 316L stainless steel pump reservoir. The pump bowl assembly components are made of Inconel 625. A Nitrogen flushed carbon ring seal is used to keep salt vapors from interacting with external ambient air. Preliminary testing shows a promising pump design for use in the Concentrating Solar Power Gen 3 systems and sets the ground for further development of the HybrimetTM NiWC3b as a competitive alternative to available superalloys.