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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 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
NASA NTRS1991

Space nuclear power systems; Proceedings of the 8th Symposium, Albuquerque, NM, Jan. 6-10, 1991. Pts. 1-3

The present conference discusses NASA mission planning for space nuclear power, lunar mission design based on nuclear thermal rockets, inertial-electrostatic confinement fusion for space power, nuclear risk analysis of the Ulysses mission, the role of the interface in refractory metal alloy composites, an advanced thermionic reactor systems design code, and space high power nuclear-pumped lasers. Also discussed are exploration mission enhancements with power-beaming, power requirement estimates for a nuclear-powered manned Mars rover, SP-100 reactor design, safety, and testing, materials compatibility issues for fabric composite radiators, application of the enabler to nuclear electric propulsion, orbit-transfer with TOPAZ-type power sources, the thermoelectric properties of alloys, ruthenium silicide as a promising thermoelectric material, and innovative space-saving device for high-temperature piping systems. The second volume of this conference discusses engine concepts for nuclear electric propulsion, nuclear technologies for human exploration of the solar system, dynamic energy conversion, direct nuclear propulsion, thermionic conversion technology, reactor and power system control, thermal management, thermionic research, effects of radiation on electronics, heat-pipe technology, radioisotope power systems, and nuclear fuels for power reactors. The third volume discusses space power electronics, space nuclear fuels for propulsion reactors, power systems concepts, space power electronics systems, the use of artificial intelligence in space, flight qualifications and testing, microgravity two-phase flow, reactor manufacturing and processing, and space and environmental effects.

El-Genk, Mohamed S.↗
DOE OSTI2024

A Molten Salt Aerosol Spectroscopy Approach to Online Process Monitoring in Molten Salt Reactors

Recent research and development in molten salt reactors (MSRs) has led to many advanced reactor designs, many of which have demonstrations planned in the next decade. However, a significant challenge that must be overcome before liquid fueled MSRs become viable globally is nuclear material accountancy (NMA) and safeguards. Liquid-fueled MSRs are problematic to safeguards because the nuclear material is in the bulk phase, inventories are constantly changing, and there are potentially many outlet pathways that must be monitored, especially in designs with online chemical processing. At the Idaho National Laboratory, we have developed a process monitoring tool to circulate molten salt through a nebulizer and optical cell, where laser-induced breakdown spectroscopy (LIBS) and ultraviolet-visible (UV-VIS) are applied to track the elemental and oxidation states of the constituents in the salt aerosol. This approach has the potential to provide information on the actinides, fission products, and corrosion in the reactor or process. Preliminary testing in aqueous praseodymium chloride (PrCl3) and neodymium chloride (NdCl3) showed good sensitivity detection limits n developed calibration curves. Additional tests and analysis or ongoing in aqueous solutions while equipment is being scaled up for high temperature molten salt testing. Results for continued testing as well as a status of the molten salt design and anticipated commissioning will be presented.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2024

Demonstration of Temperature Compensation Techniques for SPNDs Operating in High Temperatures

This report presents the testing results of rhodium-based self-powered neutron detectors (Rh-SPND) irradiated in a furnace dry tube from ambient temperature to 850°C at the Ohio State University Research Reactor. The purpose of the experiment is to demonstrate the technique and application of a temperature compensation technique for the Rh-SPND. This is performed by characterizing the temperature effects observed in past experiments—a displacement current and a stabilized dark current—of the Rh-SPND as a function of temperature under the models of shifting space charges as a product of photoconductivity properties. Low-power irradiation at the OSURR was performed with stabilized temperatures of ambient, 550, 575, 600, 625, 650, 675, and 700°C were first performed to obtain the curve fit parameters that describes the temperature effects. The results provided further insight for the behavior of the SPND at high temperatures in accordance with available insulation conductivity models. Transition points from photoconductivity to ionic conductivity were identified in the range of 550–600°C. Additionally, transition points ionic to electric conductivity were observed in the range of 675–700°C, however, the data was not able to fully capture the transition and did not have enough resolution to provide predictive compensation based only on temperature readings.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2026

Understanding the high-temperature behavior and corrosion resistance of Cr-Nb coated cladding for BWRs

While Cr-coated Accident Tolerant Fuel has proven successful in Pressurized Water Reactors, its dissolution in high dissolved oxygen Boiling Water Reactor environments represents a major technological barrier. Here, this study presents a comprehensive evaluation of novel Cr-Nb alloys as a coating breakthrough solution for Accident Tolerant Fuel cladding in Boiling Water Reactors. We systematically investigated Physical Vapor Deposited Cr-Nb coatings with 13% and 24 at% Nb content on commercial Zr-based cladding through rigorous testing under both normal and accident conditions. Our results demonstrate Cr-Nb coatings remained completely intact during extended Boiling Water Reactor autoclave testing, while pure Cr coatings failed catastrophically due to oxide dissolution. Under Loss-of-Coolant Accident conditions at 1100°C, Cr-Nb coatings maintained protective capability for 45 min. At 800°C, the alloy coatings matched pure Cr's excellent corrosion resistance. Despite microstructural changes during prolonged high-temperature exposure, no coating delamination occurred, ensuring continued protection of the underlying Zr substrate. These findings establish Cr-Nb coatings as a viable Accident Tolerant Fuel solution for Boiling Water Reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

High-performance ECRH at W7-X: experience and perspectives

The second operation phase of W7-X (OP1.2) showed the potential of exclusively electron cyclotron resonance heating (ECRH)-sustained plasma operations in stellarators. Employing multi-pass ECRH scenario in the second harmonic O-mode (O2-ECRH), stationary densities of up to 1.4 × 10 20 m -3 could be achieved. This scenario also made stationary divertor detachment possible, which is a reactor-relevant scenario for power and particle exhaust. At high densities and with sufficiently high density gradients for an improved ion confinement, the coupling between the electrons and ions was strong enough to bring the ion temperature to values above 3 keV and to the neoclassical limit for some magnetic configurations, thus enabling to test the W7-X neoclassical optimization. The planned enhancement of the ECRH performance will enable to advance towards reactor-relevant beta values and to investigate their stability and confinement of fast particles, which is a priority goal of W7-X.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Intrinsic Junction Thermocouples For Surface Temperature Measurement

The material selection, attachment, and calibration of intrinsic junction thermocouples is discussed. Intrinsic junction thermocouples are utilized for their fast response times in surface measurements during high thermal transients. The as-is measurement and calibrated corrections are presented for both the out-of-pile pulse power system and the critical heat flux experiments in the Idaho National Laboratory Transient Reactor Test facility. The fin effect from temperature measurements on the surface of fuel cladding is corrected using a linear coefficient, m, and temperature corrections are sometimes 100 °C or more—representing 18% of the original temperature measurement.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Status report on HFIR irradiation of optimized alumina forming alloys

Properties of FeCrAl alloys under neutron irradiation are of interest because of these materials’ potential application as accident-tolerant fuel cladding in nuclear systems. In parallel, alumina-forming austenitic (AFA) alloys are of interest for use as structural materials in advanced nuclear systems for their potential higher resistance to embrittlement and high-temperature steam oxidation resistance. An irradiation campaign for fiscal year 2024 has been developed under the Advanced Fuels Campaign to perform irradiation testing of various FeCrAl and AFA alloys in Oak Ridge National Laboratory’s High Flux Isotope Reactor (HFIR). The goals of this irradiation campaign are to (1) study the impact of minor alloying elements on the neutron-irradiated mechanical properties of FeCrAl alloys and (2) collect neutron-irradiated mechanical properties on AFA alloys for comparison with those of FeCrAl alloys. This campaign will include both tensile and fracture toughness specimens tested following HFIR irradiation at temperatures representative of normal operating conditions in light-water reactors. The pre-irradiation characterization to date, the irradiation plan for the FeCrAl and AFA specimens, and the subsequent post-irradiation experimental test plan are presented in this report, along with the status of HFIR builds and scheduled insertion dates.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Assembly of MiniFuel Targets for Irradiation of U-Mo Fuel Specimens in the High Flux Isotope Reactor

To support the development of advanced low-enriched uranium for use in nuclear reactors, irradiation testing of U-Mo disk specimens was performed at the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) to collect experimental irradiation data on this type of fuel at pressurized water reactor–relevant temperatures. U-Mo is a uranium alloy that has superior dimensional stability relative to alpha-phase uranium metal and has a substantially higher uranium density compared to UO 2 . U-Mo disks specimens were fabricated at Idaho National Laboratory and inserted into MiniFuel targets for HFIR irradiation. Three MiniFuel targets were successfully assembled, welded, tested, and delivered to HFIR, along with their quality assurance documentation. The targets were inserted into HFIR’s inner vertical experiment facility within the permanent beryllium reflector. Each target contains six disk specimens and will be irradiated in HFIR for three, four, and eight cycles, with target temperatures between 250 and 500°C. This report summarizes the experiment design, test matrix, pre-characterization of specimens, and experiment assembly.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2024

Integrated Simulations of the Sabatier and Carbon Vapor Deposition Reactor to Understand Its Impacts to Operations and Performance

The carbon vapor deposition (CVD) reactor is a technology developed by Honeywell Aerospace to convert methane, at high temperatures, into hydrogen and solid carbon. This element is coupled with a Sabatier reactor to support a closed-loop environmental control and life support system with the aim of achieving nearly complete oxygen recovery (> 95%). Initial open-loop, brassboard CVD reactor tests and simulations have shown the CVD’s ability to achieve moderately high methane conversion and high hydrogen selectivity. However, in an integrated system, additional deficiencies are expected due to recycling of unreacted or extraneous species from the Sabatier reactor (e.g., carbon dioxide, hydrogen, water) and CVD reactor (e.g., hydrocarbons, methane, etc.). Sabatier and CVD reactor models were integrated and simulated to predict potential impacts to individual reactors’ and the overall system’s performance. The simulations showed that increasing the recycle of the CVD effluent hydrogen combined with decreasing the system inlet hydrogen flow rate (i.e., drawing a stoichiometric flow rate from an electrolyzer) can lead to an oxygen recovery of > 95%. However, system integration comes at a detriment to the individual reactors. The simulations show the initial conversion from the integrated system (Sabatier = 87% and CVD = 59%) to be lower than the standalone systems (Sabatier = 91% and CVD = 69%). Furthermore, transient simulations show substrate densification, leading to worsening methane conversion coupled with increasing acetylene production, which is commensurate with soot formation. Simulations predict a shortening of the maintenance interval (i.e., time until CVD methane conversion drops below 50%) in the integrated system, which would increase the consumable substrate mass. These analyses highlight the importance of long-duration, integrated tests to corroborate these findings as well as suggest potential modifications (e.g., intermediate gas separations) to improve performance.

Air Revitalization↗
DOE OSTI2020

Experimental demonstration of a 5 kW th granular-flow reactor for solar thermochemical energy storage with aluminum-doped calcium manganite particles

A two-step cycle was discussed for solar thermochemical energy storage based on aluminum-doped calcium manganite reduction/oxidation reactions for direct integration into Air Brayton cycles. The two steps encompassed 1) the storage of concentrated solar direct irradiation via the thermal reduction of aluminum-doped calcium manganite and 2) the delivery of heat to an Air-Brayton cycle via re-oxidation of oxygen-deficient aluminum-doped calcium manganite. The re-oxidized aluminum-doped calcium manganite was fed back to the first step to complete the cycle. A 5 kWth solar thermochemical reactor operating under vacuum was fabricated and tested to examine the first cycle reduction step. Reactor operating conditions and high-flux solar simulator control were tuned for continuous reactor operation with particle temperatures > 1073 K. Continuous operation was achieved using intermittent, dense granular flows. A maximum absorption efficiency of 64.7% was shown, accounting for both sensible and chemical heat storage.

25 ENERGY STORAGE↗
DOE OSTI2022

Final Status Report Confirming Completion of Mechanical Testing of HFIR Irradiated Nanocomposite Materials

This report provides a final status report on the in-cell mechanical testing of High Flux Isotope Reactor (HFIR)-irradiated nanodispersion-strengthened materials at the Irradiated Materials Examination and Testing (IMET) facility. All tension tests were performed at room temperature with a nominal strain rate of 0.018 mm/mm/min using shoulder-loading grip sets by following the standard testing procedure in ASTM E8/E8M. Round 1 (unirradiated) and round 2 (0.7 dpa) testing was completed in FY 2021. This report confirms completion of round 3 (1.4 dpa) and round 4 (2.1 dpa) tensile testing in FY 2022 with examples of test data collected on select sample conditions. Additionally, the report includes IDs of samples selected for shipment to Idaho National Laboratory (INL) for additional characterization.

36 MATERIALS SCIENCE↗
NASA NTRS1991

Review of Rover fuel element protective coating development at Los Alamos

The Los Alamos Scientific Laboratory (LASL) entered the nuclear propulsion field in 1955 and began work on all aspects of a nuclear propulsion program with a target exhaust temperature of about 2750 K. A very extensive chemical vapor deposition coating technology for preventing catastrophic corrosion of reactor core components by the high temperature, high pressure hydrogen propellant gas was developed. Over the 17-year term of the program, more than 50,000 fuel elements were coated and evaluated. Advances in performance were achieved only through closely coupled interaction between the developing fuel element fabrication and protective coating technologies. The endurance of fuel elements in high temperature, high pressure hydrogen environment increased from several minutes at 2000 K exit gas temperature to 2 hours at 2440 K exit gas temperature in a reactor test and 10 hours at 2350 K exit gas temperature in a hot gas test. The purpose of this paper is to highlight the rationale for selection of coating materials used (NbC and ZrC), identify critical fuel element-coat interactions that had to be modified to increase system performance, and review the evolution of protective coating technology.

Wallace, Terry C.↗
NASA NTRS2011

Online Oxide Contamination Measurement and Purification Demonstration

Liquid metal sodium-potassium (NaK) has advantageous thermodynamic properties indicating its use as a fission reactor coolant for a surface (lunar, martian) power system. A major area of concern for fission reactor cooling systems is system corrosion due to oxygen contaminants at the high operating temperatures experienced. A small-scale, approximately 4-L capacity, simulated fission reactor cooling system employing NaK as a coolant was fabricated and tested with the goal of demonstrating a noninvasive oxygen detection and purification system. In order to generate prototypical conditions in the simulated cooling system, several system components were designed, fabricated, and tested. These major components were a fully-sealed, magnetically-coupled mechanical NaK pump, a graphite element heated reservoir, a plugging indicator system, and a cold trap. All system components were successfully demonstrated at a maximum system flow rate of approximately 150 cc/s at temperatures up to 550 C. Coolant purification was accomplished using a cold trap before and after plugging operations which showed a relative reduction in oxygen content.

Bradley, D. E.↗
NASA NTRS2017

Non-Intrusive Velocity Measurements with MTV During DCC Event in the HTTF

Velocity profiles are measured using molecular tagging velocimetry (MTV) in the high temperature test facility (HTTF) at Oregon State University during a depressurized conduction cooldown (DCC) event. The HTTF is a quarter scale electrically heated nuclear reactor simulator designed to replicate various accident scenarios. During a DCC, a double ended guillotine break results in the reactor pressure vessel (RPV) depressurizing into the reactor cavity and ultimately leading to air ingress in the reactor core (lock-exchange and gas diffusion). It is critical to understand the resulting buoyancy-driven flow to characterize the reactor self-cooling capacity through natural circulation. During tests conducted at ambient pressure and temperature, the RPV containing helium is opened (via the hot and cold legs) to a large vessel filled with nitrogen to simulate the atmosphere. The velocity profile on the hot leg pipe centerline is recorded at 10 Hz with MTV based on NO tracers. The precision of the velocimetry was measured to be 0.02 m/s in quiescent flow prior to the tests. A helium flow from the RPV is initially observed in the top quarter of the pipe. During the first 20 seconds of the event, helium flows out of the RPV with a maximum velocity below 2 m/s. The velocity profile transitions from parabolic to linear in character and decays slowly over the rest of the recording; peak velocities of 0.2 m/s are observed after 30 min. A counter-flow of nitrogen is also observed intermittently, which occurs at lower velocities (>0.1 m/s).

Andre, M. A.↗
DOE OSTI2024

Nuclear reactor radiation and temperature effects on piezoelectric surface acoustic wave devices

Surface acoustic wave (SAW) resonators were characterized in-situ in a nuclear reactor environment at high temperature. Devices based on lithium niobate (LiNbO 3 ), aluminum nitride (sc-AlN), and thin-film aluminum nitride on sapphire substrate (AlN/sapphire) were tested up to 400 °C temperature and 1.9 × 10 12 n/cm 2 s neutron flux. Shifts in device resonant frequency were detected in response to temperature and neutron flux. Devices undergo a frequency change when exposed to neutron flux. At 300 °C, AlN/sapphire produced the strongest neutron flux response about 1.02 ppm at 1.27 × 10 12 n/cm 2 s neutron flux (5.7 × 10 4 rad-Si/hr neutron dose rate), compared to 0.30 ppm for LiNbO 3 and 0.17 ppm for sc-AlN. While the transient kinetics in response to step change in neutron flux support the defect-accumulation mechanism, the concurrent measurement of device temperature using resistive temperature sensor suggests additional heating caused by absorption of gamma rays can also play a role. These results make SAW devices attractive candidates for sensor applications in extreme environments.

Aluminum nitride↗
DOE OSTI2024

INSTRUMENTATION FOR IN-CORE REAL-TIME MECHANICAL TESTING OF STRUCTURAL MATERIALS (INCREASE) PROJECT

Idaho National Laboratory (INL), in collaboration with the Electric Power Research Institute (EPRI), Nuclear Regulatory Commission, French Atomic and Alternative Energies Commission (CEA), Joint Research Centre, Nuclear Research and Consultancy Group, and Research Center Rez, is the operating agent of the In-Core Real-Time Mechanical Testing of Structural Materials (INCREASE) joint experimental program (JEEP) project that operates within the Nuclear Energy Agency (NEA)’s Framework for Irradiation Experiments (FIDES-II) program. The first objective of the INCREASE project is to design a shared capsule that can house a variety of in-core mechanical testing instrumentation for enabling smart irradiation experiments. The second objective is to collect in core stress relaxation data pertaining to high priority stainless steel (SS) materials under light water reactor conditions with target temperature of 340°C +/-20°C and a minimum neutron damage of 2 displacement per atom (dpa). The INCREASE project will be demonstrated in the Massachusetts Institute of Technology Nuclear Research Reactor (MITR).

36 MATERIALS SCIENCE↗
DOE OSTI2023

Machine learned interatomic potential for dispersion strengthened plasma facing components

Tungsten (W) is a material of choice for the divertor material due to its high melting temperature, thermal conductivity, and sputtering threshold. However, W has a very high brittle-to-ductile transition temperature, and at fusion reactor temperatures (≥1000 K), it may undergo recrystallization and grain growth. Dispersion-strengthening W with zirconium carbide (ZrC) can improve ductility and limit grain growth, but much of the effects of the dispersoids on microstructural evolution and thermomechanical properties at high temperatures are still unknown. We present a machine learned Spectral Neighbor Analysis Potential for W–ZrC that can now be used to study these materials. In order to construct a potential suitable for large-scale atomistic simulations at fusion reactor temperatures, it is necessary to train on ab initio data generated for a diverse set of structures, chemical environments, and temperatures. Further accuracy and stability tests of the potential were achieved using objective functions for both material properties and high temperature stability. Additionally, validation of lattice parameters, surface energies, bulk moduli, and thermal expansion is confirmed on the optimized potential. Tensile tests of W/ZrC bicrystals show that although the W(110)–ZrC(111) C-terminated bicrystal has the highest ultimate tensile strength (UTS) at room temperature, observed strength decreases with increasing temperature. At 2500 K, the terminating C layer diffuses into the W, resulting in a weaker W–Zr interface. Meanwhile, the W(110)–ZrC(111) Zr-terminated bicrystal has the highest UTS at 2500 K.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Water-gas shift reaction in ceramic-carbonate dual-phase membrane reactor at high temperatures and pressures

Water-gas-shift (WGS) reaction is a critical step in integrated gasification combined cycles (IGCC) power plants with CO 2 capture. Membrane reactors made with a CO 2 -permselective ceramic-carbonate dual-phase (CCDP) membrane offers the potential to enhance hydrogen yield with simultaneous CO 2 capture for WGS reaction. The present work studies operation of WGS reaction in a tubular membrane reactor made of samarium-doped ceria infiltrated with lithium/sodium molten carbonate mixture. The WGS reaction was performed in the membrane reactor with and without a high-temperature WGS catalyst at 800–850 °C, feed pressure of 7 bar, the space velocity of 150–3000 h –1 , and a feed gas mixture of 45.7/13.1/41.3 mol% CO/CO 2 /N 2 with steam to carbon ratio of 4. The results show that the catalyst-free membrane reactor can convert 92% of carbon monoxide into CO 2 and H 2 and recover 29% CO 2 at 850 °C and a space velocity of 150 h –1 . However, in the catalyst-free membrane reactor, a significant amount of unwanted carbon deposition is observed. The side reactions can be minimized by reducing the operating pressure and increasing the operating temperature and space velocity, and completely avoided using a high-temperature catalyst at space velocity>500 h –1 . The membrane reactor with a WGS catalyst achieves CO conversion of about 85%, above the equilibrium conversion, and 40% CO 2 recovery without carbon deposition at high temperature and pressure. Finally, the membrane remains in the same structure and gas-tightness after the WGS reaction tests.

20 FOSSIL-FUELED POWER PLANTS↗